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International Conference

VVER 2013

Organizers:

Experience and Perspectives after Fukushima

Book ofAbstracts

11 – 13 November 2013Prague, Czech Republicwww.vver2013.com

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obalka_VVER_abstrakty_Sestava 2 4.11.13 13:00 Stránka 1


VVER 2013Experience and Perspectives after Fukushima

Book ofAbstracts

Prague2013

VVER abstrakty_148,5x210mm 4.11.13 13:41 Stránka 1

International Conference VVER 2013 Experience and Perspectives after Fukushima

Book of Abstracts

Published by:AF POWER agency a. s.Thámova 18, Prague, Czech Republicwww.afpower.cz

First Edition, November 2013© AF POWER agency a. s.

ISBN 978-80-260-5279-1


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BĚLÁČ Josef, ALVEL, a. s.

BURKET Daneš, Czech Nuclear Society (Česká nukleární společnost)

DOSTÁL Václav, Czech Technical University in Prague

DUSPIVA Jiří, ÚJV Řež, a. s.

HANUS Václav, Czech Nuclear Society (Česká nukleární společnost)

KAWALEC Miroslav, Czech Nuclear Society (Česká nukleární společnost)

KOVAŘÍK Petr, Research Centre Řež (Centrum výzkumu Řež s. r. o.)

KŘÍŽ Zdeněk, Research Centre Řež (Centrum výzkumu Řež s. r. o.)

LINHART Stanislav, ALVEL, a. s.

MATYÁŠTÍKOVÁ Monika, Grayling

NĚMCOVÁ Terézia, Czech Technical University in Prague

NESRSTA Vlastimil, Rusatom Overseas

ŠÍMA Zdeněk, Rusatom Overseas

TUŠA Norbert, AF POWER agency a. s.

VLACH Roman, Rusatom Overseas

ZDEBOR Jan, ŠKODA JS a. s.

Programme Committee:


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PAPERS

Design Implementation of Stress-test Measures on Czech and Slovak NPPsANDĚL Jan, FABIÁN Karol . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .7

Design Issues of Fuel Rod Cladding Made of Composite Material Based on Carborund (SiC) forConcept of Water Cooled Reactors Safety under Accident Conditions BEZUMOV V. N., NOVIKOV V. V., KABANOV А. А., ZAKHAROV R. G., PIMENOV Y. B. . . . . . . . . . . . . . . . . . . . . . . .8

NPP Temelín Power Uprate ProjectBÍCA Martin . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .9

Stress-tests Design Modifications at ČEZ NPPsBÍCA Martin . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .10

Activities of TES s.r.o. in Frame of Temelín NPP Power up Rating Project PreparationBLÁHA Martin . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .11

EC Stress Tests Conclusions and Follow up Regulatory ActivitiesBRANDEJS Petr . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12

Use of Surveillance Data in RPV Integrity AssessmentBRUMOVSKÝ Milan, KYTKA Miloš . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .13

Past, Present and Future of the Paks NPP in HungaryCSERHÁTI András . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14

Radioactive Waste Free Release Measurement ExperienceDUBSKÁ Larisa . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .15

MELCOR Calculation for Support of Test on Accident Management Strategy on Debris Bed RefloodingDUSPIVA Jiří . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .16

Addressing I&C Post-Fukushima Requirements: From Regulatory Requirements Analysis to SystemsDesign and Hardened InstrumentationDUTHOU Arnaud, IKAZAKI Silvain, KUBÍNOVÁ Jana . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .17

Three Years of Experience with TVSA-T Fuel at NPP TemelínERNST Daniel, MILISDÖRFER Lukáš, SOUKUPOVÁ Marta . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .18

Compact Turbogenerators for Saturated Steam Application on Large Pressurized Water Reactors KAPIC Miroslav, PRCHLÍK Luboš, ZOUBEK Martin, DUCHEK Karel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .19

The Development of House Load Power Supply of Kozloduy NPP Before and After FukushimaKRASTEV Emil Borissov . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .20

SAFIMon – Monitoring System for Severe Accidents at Nuclear Power StationsKÖNIG Wolfgang . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .21

Possibilities for Regulatory Cooperation in Licensing New VVER Type PlantsLAAKSONEN Jukka . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .22

Nuclear Energy Research and Development: Perspective of the Company, the Czech Republic and the European UnionLACIOK Aleš . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .23

Possibilities Offered by WANO-MC in Supporting Emergency Management at VVER Type PlantsLUKYANENKO Andrey . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .24

Content:


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Use of Best Estimate Methods in Licensing of VVER ReactorsMACEK Jiří, MECA Radim . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .25

Safety Analyses for Power Uprate of VVER 1000/320MACEK Jiří . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .26

Post-irradiation Inspections on TVSA-T Fuel Assemblies at Temelín NPPMALÁ Martina, NERUD Pavel, MIKLOŠ Marek . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .27

Diagnostic Systems of VVER Primary Circuit ComponentsMATAL Oldřich, MATAL Oldřich, jr., ŽALOUDEK Josef, ŠIMO Tomáš, NESVADBA Lukáš, VÁVRA Michal . . . . .28

Zero Defect Level in VVER-1000MEČÍŘ Václav . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .29

The SCORPIO-VVER Core Monitoring and Surveillance System with Enhanced CapabilitiesMOLNÁR Jozef, VOČKA Radim . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .30

The "Sandra Z100" in Position Evaluation System for Control Rods in VVER1000NOVÝ Ladislav . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .31

LTO – IAEA Safety Standards, SALTO Peer Review Service, Results of SALTO Missions at VVER PlantsPOLYAKOV Oleksiy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .32

Application of the Dry Filter Method for Containment Filtered VentingSASSEN Felix . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .33

Equipment for the Disposal of Neutron Flux Sensors and Thermocouples for VVER 440 and VVER 1000RUDOLF Antonín . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .34

Support of Nuclear Safety by Modeling and Study of Hypothetical Severe Accidents of VVER Nuclear ReactorsSTREJC Martin, KISELOVÁ M., OTCOVSKÝ T., SÁZAVSKÝ P., ŠRANK J., Udalov Y. P. . . . . . . . . . . . . . . . . . . . . . .35

Evaluation of Operating Experience with Use of PSAŠTVÁN František . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .36

Verification of the KNI-LM Assembly Funcionality for the Reactor Coolant Level Monitoring SystemTANZER Michal, RINKE Lenka . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .37

Renovation of the NPP Dukovany. Technical Opportunity and Economical ExpediencyTOSHINSKY G. I., PETROCHENKO V. V., BÍŽA Karel, JOHN Aleš . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .38

Upgrading and Operational Experience at the Loviisa NPPTUOMISTO Harri . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .39

Nuclear Fuel for NPP: Current Status and Main Fields of the DevelopmentUGRYUMOV Alexander . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .40

Application of the Filtered Containment Venting System for Different Types of ReactorsWISNIEWSKI Susanne, ECKARDT Bernd . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .41

ŠKODA JS a.s. and its Cooperation with Czech UniversitiesZDEBOR Jan, PAVLIS David . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .42

Implementation of Configuration Management Information System (CMIS) in ŠKODA JS a.s.ZDEBOR Roman . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .43

Content:


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The IVR Strategy for VVER 1000 type 320ŽĎÁREK Jiří, KRHOUNEK V., BATĚK I. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .44

Temperature and Radiation Effect on the RPV Concrete CavityŽĎÁREK Jiří, BRABEC P. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .45

POSTERS

Calculation of Thermal Reactivity Coefficients for NPP Mochovce-3,4, Start-up Conditions by MCNP5FARKAŠ Gabriel, VRBAN Branislav, LÜLEY Jakub, HAŠČÍK Ján, SLUGEŇ Vladimír, PETRISKA Martin, HINCA Róbert, ŠIMKO Juraj . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .46

Research of Advanced Nuclear Technologies in CANUTJIŘIČKOVÁ Jana, GLASBERGER Tomáš, PEROUTKA Zdeněk . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .47

Comparison of Burnable Absorber Elements for VVER Nuclear FuelLOVECKÝ M., PITERKA L., PREHRADNÝ J., ŠKODA R. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .48

Practical Acquiring of the PARCS Code for 3D Analyses of Neutronic Behavior of VVER1000/V320MAZZINI Guido, MIGLIERINI Bruno, RUŠČÁK Marek . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .49

Content:


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Design Implementation of Stress-testMeasures on Czech and Slovak NPPs

ANDĚL Jan1, FABIÁN Karol2

1ÚJV Řež, a. s., ENERGOPROJEKT PRAHA division , [email protected]ÚJV Řež, a. s., ENERGOPROJEKT PRAHA division

In the aftermath of Fukushima event, the European Nuclear Safety Regulators Group (ENSREG) askedfor an evaluation of the plant robustness in case of extreme external events, loss of power supply orultimate heat sink and also for ability to mitigate consequences of severe accidents.This task was resolved in the “Stress tests” frame. The results were, besides other, requirements foradditional safety measures to improve NPP robustness. These inputs were transformed into designconcepts and later into documents of feasible design. ÚJV Řež, a. s. ENERGOPROJEKT PRAHA divisiontook significant part in these activities together with NPP operators in Czech Republic and SlovakRepublic and other subjects.This paper deals with conceptual design considerations and compares features of various possibletechnical solutions that further resulted in proposal of design solution. The paper describes theseissues for NPP Dukovany and Temelin in Czech republic and for NPP Mochovce, units 3 and 4, inSlovak Republic.


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Design Issues of Fuel Rod Cladding Made of Composite Material Based on Carborund(SiC) for Concept of Water CooledReactors Safety under Accident Conditions

BEZUMOV V. N.1, NOVIKOV V. V.2, KABANOV А. А.3, ZAKHAROV R. G.4, PIMENOV Y. B.5

1JSC «A.A. Bochvar High-technology Research Institute of Inorganic Materials» (JSC «VNIINM»), [email protected], 3, 4JSC «A.A. Bochvar High-technology Research Institute of Inorganic Materials» (JSC «VNIINM»)5JSC «TVEL»

Use of zirconium alloys for fuel rods claddings provides reactors cores safety. Zirconium alloys havelow capture cross sections, satisfactory thermal conductivity, strength at normal operatingconditions. In the other hand, emergency temperature rise of fuel rods claddings results in thebeginning of the zirconium-steam reaction at 900 °С and the rapid oxidation at 1200 °С with therelease of hydrogen from water. To avoid reactions with hydrogen formation, use of materials, whicheliminate the possibility of forming an explosive mixture, have a low oxidation kinetics and highmelting points, is obviously necessary. Particularly, a lot of attention around the world is given to composite materials based on thecarborund (SiC), heat-resistant alloys and steels. The paper presents the following: the main problemof design a fuel rod cladding made of a material based on the carborund (SiC) that allows to achievethe mechanical, strength, thermophysical and other properties, which provide the possibility of fuelrods long-term operation in cores of WWER type reactors; results of the study test models of thin-walled carborund tubes that were produced by different technologies; results of the study tubesamples of SiC-SiC composites (SiC-SiC-fibre and SiC-SiC powder, CVD-process). The results ofphysical-chemical studies and performance of WWER type tubes test models, produced by differenttechnologies. Moreover, density and void volume, mechanical properties (compressive and flexural)at temperatures below 600°С and thermal coefficient of linear expansions were studied. Structuralresearch of carborund tubes by the light-microscopical method and the scanning electronmicroscope method and corrosion research were carried out.


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NPP Temelín Power Uprate Project

BÍCA MartinČEZ a.s., [email protected]

In 2010 ČEZ a.s. launched the power uprate project of NPP Temelín (2 x VVER-1000). The upratednominal reactor power level is 3120 MWt (104% of original RTP). The project shall be completed inSept. 2013. Together with the subsequent upgrade of the turbine low pressure parts (independentproject), it will render the net increase of unit electric output of approx. 64 MWe.The project is based upon exploitation of the reserves of the original design, i.e. it does not requireany extensive upgrade or modifications of plant systems / equipment. This is a key factorcontributing to the outstanding economical effectiveness of the project. The scope of the projectconsisted of the completion of new safety analysis and plant relicensing, analysis of the uprate impact on systems and equipment(incl. strenght and lifetime evaluation), plant operation (incl. determination of new I&C settings),effluent limits, etc. and execution of plant startup tests.Main contractors for the project were NRI Řež and JSC TVEL (fuel vendor) with the mainsubcontractor OKB Gidropress (RCS designer). The conference presentation gives the basic overview of the project progress and results.


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Stress-tests Design Modifications at ČEZ NPPs

BÍCA MartinČEZ a.s., [email protected]

Based on the stress-tests outcome, various safety upgrades are being implemented on both NPPsoperated by ČEZ (i.e. Temelín 2 x VVER-1000 and Dukovany 4 x VVER-440) in the period from 2012 to2015. The conference presentation gives the basic overview of these upgrades. The additionaltechnical provisions for design basis conditions address potential weak points of the existing design.For NPP Dukovany they include seismic reinforcement of vital buildings, new independent ECW fancooling towers for UHS and upgrade of PAMS (addition of several cat. 1 parameters). For bothDukovany and Temelín, reinforcements against floodings have been implemented. The other group isrepresented by new technical provisions for design extension conditions as defined by the stress-tests submission, i.e. SBO and loss of UHS. These are called diverse systems and increase diversity andredundancy in execution of key safety functions. They are composed of stationary diverse powersources and stationary means for RCS/SFP makeup and secondary heat sink. The subsequent layer isrepresented by alternative technical means to provide alternative power source and RCS/SFPmakeup. These are mobile means based on the FLEX approach. Furthermore, ČEZ implements alsoseveral technical provisions for mitigation of severe accident consequences, specifically hydrogenrecombiners for both Temelín and Dukovany and in-vessel retention means for Dukovany. Veryimportant role for effective emergency response play the means for communication amongemergency response personnel. The communication infrastructure has been amended by severaldevices like field phones or satellite phones to cope with complete loss of original infrastructure.


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Activities of TES s.r.o. in Frame of TemelínNPP Power up Rating Project Preparation

BLÁHA MartinTES s.r.o., [email protected]

TES s.r.o. is an expert engineering company based in the Czech Republic. Since its foundation in 1992the company was focused on engineering services and technical support of nuclear power plantsoperation. TES Company experts gained deep and extensive experience in frame of many largeprojects related to NPP commissioning, technology and I&C systems refurbishment, power up-rating,computational analyses of operational and accident transitions for both VVER-1000 and VVER-440NPPs located in Czech Republic. TES Company also won several European Union’s tenders for TACISprojects aimed at improving the nuclear safety of Russian nuclear power plants. The main activities ofTES Company in frame of Temelín NPP power up-rating project preparation to 104% of nominalpower (3120 MWt) are described with respect to similar upcoming projects for VVER-1000 plantssuch as in Bulgaria and Ukraine. Specifically, the contribution of TES Company’s experts is presentedin frame of startup programs and tests preparation for Temelín NPP first startup after power up-rating. The extent of the work and some of main results are also discussed related to thermalhydraulic analyses performed in frame of new set-points setting verification of limitation system andsystems of normal operation affected by power up-rating at Temelín NPP.


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EC Stress Tests Conclusions and Follow upRegulatory Activities

BRANDEJS PetrState Office for Nuclear Safety, [email protected]

In response to the accident at the Fukushima Dai-chi NPP, caused by the earthquake and subsequenttsunami, the European Commission launched the so-called Stress Tests (STs) on all operating nuclearpower plants in the European Union.The organization of European Nuclear Regulators (ENSREG) was responsible for setting of the technicalcontent of STs. It was practically re-evaluation of operated NPP resistance to extreme external events.National regulatory authorities were responsible for issuing National Reports containing the resultsof the STs and proposed measures for NPP safety and robustness improving. The National Reportswere assessed by international peer review also with participation of the public. Conclusions toimprove the safety and resistance to extreme conditions has been adopted also by the extraordinaryconference of the Convention on Nuclear Safety organized by the IAEA in Vienna in 2012.All these findings were included in the National Action Plans (NacP). SONS in cooperation with theČEZ as operators of Dukovany and Temelin NPP, issued the Post Fukushima National Action Plan. Thisplan has referred to all recommendations issued by ENSREG, IAEA and resulted from internationalpeer review. NAcP has demanded implementation of all measures from operators in required terms. SONS evaluate the safety of the Dukovany and Temelín NPP continuously. Many of the measuresproposed in the NAcP have been identified during Periodic Safety Review (PSR) and at the time of STsthe part of them were already implemented or were in the implementing stage. SONS task is toensure that all corrective actions will be implemented in prescribed terms by using possibilities ofthe national legislative. A number of recommendations in the STS conclusions are beyond the scope of national legislationsand it is the task of national regulators to prepare new or revise existing legislation. SONS has startedthis process well before the announcement of the STs. SONS during the preparation of newlegislation has responded to the incentives created during the STs evaluation process and graduallyhas applied all respective recommendations.At present, the new Atomic Act proposal is in the official comment procedure. SONS aim is not onlyto ensure the implementation of all measures of STs for operated NPPs, leading to improving safety,but also to ensure that possible new nuclear source licensing process will be in compliance with thelevel required by all international requirements and the current and forthcoming EC legislation.Despite some reservations about the STs declaration, such as undefined criteria and evaluationmethods, the evaluation beyond the scope of most national legislations, short preparation time, theresult was very positive. The proposed measures clearly lead to improving safety and robustness ofoperated NPPs in Europe. The STs results showed that the WWER 440 units have sufficient timereserve to severe accidents and conservatism of the project. The STs should be taken as anextraordinary activity that should not be repeated. The obligation to regularly assess and control thelevel of nuclear safety and require maintaining the highest achievable level of the safety is a clearduty of independent national regulators, given by the national legislation.


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Use of Surveillance Data in RPV IntegrityAssessment

BRUMOVSKÝ Milan1, KYTKA Miloš2

1ÚJV Řež a. s., [email protected]ÚJV Řež a. s.

Irradiation embrittlement is the most severe degradation mechanism in RPVs during thein operation.Two ways are used for the assessment of its role – trend curves given in standards and codes and/ortrend curves obtained from testing surveillance specimens.Normative trend curves are usually applied in cases when there is not sufficient number ofsurveillance specimen data and/or for checking of the RPV real/archive material behaviour withthem. This approach requires a large database of surveillance test data obtained from many reactors.Unfortunately, fleet of WWER reactors is relatively small; moreover, the same welding materials wereused for manufacturing of several RPVs, thus the variety of chemical composition is small. In suchcase, scatter of data is large and uncertainty σ for evaluation of upper boundary trend is also largewhich can result in a large material embrittlement to be used in integrity evaluation.Direct use of real surveillance data from a given RPV requires a standard procedure for theirevaluation/fitting into the trend curve and also knowledge about the scatter of properties of RPVmaterials due to their inhomogeneities, uncertainty during testing, effect of specimen location onone side and uncertainty in determination of irradiation conditions – neutron dosimetry andirradiation temperature on the other side. IAEA NULIFE Guidelines for Assessment of Integrity and Lifetime of Components and Piping in WWERNPPs during Operation describes fully this procedure for both ways. This procedure is based onanalysis of existing databases of surveillance specimen test data, results from research on materialproperty inhomogenities in large ingots and weldments as well as uncertainties in test proceduresand neutron dosimetry.Normative trend curves are given for 15Kh2MFA(A) materials for WWER-440 and for 15Kh2NMFA(A)for WWER-1000 type reactors. Procedures for trend upper boundaries are also described includingnecessary information about the material scattering. Several examples are also given to verify thesetrend curves for both types of materials.


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Past, Present and Future of the Paks NPP in Hungary

CSERHÁTI AndrásMVM Paks NPP, [email protected]

At the end of last year we celebrated in Paks the 30th anniversary of the first power reactor start-upand of the unit initial connection to the grid. The presentation/paper gives a short historic overviewof the early operation, describes the most important achievements, as well as shows some futureplans.The robust and conservative soviet VVER-440/V313 design features served as good basis. Thedomestic weak point determination, development activities, modern safety reassessments and theimplemented upgrading measures made the plant more reliable and safe.

In particular:Safety reviews (the so called AGNES project, PSR, SAR updates, post Fukushima stress-test) and theresulting projects (e.g. increase of seismic capacity, reactor protection system refurbishment, safetyculture campaign, severe accident management).Added facilities (plant simulator and maintenance training center, interim spent fuel storage etc.).Nuclear capacity extension steps (power up-rate to 500 MW – done; lifetime extension with +20 years– in progress; introduction of 15 month fuel cycle – preparation started; new build – 5 PWRcandidates preselected, BIS composed, project company formed).


VVER 2013

15

Radioactive Waste Free ReleaseMeasurement Experience

DUBSKÁ LarisaENVINET a.s., [email protected]

The lack of radioactive waste characterization on nuclear sites produces a negative effect upon thesafety of storage and economic efficiency of radioactive waste management. The routine operation and decommissioning of nuclear facilities creates a great amount of variousmaterials that have to be removed from the site. Part of these materials can be released into theenvironment without the necessity of their long-term disposal at the repositories. The dismantlersface the problem of choosing an efficient method of measurement of these materials to have reliableinformation on waste characteristics. The great increase of disposal costs encouraged thedevelopment of the effective Free Release Measurement procedures to minimize the RW volumes.ENVINET specialists in co-operation with the Czech Metrology Institute have developed a new typeof FRM system, “MUM”, that provides the most precise measurements at very low activity. Thedevelopment was based on the company’s know-how – metal-free shielding material which allowsthe low activity measurement to be taken.The MUM system is the new equipment for measuring the RW prior to its release into theenvironment. The previous models of FRM systems and equipment built by ENVINET are currentlyused for measurements of all the RW originating from both Czech NPPs, and were also supplied forthe Free Release Facility in the Lithuanian Ignalina NPP which is now under decommissioning, alsofor several nuclear facilities in Ukraine and Russia. Reliability and effectiveness of the systems havebeen proven by many years of operation in the European nuclear power industry.


VVER 2013

16

MELCOR Calculation for Support of Teston Accident Management Strategy onDebris Bed Reflooding

DUSPIVA JiříÚJV Řež, a. s., [email protected]

UJV gained a significant experience from many previous MELCOR code applications to Quenchbundle tests (performed in KIT, Germany) resulted in the support of a preparation of the newQUENCH-DEBRIS test. Pre-test simulation was not possible without important modifications of thesource code and/or very serious simulation assumptions together with an interpretation of results.The reason for such important changes comes from the configuration of the test bundle for thisQUENCH-DEBRIS test, which uses two different materials for cladding. The inner unheated two ringsare made from the standard cladding material Zry-4 and the outer heated ring rod cladding is madefrom Hafnium (also shroud is made from Hf). Modifications of the code consist of a substitution ofthe oxidation of the Stainless Steel with Hf (oxidation kinetics, and mainly with material properties ofreactants and products of the oxidation, because most of them are included in the source code andcan not be changed from the user input). Correctness of performed modifications was validated against separate effect test performed atKarlsruhe Institute of Technology. Successful simulation of the SET test confirmed codemodifications. Then the modified code was used for the bundle pre-test simulation with the aim todefine conditions for the complete oxidation of the inner rings as an important assumption for thebundle relocation with a formation of the debris bed to be consequently reflooded.Test was performed in KIT at the end of January 2013 with new name Quench-17. UJV plan toperform post test simulation, including reflooding phase.


VVER 2013

17

Addressing I&C Post-FukushimaRequirements: From RegulatoryRequirements Analysis to Systems Designand Hardened Instrumentation

DUTHOU Arnaud1, IKAZAKI Silvain2, KUBÍNOVÁ Jana3

1Rolls-Royce Civil Nuclear I&C, [email protected] Civil Nuclear I&C, France3Rolls-Royce Civil Nuclear I&C, Czech Republic

Following Fukushima events, regulatory authorities have issued new recommendations andrequirements for severe and post-accident systems. From specification work to implementation, eachphase of the process to integrate these new Instrumentation and Control (I&C) functionalitiesrequires specific expertise:Methodology to link regulatory requirements to functionalities and systems definitionImplementing these specifications at system levelQualified hardened instrumentation and equipmentConsidering plant specificities (seismic/flooding…), interpreting the regulatory requirements toproduce the functionalities and description of the post-accident systems is a complex process. Moreover, Defence-in-Depth principles and installed equipment impose additional constraints suchas Diversity requirements or separate PAMS and SAMS.These specifications established, the design of the post and severe accident systems must beadapted to each situation. For example, Rolls-Royce offers a step-by-step approach to tailor asolution adapted to the plant specific constraints and severe environmental conditions: Digital orHardwired systems (up to 1E/Cat A qualified) for diversity, seismic-resistant equipment and hardenedinstrumentation.To complete these systems, qualified ruggedized sensors able to withstand the extreme accidentconditions must be used.

Rolls-Royce provides complete and modular solutions to address post-Fukushima I&C requirementsfor both newbuilds and upgrades, notably for VVER: regulatory guidelines analysis, specification anddesign of corresponding systems, hardened instrumentation and equipment supply.


VVER 2013

18

Three Years of Experience with TVSA-TFuel at NPP Temelín

ERNST Daniel1, MILISDÖRFER Lukáš2, SOUKUPOVÁ Marta3

1NPP Temelín, [email protected], 3 NPP Temelín

It could be said the year of 2000 was the milestone in nuclear industry. Till 2000 all VVER reactors werealways loaded with “Russian fuel”. NPP Temelin was the first VVER-1000 which was loaded with “Americanfuel” (VVANTAGE-6 fuel). However, after ten years of operation with Westinghouse fuel (the operationalexperience was presented during VVER-2010 conference) ČEZ has decided to change the fuel supplier.Russian company TVEL is now fuel supplier for both Czech NPP - NPP Dukovany (VVER 440) and NPPTemelin (VVER-1000). Full core of fresh TVEL fuel assemblies – named TVSA-T - was loaded on Unit 1 inJuly 2010 and on Unit 2 in May 2011. It means, Unit 1 finished 3rd cycle and Unit 2 2nd cycle with TVSA-Tduring 2013. Except of 1st “TVSA-T” cycle on Unit 1 all other cycles were leakers free. Existing operationalexperience shows good fuel performance and mechanical stability. The TVSA-T main design features are as follows:TVSA-Т is a shroudless hex fuel assembly with a strong skeleton.Skeleton is formed of six angle pieces of radiation-resistant zirconium alloy E635 and spacer grids welded to them. Eight spacer grids are optimized from the viewpoint of fuel rod cluster displacement effort underconditions of radiation and thermal growth. Fuel rod fastening is in the lower part by an anti-vibrationgrid and a fuel rod bottom nozzle. Important fact which helped us to monitor fuel performance is theuse of the Fuel Repair and Inspection Equipment (FRIE). FRIE was originally designed by Westinghouse tomeasure and consequently confirm compatibility of Zircaloy-4 cladding with the original VVER-1000water chemistry and also to monitor fuel assembly behavior during first four cycles on Unit 1. FRIE wasalso used for fuel repairs. Selected parameters as TVSA-T fuel assembly bow, twist and length aremeasured during outages from 2011. Measurements are performed by OAO TVEL with technical supportfrom ČEZ. Independent measurements are performed by CVŘ (Czech company from Řež). The PIIP (PostIrradiation Inspection Program) was also one of the main SÚJB requirements within the licensing processof TVSA-T fuel in 2010. From the core-design point of view all of the basic principles remain valid. Alsothe crucial parameters remain the same as they were used with VVANTAGE fuel. Nevertheless the newsafety analysis and new reload safety evaluation methodology was implemented. This methodologyrespects the Temelin Westinghouse-based I&C, the system of watched parameters (FQ, FΔH, etc.) andbasic principles of core design (CD) and reload safety evaluation (RSE) as they are applied at Temelin.However some specific modifications were done and it results in Temelin specific methodology. The system of computational codes for CD and RSE is based on conventional Russian codes in first cycles.At Temelin there are used Westinghouse system APA, which was slightly modified to be able to handlethe fuel with gadolinium. The APA system is mainly used for core monitoring system BEACON modelpreparation. BEACON system was also modified to correspond with specific requirements caused by thenew safety analysis approach. The new computational system ANDREA/HELIOS was implemented atTemelin and it is intended to be used as a default Temelin core-design code. Also the Skoda JS systembased on MOBYDICK 1000 is used for independent verification and a number of support calculations. The TVSA-T cores operation is does not indicate any serious issues and the cores behavior correspondswith predictions. Several differences (in contrary with VVANTAGE cores) were observed – like a differentaxial power oscillation stability, etc; mainly at fist cycle.


VVER 2013

19

Compact Turbogenerators for SaturatedSteam Application on Large PressurizedWater Reactors

KAPIC Miroslav1, PRCHLÍK Luboš2, ZOUBEK Martin3,DUCHEK Karel4

1, 2, 3, 4 Doosan Škoda Power

Doosan Skoda Power successfully delivered and commissioned 1000MW steam turbogenerators(STGs) for VVER1000 installed in Temelin NPP, Czech Republic. The delivery of these large 4-casingsteam turbogenerators had been a result of successful delivery of more than 20 Skoda units forreactors VVER440. All STGs on VVER440 have been successfully modernized with additional efficiencygain of minimum 3% and expected to operate additional 30 years. In addition, high pressure parts of1000 MW Skoda STG units operating on VVER1000 have been successfully modernized and shortlyflow path of LP parts will be modernized as well using modern LSB 1220 mm. This LP modernizationwill result in the increase of total electric output by ca. 22 MW per unit. Final electric power of oneblock with VVER reactor and Skoda STG will exceed 80MWs. Currently Doosan Skoda Powerdeveloped new generation of LSB54” which in addition to large fossil units can be used on saturatedsteam TG with the maximum electric output of 1200MW. Modern design and comparison with halfspeed units is discussed in detail.


VVER 2013

20

The Development of House Load PowerSupply of Kozloduy NPP Before and After Fukushima

KRASTEV Emil BorissovKozloduy NPP, [email protected]

Kozloduy NPP continuously improves and modernizes the power supply for its House loads. I will show the principal idea and the development steps of the scheme for KNPP's House LoadsSupply. At the moment of the accident in Fukushima NPP, KNPP House Loads were supplied from 4 differentincoming feeders (most of the NPP in the world have only 2). On 13.5.2011, only two months afterthe Fukushima accident, we held a stress test for power suppling of all the first category DC and ACconsumers just from the own Safety System battery. The test showed us the time and thecharacteristics of the battery discharge. I will present diagrams of the stress tests results. On14.5.2012 we held another big stress test for power suppling of all 6kV and 0.4kV consumers of oneSafety System just from the additional stationary diesel generator 5.4MW. Our test proved that incase of accident the Operation personnel will have the opportunity to supply all the consumers ofany Safety System just from the additional DG. I will present the diagram of the stress tests results.


VVER 2013

21

SAFIMon – Monitoring System for SevereAccidents at Nuclear Power Stations

KÖNIG WolfgangAREVA GmbH, [email protected]

AREVA’s new monitoring system for severe accidents at nuclear power stations called SAFIMON will be presented. Due to its modular architecture and composition of various subsystems it can be applied to various types of nuclear plants of different manufacturers. It captures the mostsignificant parameters relevant during severe accidents, dose rates inside and outside of selectedbuildings and visual information about “hotspots” like the lower head of the reactor pressure vesselor bottle-necks at escape ways etc. A separate power supply with pure batteries or complemented by a power generator provides power independently from the plant´s grid which is essential duringstation black-outs. All measured data as well as audio-visual signals are forwarded to a specialobservation container at a safe and remote location. The integration of already existing subsystemsfrom various manufacturers is also possible as the starting with a plain basic system and its laterextension step by step.


VVER 2013

22

Possibilities for Regulatory Cooperationin Licensing New VVER Type Plants

LAAKSONEN JukkaState Atomic Energy Corporation “Rosatom”, JSC “Rusatom Overseas”

Regulatory co-operation for the safety design reviews of similar nuclear power plants being licensed in severalcountries has been organized under the umbrella of Multinational Design Evaluation Programme (MDEP). This co-operation was started with a pilot project in 2006 and converted into long-term programme in 2009 by ten nationsthat are currently operating nuclear power plants and have plans for new construction. The objective of MDEP is toidentify opportunities to harmonize and converge on licensing review practices, requirements, and acceptancecriteria. The OECD Nuclear Energy Agency provides the technical secretariat support for MDEP.Until now specific features of three reactor types have been reviewed in the MDEP programme: EPR, AP-1000 andAPR-1400. The participating members in each plant specific working group are countries that have the respectiveplant type under construction or under licensing review. It is expected further that persons attending the meetingsare experts having conducted by themselves safety review of the topics that are discussed in the group. In September 2013 the Policy Group of MDEP decided to start two new groups: ABWR and VVER-TOI. The countriesthat are founding members of the VVER-TOI group are Russia, India, Finland and Turkey. Three first of thesecountries are permanent members of MDEP and Turkey has been invited as an associated member. The work isscheduled to start in a meeting in January 2014.In the cooperation there is no possibility to review all safety features of the plant in question. In the VVER groupthere is an additional difficulty in finding topics of common interest to all parties: each of the three countriesimporting a VVER plant from Russia is getting a different version. India has built VVER-92 plants of 1000 MW powerand it is now negotiating on two additional units that may be different. Finland is planning to license an AES-2006plant (1200 MW) based on design of St Petersburg AEP. Turkey is going to build four 1200 MW plants that are veryclose to VVER-TOI type.For the co-operation work it is necessary to select design issues that are similar at all VVER types in question. Takinginto account that the lay-out of each plant is different and also the reactor containments and the passive safetysystems at the three plants are different, choice of common topics is quite limited. The presentation discusses thepossible topics for joint review. Good candidate topics for review that are common to all VVER plants are: • measures to limit embrittlement of the reactor vessel wall and to ensure adequate ductility of the steel for 60years of operation;• demonstration of break preclusion concept (including leak-before-break principle) used to ensure highreliability of the reactor coolant system pipelines;• tests and analyses conducted to demonstrate the adequate performance of systems designed to protectreactor containment after reactor core meltdown accident: core catcher, catalytic hydrogen recombiners, system forprimary pressure relief;• assessment of the compliance of I&C systems with the safety principles given in the IAEA Safety Standards andthe WENRA safety objectives for new NPPs, assuming that all exported plants have Rolls Royce designed protectionsystems;• assessment of the differences in safety characteristics and transient behavior of two different reactor cores,original AES-2006 and VVER-TOI;• assessment of adequacy of functional testing and analysis of the two different passive systems that transferheat from the steam generators to atmosphere.Participation to the plant specific Working Group would evidently be useful for all countries that have interest tobuild new VVER plants. Among the old VVER countries these are Ukraine, Czech Republic, Slovakia, Hungary,Bulgaria, and Armenia. Potential new VVER countries with earlier experience of nuclear power are South Africa,Argentina, Brazil, and the UK. The closest new entrants seem to be Belarus, Vietnam, Jordan, and Bangladesh.


VVER 2013

23

Nuclear Energy Research and Development:Perspective of the Company, the CzechRepublic and the European Union

LACIOK AlešČEZ, a. s.

The role of R&D in ČEZ is to improve operation of assets (reduction of emissions, improvement ofreliability, etc.) and to identify new opportunities in innovative technologies. Nuclear R&D currentlyforms the most important part in the overall R&D activities. Specific projects are focused on criticalareas connected with safe, economic, reliable and long-term operation of both NPP. Research ofdissimilar metal welds, aging models of cables, improvement of the radioactive waste managementprocedures, research of severe accidents manageability and consequences are some of the examples.The international cooperation in R&D is mainly materialized within the membership of CEZ in theEPRI and NUGENIA.The importance of the nuclear energy production in the Czech Republic is expressed in support ofR&D by public funds - Ministry of Trade and Industry and Technology Agency (projects in the Alfa andCentres of Competences programmes). It is expected that dedicated national energy R&D supportprogramme will be launched in the future. Ministry of Youth and Education supports internationalR&D, the participation of the Czech Republic in the Jules Horowitz Reactor project serves as anexample.European nuclear energy R&D strategy is formulated by the Sustainable Nuclear Energy TechnologyPlatform (SNE-TP) that identifies the priority areas for future institutional and financial support.NUGENIA Association was mandated by the SNETP to coordinate R&D in the Gen II/III reactorsystems. As an international association it is established as the principal integrator among utilities,technology providers, research and engineering organisation, SMEs, academia and technical safetyorganisations.


VVER 2013

24

Possibilities Offered by WANO-MC inSupporting Emergency Management atVVER Type Plants

LUKYANENKO AndreyWANO MC, [email protected]

WANO-MC presentation “Possibilities Offered by WANO-MC in Supporting Emergency Managementat VVER Type Plants” discusses the WANO-MC effort to improve emergency management. Thepresentation consists of two parts. The first part of the presentation “WANO Severe AccidentManagement (SAM) Project” presents the purpose of the SAM Project, SAM Project deliverables andexpected perspectives.The second part of the presentation “WANO-MC Regional Crisis Centre (RCC) Project” givesinformation on WANO-MC efforts to provide advice and technical assistance in the event of site areaemergency, general site emergency at WANO-MC VVER plants. Another RCC purpose is todisseminate information on safety relevant events at NPPs among its members. The presentationpresents the principles on which the RCC is based, RCC operation modes, utilities level ofparticipation, RCC status, including forthcoming activities.


VVER 2013

25

Use of Best Estimate Methods in Licensingof VVER Reactors

MACEK Jiří1, MECA Radim2

1ÚJV Řež a. s., [email protected]ÚJV Řež a. s.

Czech Republic operates 4 VVER 440 units and 2 VVER 1000 units. Their Safety Analyses areperformed with advanced best estimate computer codes of RELAP, ATHLET, CATARE type which weredeveloped for western PWRs. Up to now, these codes, while applied for licensing purposes, wereused with conservative boundary and initial conditions which required a number of sensitivityanalyses. The current state of uncertainty analysis methods and their incorporation into computercodes utilised for thermo hydraulic computations are best presented in the OECD studies andprojects (eg. BEMUSE, SM2A) prepared following recommendations of Committee of the Safety ofNuclear Installations with the purpose to forward development of advanced thermo hydrauliccomputer codes.Under preparation is a proposal of the methodical procedure to be applied for thermo hydraulicanalyses of some selected initiating events for VVER 440/213 and VVER 1000/320 reactors, whichtakes into account the mentioned trends and especially OECD recommendations. Considered is, forinstance, application of this method for the evaluation of such events as LB LOCA,r SB LOCA or otherinitial events, using uncertainty and sensitivity analysis use e.g. GRS nonparametric method base onWilks formula. The important part of the best estimate method is also validation of computer code onintegral experimental facilities.In the paper, the use of the GRS best estimate method is presented and the performed validations ofcomputer code ATHLET are summarized with description of dominant phenomena corresponding tothe analysed initiating events.


VVER 2013

26

Safety Analyses for Power Uprate of VVER 1000/320

MACEK JiříÚJV Řež a. s., [email protected]

In the Czech Republic, there are 4 WWER440 units and 2 WWER 1000 units in operation. At present, one of the current problems is feasibility of power uprate of these nuclear power plants.Specifically considered is the possibility to increase the core heat output by 4 – 8 %. The actualproposal was an increase of the core heat rate by 4 %, which corresponds to the 104 % of thenominal power of VVER 1000/320. Obviously, after the necessary changes, it is requisite todemonstrate that thus modified nuclear power plant is safe. Issuance of the subsequent newoperation license is contingent on the results of the Safety Report revision.The paper presents the implementation of safety analyses of the power uprate of a nuclear powerplant with VVER 1000 reactor and describes possible changes of the plant basic parameters.Discussion of these parameters impact on the method applied for the safety analyses performancewithin Chapter 15 (Safety Analyses) follows.Proposed is also a procedure applied for the selection of limiting initiating events and then -theactual solution. Briefly is evaluated possibility to apply the Best Estimate approach, taking intoaccount uncertainties of the input data as well as that of the computer codes used.


VVER 2013

27

Post-irradiation Inspections on TVSA-TFuel Assemblies at Temelín NPP

MALÁ Martina1, NERUD Pavel2, MIKLOŠ Marek3

1Centrum výzkumu Řež, s.r.o., [email protected], 3 Centrum výzkumu Řež, s.r.o.

Pool-side inspections on irradiated fuel are mostly provided in western countries, but theimplementation of higher burnup and longer cycles bring other needs of fuel inspections duringreactor operation worldwide. The post-irradiation pool-side inspections of fuel assemblies in theCzech Republic began in 2003 at NPP Temelín´s Unit 1 and in 2004 at Unit 2 after the first cycles withAmerican fuel VVantage-6 and they are a part of a long-term monitoring of fuel behavior. Thesemonitoring still continues on Russian fuel TVSA T. This paper describes the past experience with thefuel inspections and summarizes the results given by these inspections and measurements onselected fuel assemblies TVSA-T after three years of operation at Unit 1 and after two years at Unit 2.The post-irradiation inspections are provided primarily by the vendor JSC TVEL, CV Řež is theindependent fuel inspector that performs inspections in the cooperation with the fuel vendor.


VVER 2013

28

Diagnostic Systems of VVER PrimaryCircuit Components

MATAL Oldřich1, MATAL Oldřich, jr.2, ŽALOUDEK Josef3, ŠIMO Tomáš4, NESVADBA Lukáš5, VÁVRA Michal6

1, 2, 3, 4, 5, 6, Energovýzkum, Ltd., [email protected]

Structure of used diagnostic systems and their relation to life-time evaluation programs of VVERprimary circuit components is described in the first part of the paper. The second part of the paper deals with selected details of diagnostic tools applied at VVER 440 and1000 steam generators. Some examples of results obtained from diagnostic systems and furtherimplemented into practice show data flow needed for the component life-time/aging evaluationprocess.


VVER 2013

29

Zero Defect Level in VVER-1000

MEČÍŘ Václav ČEZ, a. s., [email protected]

Prezentation „Zero defect level in VVER-1000“ provides history of the project setup and current projetoutline along with scope of the Project. Basic approach to acheive goal along with basic steps andcurrent status is discussed.


VVER 2013

30

The SCORPIO-VVER Core Monitoring andSurveillance System with EnhancedCapabilities

MOLNÁR Jozef1, VOČKA Radim2

1ÚJV Řež, a. s., [email protected]ÚJV Řež, a. s.

The SCORPIO-VVER core monitoring system has proved since the first installation at Dukovany NPP in1999 to be a valuable tool for the reactor operators and reactor physicists. It is now installed on fourunits of Dukovany NPP (Czech Republic), on two units of Bohunice NPP (Slovak Republic) replacingthe original Russian VK3 system and on the full scale plant training simulator at the Centre fortraining and education of the reactor operators and reactor physicist in Trnava (Slovak Republic). Byboth Czech and Slovak nuclear regulatory bodies the system was licensed as a Technical SpecificationSurveillance tool. Since it’s first installation, the development of SCORPIO-VVER system continues along with thechanges in VVER reactors operation. The system is being adapted according the utility needs andseveral notable improvements in physical modules of the system were introduced. The latest mostsignificant changes were done in connection with implementation of a new digital I&C system,loading of the optimized Gadolinium bearing Gd2 fuel assemblies, improvements in the area of coredesign (neutron physics, core thermal hydraulics and fuel thermal mechanics), adaptation of thesystem to up-rated unit conditions, in design and methodology of the limit and technicalspecifications checking (implementation of the on-line shutdown margin calculation to the system)and improvements in the predictive part of the system (Strategy Generator).After the currently finished upgrades the SCORPIO-VVER is still in focus of Central European nuclearpower plants with the roadmap of modification and implementation up to 2016. In October 2012 the Upgrade 3 for the support of the new type of fuel with 4.87% enrichment of235U loaded into the reactors at EBO Slovakia has been finished successfully. Thanks to the upgradedsystem all units could reach their 100% of nominal power. During June 2013 the Upgrade 6 was started at EDU NPP in Czech Republic focused to the SCORPIO-VVER enhancement and new function development up to 2016.The Dukovany’s Upgrade6 presents the most advanced upgrade in the system’s history. TheSCORPIO-VVER system will be completely renewed, re-hosted to the new hardware, implemented thelatest advanced calculation codes with enhanced accuracy and adapted to the local, national andinternational IEC standards and requirements too.With the upgrade on the way and new installation under the preparation, the SCORPIO-VVER projectis very active and prepared to meet new challenges in future.


VVER 2013

31

The "Sandra Z100" in Position EvaluationSystem for Control Rods in VVER1000

NOVÝ LadislavZAT, a. s.

Principles used in ZAT Primis control systems based on Intel-196 controllers succesfully aplied from 90th

of 20th century gave strong base for modernized family of ZAT control systems "Sandra Z100". Thisplatform was used for demanding aplication in position evaluation system for control rods duringreconstruction at Temelin NPP in the year 2011-2012. The ZAT company has developed, manufacturedand delivered the modular system consist of 260 single electronic parts which consists more than 210microcontrollers of Sandra Z100 family. The modernized position evaluation system for control rods cancomfortably operate both types of positions sensors, UP-2 and stepping UP-3 as well, including theirwide diagnostic. The delivered system so can operate with both diferent sensors principle. Requirementsfor scaning and position evaluating as well as slow and fast processes on reactor control rods given byend user was succesfully converted in to practice.Aplication of requests can now make shorter time needed in reactor starting process in order hours andclosely monitor the overall behavior of the controlled technology.


VVER 2013

32

LTO – IAEA Safety Standards, SALTO PeerReview Service, Results of SALTO Missionsat VVER Plants

POLYAKOV OleksiyInternational Atomic Energy Agency (IAEA), [email protected]

Long term operation (LTO) of a nuclear power plant (NPP) is operation beyond established time frame setforth by the licence term, design, standards or regulations. Long term operation needs to be justified bya safety assessment considering life limiting processes and features for structures, systems andcomponents. International peer reviews are a useful tool for Member States to exchange experience, learn from eachother and apply good practices in dealing with LTO of NPPs. The SALTO peer review is a comprehensive peer review service directly addressing strategy and the keyelements of a safe LTO of NPPs including ageing management objectives and complements OperationalSafety Review Team (OSART) reviews. The SALTO peer review service is designed to assist NPP operatorsin adopting a proper approach to LTO of their plants and ensure that plant safety will be maintainedduring the LTO period.


VVER 2013

33

Application of the Dry Filter Method forContainment Filtered Venting

SASSEN FelixWestinghouse Electric Germany, [email protected]

If long term decay heat removal from the containment is lost during a severe accident, means to limitthe pressure inside the containment are essential to prevent catastrophic failure of the containmentand maintain fission product retention capabilities. For this reason containment venting is discussedin the nuclear society. Various implementations of containment venting are under discussion rangingfrom hardened venting to filtered venting. Westinghouse offers a range of different technologiessuitable for different plant applications in venting. The different Westinghouse technologies forcontainment venting will be presented together with the information how to chose the matchingtechnology for the plant specific needs and regulatory requirements.


VVER 2013

34

Equipment for the Disposal of NeutronFlux Sensors and Thermocouples for VVER 440 and VVER 1000

RUDOLF AntonínŠKODA JS a. s., [email protected]

Paper for the conference VVER 2013 is about the Equipment for disposal of neutron flux sensors andthermocouples for VVER 440 and VVER 1000. Device for VVER 440: The device is designed forremoving of radioactive sensors portion with max. length of 5 m from the block of protective tubesof VVER-440 reactor, for their transport over the storage cell located in the reactor hall and finally fortheir cutting into pieces for storage. Equipment consists of 2 modules, transport module and cuttingmodule. The transport module is consisting of a container with extracting mechanism and withfixture for transport by a crane. Container serves as a shielding body and is made of lead and steel.The transport module is used to remove the sensors from the block of protective tubes of VVER-440reactor, their transport and lowering into storage cell. The transport module can be usedindependently from cutting module. Cutting module is independent accessory that is used forcutting the sensor into the pieces for storage in the cell, if cutting is requested by NPP operator.Shielding body of cutting module is made of steel construction. The equipment is controlledremotely by the operator from a distance of about 8 m because the requested dose rate in place ofoperation is 14 microSv /h. Device for VVER 1000: The device is designed for removing of radioactive sensors from the block ofprotective tubes of VVER-1000 reactor, and for their cutting into pieces for storage. The deviceconsists of the shielding body with the mechanism for sensor drawing-in and with the cutting head.The cut sensors are collected in the removable can, which is later inserted into the shipping cask forfinal disposal outside power plant. Number of the liquidated sensors stored in one can is 5 – 6(amount required by customer ČEZ-ETE). The final number of sensors stored in one can is determinedwith regard to radiation limits and filling. The inner area of sensor liquidation is monitored by 2 videocameras. The cameras´ output is shown on a display which is located on the switchboard. The deviceis controlled by a portable controller with circa 5-meter cable. The controller contains controlling andsignaling elements. Detailed information about the equipment condition, including error messages,are shown on the display on the main switchboard of the device.


VVER 2013

35

Support of Nuclear Safety by Modelingand Study of Hypothetical SevereAccidents of VVER Nuclear Reactors

STREJC Martin1, KISELOVÁ M.2, OTCOVSKÝ T.3, SÁZAVSKÝ P.4, ŠRANK J.5, UDALOV Y. P.6

1ÚJV Řež, a. s., [email protected], 3, 4, 5 ÚJV Řež, a. s.6Saint-Petersburg State Institute of Technology (Technical University)

Nuclear energy represents a high-profile and sophisticated field of human activity. Technical supportof the reactor blocks operation is being developed. Nevertheless, it is necessary to pronounce thatthe risk of emergency state is non-zero. It is requisite in the maximum extent to prevent the radioactive products release into theenvironment in case of ex-vessel severe accident of nuclear reactor.During the severe accident, loss of nuclear reactor cooling means high risk of the core componentsmelting. Corium, the melt containg nuclear fuel and molten components of inner equipment ofreactor is formed. Currently, three concepts of retaining system for melt cooling withing the severe accident with thecore degradation are preferred:a) in-vessel melt retention (failed in Fukushima),b) controlled spreading of melt at the large horizontal surface of the pool, situated outside of the concrete reactor shaft, followed by the melt cooling,c) maintaining of melt inside the water cooled metal vessel, equipped from the inside by layers of the sacrificial material.At the Radioactive Waste Management Centre in ÚJV Řež, a. s. workplace we deal with the coriumissues and nuclear safety study in case of ex-vessel severe accident in the NPPs ("b" and "c" retainingconcepts mentioned above). We co-operate with experts from Russian Federation and EU. We use the"cold crucible" high-temperature melting facility for these studies.


VVER 2013

36

Evaluation of Operating Experience withUse of PSA

ŠTVÁN FrantišekÚJV Řež, a. s., [email protected]

The Loviisa Power Plant equipped with two VVER-440 reactors has gone through extensivemodification and modernization programmes during its operating life. The capability to utilize plant-specific features has been very useful when developing and implementing measures against internaland external hazards and a consistent Severe Accident Management programme to respond to theplant-specific vulnerabilities. Developing and implementing the management for internal hazardshas required continuous awareness and efforts from the plant licensee. In most cases we haveapplied integrated deterministic and probabilistic considerations to evaluate the upgrading needsand to study various possibilities and plant capabilities to resolve the raised issues. A typical feature of internal hazards is that they challenge simultaneously more than one functionallevel of the Defence-in-Depth concept or penetrate more than one of the physical barriers of thefission product releases. Internal hazards may themselves be initiating events, such as common causefailures, internal fires or floods, missiles, inhomogeneous boron dilution, or primary-to-secondaryleakage accident (PRISE). In many cases they are hazards that are created during the accidentprogression such as pressurized thermal shock, loop seal issue, boron crystallization, containmentsump clogging, or inherent boron dilution mechanisms. There has been also upgradings formanaging external hazards such as formation of frazil ice or concern about zebra mussels and darkfalse mussels. Development of Severe Accident Management programme started after the Chernobyland its implementation was complete after one and half decade. As a consequence of the Fukushimaaccident the Management of External Hazards and Severe Accident Management have been subjectto the increased attention at nuclear power plants. The lesson learnt was that there might be cases ofpaying too little attention to external hazards in comparison to the risk they might pose for the plant. The aim of this paper is to discuss experiences from the plant modifications and modernization aswell as their relation and impact on the overall operational experience. The Loviisa plantconfiguration is in many respects quite unique, since the original VVER-440 design has been addedwith ice condenser containment, specific reactor coolant pumps and many other features.


VVER 2013

37

Verification of the KNI-LM AssemblyFuncionality for the Reactor Coolant Level Monitoring System

TANZER Michal1, RINKE Lenka2

1NPP Engineering, ŠKODA JS a. s., [email protected] Engineering, ŠKODA JS a. s.

The monitoring system RVLIS (Reactor Vessel Level Instrumentation System) is intended for thecoolant reserve monitoring in the pressurized water reactors (PWR, respectively VVER). This type ofsystem became more important in the light of the accident at the NPP Fukushima. The system RVLISwas supplied and installed by the company SKODA JS during the recent modernization of the NPPDukovany (Czech Republic) and the NPP Jaslovske Bohunice (Slovak Republic).The system RVLIS acquires the signals from two special developed KNI-LM assemblies located insidethe reactor core. The KNI-LM assembly is derived from the standard internal neutron noisemonitoring assembly (the KNI assembly) but in its upper part, there are additionally placed threespecial electrical heating elements and three thermocouples for measuring of their temperature. Twomore unheated reference thermocouples monitor an ambient temperature. The monitoring principleis based on the fact that the temperature of the heating element immersed in water is significantlylower than a temperature of a heating element immersed in steam. The KNI-LM assembly 3D simulation models were created and subsequent heating analysis and realcondition simulations were carried out for further analysis. The verification of the KNI-LM assemblymathematical FEM (Finite Element Method) model was based on the comparison of the experimentalresults (normal ambient conditions, NPP real operating conditions and LOCA simulation conditions)and the results obtained from the corresponding FEM simulation model. The acquired results of experiments as well as calculation part declare good agreement and can beused for other types of tasks (simulations of different operational state or different fault state) andthese simulations are significantly important in the field of safety measurement in nuclear energy.


VVER 2013

38

Renovation of the NPP Dukovany.Technical Opportunity and Economical Expediency

TOSHINSKY G. I.1, PETROCHENKO V. V.2, BÍŽA Karel3, JOHN Aleš4

1JSC “AKME-engineering”, [email protected] “AKME-engineering”3ÚJV Řež, a. s., ENERGOPPROJEKT PRAHA Division, Czech Republic4ÚJV Řež, a. s.

There is sizeable difference in the service lifetime of nuclear steam supplying system (NSSS) (40-60 years) and rest infrastructure of the NPP (80-120 years). For that reason, there is anopportunity to replace the withdrawn capacities by renovation of power-units without constructing the new NPP. This opportunity is associated with use of modular fast reactor with lead-bismuth coolant (LBC). Based on the experience of exploiting LBC in nuclear submarines’reactors, such type reactors – SVBR-100 – have been developing in Russia. The inherent safety level of those reactors is very high. Renovation means that the required number of SVBR-100 reactor modules will be placed in thesteam-generators and main circulation pumps’ compartments after their dismantling. Altogether,these modules will produce steam with flow rate and parameters identical to those of reactorinstallation VVER-440. The technical opportunity and economical expediency renovation of NPPDukovany with four power units VVER-440 should be validated by project development.


VVER 2013

39

Upgrading and Operational Experience atthe Loviisa NPP

TUOMISTO HarriFortum Power, [email protected]

The Loviisa Power Plant equipped with two VVER-440 reactors has gone through extensivemodification and modernization programmes during its operating life. The capability to utilize plant-specific features has been very useful when developing and implementing measures against internaland external hazards and a consistent Severe Accident Management programme to respond to theplant-specific vulnerabilities. Developing and implementing the management for internal hazardshas required continuous awareness and efforts from the plant licensee. In most cases we haveapplied integrated deterministic and probabilistic considerations to evaluate the upgrading needsand to study various possibilities and plant capabilities to resolve the raised issues. A typical feature of internal hazards is that they challenge simultaneously more than one functionallevel of the Defence-in-Depth concept or penetrate more than one of the physical barriers of thefission product releases. Internal hazards may themselves be initiating events, such as common causefailures, internal fires or floods, missiles, inhomogeneous boron dilution, or primary-to-secondaryleakage accident (PRISE). In many cases they are hazards that are created during the accidentprogression such as pressurized thermal shock, loop seal issue, boron crystallization, containmentsump clogging, or inherent boron dilution mechanisms. There has been also upgradings formanaging external hazards such as formation of frazil ice or concern about zebra mussels and darkfalse mussels. Development of Severe Accident Management programme started after the Chernobyland its implementation was complete after one and half decade. As a consequence of the Fukushimaaccident the Management of External Hazards and Severe Accident Management have been subjectto the increased attention at nuclear power plants. The lesson learnt was that there might be cases ofpaying too little attention to external hazards in comparison to the risk they might pose for the plant. The aim of this paper is to discuss experiences from the plant modifications and modernization aswell as their relation and impact on the overall operational experience. The Loviisa plantconfiguration is in many respects quite unique, since the original VVER-440 design has been addedwith ice condenser containment, specific reactor coolant pumps and many other features.


VVER 2013

40

Nuclear Fuel for NPP: Current Status andMain Fields of the Development

UGRYUMOV AlexanderJSC «TVEL»

Fuel Company (FC) of Rosatom «TVEL» is one of the worldwide leaders in the nuclear fuel production. Seventy-six (76) powerreactors (17 percent of a world market) on NPPs with VVER, PWR and BWR reactors and 30 research reactors in 15 countries,operate using the fuel produced by FC «TVEL». FC «TVEL» has been supplying its products both to the traditional foreignmarkets for Russia, where the nuclear power plants built according to the Russian (Soviet) projects operate, and to WesternEurope countries. Currently, FC «TVEL» is supplying nuclear fuel and is carrying out scientific and technical support for thenuclear fuel operation in nuclear power plants of Ukraine, Czech Republic, Bulgaria, Hungary, Slovakia, Finland, Armenia andChina. The most important partner of FC «TVEL» within unified corporate structure of State Corporation «Rosatom» is«Rosenergoatom» Concern OJSC. FC «TVEL» has been supplying the nuclear fuel to «Rosenergoatom» Concern OJSC for theNPPs with VVER-1000, VVER-440, RBMK-1000, BN-600, EGP-6 reactors, as well as for new AES-2006 nuclear power plantsbeing built. The operation of research reactors in Hungary, Kazakhstan, Uzbekistan, Poland, Czech Republic, Ukraine,Bulgaria, Vietnam and Libya is supported by Russian nuclear fuel.While producing nuclear fuel, FC «TVEL» solves the following main tasks: - satisfaction of the customers claims in what concerns operational characteristics and improvement of the nuclear fuel’s technical and economical parameters; - provision of necessary safety level during the nuclear fuel operation and fabrication; - increasing competitiveness and extending the market for the products.Constant development of existing designs of the fuel assemblies (FA) for different types of power reactors is required forachievement of the goals mentioned.The works to solve these main tasks are carried out in the following fields:1. Provision of the FA design’s geometrical stability (for VVER-1000 reactor);2. Increasing the FA operational life time;3. Improving the FA operational reliability;4. Development of the dismountable (maintainable) FA;5. Realisation of safe and economy-efficient fuel cycles including: - improving the fuel burnup; - increasing the NPP units’ thermal power; - adoption of the fuel cycles with extended duration, for example, one and a half year; - provision of preserving the nuclear fuel‘s working capacity during the maneuvering operational modes; - decreasing of fast neutron flux in the reactor vessel and the reactor internals.Large amount of R&D and technological works was carried out during several last years jointly with the leading companies inthe industry. As an outcome, new kinds of nuclear fuel for main types of power reactors were developed and adopted, andare currently successfully operating in NPPs. These types comprise the following: - modifications of TVSA and TVS-2 fuel assemblies such as TVSA-PLUS, TVSA-12, TVS-2M fuel assemblies for VVER-1000 reactor; - 2nd and 3rd generations fuel assemblies for VVER-440 reactor; - FA having enrichment which is profiled along the height, with the central fuel rods fixing and debris-filter for RBMK reactor.The nuclear fuel’s technical and economical parameters (fuel burnup, fuel cycle duration, etc.) correspond to those of thenuclear fuel fabricated by worldwide producers of power reactors.Further work on the development and adjustment of the nuclear fuel fabrication technology, fuel rods and FA designs, andon the construction and fuel materials is underway. As an outcome of this work, efficient fuel will be supplied to thecustomers. The works on the development of the fuel for new projects such as VVER-1200/1300 reactor, 4th generation FAbased on TVSA and TVS-2 FA for VVER-1000 reactor are underway. The works on the preparation of the complex of theorganizational and technical measures directed towards the elimination of the root causes for the nuclear fuel failures inVVER reactors are underway. These works are needed in order to be able to reach zero failure of nuclear fuel and provide theoperation of «clean» reactor cores.One of the most important strategic goals for fuel company «TVEL» is expanding the geographic supply span andachievement of the leadership in the nuclear fuel market.


VVER 2013

41

Application of the Filtered ContainmentVenting System for Different Types of Reactors

WISNIEWSKI Susanne1, ECKARDT Bernd2

1AREVA GmbH, [email protected] GmbH

Various studies indicate that reactor containment can possibly fail from overpressure following a coremelt accident. A filtered containment venting system can prevent such a failure. But the cost of aventing system must be low enough that it can be justified for such an unlikely event as a core meltaccident.AREVA has developed a filtered containment venting system at justifiable cost for preventingpostulated overpressure failure of PWR, BWR, CANDU, and VVER containments following core meltaccidents. This filtered containment venting system has a decontamination factor of 10.000 foraerosols and decontamination factor of 500 for elemental iodine.As pointed out, AREVA’s experience with this type of equipment is dated back nearly two decades.Our first installation of a Filtered Containment Venting System took place in 1987 at NPP Krümmel,Germany, up to now more than 50 systems have been build or are under construction in Germanyand many other countries, like e.g. in Bulgaria, Canada, China, Finland, Japan.Configuration and dimensions of the filtered containment venting system can be easily adapted tothe regulatory and technical requirements applicable for the NPP of destination.


VVER 2013

42

ŠKODA JS a.s. and its Cooperation withCzech Universities

ZDEBOR Jan1, PAVLIS David2

1ŠKODA JS a.s., [email protected]ŠKODA JS a.s.

For more than 50 years, ŠKODA JS a.s. has been active in the field of nuclear power industry and thisperiod has been filled with cooperation with the academic sphere represented, among others, byCzech and foreign universities. In particular, there has been a very close cooperation with theUniversity of West Bohemia in Plzeň which started as early as in the 1950s upon adopting nucleartechnologies in the company ŠKODA. Scientific workers of the University of West Bohemia (formerCollege of Mechanical and Electrical Engineering) participated in resolving a number of technicalproblems and the experts of ŠKODA took part in the education and training of future experts. Therehas also been a long tradition in the cooperation with the Czech Technical University in Prague,Czech Technical University in Brno and Technical University of Ostrava. In the recent years, ŠKODA JS a.s. has also cooperated with Odessa National Polytechnic University in resolving theproblem of reliability of control rod drive mechanisms at Ukrainian nuclear power plants.


VVER 2013

43

Implementation of ConfigurationManagement Information System (CMIS)in ŠKODA JS a.s.

ZDEBOR RomanŠKODA JS a. s., [email protected]

The presentation deals with the project of implementation of the SW system in ŠKODA JS intendedfor complex management of large capital projects documentation and requirements in theelectronic form throughout the life cycle of a particular system or equipment.

The entire solution consists of several components: • Basic setting of the system for future project configuration management • Document Management System – DMS • Requirements Management & Traceability – RM&T

The objective of the above project is to prepare in ŠKODA JS in advance such (information)environment for large capital projects data management, in the future including the project of theTemelín NPP Units 3&4 completion, allowing to: • reduce the costs in the management of processes, documentation and requirements, • use established and proven processes (Best Practices) in ŠJS in the implementation of capital projects• make more precise and at the same time to simplify the monitoring of the progress in implementation of capital projects • contribute to increasing the overall competitiveness of ŠKODA JS company.The paper describes the current state of practice and experience from the implementation of CMIS.


VVER 2013

44

The IVR Strategy for VVER 1000 type 320

ŽĎÁREK Jiří1, KRHOUNEK V.2, BATĚK I.3

1ÚJV Řež, a. s., [email protected], 3ÚJV Řež, a. s.

After the Fukushima accident the final strategy for treatment of the SA will be required for theexisting NPPs. Recent results from the benchmark calculation on Molten Corium Concrete Interaction( MCCI) on VVER 1000 type 320 performed within the EU SARNET 2 – WP 6.4 project, provided clearanalytical results describing radial and axial abblation rates of the corium. Based on those results,very strong arguments are now available to proceed with the IVR strategy. In our presentation statusof analytical and experimental work supporting this IVR strategy will be described.


VVER 2013

45

Temperature and Radiation Effect on theRPV Concrete Cavity

ŽĎÁREK Jiří1, BRABEC P.2

1ÚJV Řež, a. s., [email protected]ÚJV Řež, a. s.

There are more than ten concrete degradation mechanisms with influence on the containmentbuilding. Most of them are followed with augmented inspections methods. The RPV concrete cavityis exposed to temperature and radiation damage as most critical damage mechanism. Until now,effective augmented inspection for this concrete location is not proposed.In our presentation, description of our project proposal targeted on development augmentedinspection is described.


VVER 2013

46

Calculation of Thermal ReactivityCoefficients for NPP Mochovce-3,4, Start-up Conditions by MCNP5

FARKAŠ Gabriel1, VRBAN Branislav2, LÜLEY Jakub3, HAŠČÍK Ján4, SLUGEŇ Vladimír5, PETRISKA Martin6, HINCA Róbert7, ŠIMKO Juraj8

1Slovak University of Technology, Faculty of Electrical Engineering and Information Technology, Instituteof Nuclear and Physical Engineering, [email protected], 3, 4, 5, 6, 7 Slovak University of Technology, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering8SE a.s.

In general a reactor is initially started up from a precondition state by withdrawing controlassemblies or by changing the boron acid concentration in primary circuit coolant until the reactor isslightly supercritical, thus producing an exponentially increasing neutron population on a very longperiod. As the neutron population increases, the fission heating and thus the reactor temperatureincreases. This increase in temperature produces a decrease in reactivity (almost all reactors aredesigned to have a negative temperature reactivity feedback). That would lead to the slowdown ofincreasing the neutron population, or stabilization on a new power level. Requirement of negativetemperature reactivity effect could be achieved problematically, particularly for the first core loading,when all fuel assemblies are fresh. Because of the first criticality start-up of the Nuclear Power Plant(NPP) Mochovce units 3 and 4 in near future, detailed analyses of core parameters are required by theSlovak Regulatory Authority to support safe operation of the nuclear facility. The article introducesdetermination of thermal reactivity coefficients, especially summary (isothermal) and moderator(density) reactivity coefficients between 200°C and 260°C with step of 2°C. The work presentscalculated critical parameters, especially critical boron acid concentrations at given coolanttemperatures and position of the 6th control assembly group. Numerical iteration procedure wasapplied to calculate the critical parameters as a substitute for a critical experiment. Geometrical andmaterial models were created in compliance with the reactor design and the first fuel loading of theNPP Mochovce 3 and 4. All mentioned calculations were performed by computational code MCNP51.60 supported by NJOY99.364 microscopic sections processing system and our control scripts.


VVER 2013

47

Research of Advanced NuclearTechnologies in CANUT

JIŘIČKOVÁ Jana1, GLASBERGER Tomáš2, PEROUTKA Zdeněk3

1ZCU v Plzni, [email protected], 3 ZCU v Plzni

The Centre for Advanced Nuclear Technologies (CANUT) defines a system and platform of long termcooperation in research, development and innovations among leading Czech universities andresearch institutions and renowned industrial companies in different fields of nuclear technologies.Foundation and kick-off of CANUT is supported by the Technology Agency of the Czech Republic andCANUT represents one of supported national centers of competence. This contribution brieflypresents a vision of CANUT research, the fields and topics on which the researchers in CANUT workon and reports selected results of their activities. CANUT covers the research objectives from fuelincluding whole fuel lifetime cycle, control and safety systems, diagnostics technologies includingsensors, robots and manipulators up to self-consumption systems.


VVER 2013

48

Comparison of Burnable AbsorberElements for VVER Nuclear Fuel

LOVECKÝ M.1, PITERKA L.2, PREHRADNÝ J.3, ŠKODA R.4

1Regional Innovation Centre for Electrical Engineering, University of West Bohemia, [email protected] Innovation Centre for Electrical Engineering, University of West Bohemia3Regional Innovation Centre for Electrical Engineering, University of West Bohemia, Czech Technical Uni-versity, Faculty of Mechanical Engineering4Regional Innovation Centre for Electrical Engineering, University of West Bohemia,Czech Technical Uni-versity, Faculty of Mechanical Engineering, Texas A&M University, Department of Nuclear EngineeringTAMU-3133

Nuclear fuel containing burnable absorber (BA) significantly improves fuel utilization during reactoroperation. BAs compensate for the initial excess reactivity and consequently allow for lower powerpeaking factors and longer fuel cycles with higher fuel enrichments.Burnable absorber selection comprises of element selection, its weight content and spacedistribution in fuel assembly. Selection of burnable absorber element requires large amount ofcomputer time for fuel depletion analysis, therefore, in the first step of BA analysis, only selectedelements were analysed with state-of-art industry code.The contribution compares three selected elements (Gd, Eu, Er) as burnable absorbers in VVER-1000fuel assembly. Number of burnable absorber bearing pins, their location and BA weight content wereused as input variables. Multiplication coefficient during duel depletion was calculated andcompared. Fuel depletion calculations for burnable absorber evaluation was performed byTRITON/NEWT code sequence from SCALE code package.


VVER 2013

49

Practical Acquiring of the PARCS Code for3D Analyses of Neutronic Behavior ofVVER1000/V320

MAZZINI Guido1, MIGLIERINI Bruno2, RUŠČÁK Marek3

1Research Centre Řež (CVRez), [email protected], 3 Research Centre Řež (CVRez)

In response to Fukushima accident the Research Centre Rez plans to simulate accident scenarios of NPPTemelin and NPP Dukovany analyzed in the stress tests ordered by the European Commission to allEuropean NPPs. While in a number of safety analysis of various accident scenarios it is sufficient to useone point reactor kinetics there are selected types of accidents in which it is useful to model the space(3D) behavior of neutron kinetics, in particular in control rod ejections, boron dilution scenarios,including transitions from design basis to beyond design basis accidents. The aim of the presented work is to prepare a model of the core of VVER1000/V320 reactor applicable for3D modeling of neutron kinetics in selected design and beyond design basis accidents and its couplingwith the thermo-hydraulic system codes, such as RELAP or TRACE. The model is based on reference datacreating a cross-sections library starting from SCALE 6.1.2/TRITON simulations. PARCS code useshomogenized cross-sections libraries to calculate neutronic and other core parameters of the VVERreactors.


Author Index:ANDĚL Jan . . . . . . . . . . . . . . . . . .7BATĚK I. . . . . . . . . . . . . . . . . . . . .44BEZUMOV V. N. . . . . . . . . . . . . . .8BÍCA Martin . . . . . . . . . . . . .9, 10BÍŽA Karel . . . . . . . . . . . . . . . . . .38BLÁHA Martin . . . . . . . . . . . . . .11BRABEC P. . . . . . . . . . . . . . . . . . .45BRANDEJS Petr . . . . . . . . . . . . .12BRUMOVSKÝ Milan . . . . . . . . .13CSERHÁTI András . . . . . . . . . .14DUBSKÁ Larisa . . . . . . . . . . . . .15DUCHEK Karel . . . . . . . . . . . . . .19DUSPIVA Jiří . . . . . . . . . . . . . . . .16DUTHOU Arnaud . . . . . . . . . . .17ECKARDT Bernd . . . . . . . . . . . .41ERNST Daniel . . . . . . . . . . . . . .18FABIÁN Karol . . . . . . . . . . . . . . . .7FARKAŠ Gabriel . . . . . . . . . . . .46GLASBERGER Tomáš . . . . . . .47HAŠČÍK Ján . . . . . . . . . . . . . . . .46HINCA Róbert . . . . . . . . . . . . . .46IKAZAKI Silvain . . . . . . . . . . . . .17JIŘIČKOVÁ Jana . . . . . . . . . . . .47JOHN Aleš . . . . . . . . . . . . . . . . .38KABANOV А. А. . . . . . . . . . . . . . .8KISELOVÁ M. . . . . . . . . . . . . . . .35KÖNIG Wolfgang . . . . . . . . . . .21KRASTEV Emil Borissov . . . . .20

KRHOUNEK V. . . . . . . . . . . . . . .44KUBÍNOVÁ Jana . . . . . . . . . . . .17KYTKA Miloš . . . . . . . . . . . . . . .13LAAKSONEN Jukka . . . . . . . . .22LACIOK Aleš . . . . . . . . . . . . . . . .23LOVECKÝ M. . . . . . . . . . . . . . . . .48LUKYANENKO Andrey . . . . . .24LÜLEY Jakub . . . . . . . . . . . . . . .46MACEK Jiří . . . . . . . . . . . . . .25, 26MALÁ Martina . . . . . . . . . . . . .27MATAL Oldřich, jr. . . . . . . . . . .28MAZZINI Guido . . . . . . . . . . . .49MECA Radim . . . . . . . . . . . . . . .25MEČÍŘ Václav . . . . . . . . . . . . . .29MIGLIERINI Bruno . . . . . . . . . .49MIKLOŠ Marek . . . . . . . . . . . . .27MILISDÖRFER Lukáš . . . . . . . .18MOLNÁR Jozef . . . . . . . . . . . . .30NERUD Pavel . . . . . . . . . . . . . . .27NOVIKOV V. V. . . . . . . . . . . . . . . .8NOVÝ Ladislav . . . . . . . . . . . . .31OTCOVSKÝ T. . . . . . . . . . . . . . . .35PAVLIS David . . . . . . . . . . . . . . .42PEROUTKA Zdeněk . . . . . . . . .47PETRISKA Martin . . . . . . . . . . .46PETROCHENKO V. V. . . . . . . . .38PIMENOV Y. B. . . . . . . . . . . . . . . .8PITERKA L. . . . . . . . . . . . . . . . . .48

POLYAKOV Oleksiy . . . . . . . . .32PREHRADNÝ J. . . . . . . . . . . . . .48RINKE Lenka . . . . . . . . . . . . . . .37RUDOLF Antonín . . . . . . . . . . .34RUŠČÁK Marek . . . . . . . . . . . . .49SASSEN Felix . . . . . . . . . . . . . . .33SÁZAVSKÝ P. . . . . . . . . . . . . . . .35ŠIMKO Juraj . . . . . . . . . . . . . . . .46ŠKODA R. . . . . . . . . . . . . . . . . . .48SLUGEŇ Vladimír . . . . . . . . . . .46SOUKUPOVÁ Marta . . . . . . . .18ŠRANK J. . . . . . . . . . . . . . . . . . . .35STREJC Martin . . . . . . . . . . . . .35ŠTVÁN František . . . . . . . . . . .36TANZER Michal . . . . . . . . . . . . .37TOSHINSKY G. I. . . . . . . . . . . . .38TUOMISTO Harri . . . . . . . . . . .39UDALOV Y. P. . . . . . . . . . . . . . . .35UGRYUMOV Alexander . . . . .40VOČKA Radim . . . . . . . . . . . . . .30VRBAN Branislav . . . . . . . . . . .46WISNIEWSKI Susanne . . . . . .41ZAKHAROV R. G. . . . . . . . . . . . . .8ZDEBOR Jan . . . . . . . . . . . . . . .42ZDEBOR Roman . . . . . . . . . . . .43ZOUBEK Martin . . . . . . . . . . . .19ŽĎÁREK Jiří . . . . . . . . . . . . .44, 45

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