disposal high level waste including spent nuclear fuel

International Conference on the Safety of Radioactive waste Management

SESSION 3d

Disposal of

High Level Waste

Including Spent Nuclear Fuel

Declared as Waste

Session 3d– HLW IAEA-CN-242

2

ORAL PRESENTATIONS

No. ID Presenter Title of Paper Page

03d – 00 INV 03d J. Heinonen

Finland

Regulatory Experiences from the Spent Fuel

Disposal Step-Wise Implementation

4

03d – 01 111 V. Havlová

Czech Republic

Complex Safety Assessment Model of

Radioactive Waste Deep Geological Disposal in

the Czech Republic

7

03d – 02 130 A. Hagros

Finland

Preparing Posiva’s Post-Closure Safety Case

Towards the Operational Phase

11

03d – 03 145 S. Voinis

France

Andra’s Safety Options of French Underground

Facility Cigeo – a Milestone towards the

Licensing Application

15

03d – 04 172 I. Niemeyer

Germany

Bridging Nuclear Safety, Security and

Safeguards at Predisposal and Geological of

High Level Waste and Spent Nuclear Fuel

20

03d – 05 141 T. Fujiyama

Japan

Development of the NUMO Pre-selection, Site-

specific Safety Case

25

03d – 06 131 L. Bailey

United Kingdom

Development of a Generic Environmental

Safety case for the Disposal of Higher Activity

Wastes in the UK

29

03d – 07 184 D. Pellegrini

France

SITEX, the European Network of Technical

Expertise Organisation for Geological Disposal

33

03d – 08 207 A. Ström

Sweden

Research and Development Needs in a Step-

Wise Process for the Nuclear Waste Programme

in Sweden

38

Session 3c – ILW IAEA-CN-242

3

POSTER PRESENTATIONS

No. ID Presenter Title of Paper Page

03d – 09 26 F. Charlier

Germany

Germanys New Route towards a Repository for

High Level Waste – Scientific Challenges

43

03d – 10 32 J.-W. Kim

Korea

Recent Safety Assessment of a Reference

Geological Disposal System for Radioactive

Waste from Pyro-Processing in Korea

47

03d – 11 34 Y. Kovbasenko

Ukraine

Assessment of Decay Heat in Process of Spent

Nuclear Fuel Disposal

51

03d – 12 94 S. Suzuki

Japan

Assessment of Pre- and Post –Closure Safety in

the NUMO Safety Case for a Geological

Repository

56

03d – 13 96 J. Stastka

Czech Republic

Research, Development and Demonstration

Projects at the Josef Underground Laboratory

60

03d – 14 125 V. Maree

South Africa

The Management of Used (Spent) Fuel and

High Level Waste in South Africa

64

03d – 15 140 J. Leino

Finland

Regulatory Experiences in Reviewing

Construction License Application for the

Disposal of Spent Nuclear Fuel in Finland

69

03d – 16 155 L. Vondrovic

Czech Republic

Generic Underground Research Facility in the

Middle Stage of the Site Selection Process:

Bukov URF, Czech Republik

73

03d – 17 161 F. Launeau

France

Cigeo Project: from Basic Design to Detailed

Design – pursuant to Reversibility

77

03d – 18 171 B.B. Acar

Turkey

Impact of Storage Period on Safe Geological

Disposal of Spent Fuel

81


4

03d – 00 / INV 03b. Disposal of High Level Waste

REGULATORY EXPERIENCES FROM THE SPENT FUEL DISPOSAL STEP-WISE

IMPLEMENTATION

J. Heinonen

Radiation and Nuclear Safety Authority (STUK), Finland

E-mail contact of main author: [email protected]

1. Introduction

Finland began planning and preparing for nuclear waste management measures in the 1970s,

during the procurement and construction phase of the first nuclear power plants. In 1983, the

Finnish Government made a policy decision on the principles and schedules of nuclear waste

management. In 2000, the Government adopted a favourable Decision-in-Principle (DiP)

accepting the concept of a deep disposal facility for spent fuel from the Finnish nuclear

power plants in Olkiluoto and Loviisa. This DiP was confirmed by the Parliament in 2001. A

construction licence application (CLA) for the encapsulation and disposal facility was

submitted to the Government in 2012. The Finnish government granted in 12th November

2015 Posiva license to construct spent nuclear fuel (SNF) encapsulation plant and disposal

facility in Olkiluoto. Government decision was supported with STUK’s safety evaluation.

Before encapsulation and disposal process begins, Government has to issue a operating

license. Operating license application is expected to be submitted in early 2020’s and disposal

is planned to start 2023.

STUK, as the independent safety regulator, has been performing stepwise review of

developing safety case and R&D needed to demonstrate safety of SNF disposal. STUK

strategy has also been that safety regulation is developed coincide with the development of

disposal. This approach has enabled to include experiences and growing knowledge timely

into safety regulation.

STUK also decided to participate actively in pre-siting and pre-license phase. Active

participation has included pro-active public communication and step-wise evaluation of site

characterisation work and development of safety case. Based on our experiences active

participation and communication with implementer has been one of the key factors in

regulatory work to enable effective progress in disposal development and licensing.

2. Regulatory activities during the step-wise implementation

The licensing procedure for a disposal facility has several steps that are similar to all nuclear

facilities in Finland and are defined in Nuclear Energy Act and Degree. These licensing steps

are Decision-in-Principle (DiP), Construction License and Operating License.

The first licensing step towards a disposal facility for spent nuclear fuel was Decision-in-

Principle (DiP). As part of this decision Government decided that SNF would be disposed in

mailto:[email protected]


5

Olkiluoto using KBS-3 concept. In addition to the permit to proceed with the project, DiP

gave also Posiva the authorization to start to construct an underground rock characterization

facility (URCF) at the proposed site, to the depth of actual planned disposal, as required by

safety regulation.

After the DiP, STUK started work aiming for the readiness to review the construction license.

One of the major activities was the regulatory oversight of the construction work of the

underground rock characterization facility (URCF), Onkalo. STUK planned and executed the

regulatory oversight of the URCF in similar manner as it would do for nuclear facility due to

the fact that Posiva’s plan is to use the constructed URCF as part of the disposal facility in the

future.

Besides the oversight of the construction work of the URCF, STUK followed closely

Posiva’s R&D work based on the R&D program published every third year and reviewed the

draft post closure safety case documentation published by Posiva before year 2012. The aim

of the step-wise review, close follow-up and regular meetings with Posiva was to identify the

safety relevant issues and especially key safety concerns already before Posiva finalizes and

submits the construction license application. The review of safety case parts was not always

usefull in solving safety relevant issues and from this experience a need for more structured

review and assessment process for the construction license application review was seen

necessary.

In addition to the activities related to Posiva, STUK also developed it’s own resources and

competence to prepare itself for the construction license review. In 2006 STUK management

made a strategic plan to increase waste management resources before Posiva submits the

CLA. Plan was followed and the amount of people working mainly for the waste

management regulations was almost tripled during next six years. STUK made also

framework contracts with 13 external experts to support STUK during in the review of CLA.

STUK’s task in the CLA process is to review and assess the fulfillment of all applicable

radiation and nuclear safety requirements and prepare statement and safety evaluation report

for the Government. STUK submitted statement regarding Posiva’s CLA to the Govenrment

in February 2015. STUK’s main conclusion was that the planned encapsulation plant and

disposal facility can be built to be safe. Also there is sufficient reliability that there will be no

detrimental radiation effects to the public or environment neither during the operational

period nor after decommissioning and closure of the facility. In the statement to the

government STUK raised areas that need further development before specific construction

step or before submittal of operating license application.

3. Conclusions

Based on the experiences of regulating Posiva’s DGR development, we have concluded that

following aspects are important for effective regulatory work:

Development and maintaining of up-to-date requirements. Requirements can be develop

along with DGR development as more information and knowledge are gathered


6

Development of oversight strategy for each different phase. Starting for early

conceptualization and siting phase regulatory functions and focus can differentiate a lot. In

early phase regulators review can be more generic and evaluating that safety requirements

could be met. In licensing steps however regulator has to conclude if safety requirements are

met or not. This is the most challenging part of review and assessment and therefore clear

criteria should be developed.

Active interaction with implementer is needed for mutual understanding.

Regulator has a important role in communicating with public and this involvement should

start in early phase of repository development


7

03d – 01 / ID 111. Disposal of High Level Waste

COMPLEX SAFETY ASSESSMENT MODEL OF RADIOACTIVE WASTE DEEP

GEOLOGICAL DISPOSAL IN THE CZECH REPUBLIC

V. Havlová1, D. Trpkošová

1, A. Vokál

2

1ÚJV Řež, a.s., Husinec, Czech Republic

2SURAO, Dlážděná 6, Praha, Czech Republic


Abstract. A complex safety assessment (SA) model employing GoldSim software has been under

development in the Czech Republic since 2006 aimed at proving the long-term safety of the country’s future

deep geological repository (DGR) over a period of 1 million years. The input data for each of the components of

the model has been compiled from archive sources, expert literature and supporting research. The main concern

with respect to the model is to adhere as closely as possible to conditions which will prevail within the real DGR

by means of performing either laboratory or in-situ research. The paper includes a description of the model and

examples of supporting research concerning both the near- and far-fields.

Key Words: safety assessment, deep geological repository, GoldSim, radiological impact

1. Introduction

The main aim of deep geological repository (DGR) safety assessments (SA) is to consider the

performance of the repository system in terms of radiological impact or other global

indicators of the impact on safety. Such assessments may vary in terms of the relevant time

frame(s), the level of detail, the range of issues considered, the degree of precision required

with respect to the input data and the resulting calculations. The reason for the safety case as

well as the programme development stage often dictate both the scope of and degree of detail

required in the safety assessment [1].

Consequently, a complex SA model employing GoldSim software has been under

development in the Czech Republic since 2006 the purpose of which is to illustrate the long-

term safety of the future Czech deep geological repository (DGR) over a period of 1 million

years. The input data required for SA modelling purposes, consisting of results obtained from

both archive sources and limited own research, was collected in 1999 and 2011. Currently,

with respect to the performance of SA supporting research, the main concern is to adhere as

closely as possible to conditions which will prevail within the real DGR. Therefore, it is

essential that the research includes both laboratory and in-situ experimentation.

2. Czech disposal concept

The Czech deep geological repository (DGR) concept assumes that waste packages

containing spent nuclear fuel (SNF) assemblies will be enclosed in steel-based canisters

placed in vertical or horizontal boreholes at a depth of ~ 500m below the surface. The void

between the canisters and the host crystalline rock will be backfilled with compacted

bentonite which will make up the final engineered barrier. The reference SNF canister

consists of two protective layers, an outer layer of carbon steel which will corrode very

slowly under anaerobic conditions and a second inner layer of stainless steel which will

corrode at an almost negligible general corrosion rate and exhibit a low tendency to local

corrosion under anaerobic conditions. It is presumed that the buffer material will originate



8

from Czech Republic bentonite deposits; currently, so-called Rokle bentonite (Ca-Mg

bentonite) is being employed for experimentation purposes.

In addition to SNF and high-level waste, intermediate-level waste (ILW) containing long-

lived radionuclides such as decommissioned reactor core parts and serpentinite concrete

which does not meet the criteria for disposal in near-surface repositories will also be disposed

of in the future DGR. However, ILW will be disposed of in a separate section of the

repository to that of SNF assemblies since it is essential that the potential for the SNF and

ILW to exert an impact on each other be avoided. ILW will be emplaced in concrete canisters

in specially excavated chambers that will then be back-filled with a bentonite-based material.

3. Near-field

The near-field SA model assumes the disposal of a total of 5800 carbon-steel canisters

containing spent fuel (SF) with a minimum canister life-time of 10,000 years and a median

canister life-time of 110,000 years. It is assumed that the release of radionuclides will occur

following the degradation of the canisters. SF canister degradation is simulated by means of a

distribution curve obtained via the application of the Weibull distribution. The original

version of the model assumed a uniform inventory, however the latest version enables the

inventory to be divided according to a number of defined preferential transport directions,

each characterised by an individual transport pathway towards the surface [2].

The data which characterises carbon-steel canister material corrosion rates was obtained from

the results of previous projects involving a limited number of laboratory experiments [2].

Current research projects however include both laboratory and real rock massif scale

experimentation. Carbon steel, titanium and copper corrosion on contact with Rokle bentonite

has been extensively investigated at the UJV’s laboratories under defined anaerobic

conditions as part of a previous project [3] - see Fig. 1. Further research in this respect was

conducted in the context of the international Material Corrosion Test (MACOTE) project

performed at the Grimsel test site [5], as part of which in 2015 five heated modules (of UJV

construction) containing corrosion samples (steel, copper; Czech Ca-Mg bentonite and MX-

80 bentonite) were inserted into the rock massif up to a depth of 5 meters (anaerobic

conditions). The modules will be extracted over a defined time-line of 1, 2, 3, 5 and 7 years.

The results of both projects will subsequently be combined.

Fig. 1. Carbon-steel sample in contact with Ca-

Mg bentonite [3]. Fig. 2. Bentonite layer representation in

the GoldSim model.

It is currently supposed that Rokle bentonite (Ca-Mg bentonite) will be used as the buffer

material surrounding the disposal canister. For modelling purposes, the rock diffusion layer is

considered to be the bentonite/rock compartment interface thus eliminating the influence of

advection within the bentonite layer. The bentonite buffer layer is modelled in the form of 15

concentric layers (see Fig. 2), the outer layer of which represents the interface with the rock

Bentonite_Cell1_440 Bentonite_Cell2_440 Bentonite_Cell3_440 Bentonite_Cell4_440 Bentonite_Cell5_440




9

massif enclosing the repository (the rock compartment). Radionuclides are transported by

means of diffusion through the bentonite layers towards the rock compartment. Subsequently,

the radionuclides are transported by means of groundwater flow from the near-field boundary

towards a preferential path within the geosphere.

Radionuclide diffusion data for safety assessment purposes is usually obtained via the

conducting of laboratory through-diffusion experiments using radioactive tracers. Through-

diffusion experiments are based on the diffusive transport of tracers through the bentonite in

the direction of the concentration gradient.

FIG. 3. GoldSim geosphere model [2]

FIG. 4. Fracture model in PAMIRE

project -preliminary results [7]

In the case of bentonite, a process description is important particularly with respect to anionic

radionuclides (e.g. I, Se, Tc) where relative retardation is anticipated due to anionic

exclusion. The phenomenon has been described as part of the research outlined in [2], [3] etc.

4. Far-field

The rock massif is modelled in the form of a compartment with dimensions of 3km 1km

10m while the geosphere is modelled using “Pipe” components which consider advection,

dispersion, diffusion into the rock matrix and sorption as the principal processes under way.

Groundwater flows into the compartment that models various processes at work in the

biosphere from the final “Pipe” (see Fig. 3).

Radionuclide migration processes have been studied under both laboratory (e.g. [2], [3], [4])

and in-situ conditions (e.g. [7], [8]). Whilst laboratory results are able to provide results for

well-defined conditions, they are not able to fully reflect the conditions of the rock massif.

Supporting in-situ research has been conducted at for example the Josef Underground

Research Laboratory (CZ) (e.g. the PAMIRE project [7]) and at the Grimsel test site (Long

Term Diffusion, LTD project [8], [9]). The PAMIRE project described a rock fracture in

detail in preparation for the conducting of a migration advective test with radionuclides in

granitic rock (see Fig. 4), whereas the Long Term Diffusion experiment project Phase III

focused principally on the matrix diffusion process involving the injection of a radioactive

“cocktail” consisting of 3H,

22Na,

133Ba,

134Cs and non-active Se(VI) into a granitic rock

massif in 2014 and the subsequent observation of tracer diffusion [9].

5. Biosphere

The biosphere is modelled using four compartments representing land (cultivatable and

forest), a lake and a river. The model represents a universal system that corresponds to the

current lifestyle of the Czech Republic. The output of the biosphere model consists of the

effective dose rate received by one member of the critical group in the environment.

geosphere_deep_pathway

geosphere_shallow_pathway

geosphere_midle_pathway

depository_closed_area


10

6. Conclusions

The SA model was not designed as a “static” model, rather the aim is to continue the

development of the model so as to eventually describe the site finally chosen for the Czech

DGR. The following aspects should be considered in the near term: the source term, the

refinement of the geosphere transport model, the construction of individual biosphere models

for each DGR candidate site, uncertainty evaluations etc. Work to date will be concluded

with an SA evaluation due to be completed in 2018 which will address in greater detail one of

the potential sites for the construction of the Czech deep geological repository.

7. Acknowledgement

The research presented in this study was funded by SURAO [2, 3, 5, 8], the Ministry of Trade

and Industry [6] and the Czech Technology Agency (TAČR) [7].

REFERENCES

[1] NEA-OECD (2012): Methods for the Safety Assessment of Geological Disposal Facilities

for Radioactive Waste (MeSA). NEA No. 6923, OECD, 2012.

[2] Scientific support of DGR safety assessment. SURAO project (2014 - 2016).

[3] Research and development of a disposal canister for SNF deep geological disposal.

SÚRAO project (2013 - 2017).

[4] Research of material properties for the safe disposal of radioactive wastes and the

development of procedures for their evaluation. MPO TIP FR TI-1/362 project.

[5] http://www.grimsel.com/gts-phase-vi/macote-the-material-corrosion-test/macote-

introduction

[6] Research on the influence of inter-grain granite permeability for safe disposal in a

geological formation; methodology and measurement device development; MPO TIP FR

TI-1/367

[7] PAMIRE - http://www.ujv.cz/cz/pamire. TA04020986 .

[8] Long-term diffusion Phase VI. project - http://www.grimsel.com/gts-phase-vi/ltd/ltd-

introduction

[9] Soler J. and Martin A. (2015): LTD Experiment Monopole 2: Predictive Modeling and

Comparison with Initial Monitoring Data. NAGRA Report 15-33. NAGRA, Wettingen,

Switzerland.

http://www.grimsel.com/gts-phase-vi/macote-the-material-corrosion-test/macote-introduction

http://www.grimsel.com/gts-phase-vi/macote-the-material-corrosion-test/macote-introduction

http://www.ujv.cz/cz/pamire.%20TA04020986


11


PREPARING POSIVA’S POST-CLOSURE SAFETY CASE TOWARDS THE

OPERATIONAL PHASE

A. Hagros1, H. Reijonen

1, B. Pastina

2, N. Marcos

1, P. Hellä

1

1Saanio & Riekkola Oy, Helsinki, Finland

2Posiva Oy, Eurajoki, Finland


Abstract. Posiva Oy is currently preparing a safety case to support the operating licence application (OLA)

for the spent nuclear fuel disposal facility under construction at the Olkiluoto site in south-western Finland. The

methodology to prepare the safety case documentation will consider the latest updates in the regulations; lessons

learned from Posiva’s previous safety case, TURVA-2012, submitted in the context of the construction licence

application (CLA) in 2012; the feedback received from the Radiation and Nuclear Safety Authority (STUK) on

the CLA, including several specific requirements for the next safety case; and new challenges related to the

implementation of repository construction and operation. This calls for a higher level of maturity in both the

safety case itself and in the design on which the safety case is based. Since the safety case work will inevitably

take several years, it is necessary to introduce requirements, design and data freezes at the beginning of the

safety case production process. The design freeze is based on the information and requirements available at the

start of the safety case work, but updates can be expected as the design matures and is optimized for

industrialization and operation. A change management process is set up to facilitate the assessment of the

impacts of the proposed changes on the safety case results. The input data used in the safety assessment and

their possible updates will be managed by using of a central database. The uncertainties in the initial state of the

components of the disposal facility will be tackled by implementing an analysis of potential deviations in these

components at the time of installation. Deviations are then screened and implemented in scenario formulation.

Defining a range of initial state parameter values and deviations for the installed components introduces some

flexibility in design and increases the robustness of the safety case.

Key Words: Spent nuclear fuel repository, long-term safety, safety case, Olkiluoto

1. Introduction

Posiva Oy is responsible for the disposal of spent nuclear fuel from the Finnish nuclear power

plants of Loviisa and Olkiluoto. In November 2015, the Finnish Government granted a

construction licence for Posiva’s disposal facility at Olkiluoto, in south-western Finland. The

construction licence application was supported by a safety case, TURVA-2012 [1], which

was evaluated by the Radiation and Nuclear Safety Authority (STUK). STUK concluded that

Posiva had developed a safety concept that is in line with the regulatory requirements [2] and

that the post-closure safety of the disposal facility has been analyzed in a sufficient manner

for the purposes of the construction licence stage [3]. At the moment, Posiva is in the process

of preparing a safety case to support the operating licence application (OLA) for the disposal

facility. Before the application can be submitted, Posiva will have to fulfil 34 requirements

formulated by STUK for the new safety case and the related research and modelling work [3].

The new safety case will also need to consider any updates in the regulations, as well as new

challenges related to the implementation of repository construction and operation.

2. Overall Safety Case Methodology

A safety case is the synthesis of evidence, analyses and arguments that quantify and

substantiate the claim that the repository will be safe after closure and beyond the time when



12

active control of the facility can no longer be assumed [4]. A safety case includes a

quantitative and a qualitative assessment of the long-term performance of the disposal

system. The quantitative assessment (a.k.a. safety assessment) is defined as the process of

systematically analyzing the ability of the disposal facility to provide the safety functions and

to meet the requirements and of evaluating the potential radiological hazards and compliance

with the safety requirements. The qualitative assessment broadens the scope of the safety

assessment to include the compilation of a wide range of evidence and arguments that

complement and support the reliability of the results of the quantitative analyses [5].

The general safety case structure builds upon the one used in TURVA-2012 [1], i.e. the safety

case will consist of a portfolio of main reports and a number of supporting reports.

2.1.Handling uncertainties in the initial state

Design development work is moving towards implementation stage and, accordingly, Posiva

has planned the disposal operation at a very detailed level, both in order to plan and optimize

the disposal operation, but also for production and large-scale implementation tests. The

experience obtained to date is used in the safety case to better constrain the uncertainties

related to the initial state of the repository system. Initial state refers to the description of the

state of various repository components after emplacement has been completed, i.e.

information which acts as a starting point for the performance and safety assessments.

The uncertainties are handled through a systematic screening of the possible deviations

through a modified failure mode and effect analysis (FMEA [6]), and further handling in the

scenario formulation work incorporating the deviations into the safety case. The FMEA for

the initial state has been modified to screen events that can lead to failure modes that are

likely to be undetected and thus remain in the repository at the time of the initial state. The

aim is to improve the description of the initial state of the repository system from the

traditional design freeze description [7] towards a description of the repository in ‘as-built’

state.

2.2.Handling uncertainties during the long-term evolution

Uncertainties during the long-term evolution of the disposal system are handled through a

systematic analysis of how the different FEPs might act on the components of the disposal

system during its evolution, followed by the formulation of scenarios and analysis of cases

giving rise to potential failures of containment and radionuclide releases and their

corresponding radiological impacts.

3. Methodology to Handle Changes

3.1.Requirements, Design and Data freezes

Since the safety case work will inevitably take several years, it is necessary to introduce

requirements, design and data freezes at the beginning of the safety case production process.

The requirements freeze allows the design to be fixed for specific purposes, such as the safety

case or large-scale tests. The design freeze is based on the information available at the time of

requirements freeze. The data freeze refers to data other than actual design data and includes,

for example, geological site data, surface environment data or time-dependent data needed in

the modelling chain, where the output of certain models will serve as input to other models.

The data freeze does not need to happen at the same time as the design and requirements

freeze, only at the time it is needed as input in the modelling chain. Once input to the safety


13

case has been approved and frozen, its documentation and change management process (see

below) is of utmost importance to ensure traceability and reliability of the results in the safety

case. The input data will be stored in an electronic central database in a traceable manner, so

that both the approved data, approval process and future potential updates are clearly

recorded.

3.2.Design Development During the Safety Case Process

Requirements, design and data freezes were already used in TURVA-2012 (see, e.g., [7, 8]).

One of the lessons learnt was that it is not possible to freeze the design completely before the

start of the safety case work, because important developments can happen during the safety

case process, which lasts several years while the design develops and operational experience

is being obtained. This is expected to be emphasized in the operating licence application

process as the design reaches full maturity and is optimized for industrialization and

operation. The long-term performance of the design solution as well as further operational

aspects, particularly related to the installation of engineered barriers in repository-like

conditions will also be studied in large-scale demonstration tests. In their Review Report [2],

STUK has concluded that although there are no direct requirements for demonstrations in any

of the regulations, the Guide YVL B.1 [9] states that the solutions and methods chosen during

the course of the design shall be based on proven technology and operating experience. In

addition, the design shall strive for simplicity and, if new solutions are proposed, they shall

be validated through tests and experiments [2]. Posiva’s plans for large-scale demonstrations

are described in the waste management programme YJH-2015 [10].

3.3.Change Management Process

As changes to the design and to other input data may be expected to take place during the

safety case work, a change management process needs to be set up to manage the traceability

and reliability of the safety case and to facilitate the assessment of impact of changes in

design on the safety case results. For this purpose, the whole modelling chain used in the

safety case is documented and linked to the approved input data.

Configuration management defines the general process to be followed in order to implement

a change in the design or requirements for the disposal facility. The heart of the configuration

management process consists of classifying each proposed change according to its impact on

operations and safety (including long-term safety). Posiva is currently developing the

methodology to assess the long-term safety impact of proposed changes within the

configuration management process. The criteria to be followed will address the impact of a

given design, requirement or process change on the initial state, on the fulfillment of the

long-term safety functions, or the overall uncertainty management.

Change management is based on expert judgment and relies on a close co-operation between

long-term safety and design from the beginning of the safety case work. The main interfaces

between these two groups of experts are the long-term safety requirements and their

verification as well as the initial state.

4. Conclusions

Requirements, design and stepwise data freezes need to be performed in a safety case that is

developed in parallel with design optimization and operational readiness activities. A safety

case supporting the operating licence application needs a higher level of design maturity than

that supporting the construction licence application. In Posiva’s case, the design is currently


14

being optimized for industrialization and operation and large-scale demonstrations are also

taking place, the handling of changes arising from these activities is a major challenge in the

safety case process.

As some uncertainties in the initial state of the repository components can be assumed to

remain, an analysis of potential deviations in these components at the time of installation is

proposed to be implemented. The uncertainties in the initial state can then be taken into

account in the formulation of scenarios.

A change management process needs to be set up to incorporate changes in a controlled way,

so that their long-term safety impacts are properly assessed. The proposed changes need to be

considered holistically, including the impact not only on long-term safety but also on the

safety case production process. The proposed changes will only be accepted if they do not

compromise long-term safety and if the safety case analyses can be updated using the new

input. Considering the long operational phase (over 100 years) of the disposal facility, further

optimization activities are expected to occur as the operational experience and knowledge

bases develops; a change management process is thus needed also after the operations have

started.

5. References

[1] POSIVA OY, Safety Case for the Disposal of Spent Nuclear Fuel at Olkiluoto –

Synthesis 2012, POSIVA 2012-12, Eurajoki (2012).

[2] RADIATION AND NUCLEAR SAFETY AUTHORITY, STUK’s Review on the

Construction License Stage Post Closure Safety Case of the Spent Nuclear Fuel Disposal

in Olkiluoto, STUK-B 197, Helsinki (2015).

https://www.julkari.fi/bitstream/handle/10024/127160/stuk-b197.pdf

[3] RADIATION AND NUCLEAR SAFETY AUTHORITY, Safety Case for the Disposal of

Spent Nuclear Fuel in Olkiluoto: Decision, Presentation Memorandum, 1/H42252/2015,

Helsinki (2015).

[4] NUCLEAR ENERGY AGENCY, Post-Closure Safety Case for Geological Repositories:

Nature and Purpose, Report 3679, Paris (2004).

[5] POSIVA OY, Safety case plan 2008, POSIVA 2008-05, Eurajoki (2008).

[6] STAMATIS, D.H., Failure Mode and Effect Analysis: FMEA from Theory to Execution.

ASQ Quality Press (2003).

[7] POSIVA OY, Safety Case for the Disposal of Spent Nuclear Fuel at Olkiluoto –

Description of the Disposal System 2012, POSIVA 2012-05, Eurajoki (2012).

[8] POSIVA OY, Safety Case for the Disposal of Spent Nuclear Fuel at Olkiluoto – Design

Basis 2012, POSIVA 2012-03, Eurajoki (2012).

[9] RADIATION AND NUCLEAR SAFETY AUTHORITY, Safety Design of a Nuclear

Power Plant, Guide YVL B.1, Helsinki (2014).

http://plus.edilex.fi/stuklex/en/lainsaadanto/saannosto/YVLB-1

[10] POSIVA OY, YJH-2015 Nuclear waste management at Olkiluoto and Loviisa power

plants: Review of current status and future plans for 2016–2018 (in Finnish), YJH-2015,

Eurajoki (2015).

https://www.julkari.fi/bitstream/handle/10024/127160/stuk-b197.pdf

http://plus.edilex.fi/stuklex/en/lainsaadanto/saannosto/YVLB-1


15


ANDRA’S SAFETY OPTIONS OF FRENCH UNDERGROUND FACILITY CIGÉO –

A MILESTONE TOWARDS THE LICENSING APPLICATION

S. Voinis, M. Rabardy, L. Griffault

Andra, French National Radioactive Waste Management Agency, Parc de la Croix Blanche,

92298 Châtenay-Malabry, France


Abstract. Following the publishing of the Dossier 2005 Argile, the 28th June 2006 Act entitled “Programme

National de Gestion des Matières et Déchets Radioactifs” (National program for radioactive waste and nuclear

material management) [5] has set the deep geological repository in clay host rock as the selected solution for IL-

LL and HL radioactive waste disposal in France. According to this 2006 Act, reversible waste disposal in a deep

geological formation and corresponding studies and investigations shall be conducted with a view to selecting a

suitable site and to designing a repository. Since 2011, the project has entered an industrial design development

phase and has become the Industrial Center for Geological Disposal “Cigéo”.

In view of the licensing application, two main milestones for safety are identified for Cigéo: a Safety Options

Dossier “DOS” early 2016 and the safety case to support Future License Application of Cigéo “RPs” in 2018.

According to the 2007 French Act, the Safety Options is an opportunity for the operator to send in advance a

first safety case to the French Safety Authority in order to stabilize the safety strategy, the safety requirements,

the safety methods, key safety and design options, the list of safety scenario that are selected and a preliminary

impact of a few margin scenarii. The Safety options don’t present the overall safety demonstration that needs to

be presented in the safety case supporting the licensing application. The Cigeo geological disposal facility

project is designed to cater for all the HLW and ILWLL that has been produced and will be produced by

existing nuclear facilities.

Andra has conducted in the frame of the safety options a parallel and coordinated operation and post-closure

safety analysis. Those safety options take into account the particularity of Cigéo: HLW and various types of

ILWLL waste; the step by step development of Cigeo and the balancing approach between safety, technology

and economics. Considering the various families and nature of the ILWLL waste, the Safety Options consist in

establishing “dimensioning characteristics “for design and operational safety as well as “envelop inventory” to

manage the knowledge acquired at this stage. In addition, the classification of scenario and the safety approach

are adapted to the operational and post-closure context. The safety options identify the links between the two

phases.

Key Words: nuclear safety, disposal, waste management, safety options

1. Introduction

The purpose of the Cigeo geological disposal facility for HLW and ILW-LL is to allow the

safe disposal of IL-HL LL radioactive waste in order to eliminate or reduce the burden to be

borne by future generations, in accordance with Article L542-1 of the Environment Code.

Since 1991, successive safety milestones were implemented, based on the acquisition of

scientific and technical knowledge and the development of safety methods appropriate to

deep geological disposal. Since 2011, Cigéo has entered an industrial design development

phase. In view of the licensing application, as a key milestones for safety, the Andra Safety

Options Dossier “DOS” early 2016 is submitted to a national review and an international



16

review1. According to the 2007 French Act, the Safety Options is an opportunity for Andra to

send in advance a first safety case and aims to stabilize the safety strategy, the waste

inventory, safety requirements, the safety methods, key safety and design options, the list of

safety scenarios and a preliminary impact of a few margin scenarios. The safety options apply

to the disposal of high-level waste (HLW) and intermediate-level long-lived waste (ILW-LL).

The Safety options don’t present the overall safety demonstration that will be presented

in the safety case to support Future License Application of Cigéo in 2018 according to

the recent French Act of July 2016.

FIG. 1. a step-by-step iterative process as regards safety since 1991

2. Safety options and the incremental development of Cigéo

The safety options consider the duration of operation for over a hundred years with

successive phases (construction/operation); it has to be flexible enough to adapt to possible

changes in France's energy policy. There are three main phases in the life of Cigeo: (i) an

initial design and initial construction phase, (ii) an operation phase (including an industrial

pilot phase) and (iii) a post-closure phase. Cigeo will be closed in stages and the post-closure

phase will begin when the final closure of Cigeo has been authorised by a law.

FIG 2. Diagram showing the main phases in the life of Cigeo

Following final closure, the underground facility after its final closure, will be the facility as

built. At the stage of the safety options, the underground layout is only an illustration of what

Cigéo might be, based on the technical options chosen at this stage. According to the

1 International review by expert from regulatory and IAEA on behalf of the French Safety Authority


17

incremental development of Cigéo, if a new technological solution is suggested, it will be

checked that the operational and post-closure safety functions are still fulfilled (safety

indicator assessment) and the radiological impact remains as low as reasonably possible

given economic and social factors.

3. The disposal system (natural and engineered components) relies on both operational

and post-closure safety principles and safety functions

Protecting people and the environment is primarily based on the performance of safety

functions during operation comparable to those performed at all nuclear facilities, and on

safety standards (national and international), safety requirements and safety options adapted

to the specific underground environment of the Cigéo facility.

Andra has implemented, from the design stage (since the 90’s), a safety approach and process

(including siting), which relies on the specific characteristics of a repository as such: (i) the

choice of the Callovo-Oxfordian formation in which the underground facility of Cigeo is

located that meets the site technical criteria of the 2008 ASN nuclear safety guide, (ii) an

underground facility located at a depth of around 500 m, of reduced geometry and long

connecting drifts, requiring specific operating, intervention and evacuation conditions; (iii) a

disposal facility being developed in successive phases, implying a need to consider the risks

related to performing underground construction work and nuclear operations in parallel; (iv) a

coordinated approach encompassing operating safety and post-closure safety. The approach

integrates the successive iterations of Cigéo milestones including design and scientific

knowledge evolutions with the objective of ensuring post-closure safety throughout the entire

development cycle of the Cigeo project.

An appropriate level of monitoring and control will be also applied to Cigeo from its

construction and during its operation, to ensure the protection and preservation of the passive

post-closure safety features, as necessary, so that they can fulfil their safety functions once

the repository is closed.

During operational phase, five safety functions are applicable to Cigeo throughout the

operating phase and must be maintained in all incident or accident situations of internal or

external origin or, at least, restored within time limits consistent with the objectives of

protecting people and the environment defined for the Cigeo project. They are: (i) contain

radioactive substances to protect against the risk of their dispersion; (ii) protect people from

exposure to ionising radiation; (iii) manage safety with regard to the criticality risk; (iv)

remove the heat produced by waste and (v) remove gases formed by radiolysis in order to

manage explosion risks.

For the post-closure phase, the Cigéo aims to isolate the waste from humans and the

biosphere and to confine it within a deep geological formation to prevent dissemination of the

radionuclides contained in this waste (see table 1). The post-closure disposal system relies

particularly on the Callovo-Oxfordian that plays the main role, and the closure structures of

the surface-to-bottom connections (sealed shafts and ramps). The global approach to post-

closure safety assessment is based on practical expression of the safety functions and

associated requirements, analysis of component performance and analysis of the uncertainties

related to the scientific and technological knowledge underpinning the design. To fulfil the

post-closure safety functions, design principles for Cigeo and for the choice of site (see

examples in table 1) are established.


18

TABLE I: Example of Safety Principles for Cigéo

Post-closure safety functions General principles in terms of choice of site and design

Isolating waste from surface phenomena

and human actions Location of Cigeo at a depth and in an area of low, uniform

geodynamic

Preventing the circulation of water Low water flows in the Callovo-Oxfordian due to its low

permeability and the low hydraulic head gradient;

Consolidation and sealing of the surface-to-bottom connections

Restricting the release of radionuclides

and toxic elements and immobilising

them in the repository

Cells (particularly the materials used for them) designed to protect

the waste and packages from a physicochemical point of view

Delaying and reducing radionuclide

migration

Thickness of Callovo-Oxfordian (at least 130 m), high retention

capacity…

Optimised geometries of the cells and drifts in the underground

installation, particularly in terms of length.

Whether the disposal system is functioning correctly and more specifically whether the safety

functions are being fulfilled (operation normal functioning and post-closure normal

evolution,) the design options relies also on the results of the risks analysis during operational

phase adapted to Cigéo context (mainly transfer of waste package, co-activity of works and

operation..) and the subsequent scenario (e.g. dimensioning waste characteristics and

scenarios). It also relies on the scientific and technological uncertainties analysis after closure

and the resulting normal evolution, altered evolution and what-if scenarios assessment. In the

case of Cigéo, the safety options present a series of scenarios considering the dysfunction of

sealing, the dysfunction of vitrified waste canister, as well the occurrence of inadvertent

human intrusions (mostly borehole for Cigéo). Quantitative evaluations aimed at considering

“envelop” situations of those scenarios.

FIG 3. Diagram showing the coordinated approach to operating safety and post-closure safety


19

REFERENCES

[1] Act 91-1381 of 30 December 1991 on radioactive waste management research. (1992).

Official Journal of the French Republic Acts and Decrees No. 1, 10 p.

[2] Act 2006-739 of 28 June 2006 on the sustainable management of radioactive material and

waste. (2006). Official Journal of the French Republic. Acts and Decrees No. 93, 9,721 p.

[3] Délibération du conseil d'administration de l'Agence nationale pour la gestion des déchets

radioactifs du 5 mai 2014 relative aux suites à donner au débat public sur le projet

CIGEO. Ministère de l'écologie, du développement durable et de l'énergie (2014). Journal

Officiel. Lois et décrets, n°108, pp.7851-7854.

Safety Options Report – Post-Closure Part (DOS-AF). Andra. (2015). °

CGTEDNTEAMOASR20000150062.

[4] Safety Options Report – Operation Part (DOS-Expl). Andra. (2015). °

CGTEDNTEAMOASR20000150080.

[5] Act No. 2006-686 of 13 June 2006, as amended, on transparency and security in the

nuclear field. Consolidated version dated 12 July 2014. (2006).

[6] Order of 7 February 2012 laying down the general rules on basic nuclear installations

Consolidated version dated 05 July 2013. (2012).

[7] NEA IGSC Scenario Development Workshop, 1-3 June 2015, Issy-les-Moulineaux,

France , to be published, OCDE.

[8] Radioactive Waste Disposal Facilities Safety Reference Levels v2.2. (Wgwd),

W.G.O.W.a.D. Western European Nuclear Regulators Association (WENRA). (2014). 81

p.

[9] Fundamental safety principles. Safety fundamentals. IAEA. (2006). IAEA safety

standards series n°SF-1. 37 p.

[10] Disposal of Radioactive Waste. Specific Safety Requirements. IAEA. (2011). IAEA

Safety Standards Series n°SSR 5. 62 p.

[11] The management system for facilities and activities. Safety Requirements. IAEA.

(2006). IAEA Safety Standards Series n°GS-R-3. 27 p.

[12] Monitoring and Surveillance of Radioactive Waste Disposal Facilities. Specific Safety

Guide. IAEA. (2014). IAEA Safety Standards Series n°SSG-31. 96 p.

[13] The Safety Case and Safety Assessment for the Disposal of Radioactive Waste.

Specific safety guide. IAEA. (2012). IAEA Safety Standards Series n°SSG-23. 140 p.

[14] Geological Disposal Facilities for Radioactive Waste. Specific Safety Guide. IAEA.

(2011). IAEA Safety Standards Series n°SSG 14. 104 p.

[15] The management system for the disposal of radioactive waste. Safety guide. IAEA.

(2008). IAEA safety standards series n°GS-G-3.4. 85 p.


20


BRIDGING NUCLEAR SAFETY, SECURITY AND SAFEGUARDS AT GEOLOGICAL

DISPOSAL OF HIGH LEVEL RADIOACTIVE WASTE AND SPENT NUCLEAR

FUEL

I. Niemeyer, G. Deissmann, D. Bosbach

Forschungszentrum Jülich GmbH, IEK-6: Nuclear Waste Management and Reactor Safety


Abstract. In order to consider geological disposal of high-level radioactive waste and spent nuclear fuel in all

its complexity, related nuclear safety, security and safeguards issues have to be taken into account. By

identifying both synergies in overlapping methods or techniques and differences in the requirements with

respect to safety, security and safeguards, advantage of inherent synergies and conflicting requirements can be

taken at the same time. While there is a general understanding of the potential benefits of the 3S concept, neither

the interfaces and synergies between safety, security and safeguards nor their practical implementation are yet

fully understood. This paper discusses the role and importance of safety, security and safeguards regarding the

geological disposal of high-level radioactive waste and spent fuel.

Key Words: Safety; security; safeguards; 3S

1. Introduction

The use of the terms ‘nuclear safety’, ‘nuclear security’ and ‘nuclear safeguards’ is often not

sharply delimited from each other, though definitions for each of these issues exist.

According to IAEA definitions, ‘nuclear safety’ refers to “[t]he achievement of proper

operating conditions, prevention of accidents or mitigation of accident consequences,

resulting in protection of workers, the public and the environment from undue radiation

hazards” [1], and therefore stands for the safe operation of nuclear installations.

‘Nuclear security’ implies “[t]he prevention and detection of, and response to, theft,

sabotage, unauthorized access, illegal transfer or other malicious acts involving nuclear

material, other radioactive substances or their associated facilities” [1] and is aimed at the

physical protection of nuclear installations.

‘Nuclear safeguards’ are “designed to ensure that special fissionable and other materials,

services, equipment, facilities, and information made available by the Agency or at its request

or under its supervision or control are not used in such a way as to further any military

purpose” [2] or, in short, to ensure the peaceful use of nuclear material.

The interaction or intersections of the three components depend on the context, and the

significance of each of the components may vary for different types of nuclear installations.

In order to consider geological disposal of high-level radioactive waste and spent nuclear fuel

in its full complexity, all related nuclear safety, security and safeguards issues must be taken

into account. While safety can benefit from some provisions regarding safeguards and

physical protection (security), it may also be contravened by others. Some techniques for

monitoring geological repositories, such as environmental sampling, could provide relevant

data for safety, security and safeguards. Other techniques, such as geophysical measurements

for safeguards verification, are to be implemented in a way that does not infringe long-term

safety requirements. Therefore, identifying both synergies in overlapping methods or



21

techniques or with respect to their future development as well as differences in the

requirements with respect to safety, security and safeguards may help to take advantage of

inherent synergies and conflicting requirements at the same time.

The need of integrating the three ‘S’s’, also referred to as the ‘3S concept’, to the extent

possible throughout all the stages of the nuclear installations’ life cycle, was recognized by

the IAEA in 2008 [3,4] and at the same time, the G8 countries declared to support the 3S

concept [5,6] . Since then, a number of papers discussed the benefits of considering a 3S

approach [e.g., 7-9] in designing and operating nuclear facilities, but only a few addressed the

issue of applying 3S to geological disposal of high-level radioactive waste and spent nuclear

fuel [10-12].

While there is a general understanding of the potential benefits of the 3S concept, neither the

interfaces and synergies between safety, security and safeguards nor their practical

implementation are fully understood to date. This also applies to the geological disposal of

high-level radioactive waste and spent nuclear fuel. Numerous legislations, regulations and

other documents have emphasized that safety is the primary requisite in all life cycle stages

of geological repositories. But what is the significance of security and safeguards with respect

to geological disposal?

2. Role and importance of safety, security and safeguards regarding the geological

disposal of high-level radioactive waste and spent nuclear fuel

2.1.Legal and organizational framework

Nuclear safety, security and safeguards legislations are laid down in a series of national and

international agreements, conventions and regulations [13]. With reference to the 3S concept,

the IAEA noted the need for nuclear legislation that reflects the interrelations between safety,

security and safeguards in a comprehensive and synergetic manner [14]. Accordingly, any

new or revised nuclear legislation on geological disposal of high-level radioactive waste and

spent nuclear fuel should also take 3S conflicts and interfaces into account.

Safety and security are mainly based on an appropriate national legal and organizational

framework, including national regulatory oversight of safety and national law enforcement in

case of security threats. Safeguards, however, represents an international legal commitment,

determined by safeguards agreements and additional protocols between States and the IAEA

[15]. States under safeguards verification by the IAEA usually have a national or regional

Safeguards Regulatory Authority (SRA) in place that acts as interface between the State and

the IAEA. Some States, such as Finland and Japan, have established national regulatory

bodies that cover safety, security and safeguards issues of their nuclear installations and

programmes, including geological disposal, in a single organization [10].

2.2.Material concerned

Safe geological disposal requires a stable geological formation to provide for the long term

containment of radionuclides and their isolation from the biosphere. Safety therefore

addresses all types of radionuclides, in particular the long-lived ones (with half-life periods in

the order of up to 107 years), i.e. actinides and long-lived fission and activation products.

Security considers nuclear material and other radioactive material [1], and safeguards are

principally applied to all source (uranium, thorium) or special fissionable material containing

uranium or plutonium [2]. The lowest common denominator of a 3S control of nuclear


22

material in high-level waste and spent nuclear fuel would therefore include uranium,

plutonium and thorium.

2.3.Timelines

The safety case and safety assessment for geological disposal facilities consider the three life

cycle stages, i.e. the pre-operational period, the operational period and the post-closure

period, spanning over periods in the order of thousands of years and potentially longer (i.e. up

to hundreds of thousands of years) [16]. Security measures do address the three life cycle

stages as well, with a focus on the pre-operational and operational periods, although a

generally care and maintenance free post-closure phase is stipulated in the regulations in

various countries. The timeline for safeguards activities is bound by the duration of the

safeguards agreements and, in the end, will be applied as long as the Nuclear Non-

proliferation Treaty (NPT) remains in force.

A 3S assessment should thus be based on the longest timeline of the single ‘S’-components,

while the role and importance of each of the three ‘S’s’ would vary or decrease over time. If a

‘3S case’ was to be prepared instead of the safety (1S) case, the long-term post-closure period

would mainly be assessed from the safety perspective.

2.4.Control measures

Safety, security and safeguards activities include similar or complementary measures for

documenting, measuring and monitoring the inventory of radionuclides, in particular with

regard to uranium, plutonium and thorium. In order to avoid redundancy or duplication of

work and equipment, a material control and accountancy system should include practices and

procedures, as well as techniques for measurement, sealing and surveillance that fulfil the

requirements as to safety, security and safeguards to the extent possible.

2.5.Facility design

The IAEA generally considers safety, security and safeguards as essential elements in all life

cycle stages of nuclear facilities. In this context, the IAEA has issued a guidance document

[17] aimed at informing stakeholders how to design facilities for nuclear waste management

by early consideration of safeguards in the planning stage so that provisions can be better

integrated with other design requirements as to safety and security.

This approach, also referred to as ‘safeguards by design’ (SBD) should be more closely

interlocked with the 3S concept. ‘Safety, security, safeguards by design’ (3SBD), as generally

proposed by [18,19], can help to reduce efforts and costs related to nuclear waste

management and disposal.

3. Findings

Safety, security and safeguards aspects regarding the geological disposal of high-level

radioactive waste and spent fuel should be addressed and managed in a coordinated,

complementary approach. Further R&D will be needed to identify methods and technologies

(a ‘3S toolbox’) that would be best suited for the holistic consideration of safety, security and

safeguards provisions. By early consideration of conflicting requirements as to safety,

security and safeguards, their impacts on all three life cycle stages of geological disposal can

be minimized. The 3SBD toolbox should include methods and technologies for material

accountancy, nuclear measurements, containment and surveillance, environmental


23

monitoring, continuity of knowledge, as well as design implications to the benefit of all

safety, security and safeguards at geological disposal.

REFERENCES

[1] INTERNATIONAL ATOMIC ENERGY AGENCY, IAEA Safety Glossary.

Terminology used in Nuclear Safety and Radiation Protection, Vienna (2007).

[2] INTERNATIONAL ATOMIC ENERGY AGENCY, IAEA Safeguards Glossary, Vienna

(2001).

[3] INTERNATIONAL ATOMIC ENERGY AGENCY, 20/20 Vision for the Future,

Background Report by the Director General for the Commission of Eminent Persons,

Vienna (2008).

[4] INTERNATIONAL ATOMIC ENERGY AGENCY, Reinforcing the Global Nuclear

Order for Peace and Prosperity – Role of the IAEA to 2020 and Beyond, Vienna (2008)

[5] G8 HOKKAIDO TOYAKO SUMMIT LEADERS DECLARATION, “World Economy,”

Paragraphs 28 and 65, Hokkaido Toyako, Japan (2008).

[6] TSUTOMU, A., NAITO, K. “The New Nexus, 3S: Safeguards, Safety, Security, and 3S-

Based Infrastructure Development for the Peaceful Uses of Nuclear Energy”, Journal of

Nuclear Materials Management (JNMM) 34(4) (2012), 6-10.

[7] KIM, H., et al., “3S (Safety, Security, and Safeguards)-by-Design for Engineering-Scale

Pyroprocessing Facility,” Proc. ESARDA Annual Meeting, 35th Annual Meeting, Bruges

(2013).

[8] LEE, N.Y., et al., “3S Culture, Its Meaning and Future Direction,” Proc. INMM 55th

Annual Meeting, Atlanta, GA (2014).

[9] SANDERS, K.E., et al., “Interfaces among Safety, Security, and Safeguards (3S) -

Conflicts and Synergies,” Proc. INMM 56th Annual Meeting, Indian Wells, CA (2015).

[10] VAJORANTA, T., “Finland’s Integrated Approach to Safety, Security, and

Safeguards,” IAEA Technical Meeting on Safety, Security and Safeguards, Vienna

(2012).

[11] MARTIKKA, E., et al., “Safeguards for a Disposal Facility for Spent Nuclear Fuel – a

Challenge for 3S,” Proc. INMM 55th Annual Meeting, Palm Desert, CA (2013).

[12] HADDAL, R., et al., “Geological Repository Safeguards: Options for the Future”,

Proc. IAEA Symposium on International Safeguards: Linking Strategy, Implementation

and People, Vienna (2014).

[13] INTERNATIONAL ATOMIC ENERGY AGENCY, Handbook on Nuclear Law,

STI/PUB/1160, Vienna (2003)

[14] INTERNATIONAL ATOMIC ENERGY AGENCY, Handbook on Nuclear Law:

Implementing Legislation, STI/PUB/1456, Vienna (2010).

[15] CHERF, A., “Legal Framework for Safety, Security and Safeguards”, IAEA

Technical Meeting on Safety, Security and Safeguards, Vienna (2012).

[16] INTERNATIONAL ATOMIC ENERGY AGENCY, The Safety Case and Safety

Assessment for the Disposal of Radioactive Waste, IAEA Safety Standards Series No.

SSG-23, Vienna (2012).


24

[17] INTERNATIONAL ATOMIC ENERGY AGENCY, International Safeguards in the

Design of Facilities for Long Term Spent Fuel Management, IAEA Nuclear Energy

Series No. NF-T-3.1, Vienna (in print).

[18] STEIN, M., MORICHI, M., “Safety, Security, and Safeguards by Design: An

Industrial Approach,” ANS Nuclear Technology 179(1) (2012) 150-155.

[19] NUCLEAR DECOMMISSIONING AUTHORITY, Geological Disposal Safety,

Environmental, Security and Safeguards Principles for the Design Process, NDA

Technical Note no.13472678


25


DEVELOPMENT OF THE NUMO PRE-SELECTION,

SITE-SPECIFIC SAFETY CASE

T. Fujiyama, S. Suzuki, A. Deguchi, H. Umeki

Nuclear Waste Management Organization of Japan (NUMO), Tokyo, Japan


Abstract. NUMO has developed a safety case for co-disposal of HLW and TRU waste to reflect current

boundary conditions in Japan. In particular, this involves addressing public concerns in the wake of the

Fukushima Dai-ichi nuclear power plant accident and a move by the Government to more strongly support

moving forward with siting a geological repository, involving suggesting locations that are considered to be

scientifically more suitable. This paper will provide a brief overview of this Safety Case, with a focus on

advances from the old “H12 Report”, which is considered the first generic safety case in Japan. “The NUMO

pre-selection, site-specific safety case” has been developed to provide a basic structure for subsequent safety

cases that would be applied to any selected site, emphasising practical approaches and methodology, which will

be applicable for the conditions/constraints during an actual siting process.

Key Words: Geological disposal, Safety case, Vitrified waste, TRU wastes.

1. Introduction

The “H12 Report” [1] published in 1999 by the Japan Nuclear Cycle Development Institute

(now the Japan Atomic Energy Agency, “JAEA”) demonstrated the feasibility of safe and

technically reliable geological disposal of high level waste (HLW), based on a generic study.

On the basis of the H12 Report, “the Final Disposal Act” came into force and NUMO was

established as the implementing body in 2000. Intermediate-level waste generated from

reprocessing of spent nuclear fuel and mixed-oxide fuel fabrication (termed “TRU waste” in

Japan) was also included in the inventory for disposal in 2007. NUMO has been developing

key technologies required for the safe implementation of the geological disposal project since

its establishment and initiated the siting process by open solicitation of volunteer

municipalities in 2002. So far, however, no volunteer municipality has appeared and no

candidate host rock type can be specified.

The Fukushima-Daiichi NPP accident in 2011 increased nationwide concerns about the

feasibility and reliability of safe geological disposal in Japan. After re-thinking the

implementation process, “the Basic Policy on the Final Disposal of Specified Radioactive

Waste” was amended in 2015, so that the Government now leads the search for volunteer

sites. This procedure involves nominating more suitable areas from a geo-scientific point of

view to initiate discussions and cooperation with local municipalities, finally leading to

acceptance of a site investigation, which will be carried out by NUMO.

Taking such changes in boundary conditions into account, it is important at this time for

NUMO to integrate required technologies, including the latest R&D output, in order to

confirm the feasibility and safety of geological disposal in Japan. Thus NUMO has developed

the “NUMO pre-selection, site-specific safety case” which, with the site descriptive models

(SDMs) recently developed, provides a more advanced site-specific basis than the generic

safety case in the H12 Report. This has been developed in a manner that will allow it to

provide the basic structure of future safety cases that would be applicable to any site that

might arise from the site selection process.



26

2. Basic Strategy of the NUMO Safety Case

Despite the fact that there has not been a site or specific host rock identified, the NUMO

Safety Case has developed detailed geological and hydrogeological models for potential host

rock environments. Repository design and safety assessment have been thus performed for

these geological models, thereby providing underpinning evidence to demonstrate the

technical feasibility and the safety for the various types of Japanese geological environments.

More background is provided in the companion paper by Suzuki et al. [2]

3. Site Characterisation and Synthesis into SDMs

NUMO needs to prepare reliable investigation and evaluation methodologies and an approach

to synthesise their output in order to form the basis of selecting suitable repository sites. In

suitable setting, the key safety functions (isolation and containment) of the host geological

environment will persist for a long period of time. Advanced methodologies for precluding

the potential impacts of natural disruptive events and processes are shown. Key concepts,

technical knowledge bases, and basic methodology for geosynthesis of relevant information

into representative spatial and temporal models of site evolution are also documented.

The illustrative site descriptive models (SDMs) are developed for subsequent repository

design and safety assessment. Generic repository design and safety assessment were

performed in the H12 Report for two illustrative geological settings, namely crystalline rock

and sedimentary rock. However, geoscientific knowledge has expanded significantly since

then due in particular to multidisciplinary research programmes; for example, JAEA’s

underground research laboratory projects. It is thus very important to update the previous

repository design and safety assessment on the basis of the current best understanding of

Japanese geological environments in the NUMO Safety Case.

Following the categorization of all potential host rock environments, rock types are grouped

by identifying key characteristics/properties relevant to geological disposal. As a result, three

types of potential host rock environments, ‘Igneous rocks’, ‘Neogene sedimentary rocks’, and

‘Pre-Neogene sedimentary rocks’ are examined in the NUMO Safety Case.

FIG. 1. An example of the nested SDMs for Igneous rocks, including underground panel layout

(bottom, left) and EBS of HLW (bottom, right)

Realistic geological and hydrogeological models are developed in a stepwise manner for the

three types of potential host rock environments: at scales of several kilometres (repository

1km

Active fault Active fault

Granite

Highly fractured (weathered) domain

Sedimentary overburden

100~200m100~200m

GW flow

Illustrative geological setting

3 km

3 k

m

Repository scale

Regional scale

50 km ×50 km

Reserved area

Reserved area

Reserved area

Unpreferable area

Short travel time

Panel scale

Ap

pro

x．

80

0m

EBS + Rock

100m

10

0m

Near-field scale

Faults

(Length > 1km)


27

scale), for defining the location and layout of a repository and assessing groundwater flow

through the potential host rock; then at several hundred metres (panel scale) and a hundred

metres (near-field scale), for more precisely describing hydraulic properties. FIG. 1 shows an

example of the geological and hydrogeological modelling for ‘Igneous rocks’ at nested

scales. For geological modelling, key geological structures that control groundwater flow and

have a major influence on solute transport, such as faults, fractures and sedimentary

structures, are represented by a combination of deterministic and stochastic modelling

approaches.

4. Repository Design

Design methodologies should be developed so as to maintain flexibility for the range of

potential geological conditions encountered in Japan. In the NUMO Safety Case, alternative

repository concepts are presented, which are applicable for a wide range of potential

geological conditions. The designed repositories should be technically feasible to construct

and fulfil the safety functions required to isolate and contain radionuclides.

The design requirements and specifications of the engineered barrier system (EBS), disposal

tunnel, panel layout and sealing system (tunnel back-filling and plugs) have been defined.

The methodology is demonstrated by carrying out a full repository design study, tailored to

the SDMs of three types of potential host rock. The engineering feasibility of construction,

operation and closure of the repository is evaluated based on techniques demonstrated in

domestic or overseas underground laboratories and related R&D facilities. The diagram in

FIG. 1 (bottom, left) illustrates an example of an underground layout tailored to the

geological and hydrogeological model for Igneous rocks. The single level emplacement panel

is applicable in this case, avoiding faults with lengths greater than 1 km (the minimum length

that can be identified by surface-based investigation), and avoiding any less preferable areas

where calculated groundwater travel times are relatively short. Optimal operational processes

and material flow logistics, ventilation and water drainage systems for the underground

facility are also considered while determining the layout. Such site-specific design

demonstrations show progress in practicality of design methodologies.

5. Safety assessment

During the siting stages, both pre- and post- closure safety will proceed in an iterative manner

and the resulting output will support decisions made at the end of each stage from the

perspective on safety. The required safety assessment technology for scenario development,

modelling, database development, etc. will be maintained at the state-of-the-art.

For pre-closure safety, it is needed to demonstrate the feasibility of radiological protection for

local residents and workers during repository operation. Learning from the Fukushima-

Daiichi NPP accident, the regulatory guidelines for nuclear facilities have been revised, but

those for geological disposal have not been discussed in detail so far. In developing

methodology for operational safety assessment of geological disposal, relevant guidelines for

other nuclear facilities, such as those for vitrified waste storage, are considered in the NUMO

Safety Case. An the first stage, radiological safety is highlighted, focusing on activities

relevant to HLW handling and transport, based on specific repository designs and defined

procedures of operation.

For post-closure safety, it is needed to develop an assessment approach and methodologies

which can be applied to specific sites and the repository design concepts tailored to them. In

the NUMO Safety Case, procedures and methodologies to assess long-term safety are


28

demonstrated. A risk-informed approach is introduced, based on international guidelines as

well as recent national discussions on safety regulations. Scenarios are developed and

classified with consideration of their probability of occurrence and target doses are defined as

illustrated TABLE I. Referring to the guidelines of international organizations on assessment

timescales, dose calculations are carried out for up to one million years after closure.

TABLE I: SCENARIO CLASSIFICATION AND TARGET DOSE

Scenario classification Definition Target dose

Likely Scenario

The scenario is used to assess the performance of the

geological disposal system based on best

understanding of the probable evolution, as a

reference for the optimisation of protection.

10 μSvy-1

Less-likely scenario

The scenario is used to assess the safety of the

geological disposal system in view of uncertainties

in scientific knowledge supporting likely scenarios.

0.3 mSvy-1

Very unlikely

scenario Possible scenarios with extremely low likelihood. 1-20 mSvy

-1

Human intrusion

scenario

The scenario is used to check whether the geological

disposal system is robust with assumption of the

human intrusion after loss of institutional control.

Residents:1-20 mSvy-1

Intruder:

20-100 mSv per event

A hybrid methodology of scenario development is introduced, which combines a more

conventional, bottom-up, FEP-based approach and a top-down method based on safety

functions, appropriate to a risk-informed assessment. Safety assessment is being conducted

by applying a approach and methodology of realistic radionuclide transport modelling, as

needed to allow comparison of sites and also possible repository concepts that could be

tailored to them. This advanced modelling includes more realistically representing the

geometry of all components of the engineered barriers (essential for distinguishing between

different repository design options) and realistically representing the 3-dimensional geometry

of the geosphere, with particular emphasis on the solute transport characteristics of all

relevant formations (shown in FIG.1, bottom right). The estimated doses of scenarios in

different categories are smaller than the assumed criteria in the NUMO Safety Case. This

provides a basis for more comprehensive demonstration of post-closure safety at this stage.

The outline of assessment pre- and post- closure safety in the NUMO Safety Case is

presented in the companion paper. [2]

6. Conclusions and a look to the future

The safety case developed by NUMO is inherently limited by the lack of an actual site to

focus on, but the SDM-based approach provides critical experience in integrating the

activities of site characterisation and engineering design teams, focused by the fundamental

requirement to robustly assure safety. This will prove invaluable in the next phase when

parallel characterisation of potential sites may occur.

REFERENCES

[1] Japan Nuclear Cycle Development Institute, H12: Project to establish the scientific and

technical basis for HLW disposal in Japan, JNC-TN1410-2000-001～004, (2000).

[2] S. Suzuki, et al., “Assessment of pre- and post-closure safety in the NUMO safety case

for a geological repository”, International Conference on the Safety of Radioactive Waste

Management, IAEA, Vienna (2016) (in press).


29


DEVELOPMENT OF A GENERIC ENVIRONMENTAL SAFETY CASE FOR THE

DISPOSAL OF HIGHER ACTIVITY WASTES IN THE UK

L.E.F. Bailey1, T.W. Hicks

2

1Radioactive Waste Management, Building 587, Curie Avenue, Harwell, Oxford, Didcot

OX11 0RH, UK 2Galson Sciences Limited, 5, Grosvenor House, Melton Road, Oakham, LE15 6AX, UK


Abstract. The UK is committed to implementing geological disposal for the long-term, safe management of

higher activity radioactive wastes [1]. Higher activity waste includes low-level waste not suitable for near-

surface disposal, intermediate-level waste and high-level waste. As yet, no site has been selected for a

geological disposal facility (GDF) in the UK, but it has been agreed that a site will be sought using a consent-

based approach, preceded by a national geological screening process.

Radioactive Waste Management Ltd (RWM) is responsible for the delivery of the GDF. RWM also has a role

to support the ongoing packaging of radioactive wastes and to provide disposability assessments for waste

producers to provide confidence that packaged wastes will be suitable for eventual disposal in the GDF.

To underpin its role, it is essential that RWM can demonstrate its confidence in the safety of a GDF. To this

end, RWM maintains a generic Disposal System Safety Case that addresses all safety issues concerning the

transport of radioactive wastes to a GDF, the construction and operation of a GDF and the long-term,

environmental safety of a GDF after it has been sealed and closed. It is a challenge to present a meaningful

safety case when the location and hence the design of the GDF are not known. This is particularly pertinent for

the long-term, environmental safety case (ESC), which depends significantly upon an understanding of the

geological setting of the GDF and its evolution.

This paper explains how RWM has developed a generic ESC based on an understanding of the environmental

safety functions provided by a multi-barrier disposal system and the features, events and processes (FEPs) that

support them [2]. It explains how an understanding of the basic physics and chemistry underpinning generic

GDF concepts can be used to develop ‘insight models’ to build understanding of the long-term performance of a

GDF, to support a safety narrative. The paper also explains the role of probabilistic total system models in

providing illustrative calculations to support the generic ESC and RWM’s approach to addressing the inevitable

uncertainties associated with the long timescales that need to be considered within an ESC.

Key Words: radioactive, waste, disposal, safety

1. Introduction

The environmental safety of geological disposal is achieved by isolating the wastes in a

facility constructed deep underground and ensuring that the radionuclides and non-

radiological contaminants are contained such that long-term safety is provided by passive

means. To support the development of a GDF concept that achieves environmental safety,

RWM has defined a set of long-term safety requirements that are consistent with regulatory

expectations on environmental safety. In the absence of a GDF site, while geological

screening progresses in the UK, the strategy for GDF design to meet the safety requirements

is focused on the development of illustrative concepts for radioactive waste disposal in three

generic rock types (higher strength rock, lower strength sedimentary rock and evaporite

rock). For each host rock type, illustrative GDF designs have been identified for high-heat-

generating wastes (HHGW) and low-heat-generating wastes (LHGW), based on multi-barrier

concepts that have been developed in the UK or overseas.



30

2. Generic ESC strategy

The generic ESC [Error! Bookmark not defined.] focuses on a narrative that presents

RWM’s understanding of safety in the context of the illustrative disposal concepts and their

barrier system components. RWM has defined a general set of ‘environmental safety

functions’ that could be provided by different barrier system components at different times

after disposal. These environmental safety functions define how the geological environment,

wasteform, container, buffer/backfill, plugs and seals of a GDF combine to isolate and

contain the wastes and limit the transport of contaminants to the surface environment in

groundwater or gas. They also relate to how the stability of the barrier system is maintained,

how disruption of the barrier system through gas-pressurisation is avoided, and how the

potential for post-closure criticality is minimised. Each environmental safety function will be

influenced by various FEPs after disposal. For RWM’s generic ESC, the OECD NEA FEP

database [3] was reviewed to identify FEPs of potential relevance to the different GDF

concepts and the safety functions provided by the barrier systems. For example, FIG. 1

shows the waste package FEPs listed in the OECD NEA international FEP database that

could influence how a wasteform limits the release of contaminants from a waste package.

Having defined how different barriers provide environmental safety functions after disposal,

the overall environmental safety of a GDF needs to be demonstrated for all relevant scenarios

of disposal system evolution. These scenarios are identified through an analysis that

considers the timescales over which each barrier’s environmental safety functions are

expected to be effective and the situations in which barrier performance may be challenged or

compromised by probabilistic events and processes.

FIG 1 Illustration of the waste package FEPs listed in the OECD NEA international FEP database

that could influence how a wasteform limits the release of contaminants from a waste package


31

By considering barrier behaviour for each illustrative disposal concept, a base scenario has

been defined that represents understanding of expected GDF evolution, with reference to the

environmental safety functions to be provided by each barrier component. Variant scenarios

for GDF evolution have been identified based on consideration of probabilistic FEPs that

may or may not occur. By assessing these scenarios, different safety states are analysed in

which wastes are contained in the engineered barrier system or in the geological barrier, or

are returned in residual amounts to the accessible environment at regulated, acceptable levels.

2.6. Evaluation strategy

It is possible to gain understanding of the post-closure performance of a GDF by considering

the basic physics of the disposal system through ‘insight modelling’. Such modelling can be

used to inform the development of appropriate disposal concepts for different generic rock

types and is particularly helpful at the generic stage because it does not require large amounts

of data. For example, the peak radiological risk arising from the migration of radionuclides

via the groundwater pathway in an advection-dominated geological environment can be

estimated using a simple one-dimensional model of radionuclide transport in a porous

medium. Such a model can be used to illustrate in terms of a set of dimensionless parameters

how peak radiological risk is low if there is a long retarded travel time in the geological

barrier relative to the rate of radionuclide decay, significant longitudinal dispersion along the

transport path through the geological barrier, or slow leaching of radionuclides from the

disposal region relative to the rate of radionuclide decay.

Insight modelling complements the more detailed probabilistic total system modelling of the

behaviour of radionuclides and non-radiological species in a disposal system that takes

account of uncertainty in system evolution. These models enable the risks associated with

geological disposal, as well as complementary indicators of safety such as activity fluxes

across barriers, to be evaluated, thus supporting understanding of the different safety states of

a GDF. Natural and archaeological analogue evidence of how barrier materials behave under

expected disposal conditions presents other lines of reasoning that support safety evaluations.

2.7. Assessment timescales

In considering appropriate assessment timescales it is relevant to consider the hazard

presented by the wastes in the GDF and the uncertainties associated with the GDF and its

environment. For the period in which a disposal system is expected to be relatively stable

and uncertainties in the behaviour of radionuclides and non-radiological contaminants can be

quantified more reliably, it is appropriate to undertake probabilistic calculations of GDF

performance. However, for periods in excess of a few hundred thousand years after GDF

closure, the geological environment could be affected by large-scale natural processes, such

as tectonism, subsidence, uplift and erosion, permafrost development and periods of

glaciation. Until specific sites have been identified as potential locations to host a GDF in the

UK, RWM considers that it is not appropriate to undertake a detailed assessment of the

impacts of large-scale natural processes on GDF performance. Thus, the generic ESC

includes probabilistic calculations for an assessment period of 300,000 years after GDF

closure to provide an indication of the barriers that are likely to be of most importance to

GDF performance on the timescales over which some large-scale natural processes could

occur. In this period, the activities of the majority of radionuclides are expected to become

insignificant as a result of radioactive decay whilst the radionuclides are contained within the

disposal system. The behaviour of relatively soluble and mobile radionuclides, such as

chlorine and iodine, and gaseous releases of radionuclides are likely to be of most


32

significance to environmental safety in this period. A range of illustrative calculations of

radionuclide transport and containment have been undertaken to assess base and variant

scenarios over the 300,000-year period for the different illustrative disposal concepts.

3. Summary

The generic ESC explains how the geological disposal of higher activity wastes can be

accomplished in a way that ensures environmental safety in the long term after wastes have

been emplaced and the disposal facility has been closed. Underpinning the ESC are:

A safety concept that is based on ensuring that the long-term safety requirements for

the GDF are met.

A demonstration of how environmental safety can be achieved by implementing

disposal concepts that are based on systems of multiple engineered and natural

barriers that provide multiple safety functions. These barriers are designed to ensure

that the wastes are isolated and contained for the long term after disposal by passive

means.

An understanding of expected barrier performance and how conditions in a disposal

system will evolve, based on research findings presented in RWM’s knowledge base.

An approach to safety assessment based on multiple lines of reasoning, involving both

qualitative and quantitative analysis. Insight modelling and total system modelling

have been used to develop an understanding of how different components of the

engineered and natural barrier system contribute to environmental safety.

At the current time, no site is available for a GDF in the UK and therefore the ESC is

necessarily generic. The high-level generic safety arguments presented in this ESC provide

the understanding that will underpin the future development of a site-specific ESC. In

particular, at each stage of the development and design of the GDF, demonstration of the

post-closure safety of a disposal concept will be founded on an understanding of the

environmental safety functions that will be provided by the specific engineered barriers

defined for a particular combination of host rock and wasteform and the natural barriers

provided by the geological environment.

REFERENCES

[1] DEPARTMENT OF ENERGY & CLIMATE CHANGE, Implementing Geological

Disposal – A Framework for the Long Term Management of Higher Activity Waste,

URN 14D/235, July 2014.

[2] RADIOACTIVE WASTE MANAGEMENT, Generic Environmental Safety Case Main

Report, DSSC/203/01, in publication.

[3] NEA, Updating the NEA International FEP List: An IGSC Technical Note, Technical

Note 2: Proposed Revisions to the NEA International FEP List, NEA/RWM/R(2013)8,

OECD NEA, Paris, September 2012 (published June 2014).


33


SITEX, THE EUROPEAN NETWORK OF TECHNICAL EXPERTISE

ORGANISATIONS FOR GEOLOGICAL DISPOSAL

D. Pellegrini1, F. Bernier

2, V. Detilleux

3, G. Hériard Dubreuil

4, A. Narkuniene

5, J. Miksova

6,

M. Rocher1

1Radiation Protection and Nuclear Safety Institute (IRSN), France

2Federal Agency for Nuclear Control (FANC), Belgium

3Bel V, Belgium

4MUTADIS, France

5Lithuanian Energy Institute (LEI), Lithuania

6Research Centre Rez (CV-REZ), Czech Republic


Abstract. A European SITEX network is being prepared to ensure a sustainable capability for developing and

coordinating joint and harmonized activities related to the independent Expertise Function in the field of deep

geological disposal safety. Two successive EURATOM projects devoted to the preparation of this network

worked on the needed set of activities, which entails strengthening the review of safety cases, developing a

research strategy, interacting with civil society and training. This paper presents the main outlines of the on-

going second project called SITEX-II, with a specific focus on the Strategic Research Agenda issued recently.

Key Words: Expertise Function Network, Geological Disposal, Safety Case Review, Civil Society

Involvement.

1. Introduction

The European Council Directive 2011/70/EURATOM of 19 July 2011 establishes a

Community framework for the responsible and safe management of spent fuel and

radioactive waste. In line with this Directive and in consistency with international high level

safety standards issued by IAEA and WENRA, waste management organisations (WMOs)

are developing a safety case for presenting the technical and organisational arguments that

support the development of the national geological repository in each concerned country.

As safety cases develop, the safety case review by regulatory bodies in the framework of the

decision making process develops as well. In that context, organisations in charge of

reviewing the safety case must in particular evaluate whether the elements of safety, and in

particular that supported by scientific and technological results, are sufficiently convincing to

be accepted by the regulatory authority as a basis for proceeding with the decision making

process.

In that context, there is a need at the international level for developing and coordinating

activities associated with the regulatory review process of deep geological disposal. In 2012,

the EURATOM FP7 SITEX (“Sustainable network for Independent Technical EXpertise of

radioactive waste disposal”) project was launched in order to complement existing initiatives

(ENSREG, WENRA, NEA/RWMC/Regulator Forum…) with the view to characterize the

Expertise Function (see Figure 1) devoted to the technical review of a safety case at national

level for deep geological disposal of radioactive waste.



34

FIG. 1. The Expertise Function and its interactions [1].

The SITEX-II Project (2015-2017), a EURATOM Horizon2020 Coordination and Support

Action, is aimed at practical implementation of the activities defined by the former SITEX

project using the interaction modes developed by that project and with a view to further

prepare the future Expertise Function network. SITEX-II brings together, as partners,

representatives from 18 organisations involving National Regulatory Authorities (NRAs),

Technical Support Organisations (TSOs), Research Entities (REs), Non-Governmental

Organisations (NGOs), specialists in risk governance and an education institute, and involves

interactions with a wider group of Civil Society (CS) participants. Its tasks include

programming R&D, developing a joint review framework, training and tutoring for reviewing

safety cases and interacting with CS, as detailed below, together with preparing an Action

Plan that will set out the content and practical modalities of the future Expertise Function

network.

2. Programming R&D

The 2011/70/EURATOM Directive requires the Expertise Function to carry out its own

horizontal and R&D activities, so that it is not dependent on those developed by the

Implementer Function to make its own judgement. It is also stressed in IAEA safety guides

that the Regulatory Body, and thus its supporting organisations (see Figure 1), may need to

conduct or commission R&D in support of regulatory decisions (see IAEA GS-G-1.1 [2] (see

§3.33) and IAEA GS-G-1.2 [3] (see §3.68)).

SITEX-II therefore includes a task which pursues the general objective of further defining the

Expertise Function’s R&D programme necessary to ensure independent scientific and

technical capabilities for reviewing a safety case for geological disposal. In this perspective, a

Strategic Research Agenda (SRA) [4] has been issued, which will be completed by the Terms

of Reference (ToR) for its implementation. This SRA has been also an input to the JOPRAD

project (EURATOM Horizon2020 Coordination and Support Action “Towards a Joint

Programming Project on Radioactive Waste Disposal”, 2015-2017), which aim is to assess

the feasibility and, if appropriate, to generate a proposal for Joint Programming activities that

could be developed by WMOs, TSOs and/or REs in the field of Radioactive Waste

Management (RWM), in particular geological disposal.

The commitments of SITEX-II members for the development of the Expertise Function SRA

are the following:

– the SRA is developed by applying a transparent methodology;

– the SRA addresses the needs associated with the different states of advancement of

geological disposal programmes;


35

– the concerns of CS participants are duly taken into consideration.

The current version of the SRA entails topics relevant to the Expertise Function to assess

whether geological disposal facilities are developed and will be constructed, operated and

closed in a safe manner, for which a sufficient level of common interest has been expressed

amongst the SITEX-II members. So, seven main topics related to pre and post-closure safety

are considered in the SRA (1. Waste inventory, 2 Transient THMBC conditions in the near-

field, 3. Evolution of EBS material properties, 4. Radionuclide behaviour in disturbed EBS

and HR, 5. Safety relevant operational aspects, 6. Managing uncertainties and the safety

assessment, 7. Lifecycle of a disposal programme and its safety case). In addition to R&D

activities, the needs for knowledge transfer (e.g. training, tutoring), for developing states of

the art and for exchanging on practices and developing common positions are also identified.

One particularly innovative development of the SRA relates to the introduction in the main

topic n°7 of several holistic (complex) topics, for which both technical and societal aspects

need to be investigated in an integrated manner, using specific interdisciplinary

methodologies and involving CS participation. Also, regarding the other main topics, that are

mainly technical, it came out essential to embed CS participation through the involvement of

trained individuals, therefore offering the public the possibility to follow the development of

this technical research, and to perform Knowledge Sharing and Interpretation (KSI) activities

along the development of R&D results

Figure 2 illustrates the associated issues and activities of common interest for Main Topic 1.

SRA Main Topic and associated issues

Research activities (experiment

and/or modeling works)

Horizontal activities

Exchange on practices, develop common positions

Develop states of the art

Transfer of knowledge

(eg. training)

Main Topic 1: Waste inventory and source term #1. Uncertainty about databases and methodologies used for

defining waste inventories(including historical waste)

#2. Evolution of the waste inventory due to possible neutron activation

#3. Understanding of the release processes and speciation of the radionuclides for spent fuel, vitrified and cemented waste

#4. Waste acceptance criteria

FIG. 2. Associated issues and activities for Main Topic 1 of the Expertise Function SRA.

3. Developing a joint review framework

High-level safety requirements and regulatory expectations for the safety case at different

phases of geological disposal facility development (conceptualization, siting, reference

design, construction, operation, post-closure) are addressed by the EU Directive and

international standards and recommendations (IAEA, ICRP, WENRA, etc.). This leads to a

key objective for a second SITEX-II task to further develop a common understanding of the

interpretation and proper implementation of safety requirements in the safety case for the six

phases named above of geological disposal facility development. Position paper on the

selected topics and technical guides related to the review of a safety case will be elaborated,

accounting for existing initiatives and building upon return of experience at the international

level. To date, SITEX-II participants have exchanged their views and experience on how to

implement in practice the requirements and expectations related to “optimisation of

protection” and to “waste acceptance criteria”. The next topics will be “operational issues in

regards with post-closure safety”, with an introductory presentation by a GEOSAF2

representative, and “programme for site characterization”.


36

4. Training and tutoring for reviewing the Safety Case

The third task of the ongoing project aims to implement a practical demonstration of training

services that may be provided by the foreseen SITEX network. The training will be

undertaken within the existing institute for expert training in nuclear safety (the European

Nuclear Safety Training and Tutoring Institute, ENSTTI). The development of a training

module at a generalist level with emphasis on the technical review of the safety case is on-

going. The module will be presented and evaluated in the pilot training session in 2017.

5. Interactions with Civil Society

The quality of the decision-making process, and its compliance to international rules and

conventions, includes several requirements such as maintaining over time consultation and

interaction with interested parties in the decision-making process. It is therefore crucial for

the consistency of the SITEX-II project that interaction with CS is embedded all along the

development of the future SITEX network. This is expected to contribute to transparency in

the specific area of expertise, supporting the development of interactions between Expertise

and Society functions at different levels of governance and at different steps of the decision-

making process. SITEX-II involves, as partners, representatives of NGOs and interacts with

CS participants through workshops covering three thematic tasks, namely: R&D, safety

culture/review and governance. The results will be integrated in one deliverable addressing

the conditions and means for developing interactions with CS in the framework of the

foreseen SITEX network. The constructive discussions that took place to date within SITEX-

II allowed both institutional and CS participants to exchange and challenge their views,

fostering mutual understanding, notably through the elaboration of the SRA. The need for

building mutual understanding led to the development of – and was in turn dynamized by –

an innovative multi-stakeholders evaluation process and tool, allowing for a participative and

comparative discussion of alternative scenarios of long-term RWM which target passive

safety as their end point.

6. Conclusion

The progress to date of the EURATOM Horizon2020 SITEX-II project shows that

developing and coordinating joint and harmonized activities at the international level

supporting the independent Expertise Function is achievable and promising in the field of

geological disposal safety. Particularly, the involvement of of NGO representatives linked

with a wider group of CS participants within SITEX-II should allow the Expertise Function

to better account for societal concerns in its future networking activities, thus strengthening

the decision making process. At this stage, the launching of the Expertise Function network is

foreseen in 2018-2019, together with the Joint Programming if supported, meaning an

ambitious calendar that calls for interested organizations that are not part of the SITEX-II

project to express their interest now.

This project has received funding from the EURATOM research and training programme 2014-2018 under grant agreement

No 662152.

References

[1] EC FP7 SITEX project, “D6.1 Conditions for establishing a sustainable expertise

network”, 2014.

[2] IAEA, “Safety guide GS-G-1.1-Organization and staffing of the regulatory body for

nuclear facilities”, Vienna, 2002.


37

[3] IAEA, “Safety guide GS-G-1.2-Review and assessment of nuclear facilities by the

regulatory body”, Vienna, 2002.

[4] DETILLEUX, V. et al., “Overview of the Strategic Research Agenda in the field of

geological disposal of radioactive waste developed by the Expertise Function in the EC-

H2020-SITEX-II project”, EUROSAFE Forum 2016, 7th & 8th November 2016, Munich,

Germany (in preparation).


38


RESEARCH AND DEVELOPMENT NEEDS IN A STEP-WISE PROCESS FOR THE

NUCLEAR WASTE PROGRAMME IN SWEDEN

A. Ström 1, K. Pers

2, J. Andersson

1, E. Ekeroth

1, A. Hedin

1

1 Swedish Nuclear Fuel and Waste Mgmt. Co. (SKB), Stockholm, Sweden

2 SKB International AB, Stockholm, Sweden


Abstract The license holders have formed Swedish Nuclear Fuel and Waste Management Co. (SKB) to on

their behalf develop and manage a programme for the research and development activities needed to manage

and dispose of nuclear waste and spent nuclear fuel in a safe manner. The disposal of waste from

decommissioning and dismantling of nuclear power plants is also part of SKB´s assignment. Such a programme

(RD&D Programme) has since 1986 been submitted every third year to the Swedish Radiation Safety Authority

(SSM) for review as preparation for a Government decision on the programme.

After more than 30 years of research and development regarding final disposal of spent nuclear fuel, an

application under the Nuclear Activities Act for final disposal of spent nuclear fuel and an application under the

Environmental Code for the KBS-3 system was submitted in March 2011. These license applications provided a

summary of the current status of the development of the KBS-3-system and included a safety assessment. An

application to extend the existing final repository for short-lived radioactive waste was submitted in 2014.

The licensing processes are under way for both these repositories. Even though a large number of issues may be

considered resolved regarding the systems there are still substantial technology development and demonstration

efforts planned before disposal can begin and the facilities be operated as an industrial enterprise. Furthermore,

SSM's regulations specify that development and licensing of nuclear facilities will be achieved through a

stepwise process in which the requirements of the facility, its design and technical solutions is gradually

established based on research, technology development and evaluation of safety after closure.

RD&D programme 2016 was submitted in September. Adjusted to the current situation, needs for future

research and technology development is based on the stepwise decision process described above. The milestones

that are linked to major decision steps for new and extended facilities determine the required level of knowledge

and development of technology. The safety reports together with the comments made by SSM in connection

with the review of the applications, as well as audits of previous RD&D programmes, are the basis for the

programme.

Key Words: Waste disposal, spent fuel, research, technology development, safety analysis

1. Background and introduction

The Swedish power industry has been generating electricity by means of nuclear power for

more than 40 years. During this time, a large part of the system for management and disposal

of the radioactive waste and the spent nuclear fuel has been built up. The system consists of

the interim storage facility for spent nuclear fuel (Clab), the final repository for short-lived

radioactive waste (SFR) and a system for transportation of nuclear waste.

What remains to be done is to build and commission the system of facilities, the KBS-3

system, needed for final disposal of spent fuel shown on Figure 1. This work includes

building a facility part for encapsulation of the spent nuclear fuel, developing transport casks

for shipping canisters, and building a final repository where the canisters will be deposited.

For disposal of short-lived low- and intermediate-level waste, the existing repository SFR

will be extended, containers will be developed for transportation of long-lived waste, and



39

eventually a final repository for long-lived waste will also be built.

The process of construction and commissioning a new facility, or extending an existing

facility, consists of several phases. In 2011, after an extensive siting phase, SKB submitted an

application under the Nuclear Activities Act for final disposal of spent nuclear fuel and an

application under the Environmental Code for the KBS-3 system (encapsulated spent nuclear

fuel in copper canisters with an insert of cast iron, embedded in bentonite clay at 500m depth

in crystalline bedrock) adjacent to the Forsmark nuclear power plant site. In order to be able

to dispose of all additional short-lived operational waste from dismantling, SKB submitted an

application in 2014 for the extension of the SFR facility.

FIG. 1. The Swedish system for management of nuclear waste.

Under the Nuclear Activities Act, the nuclear power companies shall draw up a programme

for the research and development activities and other measures needed to manage and dispose

of the nuclear waste and the spent nuclear fuel in a safe manner and to decommission the

nuclear power plants. The license holders have formed SKB to on their behalf develop and

manage such a programme. The RD&D Programme [1, 2] has since 1986 been submitted

every third year to the regulator SSM for review as preparation for a Government decision on

the programme. This process for regular reporting and review of results and plans has

contributed significantly to the development of a high scientific quality of the work and an

open and transparent review mechanism. The regular review of the RD&D-programmes

every third year has had a significant influence on the programme.

2. Implementation plan

The implementation plan describes the measures needed to meet SKB’s obligations and when

applications and other legally mandated reports for the facilities are planned to be submitted.

During 2015, decisions have been taken on an early shutdown of four reactors in Sweden.

This means that the total amount of fuel that will be managed within the programme

decreases. The remaining six reactors are planned to be operated for 60 years. Assuming that

all the reactors have been taken out of service by 2045, SKB’s three final repositories (the

Nuclear Fuel Repository, SFR and SFL) can be closed in about 60 years. These times are

important premises in the planning.

The estimated start of construction for the Nuclear Fuel Repository is 2020 and that for the

Encapsulation part of Clink is 2022. These facilities will be in operation simultaneously in


40

2030. The Encapsulation Project includes planning, design, construction and commissioning

of the integrated facility for interim storage and encapsulation in Oskarshamn. For the

Nuclear Fuel Repository Project the final phase of system design of the final repository’s

facility parts and technical systems has recently been completed.

The extended repository for low and intermediate-level waste, SFR, is expected to be ready

for operation in 2028 to meet the needs of the nuclear power industry to dispose of nuclear

waste from operation and decommissioning of the nuclear power reactors. SKB plans to

apply for a licence to build the next final repository for long lived waste, SFL, in around

2030.

The licensing processes are under way and are expected to take several years. The relatively

long time horizon covered by SKB’s planning means that the planning premises may change

in the meantime and be handled accordingly.

Work is under way with safety analysis reports for the facilities which have to be submitted

prior to the start of construction. This work is based on experience from the preparation of the

safety analysis reports submitted with the applications, and from the reviews conducted

within the on-going licensing processes. The construction projects and the work with safety

analysis reports are primary beneficiaries of the technology development and scientific

research that is being carried out.

3. Continued research and technology development

For establishing nuclear facilities, planning is based on the stepwise decision process in the

Nuclear Activities Act and SSM regulations. The safety analysis report (SAR) is central and

should provide an overall view of how the safety of the facility is arranged in order to protect

human health and the environment against nuclear accidents. The report shall reflect the

facility as built, analysed and verified, as well as show how the requirements on its design,

function, organisation and activities are met. The implementer needs to provide successively

refined safety reports to the regulator.

The planning and milestones related to decision steps in the form of applications and safety

analysis reports determine when knowledge and development of the technology needs to have

reached a certain level, while SSM’s approval determines when SKB can commence

construction and operation of the facilities.

SKB has, as part of the applications of new and extended facilities now in progress, produced

collective accounts of the state of knowledge and the status of technology development. In

conjunction with this, the importance of remaining uncertainties regarding the ability to fulfil

the requirements on protection of humans and the environment against radiation after closure

of the repository has been evaluated. These reports, together with the viewpoints that have

been submitted in conjunction with the license review and reviews of previous RD&D

programmes, form a basis for the planned activities for research and technology development

within various disciplines.

The need for research and development activities can be divided into three main categories:

The need for an increased process understanding, i.e. scientific understanding of the

processes that influence the final disposal and thus the basis for assessing their

importance to safety after closure.


41

The need for knowledge and competence regarding design, construction, manufacture

and installation of the components included in the system.

The need for knowledge and competence of inspection and testing to verify that the

barriers and components are produced and installed according to approved

specifications and thereby satisfy the requirements.

Based on this, the research and technology development needed to solve the design and

construction issues relating to the repositories, and the research needed to carry out

assessments of the safety of the repositories post-closure, has been identified and justified.

4. Examples of important issues - research and technology development

The comprehensive research, development and planning work conducted over four decades

has led to many issues of importance for the nuclear waste programme being treated and

resolved. Here, very brief accounts of the need for the research and development being

identified for the remaining parts of the nuclear waste programme are exemplified.

For the KBS-3 method copper canister is the containment barrier. Continued work concerns

both the research on copper canister properties in the repository environment and

technological development in order to be able to produce canisters, verify them against

stipulated requirements and handle them in the KBS-3 system. For the assessment of post-

closure safety, there are issues regarding corrosion and creep that require further research.

Sulphide is the dominant long-term copper corroding agent in a KBS-3 repository. A better

understanding of the details concerning sulphide corrosion strengthens the scientific basis for

the safety assessment. The understanding of copper creep in the presence of mechanical loads

is incomplete. In order to be able to improve the modelling of creep in the assessment of

canister strength, among other things the understanding of how admixture of phosphorus

leads to favourable creep properties needs to be strengthened.

Clay materials are used in all three repositories: as buffer and backfill in the KBS-3

repository, in silo filling in SFR and as a barrier in the rock vault for the legacy waste in SFL.

For the KBS-3 the design of the buffer, backfill and closure needs to be further developed

prior to the continued design of the final repository as well as the production system for

bentonite components. The need for measures for quality assurance during manufacturing,

handling and installation needs to be further detailed.

5. Concluding remarks

It is of utmost importance to address the issues within research and technology development

that are most relevant for the development of new facilities at the time when they are needed

and in a cost efficient way. The SKB RD&D Programme 2016 includes an up-to-date

planning and presents those issues in a structured and step wise procedure based on the

milestones for all new facilities and for other measures needed.

REFERENCES

[1] SKB. RD&D Programme 2013. Programme for research, development and demonstration

of methods for the management and disposal of nuclear waste, SKB TR-13-18, Svensk

Kärnbränslehantering AB, Sweden, www.skb.se.


42

[2] SKB. RD&D Programme 2016. Programme for research, development and demonstration

of methods for the management and disposal of nuclear waste, SKB TR-16-NN, Svensk

Kärnbränslehantering AB, Sweden, www.skb.se – translation in progress.


43


GERMANYS NEW ROUTE TOWARDS A REPOSITORY FOR HLW – SCIENTIFIC

CHALLENGES

F. Charlier, B. Thomauske

RWTH Aachen University, Institute of Nuclear Engineering and Technology Transfer (NET),

Aachen, Germany


Abstract. Since 2011 Germany is pursuing a phase out strategy concerning the use of nuclear power for

electricity production. This decision was influenced by the Fukushima event.

In 2013 the federal government announced that they also had achieved an agreement with the Federal States in

Germany on a law to restart the site selection for a repository for spent fuel and high active heat producing waste

from scratch. The consequence of this law is a delay of at least two decades to start operation of a final disposal

site.

At first a commission had been installed to evaluate the Site Selection Law and to develop basic principles for

site selection, including safety requirements and selection criteria for rock formations.

The site selection then might start after the next federal election in 2017 at the earliest probably based on a new

site selection law.

A new repository site should be determined till 2031 and for this site the more detailed site investigation will

take place followed by a detailed safety analysis, before the erection of the repository can start.

Based on the present procedural steps, it seems to be rather unlikely to determine a repository site till 2031.

There will be a delay of at least 20 years compared to the schedule given in the site selection law until a

repository site can be determined.

Therefore it is important to think about possibilities to accelerate the process without any reduction in safety.

This paper presents main future needs for research and development on the German path towards a repository

site for HLW.

1. Final Disposal of Radioactive Waste in Germany

From 1979 until 2013 the salt dome of Gorleben was investigated for the disposal of high

active heat generating waste. This site investigation was stopped in 2013 after a new site

selection act came into power.

This site selection act was evaluated by a commission. It is now intended to start a new site

selection procedure from scratch including salt, clay and crystalline as host rocks.

Besides for negligible heat generating waste the iron ore mine Konrad had been licensed in

2002. Since then it is transformed into a repository. It is expected that Konrad will start in

operation around 2022.

An overview over the German disposal situation is given in TABLE I.



44

TABLE I: Disposal Projects in Germany

Project Geological

Formation

Purpose Actual Status Waste

Gorleben

1979 -2013

Salt dome Repository for all

types of radioactive

waste especially

high-level and heat-

generating waste

All investigations

are stopped in 2013

But will take part in

the new site

selection

17,000 t

HLW/spent fuel

New site selection

2017-≥ 2050

Salt

Clay

Crystalline

Repository for high-

level and heat-

generating waste

Evaluation of the

site selection act

17,000 t

HLW/spent fuel

Konrad

since 1982

Iron ore Repository for long

lived waste with

negligible heat

generation

Licence issued 2002

Start of operation

≥ 2022

Operation: ≈ 35

years

300,000 m3

LLW/ILW

2. Site Selection Process

The procedural steps to determine a repository site are [1]:

1. A first stage to evaluate the legal regulations and to determine general criteria.

2. Investigation of potential siting regions.

3. Exploration from above ground.

4. Exploration of the underground area.

5. Comparison of sites.

6. Recommendation of one site.

7. Determination of a site by federal law.

8. Licensing procedure for the proof of safety at the defined site based on a detailed

underground exploration.

9. Construction of the facility after legal verification of the approval decision, if

applicable.

This stepwise approach - including the underground exploration - is based on the German

final disposal concept from earlier times.

At first starting from a “white” German map exclusion criteria will be applied. For the

remaining areas, minimum criteria and weighing criteria will be adopted and result in regions

or sites which may be suitable.

Among these using safety analyses several regions will be selected which turn out to be the

most suitable candidates for a site investigation from above ground. Based on the results of

the site investigations from above ground, a few sites will be identified as the candidates with

the highest expectations with respect to suitability.


45

After a site investigation of the host rock from below ground one site will be selected after

safety analyses and proposed to become the site for which the licensing procedure should be

performed. The site selection process leads to one site for which the licensing procedure will

be initiated.

The target of the site selection process is to find in a transparent way criteria based one site

which is expected to then be the best possible solution.

If it would turn out within the licensing process that the selected site cannot be licensed due to

safety reasons based on new findings a setback has to be initiated and one has to go back one

or two steps in the process depending on the new insights.

3. Paths Forward

The commission has analysed the different potential solutions to dispose of high active heat

generating nuclear waste. The preferred solutions – called path - based on the present state of

the art is the final disposal in deep geological formations in a mine [1].

Besides there are other potential solutions, where the technologies are not yet available but

which may turn out as possible technologies for the treatment or disposal of theses waste

stream. They should be analysed repeatedly after certain time steps.

Especially the final disposal in deep boreholes offers an alternative to the disposal in a mine.

But at the moment questions like recoverability or what if the disposal process fails are not

yet answered. Here it is intended to watch the technology development.

4. Criteria

The commission discussed geological and societal criteria but agreed in the main principle

that safety has priority. All other criteria are seen as secondary with regard to this main

important criterion. The criteria are differentiated between [1]:

Exclusion criteria

Minimum requirements

Weighing criteria

The main principles for the site selection process are:

Safety is of priority;

Recoverability, reversibility;

Step by step approach;

No right of veto of the regions/sites but they should have the possibility that the

process have to be iterated by one step;

Transparency, Public participation and Stakeholder involvement.

5. R&D needs and scientific challenges

The German “Entsorgungskommissin (ESK)” [2] describes the R&D demand as follows:

1. Specific R&D on host rocks:

- Clay, Crystalline, Salt

2. R&D, independent of host rocks


46

- Safety concepts

- Repository concepts

- Interaction between repository, barriers and host rocks

- Site evaluation

- Characterization and comparison of sites

- Retrievability and recoverability

- Safety analyses and concepts for long-term safety

In addition, R&D in the following topics is necessary:

[The list should not be seen as being complete.]

- Fleshing out the steps of the site selection process,

- Development of canister requirements

- New or further development of canister concepts

- Application of the “isolating rock zone”-concept

- Development and demonstration of handling the waste / the canisters

- Development of concepts for following the site selection path transparent

- Development of concepts for public participation

- Development of concepts for stakeholder involvement

6. Conclusion

The stepwise approach for finding a suitable site for a repository for Germany’s heat

generating waste will start in 2017.

The preferred solution based on the present state of the art is the final disposal in deep

geological formations in a mine.

The scientific challenges and the upcoming R&D program are complex as shown in chapter 5

and due to the purposed “step-back-option” (learning process), the R&D program has to react

flexible to new findings, demands or modifications of criteria or requirements.

All R&D has to be identified and to be prioritized in the now starting site selection process.

REFERENCES

[1] Kommission Lagerung hoch radioaktiver Abfallstoffe, Abschlussbericht der Kommission

Lagerung hoch radioaktiver Abfallstoffe, K.-Drs. 268 (2016).

[2] Endlagerkommission (ESK), Stellungnahme der Entsorgungskommission,

Endlagerforschung in Deutschland: Anmerkungen zu Forschungsinhalten und

Forschungssteuerung, ESK (2016).


47


RECENT SAFETY ASSESSMENT OF A REFERENCE GEOLOGICAL DISPOSAL

SYSTEM FOR RADIOACTIVE WASTE FROM PYRO-PROCESSING IN KOREA

J.-W. Kim, D.-K. Cho, J. Jeong, M.-H. Baik, K. Kim

Korea Atomic Energy Research Institute (KAERI), Daejeon, Korea


Abstract. For a long-term safety assessment to be comprehensive, complex scenarios should be assessed

systematically by combining various scenarios with aleatory uncertainty. A methodology for a risk-based

safety assessment of complex scenarios considering the long-term complementary impacts on the disposal

system has been newly suggested by KAERI. This new methodology was recently implemented in an

upgraded version of KAERI’s TSPA model (K-PAM). KAERI’s current TSPA model contains many

necessary abstractions and a limit in associating the key physical processes. As a further study, the TSPA model

will be moved to the process model level by utilizing a high-performance computing system.

Key Words: Complex scenario; Risk-based safety assessment; K-PAM.

1. Introduction

Since 2007, the Korea Atomic Energy Research Institute (KAERI) has studied the geological

disposal of radioactive waste generated from the pyro-processing of PWR spent nuclear fuel

[1]. The study mainly includes the characterization of geological media, the design of a

reference disposal system, and the overall safety assessment of the disposal system. The

characterization of geological media at different scales has been mainly conducted at the

KAERI Underground Research Tunnel (KURT) area, the host rock of which is granite. The

conceptual design of the reference disposal system is basically based on the Swedish KBS-3

concept. For the safety assessment of a hypothetical disposal system, a total system

performance assessment (TSPA) model was developed using GoldSim. For a long-term safety

assessment to be comprehensive, complex scenarios should be assessed systematically by

combining various scenarios with aleatory uncertainty. In this study, a methodology for a

risk-based safety assessment of complex scenarios considering the long-term complementary

impacts on the disposal system is presented and implemented in an upgraded version of

KAERI’s TSPA model (K-PAM). For an illustration, a statistical analysis of historical seismic

events and well exploitation in Korea was utilized to generate a complex scenario for a risk-

based safety assessment.

2. Reference Disposal System

KAERI presented a preliminary conceptual design of a geological disposal system for the

radioactive wastes generated from the pyro-processing of PWR spent nuclear fuel. The

radioactive wastes were classified into two groups: (1) low & intermediate-level metal waste

which consists of hull materials and support frames and (2) high-level ceramic waste which is

vitrified molten salt from the electrowinning process. The metal wastes are emplaced in a

storage canister (stainless steel) and then the storage canisters are packaged by polymer

concrete, so-called metal waste disposal package (MWDP). MWDPs are supposed to be

stacked up with buffer materials in the tunnel at 200 m depth. The ceramic wastes are

emplaced in a storage canister (stainless steel) and then the storage canisters are packaged in a

disposal canister which consists of an inner container for the structural stability and an outer



48

shell for corrosion resistance. The disposal canisters are supposed to be emplaced with buffer

materials in the borehole at 500 m depth (FIG. 1).

< Reference Disposal System >

< Metal Waste >

< Ceramic Waste >

FIG. 1. Conceptual design of a geological disposal system presented by KAERI.

3. Risk-based Safety Assessment

3.1.K-PAM Methodology

The risk-based safety assessment methodology consists of 5 steps as shown in FIG. 2.

The external events include natural disruptive events, such as earthquake, etc., and human

intrusion. In the 1st step, the properties of those events related to the performance of the

disposal system are digitized and represented by probability density functions (PDFs). In the

case of an earthquake, for example, the properties can be the event occurrence rate,

magnitude, distance from the hypocenter, etc. The PDFs of each property have to be carefully

determined based on the historical records, a statistical analysis, expert judgments, etc. The

PDFs of each property are converted into cumulative density functions (CDFs) for a scenario

combination.

In the 2nd step, how the external events will affect the disposal system is defined and the

complex scenario generation criteria are determined. The external events will discriminatorily

affect each part of the disposal system, such as an engineered barrier system (EBS), natural

barrier system (NBS), and the biosphere. The impacts on the disposal system are also

dependent on the properties of the external events. Some impacts can be irreversible so that

the influence continues during the period of assessment, and some impacts can be reversible

so that the disrupted parts are recovered after some time, or the influence of some repeating

impacts can be increased gradually. This process also has to be carefully conducted based on

the analogical interpretations of the experimental results and the relevant field data.


49

In the 3rd step, a complex scenario is

generated based on the criteria defined in the

previous step. Monte-Carlo sampling method

is utilized as random numbers are

independently generated and converted into

the occurrence times and/or the values of the

properties using the predefined CDFs for each

property of the external events. The types of

impacts by the external events are then

determined based on the criteria. As all

impacts on the disposal system are arranged in

the process of time, a complex scenario is

finally completed. For every iteration, a new

complex scenario is preliminarily generated

through this step.

In the 4th step, each complex scenario

developed in the 3rd step is simulated using

the user-defined TSPA model. As the results

of the scenario assessments, the exposure dose

rates to the representative person are computed

for each scenario. Because the complex

scenario was randomly generated based on the

criteria and their probabilities, the resulting

exposure dose rates already involve the

probability of the scenario. In other words, if an exposure dose rate is obtained often from the

iterations of scenario assessments, it implies that the scenario related to the exposure dose rate

has a high occurrence probability. After each iteration, the exposure dose rates are

cumulatively averaged and converted into the total risk using a dose-to-risk conversion factor.

As the number of iterations increases, the results will be statistically stabilized. If the

difference between the risks calculated in each iteration is less than the user-defined

convergence criteria, it is assumed that the number of iterations is sufficient to consider

exhaustively all possible scenarios.

In the final step which is a post-process step, the final risk, the occurrence probabilities of

each scenario, and the complementary safety indicators are computed as ordered.

Additionally, sensitivity analysis can also be conducted in this step.

3.2.K-PAM Modeling System

The methodology above was numerically implemented in an upgraded version of KAERI’s

TSPA model (K-PAM). The overall computing steps in FIG. 2 are conducted using Matlab

except the scenario assessment (4th step) which is conducted using GoldSim. That is, a

Matlab-based overall computing system is equipped with a scenario assessment module

which was developed using GoldSim (FIG. 3). The GoldSim-based safety assessment model

explains the source term, radionuclide transport in the EBS and far-field host rock (NBS), and

radionuclide transfers in the biosphere. The radionuclide transport in the EBS includes the

radionuclide release from a MWDP (metal waste) or disposal canister (ceramic waste),

diffusive transport through buffer material, sorption, precipitation, and radioactive decay in

the EBS. In the far-field host rock, radionuclide transports through the fractured rock

undergoing sorption, precipitation, matrix diffusion, and radioactive decay.

FIG. 2. Flowchart of risk-based safety

assessment.


50

FIG. 3. GoldSim-based TSPA model and an illustrative result of the risk-based safety assessment.

3.3.Illustration

An illustrative result of the risk-based safety assessment is depicted in FIG. 3. In the

illustration, two external events, earthquake and well intrusion, were considered in the

complex scenario generation. From the results, the computation was successfully converged

into less than 1% risk-change after about 400 iterations. The time-series of dose for each

iteration are depicted with gray line, and the median dose and the risk are depicted with black

and red lines, respectively.

4. Concluding Remarks

A methodology and a modeling system (K-PAM) for a risk-based safety assessment of

complex scenarios considering the long-term complementary impacts on the disposal system

were developed in this study. The results reasonably confirm the efficiency and stability of

the modeling system. From the risk-based safety assessment of complex scenarios, the

reliability, safety and public confidence of the disposal system is expected to be convinced

more efficiently.

KAERI’s current TSPA model contains many necessary abstractions and a limit in associating

the key physical processes. As a further study, the TSPA model will be moved to the process

model level by utilizing a high-performance computing system.

REFERENCES

[1] KOREA ATOMIC ENERGY RESEARCH INSTITUTE, Geological Disposal of

Pyroprocessed Waste from PWR Spent Fuel in Korea, KAERI/TR-4525/2011, KAERI,

Korea (2011).

102

103

104

105

106

10-10

10-9

10-8

10-7

10-6

10-5

10-4

10-3

10-2

10-1

100

Dose (

mS

v/y

r) / R

isk (

/yr)

Time (yr)


51


ASSESSMENT OF DECAY HEAT IN PROCESS OF SPENT NUCLEAR FUEL

DISPOSAL

Y. Kovbasenko

State Scientific and Technical Centre on Nuclear and Radiation Safety (SSTC NRS), Kyiv,

Ukraine


Abstract. Residual energy release of standard VVER-1000 spent fuel assemblies was calculated with the U.S.

SCALE code package for a storage period of 50–1000 years. WESTINGHOUSE (USA) and TVEL (Russia) fuel

assemblies operating currently in the reactor cores of Ukrainian NPPs were considered. The calculations are

provided for average geometrical, material and operating parameters of fuel assemblies.

Upon the results obtained, empirical relations based on the sum of two exponential functions are proposed. They

describe well the dependence of residual power release in spent fuel on the storage time from 100 to 1000 years.

For the case of final disposal of spent fuel in sealed containers in geological rock formations, it is assumed that

thermal radiation may be the only mechanism for heat removal from spent fuel. Based on the balance of power

release in spent fuel and thermal radiation power of the blackbody, the time required for interim storage of spent

fuel assemblies was conservatively assessed so that fuel temperature in final disposal would not exceed limiting

values (350-500 С). In this case, conservative assessments of the minimal required time of interim storage are

from 100 to 200 years.

Key Words: spent fuel assemblies, residual energy release, storage of spent fuel.

1. Introduction

In the final disposal of spent nuclear fuel, it is assumed that there will be no monitoring

operations or, if any, they will be minimized. Hence, it is very important to assess correctly

the influence of the processes that occur in fuel on its storage parameters. One of these

processes is residual power release in spent fuel.

Storage of spent nuclear fuel in sealed cavity in deep geological formations is commonly

considered as the main option for its final disposal. In this case, heat removal from fuel will

be very limited and even insignificant residual power release can lead to substantial fuel heat-

up in the storage process.

The final disposal is preceded by two stages: cooling a spent fuel in reactor pool and interim

(often dry) storage. Their objective is to decrease residual heat release to an acceptable level.

This paper provides preliminary assessments of the time required for cooling of spent fuel

prior to its final disposal.

6. Determination of residual heat of spent fuel assemblies

Consider the residual heat release in typical fuel of Ukrainian VVER-1000 NPPs produced by

the Russian TVEL and U.S. Westinghouse fuel companies.

To determine the residual heat of spent VVER-1000 fuel assemblies – US SCALE code

package was selected. The SCALE package includes computer modules, which combining

programs and libraries to calculate one or another problem (criticality analysis, radiation

safety, heat transfer, isotopic composition vs. burnup). The most complete description of the

programs included in the SCALE is provided in [1]. The applicability of the SCALE code



52

package and its libraries of neutron-physical constants for modeling VVER fuel are

considered in [2]. The calculations were performed with the use of standart 44GROUPNDF5

library of neutron-physical constants.

Calculations were made for reactor cells of VVER-1000 fuel under the burnup level up to 50

GWt*day/tU in 4 year fuel cycle. This cells were composed of the typical modern fuel

assemblies TVS-A of Russian TVEL suppliers (Fig.1) and new fuel assemblies FA-WR of

Westinghouse company (Fig.2). The main features and differences in geometrical and

material parameters of TVS-A and FA-WR used in the calculations are presented in Table I.

The results of these calculations are shown in Fig. 3. The results demonstrate that residual

heat in fuel assemblies of both types is quite close. For a period of 50-1000 years, the

numerical values are described well by the following empirical dependence:

P (Wt/t) = 424.4*(1.4*exp(-0.02*t+1.0)+0.6*exp(-0.003*t+0.15))+65,

where t (years) is post-operational period.

TABLE I: SOME DIFFERENCES IN GEOMETRY AND MATERIAL PARAMETERS OF TVS-A

AND FA-WR

Parameter TVS-А (TVEL) FA-WR (Westinghouse)

Fuel stack length 3530 mm 3530 mm

Central Zone length (nom.) 3530 mm 3225.2 mm

Axial Blanket length (nom.) - 2 zone x 152.4 mm

Fuel mass (UO2), kg 494.54.5 550.6 ± 5.0

Fuel pin (312 pieces)

Enrichment, wt% 306*4.4%+

6*3.6%(BA)

240*4.2%+60*3.9%+

6*3.6%+6*3.0%(BA)

0.714% (blanket)

Pellet ID / OD, mm 1.4 / 7.57 - / 7.84

Cladding ID / OD, mm 7.73 / 9.1 8.0 / 9.14

Cladding material/ density, g/ccm alloy Э110 (an alloy

of zirconium) / 6.45 alloy ZIRLOTM / 6.55

Central tube

ID / OD, mm 11.0 / 13.0 11.0 / 12.6

Material / density, g/ccm alloy Э635(an alloy

of zirconium) / 6.45 alloy ZIRLOTM / 6.55

Guide tube (18 pieces)

ID / OD, mm 10.9 / 12.6 11.0 / 12.6

Material alloy Э635 alloy ZIRLOTM

Spacer grid (13 pieces in fuel zone)

Mass, g 550 830

Material / density, g/ccm alloy Э110 / 6.45 alloy 718 / 8.18

Ribs (6 stiffener corners)

Width / thickness, mm 52 / 0,65 -

Material alloy Э635 -


53

FIG.1. TVS-A model

FIG.2. FA-WR model

FIG.3. Decay heat vs. time

7. Determination of time required for interim storage of spent fuel

In the final disposal of spent fuel in closed underground compartments, thermal radiation will

be the main processes of heat removal from the fuel. If fuel assemblies are arranged in several

layers and there is no good thermal contact between them, heat exchange between FAs will

also mainly proceed through radiation.

As a model to assess the amount of heat removed through radiation, use the well-known

Stefan–Boltzmann equation for a gray body:

P=· α σ·(T14 -T2

4)·S, where:

0

300

600

900

0 200 400 600 800 1000

ТВСА-4386ТВС-WR382RR424.4*(1.4*exp(-0.02*x+1.0)+0.6*exp(-0.003*x+0.15))+65

Time, year

De

ca

y h

ea

t, W

/t


54

α - radiation coefficient (degree of blackness);

σ = 5,67*10-8·W / (m2 ·K

4) - the Stefan–Boltzmann constant;

T1 – temperature of the emitting surface; T2 – temperature of the compartment wall ;

S – area of the emitting surface.

scheme 1, 1 FA scheme 2, 7 FA scheme 3, 19 FA scheme 4, 37 FA

FIG.4. FA arrangement in final disposal

According to the published data, the typical radiation coefficient of polished metals is α=0.3 –

0.7 [3, 4]. Using simple geometrical calculations, find S=0.47 m2 – area of one FA face.

Temperature of the compartment wall is assumed to be Т1 =50оС=323 K. The limiting

temperature of FA emitting surfaces in the storage process is accepted to be

ТFAlimit = 300оС = 573K. In accordance with the above results, FA power release will be

Р~450 W after 50 years of cooling, Р~250 W after 100 years, Р~180 W after 150 years and

Р~150 W after 200 years of cooling.

If we assume that radiation comes from the surface of one FA – 6 faces (configuration 1),

then calculation with the Stefan–Boltzmann equation for FA surface temperature gives:

Р=σ(Т24-Т1

4)*6S⇒ Т2

4= Т1

4+Р/(6 α σ S)=(108,8 + 28,14)Е+8⇒ Т2=349 К

If we assume that radiation comes from the surface of 7 FAs – 18 faces (configuration 2),

then the calculation with the Stefan–Boltzmann equation for surface temperature of the

central FA gives:

7Р= α σ(Т34-Т2

4)*18S ⇒ Р= α σ(Т3

4-Т2

4)*2,57 S ⇒ Т3

4= Т1

4+Р/(6 α σ S)+Р/(2,57 α σ S)=

(108,8 + 28,14 + 65,71)Е+8=202,65Е+8 (242,87Е+8) ⇒ Т3=395К

Continuing the calculations for dense packing of greater number of fuel assemblies, we

obtain:

TABLE II. FA TEMPERATURE

Number FA / layers of FA 1/1 7/2 19/3 37/4 61/5 91/6

Central FA temperature after 50 years (α=0.7) 349 К 395 К 446 К 496 К 545 К 590 К






55

8. Conclusions

Thus, for safe final disposal up to 19 fuel assemblies (3 layers) in an underground

compartment 50 years is a sufficient period.

For safe final disposal of 37 fuel assemblies (4 layers) in an underground compartment may

require 100 years, of 61 fuel assemblies (5 layers) - 150 years and of 91 fuel assemblies may

require 200 years of interim storage. Otherwise, the fuel assemblies may be overheated after

sealing of the compartment and their integrity and configuration will be affected as a result.

REFERENCES

[1] SCALE User’s Manual. NUREG/CR-0200 Revision 6. RNL/NUREG/CSD-2/V2/R6.

[2] Y.Kovbasenko, V.Khalimonchuk, A.Kuchin, Y.Bilodid, M.Yeremenko, O.Dudka,

NUREG/CR-6736, PNNL-13694 “Validation of SCALE Sequence CSAS26 for

Criticality Safety Analysis of VVER and RBMK Fuel Designs”, Washington, U.S.

NRC, 2002.

[3] Neuer G., Thermal conductivity and thermal radiation properties of UO2, J. Non-

Equilib. Thermodyn. 1, 3-23 (1976).

[4] Siegel Robert, Howell John Thermal Radiation Heat Transfer, Fourth Edition, Taylor

&Francis, NY, 2002


56


ASSESSMENT OF PRE- AND POST-CLOSURE SAFETY IN THE NUMO SAFETY

CASE FOR A GEOLOGICAL REPOSITORY

S. Suzuki, K. Fujisaki, S. Kurosawa, K. Yamashina, A. Deguchi, H. Umeki

Nuclear Waste Management Organization of Japan (NUMO), Tokyo, Japan


Abstract. The NUMO safety case is established to improve the confidence of pre- and post-closure safety in

the Japanese geological disposal programme at the current stage prior to selection of a site.

The pre-closure safety case aims to assure both radiological and non-radiological protection of the public and

workers. Radiological protection requires radiation shielding and radionuclide containment within the disposal

facilities in case of operational perturbations. Operational perturbations, such as physical or thermal impacts on

the waste-form, are analysed using an event tree method and possible, cost-effective counter-measures identified

that would reduce their likelihood or mitigate their impact. Potential vulnerabilities of operational processes have

been considered: most of these would pose little risk to the public, but the complexity of recovery operations and

risks to workers could be significant. For protection from non-radiological hazards, the working environment

will be maintained to ensure worker comfort and safety during normal operations. In many cases, requirements

are set out in regulatory guidelines – e.g. for the ventilation system. Further, underground tunnels and ventilation

shafts should be laid out to facilitate ventilation pathways, taking transport routes for excavated rock and waste

and required active / inactive zoning into consideration.

Long-term, post-closure performance assessment is required to evaluate safety functions of specific repository

systems, with consideration of uncertainties in a realistic and rational manner, excluding excess

conservativeness. This is particularly required during site investigation to allow the pros and cons of potential

sites to be identified and the appropriateness of particular repository concepts for such sites to be evaluated.

Based on these requirements, an appropriate methodology has been developed for long-term performance

assessment in this safety case. The methodology of scenario development, which results from a desire to

combine a more conventional, bottom-up, FEP-based approach and a top-down method based on safety

functions, is appropriate to this risk-informed assessment approach. This methodology, including overall

procedures and associated toolkits, aims to increase traceability and transparency. Additionally, by clearly

reflecting the purpose and context of the safety case and state-of-the-art knowledge, it assures appropriate

degrees of completeness, comprehensiveness and sufficiency within the scenario development process. The

methodology of safety analysis, which reflects the characteristics of site and repository design as faithfully as

possible, has been improved. In particular, a radionuclide migration model for “near-field scale” (≈ several

hundred meters) has been developed based on three-dimensional mass transport analysis that reflects key

characteristics of the site and the associated repository design.

Key Words: Geological disposal, vitrified waste, TRU wastes, safety case

1. Introduction

NUMO has developed a safety case for co-disposal of HLW and TRU waste to reflect current

boundary conditions in Japan, in particular siting based on an initial open call for

communities to volunteer for initial site assessment. In particular, this involves addressing

public concerns and actions by the Government to more strongly support moving forward

with siting a geological repository, involving suggesting locations that are considered to be

more scientifically suitable.

The current Safety Case advances from the previous “H12 Report” [1], which formed the

basis for establishing NUMO in 2000 as the implementing organisation and is considered the

first generic safety case in Japan. The NUMO Safety Case has been developed to provide a



57

basic structure for subsequent safety cases that could be applied to any selected site,

emphasising the practical approaches and methodology, which will be applicable for the

conditions/constraints during an actual siting process. The NUMO Safety Case has been

extended in key areas, including assessing extreme geological events during long-term

repository evolution, widening discussion of both operational and post-closure safety,

scenario development based on a risk-informed approach, etc. This paper describes the central

issues of the safety case concerned with assessment of pre- and post-closure safety.

2. Assessment of pre-closure safety

The reference inventory includes vitrified waste produced as a result of the reprocessing of

spent fuel and “TRU waste”, which contains various types of intermediate level (but long

lived) radioactive wastes produced by reprocessing and MOX fabrication. According to the

final disposal plan [2], 40,000 packages of vitrified waste and a volume of 19,000 m3 of TRU

waste will be need to be disposed of. Radioactive protection of public and workers and non-

radiological, conventional safety for workers during construction, operation and closure of

repository are discussed.

2.1. Facility design for the radiological protection of the public and workers

Radiological protection requires radiation shielding and radionuclide containment within the

disposal facilities for all operations, extended to additionally cover potential operational

perturbations. Radiation control and facility design are based on guidelines for other nuclear

facilities [3]. Within radiation-controlled zones, most operations will be remote-handled or

will involve appropriate shielding, avoiding any significant dose to workers. Under normal

operations, radiological exposure of the public results only from highly penetrating radiation

at or beyond the site boundary. Even assuming maximum exposure times, the expected dose

beyond the boundary from the HLW handling facilities would be far below the upper limit of

radiation exposure to the general public.

To design safety measures, hazard scenarios were developed to identify operational

perturbations resulting in physical or thermal impacts on the waste-form. The scenarios were

made using event tree methodology and from this, possible, cost-effective counter-measures

identified that would reduce their likelihood or mitigate their impact, on the basis of defense-

in-depth. TABLE 1 shows the multiple measures for the fire incident.

TABLE 1: MITIGATION MEASURES FOR IDENTIFIED HAZARDS

Level in event sequence diagram Measures

Prevention of incident initiating fire Prevention of incidents providing ignition

Elimination of combustible materials

Prevention of fire propagation Elimination of combustible materials

Detection of fire (e.g. Thermal/smoke detector)

Fire extinguishing equipment

Mitigation of radionuclide release

accidents due to fire incident

Emergency exhaust filter system (if radionuclide

release is detected)

Safety of workers (linked to

conventional safety issues)

Evacuation routes

Emergency shelters

Measures such as those mentioned above are designed to provide sufficient safety margins;

however, the assessment conservatively assumes if all safety measures could fail. In practice,

the mechanical robustness of metal packages effectively assures no release of radionuclides as

a result of credible incidents in the underground facility. Potential vulnerabilities of


58

operational processes have been considered: most of these would pose little risk to the public,

but the complexity of recovery operations and risks to workers could be significant.

2.2. Facility design for the conventional safety of workers

For non-radiological protection, the working environment will be maintained to ensure

worker comfort and safety during normal operations. In many cases, requirements are set out

in regulatory guidelines – e.g. for the ventilation system. Further, underground tunnels and

ventilation shafts should be laid out to facilitate ventilation pathways, taking transport routes

for excavated rock and waste and required active / inactive zoning into consideration. For

accident situations, such as a fire underground, the evacuation pathways would be routed

along the air intake shaft, with emergency shelters provided at appropriate locations.

To fulfill such requirements, we developed a simpler concept: involving a twin emplacement

panel layout concept based on dead-end tunnels. In this concept, two horizontal connecting

tunnels are utilized (FIGURE 1), with each tunnel operated independently for construction or

operation. After finishing the construction of a disposal panel, the connecting tunnel and the

constructed area are used for waste emplacement, while new panel excavation starts from the

other connecting tunnel. Thus, the operation, ventilation and water drainage system will

switch from normal area to a radiation-controlled area in a cyclic manner. This concept may

also provide a simple evacuation pathway for emergencies such as fires.

FIG. 1 Schematic view of the twin emplacement panel layout concept.

3. Assessment of post-closure safety

3.1. Framework for post-closure safety assessment

Adopting a risk-informed assessment approach, assessment scenarios related to natural events

and processes are classified into three categories related to the probability of their occurrence

– i.e. “likely”, “less-likely” and “very unlikely”. Scenarios related to human intrusion are

treated based on a stylized approach, in line with the principle that such human intrusion

scenarios are evaluated primarily to assess the robustness of the disposal system [4].

3.2. Scenario development

NUMO developed a hybrid scenario development methodology combining top-down (safety

functions) and bottom-up (FEP-based) approaches in a complementary manner [5].

Specifically, the variables which influence a safety function allocated to a component of the

system are defined, and the factors which influence these variables are selected from the FEP

database (FIGURE 2). The treatment of each factor in a specific scenarios is determined by

assessing the probability and significance of its occurrence.

<under construction>Access ramp

Exhaust shafts

Bottom gallery

Disposal tunnel

<operational>

Connecting tunnelIntake shafts

Exhaust shaft

Emplacement zone

Disposal tunnel （TRU）

Connecting tunnel （TRU）

<emplacement

completed>

Emplacement direction

Excavation completion tunnel

Co

nn

ectin

g tu

nn

el A

Emplacement zone I

(operational)

Emplacement zone III (plan) Emplacement zone IV

(plan)

Co

nn

ectin

g tu

nn

el B

Excavation direction

Backfilling direction

Backfill completion tunnel

Mechanical Plug To access ramp

Ventilation air duct

Air Intake

Construction

material intake

Excavated rock removal

Exhaust

Excavation completion tunnel

Exhaust

Air intake

Backfill intakeEmplacement Zone II

(under construction)


59

FIG. 2 A fishbone diagram which shows the relationship between a safety function and

influencing factors.

3.3. 3.3 Modelling of radionuclide migration

The safety analysis methodology has been improved to reflect the characteristics of site and

repository design as faithfully as possible. In particular, a radionuclide migration model for

the “near-field scale” (≈ several hundred metres) has been developed based on a three-

dimensional mass transport analysis that represents key characteristics of the site and the

associated repository design. 3-D solute transport pathways are evaluated by a particle

tracking method. The various calculation cases for the safety assessment scenarios should be

carried out flexibly and efficiently, so radionuclide migration analysis taking account of

retardation processes is conducted by using 1-D model. To better represent the case

examined, the 1-D radionuclide model is fit to the solute transport properties obtained through

3-D particle tracking to create a 1-D multi-channel model.

4. Summary

The pre- and post-closure safety cases were demonstrated. The R&D will be continued to

improve the confidence in Japan throughout the siting and development of repository.

REFERENCES

[1] Japan Nuclear Cycle Development Institute, H12: Project to establish the scientific and

technical basis for HLW disposal in Japan, JNC-TN1410-2000-003, (2000).

[2] Ministry of Economy, Trade and Industries: Policy of the Final Disposal of Designated

Radioactive Waste (Cabinet Decision on May 22, 2015) (In Japanese), (2015).

[3] Nuclear Regulation Authority, The new regulatory guideline for the HLW storage, (2013).

[4] ICRP, Radiological Protection in Geological Disposal of Long-lived Solid Radioactive

Waste, ICRP Publication 122, Ann. ICRP 42 (3), (2013).

[5] Kurosawa, S., et al., Advances in scenario development for a deep geological repository in

Japan, Proceedings of Global 2015, Paris, September 20-24 (2015).

Safety function of certain component

State variable 1 State variable 2

State variable 3

・・・・・

・

Influencing factor 1-1





FEP Database

Top Down


60

03d – 13 / ID 96. Disposal of High Low Level Waste

RESEARCH, DEVELOPMENT AND DEMONSTRATION PROJECTS AT THE JOSEF

UNDERGROUND LABORATORY

J. Stastka, J. Pacovsky

Czech Technical University in Prague, Prague, Czech Republic


Abstract. The Centre of Experimental Geotechnics (CEG), the department of the Faculty of Civil

Engineering, Czech Technical University in Prague, is a full member of the Underground Research Facility

Network for Geological Disposal (IAEA, URF Network). The main role of the URF Network is to establish a

community for sharing experience and learning in the field of the geological disposal of radioactive waste. The

CEG operates the Josef Underground Laboratory, and the extensive underground laboratory space available at

this facility provides a unique background for experimental research, education, training and demonstration

activities relating to the geological disposal of radioactive waste. Although the Josef facility is not intended for

waste disposal, it does play a very important role in terms of the early stage of the site selection process for the

Czech deep geological repository.

The Expert Cooperation in the Construction of the first Czech Underground Migration Laboratory with the

Potential Application of Active Tracers project makes up one of the most important research projects currently

underway at the Josef facility. The objective is to obtain the knowledge from foreign partners necessary for

putting the first in-situ underground laboratory with the potential application of active tracers into operation in

the Czech Republic. As such research has not yet been conducted in the Czech Republic, it was essential to

engage the involvement of foreign specialists. The participation of Swiss experts from the NAGRA organisation

allows both the design and subsequent implementation of the experimental programme at the Josef facility’s

laboratories in such a way that the research and training processes are effective and so as to avoid the repetition

of outdated experimental procedures and research topics. The ongoing Mock-up Josef experiment, which

consists of an in-situ physical model simulating the vertical emplacement of a container with spent nuclear fuel,

provides a further example of one of the more important experiments underway at the facility. The in-situ

experiment involves research into the effects of heat and groundwater on the bentonite buffer surrounding a

heater which simulates a spent nuclear fuel container emplaced in an underground repository.

This article provides information on the Josef Underground Laboratory and its rich history of RD & D projects

concerning the development of the Czech deep geological repository.

Key Words: Josef Underground Laboratory, geological disposal, migration laboratory,

Mock-up Josef

1. Introduction

The Centre of Experimental Geotechnics (CEG) represents one of the most unique

departments at the Faculty of Civil Engineering, CTU in Prague. In addition to providing

teaching courses, mainly of a practical nature, in the field of geotechnics, it also specialises in

the conducting of complex RD&D projects. One of the research facility’s most important

roles is to provide practical in-situ instruction in the fields of geotechnical engineering,

geology, geochemistry, radiochemistry and radioecology. The training of future experts in this

authentic underground setting also frequently involves the participation of other Czech

universities and experienced specialists from outside the academic sphere. The IAEA

(International Atomic Energy Agency) has added the CEG to its prestigious list of

international training centres. In addition to teaching and training, the CEG is heavily

involved in a wide range of research and development activities; indeed, the Josef

Underground Laboratory, operated by the CEG, is currently being used for research purposes



61

in connection with a number of European Union-supported international experimental

projects addressing a wide range of issues related to deep repository radioactive waste

disposal (TIMODAZ - FP6, FORGE - FP7, PETRUS II - FP7, DOPAS, etc.) as well as

several domestic projects (Mock-up Josef, etc.) supported by the Czech Ministry of Industry

and Trade, the Czech Science Foundation and the Czech Radioactive Waste Repository

Authority (SURAO).

2. Research, Development and Demonstration Projects

The Josef Underground Laboratory offers more than 5km of a total of 8km of galleries

(driven during the investigation of the Mokrsko – Čelina gold deposits in the period 1981 –

1993) for teaching and research purposes. No less than seven international and domestic

research projects are currently underway at the Josef facility and a further four projects are in

the preparation stage. Whilst the CEG is gradually extending the range of research and

educational activities into other scientific fields, the main theme of both research projects and

educational courses involves issues concerning the safe disposal of spent nuclear fuel in deep

geological repositories including research into migration processes underway in real rock

environments currently being conducted by the CEG in cooperation principally with the

Nuclear Research Institute - Řež (ÚJV Řež), but also including a number of other partner

institutions.

Since the first section (Čelina West) of the Josef underground complex of galleries, with a

total length of 650m (Fig. 1), was opened for educational and research purposes in 2007, the

total length of reconstructed galleries has been gradually extended to over 5km (Čelina West,

Čelina East and Mokrsko West). The granitic rock complex (the Mokrsko West section) has

been equipped with core-forced air ventilation, a power distribution network, water supply

systems and a high-speed internet optical cable network. The various core distribution

systems installed at the facility will serve for the connection of the niche selected for the

construction of the migration laboratory for research involving the application of active

tracers.

In 2010, the first student laboratory for teaching in the field of the disposal of hazardous

substances and gases was constructed in the granitic rock medium section of the underground

complex. In 2013, the CEG FCE CTU opened a migration laboratory for research involving

non-active tracers in the Čelina West underground section (in the vicinity of the entry portals

to the underground complex). The migration laboratory was built as part of the TA CR

Determination of the Migration Parameters of Minerals with Fissure Permeability using

Fluorescent Solutions project. Clearly, therefore, the CEG has extensive experience in terms

of the reconstruction of underground galleries for the needs of specialised laboratories.

In 2011, the CEG opened a new facility within the Josef complex – the Josef Regional

Underground Research Centre (Josef URC) which involved the complete reconstruction of a

surface building to include an experimental hall, laboratories and other support facilities. This

building provides the necessary backup services for the experimental research conducted in

the underground complex.


62

FIG 1. – Horizontal layout of the Josef Underground Laboratory (left) and pictures of the entrance to

the underground complex; the upper-right picture shows the entrance to the underground complex

prior to reconstruction in 2006; the lower-right picture shows the entrance portals today

At the beginning of 2013, the CEG submitted the Inter University Laboratory for the In-situ

Teaching of Transport Processes in a Real Rock Environment development project for

approval to the Ministry of Education, Youth and Sport of the Czech Republic; the co-

researcher consists of ICT Prague.

In December 2013, together with a number of research partners (with ÚJV Řež as the senior

researcher), the CEG applied to the Technology Agency of the Czech Republic (TA CR) to

conduct the PAMIRE project (Transfer of Granitic Rock Parameters from the Micro Scale to

the Real Rock Massif Scale).

However, since neither of the above projects provide for financial support for the participation

of foreign experts, the CEG subsequently decided to apply for a grant from the Partnership

Fund of the Swiss-Czech cooperation programme. Consequently, a new project entitled

Expert Cooperation in the Construction of the first Czech Underground Migration Laboratory

with the Potential for the Application of Active Tracers commenced in 2014. Up to this time,

research into migration processes in the real environment of an underground laboratory using

active tracers was allowed at just two other European facilities – SKB, Sweden and NAGRA,

Switzerland. Importantly, the CEG (and ÚJV Řež) has enjoyed extensive cooperation with

both these facilities in the past in the context of the EU FP6 and FP7 (Euratom) research

projects. It is generally recognised that the Swiss organisation NAGRA, which runs the

Grimsel Test Site (GTS) underground laboratory, employs the most experienced experts in

the field of migration research. Therefore, the objective of the project, which will include

consultation with Swiss experts, mutual visits and a series of bilateral workshops, is to obtain

the knowledge necessary for putting into operation the Czech Republic’s first in-situ

underground laboratory with the potential for the application of active tracers.


63

3. Mock-up Josef Experiment

Since the geological disposal of high-level radioactive waste is based on the multi-barrier

concept, including the use of bentonite, the Centre of Experimental Geotechnics decided to

construct the first Czech in-situ mock-up model of a disposal place employing a bentonite

barrier. The experimental model, named Mock-up Josef, enjoys the active support of the

Czech Radioactive Waste Repository Authority (SURAO). The project, which commenced in

2012, was planned to run for four years, i.e. to 2016 or up to the time the bentonite in the

model reached full saturation. The physical model, which is situated in the Josef Underground

Laboratory, is being loaded with underground water and features a heater which simulates the

heat produced by the container with spent nuclear fuel enclosed by the bentonite layer. The

model consists of a barrier made up of bentonite blocks, a heater, a comprehensive monitoring

system and stainless steel construction equipment. The model was constructed in the Josef

surface laboratory and subsequently transported to the selected niche in the Josef underground

complex. The model was placed within a vertical disposal hole with a diameter of 750mm and

a depth of 2500mm in December 2012. The experiment is located in the granitic section of the

Josef facility (the Czech DGR development programme assumes that the future DGR will be

constructed in granite host rock).

4. Conclusions

Cooperation with international institutions provides an effective way in which to advance the

Czech geological disposal of high-level radioactive waste programme. The complex research,

development and demonstration projects conducted at the Josef facility provide information,

experience and important data relating to the various components of the disposal system.

5. Acknowledgements

Part of the work reported herein was supported by funding from Switzerland through the

Swiss Contribution to the enlarged European Union.

REFERENCES

[1] PACOVSKÝ, J.; VAŠÍČEK, R.; Josef Regional Underground Research Centre - a New

and Attractive Location for Interdisciplinary Teaching, Research and Training in the Field

of Nuclear Engineering; In: Proceedings of the 17th Pacific Basin Nuclear Conference.

(2010). ISBN 978-607-95174-1-0.

[2] SVOBODA, J.; VAŠÍČEK, R.; The Josef UEF - a new location for “in-situ” physical

modelling; In: ICPMG 2010 - 7th International Conference on Physical Modelling in

Geotechnics. (2010). ISBN 978-0-415-59288-8.

[3] ŠŤÁSTKA, J.; Mock-up Josef Demonstration Experiment; In: Tunel, vol. 23, no. 2, pp. 65

- 73, (2014). ISSN 1211-0728.


64


THE MANAGEMENT OF USED (SPENT) FUEL AND HIGH LEVEL WASTE IN

SOUTH AFRICA V Maree

1, A Carolissen

2

1National Nuclear Regulator (NNR), Cape Town, South Africa

2 National Radioactive Waste Disposal Institute (NRWDI), Pretoria, South Africa


Abstract. As a country with a nuclear power program and radioisotope production facility, the Republic of

South Africa (RSA) generates Used Nuclear Fuel (UNF) and radioactive waste through numerous activities. The

cornerstone of South Africa’s approach to addressing radioactive waste management is the Radioactive Waste

Management Policy and Strategy for the Republic of South Africa. The Policy and Strategy serves as a national

commitment to address radioactive waste management in a coordinated and cooperative manner and represents a

comprehensive radioactive waste governance framework by formulating, in addition to nuclear and other

applicable legislation, a policy and implementation strategy developed in consultation with all stakeholders. In

accordance with the Policy and Strategy, final disposal is regarded as the ultimate step in the radioactive waste

management process, although a stepwise waste management process is acceptable. Long-term storage of

specific types of waste, such as High-Level Waste (HLW), long-lived waste and high activity disused

radioactive sources, may be regarded as one of the steps in the management process. This poster presents the

South African National Radioactive Waste Management Model with a description of: the radioactive waste

management governance framework; the current HLW and UNF management, the management option and UNF

strategies. Also the poster addresses consideration of the lessons learnt from the Fukushima accident and its

impact on future radioactive waste management strategies and options, plans related to possible long term

operation of the existing nuclear power plants, introduction of new nuclear power plants and public acceptance

and challenges from anti-nuclear groups.

Key Words: Used Nuclear Fuel; High Level Waste; South Africa’s Management Strategies;

Challenges.

1. Introduction

The Republic of South Africa (RSA) recognizes the importance of the safe management of

spent fuel and radioactive waste, for this reason the country is a contracting party to the

International Atomic Energy Agency (IAEA) Joint Convention on the Safety of Spent

Nuclear Fuel Management and Safety of Radioactive Waste Management [1]. The Joint

Convention provides for the establishment and maintenance of a legislative and regulatory

framework to govern the safety of spent fuel and radioactive waste management. South Africa

fulfills its obligations under the Joint Convention by the establishment of a Radioactive Waste

Management Policy and Strategy for the Republic of South Africa (RWMP&S) [2] and has

invited the IAEA to conduct the Integrated Nuclear Infrastructure Review (INIR) mission in

2013. The INIR mission has recommended that South Africa develop an integrated national

Nuclear Fuel Cycle strategy, including Used Nuclear Fuel (UNF)/High Level Waste (HLW)

disposal [3]. South Africa already has in place a strong national radioactive waste

management model and is considering different options and strategies to address the long

term management of the UNF and HLW as recommended by the INIR.

2. Background

The past strategic programs and the current nuclear programs contribute to the generation of

HLW and UNF. HLW for legal and regulatory purposes, is defined as ‘waste with levels of



65

activity concentration high enough to generate significant quantities of heat (>2kW/m3), or

waste with large amounts of long lived radionuclides’.

In 1991, South Africa signed the Nuclear Non-Proliferation Treaty and in 1993 voluntarily

announced the dismantling of its nuclear weapons programme, HLW was generated. In the

South African context, HLW doesn’t include fuel coming from the irradiated fuel reactor

cycle. The term “used fuel” is used instead of “spent fuel” because used fuel is considered to

have useful material and is not classified as radioactive waste. UNF is produced in two main

nuclear facilities: The South African Nuclear Energy Corporation (Necsa) and Koeberg

Nuclear Power Station (KNPS). Necsa, located at Pelindaba 30 km west of Pretoria operates a

20 Megawatt tank-in-pool type nuclear research reactor: SAFARI-1 (Fig.2.). The research

reactor has been in operation for 50 years and is used in the production of medical

radioisotopes and nuclear research. KNPS is the only nuclear power plant in Africa and is

comprised of 2 Framatome PWR reactors of 900 Mwe each operated by the State Own

Company Eskom. KNPS is in operation since 1984 and situated on the Atlantic coast 40

kilometers north of Cape Town (Fig.2.).

3. South African Radioactive Waste Management Model

The overarching objective of radioactive waste management is to deal with radioactive waste

in a manner that protects human health and the environment now and in the future without

imposing undue burdens on future generations.

3.1.Radioactive Waste Management Governance Framework

The following diagram depicts the Governance Framework for radioactive waste in RSA:

Fig.1. Schematic Governance Framework for Radioactive Waste Management.

It is imperative to note that the legislative and regulatory framework for radioactive waste

management and disposal is informed by and gives effect to:

Joint Convention on the Safety of Spent Fuel Management and on the Safety of

Radioactive Waste Management [1];

National Radioactive Waste Disposal Institute Act (NRWDIA) [4];

Nuclear Energy Act [5];

RWMP&S.


66

In order to ensure that radioactive wastes are managed safely, the governance framework

makes provision for: an established legislative and regulatory framework, the necessary

organisations for implementation and providing oversight of waste management operations

and facility development. Independence between the Regulator, Waste Generators and

repository operator is the key to ensure that the RSA has an integrated and sustainable

approach to ensure that the long term management is executed. RSA’s approach to addressing

radioactive waste management issues is RWMP&S which sets a general policy for dealing

with all radioactive waste from the nuclear fuel cycle. RWMP&S developed in consultation

with all stakeholders and transparency’s principle serves as a national commitment to address

radioactive waste management. RWMP&S puts forward the following hierarchy of waste

management options to be followed, where practicable: avoiding waste and minimisation;

reuse, reprocessing and recycling; storage; conditioning and final disposal. Final disposal is

regarded as the ultimate step in the radioactive waste management process.

The RWMP&S establishes the National Committee on Radioactive Waste Management

(NCRWM). This committee is constituted by representatives from different organs of state.

One of the committee responsibilities is to evaluate the radioactive waste plans submitted by

radioactive waste generators and to provide recommendations to the Minister of Energy.

The RWMP&S also makes provision for a National Radioactive Waste Management Fund

managed by the South African Government to ensure sufficient provision for the long term

management of radioactive waste with the principle that the ‘Polluter pays’.

The NRWDIA became effective in December 2009. The NRWDIA endorsed the

establishment of the National Radioactive Waste Disposal Institute (NRWDI) which is a

national public entity. The Institute is mandated to discharge a Ministerial institutional

obligation with respect to the management of radioactive waste disposal and related waste.

The RSA acceded to the Joint Convention on the Safety of Spent Fuel Management and on

the Safety of Radioactive Waste Management (Joint Convention) in 2006. One of the objects

of the NRWDI is to fulfil national obligations in respect of the long term management of

radioactive waste disposal and related waste management activities as dictated by the Joint

Convention.

According to Section 5 of NRWDIA, the Institute must, inter alia, —

(a) perform any function that may be assigned to it by the Minister in terms of Section 55(2)

of the Nuclear Energy Act, in relation to radioactive waste disposal;

(e) manage, operate and monitor operational radioactive waste disposal facilities, including

related storage and predisposal management of radioactive waste at disposal sites;

(g) investigate the need for any new radioactive waste disposal facilities and site, design and

construct such new facilities as may be required;

(h) conduct research and develop plans for the long-term management of radioactive waste

storage and disposal.

3.2.The Current HLW and UNF Management

At present nuclear installations in South Africa use a combination of wet and dry storage for

used nuclear fuel. UNF from the KNPS is currently stored in pools on the site as well as in

casks designed and constructed for storage of used nuclear fuel. The used nuclear fuel from

the SAFARI-1 Research Reactor is initially stored in the reactor pool for at least two years to

facilitate cooling of the used fuel prior to it being transferred to an authorised dry storage

facility on the Pelindaba site. Some HLW is stored on the same site.


67

3.3.The Management Option and Strategies for HLW and UNF

The RWMS&P clearly indicates that storage on these sites is not sustainable in the long term

and considers the following waste management options for UNF and HLW: long-term above

ground storage on an off-site licensed facility; reprocessing, conditioning and recycling; direct

deep geological disposal and transmutation. Regardless of any UNF/HLW management

strategy chosen, a Centralized Interim long term off-site Storage Facility (CISF) and Deep

Geological Repository (DGR) for final disposal will be required. Like any option chosen for

the UNF, the DGR needs to be technically sound, socially acceptable, environmentally

responsible and economically feasible. Transmutation requires major investment, two cycles

options for the management of UNF can be considered: open cycle and closed cycle.

Open cycle without recycling/ reprocessing:

The fuel will be stored at the reactor site and will be transferred to a centralised off-site

storage pending the final decision or, after the transfer to a centralised off-site storage, the

UNF will be directly disposed.

Closed cycle with recycling/ reprocessing:

Firstly, the UNF will be stored at the reactor site for a specific period, secondly will be

reprocessed and finally the UNF will be disposed or, after the transfer to a centralised off-site

storage, the UNF will be reprocessed were the useful material will be reused and the waste

will be disposed.

Currently the solid low level and intermediate level waste from KNPS and Necsa are disposed

of at the national radioactive waste disposal facility Vaalputs located in the Northern Cape

Province (Fig.2.). Preliminary investigation in the early 90s has indicated that Vaalputs has

suitable characteristics that would make this site a favourite candidate to host the CISF and a

DGR.

FIG. 2. Geographical location of selected nuclear facilities.

The RSA has addressed the INIR mission’s recommendation by drafting a new policy

detailing options and strategies for UNF/ HLW management. The document was finalized and

in currently under review by the Cabinet of the RSA.

4. Key Challenges

One of the lessons learnt from the Fukushima accident was the importance to limit the UNF

inventory on-site. On-site storage should only be for cooling purposes of the UNF. The RSA

Government has committed to establish and operate a centralized off-site interim storage


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facility by 2025 and a deep geological repository by 2065 [6]. In addition, centralized off-site

interim storage facility will provide South Africa with the flexibility to make an informed

decision with regard to fuel cycle strategy (open or closed) Provision must be made for

additional waste storage/ disposal due to potential operation of new nuclear power plants as

the RSA is considering a new nuclear build programme and the possible long term operation

of the KNPS. The Fukushima accident has eroded the confidence of the public in nuclear

power and safe radioactive waste management. Hence, to be successful the waste

management programme must overcome this negative perception as technical competence is

not enough to ensure and instill stakeholder trust and acceptability. Waste Generators in RSA

must still submit their radioactive waste management plans for review and to the NCRWM.

This committee will determine the funding strategy and requirements for sustainable long

term operation of NRWDI. Funding is required for disposal activities, research and

development including investigations into waste management/disposal options. The process

of the site’s selection for the centralized off-site interim storage facility and deep geological

repository must be developed for licensing.

5. Conclusion

South Africa has an integrated and extensive national radioactive waste management model

which considers the different options and strategies to address the long term management of

the UNF and HLW as recommended by the INIR. Final disposal is regarded as the ultimate

step in the radioactive waste management process. In spite of challenges and irrespective of

the fuel strategy chosen, it is inescapable that South Africa needs the following waste

management infrastructure namely (i) CISF and (ii) a DGR. Finally, the RSA must develop

and implement a comprehensive communication strategy and plan to demystify and decipher

the public’s fears regarding the management of radioactive waste and to deepen and

strengthen stakeholder acceptance, confidence and trust.

REFERENCES

[1] Joint Convention on the Safety of Spent Fuel Management and on the Safety of

Radioactive Waste Management (IAEA, 1997).

[2] Radioactive Waste Management Policy and Strategy for the Republic of South Africa

(2005).

[3] Department of Energy, Media Statement: Nuclear Procurement Process Update, 14 July

2015.

[4] National Radioactive Waste Disposal Institute Act (NRWDIA), Act 53 of 2008.

[5] Nuclear Energy Act, 1999, (Act No. 46 of 1999).

[6] South African National Report on the Compliance to Obligations under the Joint

Convention on Safety of Spent Fuel Management and on the Safety of Radioactive Waste

Management, September 2014.


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REGULATORY EXPERIENCES IN REVIEWING CONSTRUCTION LICENSE

APPLICATION FOR THE DISPOSAL OF SPENT NUCLEAR FUEL IN FINLAND

J. Leino

Radiation and Nuclear Safety Authority, STUK, Helsinki, Finland


Abstract. Finland is one of the first countries in the world in developing a disposal solution for spent nuclear

fuel (SNF). The Construction License Application (CLA) for the Olkiluoto SNF encapsulation and disposal

facility was submitted by Posiva, the implementer, to the authorities at the end of 2012 and the Government

granted construction license in November 2015. The post-closure safety case submitted as part of the CLA was

reviewed during 2013-2015. The CLA covered both operational safety (PSAR) and post-closure safety.

In this paper, experiences gathered during the review process post-closure safety case are discussed. During the

review process some practices proved to be good but the process revealed also some needs for improvements for

the next licensing phase.

Key Words: nuclear waste, spent nuclear fuel, disposal, review of post-closure safety case.

1. Introduction

The safety case submitted as a part of the CLA was reviewed during 2013 - 2015. The safety

case covered both operational safety (PSAR) and post-closure safety documentation. The

actual review process consisted of three phases: the initial phase, detailed review and

assessment phase, and then the finalizing phase. The last phase consisted of preparation of

conclusions, writing and finalization of statements, decisions and review reports.

During the pre-licensing phase STUK started the work aiming to the readiness to review the

construction license application. STUK also increased the own competence and resources to

be prepared for the review. A strategic resource plan was made and the amount of people

working mainly for the waste management was increased. STUK made also framework

contract with 13 external experts to support STUK during the review of post-closure safety

case.

2. Regulatory experiences during pre-licensing phase

In the pre-licensing phase before the actual review phase STUK assessed Posiva’s R&D work

and draft documentation, planned the actual review process and increased its’ regulatory

resources and competence. While assessing Posiva’s R&D work and draft documentation

STUK created a list of key safety concerns and had an active communication with Posiva.

Active communication consisted of giving feedback to Posiva, discussions concerning the

content and structure of the safety case, expressing regulatory expectations, hearing of

Posiva’s expectations and forming of common understanding of the regulatory requirements.

The regulatory assessment of safety is, of course, done against regulatory safety requirements

i.e. the government decree and regulatory guidance (STUK YVL guides). In this phase

STUK’s approach was initially safety issue oriented and bottom- up assessment. This was

partly because of the regulatory safety requirements were not detailed enough. Thus, STUK

started to develop a more structured review and assessment process for the safety case review



70

and prepared a so called review plan that would be a strong basis for the review and would

guide the review. The review plan changed STUK’s approach to more regulatory requirement

oriented and safety related review basis. The review plan was seen necessary since it was

acknowledged that addressing single technical concerns in many cases did not lead to better

overall understanding of safety and often the linkage to safety was not very clear thus the

review plan collects earlier regulatory observations and expectations for the safety case.

Regulatory safety requirements given in the government decree on the safety of disposal of

nuclear waste and regulatory guides were linked to these observations and expectations. The

review plan was used as guidance both for internal and external experts participating in the

review of the safety case. It also formed a basic structure for STUK’s safety evaluation

reports for operational and post-closure safety.

3. Regulatory experiences during initial review phase

In the initial review phase STUK performed the completeness review of the safety case during

the first three months. The aim of this phase was to verify that the safety case contained all

main elements requested by the YVL guides. The conclusion from this completeness review

was that Posiva had delivered most of the documentation required and STUK could continue

the review despite some missing parts of the safety case. Nevertheless, STUK requested

Posiva to deliver the missing parts of the documentation to STUK and to update some parts of

the documentation which were considered to be too general. Based on the completeness

review observations, STUK made several requests for additional information because of the

adequacy and the quality of the safety case documentation. The extent and structure of the

post-closure safety case documentation were such that it was challenging to check the

adequacy and the quality of the elements during the initial review phase.

However, this kind of completeness review proved to be useful tool to identify the

completeness of the safety case. It also helped to identify any shortcomings and allowed

STUK to continue its review and move on to the detailed review phase on areas that were

found to be complete.

4. Regulatory experiences during detailed review phase

The detailed review phase took place three to four months after the initial review phase

depending on the documentation. The review plan which facilitated the review process of post

closure safety documentation can be considered as a good practice. The review plan that

STUK developed beforehand was seen as a strong basis for the review of CLA. Although, in

the last phase of the review it was identified that even more detailed review guidance was

needed in the next licensing phase. The more detailed guidance or review plan should consist

of guidance on how to review and assess all the elements collected in the review plan, and

elements of setting up the level of adequacy and also what are the most important and safety

significant parts in the documentation i.e. to what issues the review should be concentrating.

In the detailed review STUK did have difficulties to assess the post-closure safety regarding

some of the elements in the safety case because of the extent and structure of the safety case

and also explicit safety argumentation that was partly missing in the safety case. This was

seen understandable due the uniqueness and first-of-a-kind post-closure safety case

concerning disposal of spent nuclear fuel in Finland. Furthermore, because of missing of

some key parts in the documentation those were noticed in the completeness review the

detailed review of these parts of the documentation was delayed.


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The explicit safety argumentation has been intrinsic value and has long tradition and lot of

experience in the nuclear reactor safety assessments. Similar explicit safety argumentation

and precise minimum content and structure of the safety case that have been agreed before

hand between the regulator and the license applicant should be introduced to the nuclear

waste post-closure safety cases also. That would make post-closure safety case easier to

review for the regulator and increase traceability and transparency.

STUK decided to reform the regulatory guidance for nuclear facilities in 2005. The new

guidance was planned to be published well before Posiva was going to submit the CLA.

Publication was postponed because of delays in the project and Fukushima accident and thus

the new YVL guides were released after the submittal of the CLA. Despite of this, CLA was

written based on the new guidance. This was possible because STUK agreed with Posiva that

Posiva would use final draft versions of the guidance for the CLA documentation. A few late

changes in the guidance caused some challenges for the license applicant to prepare the

documentation as well as for STUK for the reviewing it. Thus it can be concluded that the

guidance should be ready for use for the license applicant well in advance before the

submission of the license application and any major changes in the guidance should not be

done during the review but afterwards. As a whole it was considered important that the CLA

was reviewed against the latest regulations. It should be also clear that there is a mutual

understanding between the regulator and the license applicant concerning the requirements.

During the detailed review phase STUK made approximately 30 requests for additional

information in areas where further information or clarification was needed. Requests for

additional information concerned mainly main documents e.g. PSAR and post-closure safety

case. As a result of the detailed review phase STUK accepted main documents e.g. PSAR and

post-closure safety case and submitted statement and safety evaluation report [1] to the

Government. STUK’s main conclusion was that encapsulation plant and disposal facility can

be built to be safe. Also there is a sufficient reliability that there will be no detrimental

radiation effects to the public or environment neither during the operational period nor after

decommissioning and closure of the facility.

In the statement to the government STUK raised areas that need further development before

specific construction step or before submittal of operating license application. These areas are

related for example to process for selecting suitable disposal tunnel location, further R&D and

assessment of engineered barrier system performance and development in post-closure

scenario analysis and presentation of post-closure safety case. These areas have been further

specified in PSAR and post-closure safety case decisions [2, 3] that STUK has send directly

to Posiva. Raising these areas in the statement and decisions STUK created more steps to

already step-wise licensing process. This is justified due the uniqueness and first-of-a-kind

nature of this kind of nuclear waste disposal concept.

5. Conclusions

A well planned preparation for the review during pre-licensing phase is a key factor to a

successful review. Plan for the review should be prepared and resources assured before the

review. The review should be guided as detailed as possible.

Completeness review before the detailed review phase outlines the scope of the safety case

and identifies the shortcomings in it.

The review process revealed that more thorough discussions between the regulator and the

license applicant before the review would have been needed. Especially more thorough

discussions concerning the structure of the documentation and the level of details would have


72

simplified and facilitated the review process. Regulations that are up to date are of importance

for this kind of new nuclear waste disposal concept. Also discussions concerning the mutual

understanding on the regulatory guidance are essential for avoidance of any

misunderstandings between the regulator and the license applicant. However, despite the

appeared challenges the review could be concluded because of the active communication with

the implementer during the review phase. Lessons learned during the review process have

been analyzed and development of review practices and safety guidance have been started.

REFERENCES

[1] STUK, STUK’s statement and safety assessment on the construction of the Olkiluoto

encapsulation plan and disposal facility for spent nuclear fuel, STUK-B 196, Helsinki

2015.

[2] STUK, STUK’s decision on the PSAR of the encapsulation plant and disposal facility for

spent nuclear fuel (in Finnish), 10.2.2015, Helsinki.

[3] STUK, STUK’s review on the construction license stage post closure safety case of the

spent nuclear fuel disposal in Olkiluoto, STUK-B 197, Helsinki 2015.


73


GENERIC UNDERGROUND RESEARCH FACILITY IN THE MIDDLE STAGE OF

THE SITE SELECTION PROCESS: BUKOV URF, CZECH REPUBLIC

L. Vondrovic, I. Pospíšková, J. Augusta, J. Slovák, A. Vokál

Radioactive Waste Repository Authority


Abstract. The Czech Republic’s radioactive waste disposal concept assumes the construction of a deep

geological repository in crystalline host rocks (granitic and metamorphic) at a depth of 500m below the earth’s

surface. The current stage of the site selection and evaluation process requires that the characteristics of the

geosphere be determined at a depth envisaged for the future repository. This abstract addresses the current state

of construction and preparation of the R&D programme for the new Bukov generic underground research

facility. This facility will, over the next 10 years or so, provide invaluable support for the current siting process

and the safety evaluation of the disposal concept in the Czech Republic by providing the depth calibration

parameters required to supplement the data acquired from surface exploration. The Bukov URF (located within

the Rožná uranium mine complex) is located at a depth of 600m in metamorphic rocks in the proximity of a

potential site for the construction of the future DGR. The laboratory itself is currently approaching the end of the

construction phase which commenced in 2013. The intensive characterization programme which was conducted

during the construction phase focused on the characterization of the site from the geological, geomechanical and

hydrogeological aspects. The data set acquired from this initial scientific programme will serve as input data for

the construction of synthetic geosphere models which, in turn, will serve for determining the precise location of

the various experiments and the development of specialised site description methodologies; moreover, it will

provide essential information concerning the design of the future repository. One of the most important parts of

the characterization programme will consist of the long-term monitoring of the geological processes that take

place at repository depth within the Bohemian Massif. The future experimental programme will focus on the

following principal research areas: the geosphere and materials and techniques. It is anticipated that the

geosphere part of the programme will provide a description of the characteristics of rock mass behaviour in

terms of migration properties, fracture connectivity and the long-term stability of the rock mass as well as the

stability of the underground construction itself. The materials part will focus on the long-term stability of the

various materials employed and their degradation rates. Finally, the technology part will provide valuable

information concerning the preparation of individual components of the emplacement system and the support

infrastructure.

Key Words: URF, disposal, experimental programme

1. Introduction

The SURAO generic research programme is focused on the detailed testing of the crystalline

rock concept. Generic laboratories serve as training centres for staff members,

experimentation involving mock-up experiments and the development of methodologies for

the study of rock conditions in underground environments. One of the most important aspects

of generic research consists of the testing of the validity of data collected from the earth’s

surface and the approximation of such data to depths at which the construction of the

repository is envisaged. SURAO has close connections with three underground research

centres in the Czech Republic: the Josef Gallery, the Bedřichov Water Supply Tunnel and the

Bukov Underground Research Facility (Bukov URF). The Bukov underground generic

laboratory is located in the eastern part of the Czech Republic near the Kraví hora candidate

repository site and adjacent to the Rožná uranium mine at a depth of 600m below the earth’s

surface. From the geological point of view the facility is located in the north eastern part of

the Moldanubian Zone of the Variscan orogen and is composed of migmatitised paragneisses



74

with amphibolite layers. The felsic granulites display the same deformational history as that

of the nearby Kraví hora candidate locality.

FIG. 1. Scheme of URF Bukov

2. Construction

Construction commenced in 2013 with the blasting of the main access tunnel. Following an

intensive drilling campaign (the total length of the boreholes amounted to 500m) two suitable

rock blocks were defined for testing purposes: a consolidated block made up of high quality

rock intended for diffusion and demonstration experiments and slightly fractured rock for the

performance of migration, hydraulic and material tests. The laboratory itself consists of a

300m-long connecting cross gallery with a profile of 9.2m2 leading from the access shaft and

the underground facility itself consisting of a 90m-long large-profile chamber and a gallery

niche system with a total length of 40m (see Fig 1). The second test chamber section consists

of a 20m-long niche in the front part of the access tunnel. Rock bolts will be used to provide

support for the underground sections supplemented with yieldable TH arches in areas

exhibiting more complicated geological conditions. The intensive drilling campaign consisted

of the drilling of a series of exploration boreholes of a total length of 500m into the walls of

the access tunnel for geophysical and hydrogeological monitoring purposes. The conventional

blasting method was used for the excavation of the access tunnel, whereas the smooth blasting

method was applied with respect to the laboratory niches.

3. Characterization phase

The scientific programme conducted during the construction of the facility concentrated on

the characterization of the site from the geological, geomechanical and hydrogeological points

of view. The results will serve as input material for the construction of synthetic geosphere


75

models which will, in turn, serve for the precise positioning of the various experiments

included in the research programme. The characterization programme will include the

following research areas:

Complex geological characterization

The application of a range of geological methods will be aimed at obtaining a

multidisciplinary description of the host rock in order to assist in determining the optimum

location for the performance of the experimental programme. Geological characterization will

comprise geological and structural mapping and the deciphering of the temporal, spatial and

thermal evolution of the ductile and brittle pattern. Subsequent more detailed characterization

will concentrate on more specialized study fields e.g. the radiometric dating of the fault

system, the evolution of micro-fractures within the rock, etc.

Geotechnical characterization

The geotechnical programme will be made up of three specific areas: (i) stress monitoring, (ii)

geotechnical laboratory testing and (iii) seismic monitoring. The stress measurements will

allow for the prediction of the stability of the rock mass as well as for the determination of

stress changes during the excavation process. Geotechnical testing will comprise a range of

methods that will serve for initial rock mass characterization purposes and for the provision of

input data for further geotechnical modelling. Seismic monitoring will be concerned with the

potential reactivation of the fault system during blasting and the identification of any induced

seismic activity that might occur as a result of local mining operations.

Transport properties of the rocks

The determination of the transport properties of the surrounding rock will serve for the

laboratory testing of radionuclide sorption and migration from a depth at which the

construction of the repository is envisaged.

Hydrogeological properties of the rock mass

An understanding of the behaviour of water within the repository system is crucial in terms of

safety case considerations. Hydrogeological studies therefore include the monitoring of water

influx and the evolution of the chemical and physical properties of water collected from the

surrounding rock. Borehole hydrogeological tests, tracer tests and water pressure tests will be

conducted during the experimental phase.

Synthetic geosphere models

The application of the methods described above will result in the construction of the

following synthetic geosphere models:

3D structural-geological model

3D hydrogeological model

3D geotechnical model

4. Experimental programme

The underground research facility research and experimental (R&E) programme will be

conducted in very similar conditions to those expected at the location of the future deep

geological repository. The Bukov URF will serve as a test site for assessing the behaviour of

the rocks at the candidate sites at a depth matching the expected depth of the deep geological

repository until the final site is selected and the confirmation underground laboratory is built

at that site.


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The experimental programme consists of 7 basic areas:

R&E Programme 1: Pilot characterization of the rocks in order to test the methodology for

setting up 3D Geo / GT / HG models of the site

R&E Programme 2: Testing of long-term monitoring methods for processes occurring at

repository depth

R&E Programme 3: Testing of groundwater flow / radionuclide transport models of the

fracture environment of the DGR

R&E Programme 4: Testing of the effect of the rock at repository depth on the properties of

the engineered barriers

R&E Programme 5: Testing of the development of excavation disturbed/damaged zones in

crystalline complex rocks at repository depth

R&E Programme 6: Investigation of the effect of the rock massif on the underground

structures of the DGR

R&E Programme 7: Demonstration experiments

5. Conclusion

The construction of the Bukov Underground Research Facility is fundamental in terms of the

characterization of rock masses in which it is intended that the future Czech radioactive waste

repository will be constructed. The facility is ideally located for this purpose, i.e. it is 600m

beneath the earth’s surface and situated in a crystalline rock environment. The research to be

conducted at the facility will make a significant contribution towards forming a more detailed

understanding of the processes that will take place within the repository over its lifetime.


77


CIGEO PROJECT:

FROM BASIC DESIGN TO DETAILED DESIGN – PPURSUANT TO REVERSIBILITY

F. Launeau, G. Ouzounian

Andra, French National Radioactive Waste Management Agency, Parc de la Croix Blanche,

92298 Châtenay-Malabry, France


Abstract. The Cigeo project has been in the works for 25 years. Numerous studies have been conducted, with

further specific research thanks to direct access to the Callovo-Oxfordian clay formation from the underground

laboratory of Bure-Saudron. These studies and research initially aimed to demonstrate the feasibility of the

repository. They also helped gain a high level of understanding of phenomena to support design studies and

demonstrate safety. Transition to the industrial phase began with the development of a plan for delivering waste

to the facility for disposal. The plan introduced sequencing for the various types of waste to be disposed of, and

was optimised to determine the size of inspection, transfer and handling facilities. In describing the life of the

repository and therefore the vision for its operation, it has become obvious that our generation should not impose

choices on future generations. We must provide them with reference technical solutions, with the financial

resources to implement them. It is also our duty to begin the construction and initial operating phases. However,

because the facility will operate over more than 5 generations, we must leave a degree of flexibility so that they

may reassess the options that we define and adopt their own solutions, as necessary. They will also benefit from

operational experience gathered as facility operations develop. This is the context in which the preliminary

design phase was finalised in preparation for the detailed design phase, with the aim of gradual commissioning

during the latter part of the next decade.

1. Introduction

Since the Act of December 1991 concerning research into the management of radioactive

waste, Andra has been conducting the programme for geological disposal in compliance with

the objectives set forth. The initial 15-year phase was mainly dedicated to research, including

research into alternatives to geological disposal. Following the various bids for the creation of

an underground laboratory, in 1998 the French Government selected the Bure-Saudron

facility in the Meuse and Haute-Marne departments of north-eastern France. In 2005, Andra

compiled the results and analysed them in the Dossier 2005 Argile report. The main finding of

Dossier 2005 was that geological disposal is feasible in the clay formation studied (Callovo-

Oxfordian clay) and that its safety could be proven. Based on the various results, French

Parliament passed the Planning Act in 2006, establishing geological disposal as the reference

solution for managing high-level waste (HLW) and intermediate-level long-lived waste (ILW-

LL). The facilities should be planned in a formation previously studied using an underground

laboratory, which indicates the Callovo-Oxfordian near Bure-Saudron. More detailed

investigations therefore focused on this region and in 2009, Andra proposed the location for

underground facilities. Upon completion of a series of assessments and opinions, the French

Government validated the location for the underground repository in March 2010. This began

the industrialisation process for the Cigeo project, followed by a public debate in 2013, which

became useful for later deliberations. When the preliminary design phase was completed and

before beginning the detailed design phase, the life and operation of the disposal facility were

reviewed using updated information to bring a new perspective to the industrial project. Due

to changes to regulatory requirements in France, Cigeo’s detailed design must be used for the

repository construction license application. The construction license application will therefore



78

be submitted progressively between late 2015 and mid 2018 in agreement with safety

authorities.

9. Development of Cigeo Project

Based on this initial research, Andra proposed an initial project in 2001, which was followed

by a detailed safety assessment. This was submitted for international review and created the

basis for Dossier 2005. The demonstration provided was supported by an understanding of the

phenomena affecting the behaviour of the repository gained from a sustained research effort.

The repository was no longer viewed as a single object placed in the geological environment,

but rather as a group of structures and components developing over time and subject to

relatively complex physical-chemical and sometimes combined phenomena. The approach,

now called Phenomenological Analysis of Repository Situations, has demonstrated an

unparalleled ability to describe repository operation.

Based on this analysis, new developments and improvements to the characteristics of the

structures and components were made. An overall architecture was developed as a working

basis to begin the initial industrial development phases. Once the location of the future

repository was known, more detailed drawings were produced, thus validating the overall

architecture comprising:

Surface nuclear facilities used for receiving, inspecting and conditioning waste, then

transferring packages underground via a funicular;

An approximately 4.2 km long ramp to transfer surface waste packages underground;

A surface mining facility, including access shafts to underground facilities;

An underground facility with a disposal area for ILW-LL, and a disposal area for

high-level vitrified waste.

In 2010, this overview of the main options was confirmed. Based on these main options, the

design phase began, particularly with the preparation of the preliminary design. A first draft

was submitted for public debate in 2013. It was used as the basis for later discussions with

local and regional representatives concerning the location of surface facilities. After public

debate, the location was decided.

Several possible zones were identified directly below surface facilities for mining activities.

Local representatives preferred wooded areas in order to avoid encroaching on farmland. For

nuclear facilities, the planned sector is located directly next to the underground laboratory,

straddling the border between the Meuse and Haute-Marne departments.

3. Launch of the design phase

The technical feasibility of the geological repository relied on simple, robust technical

concepts. Studies and research conducted since have explored avenues for optimisation and

provided more specific details for the basic options in order to develop a preliminary design

for a disposal facility.

The Cigeo geological repository must be able to hold a wide variety of waste packages,

particularly those generated from decades of research and development of industrial

processes. Packages will include cemented intermediate-level waste, bituminised waste, and

packages in various forms with different characteristics. To simplify operations, the various

packages were divided into types for which disposal packages had to be developed. System

standardization has been implemented via use of disposal containers.


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The inventory of waste to be disposed of in Cigeo includes 10,000 m3 of vitrified high-level

waste and 70,000 m3 of ILW-LL. The repository is therefore designed to be large enough to

hold this inventory, and operating facilities must be capable of handling the waste and

emplacing it in the repository.

The repository architecture groups together the disposal cells for different waste categories

within specific repository zones. ILW-LL and HLW repository zones will therefore be

physically separated from one another. This will ensure phenomenological independence

between each zone over the long term. Disposal zones will be built gradually in successive

phases, as new packages are received. They will therefore be designed in modules.

During the operation of the repository, surface facilities will manage waste packages before

they are transferred to underground disposal facilities. They will also support underground

operations. These facilities are designed to be decommissioned when the closure decision is

made.

4. Cigeo lifecycle phase and governance

The main, successive phases of the Cigeo project are as follows:

1. facility "design", including the technical specification of the facility structures,

buildings and procedures. This phase ends with the completion of detailed design and

the construction license application;

Subject to authorisation by decree (construction license):

2. “initial construction” of Cigeo when the first part of the facility is built. This includes

surface buildings associated with operation of the surface nuclear facility, surface-to-

bottom connections and underground structures to receive the first waste packages;

3. following issue of the operating license for Cigeo, “operation” by successive phases

over around one hundred years with package acceptance and disposal carried out in

parallel to underground facility extension work, in order to continue acceptance of

packages in the inventory. Partial closure work (moving to Stages 3 and 4 on the

International Retrievability Scale) is also carried out in addition to construction,

adaptation and regeneration work on surface buildings;

4. the “pilot industrial phase” planned for the launch of Cigeo operation before the

switch to normal operation. This pilot industrial phase will include tests designed to

demonstrate the ability to remove waste packages disposed of in Cigeo under real

conditions;

after operation has finished, the decommissioning and final closure of Cigeo, which can only

be authorised by the passing of an Act of Parliament. Cigeo then enters its “monitoring

phase”.

Construction and operation will be gradually developed in line with the forecasts for waste

package delivery.

5. Gradual development

Pursuing the process of creating a deep geological disposal facility is an ethical obligation for

our generation as important as ensuring that coming generations are able to reconsider any

decisions taken. In both instances, it is about not committing these generations to the choices

we make or fail to make. It is our generation and the previous one which built nuclear power

plants and enjoyed the benefits in terms of development and lifestyle. We must therefore bear

the investment cost for managing the waste produced. The technology and financial resources


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required to carry out the first stages of Cigeo development are now available. Nuclear power

plants are still in operation and will continue to support the funding of future investment

phases in the medium term.

By gradually implementing Cigeo, it is possible both to prepare for disposal of the HLW that

produces the most heat and to avoid any time gaps in waste management throughout the

Cigeo operation period. It should be noted that the very first vitrified waste packages

produced in the 1970s will be sent for initial highly instrumented disposal, in order to prepare

for the highly exothermic vitrified waste packages from 2080.

6. Reversibility and tools

The ethical concern for reversibility comes from the time scale required for managing the

most harmful radioactive waste. Particularly given the planned duration of approximately 120

years for the geological disposal facility operation, it is our generation’s responsibility to

design and provide future generations with a safe facility that they will be able to modify or

improve in accordance with their own objectives and requirements, or even replace by other

management facilities if other choices become available, particularly due to technical

advances. The reversibility of disposal is considered to be the ability to leave the next

generation choices concerning the long-term management of radioactive waste, including the

choice of reconsidering the decisions made by the previous generation.

In practice, reversibility is based on governance tools and technical project management tools

Governance tools: continuous improvement of understanding of radioactive waste

management, transparency and passing down of information and knowledge, the involvement

of society and checks by the government and assessment bodies.

Project management tools: incremental development and gradual approach to the construction

of Cigeo facilities, flexible operation, adaptability of facilities and retrievability of packages.

These tools support decision-making for radioactive waste management. In particular, they

ensure that the various choices available are preserved or unlocked over time.

With this new understanding of operation, retrievability is simply a technical possibility given

to the following generations so that they can implement their own options. To this end, our

responsibility is to provide facilities that are designed from the offset to be able to reconsider

our choices at a later time if required. As well as passing down high-quality options, we are

offering the necessary funds for their implementation. However, future generations will have

to bear the cost of any changes in direction.

7. Conclusion

The vision of the Cigeo project had long remained fairly static. It had been about creating an

overview with the aim of carrying out phenomenological studies and many safety analyses in

the long term. These steps have been completed, in particular between the promulgation of

French Acts of 1991 and 2010. As the industrial phase approaches, the vision is becoming

increasingly dynamic, incorporating designers in the disposal lifecycle. Disposal operation

will be carried out very gradually, in the frame of pilot industrial phase starting by trials in the

mid-2020s and a completion of commissioning in the mid-2030s.


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IMPACT OF STORAGE PERIOD ON SAFE GEOLOGICAL DISPOSAL OF SPENT

FUEL

B.B. Acar1, H.O. Zabunoğlu

2

1Turkish Atomic Energy Authority, Ankara, Turkey

2Department of Nuclear Engineering, Hacettepe University, Ankara,Turkey


Abstract. Geological disposal is the widely accepted method for safe final disposal of spent fuel (SF) and high

level waste (HLW). Currently, there are no active deep geological repositories. However, various geological

disposal projects are under way in many countries. In geological disposal, canisters containing SF/HLW are

simply placed into boreholes in a geological formation deep underground, specifically selected for final disposal

of nuclear wastes. The main factor affecting the geological repository design is the amount of waste that can be

safely emplaced per unit area of the repository (waste disposal density) and it strongly depends on the

characteristics (amount, isotopic composition, heat generation rate etc.) of the waste. The isotopic composition

and heat generation rate of SF discharged from reactor change during storage. This study aims to assess the

effect of interim storage period on disposal density of SF in a geological repository. In the first part of the study,

utilizing the code Monteburns, relevant compositions and decay heats of SFs discharged from a reference PWR

(A 1000-MWe PWR loaded with 3.3 w/o enriched UO2 fuel, with a discharge burn up of 33000 MWd/tU and

with an irradiation time of 1000 days) are obtained for selected cooling times. Then, using the code ANSYS,

thermal analyses are performed for a reference repository concept and disposal areas needed for SFs with

different ages are determined by ensuring that thermal criteria limiting the canister surface temperature is

satisfied. Results of the analysis are used to assess the effect of storage period of SF on disposal layout and to

derive the correlation between storage period and safe disposal capacity of geological repository.

Key Words: Spent fuel, geological disposal, storage, disposal density.

1. Introduction

Heat dissipation from a radioactive waste is one of the most important factors in geological

repository design and it depends on the waste type and composition. Waste composition is a

function of enrichment and burnup of the fuel, reactor power and cooling time of waste.

Disposal density calculations have two major parts: (1) determination of compositions and

decay heat profiles of wastes and (2) determination of disposal area through thermal analysis.

2. Determination of Characteristics of Spent Fuels

In this part of the study, isotopic compositions and decay heat profiles of SFs with different

storage periods are evaluated for a reference PWR by using Monteburns code. Monteburns is

a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive

decay and burnup code Origen2. Monteburns produces a large number of criticality and

burnup results based on various material feed/removal specifications, power(s), and time

intervals. In this study, inputs are prepared with reference reactor technical data and by unit

cell approximation.

2.1.Reference Reactor

A 1000-MWe PWR loaded with 3.3 w/o enriched UO2 fuel, with discharge burnup of 33000

MWd/tU and with an irradiation time of 1000 days is taken as the reference. SF discharged in



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the reference case consists of about 95.5 w/o U, 1 w/o Pu, 3.5 w/o fission products and other

actinides. The U in SF contains around 0.85 w/o U-235. About 70 w/o of Pu in SF is

composed of fissile isotopes (~59 w/o Pu-239 and ~11 w/o Pu-241).

2.2.Decay Heat Profiles

Decay heat profile of SF is obtained from Monteburns output for 106

years decay period. This

decay heat profile is shown in Figure 1 and is used as source term in thermal analysis.

FIG. 1.Decay heat of SF with 33000 MWd/tHM burnup

In order to obtain heat generation rate equations for SFs with different storage periods (40, 50,

60, 80 and 100 years) which are to be used as heat source terms in thermal analyses, time

dependent decay heat curve is fitted to sum of four exponential terms «Put’ s formula»[1]:

Where Q is decay heat in W/tHM; t is time in year elapsed since the production of the SF.

Values of the coefficients to be used in Put’s formula are given in Table I.

TABLE I: VALUES OF THE COEFFICIENTS IN PUT'S FORMULA

A1 A2 A3 A4 b1 b2 b3 b4

990.18 120.73 14.27 11.60 0.02325 0.00166 0.00013 3.1375E-5

3. Disposal Density Calculations

Once SFs disposed in the repository, temperatures of the repository components increase due

to the heat generation. Temperature affects many processes occurring in the repository, thus,

during the repository design, it is necessary to determine an appropriate density of

emplacement of heat-generating wastes and investigate the resultant time-dependent

temperature distributions.

3.1.Reference Repository Concept

The KBS-3 concept developed by Swedish Nuclear Fuel and Waste Company is taken as the

reference repository. In the reference disposal concept, SF is placed into copper canisters with

a cast iron insert. The canisters are surrounded by bentonite buffer and placed vertically into

holes in parallel tunnels at a depth of 500 m in granite rock. The depth of hole for SF canister

is 7.55 and the diameter of hole is 1.75 meters. Tunnel diameter is 5.5 meters. The distance

between the tunnels is 40 meters [2]. Four SF assemblies would be packaged within a copper

canister. Each SF assembly has a square cross-section 0.214 m by 0.214 m and 4.1 m long.

tb

i iieAtQ


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Disposal canister is 4.5 m long and 0.9 m in diameter [2]. Figure 2 shows reference repository

concept.

FIG. 2.Reference repository concept [2] and SF disposal canister

3.2.Thermal Analysis

Once disposal canisters are disposed in the repository, a transient heat diffusion phenomenon

gives rise because of the heat generated in disposal canisters. Heat transfer in the repository is

mainly by conduction. ANSYS finite element code is used to develop a 3-D thermal model of

the repository. It is assumed that the repository contains infinite number of tunnels filled with

infinite number of canisters with the same thermal output. Due to the geometrical and loading

symmetry of the repository, thermal model is simplified to one quarter of a deposition hole

with three symmetry surfaces. Vertical symmetry planes passing through the center of the

holes, half distance between the adjacent holes and half distance between the adjacent tunnels

constitute the lateral boundaries of the model. Figure 3 shows ANSYS model of repository.

FIG. 3.ANSYS model of repository

Constant temperature boundary conditions are applied at the top and bottom boundaries of the

model. All symmetric boundaries are assumed to be adiabatic. The heat-source term is applied

as volumetric heat generation in the waste region. Thermal analyses are performed for various

spacing values and the minimum distance between boreholes is determined with reference to

the thermal constraint. The thermal constraint is that the temperature at the canister surface

must not exceed 100 ºC. Bentonite will remain chemically intact for more than one million

years as long as the temperature does not exceed 100 ºC [3]. In this study, the temperature

limit is reduced to 80 ºC, in order to include a margin of 10 ºC to cover for natural deviations

in environmental parameters and another 10 ºC to cover the risk of occurrence of an air gap


84

between the canister and the buffer [4]. Figure 4 shows temperature as a function of time on

the canister surface and at the interface between bentonite and rock at minimum canister

spacing.

FIG. 4.Temperature as a function of time on the canister surface at minimum spacing

3.3.Disposal Densities

Disposal area needed to safely dispose one ton of SF (cooled for 40, 50, 60, 80 and 100 years)

in the reference repository is calculated from the minimum distance between boreholes,

distance between tunnels, and amount of waste loaded into a canister. Results are given in

Table II. TABLE II: DISPOSAL DENSITIES FOR SFs WITH DIFFERENT AGES

4. Conclusion

As seen in Table II, on the basis of per unit mass in the form ready to be buried in the

repository, SF with 40 years storage period requires the greatest disposal area. For longer

storage periods, disposal area required decreases significantly.

REFERENCES

[1] PUT, M., HENRION, P., “Modeling of Radionuclide Migration and Heat Transport

from an HLW-Repository in Boom Clay”, EC, Report EUR 14156 (1992).

[2] NIREX LTD., Outline Design for a Reference Repository Concept for UK High Level

Waste/Spent Fuel, Number:502644 (2005).

[3] CHOI, H.J., CHOI, J., “Double-layered buffer to enhance the thermal performance in a

high-level radioactive waste disposal system”, Nuclear Engineering and Design 238

(2008), 2815–2820.

[4] SWEDISH NUCLEAR FUEL AND WASTE MANAGEMENT COMPANY, Heat

Propagation in and around the Deep Repository Technical Report TR-99-02 (1999).

Cooling time

(years)

Canister spacing

(m)

Disposal area per canister

(m2/canister)

Disposal area

(m2/ ton)

40 5.2 208 107.63

50 3.9 156 80.73

60 3.0 120 62.10

80 2.3 92 47.61

100 1.6 64 33.12

disposal high level waste including spent nuclear fuel

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