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International Conference onManagement of Spent Fuel

from Nuclear Power Reactors:An Integrated Approach to theBack End of the Fuel Cycle

Report of Contributionshttps://conferences.iaea.org/e/69

International Co … / Report of Contributions The use of Nuclear Fusion Energy

Contribution ID: 0 Type: POSTER

The use of Nuclear Fusion Energy

This is a test

Country/ int. organization

IAEA

Primary author: KAMENDJE, Richard (IAEA)

Co-author: Ms MORRISON, Karen (IAEA)

Presenter: KAMENDJE, Richard (IAEA)

January 27, 2021 Page 1

International Co … / Report of Contributions MANAGING SPENT NUCLEAR F …

Contribution ID: 2 Type: ORAL

MANAGING SPENT NUCLEAR FUEL FROMGENERATION TO FINAL DISPOSAL: INTEGRATIONOF THE BACK-END OF THE NUCLEAR FUEL CYCLE

Management of spent nuclear fuel (SNF), produced from nuclear reactors, consists of three maincomponents: storage, transportation, and disposal. Of these three components, disposal is notoccurring in the United States (US) and transportation has occurred on an intermittent basis. Short-term storage of commercial SNF generally occurs in either the reactor pool or dry storage casks onthe operating or former site of the nuclear reactor. No consolidated interim storage facility (ISF)for commercial SNF has been implemented yet. More importantly, the designs for the current at-reactor dry storage casks at commercial nuclear reactors have evolved in an ad-hoc fashion. Forexample, the current size of the canisters in the dry storage casks could severely limit disposaloptions, specifically, the geologic media that could viably directly dispose of the SNF without verylong-term storage or reopening of the canisters. Consequently, storage is not integrated withdisposal to form a coherent system for managing SNF.

The lack of substantial integration between storage, transport, and disposal of SNF does not meanthat SNF is not currently stored safely. SNF is, and can continue to be, stored, transported, anddisposed of safely in accordance with US Nuclear Regulatory Commission regulations. An impor-tant question, however, is the cost to provide the required level of safety. The lack of integrationmeans that costly solutions may have to be implemented to address problems that could have beenavoided with deliberate integration, or that some aspect of SNF management is delayed substan-tially because of the lack of integration.

This work reviews (1) the current state of the three components for managing SNF in the US, (2)past recommendations for moving toward an integrated storage, transport, and disposal systemfor SNF, and (3) progress – or lack thereof – in implementing those recommendations and wherefurther steps might be taken. A key conclusion from this work is that consolidated interim storagecan provide an important integrating function within the waste management system (although notsufficient by itself), in addition to the currently functioning storage of SNF on-site at commercialnuclear reactors.

Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Cor-poration, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department ofEnergy’s National Nuclear Security Administration under contract DE-AC04-94AL85000.

SAND2014-17936A


USA/Sandia National Laboratories

Primary author: Dr BONANO, Evaristo Jose (Sandia National Laboratories)

Co-authors: Dr KALININA, Elena A. (Sandia National Laboratories); Mrs PRICE, Laura L. (SandiaNational Laboratories); Mr RECHARD, Rob P. (Sandia National Laboratories)

Presenter: Dr BONANO, Evaristo Jose (Sandia National Laboratories)


International Co … / Report of Contributions Investigations of SCC on Spent Fu …


Investigations of SCC on Spent Fuel Dry StorageCanisters used for Long Term Storage

In 2009 the United States made a decision to discontinue pursuing a license for a long-term geologicrepository for used nuclear fuel. Until another license is pursued, used nuclear fuel will accumulateand remain in dry storage for longer than originally planned. At the end of 2013, the US hadover 22,000 Metric Tons Uranium (Initial) of used nuclear fuel in 1850 dry storage casks, storedin over 60 sites at reactor sites across the United States (Carter & Vinson, 2014). Most of thisfuel is placed in metal canisters (usually welded 304SS) which are either stored horizontally ina concrete bunker or vertically in individual concrete over packs. Each canister/cask system ispassively cooled using open vent ports to allow natural convection to cool the canister inside theover pack. These open vents also allow dusts and other particulates from the outside air to enterand settle on the canister surface. The US Department of Energy and the US Nuclear RegulatoryAgency are interested in understanding potential failure mechanisms for these casks during longterm dry storage; one being stress corrosion cracking (SCC). Few visual inspections have occurredbecause 1) it is difficult to remove the canister once it is inside the overpack, 2) there is very littlespace between the canister and the overpack, and 3) there are high radiation levels. Because ofthese limitations, the DOE is working to understand the mechanisms for SCC of these canisters.In order for stress corrosion cracking to occur, three conditions must be met. There must be:1. A susceptible material: 304SS is known to be susceptible to SCC.2. A corrosive environment: Samples of the dust ,salts, and temperature on canisters in variouslocations have been collected.3. Tensile Stress: A full-scale diameter cylindrical mock-up is being used to measure residualstresses near the welds and heat affected zones.

The analysis of these experiments will be discussed in this paper.It is very important to maintain leak tight canisters; therefore the US Nuclear Regulatory Agency isconsidering visual inspections of individual canisters while in storage. The R&D being performedin this project will be used to help determine the need for this testing, potential start times, loca-tions, and frequencies of that monitoring to help ensure that the entire fleet of canisters remainsas protective as designed for the duration of the storage life.


Sandia National Laboratories

Primary author: Ms SALTZSTEIN, Sylvia (Sandia National Labs)

Co-author: Mr SORENSON, Ken (Sandia National Labs)

Presenter: Ms SALTZSTEIN, Sylvia (Sandia National Labs)


International Co … / Report of Contributions The IAEA coordinated research pr …


The IAEA coordinated research programmes BEFASTand SPAR – thirty-four years of experience with the

storage of spent fuel and facility components

The IAEA Coordinated Research Programmes (CRP) on the behaviour of spent fuel and storagefacility components during long term storage were initiated in 1981 under various titles. The firstprogramme was designated BEFAST-I (BEhaviour of spent Fuel Assemblies in STorage), followedby BEFAST-II in 1986 and BEFAST-III in 1991. A follow-on CRP was called SPAR-I to -III (Spentfuel Performance Assessment and Research), respectively.During the CRP, the participating countries contributed their R&D results on fundamental ques-tions regarding spent fuel storage. The overall objectives of the CRPs were to develop a technicalknowledge base on long term storage of commercial power reactor spent fuel through evaluationof operating experience and research by participating Member States, and to extrapolate predic-tions of spent fuel behaviour over long periods of time.A final report for each stage of the CRPs was prepared and published as an IAEA Technical Docu-ment (TECDOC).Towards the end of the BEFAST-III project, it became apparent that the R&D component of theproject was decreasing steadily; more emphasis was being placed on the operation and implemen-tation of storage technology. The storage technology (particularly dry storage) was undergoinga rapid evolution: new fuel and material design changes were coming on stream and target dis-charge burnup were steadily increasing; in addition, storage durations were predicted to be muchlonger than anticipated.Overall, 20 countries and one international organization participated in one or more of the CRPs.The paper describes the research objectives and results achieved during the course of the CRPs.


Hungary

Primary author: Mr TAKATS, Ferenc (TS Enercon Kft.)

Co-authors: Mr MACHIELS, Albert (EPRI, Palo Alto); Mrs JUSSOFIE, Astrid (GNS, Essen); MrISSARD, Herve (Areva TN); Mr STANDRING, Paul (IAEA, NEFW)

Presenter: Mr TAKATS, Ferenc (TS Enercon Kft.)


International Co … / Report of Contributions Ageing Management of Spent Nuc …


Ageing Management of Spent Nuclear Fuel Facilities

Argentina has three nuclear Power Plants in operation, Atucha I (CNA I), Embalse (CNE) andAtucha II (CNA II). All of them are based on natural uranium as fuel with heavy water moderationand refrigeration.Atucha I (370 Mwe) and Atucha II (750 Mwe) are almost unique in their type, both are PressureVessel type. The Embalse Nuclear Power Plant ( 600 Mwe) is a CANDU type.In this work we will refered to the Atucha I spent fuel. The fuel of Atucha is 6 meter long and has36 bars of zircaloy-4. For these fuels there are deposits located in the places where the reactorsare situated. Atucha I has two pools houses where about 11.000 spent fuels is accumulated so far.From middle of the year 2000 the whole nucleus are of fuels with 0,85% enrichment and therefore,the consumption of fuel will be of 210 fuels per year.One of the alternatives is when the Atucha Nuclear Power Plants concludes their operation to passthe spent fuel to dry storage and another alternative is to continue with the operation of the poolsand pass to dry a group of select spent fuel. The main objective is to elaborate a life managementprogramme of the deposits of spent fuel inside the general life management programme of nuclearfacilities. The life management plan of a facility is permanent and it is adjusted to the changesthat may occur in the ageing mechanisms and the technological advances that allow them to bemitigated or solved.For the selection of the technology it should be kept in mind, those characteristic of the fuel elementAtucha, their longitude and their burn up one that it will be different according to the time that isconsidered. The construction should also be to modulate, allowing toadd sectors according to thenecessities. It should also be considered the conditions in which it can be carried out the transferfrom the pools to the deposits in dry.


Argentina-Comision Nacional de Energia Atomica

Primary author: Dr VERSACI, Raul (Comision Nacional de Energia Atomica)

Presenter: Dr VERSACI, Raul (Comision Nacional de Energia Atomica)


International Co … / Report of Contributions THORIUM FUEL CYCLES IN A VE …


THORIUM FUEL CYCLES IN A VERY HIGHTEMPERATURE HYBRID SYSTEM

The current growth of the energy demand, the perspective of a pronounced increment for the nextfuture, added to the near depletion of the fossil fuels has made finding sustainable alternatives ofenergy supply, a challenge to the international scientific community. Nuclear Energy is presentedas a prominent energy source because nuclear energy is a clean, safe, and cost-effective energysupply. However, nuclear energy faces substantial challenges to be successful as a sustainableenergy source: manageable nuclear waste, effective fuel utilization, and increased environmentalbenefits, competitive economics, recognized safety performance and secure nuclear energy sys-tems and nuclear materials. Nowadays an innovative generation of nuclear energy systems andfuel cycles are investigated in order to solve these challenges. The Generation IV of nuclear reac-tors is expected to solve the problems of the nuclear energy. Pebble Bed Very High Temperatureadvanced systems together with fuel cycles based in Thorium has significant perspectives to takeon the future nuclear energy development challenges and to increase the development possibilities.In this paper the main advantages of the use a Very High Temperature hybrid system using a vari-ety of fuel cycles based on Thorium (Th-U233, Th-Pu239 and Th-U) under a deep burn scheme arestudied. The conceptual design of the Very High Temperature hybrid system composed of a VeryHigh Temperature Pebble Bed Reactor (VHTR) and two Pebble Bed Accelerator Driven Systems(ADSs) is analyzed. The VHTR and the ADSs are designed to work in a thermal neutron spectrum,moderated by graphite and cooled by Helium. The fuel elements are TRISO coated particles con-fined in graphite pebbles. Parameters related to the neutronic behavior like nuclear fuel breeding,Minor Actinide stockpile, the energetic contribution of each fissile isotope and the radiotoxicityof the long lived wastes are used to study Th+U233, Th+Pu239 and Th+U fuel mixtures based onThorium in the hybrid system, using the MCNPX ver. 2.6e computational code.


Cuba/The Higher Institute of Technologies and Applied Sciences

Primary author: Prof. RODRIGUEZ GARCIA, Lorena Pilar (The Higher Institute of Technologiesand Applied Sciences, Faculty of Nuclear Sciences and Technologies)

Co-authors: Dr BRAYNER DE OLIVEIRA LIRA, Carlos Alberto (Departamento de Energia Nu-clear—UFPE); Dr GARCÍA HERNÁNDEZ, Carlos Rafael (The Higher Institute of Technologies and Ap-plied Sciences , Faculty of Nuclear Sciences and Technologies); Dr MILIAN LORENZO, Daniel Evelio(The Higher Institute of Technologies and Applied Sciences, Faculty of Nuclear Sciences and Technolo-gies); Prof. MILIAN PEREZ, Daniel (The Higher Institute of Technologies and Applied Sciences , Facultyof Nuclear Sciences and Technologies)

Presenter: Prof. RODRIGUEZ GARCIA, Lorena Pilar (The Higher Institute of Technologies andApplied Sciences, Faculty of Nuclear Sciences and Technologies)


International Co … / Report of Contributions COMPARATIVE ANALYSIS OF IS …


COMPARATIVE ANALYSIS OF ISOTOPECOMPOSITION OF VVER 1000 WESTINGHOUSE

AND TVEL SPENT FUEL

Determination of isotope composition of spent fuel is necessary to solve the tasks related to:- Account and control of quantity of nuclear hazardous material;- Determination of source terms during analysis of radiation safety;- Using burnup as the nuclear safety parameter while substantiating safety of spent fuel manage-ment systems (“burnup credit” principle).Isotope composition of spent fuel is determined by not only of its burnup level, but also thoseconditions, or, more exactly to say, by that neutron spectrum under which this burnup occurred.Spent nuclear fuel with the same burnup value can have different isotope composition dependingon neutron spectrum in which this burnup took place. The more hard was neutron spectrum, themore U238 is involved into the burnup process (mainly, due to generation of Pu239), and the moreU235 is remained in spent fuel under the same burnup level. Therefore, this work considers thoseoperational parameters, which changes are capable of influencing upon in-core neutron spectrumhardening.This work has analyzed the impact on VVER 1000 spent fuel isotope composition caused by thedifferent operational conditions, such as the presence or absence of absorber-rods, oscillating theconcentration of boric acid, dissolved in the moderator (water) during the campaign, fuel and/ormoderator temperature, as well as changes in water amount at the periphery of an assembly dueto its location in the central or periphery part of the core and/or due to changes in inter-assemblygaps. Also, impact caused by technological allowances applied while manufacturing fuel assemblywas analyzed by weight of fuel and by its enrichment.Calculations were made for reactor cells of fuel assemblies for VVER-1000. They were composedof the new fuel assemblies of USA Westinghouse and the typical fuel assembly of Russian TVELsuppliers.


Ukraine/SSTC NRS

Primary author: Dr KOVBASENKO, Yurii (SSTC NRS)

Presenter: Dr KOVBASENKO, Yurii (SSTC NRS)


International Co … / Report of Contributions The Increasing Importance of Stor …


The Increasing Importance of Storage AgingManagement in Assuring the Sustainability of the

Back End of the Fuel Cycle

On September 19, 2014 the United States Nuclear Regulatory Commission (NRC) published itsfinal rule and supporting Generic Environmental Impact Statement (GEIS) on Continued Storageof Spent Nuclear Fuel. This action, motivated by a U.S. Court mandate following the terminationof the Yucca Mountain Repository project, marked a significant shift in U.S. regulatory policy.Previously, since 1979, NRC had addressed the long-term environmental impacts of the growinginventory of spent fuel at reactor sites through a Waste Confidence rule that was largely predicatedon assumptions about the availability of a repository for geologic disposal in some reasonabletimeframe. However, in a world where the U.S. no longer had a repository program, and might nothave one for the foreseeable future, the Court found that approach to be inadequate – remandingand vacating the Waste Confidence rule.

The shift from Waste Confidence to Continued Storage marks the beginning of a new paradigm inspent fuel management in the U.S. Under Waste Confidence, the schedule for the development of ageologic repository was the central focus toward assuring the sustainability of the back end of thefuel cycle. But Continued Storage contemplates the environmental impacts of indefinite storageeither at reactor sites or centralized locations – shifting this focus to the aging management ofspent fuel storage systems. While the U.S. will continue to pursue geologic disposal, it is nowa given that final disposal decisions will be made by future generations. The responsibility nowplaced on the current generation is to provide long-term assurance that spent fuel in storage willremain safe until future disposal decisions can be made.

This paper will focus on what is being done in the U.S. to meet this responsibility. It will describeaging management programs being instituted at U.S. nuclear facilities to support the renewal ofNRC storage licenses along with the research and development efforts being conducted to supportthese programs – highlighting industry’s response to specific postulated age related degradationmechanisms and efforts to develop enhanced inspection technologies. Finally, the paper will offera forward-looking perspective on what additional technologies might be needed in the future toassess and mitigate age related degradation as well as to facilitate repackaging prior to ultimatedisposal.


US/Nuclear Energy Institute

Primary author: Mr MCCULLUM, Rod (Nuclear Energy Institute (U.S.))

Presenter: Mr MCCULLUM, Rod (Nuclear Energy Institute (U.S.))


International Co … / Report of Contributions Long-term Storage of Spent Nucle …


Long-term Storage of Spent Nuclear Fuel and HLWin Dual Purpose Casks towards Disposal – Challenges

and Perspectives

The safe and secure storage of spent nuclear fuel and high-level waste from nuclear power reactorsis one of the major long-term issues to be solved in the nuclear business. Today no country has arepository available for final disposal or in some countries reprocessing capacities for spent fuelare not sufficiently available as planned. In the meantime the safe interim storage for the long-term - may be 80 years or even longer - is the only available option from today’s perspective anddry cask storage is increasingly favoured throughout the world for various reasons.This paper describes and discusses the major challenges of spent fuel management in Germanyafter the phase-out for nuclear electricity generation was decided in 2011 and a new repository sit-ing procedure was implemented in 2013. Consequences from those decisions which were legallyfounded by amendments of the German Atomic Energy Act result in the need to transfer all re-maining spent fuel from limited reactor operation (last reactor shutdown until the end of 2022) intocasks for subsequent dry interim storage on-site. Storage licenses are generally issued site-specificconsidering specific dual purpose casks (DPC) and their inventories and they are generally limitedto 40 years so far. But the need for extending that interim storage period in the future has becomeobvious. Even though, this is not an issue to be solved today questions about additional safetydemonstrations will arise as soon as licenses need to be extended. Certainly, these questions willask for reliable data about the long-term performance and safety of structures, systems, and com-ponents, e. g. the long-term performance of cask components and materials like bolted closuresystems including metal gaskets, or fuel rod behaviour concerning cladding materials under stressand temperature conditions. In case of dual purpose casks for storage and transportation this in-cludes aspects on how to demonstrate transportability during or after several decades of interimstorage. Long-term investigations often require plenty of time and therefore need to be initiatedtimely.Preliminary R&D results and experiences from more than 20 years of safe interim storage in Ger-many are discussed. In addition, the national perspective towards potential data and R&D needsto demonstrate safety for extended interim storage periods is related to international actions asthe US Extended Storage Collaboration Program (ESCP) and several IAEA projects focussing onthat issue.


Germany

Primary author: Dr VÖLZKE, Holger (BAM Federal Institute for Materials Research and Test-ing)

Co-author: Dr WOLFF, Dietmar (BAM Federal Institute for Materials Research and Testing)

Presenter: Dr VÖLZKE, Holger (BAM Federal Institute for Materials Research and Testing)


International Co … / Report of Contributions Specific safety consideration for m …


Specific safety consideration for management ofspent nuclear fuel in different countries;

Comparative study

Spent fuel is generated continually by operating nuclear reactors. It is stored in the reactor fuelstorage pool for a period of time for cooling and then may be transferred to a designated wet ordry spent fuel storage facility, where it will await reprocessing or disposal (if it is considered tobe radioactive waste). The spent fuel storage pools of some reactors have sufficient capacity toaccommodate all the spent fuel that will be generated during the lifetime of the reactor. Specificsafety consideration of spent fuel management in ten countries considered in this paper. Four ofthese countries reprocess their spent fuel (France, Japan, Russia, and the United Kingdom) andfive are planning on direct disposal (Canada, Germany, the United States, Finland and Sweden).South Korea’s disposal plans are currently a subject of discussions with the United States in con-nection with the renewal of their bilateral agreement on nuclear cooperation. Law number 7 forthe year 2010 (Egyptian nuclear law) addressing in some articles the safety consideration of spentfuel, policy and strategy and it will be discussed in the text. Egyptian Nuclear and RadiologicalRegulatory Authority (ENRRA) established draft regulation for management of spent nuclear fuelstorage. The Egyptian regulations are guided by the recommendation of IAEA safety standard.More details about the Egyptian regulations on the safety of spent fuel storage will be discussedin the full text.


Egypt/Nuclear and Radiological Regulatory Authority, Nasr city, Cairo, Egypt

Primary author: Prof. ABDEL GELEEL, Mohamed (Head of Nuclear Fuel Cycle Department)

Presenter: Prof. ABDEL GELEEL, Mohamed (Head of Nuclear Fuel Cycle Department)


International Co … / Report of Contributions An Integrated Approach to Closin …


An Integrated Approach to Closing the TechnicalData Gap for High Burnup Spent Fuel Performance

during Normal Conditions of Transport (NCT)

The United States Department of Energy (DOE), Office of Nuclear Energy (NE) initiated a programin 2009 to develop the technical basis to support the licensing for long term storage and subsequenttransportation of high burnup spent fuel. An initial focus of this program was development of atechnical data gap analysis that identified the data gaps that needed to be addressed to supportlicensing. Given the number of gaps that were identified, a prioritization of gaps was developedin order to identify the high priority gaps where a focus would result in high impact in termslicensing support.

From the gap analysis and prioritization work, performance of high burnup fuel during transportwas identified as high priority. This paper summarizes the work that has subsequently been doneto address the issue and to understand performance characteristics of high burnup fuel duringNCT operations. The work that has been done includes laboratory and field testing, as well asmodeling and simulation. Further, this work addresses material property issues associated withcladding strength and ductility, cladding and fuel interaction related to the fuel rod response tocyclic loadings, and loading functions on the fuel generated from shock and vibration conditionsthat are representative of Normal Conditions of Transport. This integrated approach builds confi-dence in our overall understanding of high burnup spent fuel system response to transportationloadings.

This paper will cover the technical issues associated with addressing the technical gap of highburnup spent fuel performance when subjected to Normal Conditions of Transport. The theme ofthe paper will show how integration of engineering tools (e.g., testing and analysis), coupled withunderstanding of the basic science phenomena germane to this problem, is used to demonstrateunderstanding of how high burnup fuel will perform during Normal Conditions of Transport.

Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Cor-poration, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department ofEnergy’s National Nuclear Security Administration under contract DE-AC04-94AL85000.


Sandia National Laboratories

Primary author: Mr SORENSON, Ken (Sandia National Laboratories)

Presenter: Mr SORENSON, Ken (Sandia National Laboratories)


International Co … / Report of Contributions MCNP MODEL FOR RESEARCH R …


MCNP MODEL FOR RESEARCH REACTOR(TRR-1/M1) FUEL STORAGE RACK CRITICALITY

ANALYSIS

The spent fuel storage of TRIGA reactor (TRR-1/M1) was designed. Detailed criticality safety anal-ysis was conducted according to the basic safety standard and within the framework of the regula-tory aspect. The Monte Carlo computer code (MCNP) was used for the detailed geometry modeling.Two types of fuel elements were considered for the calculation including 8.5 wt% and 20 wt% fuelelements. The critical number of the fuel elements was determined. The dependence of multipli-cation factor (keff) on the distance between the fuel elements in the rack (pitch) and the waterdensity which varied from 0.01 g/cm3 to 1 g/cm3 were calculated. The MCNP results shown thatthe keff strongly depend on the lattice pitch, moderation effect and the water density. At present,there is limited experience in the safety management of spent fuel for research reactor. The strat-egy development and implementation of the integrated approach to management of the spent fuelfrom research reactor are important for the long term strategy of the ageing management program.


Thailand Institute of Nuclear Technology

Primary author: Dr TIYAPUN, Kanokrat (Thailand Institute of Nuclear Technology)

Presenter: Dr TIYAPUN, Kanokrat (Thailand Institute of Nuclear Technology)


International Co … / Report of Contributions Developing Regional Spent Fuel St …


Developing Regional Spent Fuel Strategies

Despite well-developed repository programs in Sweden, Finland and France, and encouragingprogress in several other countries including Canada, the vast majority of states – including the U.S.– continue to struggle developing and implementing timely plans to safely and securely dispose ofspent fuel and high level waste produced by nuclear power programs. Spent fuel continues to ac-cumulate in cooling pools with limited storage capacity. On-site dry cask storage is a longer-term,but still temporary, solution and reprocessing generates significant waste streams that require ge-ologic disposal.The lack of spent fuel management options strains the credibility of the nuclear community andundercuts public and political acceptance for current and future nuclear activities. It is unethical,irresponsible and ultimately unsustainable to push the problem onto future generations. It is alsoa proliferation and security issue. The decisions that countries make on the back end of the fuelcycle (activities that could include dry cask storage, shipping spent fuel to away-from-reactor stor-age or to reprocessing plants and, eventually, the final disposal of spent fuel or high level waste)can include sensitive nuclear technologies such as plutonium separation.Drawing on two years of work funded by the Hewlett and MacArthur Foundations, the principalinvestigators offer recommendations for developing spent fuel pathways that address broader fuelcycle concerns. Key findings include:• many countries face storage and disposal siting challenges. This is particularly true in countriessuch as Japan and South Korea, where the need is urgent and growing.• cooperative networks and regional frameworks for storage and disposal could be a productiveway to address these problems in Asia and elsewhere. Such partnerships can enhance regionaltransparency and flexibility as well as improving global security and fortifying nonproliferation.• the possibility of multinational options should not be used as an excuse for countries to neglectdomestic responsibilities; all countries must have a national spent fuel management program. Con-versely, national programs should not oppose concerted exploration of multinational approaches.• geological disposal can be both a business venture and a public service. Even one success storycould be a game changer, prompting regional waste management solutions in other parts of theworld.• many nations with spent fuel management issues state that sustained U.S. engagement in re-gional fuel cycle/spent fuel initiatives is necessary.• many productive avenues of discussion can be pursued immediately that promote cooperationwell before the difficult phase of repository siting.


United States/Nuclear Threat Initiative

Primary author: Dr NEWMAN, Andrew (Nuclear Threat Initiative)

Co-author: Prof. ISAACS, Thomas (Nuclear Threat Initiative)

Presenters: Dr NEWMAN, Andrew (Nuclear Threat Initiative); Prof. ISAACS, Thomas (NuclearThreat Initiative)


International Co … / Report of Contributions Safety of the Dry Spent Nuclear Fu …


Safety of the Dry Spent Nuclear Fuel Storage inUkraine: Scientific Approach and Results

More than 250 tons of spent nuclear fuel is produced by Ukrainian NPPs each year. Currentlythe problem of spent nuclear fuel storage in Ukraine is under process of solving. The temporarystorage of SNF until accepting decision about disposal or reprocessing today is used for six WWER-1000 reactors of Zaporizhska NPP.The Dry Spent Nuclear Fuel Storage Facility (DSNFSF) on Zaporizhska NPP is operated more than10 years. The safety of DSNFSF consists of three main components – Nuclear Safety, RadiationSafety and Safe Thermal Condition. Therefore the problem of determination of safe thermal con-dition for the storage facility is one of main problems at operation. Usually in engineering calcu-lations is used conservative approach which doesn’t allow identifying detailed processes in equip-ment and gives approximate results. But for the increasing safety level on DSNFSF it is necessaryto know all details and features of equipment operation especially in field of thermal processes.So scientific approaches are extremely necessary on all stages of facility operation: from construc-tion to decommissioning. The long and the safe operation of DSNFSF on Zaporizhska NPP can beensured only with full scientific support also and it is requires the developing new methods andmethodologies for calculations.For detailed definition of thermal state of spent fuel assemblies during all period of storage themulti-stage approach was developed. On each level of multi-stage approach the solving of conju-gate heat transfer problems were used. This approach was used for:1. The definition of thermal state of containers with spent nuclear fuel with taking into accountthe outer factors influence and spent fuel assemblies heat generation capacity changing.2. The definition of containers group’s thermal state on open storage platform with taking intoaccount their mutual influence, the influence of outer factors and heat generation capacity chang-ing.In addition all calculations which are mentioned above required solving the inverse heat transferproblems. In particular the inverse heat transfer problems were used for:1. The placement of spent fuel assemblies in storage basket for minimization of maximal tempera-tures.2. The identification of equivalent thermal physical properties of compound bodies (storage basket,spent fuel assemblies, fuel rods) for using its in simplified models.The results which were presented in this work are used for estimation and the improving of safetylevel of DSNFSF on Zaporizhska NPP.


National Academy of Sciences of Ukraine

Primary author: Dr ALYOKHINA, Svitlana (National Academy of Sciences of Ukraine)

Presenter: Dr ALYOKHINA, Svitlana (National Academy of Sciences of Ukraine)


International Co … / Report of Contributions Consequences of Long Term Black …


Consequences of Long Term Black Out Accident onSpent Fuel Pool for VVER-1200

Natural disaster occurred at Japan Fukushima Daiichi NPP in March 2011 showed that combinationof natural disasters with a long term black out which is not accepted as a design basis accidenttill that day could be how dangerous. In this study, inventory calculation in the spent fuel pool,transmission from inventory to source term and atmospheric dispersion calculation of source termwere investigated, in a case of long term black out accident that can occur in the AES 2006 designVVER-1200 type of NPP planned to be build in Turkey.


Turkey/TAEK

Primary author: Mr DOĞAN, KEMAL (Turkish Atomic Energy Authority)

Presenter: Mr DOĞAN, KEMAL (Turkish Atomic Energy Authority)


International Co … / Report of Contributions Desirable Back-End of Nuclear Fue …


Desirable Back-End of Nuclear Fuel Cycle

After the irradiation of the nuclear fuel, it is necessary to storage the fuel for cooling into thereactor pools, this is the first step of a series of process before reach his final destination.Up to now there are two options more commonly adopted for the nuclear fuel cycle, one is theopen cycle which requires a deep geological repository for final disposition of fuel. the other willbe the reprocessing of the fuel to extract uranium and plutonium as the two valuable materialremaining into the spent fuelThe back end of the fuel cycle should be carefully planned to avoid delays in the implementationof any of the options available to complete the activities related with each option. also importantare the time steps for each option and interfaces between them.First it is important to decide if the nuclear fuel cycle will be: open cycle or closed cycle to focus theactivities, facilities and technology necessary to complete the fuel cycle selected. so, it is importantto plan an integrated approach for the management of the back end of the nuclear fuel cycle.The steps for the back-end are as follow: After irradiation there is a step for cooling the spent fuel,this step will last 5 years, as the cooling takes place in the reactor pool, eventually will be saturatedand a storage step out of the rector pool will continue for part of the total fuel discharged from thereactor.A Short Term Option is available if there is not a policy that indicates clearly if the nuclear fuelcycle will be open or closed. in such a case the option will be: Interim Wet or Dry storage, Theseoptions will provide at least, 100 years to think over what will be the final destiny of the fuel.This paper will show some of the alternatives available and interfaces between them, as show inthe following diagram.


MEXICO/Instituto Nacional de Investigaciones Nucleares

Primary author: Dr RAMIREZ SANCHEZ, JOSE RAMON (ININ)

Co-author: Dr PALACIOS HERNANDEZ, JAVIER C. (inin)

Presenter: Dr RAMIREZ SANCHEZ, JOSE RAMON (ININ)


International Co … / Report of Contributions Challenges in developing the basic …


Challenges in developing the basic design of theKBS-3 system into a qualified and industrially viable

operation

The programmes for final disposal of spent nuclear fuel are similar in Sweden and Finland, andthere has been extensive cooperation between the waste management organisations in the twocountries over the years. Since both programmes now enter a stage of final design and implemen-tation this cooperation will be deepened, aiming when possible for the same technical design.

The repositories in Sweden and Finland will be constructed according to the KBS-3 method. Whilea technically feasible reference design and layout is presented, detailed designs adapted to an in-dustrialised process designed to fulfilling specific requirements on quality, cost and efficiency needstill be developed. Also the repository layout needs to be adapted to the local conditions foundwhen constructing the repository at depth.

The repository design must be such that it results in a safe repository. For these reason, and inaccordance with regulations, both SKB and Posiva have developed design requirements and otherconditions and presented these to the designer. However, the formulation of what requirementsshall be put on design such that it meets long term safety is not trivial. Safety Assessment, usuallystudies a few specific designs, and would generally not say if there are other designs that may alsolead to safety. It is also easy to formulate rules that would lead to safety, but are impossible toimplement and verify. The design requirements developed typically concern specification on whatmechanical loads the barriers must withstand, restrictions on the composition of barrier materialsor acceptance criteria for the various underground excavations.

Essentially the detailed technical design need to be completed in time for the detailed design ofthe planned facilities in the KBS-3 repository system, i.e. the encapsulation plant, the facility forbuffer and backfill bentonite component production and the underground repository. However,technology development support will also be needed during implementation and start operationof these facilities. A technology development plan spanning the time from now until the license tostart operation has been developed. This plan aims for a common holistic view and understandingof what is needed to reach the target operating facilities and identifies the various developmentefforts needed in relation to the program plan for nuclear fuel program with regard to the timeand resources.


Sweden and Finland

Primary author: Dr ANDERSSON, Johan (Swedish Nuclear Fuel and Waste Management Co.(SKB))

Co-author: Ms JALONEN, Tiina (Posiva Oy, Finland)

Presenter: Dr ANDERSSON, Johan (Swedish Nuclear Fuel and Waste Management Co. (SKB))


International Co … / Report of Contributions Spent Nuclear Fuel Management i …


Spent Nuclear Fuel Management in Switzerland:Perspective for Final Disposal

Spent nuclear fuel management in Switzerland must be considered from the perspective of thedevelopment and implementation of a safe geological repository for both high-level waste andspent fuel assemblies. This paper provides a brief overview of the most relevant issues relatingto the management and disposal of the spent fuel arising from the five operational Swiss nuclearpower plants.

The principal strategy for spent fuel disposal in Switzerland is to enclose it in canisters that areembedded in bentonite, surrounded by the host rock, Opalinus Clay. A prerequisite for the licenc-ing of the geological repository is the development of the scientific basis to ensure, with specifiedsafety margins, the handling of the spent fuel and its encapsulation in repository canisters. Theperformance of the spent nuclear fuel has to be assessed by accurate characterisation of the fuelproperties, such as radionuclide inventory, source term, fission gas release and decay heat. Specificissues such as opening the transport/storage cask, handling of spent fuel in the repository surfacefacility and repackaging of the fuel into the canisters for geological disposal must be properly ad-dressed. Taking into account handling operations, there are numerous safety-relevant issues to beconsidered, starting from material ageing to possible radioactivity release from the cask/fuel. Thecurrent technology for handling and encapsulation of damaged fuel in special containers at NPP’sis also addressed in the paper.

One of the most important considerations for the near-field of the geological repository is the heattransfer between the canisters and the surrounding bentonite and host rock, which sets a boundingvalue for heat production per canister. Because of the very high values for burnups and decay heatof the Swiss spent fuels, the loading of the canisters must be optimised according to the spent fuelproperties at the time of emplacement. Some relevant aspects of this optimised loading processare discussed in the paper.

The paper also examines aspects that are relevant for the assessment of long-term safety; theserequire comprehensive analysis and development of appropriate programs and strategies. A briefoverview of the most significant NAGRA programs under development is also given: the burnupcredit program for the assessment of criticality safety and the Full-Scale Emplacement Experimentin the Mont Terri Rock Laboratory for geophysical monitoring of the changes in the materialproperties of the bentonite backfill with simulated decay heat.


Switzerland

Primary author: Dr CARUSO, Stefano (NAGRA)

Co-author: Dr PANTELIAS GARCES, Manuel (NAGRA)

Presenter: Dr CARUSO, Stefano (NAGRA)


International Co … / Report of Contributions How innovative approaches and te …


How innovative approaches and technologiesthroughout the Fuel Cycle are supporting NPPoperations while anticipating future back-end

challenges

Used Fuel Management is more and more one of the major topics that nuclear countries and utilitieshave to face when managing existing nuclear power plant as well as new build.

Stakeholders concerns are growing in parallel of the steady increase of used fuel inventories dueto fading away of the back-end endpoint in most nuclear countries.

Utilities are additionally being challenged further with new operation conditions to accommodateFukushima-driven safety requirements and to adapt to new grid demand due to growing shareof renewables, intermittent energy by nature, while anticipating the ever-growing competition inelectricity liberalized market.

To respond to this challenging environment and combination of constraints, AREVA has neverstopped over the past few years developing and implementing new technologies at every step ofthe fuel cycle.

Starting from the Front-End, through Enrichment process and innovation in Fuel design and fab-rication as well as Core design and management optimization with drastic reduction of used fueldischarged per electricity production and even direct re-use of earlier discharged used fuel withboth corollary uranium savings and decrease of used fuel inventory and environmental impact.

If the innovation and new technologies in the very reactor operations have helped also respondingto the above mentioned combination of constraints and requirements, innovations in the Back-Endof the Fuel cycle have also helped to respond to the challenging environment whereas throughadvanced innovative solutions in a specific area, where used fuel inventory as well as plutoniumbalance of a utility are both brought to a minimum down to zero.

This paper/presentation will briefly describe challenges and issues at stake and present and illus-trate through industrial scenarios how some AREVA’s innovative approaches and technologiesthroughout the Fuel Cycle are safely and economically supporting NPP operations while anticipat-ing future back-end challenges.


AREVA

Primary author: Mr CHIGUER, Mustapha (Conference Participant)

Presenter: Mr CHIGUER, Mustapha (Conference Participant)


International Co … / Report of Contributions Valuable Knowledge from Canad …


Valuable Knowledge from Canada’s UndergroundResearch Laboratory (1980-2014) to Support the

Development of a Safe Geological Disposal Facility

A comprehensive research and development program of geologic characterization and large-scalegeotechnical experiments in granite was undertaken at Canada’s Underground Research Labora-tory (URL) starting in 1980 to evaluate the concept of permanent disposal of used CANDU® fuelin a deep geological repository (DGR) constructed in Canadian Shield granite. Before the under-ground facility was permanently closed in 2010, the URL was a member of the International AtomicEnergy Agency (IAEA) Network of Centers of Excellence for underground laboratories. Much col-laborative research with international organizations was undertaken at the URL. Currently, evenafter the demolition of the surface facility at the URL in 2014, collaboration with internationalorganizations is still continuing through the Enhanced Sealing Project (ESP). The ESP consists ofinstrumentation and monitoring of a full-scale shaft seal installed to permanently seal the accessshaft for Canada’s URL. The purpose of the seal is to limit the mixing of the saline groundwater be-low fracture zone and fresh groundwater above it. The ESP is a unique opportunity to observe theperformance of a full-scale composite shaft seal, relatively comparable to that likely to be installedat an actual DGR on closure, during the period of groundwater recovery after a repository is closed.Currently the ESP is jointly funded by CNL (Canada), POSIVA (Finland), and ANDRA (France) andmonitoring is planned to be continued until the end of 2016. Over its lifespan, Canada’s URL expe-rienced major organizational and employment changes. Many experts have left the organizationfor other opportunities, retirements, or other reasons. Organizational continuity is important, butit was not the case in Canada’s URL. Many of the technical lessons learned from the URL havebeen lost in the transition of used fuel disposal management responsibilities between differentorganizations. The current DGR design, such as the one in Finland, includes up to 100 years ofoperation before it will be permanently closed. Consequently, successful demonstration of trans-ferring knowledge from one generation or organization to the next will be critical part of a safeoperation of a DGR. This paper is part of initiatives to minimize further knowledge loss at Canada’sURL. It describes some technical lessons learned from Canada’s URL (1980-2014) that can be usedto support the development of a safe geological disposal facility for disposal of highly radioactivewaste.


Canada/ Canadian Nuclear Laboratories (CNL)

Primary author: Dr PRIYANTO PUTRO, Deni G. (Canadian Nuclear Laboaratories)

Co-author: Mr THOMPSON, Paul (Canadian Nuclear Laboratories (CNL))

Presenter: Dr PRIYANTO PUTRO, Deni G. (Canadian Nuclear Laboaratories)


International Co … / Report of Contributions Public Interactions in the U.S. Dep …


Public Interactions in the U.S. Department ofEnergy’s Design of a Consent-Based Approach toSiting Used Nuclear Fuel Interim Storage Facilities

and Associated Transportation Activities

Over the past several decades, the United States has generated used nuclear fuel and high-levelradioactive waste from the commercial production of nuclear energy and from defense relatedactivities. These materials must be properly managed and eventually permanently disposed. TheSecretary of Energy, in 2010, chartered the Blue Ribbon Commission (BRC) on America’s NuclearFuture to recommend a new strategy for managing the back end of the nuclear fuel cycle. The BRCstudied the issues for two years, received input through public meetings and a series of issue papers,and published its final report and recommendations in January, 2012. The BRC recommended anapproach to site nuclear facilities that is consent-based, transparent, phased, adaptive, standards-and science-based, and governed by legally-binding agreements between the federal governmentand a jurisdiction willing to host such facilities.

In January, 2013, the Secretary of Energy issued the Administration’s Strategy for the Manage-ment and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste. The Strategy providesa framework for moving toward a sustainable program to develop an integrated system capableof transporting, storing, and disposing of used nuclear fuel and high-level radioactive waste fromcivilian nuclear power generation, defense, national security and other activities. The Strategyendorsed the proposition that prospective host jurisdictions must be recognized as partners withthe implementing organization. Public trust and confidence are prerequisites to the success of theoverall effort, as is a program that remains stable over many decades. Therefore, public percep-tions and concerns about spent nuclear fuel must be addressed regarding the program’s ability totransport, store, and dispose of used nuclear fuel and high-level radioactive waste in a manner thatprotects the public’s health, safety and security, and the environment.

The Department of Energy (DOE) is undertaking activities within existing Congressional autho-rization to plan for the eventual transportation, storage, and disposal of used nuclear fuel. Key toimplementing the Strategy is the recognition that success will depend on openness, transparency,fairness, inclusiveness, and trust-worthiness. Consistent with the need to move forward underexisting authorization pending enactment of new legislation, DOE intends to engage interestedstate, Tribal, and local governments, other organizations, and the public in discussions about howto design and implement consent-based siting processes, with initial focus on an interim storagefacility as a pilot project to remove used fuel from shutdown reactors.


United States/Department of Energy

Primary author: Mr JONES, Jay (U.S. Dept. of Energy)

Co-author: Mrs HOLM, Judith (North Wind Services)

Presenter: Mr JONES, Jay (U.S. Dept. of Energy)


International Co … / Report of Contributions State of the art Storage Facilities f …


State of the art Storage Facilities for Spent Fuel atIgnalina Nuclear Power Plant, the World’s largest

RBMK Reactor

One pre-condition for Lithuania to join the European Union (EU) in 2004 was to shut down theIgnalina Nuclear Power Plant (INPP). The power plant has two RMBK type water cooled graphitemoderated pressure tube reactors each of design capacity 1500MW(e) which started operationin 1983 respective in 1987. To fulfil this requirement, it was decided that Ignalina NPP severaldecommissioning projects had been initiated.

In January 2005 NUKEM Technologies GmbH in consortium with GNS mbH was awarded a con-tract for an Interim Spent Fuel Storage Facility (B1- ISFSF). As turn-key project the B1-ISFSF coversthe design, procurement, manufacturing, supply, erection, installation, setting-to-work and com-missioning.

Project Description

The INPP reactors were shut down for decommissioning by end 2004 respective 2009. Currentlythe spent fuel from the reactors is still stored in the cooling ponds. As for the decommissioning thereactor units have to be de-fuelled completely and approximately 18,000 fuel assemblies have tobe stored, a new Interim Storage Facility has to be erected. In addition the scope includes also allnecessary equipment for spent fuel retrieval, packaging, sealing, transport and other equipmentappropriate to the design solution and required for the safe removal of the existing spent fuel fromthe storage ponds and insertion into the new ISFSF.

A small proportion of the spent fuel has suffered damage which is minor (without loss of integrityof the cladding) or major (potential for rupture of the cladding). It is necessary that this damagedfuel is also retrieved from the storage pools and stored in the new B1-ISFSF.The following main design criteria for the new B1-ISFSF have be fulfilled:- Interim Storage of spent fuel assemblies for at least 50 years- Possibility of retrievability of spent fuel- Two independent barriers to prevent the release of radioactive material to the environment,whereas the fuel cladding shall not be one of the barriers- Criticality safety keff lesss than 0.95- Nuclear and radiation safety in accordance with national and international standards

The main areas of the B1-ISFSF project are consisting of Fuel Bundle Handling and loading, StorageCasks Type CONSTOR® RBMK1500/M2, Damaged Fuel Handling System (DFHS), Casks HandlingEquipment, transportation equipment, Interim Storage Building with infrastructure, Fuel Inspec-tion Hot Cell (FIHC).

The presentation will describe the technical details, project challenges and actual status of theproject.


Germany

Primary author: Mr RICHTER, Manfred (NUKEM Technologies Engineering Services GmbH)


International Co … / Report of Contributions State of the art Storage Facilities f …

Presenter: Mr RICHTER, Manfred (NUKEM Technologies Engineering Services GmbH)


International Co … / Report of Contributions Criticality Safety and Burnup cred …


Criticality Safety and Burnup credit Analysis forMOX Fuel

Burnup credit is defined as the consideration of the reduction in reactivity associated with theuse of the fuel in power reactors. Changes in the isotopic composition during fuel burnup whichresult in a reduced reactivity can be conveniently characterized by the reduction of the net fissilecontent, the build-up of actinides, the increase of the concentration of fission products, and thereduction of burnable absorber concentration where applicable. In practice, the conservative use ofburnup credit requires consideration of all fissile isotopes, and allows consideration of any neutronabsorbing isotopes for which properties and quantities are known with sufficient certainty[1].

The present research analyse the burn up and neutronic parameters of an assembly of MOX fuels. The discharge burnup is extended up to 70 GWd/T. In the assembly benchmark problems , important parameters for in core fuel management such as local power peaking factors and reactivity coefficients were included in the analysis. A PWR MOX fuel assembly is the same geometrical configuration as 17x17 type PWR fuel design. The assembly pitch 21.505 cm , fuel rod pitch 1.265 cm , pellet outer diameter 0.824 cm , cladding outer diameter 0.952 cm The average Pu fissile content is 11 \% wt assuming 21 full effective power months operations using three batch loading strategy . The assembly is composed of low , middle and high Pu content fuel rods[2].A typical compositions of low , medium and high Pu and the structure content can be found at reference [1]. The power density is 36.6 MW/TH. The density of water coolant at hot and cold conditions are given at reference [2]. MCNPX computer code package ( which is based on Monte Carlo method reference [3] ) is used to model the assembly and a three dimensional computer model has been designed to simulate the burn up of the assembly in a typical operation conditions of PWR reactor

The results indicate burnup dependency of multiplication factor and local peaking factor, burnup dependency of fission rate distributions , Figure 2 illustrates the assembly multiplication factor versus fuel burn up (MWd/T)


Egypt/Nuclear and Radiological Regulatory Authority

Primary author: Mr IBRAHIM, Moustafa aziz (head of nucleae safety department)

Presenter: Mr IBRAHIM, Moustafa aziz (head of nucleae safety department)


International Co … / Report of Contributions Aging Mechanisms Influence on T …


Aging Mechanisms Influence on Transport Safety ofDual Purpose Casks for Spent Nuclear Fuel

Abstract: When storage of spent nuclear fuel (SNF) is done in dual purpose casks (DPC), the ef-fects of aging on safety relevant DPC functions and properties have to be managed in a way thata safe transport after the storage period of several decades is capable, and can be justified and cer-tified permanently throughout that period. The effects of aging mechanisms (like e.g. radiation,different corrosion mechanisms, stress relaxation, creep, structural changes and degradation) onthe transport package design safety assessment features have to be evaluated. The considerationof these issues in the DPC transport safety case will be addressed. Special attention is given to allcask components which cannot be directly inspected or changed without opening the cask cavity,what are the inner parts of the closure system and the cask internals, like baskets or spent fuel as-semblies. The design criteria of that transport safety case have to consider the operational impactsduring storage. Aging is not subject of technical aspects only, but also of “intellectual” aspects,like changing standards, scientific/ technical knowledge development and personal as well as in-stitutional alterations. Those aspects are to be considered in the management system of the licenseholders and in appropriate design approval update processes. The paper addresses issues whichare subject of an actual IAEA TECDOC draft “Preparation of a safety case for a dual purpose caskcontaining spent nuclear fuel”.


Germany

Primary author: Dr DROSTE, Bernhard (BAM Federal Institute for Materials Research and Test-ing)

Co-authors: Dr ROLLE, Annette (BAM); WILLE, Frank (BAM); Dr KOMANN, Steffen (BAM); MrSCHUBERT, Sven (BAM); Mr PROBST, Ulrich (BAM)

Presenter: Dr DROSTE, Bernhard (BAM Federal Institute for Materials Research and Testing)


International Co … / Report of Contributions China Spent Fuel Management Str …


China Spent Fuel Management Strategy and R&D

China has draw up a great nuclear power programme to satisfy the clear energy requirementsof the sustainable development of the economy and society. Currently, there are about 22 nuclearpower plant reactors are opening and the installed capacity is about 21000MWe. The total installednuclear power capacity will achieve 58000MWe in 2020 and the installed capacity of nuclear powerplants under constructing is about 30000MW at that time. If the onsite storage time of the spentfuel is five years, the amount of spent fuel assemblies that should be transported to the reprocessingplant will increased rapidly from about 100 to more than 1000, and nearly about 3000 in the yearof 2015, 2020 and 2025, respectively.Because the spent fuel management is very important for the safely use of nuclear power, chinagives a great efforts to promote the technology and the industry abilities in the fields of spent fuelstorage, transport and reprocessing. Based on the new technical standards, especially after theFukushima disaster, measures such as water position monitoring, automatic water filling systems,and£¬if it is needed£¬additional neutron poison that avoiding critical accident, are used at theonsite spent fuel storage pools. R&D Programmes developing spent nuclear fuel transport cask aswell as dry storage technology are carrying out. According to the close nuclear fuel circle policy,china develop reprocessing technology actively focused on the advanced PUREX with a moresimplified process, no salts reducing agent, and achieving the nonproliferation goals. Through theindependent design and construction or international cooperation, it is expected that a commercialreprocessing plant will be completed in 2030.Under the IAEA’s safeguards framework, china is willing to cooperate and exchanges extensivelywith other countries in the field of spent fuel management.


CHINA NATIONAL NUCLEAR CORPORATION

Primary author: Mr QI, ZHANSHUN (CHINA NATIONAL NUCLEAR CORPORATION)

Presenter: Mr QI, ZHANSHUN (CHINA NATIONAL NUCLEAR CORPORATION)


International Co … / Report of Contributions Development of an Execution Stra …


Development of an Execution Strategy Analysis(ESA) Capability and Tool for Storage of Used

Nuclear Fuel (UNF)

The Nuclear Fuel Storage and Transportation Planning Project (NFST), under the U.S. Departmentof Energy, Office of Nuclear Energy, Office of Fuel Cycle Technology, is developing foundationalcapabilities to support the application of system engineering and decision analysis principles toinform future decisions regarding the deployment of a consolidated interim storage facility (ISF)for managing used nuclear fuel (UNF) and High Level Waste (HLW). The development of an Execu-tion Strategy Analysis (ESA) capability and tool is a key part of evaluating alternative strategies forfuture deployment of a consolidated ISF using a consent-based siting process per the Administra-tion’s Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level RadioactiveWaste (Strategy). The ESA tool allows for on-going performance assessment of the evolving projectplan/strategy that takes into account significant assumptions, risks, and uncertainties throughoutthe project lifecycle.

Deploying a consolidated ISF will be a complex social and technical endeavor. There will be uncer-tainties regarding many aspects associated with meeting key milestones for full implementation;and alsoboth programmatic and technical risks associated with these milestones.

The formal ESA approach being applied by the NFST goes beyond traditional project analysistools. NFST developed a dynamic simulation tool that explicitly models and assesses the impactsof uncertainties (activity durations and costs), constraints (policy, legislation, regulatory), risks(technical, non-technical), and opportunities.

The ESA process has provided a single coherent framework that describes key milestones that mustbe achieved and the activities that must be completed to deploy a consolidated ISF. The process hasalso developed a comprehensive registry of risks and opportunities associated with deployment.In total, these provide a common understanding of the overall effort required.

In addition, the ESA process and the results of the dynamic, probabilistic simulation model canbe used to support the development of plans, budgets, and alternative strategies for meeting thegoals in the Administration’s Strategy.

The paper will describe the process by which the ESA capability and tool are developed and addressthe value of such a process in developing and implementing a long-term strategy for managingUNF and HLW.


USA

Primary author: Mr STOLL, Ralph (Predicus LLC (contractor for Argonne National Laboratory))

Co-author: Dr NUTT, William (Argonne National Laboratory)

Presenter: Mr STOLL, Ralph (Predicus LLC (contractor for Argonne National Laboratory))


International Co … / Report of Contributions Phrohydrolysis research on the flu …


Phrohydrolysis research on the fluoride salts ofTMSR

In order to make full use of the rich resources of thorium, a new nuclear reactor-TMSR (thoriummolten salt reactor) was proposed. In this reactor, the molten mixture of ThF4, UF4 and LiF-BeF2was used.Now we are trying to develop a totally new flowsheet for TMSR spent fuel processing in whichpyrohydrolytic technology is included. Pyrohydrolysis was first introduced by Warf et al. in the1950s[1]. According to this method, the insoluble fluoride slats can be converted into correspond-ing oxides for further management.The pyrohydrolysis behavior of different fluoride salts, such as ThF4 and UF4 were studied on ournew designed eqiupment. The results showed that, UF4 and ThF4 were converted into their cor-responding oxides UO2.25 and ThO2 at 300 ˚C and 350 ˚C, and the conversion efficiency is over99.0%[2].

[1] Warf, J.C., Cline, W.D., Tevebaugh, R.D. (1954) Pyrohydrolysis in determination of fluoride andother halides. Anal. Chem. 26(2): 342-346.[2] Xiaoyu, D., Xiaobei, Z., Yulong, S., Yuxia, L., Lan, Z. (2014) Pyrohydrolysis of uranium tetraflu-oride and thorium tetrafluoride. J. Nucl. Radiochem. 36(3):181-185.


China

Primary author: Dr LIU, Yuxia (Shanghai Institute of Applied Physics, Chinese Academy of Sci-ences)

Co-author: Prof. ZHANG, Lan (Shanghai Institute of Applied Physics, Chinese Academy of Sci-ences)

Presenters: Prof. ZHANG, Lan (Shanghai Institute of Applied Physics, Chinese Academy of Sci-ences); Dr LIU, Yuxia (Shanghai Institute of Applied Physics, Chinese Academy of Sciences)


International Co … / Report of Contributions Performance of Elastomer Seals in …


Performance of Elastomer Seals in Transport andStorage Casks

Elastomer seals are widely used as barrier seals in containers for low and intermediate level ra-dioactive waste and for spent fuel transportation casks. In addition, they are also used for spentfuel storage and transportation casks (dual purpose casks (DPC)) as auxiliary seals to allow leakagerate measurements of metal barrier seals for demonstration of their proper assembling conditions.Depending on the area of use, the rubber materials have to demonstrate proper sealing perfor-mance with regard to mechanical, thermal and environmental conditions as well as irradiationduring the entire operation period. Concerning DPC, degradation effects should be limited in away that, for example, effects from potentially released decomposition elements may not harme. g. metal barrier seals. Leakage rate measurements should be possible also after long interimstorage periods prior to subsequent transportation.

Because of the complex requirements resulting from the various applications of containers for ra-dioactive waste and spent nuclear fuel, BAM has initiated several test programs for investigatingthe behaviour of elastomer seals. Experiments concerning the low temperature performance downto -40℃ and the influence of gamma irradiation have been started first. Currently also the thermalaging behaviour of elastomer seals is examined. The applied methods are e.g. Dynamic MechanicalAnalysis, measurement of hardness, Compression Set and Compression Stress Relaxation. Exem-plary test results are shown for selected rubber materials. Their major relevance concerning safetyaspects is discussed in this paper.Furthermore, materials testing is accompanied by the development of finite element (FE) models tosimulate the seal behaviour by using the FE code ABAQUS®. At first this shall enable the simula-tion of specific laboratory test configurations containing elastomer seals and finally the simulationof complete lid closure systems under specific operation or accident conditions. In a first step, ba-sic compression and tension tests were carried out. Test data were used to identify the constitutivebehaviour and find parameters for material models already implemented in the computer code. Ba-sically, the investigated rubber-like materials show hyperelastic behaviour with additional effects.


BAM Federal Institute for Materials Research and Testing

Primary author: Dr JAUNICH, Matthias (BAM Federal Institute for Materials Research and Test-ing)

Co-authors: KÖMMLING, Anja (BAM Federal Institute for Materials Research and Testing); DrWOLFF, Dietmar (BAM Federal Institute for Materials Research and Testing); Dr VÖLZKE, Holger(BAM Federal Institute for Materials Research and Testing); WEBER, Mike (BAM Federal Institute forMaterials Research and Testing); Dr ZENCKER, Uwe (BAM Federal Institute for Materials Research andTesting)

Presenter: Dr JAUNICH, Matthias (BAM Federal Institute for Materials Research and Testing)


International Co … / Report of Contributions Technological development to sup …


Technological development to support a change inUnited Kingdom’s strategy for management of spent

AGR oxide fuel

The United Kingdom’s Nuclear Decommissioning Authority has announced its preferred strategyfor managing spent oxide fuel arising from the EdF Energy fleet of Advanced Gas-cooled Reactors(AGR). This is to reprocess the optimum quantity of fuel at the Thermal Oxide Reprocessing Plant(THORP), Sellafield and place the remaining fuel and all future arisings into interim storage inthe THORP Receipt and Storage (TR&S) facility fuel ponds pending geological disposal; storageperiods up to 80 years are likely. Sellafield Ltd. is responsible for completing the technical basisfor interim storage of AGR fuel and making a safety case to deliver this strategy.

The imminent transition from buffer storage of fuel pending reprocessing to interim storage with-out the contingency of reprocessing of fuel that might fail (defined as penetration of the fuelcladding) during storage requires a more thorough technical basis. Therefore, Sellafield Ltd. hasdeveloped a programme of technical work to: increase confidence in the storage regime; demon-strate that corrosion models are applicable to cladding after higher irradiation and to changes infuel design; understand the efficacy of corrosion inhibition at higher temperatures than previouslystudied; characterise failed fuel; and develop methods for fuel condition monitoring.

The current technical underpinning for interim storage of AGR fuel is primarily empirical, based onsafe storage in sodium hydroxide dosed pond water of a large quantity of fuel, including some fuelstored for a period of nearly 25 years. Post storage hot cell examination of a sample from this longstored population is currently ongoing as a demonstration of the storage regime. Furthermore, dueto the continued evolution of the AGR fuel design and reactor operations, it is also necessary toensure that the storage experience accrued is applicable to modern fuel discharges that will makeup the majority of the storage inventory. Post irradiation examination of modern high burn-upfuel has also been completed to address changes in irradiated cladding properties resulting fromchanges in reactor operation.

This paper summarises the existing technical basis to support the transition from reprocessing tointerim storage, the process of developing a programme of work to identify and address gaps, andthe latest results available from the ongoing post irradiation and post storage examination work.


United Kingdom

Primary author: Mr KYFFIN, John (Sellafield Ltd)

Co-author: Mr HILLIER, Andy (Sellafield Ltd)

Presenter: Mr KYFFIN, John (Sellafield Ltd)


International Co … / Report of Contributions Aspects of Spent Fuel Behavior As …


Aspects of Spent Fuel Behavior Assessment forTransport Packages

Transport packages for spent nuclear fuel have to be assessed with respect to specific transportconditions which are defined in the safety regulations of the International Atomic Energy Agency.In general, gastight fuel rods constitute the first barrier of the containment system. The physi-cal state of the spent fuel and the fuel rod cladding as well as the geometric configuration of thefuel assemblies are important inputs for the evaluation of the package safety under transport con-ditions. The objective of this paper is to discuss the methodologies developed by BAM for theassessment of spent fuel behavior within the approval procedure of German spent fuel packagedesigns. In particular, cracks or failures in the fuel rod cladding can occur under transport con-ditions. These defects can cause the release of gas, volatiles, fuel particles or fragments into thecavity and have to be considered properly in safety analysis. Another spent fuel related issue isthe transport of defective fuel rods. One concept is to use special quivers which can be handledlike fuel assemblies. This concept requires additional assessments concerning drying, sealing andthe mechanical and thermal design of such quivers. The package as a mechanical system is char-acterized by a complex set of interactions, e. g. between the fuel rods within the assembly as wellas between the fuel assemblies, the basket, and the cask containment. This complexity togetherwith the limited knowledge about the material properties and the variation of the fuel assembliesregarding cladding material, burn-up and the operation history makes an exact mechanical analy-sis of the fuel rods nearly impossible. The simplified approaches to consider conservatively spentfuel behavior currently accepted by BAM are presented here.


Germany

Primary author: Dr LINNEMANN, Konrad (BAM Federal Institute for Materials Research andTesting)

Co-authors: Dr ROLLE, Annette (BAM Federal Institute for Materials Research and Testing); DrDROSTE, Bernhard (BAM Federal Institute for Materials Research and Testing); Dr WILLE, Frank (BAMFederal Institute for Materials Research and Testing); Dr MÜLLER, Lars (BAM Federal Institute for Ma-terials Research and Testing); Dr BALLHEIMER, Viktor (BAM Federal Institute for Materials Researchand Testing)

Presenter: Dr LINNEMANN, Konrad (BAM Federal Institute for Materials Research and Testing)


International Co … / Report of Contributions Development of a U.S. Rail Transp …


Development of a U.S. Rail Transport Capability forSpent Nuclear Fuel and High-Level Waste

The Department of Energy (DOE) is laying the groundwork for implementing an integrated nuclearwaste management disposition system. This includes preparing for future large-scale transport ofspent nuclear fuel (SNF) and high-level radioactive waste (HLW) since transport will be a necessarycomponent of any integrated nuclear waste management disposition system. DOE continues toplan for and develop options for decision-makers on the design of an integrated nuclear wastemanagement disposition system. A significant component of this integrated disposition system isthe development of a rail transport capability for SNF and HLW as described in this paper.

The Association of American Railroads (AAR) has published a technical standard developed specif-ically for railcars used during transport of High-Level Radioactive Material (HLRM): PerformanceSpecification for Trains Used to Carry High-Level Radioactive Material, Standard S 2043. AARdefines the term HLRM to include SNF and HLW.

HLRM will be shipped in transport casks certified in accordance with 10 CFR Part 71 by the NuclearRegulatory Commission (NRC). The NRC has certified transport cask designs supplied by variousmanufacturers. These rail transportation casks will weigh between approximately 82 and 156 tons(74 and 141 metric tons) when loaded; additionally, each cask will be attached to the railcar by acradle (often called a “skid”) that is expected to weigh between 10 and 20 tons (9 and 18 metric tons).No existing railcars have been approved as AAR S-2043 compliant for shipping these commercialNRC certified casks. Therefore, new railcars that meet S-2043 will need to be designed, tested,approved and fabricated to transport HLRM over the railroad infrastructure in the United States.

The project is being implemented in three phases. Phase 1 is Mobilization and Conceptual Designduring which conceptual designs for both the cask and buffer railcars will be developed. Phase2 is Preliminary Design during which the preliminary designs of the cask and buffer railcars willbe completed. Phase 3 is Prototype Fabrication during which one prototype cask railcar and twoprototype buffer railcars will be fabricated. The total duration for all 3 phases is estimated to beabout 4 years. A subsequent project is planned to perform the testing and obtain approval of thecask and buffer railcars from the AAR.


USA/ Department of Energy

Primary author: Dr SCHWAB, Patrick (U.S. Department of Energy)

Co-authors: Dr FELDMAN, Matthew (Oak Ridge National Laboratory); Ms BATES, Melissa (U.S.Department of Energy); Mr DAM, Scott (TechSource, Inc.); Dr MAHERAS, Steven (Pacific NorthwestNational Laboratory); Mr REICH, Willaim (Oak Ridge National Laboratory)

Presenter: Dr SCHWAB, Patrick (U.S. Department of Energy)


International Co … / Report of Contributions Major consideration and scenario …


Major consideration and scenario development insafety assessment for interim spent fuel storage

facility on Beyond DBA

Korea has 19PWR reactors and 4 PHWR reactors, and the storage capacities of spent nuclear fuelwill become insufficient. Korea is performing the expansion project through high density storagerack, its trans-shipment to the adjacent reactor facility. Korea government has the plan to com-plete the construction of interim storage facility by 2024 for spent fuel management.After Fukushima accident, the safety and integrity of spent nuclear fuel storage facility is moreimportant. It is important to develop the various accident scenario consequences of different typespent fuel storage based on Fukushima Accident, that is Beyond DBA. Fukushima Nuclear powerplant has three type spent fuel storage facility - on site wet-storage facility, off site wet-storagefacility, dry storage facility. According to the storage type, the impact of earthquake and thunamiis different.In this paper, it will be analysis and focus on the storage facility of spent nuclear fuel in Fukushimapower plant. The on-site wet type storage facility of 1~4 Fukushima power plant, it will investigatethe different reason on the impact of accident.In addition, we have to take the provision and safety assessment for Beyond DBA like Fukushimaaccident, because Korea has the plan to construction interim spent fuel storage facility by 2024,Performing safety assessment, the major factors are the nuclide leakage radius for safety assess-ment, the damage rate of spent fuel, the extinction rate outside of plume in safety assessment byGoldSim Module.The result of safety assessment is different about 2~10 times depending on the major factor above.So we have to discuss about this major considerations and develop this project after taking reliabledata for safety assessment.


Republic of Korea, KEPCO E&C

Primary author: Mr LEE, Dongjin (KEPCO E&C)

Presenter: Mr LEE, Dongjin (KEPCO E&C)


International Co … / Report of Contributions The Status of Spent Fuel Managem …


The Status of Spent Fuel Management and theProcess of Building Public Opinion in Korea

This study examined the current status of spent fuel management in Korea and the process ofbuilding public opinion that is currently underway to determine the final national policy for spentfuel.Spent fuels generated in the PWR plants are stored in a spent fuel pool for each unit. To date, al-most all PWR plants continue to implement measures to address the lack of storage capacity suchas installing storage racks additionally, installing high-density storage racks or transferring spentfuels to the spent fuel pool of other neighboring units. Spent fuels generated in PHWRs which areWolsong Units are stored in spent fuel pools for longer than six years and then transferred to thedry storage facility on site.The Nuclear Energy Promotion Commission decided on the “Execution Plan for Spent Fuel Man-agement Plan” in the 2nd meeting held on November 20, 2012 and setting a direction for developinga spent fuel management plan as follows: (1) put safety first (2) develop short-, mid- and long-termmanagement plans (3) prepare supporting measures which can be accepted by the general publicto ease the burden borne by future generations and local residents. At the same time, it was de-cided to form and run a public engagement commission so as to develop a management plan whilesecuring the highest possible level of acceptance.On October 30, 2013, the Public Engagement Commission on Spent Nuclear Fuel Management waslunched and embarked on public engagement activities with an aim of presenting recommenda-tions on the method for managing spent fuel to the government by late 2014. Therefore, a nationalpolicy, strategy and management plan for spent fuel will be presented in detail in the Basic Planfor Radioactive Waste Management where the results of above-mentioned public engagement ac-tivities will be incorporated.This study is expected to transfer experience of and lessons from determining national policy tocountries that have not established their final management policy for spent fuel, based on Korea’ssolutions to shortage of spent fuel pools and the formation of national consensus.


Korea/KINS

Primary author: Dr KIM, SUNG IL (Korea Institute of Nuclear Safety)

Co-authors: Mr LEE, Byungsoo (KINS); Mr YOOK, Dae-Sik (Korea Institute of Nuclear Safety); MrCHEONG, Jaehak (KINS)

Presenter: Dr KIM, SUNG IL (Korea Institute of Nuclear Safety)


International Co … / Report of Contributions Zr recovery process using electror …


Zr recovery process using electrorefining for thetreatment of cladding hull wastes

Zirconium (Zr)-alloys such as Zircaloy and Zirlo have been widely used as cladding tube mate-rials for nuclear fuel assembly due to its low absorption cross-section to thermal neutrons, highhardness, ductility, and corrosion resistance. The Zr-alloy cladding hull wastes are expected to begenerated from the pretreatment step of recycling technologies such pyroprocess or of long-termstorage technologies for used nuclear fuel. In the case of PLUS-7, which is a fuel assembly usedfor Korean standard PWR, the amount of the cladding hull waste will reach about 2.5 tons for thetreatment of 10 tons of used nuclear fuel. In addition, the waste level will supposedly exceed anintermediate-level due to the fuel residue on the surface, fission products by pellet-cladding inter-action (PCI), and irradiation by neutron.In this regard, this paper demonstrates Zr electrorefining process for the treatment of cladding hullwastes, thereby minimizing the amount of intermediate-level waste and recycling the recoveredZr. The Zr electrorefining has been mainly studied for LiCl-KCl based molten salts. However, apowder-type deposition characteristic of Zr in chloride-based molten salts is regarded as an obsta-cle for the realization of electrorefining process for a commercial-scale hull treatment because ofsignificant salt incorporation in the deposit. On the other hand, all-fluoride based salts are knownto form a coherent or dendritic Zr, which is preferred for a high throughput Zr production. How-ever, there exist several challenges associated with high corrosivities, salt waste treatment, etc.Therefore, we examined the electrochemical behavior change of Zr in chloride-based molten saltsby adding fluoride compounds. Also Zr electrorefining experiments were performed to investigatethe morphological feature of Zr deposit in the mixed chloride-fluoride molten salts.


South Korea/Korea Atomic Energy Research Institute

Primary author: Dr LEE, Chang Hwa (Korea Atomic Energy Research Institute)

Co-authors: Mr KANG, Deok Yoon (Korea Atomic Energy Research Institute); Dr PARK, GeunIl (Korea Atomic Energy Research Institute); Dr KANG, Kweon Ho (Korea Atomic Energy ResearchInstitute); Dr JEON, Min Ku (Korea Atomic Energy Research Institute); Mr CHOI, Yong Taek (KoreaAtomic Energy Research Institute)

Presenter: Dr LEE, Chang Hwa (Korea Atomic Energy Research Institute)


International Co … / Report of Contributions PRELIMINARY THERMAL MARG …


PRELIMINARY THERMAL MARGIN COMPARISONBETWEEN A SIMPLE SF MODEL AND A COMPLEX

SF MODEL FOR A DRY STORAGE CASK

It is expected that amount of nuclear spent fuels will exceed the capacity of temporary storagetanks in each nuclear power plant in Korea after several years. Accordingly, the industry demandfor storage cask commercialization has been increased. According to the demand, KINS(Korea In-stitute of Nuclear Safety) started to develop a thermal technical guidelines including CFD analysisas a development of technical guidelines for storage cask licensing. Given the growing industryneed to store nuclear spent fuel of increasingly higher burn-ups and heat loads in dry storagecasks, a conservative calculation model, for example, a homogenized model, is being substitutedfor a complex model, for example, rod based model, to secure more thermal margin.In this study, there were several assumptions that the PWR nuclear fuel assembly model was de-signed based on a CE-type-16-by-16 PWR nuclear fuel and the fuel assembly was cooled for 10years in a temporary storage tank, so the decay heat generation of the spent fuel assembly was938.8 W, and the PWR dry storage cask model was designed based on a commercial cask whichloads 24 PWR spent fuel assemblies. To compare the thermal margin between the homogenizedmodel and the rod model, 4 cases CFD model, a 1/4 cask homogenized model, a 1/8 cask rod model,one spent fuel assembly homogenized model and one spent fuel assembly rod based model, werecalculated.Consequently, the comparison of thermal margins between two calculation models was reviewedand valuable physical parameters were collected after this CFD thermal analysis. And it is expectedto contribute an establishment of CFD thermal analysis methodology using the complex analysismodel.


Korea / Korea Institute of Nuclear Safety

Primary author: Dr CHA, Jeong Hun (Korea Institute of Nuclear Safety(KINS))

Co-author: Mr YOOK, Dae-Sik (Korea Institute of Nuclear Safety)

Presenter: Dr CHA, Jeong Hun (Korea Institute of Nuclear Safety(KINS))


International Co … / Report of Contributions Study on Temperature Estimation …


Study on Temperature Estimation Method of PWRSpent Fuel Cladding in Dry Storage

In Japan, a need for additional spent fuel storage capacity is increasing due to a decrease in theresidual storage capacity of at reactor (AR) spent fuel pools. The Japan’s first away from reactor(AFR) dry storage facility for spent fuel, the Recyclable Fuel Storage Center in Mutsu-shi, Aomori,has completed its construction, and dry storage using Dual Purpose Casks (DPC) in this facility isunder safety review based on the new regulatory requirements. Both spent PWR fuels and spentBWR fuels will be stored there.In the case of DPC storage, if there is no mechanical impact such as DPC overturning or fluc-tuation of the inter-lid pressure during the storage period, integrity of DPC and stored fuel willbe considered to be maintained throughout the storage period, and DPC will be shipped withoutopening DPC lids for pre-shipment inspection. So the evaluation of integrity of stored spent fuelis one of the key technologies for DPC storage. For dry storage, hydride reorientation and creep offuel cladding are considered as main deterioration mechanisms, and both depend on the claddingtemperature history. Although the investigations for the other aging issues are necessary, in thispaper we focus on the temperature change during the storage and explain an approach for fuelcladding temperature estimation by measuring the surface temperature of DPC.Japan has experiences of AR BWR spent fuel dry storage with the integrity check of the storedBWR spent fuels. Meanwhile, due to the lack of PWR spent fuel dry storage experience, the elec-tric utilities are preparing a long-term storage test campaign in advance of the start of actualstorage at the Recyclable Fuel Storage Center. The regulator has participated in this campaignfrom the planning phase, and developed an analysis tool with the purpose for comprehension ofstored spent fuel cladding temperature history. This tool aims to estimate the cladding tempera-ture not by direct measurement but through measurement of the surface temperature distributionof the test container. Validation of the tool by comparing the analytical results to the measuredtemperature distributions of the test container in the preliminary heat-transfer test was conducted.The analytical results well agreed with the measured temperature distributions, which support theconclusion that the fuel cladding temperature could be estimated from the surface temperaturedistribution of the test container.


Japan/Nuclear Regulation Authority

Primary author: Mr YAGIHASHI, HIDEKI (Nuclear Regulation Authority)

Co-authors: Mr MASAKIYO, HISHIDA (Nuclear Regulation Authority); Mr AKAMATSU, MIKIO(Nuclear Regulation Authority)

Presenter: Mr YAGIHASHI, HIDEKI (Nuclear Regulation Authority)


International Co … / Report of Contributions Dry Storage of Irradiated Nuclear …


Dry Storage of Irradiated Nuclear Material in DualPurpose Cask at the JRC-Ispra Site

The Joint Research Centre (JRC) is currently involved in a Decommissioning and Waste Manage-ment (D&WM) Programme to progressively eliminate its historical liabilities related to the oldnuclear installations and radiological waste used/generated in the past at Ispra site, inter alia: nu-clear research reactors, hot-laboratories, radiochemical facilities and a variety of plants to treatand store the liquid and solid radiological waste. In the frame of the Programme, the IrradiatedNuclear Material (INM) accumulated over 40 years of research activities is foreseen to be retrieved,repacked and temporarily stored on-site by means of dry storage solution in one or more “all-metal” Dual Purpose Cask (DPC).The inventory of INM is quite various; the nuclear material is mainly constituted by “exotic” fuelitems irradiated in experimental rigs with a broad range of burn-up in the ESSOR research reactorduring ‘70 and ‘80. The material is indeed characterised by different enrichments, physical forms,claddings and dimensions. In addition, pins (and segments of them) of BWR and PWR comingfrom German and Italian nuclear power plants, used to perform destructive analyses and parti-tioning of fission products and minor actinides with innovative solvent extraction processes, areincluded in the inventory of the INM to be transferred into the DPC. The INM will be retrievedfrom the current storage locations, (i.e. dry pits in hot cell facility and reactor decay pond) andrepacked in intermediated cylindrical containers before the transfer into the casks. The DCP willbe finally transferred in a dedicated facility 400 m far from ESSOR reactor in the radioactive wastemanagement area of the Centre.The JRC intends on one hand to use state-of-the-art pre-designed DPC body with a bespoke innercanister that shall be able to fit INM features; on the other hand two main options are still underdiscussion to identify the facility that will host the cask(s) at the JRC Ispra site. The options canbe briefly summarised as follows: i) the DPC will be stored in a new dedicated facility; ii) the DPCwill be stored in a facility with other LLW or ILW. In both cases the facility will be designed andbuilt taking into account the security requirements for Category II nuclear material or higher.


European Commission, Joint Research Centre, Nuclear Decommissioning Unit

Primary author: Dr MAZZUCCATO, Matteo (European Commission, Joint Research Centre, Nu-clear Decommissioning Unit)

Co-authors: Dr ZANOVELLO, Flavio (European Commission, Joint Research Centre, Nuclear De-commissioning Unit); Dr KIRCHNER, Thomas (European Commission, Joint Research Centre, NuclearDecommissioning Unit)

Presenter: Dr MAZZUCCATO, Matteo (European Commission, Joint Research Centre, Nuclear De-commissioning Unit)


International Co … / Report of Contributions Lessons Learned from a Review of …


Lessons Learned from a Review of InternationalApproaches to Spent Fuel Management

Worldwide, slow progress in the deployment of geological disposal facilities and reduced use ofreprocessing has led to the need to increase the inventory and duration of spent fuel storage. Thishas led to a range of approaches to the management of spent fuel in storage being adopted in dif-ferent countries. A review has been undertaken to identify learning from the different approachesadopted internationally with a view to informing decision making relating to spent fuel manage-ment.

The review surveyed current spent fuel storage and disposal practices, standards, trends and recentdevelopments. National strategies for spent fuel storage and disposal in 16 countries were surveyedand more detailed studies were carried out into the evolution of spent fuel storage and disposalstrategies and activities in 4 countries.

As far as spent fuel management is concerned, the review highlighted that:

• Spent fuel management should be aligned to national policy and strategies for the final disposi-tioning of the fuel.• Given the long timeframes associated with geological disposal facility site selection and the man-agement life-cycle associated with nuclear fuel, selection of national spent fuel storage arrange-ments should reflect the need for efficiency of delivery of the whole spent fuel management strat-egy; it should ensure unduly increased costs due to expedient short term focus are avoided.• Commercial and financial arrangements should ideally be constructed to ensure that, at eachstage of the spent fuel lifecycle, spent fuel management decisions do not unnecessarily precludefuture management options. This will minimise the constraints placed on future fuel handling,packaging and disposal activities.• Having the capability to allow for extended storage of spent fuel, either at reactor sites or at acentralised facility, potentially over periods of several decades, may give increased flexibility inthe design of future packaging or disposal concepts.

On a technical level, very long-term storage of spent fuel over 100 years or more using existingtechnologies, or foreseeable evolutions of them, is technically feasible and credible. The use ofmultiple approaches to fuel storage, and continued evolution of the design of storage facilities,indicates that there is no single best storage technology and that local factors such as existinginfrastructure, approach to fuel cycle management, existing experience/capability and short-termcash flow considerations all influence technology selection. Both wet and dry storage systemscontinue to receive regulatory approval and are generally considered to be acceptable.


United Kingdom

Primary author: Mr HAMBLEY, David (National Nuclea r Laboratory)

Co-authors: Dr LAFERRERE, Alice (National Nuclear Laboratory); Dr WALTERS, Steve (NationalNuclear Laboratory)

Presenter: Mr HAMBLEY, David (National Nuclea r Laboratory)


International Co … / Report of Contributions Design and construction work exp …


Design and construction work experience of InterimStorage Facility for Spent Fuels

Recyclable-Fuel Storage Company (RFS) has completed construction works of the storage buildingand fabrication works of several metal casks for spent fuels as the Japanese first off-site interimstorage operator. RFS are now undertaking the safety review under new regulation standards basedon the lesson learned by Fukushima-daiichi accident and preparing for the start of operation.This paper introduces the outline of the design and construction work experience for this facility.

This interim storage facility is intended to temporarily store the spent nuclear fuels of light waterreactors till they are reprocessed. The building has a storage capacity of up to 288 dry metal casks.The storage building has natural air cooling systems for removing heat from the filled casks. Thecenter part of the storage area has high-height exhausting part for effective ventilation. On theother hand, the other part of the storage area has low-height ceiling because metal casks are movedby the air pallet transporter instead of overhead crane.Metal casks are designed so that the basic safety feature of the casks such as containment, shielding,subcriticality and heat removal are safely maintained during storage periods of 50 years. Negativepressure is kept for inner part of casks and positive pressure is kept for the part between two lids.The casks are also filled with helium gas so as to convey heat effectively and to maintain the fuelintegrity. These casks are used for dual purpose i.e. both storage and transport. For transport,tertiary lid and impact limiters are also set on the casks. During storage, visual inspection, pres-sure monitoring between the double lids, temperature checking on the casks and monitoring ofradiation dose of inner storage building, etc. are carried out.The establishment permit for spent fuel storage and approval of design and construction methodswere issued in 2010. After the approval, RFS started the construction of the spent fuel storagefacilities. Construction works of storage building including installation of the electrical and me-chanical equipment were completed in August 2013. Fabrication of several metal casks was alsocompleted with various inspections as transport and storage casks.Currently, RFS has been undertaking NRA review under new regulation standards. As prepara-tion work for the start of the operation, RFS carried out the training of the handling casks usingmock-up cask, air pallet transporter, handling equipment and transport vehicle.


Japan

Primary author: Mr TAKAHASHI, Masahiko (Recyclable-Fuel Storage Company)

Co-authors: Mr CHIKAHATA, Hideyuki (Recyclable-Fuel Storage Company); Mr ISHIKAWA, Tat-suya (Recyclable-Fuel Storage Company)

Presenter: Mr ISHIKAWA, Tatsuya (Recyclable-Fuel Storage Company)


International Co … / Report of Contributions History and current status of spent …


History and current status of spent fuel managementat Dukovany NPP

The first Czech Nuclear Power Plant, NPP Dukovany, has almost reached 30 years of safe operation.Significant amount of spent fuel has been produced, though a number of fuel cycle improvementshelped to increase fuel performance and decrease its quantity. During the operation years, spentfuel management had to respond to many changes, both political and technical. We had to leavean option to send the spent fuel back to the country of origin and to find our own way how todeal with spent fuel. After a provisional measure of storing the fuel in neighbouring country weadopted spent fuel storage system based on dual purpose metal casks stored in the plant. Newchallenge is a future strategy - long term storing, final disposal or reprocessing for re-using.


Czech Republic

Primary author: Mr GERŽA, Jiří (ČEZ, a.s., NPP Dukovany)

Presenter: Mr GERŽA, Jiří (ČEZ, a.s., NPP Dukovany)


International Co … / Report of Contributions Transport aspects of spent fuel ma …


Transport aspects of spent fuel management

There are different concepts for management of spent nuclear fuel, including wet storage in poolsand dry storage in welded canisters or in dual-purpose transport and storage casks. But all of themrequire transport of the spent fuel, at least transport to the final destination after the interim stor-age period. This presentation analyzes how this safe transport of spent fuel after several decadesof storage is reflected in the transport regulations and how it can be achieved for different storageconcepts.

For any transport package it must be ensured that the conditions of the packaging and contentsat the beginning of shipment meet the conditions the package design safety analysis is based on.For dry storage of spent fuel this requires that the ageing and the ageing management have to beconsidered as early as in the design phase for the storage and transport arrangements and have tobe considered for the package design approval and the approval of the storage site.

The need of consideration of ageing for packages to be stored for a long time before transportcan be derived from the current IAEA transport regulations already, but these regulations or thecorresponding guidance could be improved to strengthen this requirement, as is discussed in thepresentation.

On the other hand Germany has developed and applied a concept for transport related ageing as-sessment and management during dry storage of spent fuel in transport and storage casks. Thisconcepts consists of approval of the package design for transport as a prerequisite for transport andfor the interim storage license, followed by review and renewal of the certificate during interimstorage period and use of certificate at the end of the interim storage period, including pre-definedactions to be taken before the shipment of the DPC from the facility to the final destination. Expe-riences from this regulatory concept are presented in the second part of the presentation, as wellas consequences for other storage concepts.


Germany

Primary author: Dr REICHE, Ingo (German Federal Office for Radiation Protection (BfS))

Co-authors: Mr NITSCHE, Frank (Federal Office for Radiation Protection (BfS)); Mr BÖRST, Frank-Michael(Federal Office for Radiation Protection (BfS))

Presenter: Dr REICHE, Ingo (German Federal Office for Radiation Protection (BfS))


International Co … / Report of Contributions Evaluation of Radiation Characteri …


Evaluation of Radiation Characteristics of SpentRBMK-1500 Nuclear Fuel Storage Casks During Very

Long Term Storage

The present spent nuclear fuel management concept in the Lithuania foresees dry spent nuclearfuel storage in the casks for 50 years. During this time the final management concept shall bedeveloped. Different options are under consideration, however due to various uncertainties infuture, there is a risk that these options can be not implemented in due time, therefore elongationof the spent nuclear fuel storage for a period over 50 years shall be also considered.Existing spent nuclear fuel storage facility at Ignalina Nuclear Power Plant site has been extendedfor several times and currently is filled up to its final capacity with 20 metal CASTOR® RBMK-1500and 100 reinforced concrete CONSTOR® RBMK-1500 casks.Modelling of radiation doses on the sidelong, upper and lower surface of theses dry storage casksand for certain distance during very long term storage has been performed using SCALE computercodes system. Obtained results showed that CONSTOR® RBMK-1500 cask has better shieldingproperties than CASTOR® RBMK-1500 cask. Material composition and thickness of the side wallsand bottoms of these casks are different, therefore total equivalent dose rates at casks side-wallsand bottoms differ from 2 to 7 times during storage periods up to 50 years. During very long termstorage period (up to 300 years) the total dose rate is significantly decreasing, the difference ofdose rates for different cask is increasing up to 60 times and after such period dose rate caused byneutrons is becoming dominant for the both casks.Description of the calculation model, accepted assumptions and obtained modelling results arepresented and discussed in the paper.


Lithuanian Energy InstituteNuclear Engineering Laboratory

Primary author: Mr SMAIZYS, Arturas (Lithuanian Energy Institute)

Co-authors: Mr NARKUNAS, Ernestas (Lithuanian Energy Institute); Mr POSKAS, Povilas (Lithua-nian Energy Institute)

Presenter: Mr SMAIZYS, Arturas (Lithuanian Energy Institute)


International Co … / Report of Contributions Development of long-term safety r …


Development of long-term safety requirements foran alternative design variant (KBS-3H) for spent fuel

disposal

In 2012, Posiva Oy submitted a construction licence for a spent fuel disposal facility to be con-structed at Olkiluoto, Finland. A safety case (TURVA-2012) was compiled to support the licenceapplication. The disposal concept is based on the KBS-3V method, where the spent fuel canistersare emplaced individually in vertical deposition holes. Posiva Oy is also studying, in collaborationwith its Swedish counterpart SKB, an alternative design variant, KBS-3H, where the canisters areemplaced horizontally in 100−300 m long deposition drifts. In order to compare these two alterna-tives, a safety case is being produced for the KBS-3H design. The main objective is to determinewhether KBS-3H can be shown to fulfil the long-term safety requirements with the same level ofconfidence as for KBS-3V. To this end, long-term safety related requirements specific to the KBS-3H design are being defined following Posiva’s requirements management system (VAHA). VAHAincludes five levels of requirements spanning from legal and stakeholders’ requirements (level 1)to safety functions for the individual barriers (level 2), performance targets (level 3), design re-quirements (level 4) and finally design specifications (level 5).

The level 1 requirements, since they stem from laws and regulations, are identical for both designs.At lower levels, the differences in the designs have an increasing effect on the details of the require-ments and design specifications. The set of release barriers is partly different in the two designs,as are the types and dimensions of the emplacement areas and their construction methods. Thedevelopment of the KBS-3H-specific requirements starts by defining the barriers of the KBS-3Hsystem and assigning safety functions for the individual barriers. The safety functions will thengive rise to performance targets, and subsequently to the more detailed requirements and spec-ifications at lower levels. The safety case for KBS-3H will then evaluate whether the horizontaldesign fulfils these requirements.

The requirement definition includes interesting aspects related to the fact that KBS-3H is beingdeveloped since decades in parallel to the reference design KBS-3V and it includes several novelsolutions and unique components not included in KBS-3V. The iteration among requirements for-mulation, safety assessment and design development is particularly visible in this project.


Finland

Primary author: Dr HAGROS, Annika (Saanio & Riekkola Oy)

Co-authors: Dr PASTINA, Barbara (Posiva Oy); SELROOS, Jan-Olof (Swedish Nuclear Fuel andWaste Management Company); Ms SNELLMAN, Margit (Saanio & Riekkola Oy)

Presenter: Dr HAGROS, Annika (Saanio & Riekkola Oy)


International Co … / Report of Contributions ASSESSEMENT OF 3D DEPLETIO …


ASSESSEMENT OF 3D DEPLETION CAPABILITIESOF MCNP6.1 FOR WWER TYPE FUEL BURNUP

CREDIT ANALYSIS

Currently two main approaches are applied to demonstrate criticality safety of spent fuel in theBack-End of the Fuel Cycle: fresh fuel approach and burnup credit approach. In the fresh fuelapproach, spent nuclear fuel has a pre-known precise nuclide composition that enables to developbounding approach to the criticality safety analysis in the Back-End of the Fuel Cycle. This ap-proach assures conservativeness of the analysis with respect to criticality safety, however, it limitsthe storage/cask capacity or upper limit of enrichment of the fuel assemblies to be loaded in thestorage/cask.

Burnup credit approach allows to dispose of above mentioned limitations, however, it requiresknowledge of isotopic composition of spent fuel as a function of burnup and irradiation historywith high degree of confidence. To get a desired confidence extensive verification and validationactivities based on chemical assay analysis are needed. Usually chemical assay data represent cutsfrom several axial positions of certain fuel rods, therefore 3D depletion analytical capabilities arerequired to carry out full scope verification and validation activities.

This paper discusses results of the assessment of 3D depletion capabilities of MCNP6.1 code basedon WWER-440 fuel chemical assay data.


Nuclear and Radiation Safety Center of Armenian Nuclear Regulatory Authority

Primary author: Dr BZNUNI, Surik (Nuclear and Radiation Safety Center, Armenia)

Co-authors: Mr AMIRJANYAN, Armen (Nuclear and Radiation Safety Center, Armenia); Mr RAM-SEY, Jack (US Nuclear Regulatory Comission, USA); Mr BAGDASARYAN, Nairi (Nuclear and RadiationSafety Center, Armenia); Dr KOHUT, Peter (Brookhaven National Laboratory, USA)

Presenter: Dr BZNUNI, Surik (Nuclear and Radiation Safety Center, Armenia)


International Co … / Report of Contributions The Use of System Analysis and S …


The Use of System Analysis and System Engineeringto Inform the Implementation of an Integrated Waste

Management System

The Nuclear Fuels Storage and Transportation Planning Project, under the US Department of En-ergy Office of Nuclear Energy Fuel Cycle Technologies program, is developing tools and tech-niques, gathering data and information, and conducting analyses to inform future decisions re-garding the waste management system. Spent Nuclear Fuel (SNF) management system analysis,systems engineering, and decision analysis principles are being used to develop concepts for poten-tial waste management system configurations and inform future decisions regarding integration.The application of these techniques to this complex and challenging problem have been recognizedas being essential by the Blue Ribbon Commission for America’s Nuclear Future and the US Nu-clear Waste Technical Review Board. Flexibility, adaptability, phasing, and step-wise learning arekey considerations in the overall waste management system evaluations.

These studies are informing alternatives for managing SNF generated by the current fleet of lightwater reactors operating in the US. The objectives of the effort are to:• Provide quantitative information with respect to a broad range of SNF management alternativesand considerations• Develop an integrated approach to evaluating storage, transportation, and disposal options, withemphasis on flexible implementation options.• Evaluate impacts of storage choices on SNF storage, handling, and disposal options• Identify alternative strategies and evaluate with respect to cost and flexibility• Consider a broad range of factors including repository emplacement capability, thermal con-straints, re-packaging needs, storage and transportation alternatives, and impacts on utility oper-ations.


USA/Oak Ridge National Laboratory

Primary author: Mr HOWARD, robert (Oak Ridge National Laboratory)

Co-authors: Mr TRAIL, Casey (Argonne National Laboratory); Mr WHEELER, Jack (Department ofEnergy); Mr FORTNER, Jeff (Argonne National Laboratory); Dr JARRELL, Joshua (Oak Ridge NationalLaboratory); Mr NUTT, Mark (Argonne National Laboratory); Mr JOSEPH, Robby (Oak Ridge NationalLaboratory)

Presenter: Mr HOWARD, robert (Oak Ridge National Laboratory)


International Co … / Report of Contributions The High Burnup Confirmatory D …


The High Burnup Confirmatory Data Project

The United States (US) is interested in demonstrating the ability to safely store — for many decades— and then transport spent nuclear fuel (SNF). The potential need to store SNF for many decadeswill have a near-term and potentially significant impact on US nuclear plant licensing and opera-tions. While dry storage of lower burnup SNF [less than 45 gigawatt days per metric ton uranium(GWD/MTU)] has occurred in the US since 1986, dry storage of high burnup SNF has been morerecent. As of 2012, approximately 200 dry storage casks have been loaded with at least some highburnup SNF. Furthermore, almost all SNF being loaded in the US is now high burnup. While cur-rent knowledge indicates storage and transportation will not be a problem, high burnup SNF hasdifferent mechanical properties than lower burnup SNF, and industry needs additional data onhigh burnup SNF under typical conditions.

To assist in the collection of this data, the US federal government has initiated a High Burnup Con-firmatory Data Project (CDP) project to develop and implement a plan to load an instrumentedTransnuclear (TN)-32 bolted-lid cask with high burnup fuel and store the cask and fuel on an Inde-pendent Spent Fuel Storage Installation (ISFSI) for a period of ten years. The project is led by theElectric Power Research Institute (EPRI) and includes members from the US federal government,the US nuclear industry, and US national laboratories.

A Test Plan for the CDP has been developed to establish how data will be collected from a SNF drystorage system containing high burnup fuel. The high burnup fuel to be included in this projectincludes four different cladding types: standard Zircaloy-4, low-tin Zircaloy-4, Zirlo, and M5. TheTest Plan outlines the data to be collected; the high burnup fuel to be included; and the storagesystem design, procedures, and licensing necessary to implement the Test Plan. The CDP willinclude temperature and gas sampling as well as pre-characterization of fuel rods similar to theones being stored. The main goals of the project are to provide confirmatory data on the behaviorof high burnup fuel under typical dry storage conditions that can be used for model validationand potential improvement, provide input to future SNF dry storage cask designs, support licenserenewals and new licenses for ISFSIs, and support transportation licensing for high burnup SNF.


U.S. / U.S. Department of Energy

Primary author: Ms BATES, Melissa (U.S. Department of Energy)

Co-authors: Mr WALDROP, Keith (Electric Power Research Institute); Mr LARSON, Ned (U.S.Department of Energy)

Presenter: Ms BATES, Melissa (U.S. Department of Energy)


International Co … / Report of Contributions Risk Informing Interim Storage of …


Risk Informing Interim Storage of Spent NuclearFuel

The U.S. Nuclear Regulatory Commission (USNRC) risk-informs a variety of operating and newnuclear reactor regulatory activities. This practice is rooted in probabilistic risk assessment (PRA),a highly detailed, quantitative method. Significant resources are required to develop PRA mod-els for reactors, however. The USNRC’s Division of Spent Fuel Management (SFM) is developinga qualitative/semi-quantitative risk-informed framework for interim dry spent fuel storage. Thepurpose of this framework is to better enable the USNRC to enhance the efficiency and effective-ness of its dry storage regulatory activities, including actions to improve guidance, streamlinecasework activities, help assess modifications to dry storage facility designs and operations, andevaluate requests for exemptions to the regulations while maintaining adequate safety and secu-rity. The framework is intended to be flexible enough to incorporate risk insights from pilot PRAs,human reliability analyses, observations from operating experience, and expert opinion. Differenttypes of risk assessments were considered for risk-informing interim dry spent fuel storage. Thechallenges of creating an initial framework will be discussed, as dry storage relies significantly onpassive systems, and traditional PRAs have been developed for and applied to active systems, e.g.,operating reactors. The proposed framework is intended to risk inform the first licensing periodfor dry storage, including loading, transfer and storage operations and the associated regulatoryactivities. It is expected the framework will be expanded to include transportation and longer in-terim storage periods. As part of the proposed framework, the elements of defense-in-depth aswell as appropriate metrics for risk evaluation will be defined.


U.S. Nuclear Regulatory Commission (USNRC)

Primary authors: Mr GORDON, Matthew (U.S. Nuclear Regulatory Commission (USNRC)); MrCALL, Michel (U.S. Nuclear Regulatory Commission (USNRC))

Co-author: Mr RAHIMI, Meraj (U.S. Nuclear Regulatory Commission (USNRC))

Presenter: Mr GORDON, Matthew (U.S. Nuclear Regulatory Commission (USNRC))


International Co … / Report of Contributions Evaluation of sealing performance …


Evaluation of sealing performance of metal gasketused in dual purpose metal cask subjected to an

aircraft engine missile

The CRIEPI (Central Research Institute of Electric Power Industry) has executed several studyprograms on demonstrative testing for interim storage of spent fuel, related to metal cask storagetechnology to reflect in Japanese safety requirements for dry casks.If a metal gasket used in a dual purpose metal cask was subjected to high temperature for a longterm, the residual linear loads and total spring back distance of a metal gasket might decreasedue to the creep deformation of the outer jacket made of soft metal. Therefore, when the caskwould receive the severe mechanical force under hypothetical accidental condition, the sealingperformance of metal gaskets might be considerably affected.In this paper, to investigate of the integrity of the lid structure of the metal cask during the extremeimpact loads due to aircraft crash, two impact scenarios for aircraft engine crash onto the metalcask without impact limiters are considered for both, a vertical impact onto the lid structure and ahorizontal impact hitting the cask. The horizontal impact test using scale model engine of aircrafthas been executed and leak rate from the metallic gasket in the cask also measured at the impactin the test. The vertical impact onto the head of the full-scale metal cask has been also executed.The test cask was mounted on a supporting flame structure by the specific panel. The reactionforces were measured by six load cells installed between the panel and the supporting flame. Atthe impact in the test, the leak rate, inner pressure between the lids and displacement of the lidswere measured. From these experimental results, it seems that the loss of the inner pressure of thecask cavity may be avoided in the impact event with the horizontal and vertical orientation evenif the severe impact load was applied on to the metal cask due to aircraft engine crush. Moreover,in order to evaluate the deformation response of the metal gasket subjected to the accidental loadsmeasured during experiments, the impact analysis by LS-DYNA code was executed. As a result,it was found that the opening displacement of the gasket was negligible as compared with theevaluated spring back distance of metal gasket used for 60 years.


JAPAN

Primary author: Dr SHIRAI, KOJI (Central Research Institute of Electric Power Industry (CRIEPI))

Co-authors: Dr NAMBA, KOSUKE (Central Research Institute of Electric Power Industry (CRIEPI)); MrWATARU, MASUMI (Central Research Institute of Electric Power Industry (CRIEPI))

Presenter: Dr SHIRAI, KOJI (Central Research Institute of Electric Power Industry (CRIEPI))


International Co … / Report of Contributions Design Strategies for Direct Recyc …


Design Strategies for Direct Recycling of ACR-700Spent Fuel

Increasing the nuclear fuel burnup has a particular importance to improve the uranium utilization,to reduce the high level nuclear waste and to reduce the amount of plutonium in spent fuel perunit energy which improves the plutonium proliferation resistance.This work is focused on strategies for direct recycling of 700 MWe Advanced CANDU Reactor(ACR-700) spent fuel in CANDU-6 reactor. ACR-700 discharges the fuel with a significant amountof fissile isotopes (U-235, Pu-239 and Pu-241). Three strategies are considered for recycling thespent fuel of ACR-700. First strategy is recycling the ACR-700 spent fuel bundles directly inCANDU-6, since the two reactors have the same inner diameter of the fuel channel. Second strat-egy is removing the central pin (which has residual reactivity plenty of the Dysprosium fissionablepoisons) from the ACR-700 spent fuel and re-fabricating the spent fuel bundle into CANDU-6 fuelbundles using dry processing such as the DUPIC fuel cycle. Third strategy is removing the outerfuel pins (which have the fewer amounts of fissile isotopes) and the central pin and re-fabricatingthe rest two fuel rings into CANDU-6 fuel bundles using the dry processing.The calculations using the MCNPX code showed that the recycled spent fuel in CANDU-6 givesburnup around 3, 6.75 and 13.3 MWd/kgU for the three strategies, respectively. Normalizing theburnup on the all fuel in the spent bundle, additional burnup of 3, 6.5 and 6.9 MWd/kgU can beobtained for the three strategies respectively. This means that the third strategy gives the higherburnup from the spent fuel. Moreover in the third strategy, only half of the spent fuel bundleis re-fabricated into CANDU-6 fuel bundle which decreases the cost of re-fabrication. Knowingthat ACR-700 burns fuel to 20.5 MWd/kgU, the third strategy can increase the burnup of fuel byabout 34%. The calculations give acceptable power distributions on the fuel bundles for the threestrategies and acceptable coolant void reactivity compared to reference CANDU-6.


Atomic Energy Authority, ETRR-2Cairo, Egypt

Primary author: Dr MOHAMED, Nader (Atomic Energy Authority, ETRR-2, Cairo, Egypt)

Presenter: Dr MOHAMED, Nader (Atomic Energy Authority, ETRR-2, Cairo, Egypt)


International Co … / Report of Contributions CRIEPI’s studies on the SCC of the …


CRIEPI’s studies on the SCC of the canister for spentnuclear fuel

To put the concrete cask in practical use in Japan, stress corrosion cracking (SCC) of canister mustbe coped with. It is required to take measures for one or two of the three factors, i.e. weldingresidual stress, material, and environment, to cope with the SCC that may result in loss of thecontainment function of the canister. Prevention of loss of containment due to SCC of a canisterwas evaluated either by a method of comparing the amount of salt on the canister surface duringstorage with the minimum amount of salt to initiate rust and SCC or by a method of comparingthe wetting time of the canister surface under salty-air field environment with the lifetime of theSCC fracture of the canister material. We examine an application of the zirconia shot peening asthe residual stress improvement processing that is considered for prevention of SCC occurrence ofthe canister weld effectively. For the inspection of the canister integrity during storage, there areseveral candidate methods. We develop a technique to measure chlorine deposited on the surfaceof canisters by laser-induced breakdown spectroscopy. Furthermore, we propose the monitoringsystem using the temperature deference between the top and the bottom of the canister as a heliumleak sensor for the canister in storage. In order to analytically evaluate the change of the temper-ature deference between the top and the bottom of the canister during the leakage of helium gasfrom the canister, the unsteady state thermal hydraulics model which takes the change of densityof helium gas into consideration has been developed.


Japan/CRIEPI

Primary author: Mr WATARU, Masumi (CRIEPI)

Co-authors: Mr TAKEDA, Hirofumi (CRIEPI); Dr TANI, Junichi (CRIEPI); Dr SHIRAI, Koji (CRIEPI); MrETO, Shuzo (CRIEPI); Mr FUJII, Takashi (CRIEPI); Dr SAEGUSA, Toshiari (CRIEPI)

Presenter: Mr WATARU, Masumi (CRIEPI)


International Co … / Report of Contributions Huge non-spent potential of the “s …


Huge non-spent potential of the “spent fuel”- theways of its utilization

Nowadays nuclear power is based on the usage of exothermic nuclear processes. Actually it meansthat the thermal engine principle of heat conversion into mechanical work is used - with the re-spective thermodynamic limitation of maximal efficiency (by Carnot cycle). The only difference isthat heat is obtained due to nuclear fuel “burning” - instead of traditional fossil fuel burning. As aresult, the presently running “nuclear thermal machine” is worse than a traditional one (because ofthe security and safe concerns, its operating temperature is lower as compared with the presentlyattained working temperature in the traditional thermal engine). This “thermal inertia” of nuclearpower technology leads to small achievable efficiency (of about 30%) and, respectively, enormousthermal pollution.We develop the alternative, “cold” nuclear power technologies, which could supplement exother-mic nuclear power; at present stage they need additional investigation and promotion (are not thesubject of the present paper).The current situation in the nuclear power industry is the next: most of the nuclear energy isproduced by the thermal reactors working on enriched uranium as a fuel; the resulting thermalpollution is inadmissibly large (about 70%); the output is the highly radiating “spent fuel”, whichneeds expensive storage; its reprocessing generates “nuclear wastes”.The first of the two above-mentioned shortages can be eliminated by elaboration and implemen-tation of the cutting-edge technologies for secondary (low-potential) energy utilization. As forthe second one, we consider the best solution for the “spent fuel” coming from the light waterreactors - its re-use without constructive changes in the specially designed heavy water reactors.Reprocessing (especially using presently applied technologies) should be postponed for an indef-inite period (but still could be indicated - only for the eventually damaged fuel bundles). Such asolution ensures substantial cost-effective growth of nuclear power, reduces the threat of nuclearproliferation and terrorism, stimulates development and implementation of technologies for theactive storage of the spent fuel. Perfectly (hermetically) sealed spent fuel rods represent full-valuehigh-tech products and can be technologically used as “cold” (non-equilibrium) plasma generators,sources of heat, gamma-radiation etc. In such a way, the problem of nuclear wastes disappearsdefinitely.We develop for the spent fuel: on-site use for the condensation process intensification and microbi-ological sterilization; “distributed” energy-active storage, inclusively geological one; “cold” MHDpower generation and other technologies.


Republic of Moldova

Primary author: Mr BOSNEAGA, Iurie (Institute of Applied Physics of Academy of Sciences ofMoldova)

Presenter: Mr BOSNEAGA, Iurie (Institute of Applied Physics of Academy of Sciences of Moldova)


International Co … / Report of Contributions SAFETY CONSIDERATIONS FOR …


SAFETY CONSIDERATIONS FOR DUAL PURPOSECASK

AREVA TN, part of AREVA group, provides total system solutions for used fuel or radioactivewaste management, and comprehensive transportation services for the entire nuclear fuel cycle.

Since the ‘80s AREVA TN has developed an entire set of dual purpose casks, the TN®24 family.This cask family has been designed for the safe transportation and the dry interim storage of theused fuel and the radioactive waste. Key safety issues in both fields are the safe enclosure of theradioactive material, the safe removal of decay heat, securing nuclear criticality safety, limitationof radiation exposure to acceptable levels.

The TN®24 casks have been designed to meet the Type B(U) package requirements of the transportregulations issued by IAEA (International Atomic Energy Agency).In addition these dual purpose casks have been designed to meet all safety requirements imposedby national regulations as well as the site specific safety analyses.

The purpose of this paper is to present the solutions designed by AREVA for used nuclear fuel andradioactive waste management which are in operation in several Interim Storage Facilities through-out the world. This paper presents these experiences and broaches some considerations related tothe Safety Evaluation for transportation and for interim storage as well in normal condition andin hypothetical accident conditions.


FRANCE / AREVA TN

Primary author: Mr GARCIA, Justo (AREVA TN)

Presenter: Mr GARCIA, Justo (AREVA TN)


International Co … / Report of Contributions AREVA OPERATIONAL EXPERIE …


AREVA OPERATIONAL EXPERIENCE IN INTERIMDRY STORAGE

For more than 50 years, AREVA TN, part of the AREVA Group, has offered a complete range oftransport and interim storage solutions for radioactive materials throughout the entire nuclear fuelcycle.

Interim storage of used nuclear fuel is a reliable intermediate solution while waiting for a deci-sion concerning disposal sites or recycling. Intermediate storage is safe as shown by importantindustrial feedback and the operational records.

AREVA has developed different Used Nuclear Fuel dry storage solutions worldwide. Since the‘80s AREVA TN has developed an entire set of transport and storage casks, the TN®24 family.Composed of 20 types of casks, this family has a wide range of capacities from 21 PWR to 97 BWRused fuel elements. These casks are currently in operation in Europe, the United States and Japanto safely provide interim storage of used fuel elements. In parallel, another type of interim storagesystem, the NUHOMS®, has been developed mainly for the US market. To date NUHOMS® is thedry storage solution of choice of more than 50% of U.S. nuclear facilities.

Up to now more than 1,000 dry storage systems designed by AREVA have been loaded throughoutthe world, representing a significant track record.

The purpose of this paper is to present an overview of the solutions designed by AREVA for usednuclear fuel management, and specifically the different systems currently in operation in severalInterim Storage Facilities throughout the world. This paper broaches as well some considerationsrelated to the Safety Evaluation and Management program.


FRANCE / AREVA TN

Primary author: Mr GARCIA, Justo (AREVA TN)

Presenter: Mr GARCIA, Justo (AREVA TN)


International Co … / Report of Contributions Egyptian Proposed Strategy of Spe …


Egyptian Proposed Strategy of Spent FuelManagement from Nuclear Power Reactors

A proposed policy and strategy of the Arab Republic of Egypt towards the management of spentfuel from the nuclear power reactors is presented in this paper. The proposed Egyptian strategysupports the free international trade of nuclear materials and services and adheres to Non Prolif-eration Treaty (NPT) and other institutional frameworks aimed at promoting the peaceful use ofnuclear power in all countries.The accident of the Fukushima Daiichi nuclear power plant prompted an international review ofnuclear and spent fuel safety which led to modifications of strategies and revision of internationalguidance — e.g. through the publication of a new Specific Safety Guide,Storage of Spent NuclearFuel (IAEA Safety Standards Series No. SSG-15, Vienna, 2012).This work accounting the main target of the spent fuel management to implement the applicationcodes and standards of safety, quality assurance in that regards .From the point of view planning.The Nuclear Power Plants Authority (NPPA) in Egypt is responsible for a more stable short andlong strategy of spent fuel management. Egypt has undertaken the decision to adopt an open fuelcycle for the first nuclear power plant, i.e. no reprocessing of spent fuel. NPPA would developat an early stage a conceptual plan describing all important steps leading to the final disposal ofspent fuel and radioactive waste in Egypt utilizing fully the national and international experienceand the capabilities of international cooperation.


Egyptian Nuclear Power Plants Authority (NPPA)- Cairo, Egypt

Primary author: Ms AFIFI, Madiha (Abdel Samie Abdel Rahman)

Presenter: Ms AFIFI, Madiha (Abdel Samie Abdel Rahman)


International Co … / Report of Contributions Development of the Licensing Pro …


Development of the Licensing Procedure andRegulatory Framework for the Spent Fuel Storage

Cask in Korea

The future national policy for spent fuel management of Korea will be chosen based on the resultof the public engagement, taking into consideration the national/international trends on policyand technology development. Based on this, the public engagement has started with the launchof the Public Engagement Commission on SNF management (PECOS) in 2013.Regardless of the recommendation report of PECOS, Spent fuel storage cask is necessary in Korea.Therefore, Concrete storage cask and dual purpose metal cask are now under development by theKORAD (KOrea RADioactive waste agency) for use in the interim storage facility in the future.Unfortunately, there is no independent licensing procedure on the spent fuel storage cask in nu-clear safety act because it is considered as the main safety equipment in the interim storage facility.The aim of this study is to develop the licensing procedure for the storage cask in nuclear safetyact and to develop the revision draft of nuclear safety act on the interim storage facility addition-ally. Once this independent licensing procedure on the spent fuel storage cask is introduced tothe nuclear safety act, it is expected that developer or operator can develop and commercialize thestorage cask in Korea.Independent licensing procedure for the spent fuel storage cask in nuclear safety act was developedand it was composed of the design approval of storage cask, administrative applying procedure,technical criteria, manufacture inspection, manufacture inspection criteria, periodical inspection,and periodic inspection criteria, etc. Especially, the safety case and aging management programwas also developed to introduce the nuclear safety act as increasing the interest on the transportafter long term dry storage.In order to develop the revision draft of nuclear safety act on the interim storage facility, the li-censing procedure of interim spent fuel storage facility was separated from the current ‘disposalfacility, etc.’ in accordance with the article 63 of nuclear safety act independently. This procedurewas composed of the permit for construction and operation of interim storage facility, criteria forpermit, inspection criteria, periodic safety review, and decommissioning criteria, etc. the detailcontents of these requirements were also developed to introduce the regulatory framework for theinterim storage facility.


Korea, Republic of

Primary author: Mr YOOK, Dae-Sik (Korea Institute of Nuclear Safety)

Presenter: Mr YOOK, Dae-Sik (Korea Institute of Nuclear Safety)


International Co … / Report of Contributions Development of SCC Resistant Ca …


Development of SCC Resistant Canisters for SpentFuel Storage and Transport

Mitigation of Saline Air Induced Stress Corrosion Cracking (SCC) is one of the big issues for ag-ing management of spent fuel storage canisters made of austenite stainless steel for both longstorage period and transportation after the storage. SCC is induced when three conditions, ma-terial (austenite stainless steel), saline air environment (salt deposit) and residual stress (surfacetensile stress) satisfy the SCC induced conditions. As austenite stainless steel is usually used forthe canister shell material mainly because of economy, it is hard to eliminate the material condi-tion to prevent SCC. On the other hand, SCC seldom occurs during the storage period, because theamount of salt deposit on the canister shell does not usually reach the threshold value. It is worth-while to eliminate the residual stress condition to make sure the mitigation of SCC, even if the saltdeposit seldom exceeds the threshold. SCC resistant canister is developed based on the surfacestress treatments at manufacturing factory and after lid welding to eliminate the residual stresscondition. The specification of the SCC canister is proposed and effect of surface stress treatmentsto SCC are confirmed in the presentation.


Japan

Primary author: Dr RYOJI, ASANO (HITACHI ZOSEN CORPORATION)

Co-author: Mr MASANORI, Goto (Hitachi Zosen Corporation)

Presenter: Mr MASANORI, Goto (Hitachi Zosen Corporation)


International Co … / Report of Contributions Potential Interface Issues in Spent …


Potential Interface Issues in Spent Fuel Management

Efficient spent fuel management (SFM) requires an evaluation of the potential interface issuesamong phases of the nuclear fuel life cycle: including: technical disconnects, policy considerations,and varied positions of the many stakeholders that influence management options and decisions.Because many issues affect multiple stakeholders and may require long lead times to resolve, it isimportant to identify interface issues early and solve them in a timely manner. Opportunities arelost if interfaces are not identified and addressed in the early stages of each of the Back-End ofFuel Cycle (BEFC) phases.The objective of this paper is to suggest a process for systematically identifying and evaluating thepotential interface issues in SFM, and to recommend effective management based on the experi-ence of Member States before losing timely resolution opportunities.Some conclusions that can be drawn from the system integration in BEFC tasks are:— Assuring compatibility of schedules, equipment, and acceptance criteria are key to solve inter-face issues.— Record/data keeping is an important issue for each interface. Without proper records, interfaceissues might not be able to be addressed or conservative and costly alternative approaches mightneed to be developed.— As storage periods are extended and countries plan consolidation into regional or centralizeddry storage facilities, this interface issue will take on increasing importance – particularly if in-spections and/or repackaging are needed to prepare fuels for long-term storage.— Additional pro-active efforts are needed from every participating organization in the BEFC toensure early attention to public acceptance in the siting, safety, operation, duration, oversight, andpath forward. Accurate information must be provided in a user-friendly format.The principles presented in this paper emphasize the importance of systematically identifying andmanaging interface issues within the BEFC. Because of the complexity of the issues and inter-faces, a process is provided to help ensure exact identification of applicable interface issues andconsideration of the associated issues and opportunities.


Japan/ CRIEPI

Primary author: Dr SAEGUSA, Toshiari (Executive Research Scientist, CRIEPI)

Co-authors: Mr CARLSEN, Brett (Idaho National Laboratory); Mr DEMAZY, Guy (SYNATOM); MrAARLE, Jan (NOK); Mr EINZIGER, Robert (US NRC)

Presenter: Dr SAEGUSA, Toshiari (Executive Research Scientist, CRIEPI)


International Co … / Report of Contributions Regulatory experiences from impl …


Regulatory experiences from implementation of SNFdisposal programme from site selection to

construction of disposal facility – the Finnish case

Finland is one of the foremost countries in the world in developing a disposal solution for spentnuclear fuel (SNF). The Construction License Application (CLA) for the Olkiluoto SNF encapsu-lation and disposal facility was submitted to the authorities at the end of 2012 and the facility isexpected to start operation around 2022.

In 1983 the Government made a strategy decision on the objectives and target time schedule for theresearch, development and technical planning of nuclear waste management. Decision includedthe milestones for an operating disposal facility by 2020.

The Government issued in 2000 a Decision in Principle deciding that Olkiluoto will be the site forthe SNF repository and that Posiva was allowed to proceed by constructing there the undergroundrock characterization facility, Onkalo. The disposal concept was decided to be KBS-3 and thedisposal facility was planned to be located at a depth of 400-700 meters.

After Decision in principle STUK has developed regulatory oversight approach for undergroundfacilities that has been used in oversight of Onkalo construction. The safety case development wasfirst reviewed by using bottom-up approach. Based on the experiences STUK developed for CLAreview and assessment more safety oriented approach.

Posiva submitted the construction license application and supporting documentation to the author-ities at the end of 2012. STUK started the review and assessment with an initial review in early2013. STUK has performed thorough review and assessment against safety requirements. Resultsof the review will be presented to the Finnish government around January 2015. The CLA andSTUK’s review cover aspects of safety, security and nuclear safeguards. The scope of applicationcovers both operational safety of facilities and post-closure safety of disposal.

In pre-CLA phase STUK implemented comprehensive process of preparations, which includedresource and competence build-up, preparation of internal review plan and review of Posiva’sdraft CLA documentation. In parallel with CLA review process STUK has implemented inspectionprogram focusing on applicant’s management system and readiness for construction. To supportregulatory decision making, STUK has used a wide range of national and international experts inthe CLA review process.

The paper will highlight key aspects of regulatory oversight and experiences from detailed sitecharacterization, safety case development phase and review and assessment of the constructionlicense application for Olkiluoto SNF encapsulation and disposal facility.


Finland, STUK

Primary author: Mr HEINONEN, Jussi (STUK)

Co-author: Mr HÄMÄLÄINEN, Kai (STUK)

Presenter: Mr HEINONEN, Jussi (STUK)


International Co … / Report of Contributions Fuel Rod Mechanical Behavior und …


Fuel Rod Mechanical Behavior under Dynamic LoadCondition on High Burnup Spent Fuel of BWR and

PWR

It is assumed that the characteristics of high burnup fuel, such as the increase of hydrogen contentsand hydrides radially precipitated in cladding, would affect the fuel integrity at cask drop accidentduring dry storage and transport.For the assessment of high burnup spent fuel integrity at the cask drop accident, the mechanicalbehavior of the fuel rod such as deformation, failure and pellet dispersion under the dynamic loadcondition was examined using BWR fuel rods (56GWd/t, Zry-2/Zr liner cladding) and PWR fuelrods (52-55GWd/t, MDA cladding).In order to acquire the dynamic mechanical properties of BWR and PWR fuel cladding, dynamictensile tests (strain rate: up to 10^2 s^-1) were performed to obtain axial tensile strength andelongation using cladding coupon specimens, and dynamic ring compression tests (compressionspeed: up to 4000 mm/s) were performed to obtain ring compressive strength and failure flatten-ing ratio. In order to evaluate the dynamic behavior of BWR and PWR fuel rod, the axial andlateral dynamic load impact tests were performed to obtain the failure load and mode on axial andlateral compression using fuel rodlet specimens. After the tests, fracture area and surface wereexamined by fractography and metallography, and the weight and particle size distribution of dis-persed pellets were measured to evaluate the failure behavior of the fuel rod. In the dynamic ringcompression tests and the lateral dynamic load impact tests, influence of hydride orientation wasalso evaluated using hydride re-orientation treated claddings.In the axial dynamic load impact tests, shearing breakage caused by initial impact or bucklingbreakage were observed. In the lateral dynamic load tests, different failure mode and strengthwere observed between “with” and “without” pellet. Based on these test results, the threshold offuel rod mechanical failure under dynamic load condition has been evaluated, and the pellet dis-persion data have been prepared for safety evaluation such as criticality and radiation exposure atthe cask drop accident during dry storage and transport.


Japan/Nuclear Regulation Authority

Primary author: Mr HIROSE, Tsutomu (Regulatory Standard and Research Department, Secretariatof Nuclear Regulation Authority (S/NRA/R))

Co-authors: Mr YAMAUCHI, Akihiro (Regulatory Standard and Research Department, Secretariatof Nuclear Regulation Authority (S/NRA/R)); Mr KAMIMURA, Katsuichiro (Regulatory Standard andResearch Department, Secretariat of Nuclear Regulation Authority (S/NRA/R)); Mr OZAWA, Masaaki(Regulatory Standard and Research Department, Secretariat of Nuclear Regulation Authority (S/NRA/R))

Presenter: Mr HIROSE, Tsutomu (Regulatory Standard and Research Department, Secretariat ofNuclear Regulation Authority (S/NRA/R))


International Co … / Report of Contributions Specific aspects of high burnup or …


Specific aspects of high burnup or mixed oxide fuelrods during dry storage

Beside the long term behavior of the fuel itself the cladding behavior is an important issue forinterim and long term storage of spent fuel in casks. In dry storage the fuel rod cladding is im-portant for the retention of fission products. It is still the first barrier even if the storage casksare designed to contain the fission products during the designated storage period. In Germanythe period for dry storage is limited today to 40 years starting with the loading of the cask. Toensure the cladding integrity the cladding hoop stress and the cladding hoop strain as well as themaximum temperature during drying and storage are limited to certain values. High burnup fuelsas well as mixed oxide fuels have some properties that may be especially considered in dry storagewith respect to cladding integrity especially when the storing time has to be extended. An exten-sion of the dry interim storage time beyond 40 years might be necessary due to the lack of a finalrepository.

Both high burnup fuel and mixed oxide fuel has a higher fission gas release and a slower decreaseof the post irradiation decay heat. Therefore, the number of high burnup or mixed oxide fuelassemblies is limited in most cask loading patterns. Further aspects of these kinds of fuel whichshould be addressed with respect to dry storage are the influence of the high burnup structureand the cavity pressure of the pores of the fuel. Whether these effects have an influence on fueldegradation and on the cladding integrity during the dry storage is not examined up to now.

The influence of high burnup or mixed oxide fuel on other cladding degradation mechanism likethe inner and outer oxide layer, hydrogen pick up, irradiation damages in the cladding materialcrystal structure are well understood during fuel operation and pool storage, but the long termconsequences still aren’t verified due to the lack of experience and long term experiments.

One approach to consider these effects in future is to extend existing fuel performance codes withappropriate models and the consideration of all stages of the fuel life from the more short termoperational behavior to the long term storage time. Within such assessments conservative as wellas best estimate approaches should be considered. For long term predictions deterministic andprobabilistic methods should be used in parallel.


Germany

Primary author: Mr SPYKMAN, Gerold Hermann (TÜV NORD EnSys Hannover GmbH & Co.KG)

Presenter: Mr SPYKMAN, Gerold Hermann (TÜV NORD EnSys Hannover GmbH & Co. KG)


International Co … / Report of Contributions Study on the multilateral manage …


Study on the multilateral management of spent fuelaccording to Korea’s power supply plan

Through the 6th power supply plan at 2013, Korean government announced nuclear power plants(NPP)reaches 34 units at 2024. In 2014, it announced through the 2nd energy basic plan that the 43GW ofthe power supply will be from the NPP. Just like this, although the operating NPP unit is increasing,there isn’t a specific plan on the management or disposal of the spent fuel produced by the nuclearpower generation. Recently in November, Public Engagement Commission on Spent Nuclear FuelManagement suggested the establishment of permanent disposal facility around 2055.According to the technological assessment, it is predicted that the spent fuel storage amount atNPP site will be started to be saturated from 2025. Normally, it suggests temporal expansion ofstorage facility in the plant or operation intermediary storage facility as the alternative before thepermanent disposal. However, there are more diverse ideas such as overseas reprocessing or do-mestic reprocessing. This study analyzed the pros and cons and the possibility to be realized ofeach alternatives through multilateral analysis.The interrelation between the spent fuel production amount and management plans among thetime difference and system dynamics methodology for the analysis of the pros and cons and feasi-bility study of each interrelation was used for the management plan assessment. For the interpreta-tion of the system dynamics methodology the analysis tool was made using the Goldsim programto calculate the yearly spent fuel production amount and disposal of each management plans andstorage and disposable amount.The main assumption of the assessment is as following.• Operating nuclear power plant units : 34 units (light water reactor 30 units, heavy water reactor4 units)• Calculation of the spent fuel production amount: The actual and forecast yearly productionamount of each nuclear power plant unit.• The Cooling Time in NPP : 6 ~7 years• Cooling Time before permanent disposal: 40 years• Yearly reprocessing amount: overseas- 360 ton/year, domestic- max. 600ton/year• Yearly permanent disposal amount: 660 ton/year

In case of the overseas reprocessing, as the amount of the spent fuel is more than the reprocessingamount it is not easy to be the alternative. Also in case of the domestic reprocessing, the timethat the technology is completely developed and the amount to be reprocessing can be a problem.Therefore, the temporal expansion of storage facility in the plant or operation intermediary storagefacility is thought to be necessary.


SOUTH KOREA/KYUNGHEE UNIV.

Primary author: Mr LEE, SUHONG (KYUNGHEE UNIVERSITY)

Co-author: Prof. WHANG, JOOHO (KYUNGHEE UNIV.)

Presenter: Mr LEE, SUHONG (KYUNGHEE UNIVERSITY)


International Co … / Report of Contributions Mechanical Properties of Alumini …


Mechanical Properties of Aluminium Alloys forTransport and Storage Cask after Long Term Storage

In recent dry transport and storage dual-purpose cask designs for spent nuclear fuel, various kindsof aluminium alloy are widely used for basket material as a structural member. Basket for dual-purpose cask is designed to maintain prescribed geometrical arrangements of spent fuel assembliesunder accidents during transport and storage conditions, and also designed to have thermal per-formance for removal of decay heat from spent fuel assemblies. Aluminium alloys, which have agood heat-transfer performance and material workability as well as a high specific strength, aresuitable for basket material.

Because cask may be transported to a reprocessing plant after long term storage at an interimstorage plant, it is important to maintain integrities of basket structure under transport accidentsafter long term storage. However, mechanical properties of these alloys are possibly changedduring storage since cask service temperature may affect micro structure of aluminium alloys overthe long time period. It is therefore essential in cask design qualification to identify degradation inmechanical strength of basket material both conditions for normal storage and transport accidentafter long term storage.

MHI has been conducted comprehensive evaluation of mechanical properties considering thermalageing of the aluminium alloys A6N01 containing boron carbide and normal A3004 without boronfor the basket, which resulted in construction of a data base for assessing structural integrity ofthe basket. In this paper, effects of thermal ageing on mechanical properties of these materialsare presented based on the data, and the microstructural strengthening mechanism which playimportant roles in the evolution of mechanical properties are discussed.


Japan

Primary author: Mr ISHIKO, Daiichi (Mitsubishi Heavy Industries, LTD.)

Co-authors: Mr KISHIMOTO, Junichi (Mitsubishi Heavy Industries, LTD.); Mr YAMAMOTO,Ryuichi (Mitsubishi Heavy Industries, LTD.); Mr MAEGUCHI, Takaharu (Mitsubishi Heavy Industries,LTD.); Mr KAWAHARA, Yoshiyuki (Mitsubishi Heavy Industries, LTD.)

Presenter: Mr KISHIMOTO, Junichi (Mitsubishi Heavy Industries, LTD.)


International Co … / Report of Contributions Demonstration Test Program for …


Demonstration Test Program for Long-term DryStorage of PWR Spent Fuel

In Japan, the first interim spent fuel storage facility away-from-reactor (AFR) will start its opera-tion for management of spent fuels until reprocessing. This facility stores BWR /PWR spent fuelassemblies using dry metal dual purpose casks (storage / transport) which will be transported totheir destinations after the interim storage for decades. This facility is not equipped with a hot-cellfor opening the primary lid of the cask because one of the basic concepts the facility is a simpleoperation not to handle a radioactive material directly, that reduces radiation exposure of workersand a risk of contamination troubles.

Although a visual inspection of spent fuel assemblies is usually carried out before spent fuel trans-portation, the visual inspection of spent fuel assembles is not carried out in the interim storagefacility because the interim storage facility does not have a hot-cell as noted above. Therefore, weare preparing for the demonstration test for long-term dry storage designed to confirm the spentfuel integrity during long-term dry storage by use of the test container which is able to reproducestoring the PWR spent fuel in the similar environment to actual casks.

In this presentation, we introduce the approach to store the PWR spent fuel for long-term andthe status of the demonstration test with explanation of the outline of the demonstration test, thespecification of the manufactured test container and the result of temperature evaluation of spentfuel assemblies during dry storage in the test container with a previously-verified assessment toolwhich is constructed to simulate the result of heat-transfer test for the test container.


Japan

Primary authors: Mr IRIE, Norikazu (The Kansai Electric Power Co., Inc.); Mr FUKUDA, Shin-ichi(The Japan Atromic Power Company); Mr KAWANO, Yoshinobu (Kyushu Electric Power Co., Inc.)

Co-authors: Mr KISHIMOTO, Junichi (Mitsubishi Heavy Industries, LTD.); Mr NISHI, Kosaku(Mitsubishi Heavy Industries, LTD.)

Presenter: Mr KISHIMOTO, Junichi (Mitsubishi Heavy Industries, LTD.)


International Co … / Report of Contributions Aging Management Solutions to E …


Aging Management Solutions to Ensures Safety ofExtended Dry Fuel Storage

Storing used fuel at reactor sites for long periods of time wasn’t originally planned. But given thedelays in the deployment of long term solutions such as geological repositories, license renewal be-yond the initial original license of 20 years is now necessary. In the United States, there are 72,000MTU of SNF discharged including 22,000 MTU in dry storage (about 1,900 casks/canisters for 90%in dual purpose) at 63 operating dry storage facilities. In Germany, dry storage casks are storedat consolidated storage sites and on-site as transportation of spent fuel is prohibited nowadays.With further reactor shutdowns until 31.12.2022, it is expected to have about a total of 1,500 or1,600 dry storage casks at 16 storage sites (12 on-site). Regulators are in the process of defining pe-riodic inspection and testing management program to monitor and maintain dry storage systemson site to ensure a high level of safety and security. In this early stage of license renewals, agingmanagement program includes mainly periodic inspections of the used fuel dry storage systemsand components to ensure potential aging effects are identified and effectively managed. As ourindustry learn through R&D studies and demonstration and surveillance programs, aging manage-ment programs will be evolving with advanced inspection technologies and industry operatingexperience as they become available in the future.This presentation will present innovative solutions being developed to monitor age-related degra-dation and prevent equipment failures caused by aging.


FRANCE

Primary author: Mrs SHELTON, Catherine (AREVA TN)

Co-authors: Mr GARCIA, Justo (AREVA TN); Mr NARAYANAN, Prakash (AREVA TN)

Presenter: Mrs SHELTON, Catherine (AREVA TN)


International Co … / Report of Contributions Technologies for integrated safety …


Technologies for integrated safety managementregarding gas generation in spent fuel dry storage

systems or transport systems, wet or dry

The gas build up in the casks and potential presence of hydrogen are important safety issues inour industrial activities. For wet packages, radiolysis of water in the cavity is the main mechanismfor gas and hydrogen generation whereas for dry systems, radiolysis of residual water in case ofincomplete drying is the most significant mechanism for gas production.An evaluation of hydrogen generation in packages is necessary to ensure that a flammable mix-ture will not be formed during transportation and to verify that casks do not accumulate an unsafeconcentration of hydrogen.The purpose of this paper is to give an overview of the current R&D programs to mitigate thehydrogen risk in the transportation or storage casks.The technology for wet transportation of spent fuel is a catalytic recombining system which isqualified through tests at various temperatures. Particular attention is placed on the recombiningefficiency after immersion of the catalyst in borated water, which would occur in a nuclear reactorpool during loading of used fuel.For dry storage and transportation of normal spent fuel (not damaged), water can be removed orthe water amount reduced to a very low level with AREVA TN cask vacuum drying technologyand high efficiency procedure.Concerning the transport of leaking assemblies, there is potentially generation of hydrogen con-nected to the presence of residual water within the fuel rods.

For addressing gas generation issues, R&D results and technologies are now available. Experimen-tal programs have been performed in order to obtain measurements of the gas generation rate soas the empirical data can be used in analytical models to predict H2 concentration.New technologies have been developed to mitigate the hydrogen risk. A high efficient method forthe elimination of gas generated is the introduction in the transport cask of materials able to trapthe radiolytic H2 or to buffer the hydrogen concentration far below the flammability limit: thehydrogen getters. The hydrogen gas is absorbed by these materials and chemically bound in thecrystalline structure. The materials under study belong to the class of the intermetallic compounds.Such intermetallic compounds able to store irreversibly hydrogen are known as Non EvaporableGetters (NEG).


AREVA TN

Primary author: Mr ISSARD, herve (AREVA TN)

Co-authors: Mr SAFFRE, Dimitri (AREVA TN); Mrs GHAZAL, Roula (AREVA TN)

Presenter: Mr ISSARD, herve (AREVA TN)


International Co … / Report of Contributions Radioactive materials and waste m …


Radioactive materials and waste management:taking-up present and future challenges

The reduction of the volume and radiotoxicity of radioactive waste has been one of the key driversfor the implementation of the French nuclear program. Interactions between the nuclear industry,the administration, the competent authorities and other stakeholders including the civil societyhas led to a continuous evolution of the industrial facilities, practices, solutions and products tomove towards waste minimization.Bringing a solution to the management of ultimate radioactive waste is clearly a requirement forthe sustainable development of nuclear energy whose crucial role in the world’s future energy mixhas been confirmed in the past few months by several international organizations.The first question to be answered concerns the definition of radioactive waste. This notion may beseen as self-evident. It is not. For example, used fuel is considered a High Level Waste in Finlandor Sweden, but a recyclable material in countries including China, France, India, Japan, Russia, andthe Netherlands. In France, a definition is provided by law to clarify this notion.AREVA has been designing and implementing solutions for maximizing materials reuse, improv-ing existing waste management routes and thus, mitigating risks, safety being the top priority.Innovation is also at the heart of integrated solutions and routes under development.This paper will review solutions and processes already implemented. It will practically illustratehow the ever stronger level of interactions between the involved stakeholders has led to an inte-grated approach connecting consistently nuclear power plants, fuel cycle facilities and final repos-itories. The long term management of radioactive substances, one of the most important of whichbeing used fuel, lies in defining the best compromise between reduction at source – potentialtreatment (including characterization, sorting, processing) - recycling – storage and disposal. Theefficiency, the completeness and consistency of the proposed solutions with today’s and tomor-row’s global framework will also be analyzed.An important lesson learned from the experience in countries having developed a nuclear pro-gram for decades is that the early implementation of comprehensive solutions for the long termmanagement of used fuel and radioactive waste concomitantly with the reactor fleet deploymentis a prerequisite to sustainability. The road ahead for such an implementation is long and full ofchallenges. This paper will aim at analyzing how countries which are considering the constructionof their first nuclear power plant or the expansion of a recently launched nuclear power programmay draw from this experience.


FRANCE/AREVA

Primary author: Mr ROMARY, Jean-Michel (AREVA)

Co-author: Mr GRYGIEL, Jean-Michel (AREVA)

Presenter: Mr ROMARY, Jean-Michel (AREVA)


International Co … / Report of Contributions A Unified Spent Nuclear Fuel Data …


A Unified Spent Nuclear Fuel Database and AnalysisSystem

A fully integrated waste management system involves managing the waste from the time it is dis-charged from the reactor and designated as spent nuclear fuel (SNF) to the time it is disposed ofin a geologic repository. Performing the different types of analyses required to account for thechanging nuclear and mechanical characteristics of SNF over time, and understanding how thesecharacteristics affect the different storage, transportation, and disposal options, can require manytools and types of data. To support the US Department of Energy Office of Nuclear Energy (DOE-NE) Nuclear Fuels Storage and Transportation Planning Project (NFST) planning activities to laythe groundwork for implementing interim storage, including associated transportation, per theAdministration’s Strategy for the Management and Disposal of Used Nuclear Fuel and High-LevelRadioactive Waste, an integrated data and analysis tool—the Used Nuclear Fuel-Storage, Trans-portation & Disposal Analysis Resource and Data System (UNF-ST&DARDS)—has been devel-oped. UNF-ST&DARDS provides a controlled source of technical data integrated with key analysiscapabilities to characterize the inputs to the overall US waste management system from reactorpower production through ultimate disposition. This system is a new and unprecedented capabil-ity/resource that enables automated assembly-specific and cask-specific evaluations for assessingissues and uncertainties related to the extended storage and transportability of loaded canisters;supporting safety confidence and R&D prioritization; and providing a foundational data and analy-sis capability resource for the future. Various types of data are stored within the a comprehensive,domestic SNF system database, the Unified Database, including: fuel assembly discharge infor-mation; fuel assembly design data; reactor-specific operation data; cask design and loading data;infrastructure and logistics-related data to support systems analyses; and nuclear safety analysischaracterization results for individual assemblies and SNF canister/cask systems. Key elementsof the system design include the data relations defined within the Unified Database and an appli-cation agnostic template engine that allows the large number of inputs required to characterizethe SNF for each respective site to be generated automatically. This paper provides an overviewof: the UNF-ST&DARDS architecture; automated analysis capabilities that include assembly de-pletion and decay, cask criticality and shielding via the SCALE code system, cask thermal analysisvia the COBRA-SFS code; and some of the results visualization and data interrogation capabilitiesavailable through the user interface.


USA/Oak Ridge National Laboratory

Primary author: Mr SCAGLIONE, John (Oak Ridge National Laboratory)

Co-authors: Dr RADULESCU, Georgeta (Oak Ridge National Laboratory); Dr BANERJEE, Kaushik(Oak Ridge National Laboratory); Dr ROBB, Kevin (Oak Ridge National Laboratory); Dr LEFEBVRE,Rob (Oak Ridge National Laboratory)

Presenter: Mr SCAGLIONE, John (Oak Ridge National Laboratory)


International Co … / Report of Contributions Enlargement of the Olkiluoto spen …


Enlargement of the Olkiluoto spent fuel interimstorage

The objective of this presentation is to share regulator’s experiences of regulating and having over-sight for the enlargement of a spent fuel interim storage. An overview of the current situation ofspent fuel management and near future plans in Finland will be given.In Finland, there are four operating reactors, one under the construction and two reactors that arewaiting the construction licenses to be submitted. In Olkiluoto, the two operating units and oneunit under construction, have a shared interim storage for the spent fuel. The storage was designedand constructed in 1980’s. The option for enlarging the storage was foreseen in the original design.Considering spent fuel from these three units and the fact that the final disposal begins after 2020,extra space in the spent fuel storage was estimated to be needed around 2015.The spent fuel is cooled in the interim storages before encapsulating it for the underground finaldisposal. The construction license application of the spent fuel encapsulation plant and the under-ground final repository was submitted at the end of 2012. The operation of these final disposalunits is estimated to begin after 2022.The enlargement of the interim storage was included in Olkiluoto NPP 1&2 operational licenseand it was considered as a major plant modification. To conduct the enlargement, the operatorwas required to submit the documentation similar to application for the construction license of anuclear facility.When conducting changes in an old nuclear facility, the updated safety requirements have to befollowed. The major challenge in designing the enlargement was to modify it to withstand a largeairplane crash. The operator chose to cover the pools with protective slabs and also to build alandfill embankment and concrete structures out side the interim storage. The designing of coverslab structures is an optimisation task between the safety issues that are partly opposite to eachother.The construction phase of the enlargement caused some unexpected events. Synchronization ofthe construction phases with implementation of the system modifications proved out to be morechallenging than originally considered. These experiences emphasize the importance of good, thor-ough and detailed planning of the construction phases.


Finland/Radiation and nuclear safety authority

Primary author: Ms MAARANEN, Päivi (senior inspector, Radiation and Nuclear Safety Authorityin Finland, STUK)

Presenter: Ms MAARANEN, Päivi (senior inspector, Radiation and Nuclear Safety Authority inFinland, STUK)


International Co … / Report of Contributions Increasing efficiency in waste man …


Increasing efficiency in waste management: fromthe conditioning at source to the long term disposal

Used fuel, in France, is reprocessed in order to retrieve its recyclable content. The residues arein the form of High Level Waste (HLW) or Intermediary Level Waste Long Lived (ILW-LL) andinclude on one hand fission products and minor actinides traces, and on the other hand metalstructures of the nuclear fuel assembly (cladding, hulls and end-caps). A distinction is made be-tween the wastes stemming from used fuel processing and related final waste conditioning, fromthe waste generated through the operation of nuclear power plants and fuel cycle facilities, andthe waste arising from the dismantling of nuclear installations. AREVA designs and implementssolutions for materials and waste management with the constant concern of improving the safetymitigating risk management and increasing overall efficiency of the routes.This hereby paper will lay out a review of the key factors considered for a global waste manage-ment strategy from raw waste generation, with a “zero waste” ambition, to the final waste disposal.After an overview of the different types of waste stemming coming from nuclear power genera-tion, this paper will review solutions and processes already implemented for their safe and efficientmanagement.It will stress how AREVA performs important efforts (i) to carry out radiological and chemical char-acterization, (ii) to develop suitable solution for the waste conditioning based on R&D program,(iii) to control the production and finally (iv) to pay a great attention to the long term behaviourof waste container. Concerning the CIGEO deep geological project (lead by ANDRA, the FrenchWaste Management Agency), the primary packages must constitute the first static barrier of pro-tection. The behavior assessment of the primary package is a very key point. For this purpose,AREVA deploys an important effort on characterization and address thanks to R&D programs foreach kind of waste containers (i) the study of the release of gaseous compounds, (ii) the releaseof chemical species, (iii) the interaction between the content of the wastes and the conditioningmatrices (with respect to the corrosion of the package, the long term behavior of the concrete andthe complexation of radionuclides) and (iv) the retention capacity of the matrix itself.


FRANCE/AREVA NC

Primary author: Mrs LAMOUROUX, CHRISTINE (AREVA NC)

Co-author: Mrs COCHIN, FLORENCE (AREVA NC)

Presenter: Mrs LAMOUROUX, CHRISTINE (AREVA NC)


International Co … / Report of Contributions Using of the old type transport cas …


Using of the old type transport cask C-30 for animproved fuel VVER-440

The transport cask C-30 with basket T-12 was developed in former East Germany for VVER-440fuel with enrichment 3.6%. Later was in Slovakia developed the new compact basket KZ-48 forVVER-440 fuel with enrichment 4.4% and maximal burnup 55 MWd/kgU. The present licence is forVVER-440 fuel with enrichment 4.4% and maximal burnup 60 MWd/kgU. The future licence needto be for VVER-440 fuel with average enrichment 4.87% and maximal burnup 70 MWd/kg.

In this paper are criticality, inventory and shielding analyses described .


Slovakia/VUJE

Primary author: Mr CHRAPCIAK, Vladimir (senior engineer)

Co-authors: Mr LIPTAK, Pavol (senior engineer); Mr ZAJAC, Radoslav (senior engineer)

Presenter: Mr CHRAPCIAK, Vladimir (senior engineer)


International Co … / Report of Contributions THE ROLE OF VOLUNTARY NAT …


THE ROLE OF VOLUNTARY NATIONALCONSENSUS STANDARDS IN THE TRANSPORT OF

SPENT NUCLEAR FUEL

In the United States, the American National Standards Institute (ANSI) facilitates the developmentof American National Standards by accrediting the procedures of standards developing organiza-tions, also known as SDOs. Accredited Standards Committee (ASC) N14, “Packaging and Transportof Radioactive and Non-Nuclear Hazardous Materials,” is the ANSI-accredited SDO that publishesstandards for transport of radioactive materials in the United States. Four current N14 standardsdirectly apply to spent fuel transport: N14.5 – Leakage Tests on Packages for Shipment, N14.6 –Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 kg) or More forNuclear Materials, N14.33 – Storage and Transport of Damaged Spent Fuel, and N14.36 – Measure-ment of Radiation Level and Surface Contamination for Packages and Conveyances. Additionally,several N14 standards were previously under development that were applicable to spent fuel trans-port; however, when the DOE Office of Civilian Radioactive Waste Management was shut downin 2010, work on these standards was abandoned, and those writing committees were disbanded.The United States also works closely with the International Organization for Standardization (ISO)Technical Committee 85, Subcommittee 5, Working Group 4 (ISO/TC85/SC5/WG4) – “Transporta-tion of radioactive materials,” in the development of international standards related to spent fueltransport.

As Chairman of ANSI N14 and the United States’ voting member on the ISO/TC85/SC5/WG4, theauthor is in a unique position to offer perspectives on the development of standards related tospent fuel transport. This paper discusses the role of voluntary national consensus standards inthe transport of spent fuel in the United States and internationally and looks to the future todescribe ways in which new standards could facilitate these shipments while ensuring the safetyand security of workers and the public at large.


United States

Primary author: Mr FELDMAN, Matthew (Oak Ridge National Laboratory)

Presenter: Mr FELDMAN, Matthew (Oak Ridge National Laboratory)


International Co … / Report of Contributions Laying the Groundwork for a …


Laying the Groundwork for a Large-scale Used FuelTransportation System

The US Department of Energy Office of Nuclear Energy established the Nuclear Fuels Storageand Transportation Planning Project to lay the groundwork for implementing an interim storagefacility, including associated transportation activities. Efforts include the development of a systemfor the large-scale transport of spent nuclear fuel that will be necessary in an integrated wastemanagement system. Progress is being made on long lead time, destination-independent aspectsof the transportation infrastructure. The large-scale transportation system for spent nuclear fuelis divided into three elements: institutional, operational, and hardware. The institutional elementrefers to the various forms of stakeholder interaction that must occur for this type of transportationsystem to be successful. It includes activities like development of a national transportation plan,work on policy development for Section 180(c) of the Nuclear Waste Policy Act, and identificationof a preliminary suite of national transportation routes that reflect the interests of a broad crosssection of stakeholders while meeting regulatory requirements. The operational element refers tothe activities that must be undertaken to run a large-scale transportation system. This element iscurrently focused on development of a new SNF transportation routing analysis tool, study of theinfrastructure near SNF storage sites that may be de-inventoried first, and development of tools formodeling transportation activities. The hardware element refers to the casks, railcars, and otheritems necessary to operate the system. This element currently focuses on development of railcarscompliant with Association of American Railroads Standard S-2043, as well as studies related tothe use of rail casks and their ancillary equipment. The Nuclear Fuels Storage and TransportationPlanning Project is making significant progress in all three of these areas along the path forwardto a fully operational transportation system.


United States

Primary author: Mr FELDMAN, Matthew (Oak Ridge National Laboratory)

Co-authors: Dr BICKFORD, Erica (U.S. Department of Energy); Mr JONES, Jay (U.S. Departmentof Energy); Dr WAGNER, John (Oak Ridge National Laboratory); Dr CONNOLLY, Kevin (Oak RidgeNational Laboratory); Dr SCHWAB, Patrick (U.S. Department of Energy); Dr MAHERAS, Steve (PacificNorthwest National Laboratory)

Presenter: Mr FELDMAN, Matthew (Oak Ridge National Laboratory)


International Co … / Report of Contributions Used fuel treatment : for which fuel ?


Used fuel treatment : for which fuel ?

For more than 40 years, used fuel reprocessing is an industrial reality, with tens of thousandsof tons treated in several plants around the world, allowing saving natural resources, reducingwaste volume and toxicity and easing waste storage and disposal. Reprocessing has already beenapplied successfully to a very wide range of used fuels, coming from different types of reactorswith a variety of designs and characteristics. While details of operation vary to comply with thisdiversity, technological solutions remains quite similar, allowing to progressively extend the scopeof reprocessing while retaining perfect track records in terms of safety, security and environmentalimpact.This paper will give an overview of the French experience and results from the industrial treatmentof different fuels and the associated characteristics.


France - AREVA

Primary author: Mr BRUEZIERE, jerome (AREVA)

Presenter: Mr BRUEZIERE, jerome (AREVA)


International Co … / Report of Contributions Sustainable cycle solutions


Sustainable cycle solutions

Almost four years after the Fukushima events, the nuclear industry has shown its capacity to re-spond to first additional safety requirements and continues working closely with the regulators andinvolved stakeholders for further improvements. Continued growth is still anticipated and severalinternational organizations have confirmed 2035 scenarios close to or above 600 GWe of installednuclear capacity. Securing and optimizing the fuel cycle is crucial to ensure the sustainability ofsuch a nuclear program in the medium term. Rethinking fuel cycle schemes is also decisive toprepare long term transitions to next generations of reactors. R&D and Innovation shall remain acornerstone of this sustainable development.

With the foreseen nuclear development, the used fuel backlog will continue to grow for severaldecades. Bolstered by a 40 years experience, recycling has been contributing to further increasesafety while offering main operational, environmental, global acceptance benefits, among others.However even though recycling capabilities are on track to expand in some countries there will be agrowing need, to move forward efficiently, to smartly mix proven and evolving solutions (recycling,on site dry storage, pools, centralized storage, advanced technologies). These shall be combined inan optimized manner taking into account key criteria related to non proliferation, minimization ofenvironmental impact, economics, fleet performance, responsibility towards future generations…The recommendations from Safety Authorities to deepen the question of very long term dry storagehave also reactivated the debate about the available options for the short and long terms, includingrecycling, which is therefore likely to play an increasing role, especially if flexibility is requiredto ensure a cost-effective fast reactor ramp-up. Such an approach is aimed at meeting needs thatmay differ, even diverge depending on the concerned stakeholders.

The paper will in particular illustrate how Areva’s “Sustainable Cycle Solutions” may optimizefuel cycle schemes for the short and longer terms and fulfill the stakeholders’ needs. It will alsopinpoint the related limits and conditions that have to be met to fulfill the above mentioned criteria.


France - AREVA

Primary author: Mrs DREVON, Caroline (AREVA)

Co-author: Mr BRUEZIERE, jerome (AREVA)

Presenter: Mrs DREVON, Caroline (AREVA)


International Co … / Report of Contributions Innovative Polyvalent Fuel Treatm …


Innovative Polyvalent Fuel Treatment Facility

Although many used nuclear fuel types have already been recycled, recycling plants are generallyoptimized for LWR UOx fuel. Benefits of used fuel recycling are consequently restricted to thosefuels, with only limited capacity for the others like LWR MOX, FR MOX or Research reactor fuel.In order to increase the capacity of the La Hague plant to process other fuels, an innovative andpolyvalent shearing and dissolving cell is planned to be put in operation at La Hague. This installa-tion, called TCP (French abbreviation for polyvalent fuel treatment), will be set up to accept a widerange of fuel while benefiting from the installed capacity. The TCP shearing tool and dissolvingequipment will benefit from AREVA’s industrial experience, while taking part in the next stepstowards a fast reactor fuel cycle development using innovative treatment solutions. Feasibilitystudies and R&D trials on dissolution and shearing are currently ongoing. This new installationwill allow AREVA to propose new services to their Customers, in particular in term of MOX fuel,Research Test Reactors fuel and Fast Reactor fuel treatment.


France-AREVA

Primary author: Mr BRUEZIERE, jerome (AREVA)

Presenter: Mr BRUEZIERE, jerome (AREVA)


International Co … / Report of Contributions Status of the Swedish Programme …


Status of the Swedish Programme for DeepGeological Disposal of Spent Nuclear Fuel

The licensing processes according to the nuclear activities act and the environmental code is nowgoing on. The Swedish Radiation Safety Authority has so far asked SKB for additional informationregarding e.g. copper corrosion, canister design, groundwater flow modelling and biosphere mod-elling. The Land and Environmental Court has received input from many stakeholders includingthe concerned municipalities, environmental groups and many others regarding their requests foradditional documentation or explanations by SKB. These requests cover issues concerning site se-lection, alternative disposal methods, environmental consequences due to groundwater dischargeand increased traffic, just to name a few. The Court will now make its own judgement and preparethe formal demand on SKB for additional information.It is presently estimated that the final statements to the government by SSM and the Environmen-tal court will come in 2015. Political decisions, provided that the safety authority and the courthave given a green light, by the municipalities of Östhammar and Oskarshamn could then possiblybe taken in 2016 followed by a final decision of the Swedish government, whereby construction ofthe repository could start around 2020. The local public opinion is still very positive with about 80% of the population being positive to the plans of SKB. A broad dialogue with all stakeholders willcontinue to be of high priority for SKB in order to arrive at a situation in a few years from nowwhere all necessary decisions can be taken based on a credible and transparent understanding ofthe critical issues.

While the licensing process is still under way, SKB continues to prepare for implementing the KBS3-system. Among other things, this entails building up an organization and industrial productionsystem for all parts of the final disposal process.


Swedish Nuclear Fuel and Waste Management Co

Primary authors: Mr STRÖM, Anders (SKB); Ms AHSBERG, Helene (SKB); Ms LAÂROUCHIENGSTROEM, Saida (SKB)

Presenter: Ms LAÂROUCHI ENGSTROEM, Saida (SKB)


International Co … / Report of Contributions Managing Aging Effects on Dry C …


Managing Aging Effects on Dry Cask StorageSystems for Extended Long-Term Storage and

Transportation of Used Fuel

In the United States, there is currently no designated disposal site for used nuclear fuel, whichraises the prospect of extended long-term storage (i.e., >60 years) and deferred transportation ofused fuel at operating and decommissioned nuclear power plant sites. Under U.S. federal regula-tions contained in Title 10 of the Code of Federal Regulations 72.42, the initial license term foran independent spent fuel storage installation (ISFSI) must not exceed 40 years from the date ofissuance. Licenses may be renewed by the U.S. Nuclear Regulatory Commission (NRC) at the ex-piration of the license term upon application by the licensee, for a period not to exceed 40 years.Applications for ISFSI license renewals must include (1) time-limited aging analyses that demon-strate that structures, systems, and components (SSCs) important to safety will continue to performtheir intended function for the requested period of extended operation and (2) a description of theaging management program for management of issues associated with aging that could adverselyaffect SSCs important to safety. This paper highlights issues related to managing aging effects ondry cask storage systems and ISFSIs for extended long-term storage and subsequent transportationof used nuclear fuel. In particular, it focuses on aging management issues related to the confine-ment boundary of bolted- and welded-closure storage casks and canisters. These highlights wereextracted largely from the 2014 report prepared by Argonne for the U.S. Department of Energy’sUsed Fuel Disposition Campaign for R&D on extended storage and transportation of used fuel. Thepaper will also include additional information on the update of guidance documents by the U.S.NRC on “Standard Review Plan for Renewal of Used Fuel Dry Cask Storage System License andCertificate of Compliance,” NUREG-1927, and by the U.S. Nuclear Energy Institute on “IndustryGuidance for Operations-Based Aging Management,” NEI 14-03, which was submitted for NRCendorsement in September 2014 and includes the DOE/Argonne National Laboratory aging man-agement report as a key reference. Finally, the paper will briefly discuss aging management needsfor a Pilot Interim Storage Facility and beyond, as the used fuel may need to be stored and trans-ported multiple times before final disposal at a mined repository or geological disposal facility.


United States

Primary author: Dr LIU, Yung (Argonne National Laboratry)

Co-authors: Dr DIERCKS, Dwight (Argonne National Laboratory); Dr NUTT, Mark (ArgonneNational Laboratory); Dr CHOPRA, Omesh (Argonne National Laboratory)

Presenter: Dr LIU, Yung (Argonne National Laboratry)


International Co … / Report of Contributions Lessons learned from a periodic sa …


Lessons learned from a periodic safety review,applied to the design of new nuclear fuel cycle

facilities

During the last safety reassessment of the spent fuel reprocessing plant operated by AREVA NCin La Hague (France), IRSN reviewed the experience feedback on incidents as well as the compli-ance of several equipment of the primary containment to their safety requirements. This reviewis designed to assess the level of safety of a facility in relation to the rules or the laws which areapplicable at the time of this review, updating in particular risk assessment.Further investigations are now necessary to control some risks poorly anticipated at the designstage. They give information guidance for the design of new nuclear facilities.For example, leakage detection systems associated to the primary containment in units dealingwith uranium or plutonium are designed to detect an arrival of a large amount of liquid in a sec-ondary containment. The experience feedback showed that the related systems failed to detectleakages of radioactive materials because the leaks occurred at process steps where uranium andplutonium were in solid state, or because of a very low flow of a liquid leak on warm surfaces ofthe equipment. Correctives measures consist in conducting periodic observations of the secondarycontainment below the equipment. Such observations may be difficult because of the absence ofport-hole, the difficulties to introduce cameras, the low level of light in the cells, etc.Furthermore specific tools for non-destructive testing methods are developed, in order to monitorthe residual wall thickness of equipment, in order to confirm the corrosion kinetics. But their de-ployment in highly contaminated cells is limited by the difficulties to introduce these devices intothe cells and to place them at a suitable position on the equipment.The anticipation of these controls in the design stage would have made it possible to facilitatethem and to better control the aging of facilities and the absence of leak. These lessons are alreadyapplied by IRSN for the safety assessment of new nuclear fuel cycle facilities.


FRANCE/IRSN

Primary author: Dr RACIMOR, DAVID (IRSN)

Co-authors: Ms LIZOT, Marie-Thérèse (IRSN); Dr CHARRIN, Nicolas (IRSN)

Presenter: Dr RACIMOR, DAVID (IRSN)


International Co … / Report of Contributions Reprocessing of BREST reactor mi …


Reprocessing of BREST reactor mixed nitride SNF

BREST-OD-300 is pilot demonstration lead-cooled fast reactor fuelled with mixed nitride uranium-plutonium fuel (MNIT). Nuclear fuel cycle with BREST reactors concept implies that the fuel hasan burn-up about 10 % h.a.. “Equilibrium” isotopic composition of fissile materials (FM) fuel isused in the nuclear fuel cycle and the duration of the external fuel cycle does not exceed one year.It means that within one year spent nuclear fuel (SNF) should be reprocessed and fresh MNIT-fuelshould be fabricated from the products of the reprocessing. At the present time a carbothermicsynthesis technology is chosen as a basic process for the fabrication of mixed uranium-plutoniumpowders and pellet technology – for fuel fabrication. Thus manufacture on reprocessing of MNITSNF from BREST reactor should provide reprocessing of fuel with cooling time not exceeding oneyear, FM content of 10-15%, burnup about 10% h.a. and obtain a mixture of actinides with coeffi-cients of purification about 1000000.It is clear that requirements for reprocessing of MNIT SNF are imposed, such as safety level, eco-logical issues, and economic competitiveness. In our opinion PH-process (combined (pyro+hydro)technology) of reprocessing SNF can meet all these requirements. PH-process consists of a com-bination of pyrochemical, including pyroelectrochemical, head operations and hydrometallurgicaloperations for refining of end product (U-Pu-Np-Am) and waste treatment. Development of PH-process has been run since 2011 by RIAR, Bochvar Institute, and Khlopin Radium Institute.By now following basic technological operations have been tested in laboratory conditions on realproducts:- steel cladding removal by dissolution in molten zinc;- pyroelectrochemical recovery of uranium, plutonium, and neptunium;- extraction refining of uranium, plutonium, and neptunium;- separation of rare earth elements and transplutonium elements;- separation of americium and curium;- preparation of mixed oxides of actinides by microwave denitration;- decomposition of ammonium nitrate and complexions.In 2014 manufacturing of experimental setups to test engineering solutions for equipment for py-roelectrochemical operations was started and installation for inspection innovative systems for hy-drometallurgical operations was completed. In 2015 design of reprocessing MNIT SNF and wastetreatment facility was started. Reprocessing unit will be located at the same site with the pilotdemonstration power complex with BREST-OD-300 reactor on the territory of SCC (Seversk).


Russia / Bochvar Institute

Primary author: SHADRIN, Andrei (Bochvar Institute)

Co-authors: Dr DVOEGLAZOV, Konstantin (Bochvar Institute); Dr IVANOV, Valentin (BochvarInstitute)

Presenter: SHADRIN, Andrei (Bochvar Institute)


International Co … / Report of Contributions Spent Fuel Storage Integration in t …


Spent Fuel Storage Integration in the United States –Planning for Storage and Transportation

Within the United States, the Nuclear Regulatory Commission (NRC) has the responsibility forlicensing and inspection of spent fuel storage operations. This includes storage in the spent fuelpool, typically licensed under Title 10 Code of Federal Regulations (CFR) Part 50 regulations, andwithin spent fuel cask systems licensed by 10 CFR Part 72 regulations. The spent fuel cask sys-tems are also referred to as dry cask storage systems. The spent fuel is adequately protected fromcriticality concerns and other events whether stored in the spent fuel pool or in a dry cask stor-age system. The licensee decides when to move the spent fuel to a dry cask storage system basedon operational and economic needs. Over the past twenty-nine years, the United States dry caskstorage systems have transitioned from single use storage-only cask systems to multi-use cask sys-tems, with the ability to be transported to a centralized storage facility or ultimately to the finaldisposal site. Occasionally, these multi-use cask systems experience defects during fabrication ordeficiencies during spent fuel loading operations that render the canister unable to meet storagerequirements under 10 CFR Part 72 regulations and transportation requirements under 10 CFR Part71 regulations. These defects and deficiencies may require exemptions by the NRC and trackingof the component by the licensee. This paper will briefly discuss the transition within the UnitedStates from single-use cask systems to multi-use cask systems, several strategies considered for de-ciding when to move the spent fuel from wet storage into dry cask storage, processes to transitionthe canister loaded with spent fuel from a storage system to a transportable system, and discussdeficiencies that can render a multi-use cask system into a single-use cask system.


United States/United States Nuclear Regulatory Commission

Primary author: Mr KELLAR, Ray (U.S. Nuclear Regulatory Commission)

Presenter: Mr KELLAR, Ray (U.S. Nuclear Regulatory Commission)


International Co … / Report of Contributions Interim storage of spent fuel in Ge …


Interim storage of spent fuel in Germany – history,state and prospects

Since the concept of dry interim storage of spent nuclear fuel elements in transport casks had beendeveloped by the former DWK in the late 1970s and had been first evaluated by the RSK (Reaktor-Sicherheitskommission) in 1979, development has been going on. Dry interim storage passed a lotof modifications in different fields, but the fundamentals of this first concept – on which the dualpurpose cask used for transport and storage is based on – are still up to date. They are still theessential also of the latest version of the recommendations for dry interim storage of spent fueland high active waste by the ESK (Entsorgungskommission, former RSK).Nonetheless, significant changes in the recent years affect the authorities, experts and operatorsby generation alternation of the faces working in the field of interim storage. Thus, an essentialaspect of age management is transfer of knowledge and experience, which was step by step gainedover the years, to the next generation.This presentation gives an overview over the licensing activities in the field of dry interim storageof spent fuel in Germany within the last 35 years up to now and outlining future perspectives. Thefirst licenses for dry interim storage of spent fuel in Germany were granted in 1983 and in 1987for the central storage facilities in Gorleben and in Ahaus.In October 1989, the responsibility for licensing of interim storage of spent fuel according to ar-ticle six of the atomic Energy Act went over to the newly founded Federal Office for RadiationProtection (Bundesamt für Strahlenschutz - BfS).Presently, there are 16 interim storage facilities in Germany in operation, housing more than 1000loaded transport and storage casks. All existing storage licenses cover a 40 years storage period.Within the last years, ongoing development has been demanding a respectable amount of amend-ments of the existing storage licenses. These are due to developments in the state of the art and inregulatory framework, but also cover more specialized themes.The licensing activities in the next years will be dominated by special questions resulting from thenecessity to empty the storage pools inside the NPPs. Also, storage of different kinds of wastefrom reprocessing of spent fuel in France and UK will be a substantial topic.


Germany - Bundesamt fuer Strahlenschutz

Primary author: Ms PALMES, Julia (Bundesamt fuer Strahlenschutz)

Co-author: Dr GASTL, Christoph (Bundesamt fuer Strahlenschutz -Germany)

Presenter: Ms PALMES, Julia (Bundesamt fuer Strahlenschutz)


International Co … / Report of Contributions Influence of decaying stress and te …


Influence of decaying stress and temperature onradial hydride precipitation in Zircaloy-4 claddings

Considering the back end cycle, the transportation of spent fuel from reactors pools to storagepools at the reprocessing plant of La Hague in France might influence the zirconium based claddingstrength, depending on the maximum temperature reached in the transportation cask. Fuel claddingcorrosion during irradiation in reactor induces hydrogen charging with contents depending on fuelburnup. Part of this hydrogen is dissolved in the cladding during the transportation due to thehigher temperature conditions. After transportation and during the fuel unloading in the storagepool, hydrides precipitate radially and this phenomenon strongly decreases the cladding ductility.IRSN has been involved since several years in the experimental characterization of the main pro-cess controlling the radial hydride precipitation. The thorough knowledge of this phenomenon isneeded to assess safety analysis of wet or dry storages conducted by the licensee and to evaluatemore specifically the consequences of drop loads scenarios in operation involving fuel assemblies.A preliminary step in this study consisted in characterizing the influence of a constant appliedstress to determine the stress thresholds controlling the orientation of precipitated zirconium hy-drides. A new test technique was developed to study the influence of a broad range of temporallyconstant applied stresses using a single test. Pre-hydrided samples with hydrogen contents com-prised between 10 to 600 wppm were studied. Three testing temperatures were studied: 350, 400and 450℃ and the results are used to adjust a model. The above mentioned test technique is notrelevant to address the influence of time-varying stress level.A modeling of decaying stress influence on radial hydride precipitation is derived from the resultsobtained for constant stress. Some new tests were performed with decreasing stress during thecladding cool down to provide validation data. The sample geometry of the tested pre-hydridedcladding consists in cladding rings with machined gage sections to test the influence of hoop stress.The radial hydride precipitation is quantified using post-test metallography combined with imageanalysis. The experimental results are compared to calculation results combining hydride diffusionand radial hydride precipitation modeling.


France, IRSN

Primary author: Dr DESQUINES, jean (IRSN)

Co-authors: DROUAN, Doris (IRSN); PHILIPPE, Marc (IRSN); Dr MARCH, Philippe (IRSN)

Presenter: Dr DESQUINES, jean (IRSN)


International Co … / Report of Contributions Experience of Cask Technology for …


Experience of Cask Technology for SNF Management

A new technology of spent nuclear fuel (SNF) management at the back-end of the fuel cycle hasbeen developed over the last twenty years. This technology is based on the concept of the shieldedcask ensuring containment of its contents (SNF) and compliance with all other requirements forSNF storage and transport. Radiation protection and activity containment are ensured by physicalbarriers, viz. an all-metal or composite body, body linings, internal baskets for irradiated/‘spent’fuel assemblies (SFAs) and lids with sealing systems. SFA residual heat is released to the environ-ment by natural irradiation and natural convection of air around the cask.This report considers the key issues associated with the creation of a family of dual purpose casksfor SNF from naval and nuclear ice-breaker fleet activities, RBMK-1000 and BN-350 reactors as wellas their structural peculiarities. The new development on cask for transportation of SNF VVER-1000 presents also.


Russian Federation

Primary author: Mrs MAKARCHUK, Tatiana (Rosatom/Federal Center for Nuclear and RadiationSafety)

Presenter: Mrs MAKARCHUK, Tatiana (Rosatom/Federal Center for Nuclear and Radiation Safety)


International Co … / Report of Contributions Fabrication Thermal Test. Method …


Fabrication Thermal Test. Methodology for a SafeCask Thermal Performance

One of the safety-related functions in a DPC is the thermal performance. Regulations only addressa thermal test to assure the cask thermal performance under accident conditions (fire test). As aDPC is also to be used for storage, the thermal passive behavior of the cask under normal conditionof storage shall be demonstrated.Although this thermal test is not included in storage regulations, some of the regulators requiredto perform this test as a confirmatory test for the normal condition of storage. In most of the cases,the result of this test will provide with an acceptable resolution or not of the cask design licensingapplication.It is a standard procedure in Spain to perform a fabrication thermal test to the first cask manu-factured. Two aims are intended to be confirmed: a) validation of the analytical model and b) toassure the correct thermal performance of the cask.Thermal test FEM simulation uses almost the same model used for the licensing evaluation withminor changes. The FE model considers the same heat load per fuel assembly, and therefore thesame total cask heat load as used in the licensing calculations. Modifications are on account of thedifficulties to represent the normal condition of storage in the shop.Thermal test uses the first cask manufactured. The cask is in final condition and the only part ofthe cask different from the final storage cask configuration is the cask lid system. A special lid isused for the test to be able to take the instrumentation wires out of the cask cavity. The thermaltest lid should represent the same thermal behavior of the final storage cask lid configuration. Thefuel assembly heat load is simulated by using electrical heaters. FA thermal axial profile is quitedifficult to simulate using electrical heaters as these provide a uniform heat along the length. Forthis reason, both ends of the simulated fuel assembly will provide higher temperatures than thoseobtained in the licensing evaluation.Temperature measurements shall be made once the cask reaches the thermal equilibrium. A suc-cessful thermal test is considered when the test temperatures are lower than those obtained inthe FE model. Additionally, test temperatures shall be lower than those obtained in the licensingcalculations, except for those thermocouples inside the cavity close to the ends of the simulatedfuel assemblies.


EQUIPOS NUCLEARES SA (ENSA)

Primary author: Mr GARRIDO QUEVEDO, DAVID (EQUIPOS NUCLEARES SA (ENSA))

Presenter: Mr GARRIDO QUEVEDO, DAVID (EQUIPOS NUCLEARES SA (ENSA))


International Co … / Report of Contributions SNF MANAGEMENT SYSTEM IN …


SNF MANAGEMENT SYSTEM IN THE RUSSIANFEDERATION

SNF Management Concept of the State Atomic Energy Corporation ROSATOM is in place in theRussian Federation. The Concept specifies goals and strategic vector of SNF management.The Russian Federation policy in the field of SNF management is based on the principle of SNFprocession in order to ensure ecologically acceptable handling of fission products and return ofregenerated nuclear materials in the nuclear fuel cycle. Establishment of reliable system of long-term monitored SNF storage, development of SNF processing technologies, balanced commitmentof SNF regeneration products in the nuclear fuel cycle, final disposal of radioactive waste fromprocessing are strategic approaches in SNF management.The report covers principal provisions of the SNF Management Concept and practical measuresfor its implementation


Russian Federation/State Corporaton ROSATOM

Primary author: Mrs KHAPERSKAYA, Anzhelika (State Corporation ROSATOM)

Co-authors: Mr KUDRYAVTSEV, Evgeniy (Rostekhnadzor); Mr GUSAKOV-STANIUKOVICH„ Igor(JSC FCNRS); Mr IVANOV, Konstantin (State Corporation ROSATOM); Mr KRYUKOV, Oleg (StateCorporation ROSATOM); Mrs MAKARCHUK, Tatyana (JSC “Federal Center for Nuclear and RadiationSafety”,)

Presenter: Mrs KHAPERSKAYA, Anzhelika (State Corporation ROSATOM)


International Co … / Report of Contributions MULTIPLE RECYCLE OF REMIX F …


MULTIPLE RECYCLE OF REMIX FUEL BASED ONREPROCESSED URANIUM AND PLUTONIUM

MIXTURE IN THERMAL REACTORS

The report covers REMIX fuel consumption in VVER-1000. REMIX fuel is fabricated from insepa-rated mixture of uranium and plutonium obtained during NPP spent fuel reprocessing with furthermakeup by enriched natural uranium. It makes possible to recycle several times the total amountof uranium and plutonium obtained from spent fuel with 100% loading of the WWER-1000 core.The stored SNF could be also involved in REMIX fuel cycle by enrichment of regenerated uranium.


Russia

Primary author: Mr FEDOROV, Yuriy (Khlopin Radium Institute, St-Petersburg, Russia)

Co-authors: Mrs KHAPERSKAYA, Anzhelika (State Corporation ROSATOM, Russian Federation); MrKRYUKOV, Oleg (State Corporation ROSATOM)

Presenter: Mr FEDOROV, Yuriy (Khlopin Radium Institute, St-Petersburg, Russia)


International Co … / Report of Contributions Low cost gamma and neutron radi …


Low cost gamma and neutron radiation sensors forreal-time cask monitoring

A low-cost array of modular sensors for online monitoring of radioactive waste was developedat INFN-LNS. We implemented a new kind of gamma counter, based on Silicon PhotoMultipliersand scintillating fibers, that behaves like a cheap scintillating Geiger-Muller counter. Front-endelectronics and an FPGA-based counting system were developed to handle the field data, also im-plementing data transmission, a graphical user interface and a data storage system. Tests in a realradwaste storage site have shown quite encouraging results.We also developed a low-cost technique for thermal neutron detection not making use of 3He, alsosuitable for online real time monitoring of spent fuel casks. As a neutron converter we used 6LiF,being the neutron capture cross section of 6Li very well known and with only an alpha and a tri-ton in the exit channel. We can deposit wide and thin layers of converter onto several differentsubstrates, to be placed on top of solid state detectors or scintillators capable of efficiently detect-ing the decay products. Tests with neutron sources and with neutron beams have proved the fullfeasibility of these sensors.A combination of several units of the aforementioned gamma and neutron detectors can be ex-ploited for spent fuel cask monitoring in place and/or during transportation, in order to contributein the assessment of possible cask ageing with radiation loss, and to prevent possible cask tamper-ing by ensuring the continuity of knowledge.Battery-operated versions of the sensors, with wireless data transmission, are currently under de-velopment.


Istituto Nazionale di Fisica Nucleare - Laboratori Nazionali del Sud

Primary author: Dr FINOCCHIARO, Paolo (Istituto Nazionale di Fisica Nucleare - LaboratoriNazionali del Sud)

Co-authors: Dr PAPPALARDO, Alfio (Istituto Nazionale di Fisica Nucleare - Laboratori Nazionalidel Sud); Dr SCIRÈ, Carlotta (Istituto Nazionale di Fisica Nucleare - Laboratori Nazionali del Sud); DrGRECO, Carmela (Istituto Nazionale di Fisica Nucleare - Laboratori Nazionali del Sud); Dr VECCHIO,Gianfranco (Istituto Nazionale di Fisica Nucleare - Laboratori Nazionali del Sud); Dr COSENTINO, Luigi(Istituto Nazionale di Fisica Nucleare - Laboratori Nazionali del Sud); Dr SCIRÈ SCAPPUZZO, SergioSimone (INFN-LNS); Dr GRILLO, Stefania (Istituto Nazionale di Fisica Nucleare - Laboratori Nazionalidel Sud)

Presenter: Dr FINOCCHIARO, Paolo (Istituto Nazionale di Fisica Nucleare - Laboratori Nazionalidel Sud)


International Co … / Report of Contributions The French Nuclear Fuel cycle: cu …


The French Nuclear Fuel cycle: current status andpossible future options

France has deployed from the late 1970’s a fleet of large water reactors (63 GWe within about 20years, due to French government decisions in the early seventies); and, in the same time, fuel cycleindustrial facilities have been launched to deserve a closed fuel cycle policy. The main drivers forreprocessing and recycling strategy were to take advantage of valuable content of used fuels, whileoptimizing final waste management.

This strategy provides uranium and SWU savings (about 10% of French electricity comes fromburning MOX fuels), and final waste without plutonium, which simplifies long-term management.Plutonium is a sensitive element to several respects (fissile, radiotoxic, heat emitter) and, if notrecycled and burnt, management policies would have to take into account such heavy drawbacks.Recovered plutonium is systematically recycled (no increase of separated Pu stockpile, reprocess-ing is strictly adjusted to refueling capacities); and used MOX fuels are currently safely stored inpools, as a resource for future nuclear systems.Reprocessing and recycling has been operated at commercial scale in France for more than twodecades: more than thirty thousand tons of used fuels have been reprocessed, more than two thou-sand tons MOX fuels manufactured, and fuel cycle technologies have testified industrial maturity.This can be seen as the first step toward fully sustainable nuclear systems: used MOX fuels arenot currently reprocessed (the recovered Plutonium from reprocessing would not be suitable forefficient recycle into current LWRs), so recycle is only once. The deployment in the French fleetof generation IV fast neutron reactors could allow, in the future, a complete and recurrent recycleof both uranium and plutonium, drastically extending natural uranium use (about two orders ofmagnitude), and possibly eradicating any minor actinide content in the final waste.Such options are currently investigated in the frame of a specific Act, voted by the French Par-liament “for sustainable management of radioactive materials and waste”. Different scenarioshave been drafted and will be assessed (taking into account the diverse criteria and appropriateattributes), as a joint work embedding research and industrial bodies. And a dedicated researchprogram, the ASTRID program, has been launched by CEA: it aims at designing of a generation IVsodium cooled demonstrator (both reactor and related fuel cycle facilities), which could be oper-ated from around 2025, as the foundation stone of this new step towards increased sustainability.


FRANCE / CEA

Primary author: Prof. BOULLIS, Bernard (CEA, Nuclear Energy Division, Director for back--end)

Presenter: Prof. BOULLIS, Bernard (CEA, Nuclear Energy Division, Director for back-end)


International Co … / Report of Contributions Linking Nuclear Security to Spent …


Linking Nuclear Security to Spent Fuel Management

National authorities balance technical, economic, legal and political considerations in assessingthe best options for storage, handling, processing and disposing of spent nuclear fuel (SNF) andwaste. Nuclear security and nonproliferation objectives should be integral elements of those con-siderations. In evaluating fuel cycle and R&D programs, the Blue Ribbon Commission assessedcost, safety, resource utilization and sustainability, and the promotion of nuclear nonprolifera-tion and counter-terrorism goals. The Subcommittee stated in its report that “nonproliferationand counter-terrorism are particularly important considerations that have not always received theattention they deserve in evaluations of fuel cycles and nuclear energy alternatives.”

Proliferation risks are largely more familiar than nuclear security risks when it comes to fuelcycle management because they have been discussed for decades, while experts have only justbegun to consider the nuclear security ramifications of fuel management. The nuclear securitysummit process has placed more emphasis on securing HEU rather than Pu because HEU is amore attractive target of theft by terrorists. The fact that most elements of the back end presentfairly formidable obstacles to theft, with the exception of separated plutonium and facilities thatcan be sabotaged, also contributes to the lack of attention to nuclear security on the back end.

Drawing on work funded by the MacArthur Foundation and the Carnegie Corporation of NewYork, the principal investigator will present an analysis of how fuel cycle choices might be influ-enced by greater attention to nuclear security. Examples include when to move SNF from wetto dry storage, consolidation choices, and coordination of storage and repository implementation.The paper will also offer recommendations on principles to help guide such decision-making.


USA

Primary author: Ms SQUASSONI, Sharon (CSIS)

Presenter: Ms SQUASSONI, Sharon (CSIS)


International Co … / Report of Contributions Management of Spent fuel from P …


Management of Spent fuel from PHWRs in India-AnIntegrated approach

India has adopted closed fuel cycle to complete the three stage nuclear power programme en-visaged by great founder and nuclear scientist Dr Homi J. Bhabha which would provide energysecurity for the country.The spent fuel from Pressurised Heavy Water Reactors , the mainstayof nuclear power production in the country are processed in reprocessing plants at Tarapur andKalpakkam. In order to implement stage –II of Indian NPP i.e Fast Breeder Reactors( FBRs) areplanned and fuel for these FBRs will be generated at large scale Integrated Recycle plants beingset up for the first time.India has mastered the spent fuel transportation, storage reprocessing and waste managementtechnology indigenously for PHWR fuels and has been operating plants since last three decades.The large throughput plants are based on solid-in and solid-out concept using the existing technol-ogy incorporating improvements based on the feedback from the operating facilities taking intoconsideration of economics of construction and operation.This paper describes the details of Integration approach, spent fuel transportation, storage, pro-cessing and waste management being planned for large scale recycle plant.

Keywords: Nucleat Power Programme( NPP),Spent Fuel( SF), PHWRs( Pressurised Heavy WaterReactors), high level waste (HLW), low & intermediate level (L&IL) long lived solid waste, vitrifiedwaste product (VWP), vitrified waste storage facility (VWSF)


India/ Department of Atomic Energy, Bhabha Atomic Research Centre

Primary author: Mr AGARWAL, Kailash (BARC,DAE,INDIA)

Co-author: Mr BASU, Sekhar (BARC,DAE,INDIA)

Presenter: Mr AGARWAL, Kailash (BARC,DAE,INDIA)


International Co … / Report of Contributions Closed nuclear fuel cycle with fast …


Closed nuclear fuel cycle with fast reactors anddense fuel

Success of closing of a nuclear fuel cycle (CNFC) is connected with the solution of the problem ofthe multiple recycling of actinides. One of the most promising options of closing NFC is a systemwith thermal reactors (TR) used uranium or mixed fuel and with fast reactors (FR) with high densemixed fuel. Currently, in the RF various scenarios of closing NFC are considered, one of which isthe reuse of plutonium, neptunium and possibly americium recovered from TR spent nuclear fuel(SNF) for fabrication of mixed nitride uranium-plutonium fuel (MNIT) for FR.The developed scenario of CNFC, in addition to the safety, ecological and economic requirements,allows decreasing the risk of proliferation of weapon nuclear materials by eliminating the separa-tion of plutonium during reprocessing; decreases the volume of stored plutonium; recycling notonly uranium, plutonium and neptunium, but also transplutonium elements.In given scenario of CNFC liquid metal coolant (lead or sodium) is used to safety assurance of FR,since in the case of accidents involving the loss of integrity of a primary cooling circuit with liquidmetal the remaining heat removal is guaranteed. A lead option as one of the possible coolantsleads to choice of MNIT fuel due to its high density, better compatibility fuel composition withcoolant. Besides reactor with sodium coolant if using of MNIT seems to be in the best compliancewith inherent safety demand too. Currently in Russia BR-1200 with sodium (BN-1200) and withlead (BREST-1200) coolant is being carried out. MNIT is the unified fuel for both reactors.MNIT SNF reprocessing PH-process is being developed. PH-process is combined (pyro+hydro)technology based on non-aqueous operations, allowing the separation of actinides from fissionproducts with coefficient purification 100-1000, and hydrometallurgical refining operations withtotal coefficient purification 1·106. PH-process allows reprocessing of FR SNF f with high burn-up and short cooling time; decreasing the volume of stored SNF and the amount of plutonium inCNFC; reprocessing of any type FR SNF; obtaining final uranium-plutonium-neptunium productpurified for fuel refabrication using pellet technology.At present time design of the pilot demonstration power complex with BREST-OD-300 reactor isstarted. The complex includes not only reactor but also fabrication and refabrication facility (2018and 2024), reprocessing of SNF and waste treatment facility (2020).


Russian Federation / Bochvar Institute

Primary author: Dr IVANOV, Valentin (Bochvar institute)

Co-authors: SHADRIN, Andrei (Bochvar Institute); Mr SKUPOV, Mikhail (Bochvar Institute)

Presenter: Dr IVANOV, Valentin (Bochvar institute)


International Co … / Report of Contributions REGULATORY FRAMEWORK OF …


REGULATORY FRAMEWORK OF SPENT FUELMANAGEMENT FROM NUCLEAR REACTOR IN

INDONESIA

Upon the operation of 3 research reactors and there are plans for the construction of nuclear powerplants, Indonesia needs to provide a regulatory framework governing the management of spentnuclear fuel from these two types of reactors. For that reasosn Indonesia has a set of rules thatgovern it, management of nuclear spent fuel has been regulated in the law and the supportingregulations thereunder, namely government regulations, presidential regulation, and BAPETENChairman Regulation. The provision that have been set related to the duration of storage, provisionof final repository, nuclear liability, storage requirements, environmental monitoring, licensing,recording and reporting, transportation, etc. This paper provides an overview of the regulationsin Indonesia which regulates the management of nuclear spent fuel from the reactor.


INDONESIA/Nuclear Energy Regulatory Agency

Primary author: Mr ARYADI, Bambang Eko (Nuclear Energy Regulatory Agency of INDONESIA(BAPETEN))

Presenter: Mr ARYADI, Bambang Eko (Nuclear Energy Regulatory Agency of INDONESIA (BAPE-TEN))


International Co … / Report of Contributions VALUE OF EXPEDITED TRANSFE …


VALUE OF EXPEDITED TRANSFER OF FUEL FROMAT-REACTOR WET STORAGE

Following the challenges to wet storage of spent fuel during the events at Fukushima, the U.S.Nuclear Regulatory Commission (NRC) evaluated whether the reduced density of spent fuel inat-reactor wet storage would provide a substantial enhancement in safety. The reduced storagedensity would be achieved through expedited transfer of fuel to dry storage such that only fuelwith less than 5 years decay would remain in wet storage. The effect of changes in storage densityon the frequency and consequences of an event that drains the coolant from a spent fuel pool weredrawn from a detailed consequence study evaluating the effects of a large earthquake on a spentfuel pool in a Boiling-Water Reactor with a Mark I containment design. The study compared high-density and low-density loading conditions in existing fuel storage racks and assessed the benefitsof previously implemented mitigation measures. The study found that the density of spent fuelstorage had little effect on the frequency of a radiological release, but the consequences of a ra-diological release from unmitigated damage to the spent fuel pool structure could significantlychange. The NRC combined information from past studies of spent fuel storage safety and spentfuel storage licensing activities with the consequence study results to develop a conservative esti-mate of the safety benefit that could result from reduction in the density of spent fuel storage forthe entire fleet of at-reactor storage pools. This evaluation was consistent with previous studiesthat demonstrated high-density storage of spent fuel in pools protects public health and safety.Furthermore, the evaluation found that a transition to low-density storage would provide only aminor or limited safety benefit, and that its expected implementation costs would not be warrantedwhen compared with the expected benefits.


U.S.A.

Primary author: Mr JONES, Steven (U.S. Nuclear Regulatory Commission)

Presenter: Mr JONES, Steven (U.S. Nuclear Regulatory Commission)


International Co … / Report of Contributions Safe solutions for transport and dr …


Safe solutions for transport and dry storage ofdefective fuel rods

Defective fuel management is a major challenge for nuclear operators when time comes up forfinding a long-term solution for all their used fuel assemblies, defective included.

Managing the defective fuel assemblies indeed requires implementation of specific technical solu-tions.

Reprocessing the defective fuel assemblies is the most decisive way to get rid of them: there is thenno need to assess the behavior of this specific type of fuel assemblies in the long-term. Defectivefuel assemblies are being today safely transported to and reprocessed in a treatment plant.

When reprocessing is not the preferred option, the nuclear operator must choose between two drystorage solutions for the defective fuel assemblies: interim or long-term.

This paper describes the existing solutions for transportation and dry storage of defective fuelassemblies.

Transportation of defective fuel assemblies is today being performed. Various cask designs exist forthat purpose and specific operations are implemented for preparing the defective fuel assembliesfor transport. The advantage of such solutions is that they can support the operator in getting ridof the defective fuel assemblies once and for all.The paper presents the operational experience, the new licensing methodology relative to transportof defective fuel rods.

Interim dry storage of defective fuel assemblies has been implemented for decades in the USA.Specific operations are performed for preparing the defective fuel assemblies which are then storedalong with intact spent fuel assemblies in the dry storage cask. The advantage of such a solutionis its short-term cost effectiveness.The paper presents the existing dry interim storage solutions for defective fuel assemblies, theassociated operational experience.

A long-term dry storage solution for defective fuel assemblies is being developed, compliant withdirect disposal safety requirements. The solution is based on the dry encapsulation of each de-fective fuel rod into a fuel rod capsule, followed by its transfer into a capsule canister (like a fuelassembly skeleton), which is replacing a fuel assembly position of the transport and storage cask,and the associated specific licensing approach: this is the best available technology regarding long-term storage safety requirements.The paper presents this unique capsule technology, the transport and storage casks that can beused for transportation of defective fuel rods to the storage facility and especially focusses on theassociated robust safety demonstrations that support the transport and storage licensing approachcurrently used in Europe.


France / AREVA

Primary author: Ms VO VAN, Vanessa (AREVA NC)


International Co … / Report of Contributions Safe solutions for transport and dr …

Co-authors: Ms MORLAES, Isabelle (AREVA NP); Mr GARCIA, Justo (AREVA TN); Mr MUEN-CHOW, Kay (AREVA NP GmbH)

Presenter: Ms VO VAN, Vanessa (AREVA NC)


International Co … / Report of Contributions Progress towards a solution for int …


Progress towards a solution for intermediate storageof spent nuclear fuel in Norway

Spent Nuclear Fuel (SNF) in Norway has arisen from irradiation of fuel in the NORA, JEEP I andJEEP II reactors at Kjeller, and in the Halden Boiling Water Reactor (HBWR) in Halden. In totalthere are some 16 tonnes of SNF, all of which is currently stored on-site, in either wet or dry stor-age facilities. The greater part of the SNF, 12 tonnes, consists of aluminium-clad fuel, of which10 tonnes is metallic uranium fuel and the remainder oxide (UO2). Such fuel presents significantchallenges with respect to long-term storage and disposal.Current policy is for the fuel to be stored for a period of at least 50 years. In the meantime a na-tional final disposal facility should be constructed and taken into operation. Several committeeshave advised the Government of Norway on, among others, policy issues, storage methods andlocalisation of a storage facility. Both experts and stake holders have participated in these commit-tees. As the next stage in the process, a “Choice of concept” study was completed in early 2015.This paper presents an overview of the spent fuel in Norway and a description of current storagefacilities. The prospects and plans for long-term storage are then described, including a summaryof recommendations made to government, the reactions of various stakeholders to these recom-mendations, the current status, and the proposed next steps.It has been recommended that the aluminium-clad fuel be reprocessed in an overseas commercialfacility to produce a stable waste form for storage and disposal. This recommendation is contro-versial, and a decision has not yet been taken on whether to pursue this option. An analysis ofavailable storage concepts resulted in the recommendation to use dual-purpose casks.A further recommendation was that a public organisation, independent of the producer of the spentfuel, be founded to manage the SNF and that this organization also should have the responsibilityfor managing radioactive waste in Norway. Funding and operation of this organisation should bebased on the principle that the polluter pays.


Norway

Primary author: Mr BENNETT, Peter (IFE)

Co-authors: Dr OBERLANDER, Barbara (IFE); Mr LARSEN, Erlend (IFE); Dr REISTAD, Ole(IFE)

Presenter: Mr BENNETT, Peter (IFE)


International Co … / Report of Contributions Al-clad metallic uranium legacy fu …


Al-clad metallic uranium legacy fuel after 50 years inunderground dry storage . Spent fuel aging

management

Spent, legacy fuel elements from the early 50s and 60s originating from the first Norwegian RR,JEEP I, are stored in an underground dry storage since the mid 60s. It is known that some ofthe Al-clad naturally enriched metallic uranium fuel elements had defects by the time they weredischarged from the JEEP I RR and put into the cooling pond. In the mid 60s an underground drystorage site for JEEP I fuel was opened and the spent metallic fuel elements were transferred to thedry storage site: There, each element was stored individually in an open Al-container in a storagepit. Then, in 1982 the fuel elements were repacked into closed and tight stainless steel packages andstored dry, each element in one tight package and each package inside one underground storagepit. For many years the dry storage pits were sealed and opened for IAEA inspections, only. Now,and due to the age of the storage and the knowledge of the defects in the fuel, the spent fuel agingmanagement program has to include inspection of the storage pits with regard to corrosion of thepit lining, the state of stainless steel package, the state of the inside of the package with respect tohydrogen, humidity, corrosion and signs of fuel degradation.

The paper illuminates not only the NDT examinations performed and the findings, but also thechallenges, met after 50 years of storage. Next to the NDT examination the long time dry storageinduced development of the defects in the fuel elements was studied in detail by neutronradiogra-phy and the fuel and defect microstructure by metallography and SEM.


Institutt for energiteknikk (IFE), Kjeller, Norway

Primary authors: Dr OBERLÄNDER, Barbara Charlotte (Institutt for Energiteknikk, OECD HaldenReactor Project); Ms ANDERSSON, Vendi (Institutt for energiteknikk, Kjeller, Norway)

Co-author: Mr KLEEMANN, Hans-Jörg (Institutt for energiteknikk (IFE) Kjeller, Norway)

Presenters: Dr OBERLÄNDER, Barbara Charlotte (Institutt for Energiteknikk, OECD Halden ReactorProject); Ms ANDERSSON, Vendi (Institutt for energiteknikk, Kjeller, Norway)


International Co … / Report of Contributions First experience in basic design of …


First experience in basic design of a dry storagefacility for spent fuel in Argentina

Atucha I NPP was started up in 1974 initiating the nuclear energy production in Argentina. It is aHWR pressure vessel type of 350 Mw. With the purpose of solving the current situation that bothfuel pools capacity will be soon complete (Pool Buildings I and II) and considering the future lifeextension of the Plant, CNEA was asked to develop a project integrating systems in a new buildingfor dry storage of spent nuclear fuels, called ASECQ (acronym in Spanish: Almacenamiento enSeco de Elementos Combustibles Quemados).This project was carried out considering the requirements of the National Program for RadioactiveWaste Management (CNEA) and the recommendations of IAEA through a review mission called“Peer Review Mission IAEA IFMAP CNAI Spent Fuel Dry Storage Project” held from 12 to 16March 2012 in CNAI site. This revision was asked by NA-SA in the framework of Irradiated FuelManagement Advisory ProgrammeASECQ is being built in simultaneous with the operation of the NPP and it will work in the sameline of Pool Building I. In this way ASECQ will use the existing crane of Pool Building I and othersystems, already validated and operative for transport in vertical position of spent fuel elements.It will be also possible to return the spent fuels to the Pool in order to be removed for furthermanagement. It has a total storage capacity of 2754 Spent Fuel Elements.This configuration will facilitate the Operator (NA-SA) to transport the spent fuel elements and theNational Regulatory Authority (ARN) control the operation and inventory inside the proper Plant,fulfilling the nuclear surveillance and regulated safeguards. Monitored natural convection of airin normal operation was adopted outside the second barrier of protection. In case of an abnormalevent the natural convection is replaced by an alternative cooling and treatment closed system forthe air.


ARGENTINA / COMISION NACIONAL DE ENERGIA ATOMICA

Primary authors: Ms MASET, Elvira (Comisión Nacional de Energia Atómica); Mr LUNA DAVILA,Horacio (Comisión Nacional de Enería Atomica); Mr QUIROS, Horacio (Comisión Nacional de EnergiaAtómica); Mr CASAIS, Luis (Comisión Nacional de Energia Atómica); Mr FURRIEL, Miguel (ComisiónNacional de Energia Atómica); Mr BEUTER, Oscar (Comisión Nacional de Energia Atómica)

Presenter: Mr LUNA DAVILA, Horacio (Comisión Nacional de Enería Atomica)


International Co … / Report of Contributions Potential of active neutron interro …


Potential of active neutron interrogation to controlfissile materials in “closed” fuel cycle

One of the most important characteristic of a “closed” fuel cycle is a limitation of actinide losses by0.1% level. The work demonstrates a possibility to control the above parameter via the differentialdie-away technology in case of high gamma background that may be caused by fission products ofspent fuel. The preliminary results obtained by the experimental set-up based on this technologyshows that 8 min are enough to detect 0.4 mg 235U inside 68 liters empty container. In experimentswe used the ING-07T pulsed neutron generator produced by the VNIIA with 5*10^8 n/s neutronyield. Moreover, neutron spectral analysis can be employed for measuring isotopic compositionat any step of a fuel cycle


Russia

Primary authors: Mr SKLYAROV, Sergey (All-Russia Research Institute of Automatics (VNIIA)); DrBATYAEV, Vyacheslav (All - Russia Research Institute of Automatics (VNIIA))

Presenter: Mr SKLYAROV, Sergey (All-Russia Research Institute of Automatics (VNIIA))


International Co … / Report of Contributions On the impact of the fuel assembly …


On the impact of the fuel assembly design evolutionin the spent fuel management

ENUSA, the Spanish nuclear fuel manufacturer since the 1980s, is the main specialist in Spain ofLWR nuclear fuel design and performance. During this time, ENUSA has acquired deep knowledgeand experience on the fuel assemblies burned by its customers, and currently stored in the spentfuel pools. Eventually, ENUSA has also specialized in spent fuel management, offering integratedengineering and on-site services to the Spanish plants. In fact, ENUSA currently has the lead inclassification of spent fuel for dry storage and transportation projects developed in Spain.

Spanish spent nuclear fuel is classified following the US NRC legislation, particularly, the InterimStaff Guidance, ISG, 1 Rev. 2, which fixes the damaged fuel definition based on function. Alongdecades, fuel assembly design changes were focused on performance improvement to increase theend of life burnup and extend the fuel assembly useful life. However, not enough attention waspaid on the impact of these changes and their consequences on the future spent fuel management.

Spent fuel management requires handling, drying, good behavior of the materials at higher tem-peratures, transportation, etc, at normal and accident conditions. The paper summarizes ENUSA’sPWR and BWR designs evolution identifying the main design features which have significantlyimpacted in the spent fuel management. Remarkable items are, for instance:

a)Old fuel rod designs with high initial internal pressure to avoid cladding collapse and shortplenum lengths to compensate the irradiation growth. The lower volume and the high EOL internalpressures generate hoop stresses close to 90 MPa, stress limit according to ISG 11 Rev. 3, even after30 years of cooling;

b)Stress corrosion cracking of top nozzle sleeves made of sensitized SS304 has caused handlinglimitations on thousands of PWR fuel assemblies, or

c)Zirconium oxide spalling of Zircaloy 4 fuel rods irradiated to high burnups has become a hot issuein spent fuel transportation and storage management due to local embrittlement of the claddingin the spalled areas.

Giving the impact of these particular issues in the spent fuel management, deep knowledge of thefuel assembly designs and the potential effects during spent fuel management are fundamental inorder to accomplish a correct classification for dry storage and transportation. Moreover, sincefuel assemblies design keeps continuous evolution to improve their performance and extend theiruseful life, the design process must analyze the impact of the design changes in the future backend.


Spain/ENUSA Industrias Avanzadas

Primary author: Dr GARCIA DE LA INFANTA BELIO, Juan Maria (Spent Fuel - Development andEquipments Technology - ENUSA Industrias Avanzadas)

Co-author: Ms LLORET, Miriam (ENUSA Industrias Avanzadas)

Presenter: Dr GARCIA DE LA INFANTA BELIO, Juan Maria (Spent Fuel - Development and Equip-ments Technology - ENUSA Industrias Avanzadas)


International Co … / Report of Contributions Safety tests of spent fuel storage s …


Safety tests of spent fuel storage systems in Korea

In preparation for the timely installation of interim storage facility for spent nuclear fuel (SF), KoreaRadioactive Waste Agency (KORAD) is developing domestic models of SF storage systems, and theconcrete storage cask is one of them. A concrete cask consists of a metallic canister, which confinesSF with welded closure and a concrete overpack, which provides radiation shielding and physicalprotection to the canister. The safety requirements for a SF storage cask is well established in theUS, and summarized in regulatory guides such as NUREG-1536. Korea Atomic Energy ResearchInstitute (KAERI) has been performing tests of the concrete cask to demonstrate its safety andcompliance to the regulatory requirements with high priority stipulated in NUREG-1536. The testprogram includes the structural performance tests under a tip-over and earthquake, and decay heatremoval test under normal, off-normal, and accident conditions. In this paper, a brief introductionto the test program and test results are provided.


Korea/Korea Atomic Energy Research Institute

Primary author: Dr LEE, Sanghoon (Korea Atomic Energy Research Institute)

Co-authors: Mr LEE, Ju-Chan (Korea Atomic Energy Research Institute); Dr SEO, Kiseog (KoreaAtomic Energy Research Institute); Mr BANG, Kyung-Sik (Korea Atomic Energy Research Institute); DrCHO, Sang-Soon (Korea Atomic Energy Research Institute); Dr YU, Seung-Hwan (Korea Atomic EnergyResearch Institute); Dr CHOI, Woo-Seok (Korea Atomic Energy Research Institute)

Presenter: Dr LEE, Sanghoon (Korea Atomic Energy Research Institute)


International Co … / Report of Contributions Hydride effect on cladding behavi …


Hydride effect on cladding behaviour for spent fuelstorage and transport conditions

In Spain, the fuel is stored in the spent fuel pools and in on-site interim dry storage facilities whenthe pool capacity is reached. Additionally, a centralized temporary storage facility to dry storespent fuel assemblies, based in a vault system, is under construction.

Fuel conditions (internal and external pressure, temperature, etc) differs from pool to dry storage.The morphology and distribution of the hydrogen picked-up during in-reactor operation couldevolve due to the evolution of these conditions. The dissolution and re-precipitation of the hydridesin the radial direction could affect the mechanical behaviour of the fuel rods as the clad couldbecome brittle. The clad behaviour during spent fuel interim storage and subsequent transportis a key factor to assure the fuel safety functions, taking into account that the transport will beperformed after years of storage and cladding ductility will decrease as temperature decreases.

Spanish organizations CSN, ENRESA and ENUSA have carried out research and development pro-grams to characterize the behaviour of spent fuel under transport conditions.

Fresh cladding material has been electrochemically charged at different Hydrogen concentrationsfrom 150 to 2000 ppm. The mechanical and fracture behaviour has been studied using Ring Com-pression Tests (RCT) to simulate the pinch forces generated in the contact between the claddingand the spacer grid under hypothetical transport accident conditions. The tests have been per-formed at different strain rates (0.008, 1.7, 17 mm/s and 3000 mm/s) and different temperatures (20,135 and 300℃) representatives of dry storage and transport conditions. Micro and macro fracturemechanisms have been analysed.

Additionally a finite element model, based on the experimental RCT load vs. displacement curveshas been developed to calculate the fracture energy as a function of the hydrogen concentration.

Pre-hydrided samples with homogeneous and circumferential hydrogen distribution, do not presentbrittle behaviour. Rupture is produced at displacements higher than 3 mm, even for 2000 ppm hy-drogen, the lowest temperature (20℃) and the highest strain rates.For radial hydride reoriented samples brittle fracture are obtained for low displacement values at20℃ and 135℃.


spain

Primary author: Ms LLORET, Miriam (ENUSA Industrias Avanzadas)

Co-authors: Mr FERNANDEZ, Francico Javier (ENRESA); Mr REY, Jose María (CSN)

Presenter: Ms LLORET, Miriam (ENUSA Industrias Avanzadas)


International Co … / Report of Contributions A Novel Approach for Monitoring …


A Novel Approach for Monitoring Highly ActiveWastes

A realistic and reliable monitoring of Highly Active (HA) wastes and infrastructure is one of themost important challenges currently facing the nuclear industry. Despite the constraints and po-tential dangers imposed by high intensity radiation, there exists an increasing necessity to moni-tor HA environments in nuclear reprocessing and decommissioning. Gamma imaging has alreadybeen proven to be a very successful tool for monitoring medium activity (MA) infrastructure. How-ever, the radiation fields in HA facilities can be thousands of times more intense, and therefore farmore harmful both to human resources and sensitive equipment. Monitoring HA environmentsthrough gamma imaging remains a technologically challenging area. Inspired by the challengeof the Fukushima-Daiichi accident and given our previous experience regarding MA monitoringthrough gamma imaging, we have set out to develop a novel imaging system for HA environments.In this report, we provide original results on detector dynamics in high radiation fields and a newapproach to HA inspection and monitoring. More specifically, we describe the results of a feasi-bility study regarding the development of a gamma camera for HA environments with radiationintensity up to 10,000Gy/hr. The development of this lightweight and rapidly imaging gamma cam-era includes a novel coded aperture imaging approach, a minimal use of electronic components,and an appropriate choice of materials and readouts. This Japanese government funded project isprimarily intended to help identifying radioactive debris at the Fukushima establishments and dueto its encouraging and pragmatic results, it is expected to have significant applications regardingthe surveillance, monitoring and decommissioning of ageing nuclear power stations.


United Kingdom/Createc (Create Technologies Ltd)

Primary author: Dr CHRISTODOULIDES, Kyriakos (Createc)

Co-authors: Dr SHIPPEN, Alan (Createc); Dr MELLOR, Matthew (Createc)

Presenter: Dr CHRISTODOULIDES, Kyriakos (Createc)


International Co … / Report of Contributions In-Situ Quantification of Reproces …


In-Situ Quantification of Reprocessing Residues with3D Gamma Imaging

Spent fuel reprocessing leads to accumulation of fission product residues within the reprocessingequipment. Although these are managed during operations through regular washouts, many partsof the system will continue to hold residues for a variety of reasons. As reprocessing plants reachthe end of their operating lives, the significance of these residues changes as the focus switchesfrom operations to Post Operation Clean Out (POCO) and eventual decommissioning. In particularthere is a desire to optimise the POCO process so that the wastes during decommissioning can beminimised.We present results and learning from several applications of 3D gamma imaging for assessingdistribution and activity of reprocessing residues at Sellafield in the UK. Sellafield has three gen-erations of reprocessing plant, one of which has begun decommissioning and the other two ofwhich are preparing to enter POCO within the next 5 years. The N-Visage 3D gamma imagingsystem has been applied to several areas of these plants to evaluate its performance as a POCOand decommissioning management tool. These areas include solid fuel processing and dissolution,first strip, and waste stream concentration and storage.Gamma 3D gamma imaging has been able to improve the planning process for decommissioningand POCO in a number of ways. In the case of the decommissioning of a fuel cutting cell, knowl-edge of the 3D distribution of activity of Cs 137 was used to forecast the evolution of dose rateswithin the cell during decommissioning, enabling a clear ALARP justification to be made for thepreferred decommissioning method. By comparing the ratio of isotopes with different half lives ina mixer settler, it was possible to demonstrate that residues were accumulating over time within aparticular component, identifying a potential focus area for POCO. By mapping isotopes indepen-dently a forecast of the dose rate after a ‘cool down’ period can be made on a zone-by-zone basis,to aid decommissioning planning.Finally, we will also present some initial results from Fukushima, indicating how similar processescan contribute to the management of nuclear accidents.


UK

Primary author: Dr MELLOR, Matt (Createc)

Co-authors: Dr SHIPPEN, Alan (Createc); Mr WOOLHOUSE, Chris (REACT Engineering)

Presenter: Dr MELLOR, Matt (Createc)


International Co … / Report of Contributions Renewing Dry Spent Fuel Storage …


Renewing Dry Spent Fuel Storage Certificates ofCompliance and Specific Licenses

The United States (U.S.) Nuclear Regulatory Commission (NRC) regulations for storage of spentnuclear fuel (SF) permit initial storage terms of up to 40 years. Certificates of Compliance (CoCs)for SF storage system designs, or specific licenses to store SF, can then be renewed for up to 40years. Renewal authorization requires time-limited aging analyses and aging management pro-grams (AMPs) that demonstrate that structures, systems, and components important to safety willcontinue to perform their intended functions for the requested period of extended operation.

NRC staff has recently completed the renewal review for a specific license at the Calvert CliffsNuclear Power Plant site, and is currently reviewing renewal applications for a second specificlicense and two CoCs. In addition, work is underway on a revision to the staff guidance for thesafety review of specific-license and CoC renewal applications. The guidance will define an ac-ceptable method for the NRC staff to review and determine if the applicant demonstrates that thestorage system, or independent spent fuel storage installation, will continue to meet the applica-ble regulatory requirements during the renewal duration. In addition, the guidance will provideexamples of aging management programs for (1) localized corrosion and stress corrosion crack-ing of welded stainless steel dry storage canisters, (2) reinforced concrete structures, and (3) highburn-up fuel performance.

This paper will address the recent experience of the NRC staff in reviewing renewal applicationsand in developing the revised staff guidance, including discussion of example AMPs for concretestructures, and for localized corrosion and stress corrosion cracking of welded stainless steel drystorage canisters.


United States/United States Nuclear Regulatory Commission

Primary author: DAVIS, B. Jennifer (USNRC)

Co-authors: DUNN, Darrell (USNRC); WISE, John (USNRC); BANOVAC, Kristina (USNRC); TOR-RES, Ricardo (USNRC)

Presenter: DAVIS, B. Jennifer (USNRC)


International Co … / Report of Contributions Activities of the OECD/NEA/CSNI …


Activities of the OECD/NEA/CSNI Working Groupon Fuel Cycle Safety in Spent Fuel and High-Level

Waste Management

The OECD Nuclear Energy Agency (NEA) assists member countries in ensuring the adequate safetyof existing and future nuclear installations, through maintaining and developing the knowledge,competence and infrastructure needed to regulate and support the complete life cycle. NEA’sWorking Group on Fuel Cycle Safety (WGFCS) brings together representatives of regulatory bod-ies, their technical support organisations, and operators of nuclear fuel cycle facilities, which com-prise of interrelated activities, including: uranium mining and milling; uranium refining and con-version to uranium hexafluoride; uranium enrichment; fuel fabrication and storage; spent fuelstorage; spent fuel reprocessing; decommissioning of nuclear facilities; radioactive waste manage-ment and disposal options (including for spent fuel); and the research and demonstration facilitiesthat support these activities. This paper discusses current WGFCS activities in the area of spentfuel and high-level waste (HLW) management.

Consistent with its mandate, the WGFCS, in cooperation with the NEA Working Group on FuelSafety, held a workshop on the safety of long term interim storage (LTIS) facilities in Munich, Ger-many, in May 2013. The workshop covered national approaches, safety requirements, regulatoryframework and implementation issues, technical issues, operational experience, and research anddevelopment for LTIS.

As follow-on to recommendations from the workshop, and recommendations from NEA’s Com-mittee on the Safety of Nuclear Installations, the working group is developing a Technical OpinionPaper (TOP) that will gather information presented at the workshop. This information will be com-plemented by information supplied by WGFCS members to develop a picture of the current statusof LTIS in NEA member countries.

The TOP will provide an overview of LTIS requirements and technical needs, based on approachesin member countries. The TOP will focus on: national approaches and expectations for LTIS man-agement of spent fuel and HLW; spent fuel and HLW inventories and storage systems, and longterm strategies being considered; regulatory framework, policies and regulations; licensing pro-cesses and procedures; knowledge, data and regulatory gaps and challenges, and national pro-grams to address them; Identification of needs for research and development; and considerationsof lessons learned from the Fukushima accident.

The task will not duplicate ongoing technical activities (e.g. by the International Atomic EnergyAgency, the Electric Power Research Institute, the U.S. Department of Energy). Rather, the work-ing group will coordinate with these organizations and provide a compilation and analysis of themember countries’ position on long term interim storage, licensing requirements and gap identifi-cation.


OECD Nuclear Energy Agency

Primary author: BAILEY, Marissa (U.S. Nuclear Regulatory Commission)

Presenter: BAILEY, Marissa (U.S. Nuclear Regulatory Commission)


International Co … / Report of Contributions Performance Assessment of Fuel A …


Performance Assessment of Fuel AssembliesRemoved from Shut-Down RBMK Reactor Cores for

Reburning in Operating Power Units

A peculiarity of the RBMK reactors is online core refueling. So, the core contains fuel assemblieswith burnup varying from zero to the maximum design value. When the reactor is shut down fordecommissioning, the quantity of the fuel assemblies that have not reached the maximum designburnup may be significant. In 2018 – 2035, the design lifetime of the RBMK reactors operated inRussia will expire. In case of step-by-step decommissioning, such fuel can be burnt in the RBMKunits which are still in operation. The use of the fuel of good residual performance minimizes freshfuel requirements.To continue burning the fuel assemblies removed from the core of a shut-down reactor, they shouldbe cooled in a pool to decrease the decay heat. Once the acceptable residual heat is achieved, thefuel assemblies should be delivered to the reactor. These procedures are not standard, and theirinfluence on the fuel should be taken into account in making a decision on further burning of thefuel assemblies.The paper presents an assessment of candidate fuel assemblies for further burning in operatingpower units and analyses condition of RBMK FAs of different burnup. It also addresses thinningof fuel rod claddings induced by corrosion and fretting, changes in mechanical properties of the FAstructural materials, and criteria of fuel fitness for further burning by the main operational char-acteristics, i.e. dimensions, tightness, and weld health. Corrosion effect on the fuel rod claddingsduring cooling in the pool after the reactor defueling is assessed. The assessments speak for resid-ual performance of the RBMK SFAs sufficient for their safe reburning in operating units of thesame type.


Russian Federation

Primary author: Mr KANASHOV, Boris (Sosny R&D Company)

Co-authors: Mr KOSTYUCHENKO, Anton (Sosny R&D Company); Mr KUZMIN, Ilia (Sosny R&DCompany); Mr PEREPELKIN, Sergey (Sosny R&D Company); Mr CHESANOV, Vladimir (Sosny R&DCompany)

Presenter: Mr KUZMIN, Ilia (Sosny R&D Company)


International Co … / Report of Contributions Russian Experience and Proposals …


Russian Experience and Proposals on Managementof Non-Conforming SNF of RBMK Reactors

The RBMK-1000 power reactors have been operated at Leningrad, Kursk and Smolensk NPPs inthe European part of Russia. The yearly fuel discharge from the RBMK-1000 reactors makes up3500 SFAs placed in reactor cooling pools and separate spent fuel wet storage facilities at NPPs.Since 2011, in accordance with Rosatom’s con-cept the RBMK SNF has been converted to safer drystorage in a centralized dry storage facility at the Mining and Chemical Combine in Kraskoyarskregion, Siberia. According to the approved technology leak-tight SFAs with sound grid skeletons(conforming SFAs) are subject to dry storage at NPP storage facilities and transfer to the centralizedstorage facility. Conversion of defective and leaky spent fuel (non-conforming SFAs) to dry storagein NPP storage facilities and their transfer to the centralized storage facility were not ensured. Oneof the ways of management of defective and leaky RBMK-1000 SFAs is reprocessing at Mayak PA,separation of uranium, plutonium, neptunium, and vitrification of radioactive waste. In 2011, apilot batch of leaky SNF was removed from the Leningrad NPP and reprocessed to try out andverify engineering solutions. In 2014, a batch of leak-tight non-conforming SNF was shipped fromthe Leningrad NPP to a re-processing plant. The paper addresses Russian experience and proposalson management of non-conforming SNF of RBMK reactors.


Russian Federation

Primary author: Mr PEREPELKIN, Sergey (Sosny R&D Company)

Co-authors: Mrs KHAPERSKAYA, Anzhelika (State Corporation ROSATOM, Russian Federation); MrKANASHOV, Boris (Sosny R&D Company); Mr LOZHNIKOV, Igor (Leningrad NPP); Mr STAKHIV,Mikhail (Rosenergoatom Concern); Mr SMIRNOV, Valery (Sosny R&D Company); Mr SIMONOV,Vladimir (Leningrad NPP); Mr LOBKOV, Yuriy (Rosenergoatom Concern)

Presenter: Mr PEREPELKIN, Sergey (Sosny R&D Company)


International Co … / Report of Contributions BORATED STAINLESS STEEL ST …


BORATED STAINLESS STEEL STORAGE PROJECTTO THE SPENT FUEL OF THE IEA-R1 REACTOR

The IEA-R1 research reactor operates in a regimen of 64h weekly, at the power of 4.5 MW. In theseconditions, the racks to the spent fuel elements have less than half of its initial capacity. Thus,maintaining these operating circumstances, the storage will have capacity for approximately sixyears. Whereas the estimated useful life of the IEA-R1 is around twenty years, it will be necessaryto increase the storage capacity for the spent fuel. Dr. Henrik Grahn, expert of the InternationalAtomic Energy Agency on wet storage, visiting the IEA-R1 Reactor (September/2012) made somerecommendations: among them, the design and installation of racks made with borated stainlesssteel and internally coated with an aluminum film, so that corrosion of the fuel elements wouldnot occur. This work objective is the project of high capacity storage for spent fuel elements, usingborated stainless steel, to answer the Reactor IEA-R1 demand and the security requirements of theInternational Atomic Energy Agency.Key words: Research Reactor, Storage, Spent Fuel


Brazil / IPEN-CNEN/SP)

Primary author: Mr RODRIGUES, Antonio Carlos (IPEN-CNEN/SP)

Co-authors: Dr MADI FILHO, Tufic (IPEN-CNEN/SP); Mr RICCI FILHO, Walter (IPEN-CNEN/SP)

Presenter: Mr RODRIGUES, Antonio Carlos (IPEN-CNEN/SP)


International Co … / Report of Contributions Thermal Field Modeling of Spent F …


Thermal Field Modeling of Spent Fuel TransportContainer C-30

In Slovakia we use nuclear energy for more than forty years. In the past spent fuel was trans-ported to former Soviet Union. Since 1987 we store all spent fuel in Interim Spent Fuel Facility atBohunice site. So the spent fuel is stored before its final deposition or reprocessing. For transportof spent fuel we use transport container C-30. The paper describes thermal calculations of C-30.The capability of removal of residual heat from spent fuel is very important feature for every de-vice or facility. It is therefore important to understand the process of heat removal.First calculation of thermal field of transport container C-30 was done by the manufacturer of C-30in eighties. The calculation was used as an approval for the ability of sufficient heat removal.In nineties and in the beginning of 21st century the use of new type of fuel raised question, whetherC-30 could be used also for transportation of spent fuel with higher burnup and residual heat. Inthe application for the type approval of C-30 new calculation was realized, in order to demonstratethe ability to safely divert the residual heat. The evaluation of information technology enabled touse more detailed model then in previous calculations.Despite technological advices all models used one common assumption. The inside of the con-tainer was considered to be a homogenous heat source. This assumption may be sufficient fortype approval; however, in order to get most accurate results more detailed analysis of inside ofC-30 became desirable. New calculation of residual heat removal considers the inventory of C-30as a system of separate spent fuel assemblies - separate heat sources.Paper describes the results of thermal field modeling calculations.


Slovakia

Primary author: Mr VACLAV, Juraj (Nuclear Regulatory Authority of the Slovak Republic)

Presenter: Mr VACLAV, Juraj (Nuclear Regulatory Authority of the Slovak Republic)


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