disposal of very low level waste & low level waste · id 65. disposal of very low level waste...

International Conference on the Safety of Radioactive waste Management

SESSION 3b

Disposal of

Very Low Level Waste &

Low Level Waste

Session 3b – VLLW IAEA-CN-242

2

ORAL PRESENTATIONS

No. ID Presenter Title of Paper Page

03b – 01 65 S. Viršek

Slovenia

Safety Case for Slovenian Low & Intermediate

Level Waste (LILW) Near Surface Repository

4

03b – 02 103 M. Ranft

Germany

Morsleben Disposal Facility for Low and

Intermediate Level Radioactive Waste

8

03b – 03 98 A. Sakai

Japan

Disposal project for LLW and VLLW

Generated from Research Facilities in Japan: a

Feasibility Study for the Near Surface Disposal

of VLLW that Includes Uranium

12

03b – 04 137 J. Dose

Germany

The Asse II Mine – Tasks and Challenges 16

03b – 05 144 L. Griffault -

Sellinger

(S. Soulet)

France

The Safety Case of Andra’s Low and

Intermediate Level, Short Lived Radioactive

Waste Disposal Facility in the Aube District

(CSA)

21

03b – 06 169 A. Bagheri

Iran, Islamic

Republic of

Preliminary Post-Closure Safety Assessment

and Pre-disposal Radiomonitoring of Anarak

Near Surface Repository

26

03b – 07 190 R. Abu Eid

United States of

America

The Safety Case and the Risk-Informed

Performance-Based Approach for Management

of US Commercial Low Level Radioactive

Waste (LLRW)

30


3

POSTER PRESENTATIONS

No. ID Presenter Title of Paper Page

03b – 08 30 E.J. Seo

Korea, Republic of

Regulatory Activities and Lessons Learned in

Korea for a Low and Intermediate Level Waste

(LILW) Repository

34

03b – 09 37 D. Grigaliuniene

Lithuania

Waste Zone Conceptual Model Effect on

Predicted Radionuclide Flux from Near Surface

Repository

38

03b – 10 42 I. Kock

Germany

Multi-Phase Flow in a Complex Low Level

Waste (LLW) / Intermediate Level Waste

(ILW) Repository

42

03b – 11 43 A.M. Amin

Egypt

Safe Handling of Radioactive Animal Carcasses

Waste; Disposal Options

46

03b – 12 52 K. Tanaka

Japan

A Plan and its Safety Assessment of Very Low

Level Waste (VLLW) Disposal Site in order to

Dispose of Waste Materials Generated from

Decommissioning of Tokai Nuclear Power

Plant

51

03b – 13 120 A.H. Che

Kamaruddin

Malaysia

Site Selection Study for Radioactive Waste

Repository: Study Area of Negeri Sembilan

55

03b – 14 124 S. Sarkar

Australia

Regulatory Approach for the Assessment of the

Licence Application for Radioactive Waste

Management Facilities In Australia

62

03b – 15 157 M. Boroumandi

Iran, Islamic

Republic of

Simulation and Stability Analysis of Near

Surface Disposal Trenches of Radioactive

Waste by Using Finite Element Method

66

03b – 16 175 A. Ibrahim

Nigeria

Design of a Near Surface Disposal Facility for

Low and Intermediate Level Radioactive Waste

in Zaria, Nigeria

70

03b – 17 183 T. Von Berlepsch

Germany

The National Disposal Facility for Radioactive

Waste in Bulgaria

75


4

03b – 01 / ID 65. Disposal of Very Low Level Waste & Low Level Waste

SAFETY CASE FOR SLOVENIAN LILW NEAR-SURFACE REPOSITORY

S. Viršek, J. Špiler, T. Žagar

ARAO – Slovenian organisation for radwaste management, Ljubljana, Slovenia

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

Abstract. Even tough Slovenia is a small nuclear country, we have to take care of the radioactive waste

generated on its territory. In 2003, a decision was made to start the combined siting process (now popularly

called the “Nordic approach”) to find a site for the LILW repository by using an open approach and inviting

local communities to volunteer locations. Siting was finished in 2009, when the Slovenian government approved

both the site and the disposal concept that was developed during the siting.

The concept takes into consideration the properties of the site, the amount of waste and the need for modular

repository construction. The modular construction concept that was developed can be implemented in two

scenarios: a national solution for Slovenia alone or a joint solution with Croatia according to the bilateral

agreement.

The concept is called “Engineered and natural near-surface multi-barrier disposal facility”. It links together the

properties of both surface and underground repositories. The disposal silo will be constructed from the surface,

and the disposal of the final package units will also be done from the surface with the help of a portal crane.

After the silo is filled up, it will be sealed shut and placed between 55 and 15 m below the surface in a saturated

zone.

The safety of the repository has been developed based on internationally recognized LILW disposal principles.

The multi-barrier and multiple safety function principles have been introduced to the disposal concept.

All the waste meeting the waste acceptance criteria (WAC) will be packed in concrete containers and sealed

with mortar. Containers will be placed in the disposal silo and empty spaces will be filled with concrete. Once

the silo is be full it will be covered with a concrete slab and protected with a thick layer of clay from the shallow

aquifer that lies near the surface. The silo - disposal unit will be placed in layers of silt that have very low water

permeability.

In the past years, a couple of safety assessment iterations regarding the combination of the disposal concept and

the site were made, and the results show that the repository’s impact on the biosphere is negligible.

In the first part of the paper, the siting process is presented along with the site properties and the disposal

concept. The following parts include the safety assessment and the presentation of results, while the final part

provides the latest information about the project development.

Key Words: Slovenian LILW repository, near-surface disposal concept, safety case

1. Introduction

This paper presents the safety case development for the Slovenian near-surface LILW

disposal facility. Slovenia started the combined siting procedure (including not only technical

parameters but also public involvement) in 2003 to find a site for a LILW disposal facility for

the LILW generated on its territory. During the siting, the disposal concept, the safety

assessment and the preliminary design were developed, and in 2009, the Vrbina/Krško site

and the near-surface disposal concept were approved by the Slovenian government [1]. The

siting was followed by the licensing phase, which included the development of the final

design, another iteration of the safety assessment and all other necessary documentation to

obtain the construction permit.

mailto:[email protected]


5

2. Siting and the Vrbina/Krško site

The siting process, which started in 2003, involved all Slovenian municipalities. They were

asked to cooperate and propose areas for the construction of the LILW disposal facility.

ARAO (Slovenian organisation for radwaste management) received 8 positive replies, and

screened the proposed areas taking into account different parameters. Three most promising

sites were submitted to the Slovenian government for approval [2]. For each of them, a site

characterization program was prepared. After the approval, one of the communities with a

potential site withdrew from the procedure, and the second one proposed a new site. Because

of that work on the remaining site continued, and procedures began to evaluate the

additionally proposed site. For the first site, which was part of the process from the beginning

(approval), a feasibility study was prepared [3] in which different disposal concepts were

compared and evaluated. In 2009, the Vrbina/Krško site was approved along with the

“Engineered and natural near-surface multi-barrier disposal concept” [1].

The Vrbina/Krško site is located on gravelly lowland, approximately 300 m east from the

Krško Nuclear Power Plant (NPP) in the Krško municipality, and 2.5 km from the town of

Krško (north-west of the site). The site lies on a field, which is part of a plain along the Sava

River. The distance between the repository site and the river is approximately 700 m.

Geologically, this belongs to the Krško basin, which is a Neogene syncline that includes

Quaternary layers [4]. The entire central part of the larger Vrbina site area is covered by

Holocene Quaternary sediments from the Sava River. On the site itself, this layer is around

10 m thick. Below that lies the Miocene silt, which is the host rock formation for the disposal

facility. The thickness of this formation is more than 500 m.

In the Quaternary sediments, we can find an aquifer with a saturated thickness of 5 m on the

site and hydraulic conductivity ranges from 10-4

to 10-2

m/s. Bellow that lies a Miocene

aquiclude that comprises interchanging silty and sandy silt layers. The aquiclude was

classified according to International Association of Hydrogeologists (IAH) standards as

geological layers without significant groundwater sources and the hydraulic conductivity

ranges from 10-9

to 10-7

m/s.

FIG. 1: Schematically presented geological profile through the disposal site in the north-south

direction. (Q – Quaternary, Pl,Q – Plio Quaternary, M – Miocene)

disposal silo

disposal site

low permeable

Miocene silt


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3. Disposal concept

The properties of the site (a shallow aquifer on the top of low permeable layers, vicinity of

the town, seismic properties of the area, the Krško NPP being jointly owned by two

companies – one from Slovenia and one from the neighbouring Croatia – meaning that the

waste from the NPP is joint responsibility) are reflected in the disposal concept that was

developed for the site [5]. It is called “Engineered and natural near-surface multi-barrier

disposal concept” and is a combination of both surface and geological disposal facilities. All

the facilities important for the nuclear safety on the site will be constructed on the

embankment (around 2 m high) that will protect against the PMF (Probable Maximum

Flood). All the waste meeting the WAC for disposal will be placed into concrete containers

with outer dimensions of 1.95 x 1.95 x 3.30 m and sealed with cement mortar. The reason for

the height (3.30 m) of the container is special over pack that is used in the NPP storage

facility. The weight of the container prepared for the disposal will be up to 40 t, and the

container will be placed by crane into the silo (99 containers in 10 layers). The inner diameter

of the silo will be around 27 m with the primary and secondary lining with a total thickness of

2.2 m. The silo will be excavated from the surface with the help of a diaphragm wall. The

empty space between the containers and the silo will be filled with concrete. Once the silo is

filled up it will be covered with a concrete slab and placed between 55 and 15 m below the

surface. The closed silo will be covered with a clay layer almost to the surface. During the

operation, a drainage system will be installed inside the silo in order to collect potential

percolating water. A building will be constructed above the silo to protect against

precipitation and other weather conditions. One disposal unit will be enough to dispose of the

Slovenian part of the waste. If Slovenia and Croatia find a common disposal solution

according to the bilateral agreement [6], an additional silo will be constructed to increase the

capacity of the repository.

FIG. 2: Slovenian LILW disposal concept


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4. Safety assessment

The safety assessment was performed already in the siting phase to help develop the concept

suitable for the potential site. In the last iteration, the assessment [7] was used to support

licensing action and to provide the input for design optimization. It includes operational and

post-closure safety assessment, and demonstrates that the disposal facility will be able to

comply with regulatory performance objectives in Slovenia’s radiological safety regulations

[8,9]. Furthermore, sensitivity and uncertainty analyses were performed using both

deterministic and probabilistic approaches. The safety assessment has shown that the

proposed facility meets the regulatory safety criteria with a good margin in all the analyses

conducted. This conclusion is contingent on a number of basic assumptions that form the

foundation of the performed safety assessment analyses. Another purpose of safety

assessments is to provide input for the development of WAC.

5. Project status and conclusion

Slovenian LILW disposal facility project is in the first half of the licensing phase. All the

documentation needed for the environmental impact assessment is in the final preparation

stage, and procedure for obtaining the environmental protection consent will be performed in

2017. The documentation for the construction permit (final design, safety assessment, safety

report, etc.) is also under preparation and the construction permit is planned for 2018.

Slovenia now has both the site and the concept for the LILW repository while all reports

show that the impact of the planned facility on the environment will be negligible.

REFERENCES

[1] Governmental Decree on Detailed Plan of National Importance for low and

intermediate level radioactive waste repository on location Vrbina, municipality

Krško, Off. Gaz. of the RS, 114/2009.

[2] ARAO, Prefeasibility study to identify three potential sites for the LILW disposal

facility, T-2134-3/2. 2005.

[3] Acer, Feasibility study for LILW Vrbina - Krško disposal facility, rev. 1, NSRAO –

Vrb.ŠV/ŠV01/06. .

[4] Main site characterization of the Vrbina Krško site for the LILW disposal facility,

rev.1. IRGO Consulting d.o.o., GeoZS, NLZOH Maribor, Geoinženiring d.o.o., ZAG.,

2015.

[5] Draft of Vrbina Krško LILW disposal facility final design, rev. C, NRVB---5X1025.

IBE d.d, 2016.

[6] Treaty between the Gov. of the Republic of Slov. and the Gov. of the Republic of Cro.

on the regulation of the status and other legal relations regarding investment,

exploitation and decommissioning of the Krško NPP. BHRNEK, Off. Gaz. of the RS,

23/2003.

[7] Safety Analysis and Waste Acceptance Criteria Preparation for Low and Intermediate

Level Waste Repository in Slovenia – General overview of Safety Assessment Report.

ARAO (ENCO, INTERA, STUDSVIK, FACILIA, IRGO), 2012.

[8] Rules On Radioactive Waste And Spent Fuel Management (JV7). Official Gazette of

the Republic of Slovenia, No. 49/2006. Prepared in February 2011.

[9] Rules On Radiation And Nuclear Safety Factors (JV5). Official Gazette of the

Republic of Slovenia, No. 92/2009. 2011.


8


MORSLEBEN DISPOSAL FACILITY FOR LOW- AND INTERMEDIATE-LEVEL

RADIOACTIVE WASTE

M. Ranft, J. Wollrath

Federal Office for Radiation Protection (BfS), Salzgitter, Germany


Abstract. The former Morsleben radioactive waste disposal facility in Saxony-Anhalt, near Helmstedt,

Germany, is located in a salt formation. The 525-m-deep Shaft Bartensleben connects 4 main mining levels and

the 520-m-deep Shaft Marie connects two main levels. Due to rock salt and potash production, many cavities

exist in this former mine with dimensions of up to 100 m in length, 30 m in width and in height. The total

volume amounts to about 8,000,000 m3 of underground cavities. By the end of the operational phase in 1998, a

total waste volume of about 37,000 m3 with a total activity of approx. 9.3•10

13 Bq (as of 2014) had been

disposed of. The licensing procedure for the closure of the ERAM was initiated in 1992 and the respective

documents were finally submitted to the licensing authority in 2009. A public hearing took place in 2011.

During the long-lasting licensing and yet not finished procedure risks have been realised which have led to

important set-backs and have caused a new management of the project.

Key Words: LLW/ILW-disposal, licensing procedure, set-backs

[1] Introduction

The former Morsleben radioactive waste disposal facility (ERAM) in Saxony-Anhalt, near

Helmstedt, Germany is located in a salt formation. The 525-m-deep Shaft Bartensleben

connects 4 main mining levels between 386 m and 596 m b.g.s. and the 520-m-deep Shaft

Marie connects two main levels. Due to rock salt and potash production, many cavities exist

in this former mine with dimensions of up to 100 m in length, 30 m in width and in height.

The total volume amounts to about 8,000,000 m3 of underground cavities.

In 1971 the operation of the ERAM for predominantly short-lived low-level radioactive waste

started. Different areas of the mine were used to dispose of the waste using different

techniques (dumping of solid waste and drums, stacking of drums and cylindrical concrete

containers, and in-situ solidification of liquid waste). In 1990 the Federal Office for Radiation

Protection (BfS) became the responsible operator of the disposal facility. By the end of the

operational phase in September 1998 a total waste volume of about 37,000 m3 with a total

activity of approx. 9.3•1013

Bq (as of 2014) had been disposed of.

[2] The Decommissioning Concept

The decommissioning of the ERAM disposal facility is based on a safety related concept for

the backfilling and sealing measures. The concept for the backfilling and sealing measures is

focused on preventing a potential brine intrusion which could not happen directly into the

disposal areas but might be possible into the other parts of the mine. Salt concrete will be

used as backfilling material to reduce the remaining volume of the mine openings to a wide

extent and to stabilise the geomechanical situation of the mine. Seals will be constructed in

the shafts and in drifts between the major disposal areas and the other openings of the mine.



9

[3] The Licensing Procedure

The licensing procedure of the closure of the ERAM has been initiated in 1992 and the

respective documents have been finally issued by BfS to the licensing authority in 2009. A

public hearing took place in 2011. Due to a recommendation of the German Nuclear Waste

Management Commission (ESK) issued in 2013 [1] the BfS has to update the Safety Case

documentation according to the development of the state-of-the-art.

In addition, during this long-lasting licensing procedure risks have been realised which have

led to important set-backs and have caused a new management of the project.

[4] Boundary Conditions for the Project Management

Primary objective in defining the goal of the project is the absolute primacy of quality (within

the meaning of fulfilling the safety licensing requirements of the necessary damage

precaution according to the state-of-the-art in science and technology). An aggravating

secondary condition is a limitation of the project duration already resulting from the

requirement of ensuring the feasibility of decommissioning. In the event that certain areas of

the ERAM are not backfilled in time, this threatens to result in the reduction and, possibly,

loss of the evidence of feasibility of decommissioning. The same goes for the avoidance of

further innovation leaps in science and technology. Here it is also presumed that the length of

the project may affect innovations in science and technology and that these may negatively

affect the decommissioning project.

According to the previous deadline, the licensed project “Decommissioning Plan” was to be

implemented from 2014 after it had been decided. On the basis of the ESK recommendations

of 31 January 2013 [1] for the further approach for the proof of post-closure safety in the

decommissioning project, the BfS developed a new time schedule. It provides for all

application documents being completed by 2028.

[5] New Requirements from Sub-Statutory Regulations and the State-of-the-Art in

Science and Technology

The Federal Ministry of the Interior (BMI) has implemented the “Safety Criteria for Disposal

in Deep Geological Formations” [2], but has not established a substitution despite the fact

that these safety criteria have been considered no longer representing the state-of-the art from

the 1990s on. Only on the basis of the “Safety Requirements Governing the Final Disposal of

Heat-Generating Radioactive Waste” from 30 September 2010 [3], which in the opinion of

the Federal Ministry for the Environment, Nature Conservation, Building and Nuclear Safety

(BMUB) only apply to HAW, did the BMUB task the ESK with defining a new state-of-the-

art in science and technology at the end of 2011. The consequences for the classification of

the documents and safety analyses are considerable, since now different protection goals

(0.1 mSv per year and 1 mSv per year) need to be taken into account for developments of the

disposal system with different probability. With regard to the protection goals, this leads to

the simplification of evidence; with respect to the classification, however, to a considerable

need for revision. This finally culminated in the ESK recommendation to submit a

comprehensive FEP list and revision of the scenario analysis based on it.

The long period covered by the plan-approval procedure in connection with the unfavourable

resource situation has resulted in changes in the state-of-the-art in science and technology

that also effect the verification management of damage precaution. So far the project could

not or only inadequately compensate these changes. The main points are:


10

Change of criteria for judging the geomechanical integrity of the salt barrier,

Necessity of addressing the concept of the “containment providing rock zone”

Approach in scenario analyses (FEP catalogue, scenario development),

2-phase-flux calculations (development of IT),

Protection goals and differentiation of disposal system development between expected

and less probable development,

Approach in geological and geotechnical modelling (3D instead of 2D),

Transport calculation taking into account groundwater density (3D instead of 2D).

[6] Risks Realised in Planning with Research Nature

One component of the verification management for safe decommissioning are technical

solutions representing a not yet tested state of technology or also state of science and

technology. These are in particular the elements of gallery seals of the decommissioning

concept. Due to existing uncertainties, the target for the design of the gallery seals was

selected very conservatively with 10-18

m2 for the permeability of the seals. Therefore, no

experiences were available for the performance of the entire decommissioning concept. This

ensured that the results of the consequence analyses were clearly below the protection goals,

despite of neglecting the processes conducive to the evidence (such as sorption).

With sealing structures erected on a trial basis, it has so far not been possible to furnish the

necessary evidence completely. On the one hand, based on the previously favoured

construction material “Sorel concret DBM2”, there is no accepted evidence for the use in

rock not capable of creep (anhydrite). Based on new findings, on the other hand, the

corrosion stability of the salt concrete seals in connection with a non-expected finding (crack

formation) having occurred in an in-situ test in rock salt, is not proven currently.

The in-situ structure, however, shows an integral permeability today which is even below the

targeted level of 10-18

m2. That means, the integral permeability of the in-situ structure

currently falls below the very conservatively selected limit. Furnishing the full evidence of

long-term operability of the sealing structures would additionally require evidence of absence

of cracks, in addition to evidence of sufficient corrosion stability. Further R&D work is

needed here. Due to the R&D character, associated risks require appropriate prevention

measures.

[7] Requirements of the Licensing Authority and its Experts Regarding Depth of

Evidence and Commitment of Examination Results

A basic challenge which is difficult to grasp consists of establishing and respecting a reliable

approach in the examination of documents relating to the procedure. Effective and detailed

legal or sub-legal specifications on this issue are at present not available. As federal

supervision, the BMUB did not make any decisions as to this matter, apart from the

recommendations of the German Commission on Radiological Protection (SSK) [4] and ESK

[1] published in 2010 and 2013 (after completion of the application documents). Nor have the

operator and the licensing authority made any explicit and sustainable agreements dealing

with this issue. Their decisions were always made with respect to specific events and were

then in most cases not permanent. On account of the lack of legal or sub-legal specification, it

is left to the licensing authority and its experts alone to determine how to “demonstrate that

the necessary precautions to prevent damage are taken”. If the licensing authority does not

make any own decisions in this regard, the freedom of design is with their expert only.


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[8] Available Resources (Funds, Personnel, Knowledge)

Another change of the factual boundary conditions concerns the available personnel

resources. This does not only refer to changes in the availability of resources but also their

quality and effectiveness. When the Department Nuclear Waste Management (SE) of the BfS

was reorganised in 2011 (from matrix organisation to line organisation for the disposal

facility projects), it was assumed for the project ERAM that the plan-approval decision was

made soon after the public participation (2009 - 2011) and that decommissioning would start.

Therefore, the project ERAM within BfS that is oriented towards decommissioning planning

and the licensing procedure, was granted only few personnel resources. For the processing of

tasks resulting from the technical risks that had have been realised and the changed boundary

conditions (state of science and technology) there are deficiencies in resources with regard to

both extent and content. The procedure having run for a very long time meanwhile, this

situation is aggravated by the fact that knowledge is lost and evaluations change due to new

persons responsible. In the medium term, further 50 % of senior experts will leave the project

due to age-related retirements. Therefore, to implement the decommissioning project

successfully, a reorganisation of the project and additional personnel resources are necessary.

REFRENCES

[1] GERMAN NUCLEAR WASTE MANAGEMENT COMMISSION (ESK), Long-Term

Safety Case for the Morsleben Repository for Radioactive Waste (ERAM), ESK

Recommendation (2013). (in German only)

[2] FEDERAL MINISTRY OF THE INTERIOR (BMI), Safety Criteria for Disposal in Deep

Geological Formations, Rdschr. des BMI vom 20. April 1983, RS AGK 3 – 515 790/2

(1983). (in German only)

[3] FEDERAL MINISTRY FOR THE ENVIRONMENT, NATURE CONSERVATION,

BUILDING AND NUCLEAR SAFETY (BMUB), Safety Requirements Governing the

Final Disposal of Heat-Generating Radioactive Waste, http://www.bmub.bund.de

/en/topics/nuclear-safety-radiological-protection/nuclear-safety/details-nuclear-

safety/artikel/safety-requirements-governing-the-final-disposal-of-heat-generating-

radioactive-waste/?tx_ttnews[backPid]=256&cHash=bbf5f172f9319eca690ad46518411

c1c (2010).

[4] GERMAN COMMISSION ON RADIOLOGICAL PROTECTION (SSK), Radiological

Requirements for the Long-Term safety of the Morsleben Repository for Radioactive

Waste (ERAM), SSK Recommendation, http://www.ssk.de/SharedDocs/

Beratungsergebnisse_E/2010/Radiologische_Anforderungen_Morsleben_ERAM.html?nn

=2876422 (2013).

http://www.bmub.bund.de/

http://www.ssk.de/SharedDocs/


12


DISPOSAL PROJECT FOR LLW AND VLLW GENERATED FROM RESEARCH

FACILITIES IN JAPAN: A FEASIBILITY STUDY FOR THE NEAR SURFACE

DISPOSAL OF VLLW THAT INCLUDES URANIUM

A. Sakai, M. Hasegawa, Y. Sakamoto, T. Nakatani

Japan Atomic Energy Agency (JAEA), Japan


Abstract. The radioactivity of uranium-bearing waste contaminated by refined uranium increases with the

production of its progeny on a long-term timescale. Therefore, the long-term safety concept of the near surface

disposal of uranium-bearing waste is very important. The Japan Atomic Energy Agency (JAEA) examines

safety of near surface disposal by controlling the average uranium radioactivity concentration in each section of

disposal facility and performing safety assessment for very conservative assumptions.

Key Words: Uranium-bearing waste, Near surface disposal, Very low level waste.

1. Objective

A near surface disposal facility for low-level radioactive waste (LLW) generated from

nuclear power plants is operating in Japan. However, the disposal of radioactive waste

generated from other nuclear facilities and radioisotope utilization facilities has not yet been

implemented. Therefore, the Japan Atomic Energy Agency (JAEA) was assigned the task of

implementing the near surface disposal of LLW and very-low-level radioactive waste

(VLLW) generated from research facilities and radioactive isotope users in Japan.

Accordingly, the JAEA has proceeded with activities focused on disposal.

The volume of VLLW is estimated to be 76000m3 according to the storage amount and

predictions of generation amount for next 50 years. VLLW will be disposed of in trench-type

facilities that do not have engineered barriers.

Approximately 25% of the volume of VLLW is uranium-bearing waste generated from

uranium utilization facilities (fuel fabrication facilities and uranium enrichment facilities,

etc.). Radioactivity of the refined uranium increases with the long-term production of its

progeny. Therefore, the calculated dose due to uranium and its progeny is estimated to reach

its peak value beyond 10,000 years. The peak dose is a relatively larger value. Determining

the probable conditions for a period of more than 10,000 years is difficult when carrying out

safety assessment of near surface disposal. This is because the surface is easily affected by

changes in the surrounding environment. Therefore, JAEA considered the safety concept for

near surface disposal of uranium-bearing waste taking into account the long-term conditions

at a disposal site.

2. Waste characteristics and disposal method

Low concentration uranium-bearing waste is estimated to account for the majority of the

expected total volume of the generated waste. Therefore, JAEA is considering to categorize

very low level uranium-bearing waste (VLL uranium-bearing waste) as VLLW and to

dispose of it in trench type disposal facilities. The assumed mean value of the radioactive

concentration of uranium in this waste is 10 Bq/g, which is ten times the clearance level for



13

radionuclides of natural origin and equal to the IAEA exemption level for U-238 in moderate

amounts of material. The maximum waste uranium concentration is assumed to be 100 Bq/g,

which is ten times the mean value.

JAEA plans to dispose of VLL uranium-bearing waste and VLLW generated from other

nuclear facilities together in the same trench. Therefore, 25% of the total volume of VLLW

will be VLL uranium-bearing waste.

3. Safety measures for disposal of VLLW including uranium

3.1. Radiation dose from uranium and its progeny at a long timeframe

Safety concept currently applied in Japan for near surface disposal is focused on short-lived

LLW. Therefore, the established safety assessment method is based on radioactive decay.

Dose criteria used in safety assessment of the disposal are 0.01 mSv/y for likely scenarios,

0.3 mSv/y for less likely scenarios, and 1 mSv/y for human intrusion scenarios and

unexpected natural event scenarios.

However, a safety assessment method for the near surface disposal of long-lived waste such

as uranium-bearing waste has not yet been determined. Therefore, the dose caused by trench

disposal of VLL uranium-bearing waste under generic conditions after a control period was

preliminarily calculated to discuss the safety of disposal. The preliminary calculation selected

exposure pathways and parameters from the representative calculation referred as basis of

regulation in Japan and the fundamental design of trench facilities.

The peak doses calculated from groundwater pathways related to utilization of contaminated

river water are sufficiently lower than 0.01 mSv/y. However, the peak dose calculated from a

residence scenario that includes the external exposure and the internal exposure from the

ingestion of crops grown in the disposal area where the waste layer and cover soil are mixed

by excavation of the site is much greater than the peak doses of groundwater pathways at a

long period of time. Figure 1 shows annual radiation dose as a function of time for residence

scenario. The peak dose of the likely condition from residence scenario is greater than 0.01

mSv/y beyond 10,000 years even though the dose is lower than 0.3 mSv/y. The inhalation

dose from radon-222 is not included in this calculation.

The scenario calculations assumed the disposal site retained its original shape before the site

was excavated. However, the disposal facility might lose its original form as a result of

erosion or disruption by natural events or human activities over a long period of time. The

precise prediction of topological change is almost impossible.

Therefore, the dose due to uranium and its progeny in the residence scenario was assessed

under very conservative assumptions where the waste layer lay below cover soil with the

thickness of 0.3 m and radioactivity of uranium and its progeny do not discharge from the

waste layer during assessment period. The thickness of cover soil is based on the usual

thickness of the additional cover soil when a residence is constructed on the ground including

waste, for example concrete, etc. The peak dose is reached at approximately 200,000 years

with a value lower than 0.3 mSv/y as shown in the conservative condition of Figure 1.


14

FIG.1 Evaluated dose due to uranium from a residence scenario that assumes a different site

situation.

IAEA [1] shows the dose criterion for a representative person resulting from a disposal

facility is 0.3mSv/y and a dose criterion for a human intrusion scenario is 1 – 20 mSv/y.

ICRP publication 122 [2] describes that exposure from non-design basis evolution in a

situation with no oversight over a long period of time in geological disposal would be treated

as an existing exposure situations. As shown in Figure 1, the calculated dose resulting from

conservatively stylized scenario for very long timeframes is lower than dose constraint or

reference level of existing exposure situation. Therefore, trench disposal of VLL uranium-

bearing waste is considered to be a feasible disposal option.

3.2. Control of the radioactive concentration of the waste layer

It is difficult to reliably predict the long-term conditions at a near surface disposal facility.

Therefore, JAEA is discussing measures to explain the long-term safety of disposal facilities.

ICRP publication 122 [2] describes principles and strategies of the protection from exposure

in situations with no oversight over long time periods of time in geological disposal. IAEA

[3] describes that material containing radionuclides of natural origin with a radioactivity

concentration of lower than 1Bq/g are managed under the existing exposure situation, and 1

Bq/g can be used as a clearance level for material containing radionuclides of natural origin.

JAEA is discussing a management method for uranium radioactivity concentrations in

trench facilities that takes into account the aforementioned information. Figure 2 shows the

management concept for a trench disposal facility. The disposal area of the trench is divided

into sections of a certain size. The arrangement of VLL uranium-bearing waste and VLLW is

controlled so that the average uranium radioactivity concentration in each section is lower

than 1 Bq/g. A section refers to a layer composed of VLL uranium-bearing waste, VLLW,

and soil fill. The surface area of a section takes into account the viewpoint of safety

assessment, for example, size of the floor space of a residence，etc.

This method assumes that future generations can choose a management method for a disposal

site based on the existing exposure situation, even in the extreme situation where a waste

layer was left on the surface without cover soil.

Average concentration of total uranium in waste is 10 Bq/g.

The composition condition is the enriched uranium containing 5 wt% U-235

Time of scenario occurrence (year)

An

nu

al R

adia

tion

Do

se (

Sv/y

)

1.E-02

1.E-01

1.E+00

1.E+01

1.E+02

1.E+03

1.E+00 1.E+01 1.E+02 1.E+03 1.E+04 1.E+05 1.E+06 1.E+07

Control period

(50 years)

Dose

constraint

Conservative condition

Likely condition


15

FIG.2 Management concept for the average uranium radioactivity concentration for each section of

waste layer.

4. Conclusion

The dose calculated from the trench disposal of VLLW that includes VLL uranium-bearing

waste is sufficiently low within several thousand years.

However, the dose increases because of production of uranium progeny for a period of

beyond 10,000 years. Measures to deal with this issue are discussed next.

First, dose assessment in the very conservative assumptions for long periods of time is

performed. The implementer confirms that the calculated dose does not exceed 0.3mSv/y,

which is the dose constraint, or 1 mSv/y which is lower reference level for existing exposure.

Second, the amount of uranium radioactivity disposed in a trench section is controlled to

reduce the average uranium concentration in the section to lower than 1 Bq/g. The result

considers the situation as an existing exposure situation. The possibility of a non-acceptable

exposure situation is reduced, even if an extreme long-term exposure situation is assumed.

These results show that it is possible to safely implement the trench disposal of VLL

uranium-bearing waste by adequately controlling the disposed radioactivity of uranium.

This feasible study is an example of safety measures for trench disposal of VLL uranium-

bearing waste by JAEA. The safety regulatory system of disposal in Japan will be discussed

in the future.

REFERENCES

[1] INTERNATIONAL ATOMIC ENERGY AGENCY, Disposal of Radioactive Waste,

Specific Safety Requirements No. SSR-5, Vienna (2011).

[2] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION, ICRP,

2013. Radiological Protection in Geological Disposal of Long-Lived Solid Radioactive

Waste, ICRP Publication 122, Ann. ICRP, 42(3).

[3] INTERNATIONAL ATOMIC ENERGY AGENCY, Radiation Protection and Safety

of Radiation Sources: International Basic Safety Standards, General Safety

Requirement Part 3, No. GSR Part 3, Vienna (2014).


16


THE ASSE II MINE – TASKS AND CHALLENGES

J. Dose, D. Laske, M. Mohlfeld, P.L. Wellmann

Federal Office for Radiation Protection (BfS), Salzgitter, Germany


Abstract. The Asse II salt mine near Wolfenbüttel (Germany) is an approximately 100-year-old potash and

salt mine. Salt production stopped on 31 March 1964. The decommissioned mine was bought by the federation

in 1965 and was used for the storage of low-level and intermediate-level radioactive waste. Emplacement

stopped in 1978 after the Atomic Energy Act (AtG) had been amended in 1976. Now, a nuclear law plan-

approval (licensing) procedure was required as a condition for radioactive waste disposal. The Federal Office

for Radiation Protection (BfS) was to take over the operatorship of the facility with effect of 1 January 2009.

Since 24 April 2013 the so-called “Lex Asse” (§57b AtG), the “Law on Speeding up the Retrieval of

Radioactive Waste and the Decommissioning of the Asse II Mine” has been effective. The new law is the legal

basis for the retrieval of the radioactive waste. After retrieving the radioactive waste according to § 57 b AtG,

decommissioning has to take place. Retrieving requires by law the immediate and parallel conduction of

retrieval measures.

Today, the Asse II mine faces two major problems: On the one hand, influent saline solutions enter the mine, on

the other hand the stability of the mine openings is at risk. Therefore, the BfS has developed actions for an

emergency plan for the Asse II mine. Parallel to the retrieval measures - to improve the mine’s stability and

protect the emplacement chambers as well as to minimize the consequences of potential flooding – the mine is

stabilized by backfilling remaining cavities with concrete. The emergency plan is maintained and updated on a

regular basis. For this purpose, an accompanying technical examination is carried out on the basis of

calculations; experts refer to an "analysis of consequences". With its examination, BfS aims to optimize the

developed actions for an emergency plan of the Asse II mine.

Key Words: Emergency Plan, § 57 b AtG (Atomic Energy Act), Retrieval Measures

1. Previous and Current Situation

The Asse II mine near Wolfenbüttel (Germany) was taken into operation for the production

of potash and rock salt at the beginning of the last century. The former operator used the Asse

II mine as a “research mine” for the disposal of low-level (LLW) and intermediate-level

radioactive waste (ILW)1. Between 1967 and 1987 about 47.000 m³ LLW and ILW in

different types of packaging have been stored. Today, the mine’s stability is at risk. Daily,

about 12 cubic metres of salt saturated groundwater flow into the mine. The Federal Office

for Radiation Protection (BfS) was to take over the operatorship of the facility with effect of

1 January 2009. The BfS has the task to operate the mine under nuclear law and to

decommission it without delay. Long-term safety which is required pursuant to nuclear law

can only be achieved by retrieval of the radioactive waste according to current knowledge. In

2013 the “Law on speeding up the Retrieval of Radioactive Waste and the Decommissioning

1 In Germany all kinds of radioactive waste have to be disposed of in deep geological repositories. Therefore,

there is made no difference in the characterisation between waste containing radionuclides with comparatively

short half-life. All kind of radioactive waste is divided in heat generating radioactive waste and radioactive

waste with negligible heat generation. The German classification could almost be integrated in IAEA’s waste

classification scheme [1]. In exceptional cases e.g. for the Asse II mine for historical reasons, the denomination

of low-level (LLW) and intermediate-level waste (ILW) is still used.


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of the Asse II Mine” (“Lex Asse”, § 57 b Atomic Energy Act (AtG) [2]) becomes effective.

The law provides that, among others, no plan-approval procedure has to be carried out for

retrieval and associated tasks. According to § 57 b AtG retrieval operations must stop if their

implementation cannot be justified for radiological or other safety-related risks for the

population or the staff.

Recently the organisational structure in the area of radiation protection and final disposal of

radioactive waste in Germany was rearranged [3], [4]. The following offices resp. companies

will perform the different tasks: The former BfS will concentrate on public tasks of radiation

protection, e.g. medical research, nuclear emergency management. A new founded office

within the remit of the Federal Ministry for the Environment, Nature conservation, Building

and Nuclear Safety (BMUB), called BfE (Bundesamt für kerntechnische Entsorgungs-

sicherheit, Federal Office for the Regulation of Nuclear Waste Management) will now

regulate the site selection process and support the BMUB in its activities pertaining to the

final disposal of radioactive waste. In a second step a new founded federal company BGE

(Bundesgesellschaft für Endlagerung, Federal Company for Disposal) will take over

operational tasks from BfS as well as tasks of the Asse GmbH and DBE mbH. This includes

the BGE will take over the operational tasks for the safe decommissioning of the Asse II

mine. BGE will be also responsible for operational tasks for the site selection process,

construction and operational phase of repositories. Since the rearrangement is currently in

institutional change, we use simplified in this article “BfS” for assignment of operators task

of the Asse II mine.

2. Precaution Measures for Stabilization

Precautionary measures, parallel to all planning activities for retrieval (see Chapter 3), need

to be taken for the event of an uncontrollable inflow of water – which cannot be ruled out –

and to stabilise the mine. The precautionary measures have been pursued since 2010.

Stabilization measures are to reduce the risk that it will no longer be possible to

decommission the Asse mine in an orderly manner. Among the stabilization measures are the

so-called filling of roof clefts as well as additional measures involving the backfilling of no

longer needed cavities (blind shafts, galleries etc.) in the mine with concrete.

Stabilization of the mine and emergency preparedness are prior conditions for the retrieval of

waste. To improve the mine’s stability and protect the emplacement chambers as well as to

minimise the consequences of potential flooding – the mine is stabilised by backfilling

remaining cavities with concrete [5]. Precautionary measures include the planning,

preparation and execution of measures for filling remaining cavities nearby the ILW cavity

8a/511, sealing and stabilizing the mine on the 775-m – 700-m level, such as sealing and

installation of geotechnical structures, measures to limit the generation of gas, backfilling

remaining cavities to reduce the convergence and the possible extensions of pollutants and

provision of necessary resources for complete backfilling or the production of building

materials, see [5] for more details. More than half of backfilling of the remaining cavities to

stabilize the south shoulder is finished.

At least since 1988, saturated saline solution flows into the Asse II mine. Since the mine

openings and the overburden continue to deform, it cannot be ruled out that the inflow of

water will increase to an extent where it is no longer controllable. In this case a structured

operation of the facility can no longer be guaranteed: this is called an emergency. In this case

the emergency measures will take place. These measures include the filling of the ILW cavity

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8a/511 with sorel concrete, the filling of the residual cavities containing LLW with brucite

cement, the backfilling and closure of the open shafts, the flooding of MgCl2 brine against the

influent unsaturated solution and if applicable under application of compressed air. Last step

is the withdrawal from the repository, see [5].

In order to manage the influent saline solutions a major part is stored intermediately in a

storage pond on the 490-m-level and pumped to the surface following radiological

examination and clearance. About 11.5 cubic metres of influent solutions are collected daily

in front of the former salt extraction chamber 3 on the 658-m-level. A minor part of solutions

(about 0.5 – 1.0 m²) is also collected on the 725-m and the 750-m-level near to the

emplacement chambers. It is either stored intermediately on site or used underground to

produce concrete. Influent solutions that were in contact with radioactive waste and are

collected on the 750-m-level must be treated as radioactive waste and be used or disposed of.

3. Planning of Retrieval

3.1.Trial Phase

To retrieve the radioactive wastes from the Asse II mine, uncertainties and gaps in knowledge

need to be eliminated for reliable planning. That is the only way to concretely approach the

technical implementation. Also necessary is the knowledge on the boundary conditions

during retrieval to provide safety conditions for the staff and general public. Significant

changes resulting from experiences gained over the past years of the trial phase led to an

optimisation of the exploration programme dealing with the emplacement chambers in 2015.

According to “Lex Asse“ a justification relating to single measures is no longer required.

Therefore the opening of the chambers and the recovering of the waste by way of trial are no

longer necessary steps in the trial phase. In order to gain relevant data for the planning of

retrieval, the exploration and testing in the trial phase at one chamber has been continued in

2016 with the drilling of a sixth borehole in order to investigate the emplacement chamber 7

at the 750-m level. In addition to emplacement chamber 7, emplacement chamber 12 on the

750-m-level will be examined in the scope of the trial phase. The “Lex Asse” facilitate

parallel planning of “Recovery Technique”, “Interim Storage Facility” and “Retrieval”

without waiting for the results from the trial phase. However, the parallel approach also

contains the risk of planning errors, if the trial phase does not lead to the anticipated results.

3.2.Recovery Technique

In order to recover radioactive waste from the emplacement chamber, special machines need

to be developed. In a first step an investigation over existing machines for recovery was

carried out. Building on that an examination regarding further development is actually done.

3.3.Design of Interim Storage Facility

A buffer storage facility, the conditioning plant and the interim storage facility are required

for repacking the recovered waste and to store it safely intermediately until it can be taken to

a repository. For these facilities the BfS needs to find a suitable location. After a balanced

assessment of the various factors, the BfS has come to the conclusion that precedence should

be given to interim storage facility sites in the near vicinity of the mine in order for them to

be directly connected to the mine premises. For implementing the site-selection procedure in

a transparent and objective procedure, the BfS defined selection criteria and first published

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them in a discussion paper in February 2012, which forms the basis for the discussion.

Results of the discussion are issued in the Criteria Report in January 2014. Once a suitable

site for the interim storage facility has been found, planning and construction of an interim

storage facility and conditioning plant is done. The ground surface and the building ground

will significantly influence the planning works. In the event that no suitable site can be found

in the vicinity of the mine premises, it will be necessary to carry out a national search

procedure.

3.4.Retrieval: Recovery Shaft and Infrastructure

A recovery shaft needs to be built and the necessary infrastructure at the surface and

underground in the direction of the new shaft needs to be established. The infrastructure

comprises all installations (e.g. shaft tower, mine shaft buildings, laboratories, security

installations) required for the handling of the waste, from recovery to the interim storage

facility. These steps begin as soon as the geological exploration of the determined shaft site

has concluded. With the help of the drilling the geology of the rock formations will be

explored up to a depth of around 800 metres. The first exploration drilling from the surface

started in June 2013. Two exploration drillings from the 574-m-level of the Asse-II mine

towards the site of the planes new recovery shaft (shaft Asse 5) finished in 2015. A third

borehole at the 574-m-level started in July 2016. In January 2016 an exploration drilling from

the 700-m-level started and reached a length of about 250 metres in February 2016. Due to

the results of drilling (brine in borehole) more exploration drillings at the 700-m-level are

scheduled. Should the exploratory drillings show positive results, this location could be the

site for the recovery shaft over which the waste will be recovered from the Asse mine.

The planning (plan of concept) for the retrieval of all LLW and ILW radioactive waste on the

750 and 725-m-level started in spring 2015. The bidding procedure for the retrieval concept

for ILW on the 511-m-level will start in autumn 2016.

4. Analysis of Consequences and Challenging Aspects

In case of the Asse II we can state that the deep geological disposal with an insufficient or

missing safety concept adopted in the 1960s has pressed a huge cost burden on our future

generations, see also [6]. BfS and experts developed a guideline aiming at reviewing and

optimisation of precaution measures. Long-term objective of the developed guideline is

performing a post-closure safety assessment. The guideline is structured in different work

packages and is used as a “living document”. First step in this guideline is performing

shortcoming analyses. They take already into account e.g. data which are necessary to update

and the interaction of geological and hydrogeological site characterization, ground

mechanics, numerical modelling of the release and transport. First of all, requirements of

current legislation have to be respected. In a second step the safety concept, decommissioning

concept with different measures is framed and updated, containing results of the shortcoming

analyses. In further steps scenario development, concepts for numerical modelling and the

uncertainty assessment (deterministic, probabilistic) will follow. Challenging aspects of these

examinations are the enormous amounts of interactions in regard to content (analysing and

updating site conditions) and structure of the whole project. The Asse II mine is a complex

project and, generally spoken, impacts of complex projects have to be analysed from a system

perspective.

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3. Conclusions and Outlook

This article shows tasks and challenges depending on the previous and current situation of the

Asse II mine. With its examination aiming at the analyses of consequences, BfS optimises the

developed actions for an emergency plan of the Asse II mine. BfS experience from the

conducted “analyses of consequences” for the Asse II mine emergency plan could support a

repository-operators framework for updating and improving technical examinations already

in the pre-operational and operational phase.

REFERENCES

[1] INTERNATIONAL ATOMIC ENERGY AGENCY: Classification of Radioactive

Waste, IAEA Safety Standards Series No. GSG 1, Wien, November 2009, http://www-

pub.iaea.org/MTCD/publications/PDF/Pub1419_web.pdf.

[2] ATOMIC ENERGY ACT, http://www.gesetze-im-internet.de/atg/.

[3] REPOSITORY SITE SELECTION ACT (Standortauswahlgesetz - StandAG),

http://www.gesetze-im-internet.de/standag/BJNR255310013.html.

[4] ACT REGULATING THE REORGANISATION OF TASKS AND

RESPONSIBILITIES IN THE AREA OF RADIATION PROTECTION AND FINAL

DISPOSAL (Gesetz zur Neuordnung der Organisationsstruktur im Bereich der

Endlagerung” vom 26.07.2016), http://www.gesetze-im-internet.de/aktuDienst.html.

[5] BfS: Notfallplanung für das Endlager Asse, 2010,

https://doris.bfs.de/jspui/bitstream/urn:nbn:de:0221-

2013070410956/1/BfS_2010_02_Notfallplanung_Asse.pdf .

[6] Ilg, P., Gabbert, S., Weikard, H.-P.: “Nuclear Waste Management under Approaching

Disaster. A Comparison of Decommissioning Strategies for the German Repository

Asse II”, DOI: 10.1111/risa.12648, Society for Risk Analyses, 2016.

http://www.gesetze-im-internet.de/atg/

http://www.gesetze-im-internet.de/standag/BJNR255310013.html

https://doris.bfs.de/jspui/bitstream/urn:nbn:de:0221-2013070410956/1/BfS_2010_02_Notfallplanung_Asse.pdf

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THE SAFETY CASE OF ANDRA’S LOW- AND INTERMEDIATE-LEVEL, SHORT-

LIVED RADIOACTIVE WASTE DISPOSAL FACILITY IN THE AUBE DISTRICT

(CSA)

S. Soulet, L. Griffault

French National Radioactive Waste Management Agency, Parc de la Croix Blanche, 92298

Châtenay-Malabry, France.


Abstract. The waste disposal facility in the Aube district (about 200 km on the east side of Paris) is designed

to receive low- and intermediate-level, short-lived radioactive waste (LILW-SL). Nonetheless, radioactive

elements with medium and long half-lives are often mixed with the low level waste, but in extremely limited

quantities and under strict control such that their presence respect long term safety criteria. The CSA was

commissioned in 1992. Once the authorized limit (one million cubic meters) has been reached, the CSA waste

disposal facility will be monitored for at least 300 years.

According to the French Act on Transparency and Security in the Nuclear Field, June 2006 [1], the licensee of a

basic nuclear installation has to carry out periodic safety review. Last safety report was issued in 2004. Next

CSA safety document was submitted to the French Nuclear Safety Authority in 2016. At the request of this

authority, the document “reexamines” whether or not the CSA conforms to regulations and reevaluates the

safety of the installation. The article aims at presenting how this “safety reexamination” was performed and

what are the major evolutions with respect to the 2004 safety report (2004 SR) [2].

The safety reexamination considers two life phases: 1) the operational phase, which extends from 1992 to about

2060 and the post closure phase, which starts at the end of the operational phase. For the operational phase,

safety demonstration relies upon the update of the risk analysis associated with the installation’s activities. The

approach allowed revisiting the dimensioning scenarios presented in the 2004 safety report. For the post-closure

phase, safety demonstration relies upon the improvements in the basic scientific understandings, the experience

gained in operating the installation, the consolidation of the safety functions, and international practices. It

allowed revisiting the scenarios taking into account an analysis of the risk and residual uncertainties at this

stage. Human intrusion scenarios will also be presented for post monitoring period.

Key Words: low- and intermediate-level, short-lived radioactive waste disposal facility in the Aube district

(CSA), 2016 safety reexamination.

1. Introduction

According to the French Act on Transparency and Security in the Nuclear Field, June 2006

[1], the licensee of a basic nuclear installation has to carry out periodic safety review. Last

CSA safety report was issued in 2004 [2] (2004 SR). Since, a draft decision text of the ASN

concerning the re-examination of safety has been issued [3]. Andra decided to take into

account this draft to define the safety re-examination goals of the CSA:

To conduct an examination of conformity of the installation.

To re-evaluate the safety of the installation for the operation and the post-closure

phases and analyse the impact of this safety re-examination on the safety reference

documents produced previously.



22

2. The CSA disposal site

The CSA is located at approximately 180 km in the southeast of Paris and in 50 km east of

the city of Troyes. It is implanted in the Aube department on the municipalities of Soulaines-

Dhuys and the Ville aux Bois. The CSA is situated in the north border of a forest, mainly

constituted by the forest of Soulaines-Dhuys (Figure 1). The site of the CSA was chosen

according to the Fundamental Safety Rule N°I.2 [6]. The choice concerned to a zone of

outcropping sedimentary rocks constituted by a semipermeable layer (Aptian white sands)

recovering a waterproof layer (the Aptian clays). The topography of the site corresponds to a

flat slope directed to the Noues d’Amance River draining all of the subterranean flows of the

zone.

Safety relies upon a confinement system isolating the waste for a sufficiently long period of time to ensure that the radioactivity in contact with humans no longer presents a health hazard due to radioactive decay. It relies upon three main barriers, the radioactive waste package, the reinforced concrete repository structure, and the geological medium. In addition, after closure a cap or cover (mainly formed of clay) will be placed over the structures and will recover the repository zone in order to limit the water infiltrations.

FIG.1. CSA LILW-SL waste disposal facility during operation (left) and after closure (schematic

illustration on the right)

3. The safety reexamination objectives of Andra’s LILW-SL waste disposal facility in

the Aube district (CSA)

According to the draft ASN decision text, the conformity examination identifies all the

regulation texts or acts applicable to the CSA, and the associated requirements in order to

check the conformity of the installation to those texts. It also identifies requirements specified

for the Important Element for the Protection (IEP) and verifies in operation documents if they

are respected. Conformity examination will also be analysed for the 2004 safety report in

order to list the differences and present the modalities and timeframe for their treatments.

The safety revaluation aims at appreciating the level of protection of the interests mentioned

in the article L 593-1 of the Environment code [4] and also:

Verifying the good application of the principle of in-depth defence (article 3.1 of the

February 7th, 2012 Order [5]), and verifying prevention measures taken for reduction

of the consequences.

Revaluating the safety margins and identifying the possible improvements (according

to the ALARA principle, better available techniques in acceptable economic

conditions ...).


23

The approach for safety reexamination during operation or post-closure phases aims at testing

the robustness of the disposal system considering extreme scenarios or situations. The post-

closure phase, consider for the safety re-examination the period of monitoring, fixed formally

by the RFS I.2 at 300 years after the closure [6], and then for the period up to 50 000 years

after the closure of the disposal (conventional choice taken to estimate the consequences of

long lived radionuclides) in coherence with the 2004 SR. The safety re-examination focused

on the following issues:

The re-evaluation of the waste inventory at closure of the disposal.

The updating of the safety functions which are applicable to the protection of the

interests in the broad sense.

The analysis of the consequence of the evolutions since the 2004 SR (evolution of the

scientific knowledge, and experience gained in operating the CSA) on risk analyses

and definition of scenarios.

The updating of the scenarios descriptions (relative to the 2004 SR) for quantitative

evaluation of the radiological and chemical toxics impacts, for both the operation and

post closure phases.

The re-evaluation of the impacts of the CSA on the protection of the interests. The

objective of this issue is to reevaluate the safety margins by considering extreme

situations in view of testing the robustness of the safety functions of the installation.

As such, these studies aim at highlighting any potential weak point during extreme

situations which would not have been considered in the design or in the previous

safety re-examinations.

4. The 2016 safety reexamination

With respect to the safety approach, the 2016 safety reexamination was performed on a list of scenarios. In application of the ASN draft decision text [2], scenarios were defined with particularly conservatives’ orientations in order to consider situations testing further the safety function of the components both for the operation and the post-closure phases.

The approach relies for both the operation and post-closure phases on a consolidated analysis of the risks and uncertainties raising the potential failures of the safety functions or dysfunctions of components assuring safety functions. They both aim at identifying events or processes which can affect one or more safety functions and consequently induce higher radiological and/or toxicological impacts on the man and its environment and thus potentially question the protection of the interests quoted in the article L. 593-1 of the code of the environment [4].

As exposed previously, safety analyses relied on the evolutions (scientific knowledge, material, disposed waste) having an impact on the safety. The scenarios from the 2004 SR have been modified according to those evolutions and re-evaluated in 2016 with additional scenarios identified on the basis of the consolidated analysis of risks and uncertainties.

Within the framework of the 2016 safety re-examination, focus was made on scenarios for which the evolutions since the 2004 SR have led to modify significantly their description (new situations at risk, acquisition of scientific knowledge, evolutions concerning the domain of the civil engineering) and their associated data (values of parameters..), and which can potentially have a different impact on the protection of the interests.

4.1.Safety reexamination during operation

Radiological evaluations are compared with protection objectives fixed for the man (worker, public or reference group) in terms of acceptable dose. Relative to the 2004 SR, for accidental situations, the radiological impact is now evaluated on the population living near the site, and it considers a more important number of waste packages.


24

From the consolidated risk analysis, the following list of scenarios has been considered for evaluation of the radiological consequences: i) Fall of waste package, and 2) Fire of waste package, iii) external event considering a plane crash, and iv) a seismic activity.

Results of the evaluations indicate that the radiological protection objectives are respected for all those scenarios although they may consider particularly penalizing hypotheses, especially for the dimensioning scenario corresponding to the fire of a transport of up to five package boxes.

4.2.Safety reexamination after closure

The evaluations are compared with objectives fixed for the public from hypothetical reference group in terms of acceptable dose. Protection objectives are defined according national and international references considering classification of scenarios and their qualitative likelihood according to the consolidated risk and uncertainty analysis. The following list of scenario has been considered:

The normal evolution scenario (NES) which represents the disposal as designed and

taking into account its evolution over time.

Altered evolution scenarios considering involuntary storage of degraded waste

package.

Altered evolution scenario considering a cumulus of events: ascending levels of the

water table, failure of the underground collector and alteration of the concrete

structures.

Altered evolution scenario considering another cumulus of events: local collapse of

the cover on one concrete repository structure, degradation of the top concrete layer of

that repository structure and a failure of the underground collector.

Altered evolution scenario considering the use of a well located at the edge of the

CSA repository during the monitoring phase by a hypothetical reference group.

This list is completed by conventional inadvertent human intrusion scenarios occurring after assumed loss of memory (taken at 300 years): i) construction of a road crossing the disposal, ii) construction of a residence on the disposal site after construction work, iii) Children games on waste rock from construction work on the disposal site and iv) use of a well located in the disposal site after the monitoring phase.

Two radiological inventories at repository closure were considered for the evaluations: the radiological inventory corresponding to the updated one from the 2004 SR based on exchanges between Andra and the waste producers and one more conservative corresponding to radiological capacities fixed by the technical prescriptions of the CSA. The safety model of the NES has been revisited upon the knowledge acquired since the 2004 SR, especially on concrete material and their evolution with time.

Results of the evaluations indicate that the radiological protection objectives are respected for all the scenarios. The well scenario after the monitoring phase appears the most penalizing scenario for the radiological impact (dose) but it is very unlikely. Radionuclides contributing to the impact are long lived radionuclides mixed with the waste but in limited quantities and strictly controlled. The results show that the presence of numerous waste packages having no or degraded performance does not question the safety of the CSA. As well, a faster degradation of other engineered components than planned does not question the safety of the CSA. In addition, results also indicate the role played by the cover in limiting infiltration of water.

REFERENCES

[1] Loi n°2006-686 du 13 juin 2006 modifiée relative à la transparence et à la sécurité en

matière nucléaire. Version consolidée au 12 Juillet 2014.


25

[2] Centre de Stockage de l’Aube (INB N°149). Rapport Définitif de Sûreté du CSFMA –

Année 2014. Rapport n°SURRPAEES040019.

[3] Lettre N°CODEP-DCN-2013-017854. Projet de décision de l’ASN relative au

réexamen de sûreté des installations nucléaires de base.

[4] Code de l’environnement. Partie Legislative. Livre V: Prévention des pollutions, des

risques et des nuisances. Titre iX: La sécurité nucléaire et les installations nucléaires de

base. Chapitre III: Installations nucléaires de base (Articles 593-1 à L593-38). (2015)

[5] Arrêté du 7 Février 2012 fixant les règles générales relatives aux installations

nucléaires de base. Ministère de l’écologie, du développement durable, des transports et

du logement (2012). Journal officiel de la République Française, n°08/02/2012.

[6] Règle N°I.2 (Révision 1) du 19 Juin 1984. Tome 1: Conception générale et principes

généraux applicables à l’ensemble de l’installation. Chapitre 2: Principes généraux de

conception et d’installation. Objet: Objectifs de sûreté et bases de conception pour les

centres de surface destinés au stockage à long terme des déchets radioactifs solides de

période courte ou moyenne et de faible ou moyenne activité massique. ASN (1984).

Publication n°RFS I.2.


26


PRELIMINARY POST CLOSURE SAFETY ASSESSMENT AND PRE-DISPOSAL

RADIOMONITORING OF ANARAK NEAR SURFACE REPOSITORY

S. Hasanlou, A. Bagheri, A. Taherian, M. Boroumandi, S. Moemenzadeh, H. Mohajerani

Iran Radioactive Waste Management Co. (IRWA), Atomic Energy Organization of Iran

(AEOI), Tehran, Iran


Abstract. Anarak disposal facility is the primary low and intermediate level radioactive waste disposal site

located Anarak District, in Nain County, Isfahan Province, Iran. This paper presents the Preliminary post closure

safety assessment report and pre-disposal radiomonitoring of Anarak Near Surface Repository in order to

determine levels and variability of radiological conditions prior to operation that is needed in the licensing

process for near surface disposal repository. The Preliminary post closure safety assessment has been performed

based on ISAM methodology recommended by IAEA and AMBER Code is used for simulation of each

scenario. Three scenarios have been selected, including water erosion, bath-tubbing and human intrusion. The

water erosion considered as a design scenario as for climate condition, types of cover and trench design. 1100

years after closure of the repository and in case of water erosion scenario the maximum total dose is less than

0.2 mSv y-1

for the representative person who is living near the repository. Furthermore, the maximum dose is

caused by 241

Am that is equal to 0.15 mSv y-1

. All the results showed that estimated doses of radionuclides in

each scenario are less than dose constraint established by Iran National Regulatory Authority. Periodically about

200 samples including foodstuff, feeding material, surface and ground water, soil and sediment, airborne

particulate, radon and external radiation were gathered and analyzed. By using TLDs, The maximum average

dose equivalent value measured was approximately 100 µSv month-1

. Gross alpha and beta activities were

measured in common food commodities, including animal products, meat, grain, vegetables and feeding

materials. The ambient radon concentration in the air was found to vary from 5.55 to 18.4±0.2 Bq m-3

. The

measured gamma absorbed dose rate in the air at 1 m above the ground ranged from 0.043 to 0.075 nGy h-1

with

an overall arithmetic mean of 0.068 nGy h-1

. The activity concentration of anthropogenic (90

Sr, 137

Cs) and

natural (238

U, 232

Th, 40

K) radionuclide were determined in 78 soil and sediment samples. Tritium activity, total

alpha/beta and gamma-ray spectrometry analysis has been performed in all drinking water, surface and

groundwater samples. In general, all results showed the background level of the natural and artificial

radionuclides before any operation in Anarak near surface disposal facility.

Key Words: Anarak repository, Safety assessment scenarios, Pre-disposal, Radiomonitoring.

1. Introduction

Environmental radioactivity measurements are necessary for determining the background

radiation level due to natural radioactivity sources of terrestrial and cosmic origin [1]. Pre-

disposal radiomonitoring provides a baseline for comparison with environmental data during

the operational phase and after decommissioning the facility. Background radiation is defined

in the standard as: “radiation from cosmic sources; naturally-occurring radioactive material,

including radon (except as a decay product of source or special nuclear material); and global

fallout as it exists in the environment from the testing of nuclear explosive devices or from

past nuclear accidents such as Chernobyl that contribute to background radiation and are not

under the control of the licensee [2].

Based on the national strategy of nuclear waste management, the use of near surface

repository for the land disposal of low and intermediate level radioactive waste in Iran is

considered. In line with the internationally agreed principles of radioactive waste

management, and the national Regulations on Radioactive Waste Management prepared by



27

INRA-AEOI2, the safety of this facility needs to be ensured during all stages of its lifetime,

including the post-closure period. In this direction, the radiological environmental monitoring

and Preliminary Post Closure Safety Assessment is done during pre-operational period of the

Anarak site. Anarak disposal facility is the primary low and intermediate level radioactive

waste disposal site located Anarak city. Anarak city is situated in the central part of Iran, in

Nain County, Isfahan Province. Agricultural activities are limited because of climate

conditions, lack of water resources, and water quality. The area has very low rainfall. The

annual precipitation rate is less than 100 mm. Annual mean evaporation rate is higher than

3000 mm. Annual prevailing wind direction is from ENE and SSE. This study aims to assess

the environmental radioactivity level of the Anarak site and surrounding region prior to

operation phase. In addition Preliminary Post Closure Safety Assessment is done in various

scenarios to estimate doses of radionuclides in comparison with dose constrain established by

Iran National Regulatory Authority. The results from this study are expected to serve as

baseline data of natural radioactivity level and will be useful in assessing public doses.

2. Materials and Methods

2.1.Preliminary Post Closure Safety Assessment: Software and Scenarios

The Preliminary post closure safety assessment has been performed based on ISAM

methodology recommended by IAEA and AMBER Code is used for simulation of each

scenario. Three scenarios have been selected, including water erosion, bath-tubbing and

human intrusion. The water erosion considered as a design scenario as for climate condition,

types of cover and trench design. In this scenario the water infiltrate from upper side of the

site to the trenches and dissolves 5 percent of the waste in it, passing through unsaturated

zone close to the trench, exits to the surface and by temporary rivers reaches to the manmade

pool next to the site. These events would repeat for other 5 percent of the waste in next year

that it means 10 percent of the waste has transferred to the biosphere area. Start of this event

is after end of the passive institutional control period.

2.2.Pre-Disposal Radiomonitoring: Program and Collection of Samples

Pre-disposal radiomonitoring is done by using grid and judgmental sampling patterns.

Periodically about 200 samples including foodstuff, feeding material, surface and ground

water, soil and sediment, airborne particulate, radon and external radiation were collected and

analysed. In Table I and Table II Key radionuclides and radiomonitoring program in different

media of Anarak site is shown, respectively. The nearest population centre to this facility is

Anarak city with about 1500 population. The monitoring area spans an area of 5000 km2. All

samples were taken from various points, including: Anarak site, Anarak city, Chah-Gorbe,

Ashin, Esmailan and Piyouk villages and agricultural field-Dagh-e-Sorkh (See FIG.1).

TABLE I: KEY RADIONUCLIDES IN PRE-DISPOSAL RADIOMONITORING PROGRAM

Radiation Sources Cosmic Naturally-Occurring Radioactive Material Global Fallout

Radionuclides 3H,

7Be

238U,

226Ra,

232Th,

40K,

222Rn

137Cs,

90Sr,

3H

2 Iran Nuclear Regulatory Authority- Atomic Energy Organization of Iran


28

TABLE II: PRE-DISPOSAL RADIOMONITORING PROGRAM OF ANARAK SITE

Sampling Media Sampling

Device

Sampling

Type

Analysis

Device

Sampling

Frequency

Air Particulates

High Volume

Sampler Judgmental

Low Level

Counter Quarterly

Radon RAD7 RAD7

Direct Radiation

TLDs Grid and

Judgment

TLD Reader Quarterly

RS-230

GR-135

RS-230

GR-135 Annually

Soil - Grid and

Judgmental

HPGe

LSC Annually

Sediment - Judgmental HPGe

LSC Annually

Water

Resources

Groundwater - Judgmental HPGe

LSC

Low Level

Counter

RAD7

Semi-

annually

Surface Water - Judgmental Seasonally

Biota

Samples

Foodstuff Animal

- Judgmental HPGe

Low Level

Counter

Annually Vegetable

Feeding Material - Judgmental Annually

FIG. 1. Pre-disposal radiomonitoring area, population centers, Anarak site, seasonal surface water

streams and the elevation of the studying zone

3. Results and Discussion

3.1.Preliminary Post Closure Safety Assessment: Results for Design Scenario Assuming the representative person who is living near the repository, 1100 years after closure and in

case of water erosion scenario the maximum total dose is less than 0.2 mSv y-1

. Furthermore, the

maximum dose is caused by 241

Am that is equal to 0.15 mSv y-1

.

3.2.Pre-Disposal Radiomonitoring: Results for Representative Environmental Samples


29

Seventy-eight surface soil and sediment samples at a depth of 0–10 cm range were collected

from the sampling area on grid and judgmental basis. Mean activity concentrations of 238

U, 226

Ra, 232

Th, 40

K and 137

Cs in the soil and sediment samples were about 34.12±1.39,

34.03±0.69, 35.14±3.46, 527.09±17.72 and 5.944±1.2 Bq Kg-1

, respectively. Two methods of

active and passive were applied in order to measure external background dose rate. By using

TLDs in 25 points, the maximum average dose equivalent value measured was approximately

100 µSv month-1

. The measured gamma absorbed dose rate in the air at 1 m above the ground

ranged from 0.043 to 0.075 nGy h-1

with an overall arithmetic mean of 0.068 nGy h-1

. The

ambient radon concentration in the air was found to vary from 5.55 to 18.4±0.2 Bq m-3

. Gross

alpha and beta activities in common food commodities are shown in Table III.

Table III. Activity level in common food commodities

Sample Milk Chicken Meat Feeding Material Grain Vegetable

Gross α Activity

(Bq/L or Bq/Kg) 7.5 <MDA 8.9 22.3 1.51 6.71

Gross β Activity

(Bq/L or Bq/Kg) 255.3 146.9 92 570.7 229.7 264.4

Activity level in drinking water, surface and groundwater samples was measured. Tritium

activity in all samples was below the MDA, The geometric mean for alpha and beta

radioactivity was 0.37 and 0.42 Bq/L, respectively. In addition gamma-ray spectrometry

analysis has been performed in all samples and activity level of 238

U, 226

Ra, 232

Th, 40

K and 137

Cs was below the MDA.

4. Conclusions

The results of design scenario demonstrate that the effect of surface water erosion scenario is

acceptable. The results suggest that doses would still be well below the typical acceptance

criteria, even with cautious assumptions likely to result in over-estimates of dose in surface

water erosion scenario.

The activity concentration levels of the natural and artificial radionuclides were determined in

the all samples collected from Anarak site and surrounding area using active and passive

device. All results showed the background level of the natural and artificial radionuclides

before any operation in Anarak Near Surface Disposal Facility.

5. Acknowledgements

This study was supported by the Operation Deputy Office, Dr. Ali Maleki, Mohammad Rostamnejad,

Mohsen Asadian and Bahman Soleymanzadeh. We wish to express our warm thanks to Asghar

Mohammadi, Fariba Hadian, Hamidreza Pakouyan, Nasrin Gourani, Fatemeh-mohammad

Hosseinpour, Mohammad-ali Mohammadi, Amin Hemati, Taha Khaje-naeini, Arash Amin, Ali

Iranpour and Morteza Sabeti.

REFERENCES

[1] UNSCEAR, Ionizing Radiation: Sources and Effects of Ionizing Radiation. United

Nations Scientific Committee on the Effects of Atomic Radiation, 1993 Report to the

General Assembly, with Scientific Annexes, United Nations Sales Publication

E.94.IX.2, United Nations, New York, 1993.

[2] National Council on Radiation Protection and Measurements, Design of Effective

Radiological Effluent Monitoring and Environmental Surveillance Programs, NCRP,

Report No. 169, Vienna (2010) 10-50.


30


THE SAFETY CASE AND THE RISK-INFORMED PERFORMANCE-BASED

APPROACH FOR MANAGEMENT OF US COMMERCIAL LOW-LEVEL

RADIOACTIVE WASTE (LLRW)

B. Abu-Eid, D. Esh, C. Grossman

Division of Decommissioning, Uranium Recovery and Waste Management Programs,

Office of Nuclear Materials Safety and Safeguards, US Nuclear Regulatory Commission

Washington DC 20555.


Abstract: This paper describes the US Nuclear Regulatory Commission (NRC) staff approach to safety

analysis and performance assessment for LLRW disposal. The paper presents a comparative assessment of the

approach with the IAEA safety case (e.g.; IAEA Safety Series #SSG-23; [1]) and implementation aspects being

developed in coordination with participating members, through IAEA - PRISM/PRISMA projects [2].

For the

past two decades, NRC staff developed and used comprehensive technical guidance, NUREG-1573 [3], on

performance assessment methodology for disposal of LLRW in support of 10 CFR Part 61 (NRC’s Licensing

Requirements for Land Disposal of Radioactive Waste). Currently, NRC is amending its regulations that govern

low-level radioactive waste disposal facilities to require new and revised site-specific technical analyses

(Federal Register /Vol. 80, No. 58 /Thursday, March 26, 2015; [4]). Such analyses would facilitate the

development of site-specific criteria for LLW disposal acceptance. NRC staff issued a draft guidance

(NUREG-2175; [5]) on conducting technical analyses (e.g. performance assessment, inadvertent intruder

assessment, assessment of the stability of a low-level waste disposal site, performance period analyses) to

demonstrate compliance with the performance objectives in the proposed amendment of 10 CFR Part 61. The

paper presents some aspects of NRC’s technical guidance in NUREG-1573 and the draft guidance in NUREG-

2175 and provides a brief comparison with IAEA SSG-23. In summary, we show alignment in the “Safety

Assessment” approach which is an essential constituent of the IAEA safety case. We identify overlaps with the

US NRC’s “Performance Assessment” approach, as well as with NRC’s “Risk Assessment” methodology. We

also note the important role of “Uncertainty Analysis” in the safety case and in supporting regulatory decision-

making. Although there are harmonies and many similarities, there are also some differences between the IAEA

and NRC safety case approaches. These differences are outlined and discussed briefly in the paper. Our

reviews and assessments indicate that the NRC guidance documents provide detailed technical discussion and

specific approaches to demonstrate compliance with the NRC’s proposed performance criteria required for site-

specific analysis at different analytical timeframes.

1. Introduction:

NRC regulations for shallow land disposal of LLW were promulgated in 1982 under Part 61 of

Title 10 of the U.S. Code of Federal Regulations (NRC’s Licensing Requirements for Land

Disposal of Radioactive Waste). These regulations were initially developed considering a

hypothetical reference disposal facility located within the United States. The NRC is currently

amending 10 CFR Part 61 to require new and revised site-specific technical analyses, to

permit the development of site-specific criteria for low-level radioactive waste (LLRW)

acceptance based on the results of these analyses, and to facilitate implementation and better

align the requirements with current health and safety standards. In summary, the new and

revised requirements specify: a). technical analyses for demonstrating compliance with the

public dose limits; b). technical analyses for demonstrating compliance with dose limits for

protection of the inadvertent intruder; c). requirements for development of site-specific waste

acceptance criteria; d). implementation of current dosimetry in the technical analyses; and e).



31

requirements for the “safety case” including the identification and description of defense-in-

depth protections.

2. US NRC Approaches to Risk and Performance Assessment for LLW Disposal and

Key Aspects of NRC Safety Case:

The NRC regulatory approach for ensuring the safety of LLW land disposal facilities is to

establish performance objectives that ensure protection of the general population from

releases of radioactivity (10 CFR 61.41); protection of individuals from inadvertent intrusion

(10 CFR 61.42); and stability of the disposal site after closure (10 CFR 61.44). The

performance objectives are demonstrated via technical analyses, including a performance

assessment, inadvertent intruder assessment, site stability analysis, and performance period

analysis, and compliance with technical requirements. A performance assessment (PA) is a

type of risk analysis that addresses (a) what can happen, (b) how likely it is to happen (e.g.;

including uncertainties), and (c) what are the resulting impacts (e.g.; consequences). The

requirements for a performance assessment are set forth in 10 CFR 61.13(a).

The “safety case” in 10 CFR 61.2 is defined broadly the as a “collection of information that

demonstrates the assessment of the safety of a land disposal facility.” This includes the

technical analyses discussed above as well supporting evidence and reasoning on the strength

and reliability of the technical analyses and the assumptions made therein and information on

defense-in-depth. The safety case also includes a description of the safety relevant aspects of

the site, the design of the facility, and the managerial control measures and regulatory

controls. Under 10 CFR 61.10, the information provided in a license application comprises

the key components of the safety case and must also include general and technical

information as required in the proposed rule. There are also requirements for other

information associated with institutional, financial, and monitoring activities. Thus, a safety

case for a shallow land disposal facility is envisaged by NRC to cover aspects of the

suitability of the site and the design, construction and operation of the facility, the assessment

of radiation risks and assurance of the adequacy and quality of all of the safety related work

associated with the disposal facility. The NRC staff provides detailed guidance on the

contents of a license application in NUREG-1200 [6]. The guidance in NUREG-1200 is

supplemented by detailed guidance on conducting the technical analyses in NUREG-1573

[3], which is complemented by new and revised guidance in NUREG-2175 [5]; see for

example Sections 2.0 through 6.0 of NUREG-2175. The guidance also provides acceptable

methods to identify and describe capabilities of defense-in-depth protections and develop

waste acceptance criteria.

Licensing decisions are based on whether there is reasonable assurance that the performance

objectives can be met. Defense-in-depth protections, such as siting, waste-forms,

radiological source-term, engineered features, and natural features of the disposal site,

combined with technical analyses and scientific judgment form the safety case for licensing a

LLW disposal facility. The insights derived from technical analyses include supporting

evidence and reasoning on the strength and reliability of the layers of defense relied upon in

the safety case. These insights provide input for making regulatory decisions. The licensee

must conclude that the safety case demonstrates that public health and safety will be

adequately protected from the disposal of LLW (including long-lived LLW). A clear case for

the safety of a disposal facility also serves to enhance the communication among

stakeholders.


32

The NRC staff recommends that licensees include a plain language description of the

following aspects in their safety case:

1) Strategy for Achieving Safe Disposal of Radioactive Waste: The safety strategy should

include an overall management strategy for the various activities required in the planning,

operation, and closure of a land disposal facility, including siting and characterization,

facility and disposal site design, development of the technical analyses, operations, waste

acceptance, environmental monitoring, and institutional control.

2) Description of the Disposal Site and Facility: The description of the disposal site and

facility should describe the relevant information and knowledge about the disposal system

and should provide the basis for the technical analyses.

3) Description of the Technical Analyses Demonstrating Performance Objectives: The

description of the technical analyses should summarize the performance assessment,

inadvertent intruder assessment, site stability analyses, performance period analyses, and

analyses of the protection of individuals during operations.

4) Strategy for Institutional Control of the Disposal Site: The institutional control strategy

should summarize the institutional information required by 10 CFR 61.14.

5) Description of Financial Qualifications of the Licensee: The description should

summarize the financial information required by 10 CFR 61.15.

6) Description of Other Information: Depending upon the nature of the wastes to be

disposed of and the design and proposed operation of the land disposal facility, the

description may need to summarize other information required by 10 CFR 61.16.

7) Safety Arguments: The safety arguments should draw together the key findings from the

technical analyses to highlight the main evidence, analyses, and arguments that quantify

and support the claim that the land disposal facility will ensure protection of public health

and safety.

The NRC staff envisions that the safety case for a land disposal facility would evolve over

time as new information is gained during the various phases of the facility’s development and

operation. Therefore, the NRC staff expects that the safety case will be updated as new

information that could significantly impact the safety of the facility is learned. Requirements

at 10 CFR 61.28(a) specify that the application for closure of a licensed land disposal facility

must include a final revision to the safety case that includes any updates to reflect final

inventory and closure plans. The extent of the final revisions to the safety case may vary

depending on the licensee’s operation and closure of the land disposal facility and the amount

of new information that is developed that could significantly impact safety of the facility.

3. Comparative Analysis of US NRC Approaches with IAEA-SSG-23 Safety Case:

From the IAEA perspective, the purpose of a safety case is to provide a sufficient level of

detail regarding the description of all safety relevant aspects of the site, the design of the

facility, and the managerial control measures and regulatory controls to inform the decision

whether to grant a license for the disposal of LLW and provide the public assurance that the

facility will be designed, constructed, operated, and closed safely (IAEA, [1]). NRC’s

requirements and guidance address the significant components of the safety case discussed in

IAEA SSG-23. Although not specifically addressed by the revision to 10 CFR Part 61


33

discussed herein, it is noted that communication with the public and stakeholders, as

discussed in SSG-23, is a basic practice in developing NRC regulations or key guidance

documents as well as in making licensing decisions. Further, because NRC guidance is

implementing guidance, the NRC guidance is more detailed in terms of addressing the long-

term considerations of site performance.

4. Summary& Conclusions

A comparison of NRC’s recent implementation of the safety case for LLW disposal with

IAEA’s Safety Guide SSG-23 shows harmony and consistency between the approaches.

Further, NRC’s regulatory requirements and guidance documents provide a comprehensive

analysis of safety functions and safety features of a disposal facility to satisfy the invoked

long-term site performance requirements.

REFERENCES

[1] IAEA Safety Standards Series No. SSG-23; “The Safety Case and Safety Assessment

for the Disposal of Radioactive Waste;” 2012.

[2] IAEA PRISM Project: PRISM: “Practical Illustration and Use of the Safety Case

Concept in the Management of Near-Surface Disposal;” http://www-

ns.iaea.org/projects/prism/; 2009.IAEA PRISMA Project: “Application of the Practical

Illustration and Use of the Safety Case Concept in the Management of Near-Surface

Disposal Project (PRISMA).” IAEA POC [email protected]; 2016.

[3] US NRC; “A Performance Assessment Methodology for Low-Level Radioactive Waste

Disposal Facilities: Recommendations of NRC's Performance Assessment Working

Group (NUREG-1573);” 2000.

[4] US Federal Register / Vol. 80, No. 58 / Thursday, March 26, 2015 / Proposed Rules;

16082. NUCLEAR REGULATORY COMMISSION; 10 CFR Parts 20 and 61; [NRC–

2011–0012; NRC–2015–0003]; RIN 3150–AI92; Low-Level Radioactive Waste

Disposal; Nuclear Regulatory Commission. Proposed Rule.

[5] US NRC; NUREG-2175: Guidance for Conducting Technical Analyses for 10 CFR

Part 61, Draft Report for Comment (NUREG-2175); March 2015.

[6] NRC, 1994. U.S. Nuclear Regulatory Commission, “Standard Review Plan for the

Review of a License Application for a Low-Level Radioactive Waste Disposal

Facility,” NUREG-1200, Rev. 3, Washington, DC, April 1994.

http://www-ns.iaea.org/projects/prism/

http://www-ns.iaea.org/projects/prism/



REGULATORY ACTIVITIES AND LESSONS LEARNED IN KOREA FOR A LILW

REPOSITORY

E. J. Seo, M. C. Song

Korea Institute of Nuclear Safety, Daejeon, Republic of Korea


Abstract. Korea's programs to develop a low and intermediate level radioactive waste(LILW) repository were

first launched in 1986, and about twenty-year effort, a site in Gyeongju was chosen in November 2005. The

operator of disposal facility, KORAD (KOrea RADioactive waste agency) submitted an application to the

national nuclear regulatory authority for the 1st stage license, underground cavern disposal type in January 2007

and the combined construction and operating license was issued in July 2008. After the review of follow-up

actions and implementation of pre-operational inspection during construction phase, the operation of the 1st

stage facility with a capacity of 100,000 drums is approved in December 2014. During operation phase of the

facility, as the regulatory activities, the periodic inspection and the disposal inspection are implemented to

confirm whether the structure, equipment and performance of disposal facility and operational activities are in

conformity with technical standards.

Key Words: LILW repository, Safety Review, Pre-operational, Periodic and Disposal

Inspection.

1. Introduction

Korea's programs to develop a low and intermediate level radioactive waste repository were

first launched in 1986, and about twenty-year effort, a site in Gyeongju was chosen in

November 2005. The operator of disposal facility, KORAD (KOrea RADioactive waste

agency) is responsible for construction and operation of the LILW repository (Wolsong

LILW Disposal Center, WLDC), which will have a final capacity of 800,000 drums in an

area of about 2,060,000 m2 after stepwise expansion.

In January 2007, KORAD submitted an application to the national nuclear regulatory

authority for the 1st stage license, underground cavern disposal type. The professional

regulatory agency (Korea Institute of Nuclear Safety, KINS) reviewed the license documents

and the national nuclear regulatory authority issued the combined construction and operating

license in July 2008. Based on the review results, it was recommended that the applicant,

after issuance of the license, implement follow-up actions (26 items) to address issues that

require safety demonstration or further confirmation to reduce uncertainty to be identified.

After the review of follow-up actions and implementation of pre-operational inspection

during construction phase, the operation of the 1st stage facility with a capacity of 100,000

drums is approved in December 2014.

2. Stepwise development of LILW repository

The 1st stage facility is in the operational phase through a stepwise development of the

repository from site selection to construction as shown in Figure 1.



35

FIG. 1. Development of the 1st stage disposal facility in Korea.

The regulatory process for LILW repository in Korea is stepwise development as described in

Figure 2.

FIG. 2. Regulatory Process for LILW repository in Korea.

2.1.License application

The KORAD conducted site surveys and environment surveys on the finally selected site and

submitted application for Construction Permit (CP) and Operation License (OL) of a LILW

disposal facility to the Nuclear Safety and Security Commission (NSSC) based on the survey

results in January 2007. The KINS conducted a safety review of the application attached with


36

10 documents including Radiological Environmental Report (RER), Safety Analysis Report

(SAR), and Quality Assurance Program (QAP). As a result of the review, it was concluded

that the application was in compliance with the standards for permit specified in the Nuclear

Safety Act (NSA), as technical standards for location, structure, component and performance

were complied with as well as the radiological impact resulting from operation and closure of

a disposal facility was in conformity with the standards for protection of public health and the

environment as specified by Enforcement Decree of the NSA. After deliberation and

resolution, the NSSC granted the permit to the KORAD on July 2008.

Based on the review results, it was recommended that the applicant, after issuance of CP and

OL, implement follow-up actions to address issues that require safety demonstration or

further confirmation to reduce uncertainty to be identified during the period of construction

and operation, and the KINS conduct a review of the results of implementation. The

implementation and review of follow-up actions is to reduce uncertainty over safety in the

long-term and to secure the objectiveness and transparency of safety of the disposal facility

based on the safety review reflecting site characteristics obtained in the process of

construction and operation of the disposal facility. By doing so, it is ultimately possible to

develop the Safety Case for the construction stage of the disposal facility which is in line with

international requirements including the IAEA SSR-5 [1] that stipulate establishment of

Safety Case for each development stage of a disposal facility.

2.2.Construction

The construction of the disposal facility started in August 2008 and as of June 2014, most of

the construction works including excavation for construction tunnel, operation tunnel, access

shaft, unloading tunnel and disposal storage (silo) and concrete lining have been completed.

The LILW disposal facility is divided into surface and underground facilities as shown in

Figure 1. Surface facilities consist of a receipt and storage building, radioactive waste

processing buildings, service buildings and other supporting buildings. Here, radioactive

waste is received from waste generators such as NPPs and verified to be consistent with the

waste acceptance criteria. On-site treatment or conditioning is done, if necessary.

Underground facilities include construction tunnel, operation tunnel, access shaft, unloading

tunnel, and disposal silos. At first, 6 silos will be constructed approximately 80-130 meters

below sea level to dispose of approximately 100,000 waste drums.

The KORAD should undergo pre-operational inspection in accordance with the NSA during

construction phase. The purpose of the pre-operational inspection is to check prior to

operation whether the construction of a disposal facility satisfies the related design and safety

requirements. The disposal facility, etc. should be deemed to have passed the inspection when

the construction work has been progressed according to the content of a permit given under

the NSA and when the structure, equipment and performance of the disposal facility, etc. is in

conformity with the technical standard set by the NSA. The pre-operational inspection by the

KINS started in September 2008. The pre-operational inspection of the LILW disposal

facility is conducted for the purpose of confirming the appropriateness of construction and

performance and operational readiness, which is composed of 4 steps: (1) inspection on

structure, (2) inspection on system installation, (3) inspection on system performance and (4)

inspection before operation.

After the review of follow-up actions and implementation of pre-operational inspection, the

operation of the 1st stage facility is approved in December 2014.


37

2.3.Operation

The 1st disposal facility is in the operational phase and about 4,900 drums were disposed of in

the facility in September 2016. During operation phase, as the regulatory activities, the

periodic inspection and the disposal inspection are implemented.

The periodic inspection is implemented annually to confirm whether the structure, equipment

and performance of disposal facility during operational phase and whether storage, treatment

and disposal of radioactive waste are in conformity with technical standards which is

composed of 27 items including structures, radioactive waste management system. And the

disposal inspection is implemented to confirm whether the disposal of radioactive waste is in

conformity with technical standards, which is composed of 3 items: radioactive waste

management environment, radioactive waste packages and disposal environment. As a result

of the disposal inspection, when the disposal of radioactive waste is found to be in conformity

with the standards, the disposal shall be deemed to be passed.

Also, the operator of disposal facility should re-evaluate and complement, if necessary, safety

conditions of a disposal facility based on experience and data obtained from operation of a

disposal facility and results of safety assessment.

3. Concluding remarks

The 1st stage facility of the LILW repository in Korea is in the operational phase through a

stepwise development. Additionally, there are still challenges caused by the 2nd

stage

development with a capacity of 125,000 drums as a combined disposal facility: development

of underground cavern disposal (the 1st stage facility) and engineered shallow land disposal

(the 2nd

stage facility) in the same site. Especially, there will be interference of pathways,

complex radiation exposure etc. The KORAD submitted application for CP and OL of the 2nd

facility in December 2015 and the safety review is underway by KINS.

The establishment of Safety Case at any step in the development of disposal system including

the 2nd

stage facility shall be improved reflecting the experiences from the 1st stage

development.

REFERENCES


IAEA-SSR-5, Vienna (2011).


38


WASTE ZONE CONCEPTUAL MODEL EFFECT ON PREDICTED RADIONUCLIDE

FLUX FROM NEAR SURFACE REPOSITORY

D. Grigaliuniene, P. Poskas, R. Kilda

Lithuanian Energy Institute (LEI), Nuclear Engineering Laboratory, Kaunas, Lithuania


Abstract. This paper presents an investigation into how a waste zone conceptualization approach might affect

the evaluated radionuclide concentration beneath a repository. The analysed system represents a concrete vault

of a near surface repository where two types of waste packages (concrete containers with cemented radioactive

waste) are disposed of. Three waste zone conceptual models of different levels of complexity were developed:

homogeneous, layered and detailed. The investigation revealed that the highest influence is on short-lived non-

sorbed radionuclides. The homogeneous conceptual model in this case is the most conservative. For long-lived

radionuclides, the developed conceptual models do not make a significant difference in the predicted

radionuclide release.

Key Words: near surface repository, radionuclide release, conceptual model effect.

1. Introduction

A radioactive waste repository is a complex system comprised of different waste packages

and engineered structures. When performing modelling of radionuclide releases from

radioactive waste repositories, the system has to be described in the form that allows the

system’s mathematical representation and quantitative estimations. This is achieved by

formulation of conceptual models where a number of assumptions and simplifications are

adopted. For the same system different conceptual models can be developed depending on the

purpose and resources available.

This paper presents investigation into how a waste zone conceptualization approach might

affect the evaluated radionuclide release from a near surface repository (NSR).

2. Methodology

The investigation into the role of the conceptual model in prediction of radionuclide release

from an NSR was performed by developing three waste zone conceptual models of different

levels of complexity: homogeneous, layered and detailed. Radionuclide leaching and transfer

with infiltrated water through the waste zone and bottom engineered barriers to the

unsaturated zone was modelled and the results were compared. As an indicator for the

comparison purposes, the radionuclide concentration just beneath the repository was selected.

3. System Description

The analyzed system comprises of an aboveground concrete vault of a near surface repository

where low and intermediated level short lived radioactive waste is disposed of. It is assumed

that preparatory layers of clay and concrete are arranged and the vaults are constructed on the

natural clayey soil. After the vault is loaded with waste, it is covered with a concrete slab and

a protective cap with a clay layer inside. Two types of concrete containers with different

dimensions and properties with cemented radioactive waste are considered to represent two



39

different waste streams. It is assumed that the waste packages are arranged in four layers and

backfilled with cement-based material. The first and the third layers are formed of one type of

the waste packages while the other type of the waste packages is placed in the second and the

fourth layers. Schematic representation of the analysed system is presented in FIG.1.

FIG. 1. Fragment of the cross section of the disposal vault: 1 – top engineered barriers, 2 – waste

packages, 3 – backfill, 4 – bottom engineered barriers, 5 – natural soil.

A number of radionuclides with different physical (half-life) and chemical (retention)

properties were considered in the analysis: short-lived non-sorbed (H-3), short-lived sorbed

(Sr-90, Cs-137), long-lived non-sorbed (C-14 organic), long-lived weakly-sorbed (Cl-36, I-

129) and long-lived strongly-sorbed (C-14 inorganic, Tc-99, Pu-239). Activity of each

radionuclide in the vault is set as 1 TBq.

Water flow rate through the vault depends on the performance of the engineered barriers. It is

assumed that the engineered barriers will be intact and minimize the infiltration of water into

the NSR within the period of 100 years (this corresponds to the institutional control period).

Then sudden degradation of the concrete barriers is assumed. After degradation of the

concrete, the most resistive barrier is the cap. Thus, amount of water entering the vaults

gradually increases following the degradation of the cap until a natural infiltration rate is

reached.

4. Conceptual Models

Homogeneous model. In the homogeneous model, the waste zone is modelled as the

homogeneous waste form-container-backfill mixture in a vault. The properties of the

homogenized waste zone were defined taking into account the properties of the waste

packages and backfill, and their occupied volume in the vault. It was assumed that the

radionuclides are homogeneously distributed in the waste zone.

Layered model. In the layered model, the waste zone is divided into layers to represent the

structure of the waste packages arrangement in the vault. However, the homogenization in

this case is not fully avoided as the layers with waste packages are represented as

homogeneous mixture of the waste form-container-backfill between containers in the same

layer. It was assumed that the radionuclides are homogeneously distributed in the layers with

the waste packages.


40

Detailed model. In the detailed model, such elements as waste forms, containers and backfill

are distinguished. The detailed model takes into account diffusion from the waste form to the

container walls in all three directions and distribution of flow between the materials with

different hydraulic properties.

The developed conceptual models were implemented in the computer tool AMBER [1] and

radionuclide concentration in the pore water beneath the repository was estimated.

5. Results and Discussion

Concentration of the radionuclides in the pore water of the natural soil beneath the vaults for

the non- or weakly-sorbed radionuclides is presented in FIG.2 and for the strongly-sorbed

radionuclides – in FIG.3. The maximal concentration of the radionuclides Sr-90 and Cs-137

in the pore water in all cases was less than 1E-05 % from the initial inventory and are omitted

in the figures.

It can be seen from FIG.2 that in the case of the homogeneous model, the non- or weakly-

sorbed radionuclides appear in the natural soil earlier compared to the layered and the

detailed models. This happens because in the homogeneous model, the radionuclides can be

transferred to the vault base and the natural soil already at the early time steps, while in the

other models the radionuclides should pass the non-contaminated bottom of the vault filled

with backfill at first and in the detailed model – the container walls as well. The most

significant delay is observed for the non-sorbed radionuclides H-3 and C-14org.

FIG. 2. Concentration of the non- or weakly-sorbed radionuclides beneath the vault: solid line –

homogeneous model, dotted line – layered model, dashed line – detailed model.

When comparing the maximal radionuclide concentration evaluated using the different waste

zone conceptual models, it can be seen that the largest difference is for the short-lived non-

sorbed radionuclide H-3. The maximal H-3 concentration estimated using the homogeneous

model is about 30 % higher than the concentration estimated using the layered model and

about 70 % higher than the concentration estimated using the detailed model. For the non- or

1.0E-05

1.0E-04

1.0E-03

1.0E-02

1.0E-01

1.0E+00

10 100 1000

Co

nce

ntr

ati

on

(%

fro

m t

he

init

ial

act

ivit

y/m

3)

Time after repository closure (years)

C-14org

Cl-36

H-3

I-129


41

weakly-sorbed long-lived radionuclides, the difference in the maximal concentration

estimated using different models is less than 10 %.

Similar results are obtained for the strongly-sorbed radionuclides, see FIG.3. In this case the

earliest appearance of the radionuclides in the natural soil is also observed for the

homogeneous model and selection of the waste zone conceptual model gives a rather small

difference (less than 10 %) in the evaluated maximal radionuclide concentration beneath the

vault.

FIG. 3. Concentration of the strongly-sorbed radionuclides beneath the vault: solid line –

homogeneous model, dotted line – layered model, dashed line – detailed model.

When analysing the obtained results, the assumptions related to the degradation of the

engineered barriers should be also pointed out. In the modelling it was assumed that the

concrete barriers remain intact for 100 y and then sudden degradation of the concrete occurs.

After the concrete degradation, the properties of the waste zone components are almost the

same. H-3 is the only radionuclide analyzed with the maximal concentration in the soil

beneath the vault reached before degradation of the barriers. For other radionuclides the

maximal concentration in the soil is reached after the concrete degradation, and the level of

detail of the conceptual model plays a minor role.

6. Conclusions

The investigation into how the developed waste zone conceptual model affects the predicted

radionuclide concentration beneath the repository revealed that the highest influence is for the

short-lived non-sorbed radionuclides. The homogeneous conceptual model in this case is the

most conservative. For the long-lived radionuclides, the developed conceptual models do not

make a significant difference in the predicted maximal radionuclide concentration.

REFERENCES

[1] QUANTISCI AND QUINTESSA. AMBER 4.4 Reference Guide, Version 1.0, Enviros

QuantiSci, Culham (2002).

1.0E-08

1.0E-07

1.0E-06

1.0E-05

1.0E-04

100 1000 10000 100000

Co

nce

ntr

ati

on

(%

fro

m t

he

init

ial

act

ivit

y/m

3)

Time after repository closure (years)

C-14inorg

Pu-239

Tc-99


42


MULTI-PHASE FLOW IN A COMPLEX LLW/ILW REPOSITORY

I. Kock, G. Frieling, M. Navarro, S. Hotzel

Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) gGmbH, Köln, Germany


Abstract. Our R&D project, financed by the German Federal Ministry for the Environment, Nature

Conservation, Building and Reactor Safety, was designed to assess to which degree the choice of the physical

model (single phase vs. multi-phase flow) and the choice of the geometrical model (simple vs. complex) is

significant for the simulation results of the repository system regarding

i. Fluid flow (e. g. gas and liquid fluxes),

ii. Fluid and radionuclide mixing (e. g. dilution),

iii. Radionuclide transport inside and potential escape out of the repository.

Instead of using a generic repository model, an existing repository for LLW/ILW, the “Endlager für radioaktive

Abfälle Morsleben (ERAM)” was chosen as the ideal example to examine these issues, in particular since our

project results could be compared to existing calculations.

The code we used for our calculations is a version of TOUGH2 enhanced by various processes (e. g. host rock

convergence, etc.) relevant in a rock salt based nuclear waste repository. For all enhancements strict quality

assurance procedures were carried out.

Unsurprisingly, fluid flow patterns differ strongly between each of the model whereas the difference in 14

C

release between the models is not quite as strong. Results indicate that especially initial two-phase fluid flow

parameters, e. g. the initial liquid saturation of the seals, are of importance to the repository system.

Key Words: Two-Phase Flow, ERAM, TOUGH2

1. Introduction

In recent years, two-phase fluid flow calculations have become common regarding the

disposal of radioactive waste in various host rocks [1]. The generation of gas and its

consequential effects e. g. the rise in pressure and the transport of contaminants out of the

repository are potential safety relevant processes.

Recently, we conducted a preliminary safety analysis [2] where numerical calculations

regarding the two-phase flow of gas and brine were carried out on repository scale [3,4]. For

the simulations, the code TOUGH2 [5], modified by GRS with several extensions relevant to

processes in a nuclear waste repository [6], was used. Beyond conducting a successful

analysis, many issues had been identified which suggest further development of the approach.

Especially the question if a more complex model layout is possible and applicable for a

complex mine was of interest. Moreover various extensions to the software, most notably

limiting gas generation to actual water content, were considered essential.

Consequently, our R&D project, financed by the German Federal Ministry for the

Environment, Nature Conservation, Building and Reactor Safety, was designed to assess to

which degree the choice of the physical model (single phase vs. multi-phase flow) and the

choice of the geometrical model (simple vs. complex) is significant for the simulation results

of the repository system regarding

i. Fluid flow (e. g. gas and liquid fluxes),

ii. Fluid and radionuclide mixing (e. g. dilution),



43

iii. Radionuclide transport inside and potential escape out of the repository.

We decided to select the LLW/ILW repository “Endlager für radioaktive Abfälle Morsleben

(ERAM)” as our case study. Currently, the application for decommissioning (e. g. [7]) is

being reviewed for this repository. For the application numerous calculations regarding the

long term safety had been conducted [8,9]. The fluid-flow calculations are single phase, using

a simplified, single floor geometrical model of the former salt mine’s multi floored structure.

FIG. 1. Geometric Structure of the ERAM (southern part Bartensleben). Based on [10].

2. Models, Simulations and Code

Our simulations are based on a modified version of TOUGH2 [6]. Currently, all

modifications are bundled in a quality-controlled (QC) [11] version now called TOUGH2-

GRS Version 01 [12]. For all calculations the same QC-version of TOUGH2 was used.

TOUGH2 is a two-phase fluid flow simulator, where in our case a gaseous phase and a liquid

phase are considered. Radionuclide transport in our simulations is limited to nuclides with a

half-life greater than 500 a. Both nuclides from decay chains and nuclides with induced

radioactivity are considered. Altogether 35 nuclides were taken into account where only 14

C

was relevant for transport via the gas phase (in CH4 or CO2).

To test for the effects of model complexity we build three meshes, all in 3D. The first mesh,

the so called “basic mesh” is fully based on the geometrical model of [8,9] and consists of

~10 blocks with a large volume and little space discretization. The repository areas with

highest activity (repository areas West, South and East) are sealed by geotechnical barriers

from the rest of the mine. Model complexity was then increased by incorporating actual depth

and vertical discretization in the basic mesh. The geometric structures in the “extended mesh”

therefore correspond to realistic depth of roof and floor of each horizon in the mine - but the

basic layout with 10 blocks of large volume is kept. For the third mesh realistic lengths and

depths are incorporated. This mesh shows the typical room structure as seen in FIG. 1. All

structures (rooms, tunnels and also geotechnical barriers) show realistic lengths and volumes.

For all meshes resulting volumes of repository areas and other areas are the same. Scenarios

and subsequent parameterization are also based on the information given in [8,9] with the

necessary exception of two-phase fluid flow parameters, e. g. capillary pressure functions or

initial liquid saturation in the repository. Therefore, in all simulation cases parameters are as

similar as possible, with the exception of necessary adjustments when using three different

meshes. Adjustments consequently are more abundant in the complex mesh.


44

3. Results

Two basic scenarios and therefore two reference calculation cases exist for each mesh: (1)

„dry, no brine entry“ and (2) „wet, significant brine entry“ into the repository. The numbers

of model runs with parameter variations differ slightly for each mesh, so that altogether about

750 deterministic model runs were conducted.

For the basic mesh, the “dry” cases show that gases can migrate relatively free inside the

repository as long as the necessary escape pressure of 3 MPa has not been reached. In these

“dry” cases the development of the sealed repository areas in terms of pressure (liquid and

gaseous) and transport (fluids and radionuclides) is almost fully independent of the rest of the

repository. When escape pressure is reached, transport paths change and gas (potentially

including 14

C) flows in the direction of a postulated escape location. In most cases however,

escape pressure is reached late, and 14

C output is accordingly low. In the “wet” cases the

development of the sealed repository areas plays a major role in the repository system. Due to

the chemical reactive brine, geotechnical seals corrode and the brine gets into contact with

the radioactive waste. Consequently, more radionuclides can be transported to the escape

location and output of 14

C is slightly higher as in the “dry” case. In the “wet” cases

radionuclides like 59

Ni or 99

Tc also reach the escape location.

In case of the extend mesh, the importance of the seals’ performance is emphasized. This is

true for both the “dry” and the “wet” model cases. The variation of (two-phase-flow)

parameters which have an impact on the seals initial permeability (e. g. gas entry pressure,

initial saturation) strongly influence fluid flow patterns out of the sealed areas and the

resulting radionuclide output. Interestingly, in the event of initial failure of one the seals in

the “dry” case, it seems to be of no consequence on which level (1 through 4) the failure

occurs.

For the complex mesh, “dry” and “wet” model cases reveal that the response of the model to

significant parameter variations is not very pronounced. Overall, the variability of the

simulations’ results is low, for example 14

C discharge out of the repository exhibits roughly

the same value (1.000 Bq/a) in 2/3 of all model cases. However, the fluid flow patterns for

the complex mesh differ strongly from the patterns observed in the basic and extended mesh.

4. Conclusions

We ran 750 model runs with 3 very different meshes. In our simulations, fluid flow patterns

differed strongly inside the repository. Discharge of relevant nuclides in all cases lies below

previously simulated output of single-phase simulations from [8,9].

REFERENCES

[1] S. Norris, “Synthesis Report: Updated Treatment of Gas Generation and Migration in

the Safety Case,” 2014.

[2] K. Fischer-Appelt, B. Baltes, D. Buhmann, J. Larue and J. Mönig, “Synthesebericht für

die VSG: Bericht zum Arbeitspaket 13,” Vorläufige Sicherheitsanalyse für den Standort

Gorleben, GRS-290, Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH,

Köln, 2013.

[3] J. Larue, B. Baltes, H. Fischer, G. Frieling, I. Kock, M. Navarro et al., “Radiologische

Konsequenzenanalyse: Bericht zum Arbeitspaket 10, Vorläufige Sicherheitsanalyse für

den Standort Gorleben,” GRS-289, Gesellschaft für Anlagen- und Reaktorsicherheit

(GRS) mbH, Köln, 2013.


45

[4] I. Kock, R. Eickemeier, G. Frieling, S. Heusermann, M. Knauth, W. Minkley et al.,

“Integritätsanalyse der geologischen Barriere: Bericht zum Arbeitspaket 9.1, Vorläufige

Sicherheitsanalyse für den Standort Gorleben,” GRS-286, Köln, 2012.

[5] K. Pruess, C. Oldenburg and G. Moridis, “TOUGH2 User's Guide, Version 2.0,”

Berkeley, California, USA, 1999, revised 2012.

[6] M. Navarro, “Handbuch zum Code TOUGH2-GRS.00a: Erweiterungen des Codes

TOUGH2 zur Simulation von Strömungs- und Transportprozessen in Endlagern,”

GRS-310, Köln, 2013.

[7] G. Resele, M. Ranft and J. Wollrath, “Endlager Morsleben - Nachweis der

radiologischen Langzeitsicherheit für das verschlossene und verfüllte Endlager: eine

Übersicht,” Salzgitter, 2009.

[8] D.-A. Becker, D. Buhmann, J. Mönig, U. Noseck, A. Rübel and S. Spiessl,

“Sicherheitsanalyse für das verfüllte und verschlossene Endlager mit dem

Programmpaket EMOS,” Braunschweig, 2009.

[9] M. Niemeyer, G. Resele, S. Wilhelm, J. Holocher, J. Poppei and R. Schwarz, “Endlager

Morsleben - Sicherheitsanalyse für das verfüllte und verschlossene Endlager mit dem

Programm PROSA.,” 2009.

[10] DBE TECHNOLOGY GmbH (DBETEC), “Hohlrauminformationssystem für das

Endlager für radioaktive Abfälle Morsleben,” 2014.

[11] M. Navarro, H. Seher, S. Hotzel and J. Eckel, “Quality Assurance for the TOUGH2

Family of Codes Using the Code SITA,” Proceedings of the TOUGH Symposium 2015,

Berkeley, California. 28.–30. September 2015, pp. 239–246.

[12] M. Navarro and J. Eckel, “TOUGH2-GRS Version 01 User Manual,” GRS-403, 2016.


46


SAFE HANDLING OF RADIOACTIVE ANIMAL CARCASSES WASTE;

DISPOSAL OPTIONS

A. El Kamash1, A. M. Amin

1, M. Abdel Geleel

2

1Labeled compound department, Hot Labs and Waste Management Centre, Atomic Energy

Authority, Egypt 2Nuclear Fuel Cycle Department, ENNRA, Egypt


Abstract. The aim of this work is to establish safe procedure and guides for the handling and disposal of

radioactive animal carcass. All carcasses and tissues that contain or is contaminated with radioactive materials

must be disposed of separately from regular biomedical waste in compliance with Egyptian laws and

regulations. This procedure is intended to ensure the proper and safe management of radioactive animal carcases

waste at Inshas long term storage facilities. Some animals are injected by Tc-99m (short half life ~ 6 hours) that

converted to Tc-99 (very long half life ~ 211000 y). Tc-99 causes harmful effect to the environment if

transferred to the ground water. RESRAD computer code used to calculate the effective dose received from uses

of ground water that contaminated by Tc-99. Different scenarios for the disposal options of contaminated animal

carcass were takes place to achieve ALARA principles.

1. Introduction

When radioisotopes are to be used in a biomedical facility, proper consideration should be

given to the design of the facility to ensure safe use of the material in accordance with the

requirements of the regulatory organizations. Such consideration should include planning for

processing, storage and disposal of all generated radioactive waste. Some radionuclides are

also used to label human blood components to act as tracers for sites of blood loss or sites of

infection. This typically involves removing a blood sample from the patient, radiolabelling

the blood and re-injection. The actual activity that may be re-injected is usually in the range

of a few MBq to a maximum of 200 MBq, with the highest activity typically used for 99m

Tc(1)

. It may be necessary to reassess the risk to human health following the ingestion of

the relevant isotopes, including Tc-99, because of the possibility of radiation induced

genomic instability, as well as the cancer risk. The long half-life of technetium-99 and its

ability to form an anionic species makes it a major concern when considering long-term

disposal of high-level radioactive waste.

2. RESRAD computer code

The input data for the near surface trench for dispose of contaminated carcasses at Inshas site

are:

Stratum thickness [h(1)]: 4.000000 m

Bulk soil material density [rhob(1)]: 1.500000 g/cm**3

Hydraulic conductivity [Khuz(1)]: 10.000000 m/yr

Saturation ratio [sruz(1)]: 0.802299


https://en.wikipedia.org/wiki/Anionic


47

TABLE 1: TRANSPORT TIME PARAMETERS FOR UNSATURATED ZONE STRATUM NO. 1

Radio- Distribution Retardation Transport

nuclide Coefficient Factor Time

(i) Kduz(i,1), cm**3/g Rduz(i,1) Dtuz(i,1), yr

Tc-99 0.0000E+00 1.0000E+00 1.2837E+00

Water table drop rate [vwt]: 0.001000 m/yr

Bulk soil material density [rhobaq]: 1.500000 g/cm**3

Effective porosity [peaq]: 0.200000

Hydraulic conductivity [Khaq]: 100.000000 m/yr

Soil specific b parameter [baq]: 5.300000

Saturation ratio [sruaq]: 0.677340

TABLE 2: TRANSPORT TIME PARAMETERS FOR UNSATURATED ZONE CREATED BY

THE FALLING WATER TABLE

Radio- Distribution Retardation Minimum

nuclide Coefficient Factor Transport Time

(i) Kdaq(i), cm**3/g Rduaq(i) Dtuaq(i), yr

Tc-99 0.0000E+00 1.0000E+00 3.4789E-04

Aquifer contamination depth at well (z): 2.50000E+01 m

Depth of water intake below water table (dw): 1.00000E+01 m

Infiltration rate (In): 5.00000E-01 m/yr

Distance below contaminated zone to water table (h): 0.40000E+01 m

Initial thickness of contaminated zone (T): 0.20000E+01 m

Effective porosity of saturated zone (pesz): 0.20000E+00

TABLE 3: DILUTION FACTOR AND RISE TIME PARAMETERS FOR NONDISPERSION (ND)

MODEL

Radio- Dilution Retardation Horizontal Transport Rise Decay Time

nuclide Factor Factor Time Onsite Time Parameter

(i) f(i) Rdsz(i) Tauh(i), yr dt(i), yr 1/lamda(i),yr

Tc-99 1.000E+00 1.000E+00 1.000E+01 4.000E+00 3.073E+05


48

3. Result and Discussion

3.1.Calculation of Doses from Exposure Pathways

Doses are resulted from potential exposure from contaminated aquifer and the canal. The

farmer who lives at the boundary of the disposal area is assumed to drink the well water, to

irrigate crops and feed animals with the contaminated well water, and the farmer also

consumes the contaminated crop and animals. In this calculation we consider the

consumption of ground and surface water as drinking two liter a day. The doses and risk as a

result of drinking water and other pathways are calculated using RESRAD computer code.

The contaminated carcasses are uses as source term in the code and the radionuclide release

and transfer to the aquifer at the water table are calculated, also, the radionuclide

concentration at various locations and time can be calculated. If the contaminated aquifer also

discharges into a surface water body, the flux of radionuclide into the surface water can be

calculated. If the surface water body is small flowing river, for example as Ismailia Canal, the

radionuclide concentration in the canal can be calculated.

Two scenarios are used to select the best disposal option for contaminated carcasses, one is

shallow land disposal and the other is near surface disposal.

TABLE-7- CONCENTRATION OF RADIONUCLIDES IN ENVIRONMENTAL MEDIA AT T =

0.0E+00 YEARS

Contaminat- Surface Air Par- Well Surface

ted Zone Soil* ticulate Water Water

RadioNuclide mBq/g mBq/g mBq/m**3 mBq/L mBq/L

Tc-99 1.000E+01 1.000E+01 1.693E-04 0.0E+00 0.0E+00

*The Surface Soil is the top layer of soil within the user specified mixing zone/depth.

3.2.Doses calculation from exposure to all pathways

The main exposure pathway adopted in this calculation is the ground water to man .

Assuming;

A 70 kg weight man drinks two liters of ground water from a well located 150 m distance

away from the disposal area at Inshas site;

The assumptions and the parameters used in these calculations depend on the:

1. Inshas site characteristics

2. The type of waste

3. The disposal type (shallow land disposal or engineering disposal)

4. The hydrology and the geohydrology of the disposal area

The input data and assumptions are:

Aquifer thickness: 5000E+01 m

Isotope name: Tc-99

Half-life: 212 E+06 yr


49

Retardation factor: 1230E+01

Initial inventory 10 mCi/g


50

4. Conclusion

It is clear from the radioactive dose assessment results by using RESRAD Computer code

that, the total exposure dose to the whole body as a result of drinking two liters of ground-

water or surface water (Ismailia Canal) is less than the limit; 25 millirem (0.25 mSv) (10 CFR

part 61) and IAEA GSR part 3.

REFERENCES

[1] INTERNATIONAL ATOMIC ENERGY AGENCY, Handling, Treatment, Conditioning

and Storage of Biological Radioactive Wastes, IAEA-TECDOC-775, Vienna (1994).

[2] K. Yoshihara, "Technetium in the Environment" in "Topics in Current Chemistry:

Technetium and Rhenium", vol. 176, K. Yoshihara and T. Omori (eds.), Springer-Verlag,

Berlin Heidelberg, 1996.

[3] Abd El-Aziz, M. A., “Site Assessment for disposal of low and intermediate levels

radioactive wastes Inshas area” Ph.D. thesis, Ain Shams university, 1996.


51


A PLAN AND ITS SAFETY ASSESSMENT OF VLLW DISPOSAL SITE IN ORDER

TO DISPOSE OF WASTE MATERIALS GENERATED FROM DECOMMISSIONING

OF TOKAI NUCLEAR POWER PLANT

K.Tanaka1,2

, H.Noguchi2, S.Nomura

2, H.Tanabe

2, K.Morii

2, K.Fujimura

2

1The Institute of Applied Energy (IAE) Tokyo, Japan

2The Japan Atomic Power Company (JAPC) Tokyo, Japan


Abstract. The Japan Atomic Power Company (JAPC) has planned to dispose of both metal wastes and

concrete wastes generated from decommissioning of Tokai nuclear power plant (TK1). TK1 is the first

commercial nuclear power plant in Japan that is a gas-cooling reactor and is now under decommissioning.

According to a law of Japan, the disposal site is categorized as a trench type of near surface disposal without

artificial constructions. We call such a disposal site L3 site in Japan. The site will be located in the northern part

of TK1 site. JAPC will construct two trenches named A-trench and B-trench. JAPC will dispose of metal wastes

in A-trench and concrete wastes in B-trench. We have performed safety assessments to ensure necessary

measures for maintaining the site safely. Scenarios relevant to public exposure in an operation phase and in a

post-closure phase have been considered to develop conceptual and mathematical models. In order to identify

scenarios, characteristics of meteorology, geography, hydrology, geohydrology and social environment around

the site were investigated. The assessments provide that the public exposure from the site is low enough for

limits of the requirements.

Key Words: VLLW, near surface disposal, safety assessment, decommissioning

1. Introduction

Tokai nuclear power plant (TK1), which is a graphite moderate and CO2 gas-cooling reactor

(GCR), is the first commercial nuclear power plan in Japan and is under decommissioning

[1]. The Japan Atomic Power Company (JAPC) who is an operator of TK1 has planed to

dispose of very low-level waste (VLLW) generated from decommissioning in TK1 site.

According to a law of Japan, this disposal site is categorized as a trench type of near surface

disposal without artificial constructions, which we call L3 in Japan [2].

JAPC has already submitted an application of the disposal site, which is called TK-L3, for

regulatory body. In tasks for preparing the application, we have assessed public exposure the

public exposure around TK-L3 through various pathways of various scenarios. The

assessments provided that the public exposure would be low enough both during the

operation and after a controlled period. It has met a legal limit of Japan.

2. A specification of TK-L3

TK-L3 will be located in the northern part of TK1 site as shown in FIG. 1. An area of the site

is 6.0×10+6

m2 and total amounts of waste are 1.6×10

+4 tons. JAPC will construct two

trenches named A-trench and B-trench. JAPC will dispose of metal wastes in A-trench and

concrete wastes in B-trench. Metal wastes will be mainly generated from dismantling of

Steam Rising Units (SRU) and from piping which connects a reactor and SRU. Surface of

SRU internals and inner surface of the piping are contaminated by radionuclides, which arise

from activations of corrosion products and from dispersion of fission products. Concrete



52

wastes are generated from dismantling of the primary Biological Shielding Wall (BSW) and

some parts of the secondary BSW. The concrete wastes are activated by neutron irradiation.

Radiological specification of waste materials to be disposed of is shown in TABLE 1.

In order to perform

observations about a

structure of groundwater-

flow and the stratum in

TK1 site, JAPC has dug

35 observation-wells for

boring investigations as

seen in FIG. 2. The

observation showed that

the groundwater always

flows toward the sea and

that the upper limit of the groundwater is around 5.6m in depth.

TABLE 1: RADIOLOGICAL SPECIFICATION OF WASTE MATERIALS

Radionuclide Upper limit of radioactivity concentration

(Bq/ton)

Total radioactivity

(Bq) Radionuclide

Upper limit of radioactivity concentration

(Bq/ton)

Total radioactivity

(Bq)

3H 3.0x10

9 1.4x1012 90

Sr 1.0x107 1.7x10

9

14C 5.0x10

7 1.2x1010 137

Cs 7.0x106 9.1x10

8

36Cl 1.0x10

8 4.6x1010 152

Eu 3.0x108 5.6x10

10

41Ca 2.0x10

7 3.4x109 154

Eu 9.0x106 2.5x10

9

60Co 8.0x10

9 1.3x10

11 α-nuclide 4.0x10

6 1.4x10

8

63Ni 3.0x10

9 6.6x1010

FIG.3 shows a cross-section view of a trench of TK-L3. There will be three layers of waste

materials in the trenches of TK-L3 of which the depth is 4m. Since the depth will be enough

to prevent that waste materials soak in the groundwater, radionuclides will migrate to the

groundwater by only rain infiltrating into the waste material. Cover soil of thickness 2.5 m

with pavement will reduce public exposure by Gamma-ray from waste materials to safe

FIG.2 LOCATIONS OF A WELL FOR OBSERVATION

: Location of a well for observation of the flow of the groundwater

Location of TK-L3

Tokai NPP

: Site boundary line

Distance of 400m from the shoreline

FIG.3 CROSS-SECTION VIEW OF TK-L3

Tokai Dai-ni

NPP

Tokai NPP

Location of TK-L3

: Location of monitoring post

FIG. 1: BIRD-EYE VIEW OF A LOCATION OF TK-L3 IN TK1 SITE

NPP SITE


53

enough. Retaining walls and dividers will be installed to ensure the safety of emplacement

activities of waste materials.

3. The assessment of the public exposure

In the assessments of the public

exposure from waste materials

disposed in TK-L3, we considered

three kinds of exposure scenarios.

1) Exposure by the migration of

radioactivity to the groundwater

(FIG. 4)

2) Exposure by land-reuse as farmland

(FIG. 5).

3) Exposure by land-reuse to dig the

site. (FIG. 6)

We assessed public exposures by each

pathway of each scenario by applying

appropriate procedure with appropriate

parameters [3-6]. Pathways of each

scenario are shown in text- boxes

colored in blue in FIG. 4, 5 and 6

respectively. Results of the assessment

are shown in TABLE 2.

We also evaluated exposure by

Gamma-ray from waste materials

directly and by sky-shine of Gamma-

ray in the operation period of the

emplacement by using calculation

codes [7-9]. The evaluations are also

tabulated in TABLE 2.

TABLE 2 PUBLIC EXPOSURES BY EACH PATHWAY OF EACH SCINARIO

Exposure by the migration of

radioactivity to the groundwater

Exposure by land-

reuse as farmland

Exposure by land-reuse to

dig the site

Exposure in the

operation period

Intake by drinking

Exposure by activities at

the sea shore

Intake of marine

products

Intake of farm

products

Intake of livestock products

Exposure of construction

activities

Exposure by a

residence

Direct Gamma

-ray

Sky-shine

39.0a 3.2x10

-6a 5.3

a 54.0

a 86.0

a 13.0

a 8.7

a 0.14

a 21.4

a

(1mSv/a)b (10μSv/a)

b (1mSv/a)

b (300μSv/a)

b

(10μSv/a)

b

(50μSv/a) b

a: Unit: μSv/a b: Numerical values in the parentheses are regulation limits.

FIG. 4 Exposure by the migration of radioactivity to the

groundwater

FIG. 5 Exposure by land-reuse as farmland

FIG. 6 Exposure by land-reuse to dig the site


54

4. Summary and conclusion

JAPC has performed safety assessments to ensure necessary measures for maintaining the

TK-L3 safely. Scenarios relevant to public exposure in an operation period and in a post-

enclosure period, which is after controlled period, have been considered to develop

conceptual and mathematical models. In order to identify scenarios, characteristics of

meteorology, geography, hydrology, geohydrology and social environment around the site

were investigated.

The assessments provided that public exposures by TK-L3 are low enough for regulation

limit except intake by drinking and exposure by a residence. In case of intake by drinking,

although the use of the well water at down-stream side in TK-L3 would be possible, the

possibility of its occurrence would be low. Furthermore, because the area of Tokai NPP site

is not suitable for the residence, the possibility that exposure by a residence would be low.

According to the consideration as mentioned here, the assessments show that TK-L3 is a

disposal site safe enough.

REFERENCES

[1] The Japan Atomic Power Company, Decommissioning project of Tokai NPP

http://www.japc.co.jp/haishi/tokai_haishi.html

[2] Nuclear Regulation Authority of Japan, Regulation for disposal of very low-level waste

(in Japanese)

[3] INTERNATIONAL ATOMIC ENERGY AGENCY, Generic Models for Use in

Assessing the Impact of Discharges of Radioactive Substances to the Environment,

Safety Reports Series No.19, Vienna (2001)

[4] INTERNATIONAL ATOMIC ENERGY AGENCY, Sediment Distribution Coefficients

and Concentration Factors for Biota in the Marine Environment, TECHNICAL

REPORTS SERIES No.422, Vienna (2004)

[5] INTERNATIONAL ATOMIC ENERGY AGENCY, Generic Models and Parameters

for Assessing the Environmental Transfer of Radionuclides from Routine Releases,

Exposures of Critical Groups, IAEA Safety Series No.57, Vienna (1982)

[6] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION, Age-

dependent Doses to Members of the Public from Intake of Radionuclides: Part 5

Compilation of Ingestion and Inhalation Dose Coefficients, ICRP PUBLICATION 72,

(1995)

[7] M.L.Couchman et.al., G-33 CODE, NUS-TM-NA-42, Washington D.C. (1965)

[8] W. Engle, et. al, A One Dimensional Discrete Ordinate Transport Code with Anisotropic

Scattering, CCC-254, Oak Ridge (1967)

[9] AECL Research, Point Kernel Code System for Neutron and Gamma-ray Shielding

Calculations Using the GP buildup Factor, CCC-645, AECL Research (1986)


55


SITE SELECTION STUDY FOR RADIOACTIVE WASTE REPOSITORY: STUDY

AREA OF NEGERI SEMBILAN

Che Kamaruddin, A. H.1, Tahar, K. N.

2, Wan Mohamad, W. M. N.

2

1Malaysian Nuclear Agency (MNA), Selangor, Malaysia

2Universiti Teknologi Mara (UiTM Shah Alam), Selangor, Malaysia


Abstract. Radioactive materials are used in beneficial ways such as in medical diagnosis and therapy,

scientific research and specialized industrial applications, many of these activities generate radioactive waste,

which occur either in gas, liquid or solid form. The volume and total amount of radioactive wastes is increasing

every year. From 1984 until 2012, there are more than 8000 unit of disused sealed radioactive sources, 445 m³

of solid wastes, 44 m³ of organic waste and 15,000 m³ liquid waste collected and managed by Nuclear Malaysia

as recorded in the radioactive waste inventory database Nuclear Malaysia. The government considers that the

establishment of a national near-surface repository for low level radioactive waste is a national responsibility

and therefore feasible and comprehensive strategies are needed for continuous waste management. A suitable

repository site must have long-term stability and attributes that will enable the wastes to be isolated so that there

is no unacceptable risk to people or the environment either while it is operating or after closure. Radioactive

waste should be disposed of in a controlled and proper manner by considering the fact that the waste contains

radionuclides that harmful and can bring danger to any living things. Therefore, the criteria for choosing the

suitable or potential sites is very important for an operator who was given responsibility to make sure safety

consideration in all aspects is being complied as stipulated by law from regulatory authority. A study for

screening the suitable area which covered whole state of Negeri Sembilan has been conducted using ArcGIS

software. Two techniques in Multi Criteria Decision Making (MCDM) were considered in the GIS processing

by using Boolean Overlay and Weighted Sum Overlay methods.

Key Words: National near-surface repository, ArcGIS software, Multi Criteria Decision

Making (MCDM).

1. Introduction

The government considers that the establishment of a national near-surface repository for low

level radioactive waste is a national responsibility and therefore feasible and comprehensive

strategies are needed for continuous waste management. Nuclear Malaysia has been given the

responsibility to develop national repository for low level radioactive waste as a long-term

solution for radioactive waste management programme.

A suitable repository site must have long-term stability and attributes that will enable the

wastes to be isolated so that there is no unacceptable risk to people or the environment either

while it is operating or after closure. Criteria for site selection usually consist of geological

factors (which includes soil properties, lithology, lineament, and geomorphology),

meteorological factors such as rainfall distribution, hydrogeological aspects, proximity or

distance from road or river, and land-use factors.

Establishment of screening criteria for this study was driven from many references including

Chuang, et al., (2006), Huang et al., (2006), and Risoluti, et al., (1999), and some case studies

from some countries including Australia, Russia, United Kingdom (U.K) and others as well

as understanding on environmental setting of the study area based on researcher’s judgement.

According to the International Atomic Energy Agency (IAEA), there are four stages should



56

be recognized in the siting process for a radioactive waste disposal facility consist of

conceptual and planning stage, area survey stage, site investigation stage and detail site

characterization stage.

2. Site Selection

Radioactive waste should be disposed of in a controlled and proper manner by considering

the fact that the waste contains radionuclides that harmful and can bring danger to any living

things. Therefore, the criteria for choosing the suitable or potential sites is very important for

an operator who was given responsibility to make sure safety consideration in all aspects is

being complied as stipulated by law from regulatory authority. A study case for implementing

this screening process covered for the whole state of Negeri Sembilan area has been

conducted. This study case is the first phase of screening process to identify the best suitable

sites or location for developing the disposal facility in Malaysia. There are four steps in the

site selection processes which consist of conceptual and planning stage, area survey stage,

site characterization stage and site confirmation stage (IAEA, 1994).

3. Methodology

Research methodology can be divided into five sections such as spatial data collection and

GIS layer preparation, software used, Boolean Overlay and Weighted Overlay, Criteria for

site selection and Geographical Information Science (GIS) Modelling using Model Builder in

ArcGIS 10.2.

3.1.Data collection and GIS layer preparation

Data collections were obtained from many related agencies which are from Agensi Remote

Sensing Malaysia (ARSM), Jabatan Mineral and Geosains (JMG), Jabatan Perangkaan

Malaysia. Spatial data collection used to prepare spatial layers in ArcGIS software. The

Landsat TM Scene of 126/58 and 127/58 obtained from ARSM were used in this study. The

datasets were mosaic and corrected to fit the Rectified Skew Orthomorphic (RSO) Malaysian

Projection by applying image-to-image registration technique. The image were already

enhanced to improve their appearance by using band combination of red, green, and blue

(RGB). Topological and geological features such as lineaments, geomorphology were also

interpreted from the images.

Besides, the satellite image Level 1 GeoTIFF Data of The Landsat 8 OLI/TIRS were also

obtained and downloaded from USGS Earth Explorer website for this study. Through

consideration of all the result, the RGB natural colour combination of band 4, 3, 2 is the most

appropriate combination for visual interpretation for the study area.

Thirteen (13) spatial input data have been used into Model Builder for spatial processing in

ArcGIS 10.2 Desktop software. The input data used in this study are consist of lithology, soil

properties, rainfall distribution, hydrogeology, land use, geomorphology, lineaments, road,

river, town, population, elevation, and slope. Slope was produced from Digital Elevation

Model (DEM) which also can give us information about elevation of the study area. Before

selecting the suitable sites, input layers need to be categorized and reclassified.

3.2.Software Used

ArcGIS version 10.2 is one of the main software used in this research. This GIS software

allows user to edit, update, manipulate and analyze spatial and attributes data explained in the

earlier sections. In this research, this software is used to edit and update the spatial data with


57

new information. For example, the main criteria for each parameters can be classified and be

determined as represented in the attributes table. The surface analysis or 3D Analyst, Spatial

Analyst and the Geoprocessing Tools are the main modules in the ArcGIS software used to

manipulate, process and analyze the related data. Suitable areas for radioactive waste

repository maps are generated using this software.

3.3.Boolean and Weighted Overlay Method

In this section, all required data for site selection were analyzed and reclassified before

appropriate weight value given. A new information layer with a variety of new spatial units

from spatial intersection is important to decide which newly created spatial units should be

summarized and which must be recorded separately when applying this information to

suitability analysis. Boolean algebra is used for this task. Each class in every layer were

assigned either as an output value of 0 or 1 for bitmap known as “Boolean Analysis”. It was

established by the English mathematician and logician George Boole (1815 – 1864). The

value assigned for each class of every parameter was based on their suitability for radioactive

waste repository and must followed all criterions fixed in the early study. For the site

selection study, the spatial maps produced are at the scale of 1: 600, 000.

Each class in every layer were assigned either as an output or weight value between 1

(lowest) to 10 (highest) according to “Binary Evidence Analysis”. The value assigned for

each class of every parameter was based on their suitability for radioactive waste repository

and must followed all criterions fixed in the early study.

The Weighted Sum tool provides the ability to weight and combine multiple inputs to create

an integrated analysis. Basically, there are four main techniques for the development of

weights such as ranking methods, rating methods, pairwise comparison methods and trade-off

analysis methods (Malczewski, 1999). However, for this study, ranking and rating methods

were used for assessing the importance of weights and estimating of weights on the basis of

predetermined scale.

TABLE I: THIS TABLE SHOWS DIFFERENT ASPECTS, PARAMETERS, SUB-CRITERIA,

AND SCORE FOR SELECTING THE POTENTIAL SITE OF THE DISPOSAL FACILITY.

N

o

Aspects Parameters Criteria Score (%)

1

Geology

Lithology Igneous rock 8

Soil properties Clay 5

Land use Forest, cleared land 15

Geomorphology Higher Hills,

mountains

10

Elevation 50 - 300 13

Slope 5 - 15⁰ 6

2 Hydrogeolog

y

Hydrogeology Low yield 5

3 Meteorology Rainfall 1000 -1500 mm 5

4 Accessibility

& Proximity

Lineament 5 km from fault zone 8

Main Road access Within 5 km 5

Main River 2.5 km from main

river

5

Town 5 km 10

5 Land cover Area > 100 ha

5


58

3.4.Criteria for Site Selection

Identifying the criteria for site selection process for developing a radioactive waste facility is

the most important part to be discussed. Siting criteria have been selected and determined

based on feasibility study, published information and also from “expert opinion” before

selecting the best criteria for selection study in order to develop the first National Radioactive

Waste Repository facility in Malaysia.

a) Suitable Criteria

There are thirteen (13) criteria are being considered for site selection study of Negeri

Sembilan consist of suitable criteria, exclusion criteria and preferred criteria. These criteria

influence the site selection decision. Basically, eight (8) parameters related to the suitable

criteria are as lithology, soil type, rainfall, hydrogeology, land use, geomorphology,

elevation, and slope.

b) Exclusive and Inclusive Criteria

Exclusive criteria can be defined as criteria which lead to an exclusion of unsuitable areas

and to the ascertainment of suitable areas within an investigation area. Inclusion criteria

means the subject must be included in the study area. There are three (3) exclusive criteria

and one (1) inclusive criteria have been defined for the case study as below:-

i. Distance to river: exclusion of all areas within a buffer of 2.5 km from main rivers.

ii. Distance to lineament or fault zone: exclusion of all areas within a buffer of 5 km from

lineaments.

iii. Proximity to town: The suitable area should be located at least 5 km from town (buffer 5

km).

iv. Distance to road: considering all areas within a buffer of 5 km from main roads.

c) Preferred Criteria

A preferred criteria for the site selection processes was also be determined. The suitable

area for the development of national radioactive waste repository facility shall be more than

100 hectares.

Then, the siting criteria have to be sorted into several categories or classification using

ArcGIS after determining it with previous studies and expert opinions from many local and

international agencies or companies.

3.5.GIS Modelling

GIS Modelling is used to analyze multi-layer of data spatially and quantitatively. The

accuracy and reliability of the result using GIS application could be high depending on the

available spatial data. All the parameters or layers in every aspect were overlaid to produce

an intermediate maps. These maps were overlaid with exclusion map to produce the final

suitable area. Then, the most suitable areas were determined.

4. Analysis and Results

For this study, Boolean Overlay were being presented in Model 1(a) and Model 1(b).

Analysis using Model 1(a) was not take into account town buffer as one of important criteria

for site selection studies. Model 1(b) indicates the area with concerned of buffer from town

which should be at least 5 km. Instead of using Boolean Overlay method, Model 2 was

presented with concerned of weighted overlay method in the site selection studies. However,

to get the final suitable areas, the Boolean Methods were also used to exclude buffer which

related to a specific criteria that have been discussed earlier.


59

4.1.Boolean Overlay (Model 1a and Model 1b)

To well describe Boolean Overlay Method in the ArcGIS interface, there are some steps and

processes which need to be followed before selecting the most potential area for the disposal

facility. The result maps produced after geospatial data processing in ArcGIS are shown in

Figure 1 as follows:-

Model 1(a) Model 1(b)

FIG 1: Comparison between Model 1(a) and Model 1(b). Model 1(a) before considering buffer 5 km

from town and Model 1(b) after considering buffer 5 km from town

4.2.Weighted Overlay (Model 2)

Weighted sum overlay techniques have been applied in order to overlay the map layers.

Weighted overlay is a technique for applying a common scale of values to diverse and

dissimilar input data to create an integrated analysis. To produce the final output raster or

suitability map, weighted sum overlay of the cell values for each input raster are multiplied

by the raster's weight. The maps resulting from this weighted sum overlay for site suitability

area is shown in Figure 2.

FIG 2: More potential areas were identified (in white color) by using Model 2 analysis

5. Discussion

Boolean Overlay and Weighted Sum Overlay techniques have been applied In order to

overlay the map layers. Boolean Overlay method are also known as Binary Overlay which is

a technique of grid analysis systems using AND function to combine the binary scores of

different criteria. This technique also eliminating any cells that did not score “good” in all

criteria. An OR function used to allow minimum risk approach where all cells that scored

“good” at least one criterion are selected as suitable.


60

Generally, weighted overlay is a technique that has been used by many scientists or

researcher with using a common scale of values to diverse and dissimilar input data for

creating an integrated analysis. In this study, different map layers characterizing site

suitability of the potential area were weighted using the weights derived from expert opinions

and literature review papers. In the weighted sum overlay, the cell values of each input raster

are multiplied by the raster's weight criteria weights.

From the results and analysis of Model 1(a), six (6) suitable areas were identified which is

located in Mukim Labu, Negeri Sembilan area. After considering town buffer 5 km from

town, only one (1) suitable site was identified which is also in Mukim Labu as presented in

Model 1(b).

Then, the output results from the process were added in order to produce the output raster.

These output raster need to be reclassified again into five (5) classes to show the importance

of the layers. Then, the reclassified raster layers were converted into vector maps by using

raster to polygon tools. However, weighted sum overlay used in the Model 2 also integrates

with Boolean Overlay process to erase the exclusion areas and intersect the inclusion area in

order to produce the final output raster or suitable area maps for developing the radioactive

waste repository facility.

With using Boolean Overlay and Weighted Overlay methods for Model 2, the identified

suitable site areas were scattered into several location in the east, west, and north of Negeri

Sembilan.

6. Conclusions

The suitable sites are located at the Mukim Labu area in the district of Seremban from the

west of the Negeri Sembilan state. By analyzing the potential area with coverage area of 100

hectares or 1 km2, there are three most potential sites to be selected for developing

radioactive waste disposal facility. This study provides an opportunity to explore a survey

method to find potential site for a low level radioactive waste repository in the state of Negeri

Sembilan, Malaysia using remote sensing and GIS technologies. Spatial data representing

geological, meteorological hydrogeological, and surface process were utilized for assessing

and characterizing the suitability of potential sites.

REFERENCES

[2] INTERNATIONAL ATOMIC ENERGY AGENCY, Siting of Near Surface Disposal

Facilities, IAEA Safety Series No. 111-G-3.1, Vienna (1994).

[1] Chuang, W-S., Chi, L-M., Tien, N-C., and Chang, F-L., Site Selection for the Disposal

of LLW in Taiwan. Proceeding of the Waste Management Conference, Feb. 26 –

March 2, Tucson, AZ, USA (2006).

[2] Huang, L.X., Sheng, G., and Wang, L., GIS Based Hierarchy Process for the Suitability

Analysis of Nuclear Waste Disposal Site. Journal of Environmental Informatics

Archives, Vol. 4, 289-296 (2006).

[3] Risoluti, P., Ciabatti, P., and Podda, A., The Site Selection Process Under Way in Italy

For LLW Repository and HLW Storage. Journal of Radioactive Waste Management

and Environmental Remediation, Roma, Italy (1999).


61

[4] Malczewski, J., GIS and Multicriteria Decision Analysis. John Wiley & Sons, New

York, pp. 392 (1999).

[5] Abdullah, C.H., Mohamad, A., Yusof, M.A.M., Gue, S.S. & Mahmud, M.,

“Development of Slope Management in Malaysia”, Malaysia (2007).

[6] S.A., Grant, I.K., Iskandar, Contaminant Hydrology - Cold Region Modelling. CRC

Press, London, New York.89-90 (2000).

[7] Guidelines for Slope Design – Slope Engineering Branch, Public Work Department

Malaysia, (2010).

[8] K. D., Lachman, P.G., Harrington, Retrievability as proposed in the U.S. High-level

Radioactive Waste and Spent Nuclear Fuel Repository Concept. W.M’01 Conference,

Las Vegas, Nevada (2001).

[9] Soil Survey Staff, Soil Survey Manual. United States Dept. of Agri. U. S. Government

Printing Office, Washington, D. C. 20402. pp. 136-146 (1993).

[10] Adzemi M.A., Land Evaluation System for Elaeis Guineensis Jacq. Cultivation in

Peninsular Malaysia. Universiti Putra Malaysia. pp. 6-7 (1999).


62


REGULATORY APPROACH FOR THE ASSESSMENT OF THE LICENCE

APPLICATION FOR RADIOACTIVE WASTE MANAGEMENT FACILITIES IN

AUSTRALIA

S. Sarkar

Regulatory Services Branch, Australian Radiation Protection and Nuclear Safety Agency

(ARPANSA), PO Box 658, Miranda, NSW 1490, Australia


Abstract. This paper describes the regulatory approach of Australian Radiation Protection and Nuclear Safety

Agency (ARPANSA) for assessing the operating licence application for radioactive waste management

facilities. These facilities relate to predisposal management of radioactive waste. ARPANSA is the regulatory

authority for commonwealth entities operating nuclear installations including radioactive waste management

facilities. In assessing the application for nuclear installations the ARPANSA assessors prepare a Regulatory

Assessment Report (RAR), which is a recommendation to the Chief Executive Officer (CEO) of ARPANSA

whether to issue a licence to site, construct, operate and decommission facilities. The key elements of

ARPANSA’s assessment include plans and arrangements for managing safety, safety case, use of defence-in-

depth and conservative proven design and engineering practice, operational limits and conditions, and use of

international best practice. Compliance monitoring of licenced facilities are undertaken through regular

reporting, site visits and a risk-based planned inspection program.

Key words: radioactive waste management, regulatory assessment, predisposal management

1. Introduction

In Australia the main legislation and regulatory framework governing the safety of nuclear

installation, radiation facilities and radioactive material is the Australian Radiation

Protection and Nuclear Safety Agency Act 1998 (the Act) and the Australian Radiation

Protection and Nuclear Safety Regulations 1999 (the Regulations). The objective of the Act is

to protect the health and safety of people and the environment from the harmful effects of

radiation. The Act gives the CEO power to attach conditions to the licence for operation of

the facility.

Nuclear installations that operate under ARPANSA facility licence include research reactor,

radioisotopes production facilities, radioactive waste management and spent fuel

management facilities. All of Australia’s existing nuclear installations are under the effective

control of the Australian Nuclear Science and Technology Organisation (ANSTO). Though

some other Commonwealth entities operate radioactive waste management facilities nuclear

operational wastes are managed by ANSTO. The licensed waste management facilities

operated by ANSTO include:

Low level solid waste facilities for handling, processing and storage

Intermediate level solid waste facilities including storage of vitrified waste generated

from reprocessing of research reactor fuel, and various solid waste and solid waste

containing LEU generated from production of molybdenum-99

Low level liquid waste including storage, treatment and conditioning

Storage of intermediate level liquid waste



63

Storage of nuclear materials and safeguards materials

In 2014, ARPANSA issued the licence to site and construct an intermediate level liquid waste

conditioning facility to process the intermediate level liquid waste generating from the

production of molybdenum-99. Synroc (synthetic rock) technology, developed by ANSTO,

will be used to condition such intermediate level liquid waste.

2. Discussion

The ARPANSA regulatory approach is non-prescriptive but provides guidance and therefore,

the licence holder has flexibility in developing plans and arrangements to give appropriate

safety. To help judge the adequacy of the plans and arrangements, and safety cases

ARPANSA has published Regulatory Guidelines [1], Regulatory Assessment Principles [2]

and Regulatory Assessment Criteria [3]. These have been developed drawing on past

experience, best practice and international standards such as International Atomic Energy

Agency (IAEA).

An updated safety case is required for each of the principal stages in the life of a facility. The

safety case includes the design information for the facility, including the operational limits

and conditions (OLCs) within which the facility must operate, and a safety analysis that is

documented in the safety analysis report (SAR). The extent and rigour of the SAR should be

commensurate with the hazard categorisation of the facility. The margins between the

operational limits and conditions and the relevant safety limits are included in the SAR. As

part of the safety analysis report the licence holder must categorise the hazard associated with

the facility and also categorise systems, structures and components by their safety

significance.

The safety analysis establishes the hazard of the facility according to the following

categories:

Hazard Category F1: where there is no potential for significant3 consequences outside

the facility.

Hazard Category F2: where there is potential for significant1 consequences on the site

outside the facility, but not outside the site.

Hazard Category F3: where there is potential for significant1 consequences outside the

site

The responsibility for demonstrating each relevant assessment principle rests wholly with the

licence holder. The operating organisation’s safety analysis can assist in demonstrating safety

to ARPANSA. An alternative to an ARPANSA principle may be acceptable to the CEO of

ARPANSA if the operating organisation clearly demonstrates that the alternative principle

provides a degree of safety based on the application of contemporaneous international best

practice in radiation protection and nuclear safety.

Regulatory assessment (safety evaluation) of the information described in the application is

the key aspect for granting authorisation to operate a nuclear installation. Of particular

importance in safety evaluation for a facility licence are the plans and arrangements for

managing safety. These plans require the demonstration of appropriate arrangements for:

effective control;

3 Some judgement is required on significance and guidance is given in the Regulatory Assessment Principles

and also need to be addressed in the SAR


64

safety management;

radiation protection;

radioactive waste management;

security;

emergency; and

environment protection

The plans and arrangements for managing safety are then assessed against regulatory

guidelines developed by ARPANSA [1]. In addition safety case is assessed against

ARPANSA’ assessment principles [2] and criteria [3] and the requirements of the

international standards such as those of the IAEA.

A licence may be subject to conditions as set out in the ARPANS Act and Regulations; or

imposed by the CEO. Such licence conditions are not surrogates for safety; they outline

certain additional requirements placed on the licence holder that will assure the CEO of

ARPANSA that the licence holder is undertaking the licensed activity safely.

3. Results

After assessing all the relevant information ARPANSA assessors prepared a regulatory

assessment report (RAR), which is a recommendation to the CEO of ARPANSA whether to

issue a licence to operate4 the waste management facilities. This report was based on the

results of the detailed assessment of the application and the resolution of issues resulted from

any public submission on the application. The RAR is a complete assessment of the

application for an authorisation to operate the waste management facilities. This report

demonstrates that the conduct for which the licence is sought can be effectively controlled to

provide adequate protection to the health and safety of the people and the environment. The

operation of the ANSTO waste management facilities are currently subjected to six licence

conditions as they relate to compliance reporting, periodic performance assessment and

discharge of radioactive waste, and compliance with OLCs.

4. Compliance Monitoring

Apart from regular reporting (quarterly and annually) by the licence holder and regular site

visits, a long-term schedule of inspection is used for regulatory compliance monitoring. The

complexity and risk inherent in each facility determines the scope and duration of each

inspection. The scope of the inspection is determined based on risk using a graded approach.

There are eight inspection areas covered during the baseline inspection period. For example,

a single inspection may last two weeks and involve just one of the eight areas; it may, on the

other hand, involve four areas and last only two days.

The following inspection areas, which are comprised of more specific modules, collectively

known as the Performance Objectives and Criteria, constitute the baseline schedule for each

licence. These areas are broad in scope and intended to cover all aspects of licence holder

performance.

i. Performance Reporting Verification: address the reporting culture, both internally and

externally, including discrepant or unreported performance data, performance

indicator verification, and compliance with operating limits and conditions.

4 the RAR is also prepared for the CEO whether to issue a licence for siting, construction and decommissioning

of facility


65

ii. Configuration Management: include evaluation of facility modifications, equipment

alignment, operability determinations, temporary facility modifications, and safety

system design and capability.

iii. Inspection, Testing, and Maintenance: include post-maintenance testing, in-service

testing and inspection, surveillance testing, and maintenance and work control.

iv. Training: address personnel training, the use of a systematic approach to training,

accredited operator training, etc.

v. Event Protection: include adverse weather, fire protection, flooding, bush fires, land

management, etc.

vi. Security: include aspects of security arrangements and requirements. Modules also

include infrequently conducted tests or evolutions, outage performance, etc.

vii. Radiation Protection: include access control, dosimetry, ALARA planning, radiation

monitoring instrumentation, effluent system monitoring, radioactive material

processing and transportation, etc.

viii. Emergency Preparedness and Response: include exercises and drills, emergency

response organisation testing, notification testing, etc.

There are also three cross-cutting aspects that may be addressed in each inspection, namely,

Human Performance, safety Culture, Performance Improvement.

5. Conclusion

The matters and process considered in the regulatory assessment provide an effective and

efficient regulatory approach for safe and secure operation of nuclear and radiation facilities

including radioactive waste management facilities. Australian government is planning to

construct a National Radioactive Management Facility (NRWMF) to provide a centralised

location for the disposal of low level waste and storage of intermediate level solid waste

facilities. ARPANSA will apply similar regulatory approach to the NRWMF.

REFERENCES

[1] Australian Radiation Protection and Nuclear Safety Agency, ARPANSA, Regulatory

Guide: Plans and Arrangements for Managing Safety, (2014)


Assessment Principles for Controlled Facilities, RB-STD-42-00, (2001)


Assessment Criteria for the Design of New Controlled Facilities and Modifications to

Existing Facilities, RB-STD-43-00, (2001)


66


SIMULATION AND STABILITY ANALYSIS OF NEAR SURFACE DISPOSAL

TRENCHES OF RADIOACTIVE WASTES BY USING FINITE ELEMENT METHOD

M. Boroumandi, A. Masood Taheria, A. Bagheri, S. Hasanlou, S. Momenzade

Iran Radioactive Waste Management Co. (IRWA)/ Iran Atomic Energy Organization/

Tehran/Iran


Abstract. At this paper, simulation and stability analysis of near surface disposal trenches in Anarak

Repository has been evaluated by finite element method. Emplacing of waste packages in the trenches impose

extra load to bottom and sides of trenches and it may increase deformations in the trench walls and cover. Also

corrosion of waste packages in long term may increase deformations and settlements and finally failure of

trenches may be occur. 2D model of trench containing waste packages with elasto-plastic behavior of materials

in finite element software has been designed and analyzed. Amount of deformations were measured and

considered in design of trenches.

Key Words: Disposal Trench, LILW, Finite Element Method, Stability Analysis.

1. Introduction

There are many disposal facilities for low and intermediate level radioactive wastes (LILW)

around the world. Near surface trench is one of the most common methods for disposal of

LILW [1, 2]. Effective and safe management of radioactive wastes needs an inclusive

program and consideration different criteria in trenches design [1, 3].

Trench design besides waste characteristics, site characteristics, type of barriers and

characteristics of cover layers are influencing on overall safety of disposal. LILW wastes

have different physical and chemical properties because of their different sources. For near

surface disposal trenches, suitable wastes are solid and solidified wastes with low leaching

rate of radionuclides and a small, non-degradable toxic chemical content [3, 4]. But in long

term, degradation of waste packages due to corrosion and also settlement of backfill layers

and cover can reduce safety of disposal trench. So evaluation of different aspects of trench in

design stage is an important step toward ensuring operational as well as long term safety

disposal.

Disposal of solid wastes with no treatment may effect on disposal performance and followed

by settlements due to low stiffness and high compressibility. So characteristic of wastes and

waste packages are important parameters in safety of disposal. Emplacing of waste packages

in the trench impose extra load to bottom of trenches and it may lead deformations in the

trench walls and bottom. Also corrosion of waste packages may increase deformation and

settlement. These deformations can impose cracks in cover and trenches walls and finally,

failure of trench may occur. For measurement of amount of deformations, design and

performing an elasto-plastic model performed.

Anarak Repository is Iran near surface disposal facility for LILW. Different properties of

wastes can effect on disposal safety, so these subject considered in design stage. The

objective of this paper is to study stability analysis of trenches in different conditions by

finite element methods.



67

2. Material and Methods

2.1.Study Area

The Anarak site has been selected as disposal facility for LILW in Iran. This site is located in

Isfahan province, about 24 km west of Anarak city and 90 km northeast of Naein as is shown

in Fig. 1. This site has been located in a syncline structure which consists of clay, marl and

sandstone layer that create good condition for controlling of radionuclides migration.

FIG. 1. Location of study area

2.2.Finite Element Simulation of Disposal Trenches

Numerical methods have been used in engineering project, extensively. Stability analysis of

near surface trenches in Anarak Repository has been evaluated by finite element method to

predict deformations and settlements. In primary stage of trenches design dimension of

trenches has been selected based on design criteria. 2-dimensional profile of trench with

wastes and backfill layers has been considered for convenient meshing and further analysis.

For assignment of material properties, elasticity properties of host rock, backfill layers and

waste packages should be considered. Mohr-Colomb criterion was selected for material

properties assignment. This area consist of two parts include of alluvium layers and host rock.

For determination of soil properties, direct shear test results were analyzed and for rocky part

uniaxial compression test and 3-axial compression tests were done on core samples obtained

from different boreholes. Summary of material properties were used in simulation presented

in Table I.

Cover weight take into account as surcharge load in model based on properties of different

layers and their thickness. It’s required to assigning the local body forces into the trench’s

various parts consisting body forces of barrels, cover layers, host rocks and backfill layered

soil. At this study density of host rock, waste packages and soil were considered as 2300,

2500 and 2100 N/m3, respectively.


68

TABLE I: MATERIAL PROPERTIES OF WASTE PACKAGES, SOIL AND ROCKS

Material Young Modulus (E)

GPa

Poisson ratio

(ν)

Friction Angle (φ)

Cohesion (C(

MPa

Host rock (clay and

marl stone)

1.96 0.23 40 6

Backfill and Soil layers 25*10-3 0.25 34 0.05

Waste Packages 204 0.29 - -

Waste analyzed in different alignment such as horizontal and vertical alignment and with

different properties of waste packages. Then by increment of loads of barrels, backfill layers

and cover layers, amount of deformations, stresses, strains and settlement in different parts of

trench were measured.

3. Results

3.1. Stability analysis of trenches with high stiffness and cover loads

The model was analyzed with high stiffness for consideration wastes in primary conditions.

After analysis by finite element model deformation of trenches were measured. Based on

mechanical properties of barrels and model results horizontal alignment have been chosen for

waste emplacement in trenches. The results showed that maximum stress and deformation are

0.64 MPa and 1.96 cm, respectively. On the other hand the amount of plastic strain measured

equal to 3.3*10-1

and there is no plastic strain in trenches wall, which show stability of

trenches wall. So we can conclude the trenches wall against to loads of cover and barrels

remain stable and there is no problem in stability analysis of trenches wall.

Stability parameters have been met when the barrels are solidified with concrete and they are

stiff, but after a long period of time, barrels strength will be decreases because of the

corrosion of materials. So, low elasticity parameters (Poisson's ratio and Young's modulus) of

wastes were considered for stability analysis and maximum subsidence of radioactive wastes

has been measured.

5.1 Stability analysis of trenches with low stiffness and cover loads

One of important issues in near surface disposal trenches is long term behavior of wastes in

trenches. Waste packages in trenches will corrode and baring capacity of them will be

decreased. These deformations can be followed by settlements and decrease of performance

of covers.

So at this stage model was analyzed by low stiffness of barrels (0.5 MPa elasticity modulus

and 0.4 Poisson's ratio). Results showed with assumption of deformability behavior of

barrels and backfill layers, the amount of deformations increased to 28 cm. However strain

plastic in trenches wall remain zero, but stresses have been increased. These deformations are

presented in Fig. 2.


69

Total deformations with low stiffness Maximum stress with low stiffness of barrels FIG. 2. Deformation and stresses of model for wastes with low stiffness

4. Conclusion

Weights of cover layers, waste packages and backfill layers which are considered as external

loads can influences on trench's performance. Solid wastes and solidified waste after

corrosion will have very low strength and they can't tolerate external loads and finally

settlement may occur.

Finite element simulation for trench's geotechnical model (as one of simulation method) was

performed to analyze the trench deformations under different conditions. Displacement,

plastic strain and stress concentration in the trenches evaluated in different situation.

The results showed different amounts of deformation in different situations. The maximum

deformation has been measured in surface and cover layers and it's about 28 cm. Also

maximum stress in trenches wall is about 3.2 MPa. According to long term operation of

trenches, maximum deformations were considered for detailed design. So disposal of solid

and solidified wastes with low stiffness in near surface trenches should be operated with

corrective actions.

REFERENCES

[1] INTERNATIONAL ATOMIC ENERGY AGENCY, The Principles of Radioactive

Waste Management: A Safety Fundamental, Safety Series No. 111-F, IAEA, Vienna

(1995)

[2] HAN, K.W., HEINONEN, J., BONNE, A., Radioactive Waste Disposal: Global

Experience and Challenges, IAEA Bulletin 39 1 (1997) 33–41

[3] INTERNATIONAL ATOMIC ENERGY AGENCY, Technical considerations in the

design of near surface disposal facilities for radioactive waste, IAEA-TECDOC-1256,

Vienna (2001).


Specific Safety Requirements, IAEA-SSR-5, Vienna (2011).


70


DESIGN OF A NEAR SURFACE DISPOSAL FACILITY FOR LOW AND

INTERMEDIATE LEVEL RADIOACTIVE WASTE IN ZARIA, NIGERIA

1A. Ibrahim, D. Kula

2

1Nigerian Nuclear Regulatory Authority, Abuja, Nigeria

2Department of Mechanical Engineering Ahmadu Bello University Zaria, Nigeria


Abstracts. Near surface disposal facility with the capacity for receiving, treating, managing and disposing

about 500 waste containers of low and intermediate level radioactive waste was designed. The facility is

designed to have the administrative area, operational area and the final disposal trench. The design is intended to

manage low level waste currently generated in Nigeria from the use of nuclear applications i.e. Medical,

Industrial, nuclear well logging, research and training, it’s also designed to efficiently handle intermediate level

waste expected to be generated from the expansion of nuclear technology in Nigeria, and the proposed Nuclear

Power Plant (NPP). The final disposal trench is sectioned to provide for future waste disposal, in such a way that

until a section is fully occupied and sealed before other sections are utilized. The trench is designed to be safely

operated, secured from unauthorized access to prevent been used for malicious intent and to meet Nigerians

safeguards obligations. The trench is shielded to attenuate gamma and low neutron with mathematically

determined concrete thickness of 1.4meters for surface shield. With these shield the background radiation within

the trench would not exceed 𝟎. 𝟑 𝒎𝑺𝒗/𝒚𝒓 cumulatively when operational. The cost estimate for the trench was

determined to cost N24,103, 900.00

1. Introduction

Nuclear technology and application in Nigeria is fast developing; Nigeria and many other

developing countries use nuclear application in everyday activities, in health sector,

agricultural sector, oil and gas Industries, Construction industries, Manufacturing Industries,

Education and training. [1]. Nigeria has a nuclear power program which has been going on

for quite a long time. The country hopes to have its first nuclear power plant by 2025, and

efforts to build a new research reactor of 7MW capacity are underway [2]. The issue of

nuclear waste management comes up whenever there are nuclear programs, the challenges of

nuclear waste is enormous, due to safety and security concerns. Radiological waste from the

hospital, from disuse sources and the most delicate and controversial waste from nuclear

power plant spent fuel. Under typical circumstances, a developing country using sealed

radioactive sources may generate hundreds of disused sources with low levels of radioactivity

over several years. Although Nigeria do not have nuclear power plant yet, but the presence of

thousands of radioactive sources used in the oil and gas sector also calls for concern. Low

activity sources pose the larger challenge because they exist in large quantities around the

world and in different forms and variations [3].

2. Materials and Methods

The design of the near surface facility for radioactive waste, the facility consists of various

facilities which would help in waste management operations before final disposal. It is sub

divided it two major sections, Administrative and Operational



71

2.1. Administrative Section

The Administrative section is meant for daily administrative activities, with provisions for

reception, offices for records and other administrative duties, security office for monitoring

access and physical protection of the facility and the surroundings.

2.2. Operation section

Operational section is where the predisposal operation takes place before the waste is finally

disposed in the disposal repository. It consists of Interim Storage facility, Unloading and

sorting facility, General services facility, waste condition facility and provisions for

decontamination facility which would serve as in and exit point for contamination

monitoring. The scope of the design is emphasize on the final disposal trench for low and

intermediate radioactive waste with the objective of isolating the waste from people and the

environment until natural processes of decay & dilution prevent any radionuclide from

returning in concentrations that pose a hazard [4].

2.3 Zaria Site Characteristic

It was important to consider and know the sub soil formation of Zaria and Ahmadu Bello

University were the trench was designed for, the only research reactor in Nigeria is also

located in Zaria, there are also possibilities they consider this design when they plan to build

a new radioactive waste facility. The A.B.U Zaria in being part of the kubani, basin, is

therefore underlain by Precambrian rock of the Nigerian baseman. The proposed site is

situated within the south-western part of the campus and underlain by muscovite

biotiticgnoiss. The alluvial deposal in Zaria area consists of granite, sands, silts, and clay.

Thickness ranges from 5 - 15m and the aquifer ranges from 28 – 34m [5].

2.4 Trench design Consideration

The trench is designed for disposal of low and intermediate waste to take care of current and

future radioactive waste to be generated from the expanding nuclear energy application in

Nigeria and nuclear power program. The facility is designed to accommodate more than four

hundred containers of low and intermediate level waste, low level waste in metal containers

while intermediate level waste in concrete containers, it’s expected that before the waste is

brought into the trench for final disposal it must be conditioned to meet the facility waste

acceptance criteria for solid waste. The surface radiation of each waste container would not

exceed “𝟏𝟎 𝒎𝑺𝒗 𝒉⁄ “which is the IAEA Transport Regulations for exceptional case. The

trench is designed to be partitioned into four sections, this would provide disposal medium

for future expected waste, each section is designed to accommodate 100 drums and more

waste containers arranged in two columns or more depending on the numbers of waste

containers ready for disposal, the trench is sectioned to provide for future waste disposal, in

such a way that until a section is fully occupied and caped before the other section is utilized ,

this may take long period of time depending on the quantity of waste the facility is projected

to receive, each section would be independent from the other and they shall be shielded

separately at the top and walls of the trench.

2.5 Design Parameters


72

The trench is to be dug to a depth deep beneath the ground to 25 meters before the aquifer

this is to give reasonable distance to the aquifer to prevent ground water contamination. The

depth of the containers to the top soil of the trench is about 9 meters and the depth and

distance from the drums to the cap is estimated to be 5 meters. Drainage layer of about

200mm is also incorporated into the design to prevent leaching and it’s also an engineered

feature to help the integrity of the trench from chemical reactions. The trench is to be

backfilled at all sides with red clay and Betonite. Monitoring pipe is also incorporated into

the design to help monitor for contamination, and other activities, such as concentration,

temperature, pressure and activity inside the trench.

2.6` Shield Design Calculation

"𝒏 = 𝒙

𝑯𝑽𝑳" Where; 𝒙 = Thickness of shield, 𝑯𝑽𝑳 = Thickness of concrete permit

𝟏

𝟐 of the

incident radiation to pass “ 𝒏 = 𝒍𝒏(

𝜤°𝜤

)

𝒍𝒏 𝟐=

𝒍𝒏(𝜤°𝜤

)

𝟎.𝟔𝟗𝟑 ” , “𝒙 = 𝑯𝑽𝑳 ×

𝒍𝒏(𝜤°𝜤

)

𝟎.𝟔𝟗𝟑 ”

Note that 𝑯𝑽𝑳 of 𝑨𝒎𝟐𝟒𝟏 was considered because it has higher 𝑯𝑽𝑳 for concrete at

. 𝟓 𝒈 𝒄𝒎𝟑⁄ . Shielding factor from IAEA Safety Series No: 47. “In order to solve nuclear

waste problem you have to solve the americium problem [6].

𝑯𝑽𝑳 = 𝟔. 𝟗𝒄𝒎, 𝜤° = 𝟒𝟎𝟎𝟎 𝒎𝑺𝒗 𝒉⁄

𝟏𝟎 𝒎𝑺𝒗 𝒉⁄ was considered based on IAEA Transport Regulations for exceptional cases was

considered to optimize the shield. Therefore: multiply 400 drums × 𝟏𝟎 𝒎𝑺𝒗 𝒉⁄ , i.e 400

drums considered.

𝟎.𝟑 𝒎𝑺𝒗 𝒉⁄

𝟑𝟔𝟔 ×𝟐𝟒 = 𝟎. 𝟎𝟑𝟒 𝒎𝑺𝒗 𝒉⁄ ,……𝒏 =

𝒍𝒏(𝜤°𝜤

)

𝟎.𝟔𝟗𝟑

𝒏 = 𝒍𝒏 (

𝟒𝟎𝟎𝟎 × 𝟏𝟎𝟑 𝝁𝑺𝒗 𝒉⁄𝟎. 𝟎𝟑𝟒 )

𝟎. 𝟔𝟗𝟑

= 𝟐𝟎. 𝟏𝟕 (𝟔. 𝟗 Shielding factor from IAEA Safety Series No: 47), 𝒏

𝒙 = 𝑯𝑽𝑳 × 𝒏 ,𝒙 = 𝟔. 𝟗𝒄𝒎 × 𝟐𝟎. 𝟏𝟕 = 𝟏𝟑𝟗. 𝟏𝟕𝟑𝒄𝒎 ∴ 𝒙 = 𝒕𝒉𝒊𝒄𝒌𝒏𝒆𝒔𝒔 ≅ 𝟏. 𝟒 𝒎𝒆𝒕𝒓𝒆𝒔


73

FIG. 1.Plan of the Trench Design.

3. Conclusion

A trench with a capacity of storing and receiving over 400 low and intermediate level

radioactive waste was designed for disposing low and intermediate level waste in Zaria.

Design analysis was carried out for the disposal facility. The design concept of capping the

top of the trench was developed based on years of utilizing the trench and in expectation of

waste readily available for disposal, so that after each section is fully utilized before the other

section is used. With the effective shield provided by the design and arriving at 1.4 meters

thickness of concrete for the cover of trench the environment is expected to be safe from the

harmful effect of ionizing radiation.

REFERENCES

[1] Ibrahim. A, “Minimizing the risk of Proliferation and Nuclear / Radiological Terrorism

in Nigeria” Paper presented at Center for Non proliferation Studies, Paper, 3, California

(2014)

[2] Mallam, S.P. “Over view of Radioactive Waste Management in Nigeria”, Paper presented

at the National training Course on Radioactive waste management system responsibilities

allocation, Abuja 26-30 November 2012

[3] International Atomic Energy Agency (2016). IAEA Reaches Milestone in Disposal of

Radioactive Sources. Accessed from www.iaea.org. 2016

[4] International Atomic Energy Agency Technical Document“TECDOC-1515

“Development of specification in radioactive waste management” (2006), Accessed from

www-pub.iaea.org. 5th

February 2016

[5] Jimoh; 8, L.,“Geophysical Investigation of a Sewage Treatment Site at Ahmadu Bello

University, Zaria (Main Campus) using 2-D Electrical Resistivity Tomography”

(2011).(MSC Thesis) Department of Physics , Ahmadu Bello University, Zaria, Nigeria.


74

Tom, M. Scientist discovers how to remove radioactive toxic element americium.(2016)

Accessed April 10 2016 from www.ibitimes.co.uk.

http://www.ibitimes.co.uk/


75


THE NATIONAL DISPOSAL FACILITY FOR RADIOACTIVE WASTE IN

BULGARIA

T. v Berlepsch1, E. Gonzalez Herranz

2, I. Stefanova

3, B. Haverkamp

1,

G. Nieder-Westermann1

1DBE TECHNOLOGY GmbH, Peine, Germany

2WESTINGHOUSE ELECTRIC Spain, Madrid, Spain

3Bulgarian State Enterprise for Radioactive Waste Management (SERAW), Sofia, Bulgaria


Abstract. The need for building a National Disposal Facility (NDF) is recognised by the Bulgarian

government and the Bulgarian State Enterprise for Radioactive Waste Management (SERAW). To address this

need SERAW is endeavouring to build a near-surface repository for short-lived low- and intermediate-level

radioactive waste to discharge its statutory responsibilities in waste management. The European Union finances

the establishment of the NDF through the Kozloduy International Decommissioning Support Fund (KIDSF).

SERAW placed a contract for the development of the design to the Consortium of Westinghouse Electric Spain

S.A.R, DBE TECHNOLOGY GmbH, and ENRESA. Two Bulgarian companies participate also in the project as

subcontractors – EQE Bulgaria and to a lesser extent КК–Project. After obtaining SERAW's approval for the

preferred repository conceptual design variant, the repository project usually referred to as R Project 5, focused

on developing the Technical Design and preparation of the Intermediate Safety Analysis Report (ISAR) that

demonstrates the safety and suitability of the proposed NDF design. These documents were delivered to the

relevant Bulgarian authorities. The required permits have either been received or are expected before the end of

2016. This paper summarises the main steps and achievements of R Project 5.

Key Words: R-Project 5, National Disposal Facility, Radioactive Waste Management,

Licensing.

1. Introduction

In the framework of the accession treaty to the European Union the Republic of Bulgaria

committed itself to the early decommissioning of the four WWER 440-V230 reactors at the

Kozloduy Nuclear Power Plant (KNPP). Due to this early decommissioning large amounts of

low and intermediate radioactive waste will arise much earlier than initially scheduled.

Consequently, Bulgaria has intensified its efforts to provide a near surface disposal facility

for low and intermediate level waste (LILW) with the required capacity. It is supported in this

endeavour by a compensation mechanism established by the European Union, the “Kozloduy

International Decommissioning Support Fund (KIDSF)”, aimed at alleviating the significant

impact of the early NPP phase out on Bulgaria`s economy. The fund is managed on behalf of

the European Union by the European Bank for Reconstruction and Development (EBRD).

In a series of projects the State Enterprise for Radioactive Waste (SERAW) selected a site for

the National Disposal Facility (NDF) in the vicinity of the KNPP at Radiana, which is located

on the terraces of the Danube River valley's south rim. This early work also specified

Enresa’s facility at El Cabril as the reference design for the NDF. SERAW placed a contract

for the development of the design of the NDF with the Consortium of Westinghouse Electric

Spain S.A.R and DBE TECHNOLOGY GmbH of Germany, with ENRESA, the Spanish

National Waste Management Agency providing technical review and support. Two Bulgarian

companies participate also in the project as subcontractors – EQE Bulgaria and КК–Project.



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The project's official name is “Technical Design and ISAR Preparation for the National

Disposal Facility at Bulgaria”, usually referred to as “R-Project 5”.

The NDF design work started in October 2011. Initially, two repository conceptual designs

were developed considering the particular characteristics of the Radiana site. The most

favourable variant was selected by means of a formal multi-attribute analysis with evaluation

criteria such as operational and long-term safety, environmental impact, constructability,

initial investment, and operational costs. SERAW approved the recommended Conceptual

Design on December 2012, and authorised the Consortium to begin development of the

Technical Design. Since then the Technical Design work has been completed and submitted

to SERAW. Simultaneously to the preparation of the Technical Design for the NDF, the

consortium also developed the Intermediate Safety Analysis Report (ISAR), currently

awaiting final approval.

2. The Repository Design

The Radiana Site, a quasi-rectangular 46-hectare area with approximate maximum

dimensions of 470 m x 1250 m, is located between the KNPP Administrative Road

connecting the town of Kozloduy with the NPP on the north and Road No. 11 to the south

connecting Hurletz and Kozloduy. The site is located on slopping terrane between the second

and sixth loess terraces of the river Danube.

The NDF shall be able to accept and dispose of all Category 2a radioactive waste (RAW),

corresponding to what is usually referred to as Short-Lived, Low and Intermediate-Level

Waste (SL-LILW), arising in Bulgaria from the operation and dismantling of the national

nuclear facilities. According to forecasts the NDF will receive conditioned waste packed in

18,615 cubic-shaped concrete containers (i.e., waste packages). The waste packages have a

side length of 1.95 m and a weight of up to 20 tons. The total volume occupied by these

waste packages will be 138,200 m3. The radionuclide inventory is approximately 2.4 × 10

14

Bq.

The NDF design relies on a multiple barrier isolation system. The isolation function is

guaranteed by the system as a whole so that possible deficiencies of a barrier or its

degradation over the course of time are compensated by other barriers, thus ensuring that the

protection objectives are achieved. The safety of the facility is based on a defence-in-depth

concept consisting of a system of physical barriers and administrative measures.

For practical and operational safety reasons the repository facilities have been grouped into:

Disposal zone, in which the disposal cells are located

Building zone, in which the Waste Reception and Buffer Storage (WRBS) Building,

the site administration, control room and ancillary and support buildings are located.

The NDF has 66 disposal cells for waste package disposal. These disposal cells are located on

3 equal platforms, each with 22 disposal cells and their related systems. A first disposal

platform will be constructed prior to disposal start, the second approximately after 20 years,

and the third after 40 years of operation. The disposal cells are arranged in two lanes, each

with 11 disposal cells. The disposal cells are monolithic rectangular boxes with two inner

walls made of reinforced concrete, with a capacity for 288 waste packages emplaced in 3

chambers of 96 waste packages each (8 × 3 waste packages in plan, 4 layers in height). The

external dimensions of each disposal cell are 20.15 m long by 17.05 m wide. The height is

9.45 m measured from the foundation level up to the top of a full and sealed disposal cell.

Each storage platform will host 6,336 waste packages corresponding to about 20 years of

repository operation. The total disposal capacity of the NDF will be 19,008 waste packages.


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After a disposal cell is fully loaded with waste packages, it will be closed with a reinforced

concrete slab. During the disposal process and construction of the concrete slab the disposal

cell will remain covered by a mobile roof to protect loading and closure operations from

inclement weather conditions. The mobile roof also houses the overhead crane used to

emplace waste packages into their final position within each disposal cell. Before the mobile

roof is relocated to the next disposal cell position, hydro insulation protective measures are

applied over the exposed surfaces of the closed disposal cell.

A critical component of the disposal system is the Infiltration Control Network, which

consists of a pipe system to collect and control the water that could enter a disposal cell after

closure and interact with waste. The pipes are located in an accessible underground gallery

below the disposal cells. The system includes a pipe connection coming from each disposal

cell and a collection tank. Water is exclusively driven by gravity.

The Building Zone contains the entire infrastructure needed for the efficient and safe

operation of the NDF. The most important structure in this zone is the Waste Reception and

Buffer Storage (WRBS) Building. The WRBS Building is designed to receive each

radioactive waste package delivered to the NDF by truck. The building also provides a buffer

storage capacity of 120 waste packages that allows for the regulation and optimization of the

waste package flow to the disposal cells.

The NDF has a single main access that links the Kozloduy NPP road with the Building Zone.

The waste package disposal operation begins in the WRBS Building waste package

loading/unloading area. Here an internal transport vehicle is loaded with a waste package.

This vehicle goes to the assigned disposal cell and parks under the mobile roof where the

overhead crane lifts the waste package and hoists it to its storage position within the disposal

cell.

At the end of NDF operations, a long-term multi-layer cover, specifically designed to prevent

the intrusion of water into the disposal cells during the surveillance phase will be constructed.

3. Repository Licensing

Licensing of a radioactive waste repository in Bulgaria comprises separate licenses pursuant

to different legal instruments that govern the use of land and spaces, the environmental

impact of any industrial facility and/or construction works, as well as nuclear matters.

Ancillary permits including site security, fire protection, and protection of the groundwater

are required.

The NDF is specifically defined as a nuclear facility by the "Act on the Safe Use of Nuclear

Energy" (ASUNE) and must be licensed as per the ASUNE requirements. In addition, as all

industrial facility construction or infrastructure works in Bulgaria, the requirements of the

Act on Territory Arrangement also apply. The responsible licensing authority for the NDF as

nuclear facility is the Bulgarian Nuclear Regulatory Agency (BNRA). The responsible

organisation for licensing the NDF as per the requirements of the Act on Territory

Arrangement is the Ministry of Regional Development and Public Works (MRDPW). The

investment proposal on establishment of the NDF is also subject to an Environmental Impact

Assessment (EIA). The responsible authority for issuing decisions on the EIA is the

Bulgarian Ministry of Environment and Waters (MEW).

In addition, commitments under the EURATOM Treaty have to be followed. Article 37

requires consideration of the cross-border effects, in the case of the NDF especially upon

Romania. Article 41 requires projects relating to this article to be communicated to the


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European Commission (EC) in order to allow the EC to discuss with the Investor (i.e.,

SERAW) all aspects of the investment project.

3.1.Technical Design – TD

In order to assure sufficient quality in the development and review of the Technical Design

the document has been structured following Bulgarian requirements for investment projects

into 19 separate design parts. Additionally, in each design part of the Technical Design the

documentation, which corresponds to the respective buildings and facilities in the General

Layout Plan (GPL), is arranged in up to 23 separate sub-parts. The use of sub-parts is

optional, i. e. they are only considered if necessary. For example, the design part Architecture

is subdivided into 19 subparts describing in detail the fundamental connections and

parameters for the various facilities and common areas, while there is only one subpart for the

Design Part Geodesy providing the topographical base for the project. In total the Technical

Design documentation prepared by the Consortium fills around 50 folders with

approximately 6500 pages.

3.2.Intermediate Safety Analysis Report – ISAR

As previously stated, the license for repository construction requires completion of an

Intermediate Safety Analysis Report (ISAR). The ISAR assesses the behaviour of the

disposal facility and, in particular, the NDF potential radiological impact on humans and the

environment. The report considers potential pathways for radionuclide releases into the

environment and the resulting health effects. The ISAR shall provide convincing proof that

the NDF design, as laid down in the Technical Design documents, and the planned operations

are safe in accordance with applicable regulations, taking into account:

Characteristics of the site

Characteristics of the wastes to be disposed

Planned activities and personnel involvement

Characteristics of the risks associated with the NDF

4. Project Implementation

As previously described the NDF will be constructed in three stages. The auxiliary

installations and the first platform of disposal cells will be built during the first stage and will

provide a fully compliant disposal facility, but without the full complement of disposal cells.

Subsequently, a second and third stage of construction will expand NDF to full capacity. The

design takes into consideration ongoing operations during the second and third expansion

campaigns. To minimize construction interference on operations a secondary access road

(also used as an emergency evacuation route) will be used to access construction areas.

Full implementation of the project will include the following NDF lifecycle phases:

Operation: 60 years – receiving and emplacing waste packages

Closure phase after disposal end: 15 years – building of multi-layer cover;

Institutional control after closure: 300 years – surveillance of site.

5. OUTLOOK

In principle all licensing documents have been finalised and submitted by SERAW to the

responsible authorities. In expectance of a generally positive acknowledgement of the

submitted documentation SERAW signed a contract with NUKEM for the construction of the


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NDF on July 7th

, 2016. It is expected that construction works can commence at the beginning

of next year.

disposal of very low level waste & low level waste · id 65. disposal of very low level waste...

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