Draft Safety Case for the Management of Disused
Sealed Radioactive Sources in Tunisia
NLM-REP-14/197 Rev 1
Date: 2014-12-03
Prepared by: R Swart
Nuclear Liabilities Management
Necsa P.O. Box 582 Pretoria, 0001
South Africa
Form No.: NLM-DIV-FORM-00-002 Rev: 03
REPORT No.:
NLM-REP-14/197 Rev 2
DATE:
3 December 2014
DRAFT SAFETY CASE FOR THE
MANAGEMENT OF DISUSED
SEALED RADIOACTIVE SOURCES
IN TUNISIA
TITLE:
1.0
AUTHORIZATION
NAME
1.1
SIGNED
PREPARED
R Swart
REVIEWED
L Hordijk
APPROVED
GR Liebenberg
DATE
DISTRIBUTION
NO.
1
NAME
NAME
NO.
8
15
9
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3
10
17
4
11
18
5
12
19
6
13
20
7
14
21
2
NLM QA Records
NO.
IAEA
* = Distributed via E-mail
NAME
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DRAFT SAFETY CASE FOR THE MANAGEMENT OF DISUSED SEALED RADIOACTIVE
SOURCES IN TUNISIA
1.2
REVISIONS
Revision number
0
1
2
Reason for change
First issue
Revised following review
done with CNRP and CNSTN
personnel during IAEA
mission 13-17 October 2014
Revision based on Morocco
SC review and additional
comments
Preparer
R Swart
L Hordijk
L Hordijk
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DRAFT SAFETY CASE FOR THE MANAGEMENT OF DISUSED SEALED RADIOACTIVE
SOURCES IN TUNISIA
1.3
CONTENTS
1.0
AUTHORIZATION ....................................................................................................... 2
1.1
DISTRIBUTION ........................................................................................................... 2
1.2
REVISIONS ................................................................................................................. 3
1.3
CONTENTS ................................................................................................................. 4
2.0
PURPOSE ................................................................................................................... 7
3.0
SCOPE ........................................................................................................................ 7
4.0
REFERENCES ............................................................................................................ 8
5.0
ABBREVIATIONS ........................................................................................................ 8
6.0
DSRS MANAGEMENT DESCRIPTION IN TUNISIA .................................................. 8
6.1
LEGISLATION AND REGULATIONS RELATING TO THE MANAGEMENT OF DSRS
IN TUNISIA .................................................................................................................. 8
6.2
REGULATORY BODY ............................................................................................... 10
6.3
NATIONAL SAFETY CRITERIA .................................................................................11
6.4
WASTE OPERATOR ................................................................................................. 12
7.0
GENERIC ASSESSMENT CONTEXT ...................................................................... 13
7.1
PURPOSE OF THE SAFETY CASE ......................................................................... 13
7.2
SCOPE OF THE SAFETY CASE .............................................................................. 14
7.3
DEMONSTRATION OF SAFETY .............................................................................. 15
7.4
GRADED APPROACH .............................................................................................. 18
7.5
SAFETY STRATEGY ................................................................................................ 18
8.0
FACILITY AND PROCESS DESCRIPTION .............................................................. 19
8.1
SITE CONDITIONS AND FACILITY DESCRIPTION ................................................ 20
8.2
PROPOSED FACILITY OPERATION ....................................................................... 28
8.3
DSRS INVENTORY................................................................................................... 31
9.0
SAFETY ASSESSMENT ........................................................................................... 31
9.1
SAFETY ASSESSMENT CONTEXT ......................................................................... 31
9.2
SAFETY ASSESSMENT ENDPOINTS ..................................................................... 35
9.3
DEVELOPMENT OF SCENARIOS ........................................................................... 35
9.4
DATA USED AND ASSUMPTIONS MADE FOR THE SAFETY ASSESSMENT ...... 38
10.0
SAFETY ASSESSMENT ........................................................................................... 40
10.1
BASIC ENGINEERING ANALYSES .......................................................................... 40
10.2
QUANTITATIVE DETERMINISTIC ASSESSMENT OF WORKER DOSE................ 43
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DRAFT SAFETY CASE FOR THE MANAGEMENT OF DISUSED SEALED RADIOACTIVE
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10.3
DETERMINISTIC ASSESSMENT OF WORKER AND PUBLIC DOSE FOR
ANTICIPATED OPERATIONAL OCCURRENCES: .................................................. 47
10.4
THE FOLLOWING OTHER ACCIDENT SCENARIOS WILL BE CONSIDERED IN
THE TUNISIAN SAFETY ASSESSMENT: ................................................................ 48
10.5
COMPARATIVE DOSE ASSESSMENT: SAFRAN .................................................... 50
10.6
OPTIMIZATION OF PROTECTION: ASSESSMENT ................................................ 50
10.7
NON-RADIOLOGICAL HAZARD ASSESSMENT ..................................................... 52
10.8
ASSESSMENT OF THE IMPLEMENTED WASTE MANAGEMENT PRACTICE..... 53
10.9
MANAGEMENT SYSTEM ASSESSMENT ............................................................... 54
10.10
ASSESSMENT OF UNCERTAINTIES ...................................................................... 55
10.11
ASSESSMENT OF POSSIBLE PUBLIC EXPOSURES ........................................... 56
10.12
ASSESSMENT OF POSSIBLE ENVIRONMENTAL PATHWAYS ............................. 56
10.13
WASTE MANAGEMENT ........................................................................................... 57
11.0
HUMAN RESOURCES ............................................................................................. 57
12.0
IDENTIFICATION OF FACILITY SPECIFIC LIMITS AND CONDITIONS ................. 58
13.0
INTEGRATION OF SAFETY ARGUMENTS ............................................................. 59
13.1
FACILITY DESIGN AND ENGINEERING ................................................................. 59
13.2
FACILITY OPERATION ............................................................................................. 59
13.3
OPTIMIZATION OF PROTECTION .......................................................................... 59
13.4
WASTE MANAGEMENT PRACTISE ........................................................................ 60
13.5
INTEGRATED MANAGEMENT SYSTEM ................................................................. 60
13.6
UNCERTAINTIES ...................................................................................................... 60
14.0
COMPARISON WITH SAFETY CRITERIA AND CONCLUSIONS ........................... 60
15.0
ASPECTS REQUIRING CLARIFICATION AND RECOMMENDATION ................... 61
16.0
APPENDIX A: HOTSPOT OUTPUT FOR ACCIDENT SCENARIO 1 ....................... 62
17.0
APPENDIX B: HOTSPOT OUTPUT FOR ACCIDENT SCENARIO 2 ....................... 65
18.0
APPENDIX C: NATIONAL INVENTORY (CNRP) ..................................................... 67
19.0
APPENDIX D: CNSTN INVENTORY ........................................................................ 69
20.0
APPENDIX E: SAFRAN ASSESSMENT ................................................................... 76
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DRAFT SAFETY CASE FOR THE MANAGEMENT OF DISUSED SEALED RADIOACTIVE
SOURCES IN TUNISIA
List of Tables
Table
Table
Table
Table
Table
Table
Table
Table
Table
Table
Table
Table
Table
Table
Table
Table
Table
Table
1: References ................................................................................................................................ 8
2: Assumptions and justifications for quantitative deterministic assessment ...................................... 38
3: Basic Engineering Analyses ....................................................................................................... 40
4: Collection of DSRS at Interim Stores .......................................................................................... 43
5: Transport of DSRS ................................................................................................................... 43
6: Receipt of DSRS at CNSTN ........................................................................................................ 44
7: Temporary Storage of Cat 3 Sources and general activities in the CSF .......................................... 44
8: Conditioning Campaign 1 .......................................................................................................... 44
9: Conditioning Campaign 2 .......................................................................................................... 45
10: Transfer of Conditioned Waste Packages to the Waste Store ...................................................... 46
11: Worker Dose Summary ........................................................................................................... 46
12: Accident scenario 1 ................................................................................................................ 48
13: Accident scenario 2 ................................................................................................................ 49
14: Optimization of Protection: Assessment.................................................................................... 50
15: Quantitative Assessment: Current and Future Waste Management Practices................................ 53
16: Quantitative Assessment: Integrated Management System ........................................................ 54
17: Provisional Quantitative Assessment: Safety Case Uncertainties ................................................. 55
18: Aspects Requiring Clarification/Recommendations ..................................................................... 61
List of Figures
Figure 1: Architect’s rendition of the proposed Centralized Storage Facility at CNSTN ................................. 22
Figure 2: Waste Treatment and Storage Facility Layout ............................................................................ 25
Figure 3: Proposed Radiological Area Classification of Facility ................................................................... 26
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DRAFT SAFETY CASE FOR THE MANAGEMENT OF DISUSED SEALED RADIOACTIVE
SOURCES IN TUNISIA
2.0
PURPOSE
The purpose of this document is to describe the various elements of the safety case for the
management of disused sealed radioactive sources in Tunisia.
3.0
SCOPE
Activities regarding the management of DSRS in Tunisia are at this stage limited to storage of DSRS
at user facilities. Legislation is, however, currently being developed to address various issues
regarding the management of radioactive waste in Tunisia. CNSTN is busy constructing a new DSRS
management facility at their site at Sidi Thabet. This facility may also in future be regarded as the
national centralized DSRS management facility. For the purposes of this safety case it will be
assumed that the CNSTN DSRS management facility will eventually become the national centralized
facility.
The scope of the safety case will include all the aspects or arguments that will ensure the safety
of all management activities relating to DSRS to be performed by the National Centre of Nuclear
Sciences and Technologies (CNSTN). This will include amongst others a description of the current
legislation and regulations pertaining to the safe management of DSRS in Tunisia, description of
the regulatory function as well as the appointed waste operator, site, facility and activity
description, waste inventory, the context for the evaluation of the safety case, a safety
assessment for normal and accident scenarios, a safety case compliance assessment, unresolved
issues, limiting conditions, as well as management systems and procedures required to ensure
compliance to set safety criteria and to sustain an acceptable level of safety. The safety
assessment to be performed will be of a more generic nature due to no DSRS management
activities other than storage taking place in Tunisia.
The Safety Case and associated Safety Assessment for the management of DSRS in Tunisia will take
the IAEA requirements with regards to predisposal management of radioactive waste [1] into
consideration and will be developed and performed in accordance with the IAEA requirements and
recommendations as described in [2]. The safety criteria will be taken from international safety
standards and used as a basis for evaluation of safety and protection. [4]
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DRAFT SAFETY CASE FOR THE MANAGEMENT OF DISUSED SEALED RADIOACTIVE
SOURCES IN TUNISIA
4.0
REFERENCES
Table 1: References
1
Number
Title
GSR Part 5
IAEA, Predisposal Management of Radioactive Waste, IAEA
Safety Standards Series No. GSR Part 5, IAEA, Vienna (2009).
2
GSG-3
IAEA
Safety
Standards
(2013),
Safety
Case
and
Safety
Assessment for Predisposal Management of Radioactive Waste,
Vienna.
3
NLM-REP-14/016
Mission Report – Safety Case Development in Tunisia
4
GSR Part 3
IAEA, Radiation Protection and Safety of Radiation Sources:
International Basic Safety Standards - 2014
5.0
ABBREVIATIONS
CNRP – Tunisian National Centre of Radiation Protection
CNSTN – National Centre of Nuclear Sciences and Technologies
DSRS – Disused Sealed Radioactive Sources
CSF – Central Storage Facility
RPO – Radiation Protection Officer
IAEA – International Atomic Energy Agency
6.0
6.1
DSRS MANAGEMENT DESCRIPTION IN TUNISIA
Legislation and Regulations Relating to the Management of DSRS in Tunisia
Tunisia currently has the following laws, Decrees and Orders pertaining to Radiation Protection;
Transport of Hazardous Waste and Radiological Waste Management. They are the following:
6.1.1
Tunisian Laws
-
Law 81-51 of 18th June 1981 (JORT N°42 Tunis 1981: 471-472) related to the protection
against ionising radiation;
-
Law 81-100 of 31st December 1981 (JORT N°84 Tunis 1981: 30046) related to the creation,
missions and attributions of National Centre of Radiation Protection (CNRP);
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DRAFT SAFETY CASE FOR THE MANAGEMENT OF DISUSED SEALED RADIOACTIVE
SOURCES IN TUNISIA
-
Law 88-91 of 2nd August related to the creation of the (ANPE) National Agency for
Environmental Protection
-
Law 96-41 of 10th June 1996 (JORT N°49 Tunis 1996: 1192-1196) related to dangerous waste
management materials including radioactive waste
-
Law 97-37 of 10th June 1997 (JORT N°45 Tunis 1997: 1020-1021) related to transport by road
of dangerous materials including radioactive materials
6.1.2
Tunisian Decree’s
-
Decree No. 86-433 of 26th March 1986 (JORT N° 42 Tunis 1986: 492-497) on protection against
ionizing radiation
-
Decree No. 2000-2339 of 10th October 2000 establishing the list of hazardous waste (including
radioactive waste)
-
Decree No. 2002-2015 of 4th September 2002 laying down technical requirements for
equipment and management of vehicles used to transport dangerous goods by road
-
Decree No. 2005-3079 of 29th November 2005 establishing the list of hazardous materials to be
transported by road necessarily under the control and with the support of security units
6.1.3
Tunisian Orders
-
Order of Minister of Public Health (10/09/1986) about the information and particulars to
accompany applications for approval radioactive sources and radiation devices
-
Order of Ministers of Interior and Transport (18/03/ 1999) about the model of safety record
and the instructions for the transport of dangerous goods by road
-
Order of Minister of Transport (19/01/ 2000) fixing the danger labels and markings for the
transport of dangerous goods by road
-
Order of Ministers of Interior and Transport (19/05/2000) determining the hazardous materials
for which carriage must obtain a road map, the model of the sheet and the conditions of issue.
The abovementioned laws and decrees are however outdated and is in need of being reviewed to
be in line with current International Standards and Regulations.
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DRAFT SAFETY CASE FOR THE MANAGEMENT OF DISUSED SEALED RADIOACTIVE
SOURCES IN TUNISIA
Tunisia is in the process of developing a National Radioactive Waste Management Policy as well as a
National Radioactive Waste Management Strategy. These documents are not yet finalised. An
example of a draft policy and strategy has been provided to the CNRP by the IAEA. These
documents still has to be studied by the Necsa team but it seems from the discussions that all the
important aspects required for such documents have been included e.g. waste endpoints; provision
of funds, decommissioning plan and allocation of responsibilities with regards to protection of
workers, and population etc.
Similarly a Radioactive Waste Management Agency will be constituted under the new Nuclear Law.
This entity will however be a separate agency from the CNSTN.
Orphan sources are addressed in the new Nuclear Draft law; the Tunisian Government will still
however be ultimately be responsible for orphan sources.
CNSTN is in the process of establishing a new facility for the management of DSRS at their Sidi
Thabet site. This facility, although established for the CNSTN inventory, has the capacity to feature
also as a national centralised facility.
For a CSF to be established a mandate from government will have to be obtained. Should the
CNSTN facility be decided on as the national CSF then CNSTN will have to submit a complete safety
case for the facility. This safety case could, with finalization suit the purpose.
6.2
Regulatory Body
Currently CNRP is fulfilling the role as the Regulator. The new draft Nuclear law does not make
provision for the creation of a National Authority; i.e. a National Regulator. The aforementioned is
however stipulated in a governmental Decree. The new regulatory authority will fall under the
mandate of the President of the Government. The Constitution will prescribe which competencies
will be necessary for the Regulator and CSF Operator to be able to manage and implement all
requirements as stipulated by Tunisian law in line with International Standards and guidelines.
The national waste management policy should assign the new regulatory authority the following
responsibilities in terms of radioactive waste management:
-
Translating the policy into the national legislation.
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DRAFT SAFETY CASE FOR THE MANAGEMENT OF DISUSED SEALED RADIOACTIVE
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-
Regulate and control the use of nuclear energy and ionizing radiation.

Development and maintenance of a national registry for all radiation sources and
radioactive materials that in use or imported or exported or disused sources including
orphan sources in any form (sealed or not sealed sources)
-
Enforce the implementation of the regulations on radioactive waste management and
spent fuel, and other national governmental organizations, such as the MOE.

Ensure the fulfilment of requirements of public safety, radiation protection, and
nuclear safety and security.

Granting licenses and permits for radiation institutions, nuclear facilities, and workers
in the radiation and nuclear fields.

Issue regulations related to the following :
o
The safe use of nuclear energy,
o
Safety and security of radiation sources,
o
Radiation protection,
o
Management of radioactive waste and spent nuclear fuel,
o
Transport of radioactive materials,
o
Extracting, mining and processing of the nuclear materials.
CNRP currently performs inspections at the various end-users of radioactive sources on a routine
basis. These inspections are mainly based on users of sources in the medical field. DSRS are also
conditioned (without removal from the working shield) by the CNRP by being placed in a cement
matrix and stored in drums at the premises of the end-user.
6.3
National Safety Criteria
The various regulations pertaining to the safety criteria in Tunisia are still in draft form and were
not available at the time of the development of this safety case. It was, however, confirmed that
Tunisia is following the international guidelines in this regard as described in [4]. These criteria
are as follows:
6.3.1
Protection of Workers (Planned Exposure Situations)
For occupational exposure of workers over the age of 18 years, the dose
limits are:
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DRAFT SAFETY CASE FOR THE MANAGEMENT OF DISUSED SEALED RADIOACTIVE
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a)
An effective dose of 20 mSv per year averaged over five consecutive years (100 mSv in 5 years)
and of 50 mSv in any single year;
b) An equivalent dose to the lens of the eye of 20 mSv per year averaged over five consecutive
years (100 mSv in 5 years) and of 50 mSv in any single year;
c)
An equivalent dose to the extremities (hands and feet) or to the skin of 500 mSv in a year.
Additional restrictions apply to occupational exposure for a female worker who has notified
pregnancy or is breast-feeding. Notification of the employer by a female worker if she suspects that
she is pregnant or if she is breast-feeding shall not be considered a reason to exclude the female
worker from work. The employer of a female worker, who has been notified of her suspected
pregnancy or that she is breast-feeding, shall
adapt
the
working
conditions
in
respect
of
occupational exposure so as to ensure that the embryo or foetus or the breastfed infant is afforded
the same broad level of protection as is required for members of the public.
For occupational exposure of apprentices of 16 to 18 years of age who are being trained for
employment involving radiation and for exposure of students of age 16 to 18 who use sources in
the course of their studies, the dose limits are:
a)
An effective dose of 6 mSv in a year;
b) An equivalent dose to the lens of the eye of 20 mSv in a year;
c)
6.3.2
An equivalent dose to the extremities (hands and feet) or to the skin of150 mSv in a year.
Protection of Public
For public exposure, the dose limits are:
a. An effective dose of 1 mSv in a year;
b. In special circumstances, a higher value of effective dose in a single year could apply, provided
that the average effective dose over five consecutive years does not exceed 1 mSv per year;
c. An equivalent dose to the lens of the eye of 15 mSv in a year;
d. An equivalent dose to the skin of 50 mSv in a year.
6.4
Waste Operator
The current situation in Tunisia is that DSRS is stored at the facilities of the end users. This is a
requirement written into current legislation. Most end of life sources used in the Medical
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DRAFT SAFETY CASE FOR THE MANAGEMENT OF DISUSED SEALED RADIOACTIVE
SOURCES IN TUNISIA
institutions (Category 1 DSRS) is however returned to the supplier. All users of Sealed
Radioactive Sources are authorized by the CNRP.
Although the legislation on the management of radioactive waste is in the process of
development, it is envisaged at this stage that the responsibility for the management of DSRS on
a national level will in future lie with CNSTN. CNSTN is currently also constructing a new DSRS
management facility where the following DSRS management activities will take place:.
7.0
7.1
-
Collection and transportation of DSRS form user facilities to the proposed CSF
-
Receiving of the DSRS at the CSF
-
Temporary storage of DSRS
-
Conditioning of DSRS
-
Pre-disposal storage of DSRS
GENERIC ASSESSMENT CONTEXT
Purpose of the Safety Case
A safety case is a living document or set of documents that should be developed already during the
design stages of a facility or the planning stages of an activity. This will then form the basis for
regulatory decisions as well as operational decisions.
In the case of Tunisia DSRS management activities are currently limited mainly to storage of DSRS
in their working shields (some placed in their working shields in a cement matrix) at the user
facilities. A DSRS management facility at CNSTN is in the final stages of construction and
administrative activities relating to the management of DSRS are already taking place. The purpose
of the assessment will therefore be mainly to perform a prospective assessment of the expected
centralized DSRS management activities with the view of proposing additional measures to enhance
the safety and security of the facility and the DSRS management activities.
The following specific aspects will be addressed in this safety case:
i)
Demonstration of the safety of the proposed CNSTN Waste Management Facility
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DRAFT SAFETY CASE FOR THE MANAGEMENT OF DISUSED SEALED RADIOACTIVE
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ii) Demonstration of the safety of various proposed radioactive waste management activities that
will be performed by CNSTN. These activities include collection at user’s facilities, transport of
DSRS to the CNSTN Waste Management facility, receiving and characterization of the DSRS,
temporary storage and conditioning and longer term storage.
iii) Optimization of the respective waste management activities described above.
iv) Management systems implemented in support and to ensure the safety of the respective waste
management activities described above.
v) Definition of Limits, Controls and Conditions that will be applicable to the respective activities
described above.
vi) Input to and further development of existing monitoring and programmes and activity
procedures.
7.2
Scope of the Safety Case
The scope of the safety case for Tunisia is limited due to existing lifecycle stage of the facilities i.e.
already constructed facility that is not yet operational and will therefore be focused on the as build
facility and the future operational aspects of the facility which are defined as the:

Collection and transport of DSRS to the proposed CSF at CNSTN;

Receiving, identification, characterization and handling of DSRS when it arrives at the
centralized facility at CNSTN;

Temporary storage of the DSRS at the centralized facility at CNSTN;

Conditioning of the DSRS for long term storage.

Handling and placement into final storage.
This version of the safety case will therefore not address the following:

The development of waste management options and strategies and its scientific and technical
bases.

The development of facility designs.

The siting including the site characteristics details and evaluation of possible sites.

The construction and commissioning of such facilities.

Decommissioning or decommissioning planning of facilities (Should be addressed in follow-up
revisions).
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7.3
Demonstration of Safety
Taking cognisance of the scope of the safety case as defined in 7.2 above and the application of the
graded approach as defined in 7.4 the safety of the new waste management facility will be
evaluated and demonstrated by the following:
7.3.1
Basic Engineering Analysis
A qualitative assessment will form the basis for the basic engineering analyses which will mainly
cover the following:

Basic site characteristics and credible external events have been considered in the design of the
waste management facilities to ensure structural stability.

Quality assurance has been considered in the design, construction, maintenance and
modification the waste management facilities. The following needs to be demonstrated:
-
The facilities have been designed and constructed in accordance with acceptable national
construction codes and standards.
-
Inspection and maintenance plans exist and are implemented
-
Formal processes are defined and implemented for the evaluation, approval and
implementation of modifications (Change management)

Safety and security aspects were considered in the design of the facility and the approach to
demonstration of compliance refers to mainly the existence of the following features:
-
The characteristics of the walls allow ensuring a level of dose rate that complies with the
restriction for public exposure (1 mSv/a) even for the maximum anticipated inventory and
occupancy of 400h per annum i.e. 2.5µSv/h.
-
The lighting system will be adequate and permits the performance of operations in a safe
manner.
-
Physical delineation of areas designed for storage and for the main waste (DSRS)
management operations are isolated, this way it is ensured the appropriated segregation
of materials optimizing worker’s exposure during operations.
-
Each delineated area has a sufficient physical space that ensures a minimal probability of
accident occurrence during waste management operations and package handling.
-
Storage building areas were designed under the principle of labyrinth, which contributes
to optimize the exposure of workers. (Waste operations are not taking place in the area
where DSRS are stored).
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DRAFT SAFETY CASE FOR THE MANAGEMENT OF DISUSED SEALED RADIOACTIVE
SOURCES IN TUNISIA
-
Packages with sources are stored in a manner such that packages are not in contact with
the floor or interior surface of the building walls. This limits the potential of corrosion of
packages/containers and allows for inspection and control operations.
-
Unconditioned radioactive sources are stored in storage systems ensuring normal
operation and minimizing probability of accidents. Their main characteristics are:

Storage capacity is greater than current and foreseen needs of management.

It ensures source segregation. In this way, periodic inspection and radiological
monitoring of the storage building and of the waste drums/packages is facilitated.

Its structure resists the maximum load of the sources that are intended to be
stored.
-
There is a vault with special shielding structure that minimizes worker’s exposure for the
storage of sources of greater or unknown activity that could have not been conditioned.
-
For situations of operational occurrences and accident due to internal operational factors,
the engineering systems ensuring safety are:

Floor and wall finish allow easy decontamination

The segregation of the different areas limits the potential dispersion of any
contamination.

In case of a potential surface decontamination using liquids there is a collection
system inside the facility that prevents its release to the environment. The system
has a retention tank that permits environmental monitoring before releasing to the
environment.

-
The facility has its own fire detection and fire-fighting equipment.
The facility design makes provision for physical security features commensurate with the
anticipated security threat. Design features include the following:

Robust building construction with high integrity doors and locks to the treatment
and storage areas.

Buildings are equipped with intrusion alarms.

The buildings have vehicle access points. A separate personnel door is provided to
segregate personnel from vehicle movements.

No windows are provided in the storage areas so as to improve its shielding and
security performances.

Security fence around facility.
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7.3.2
Demonstration of the safety of various radioactive waste management activities performed by CNSTN.
Quantitative and qualitative assessments will be performed to assess the impact of the waste
management activities as listed in 7.2 above. Results will be assessed in terms of the safety criteria.
The following specific assessments will be performed:

For normal operation; quantitative deterministic assessment of worker dose due to the range of
activities by various occupational groups of CNSTN;

For anticipated operational occurrences: quantitative deterministic assessment of worker and
public dose as applicable;

All other credible occurrences; a quantitative and qualitative assessment of the impact of other
occurrences and the listing of specific preventative and mitigating measures.
(At the time of the country visit to Tunisia no management activities were yet taking place. Real time
measurements could therefore not be obtained for the activities. Radiological assessment will be based
using a conservative approach combined with a realistic approach where possible. The assessment will
rely on typical exposure data collected during similar type exercises elsewhere taking cognizance of the
activities and types of DSRS mostly handled.)
7.3.3
The results from the quantitative and qualitative assessment as defined in 7.3.2 above will also be
compared to the proposed target and objectives set for the optimization of protection.
7.3.4
A qualitative assessment of the non-radiological hazards of the facilities and the listing of specific
control measures will be performed.
7.3.5
A qualitative assessment of the implemented waste management practice; – The approach to waste
management will be regarded as a contributing factor to safety.
7.3.6
A qualitative assessment of the availability, level of implementation of an integrated management
system to ensure a sustained level of safety during the operational phase of the facilities will be
undertaken. This assessment will focus on RP, work procedures, QA aspects and processes for the
management of limits and conditions.
7.3.7
Uncertainties inherent to the assumptions made in the quantitative assessments or any other
uncertainties identified during the safety assessment will be evaluated to determine its impact on
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safety. Uncertainties with a significant impact on safety will be listed with recommendation for its
management.
7.4
Graded Approach
A graded approach is applied for defining the extent and depth of this safety case. The main factors for
justification of a limited approach towards the safety assessment are the following:

The simplicity of the activities involving the management of DSRS. Most of the activities
involving DSRS entail handling of the DSRS inside robust working shields which limits external
exposure potential.
•
The radiological hazard when undertaking the various management activities involving DSRS
(especially Cat 3-5) can be regarded as low. This once again, as described above, is due to the
simplicity of the activities handling only DSRS inside their working shields. The only time that
bare DSRS will be handled during normal operational conditions in any DSRS management
facility is during source conditioning operations. In such instances the risk is reduced by
performing the work in accordance with specific works procedures and under work permit
systems where there is permanent radiation protection controls in place.

Inherent high level of passive safety associated with the DSRS management operations and the
limited reliance on active protection systems.
7.5
Safety Strategy
The strategy for safety refers to the approach that was taken in the facility design and all the
respective DSRS management activities to comply with the regulatory requirements and to ensure
that good engineering practice has been adopted and that safety and protection are optimized.
The safety case for Tunisia will take into account the following safety strategies during the
management of DSRS:
-
Defense in Depth – In this instance care is taken to ensure that multiple safety layers are
available. This principle is followed for all the respective management processes for
instance:
o
To ensure containment of DSRS during transport from user facilities all over Tunisia.
The DSRS inside their working shields will be packaged inside a specially manufactured
secondary container. This box will form a secondary barrier should the DSRS for some
or other reason gets separated from its working shield during transport.
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o
After receiving of a unit containing a DSRS at the CNSTN CSF the RPO will first take
measurements to ensure that the DSRS is still within the working shield.
o
During conditioning the RPO will continuously monitor and ensure that no exposure
levels exceed the criteria set for the respective activity.
-
Shielding – Shielding is used to ensure that doses to workers and also the public, as low as
possible. This is applicable to all activities including storage. One example is that the
transport containers mentioned above under defense in depth are also lined with lead in
order to keep the exposures to members of the public during transport at the lowest
possible levels. Shielding is selected based on the type of DSRS and the sources activity.
-
Approaches during the management of DSRS in Tunisia. The proposed national waste
management strategy will highlight the principles of waste minimization and avoidance, reuse or re-processing of waste, safe and secure storage and conditioning and final disposal
of waste. These principles will be followed by CNSTN for the management of DSRS:
o
Secondary radioactive waste is only expected when a leaking source is found or during
accident scenarios when a source is damaged. The generation of secondary waste
during such an incident will be minimized by isolating the source in a secondary
containment to prevent further contamination.
o
The draft national policy encourages a “return to supplier” principle when procuring
sources. This principle will be followed as far as practical achievable in order to avoid
radioactive waste.
o
Storage of DSRS will only take place inside proper containment such as the original
working shields or another type of suitable containment. For storage of the DSRS in
the CSF personnel shall take into account the ambient dose rates measured in the
respective storage rooms. The DSRS will be stored in a way that will provide the lowest
possible dose rate in the areas that personnel normally occupy when they enter the
respective storage rooms.
o
Recycling of sources stored at the CNSTN CSF will be encouraged and applied by
CNSTN.
8.0
FACILITY AND PROCESS DESCRIPTION
A new facility for the management of radioactive waste is currently under construction at the
“Centre National des Sciences et Technologies Nucleaires” (CNSTN) in Sidi Thabet, Tunisia. This
facility has been constructed for the management of the national radioactive waste inventory. It was
decided, however, after discussions with the regulatory authority, that the store could not currently
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be used as a national waste management facility due to the absence of applicable legislation and
regulations to this extent.
8.1
8.1.1
Site Conditions
General description of the site
The land where the Facility is constructed, is on the CNSTN site which is located in Sidi Thabet,
which is located north of Tunis and about 20 km from the Tunis city centre. About 8km from the
Mediterranean sea.
There are various building and laboratories on the site which is about 5km2. It is located in a semi
farming area.
The site is located against a slight ridge and the CSF is located highest against this ridge. Thus any
water flow from the facility will be into the site.
Elaborate and correct this data
8.1.2
Demography
The demographic data is based on census studies performed during xxxx. The main
demographical findings are the following:
Total population of Sidi Thabet as per 2004 is 8909
Distribution of population in a radius 20 km around the site:
 0 - 1 Km
: X Inhabitants
 1 - 2 km
: X Inhabitants
 2 – 5 km
: X Inhabitants
 5 – 20 km
: X Inhabitants.
Population growth: to be included
8.1.3
Meteorology
•
Local meteorological conditions: To be included
•
Wind conditions: To be included.
•
Temperatures: To be included
•
Humidity: To be included
•
Cloud cover: to be included
•
Evaporation: to be included
•
Precipitation: to be included
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8.1.4
Site Geology and Geohydrology
Include relevant Geological and hydrogeological information on the site, that was considered for
the design and construction of the waste management facility.
8.1.5
Site Seismology
Include info about seismology of the site. Specify the actual reference seismic event that was
used in the design and construction of the waste management facility.
8.1.6
Aircraft Crash Probability
To be included here
8.2
Facility Description
The purpose of the storage facility will at this stage be limited to the management and storage of
radioactive waste generated at the CNSTN facilities only. The radioactive waste inventory of CNSTN
consists mainly of disused sealed radioactive sources (DSRS).
The current situation in Tunisia is that DSRS is stored at the facilities of the users. This is a current
strategy by CNRP in awaiting the national policy and strategy on Radioactive Waste management.
Most end of life sources used in the Medical institutions (Cat 1 and 2 DSRS) is however returned to
the supplier. All users of Sealed Radioactive Sources are authorized by the CNRP. All DSRS collected
in Tunisia should however ultimately be brought to the proposed CSF for storage and further
management.
The facility at CNSTN can be described as follows:
The facility consists of a single building with an administrative block with two levels containing the
building main entrance, various offices, laboratories, parking bay for the mobile laboratory, ablution
facilities and finally the clean entrance sides to change rooms (see picture in Figure 3). The change
rooms lead into the main waste management facility.
The radioactive waste management facility consists of a concrete structure of which the main
operational areas consist of the following:
-
Receiving and treatment hall (Main Hall). The surface area of this hall is about 160 m2 and a
height of 7.1 m. The external walls of the facility are constructed of concrete (0.40 m thick up to a
height of 5 m and 0.2 m thick from there). The roof is constructed from a combination of hollow
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blocks and concrete (0.19 m hollow blocks and 8 cm concrete). Allowance has been made for
windows at a level about 0.5 m from the roof.
-
Storage room. This room has a surface area of about 46 m2 with concrete walls and is about 3.1 m
high. The external walls of the storage facility are 0.27 m thick. The internal wall bordering the
measuring laboratory is also 0.27 m thick while the internal wall bordering the main hall is 0.20 m
thick. The roof of this storage area is of the same design as the roof the main hall.
-
Laboratory (analysis, measuring etc.). This room, adjacent to the storage room and accessed from
the main hall, has concrete walls and a surface area of about 30 m2.
-
Change House and amenities and access to main hall and storage facility.
An architect’s rendition of the proposed facility currently under construction at CNSTN of the CSF is
presented below as Figure 1.
Figure 1: Architect’s rendition of the proposed Centralized Storage Facility at CNSTN
8.2.1
Facility Design and Construction.
Basic information regarding design considerations, applied design and construction codes and
standards needs to be obtained to justify the new facility design building integrity and stability. Design
and construction documentation e.g. facility layout, civil design, electrical design, etc, design review
and construction reports as well as certificates of conformance should be referenced here.
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Safety related assumptions on which the design of the facilities has been based also need to be
obtained and listed here. (e.g. The building structure and associated civil infrastructure has been
designed to cope with external environmental events. The design basis for these external
environmental events will consider events with a return frequency of 1 in 100 years using data for the
local area and will provide conservative design margins)
It should be confirmed that the seismic hazard of the region has been taken into account in the design
of the facility. Basic information regarding the ground accelerations level that the facilities would be
able to withstand and its justification should be provided. -to be confirmed.
8.2.2
Main Safety and Security Related Design Features
8.2.2.1 Building structure

The foundations, columns, walls and roof have been designed to support all super imposed
structural loads as well as all applicable dead loads; - Provide reference to design
documentation covering loads used as design inputs

The floor slab is able to support the concentrated point loads of the waste containers 5 t per
m2, and an impact load of resulting from accidental dropping of waste container of 5 t from a
height of 2 m, as well as live loads of vehicles/equipment used to load the packages; - Provide
reference to design documentation covering loads used as design inputs

The slab will be sufficiently thick around the building perimeter to support the walls and locally
around all internal stanchions; - Provide reference to design documentation covering loads used
as design inputs

Rain water will be prevented from entering the buildings by surface contouring and drainage
channels around the buildings. – Provide reference to design documentation

Resistance to water penetration from the ground will be provided by a polyethylene damp proof
membrane to the underside of the slab; - Provide reference to design documentation

The interior construction of the building will be such that the risk of any liquids being released
to the environment is minimized; - Provide reference to design documentation

The buildings will be provided with an internal floor drain system to direct any internal liquid
traces generated to a sump pit of capacity at least 1m3. The floor will be sloped to facilitate
movement of liquid away from the storage areas toward the floor drains. There will be provision
for inspection of the sump and sampling of accumulated liquid. - Provide reference to design
documentation
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
The floor slab will have a steel floated finish with an epoxy paint coating to provide a hard
wearing and decontaminatable surface; - Provide reference to design documentation

Where ducts, pipes or cables that pass through walls or the floor, suitable means to
accommodate expansion and provide fire resistance will be provided and they will be such that
the structural and fire integrity of the building is not impaired. Water proofing will be applied at
the entry point to a building - Provide reference to design documentation
8.2.2.2 Shielding

The design and construction ensure the required shielding is provided for (see dose assessment
assumptions) and that no major cracks or shine paths are present in the as constructed
building. To be confirmed. Individual packages will be shielded by other packages, internal
building structures or by concrete blocks. The external wall thickness of the storage area has
been increased to 270mm to provide optimum shielding for the stored DSRS. The radiation
outside the facility will depend on various factors, which include: external wall thickness and
density, external radiation on packages, nuclides, package design, total facility inventory and
storage configuration.
Due to the number of factors, it is not possible to determine the
maximum nuclide and activity inventory of the facility which will ensure that the radiation
outside the facility will be acceptable for possible public exposure. For this reason routine RPO
surveys outside the facility are required. These then needs to confirm that radiation levels
remains less than 10 x natural background levels (or any limit prescribed by the regulator).
8.2.2.3 Access and Physical Security

Physical security will be provided primarily by a number of passive physical barriers including a
site perimeter fence, a site security access point, strong building construction, high integrity
doors and locks to the treatment and storage areas. Buildings will be equipped with intrusion
alarms. –to be confirmed

The building has one vehicle access point. A separate personnel door will also be provided to
segregate personnel from vehicle movements. In the case of the waste treatment facility and in
the interest of security only the personnel door can be opened from outside. -confirmed

Allowance has been made for windows at a level about 0.5 m from the roof for the purposes of
providing additional natural lighting and will not impact on the shielding and security
performances. These windows cannot open.
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8.2.2.4 Waste Treatment and Storage Facility Layout
The layout of the waste treatment facility is illustrated in Figure 2 below;
Deconta
mination
bay (with
drain to
tank)
Figure 2: Waste Treatment and Storage Facility Layout
The following activities will take place in the Receiving and Treatment Hall:
-
Receiving of DSRS
-
Characterisation and conditioning of DSRS
-
Decontamination in Decontamination Bay
Areas in which DSRS are present are subject to radiation protection control measures. Interior walls
will separate the storage area from the receiving and conditioning area and provide radiation shielding.
Access to the storage area will be via a labyrinth type arrangement (wall in front of the storage room
door) to provide easy access and at the same time reduce radiation shine. The proposed classification
of the radiologically controlled areas of the facility is illustrated in Figure 3 below.
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Controlled
area
Controlled area
Controlled area
Decontamination bay:
Controlled area for both
contamination and
radiation during
conditioning campaigns
and decontamination work
All four areas: change
house, laboratory,
storage and receiving
area: always controlled
area for radiation
Figure 3: Proposed Radiological Area Classification of Facility
All the operational areas will be controlled areas. Due to the nature of DSRS, there is very small
change of leaking sources and possible spread of contamination.
However during the possible
decontamination work and conditioning of the sources there is a possibility of contamination, for this
reason access to the controlled area will at that stage be both controlled for radiation and
contamination. Thus typically personnel entering the area during conditioning of sources will require
wearing protective clothing (overall), overshoes and gloves. On exit of the area, each person will
have to be checked for contamination in the change room.
During these activities the vehicle
entrance to the receiving area will remain closed and no vehicles would be allowed to enter the
facility. All personnel access will the through the change rooms.
the facility RP programme.
This will need to be prescribed in
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Drums/packages will be placed in a manner such that packages do not contact the interior surface of
the building walls and so as to allow access to visually inspect packages and wall surfaces for
degradation and to allow for easy retrieval;
8.2.2.5 Fire protection
Fire protection is provided by the utilisation of construction materials that are not flammable and by
forbidding any flammable materials to be introduced into the store. Fire detection and fire fighting
equipment will be installed. Such equipment will be tested and maintained. High quality electrical
equipment complying with national quality standards is installed in both the buildings. The site will be
maintained clear of vegetation and combustible materials will not be stored on the site. Fire detection
equipment will be installed, fighting equipment will be provided and strict compliance will be
maintained with national and local fire regulations.
8.2.2.6 Ventilation
The waste management facility building will be provided with natural ventilation; outlets will be
located high on the building walls and covered with grids to prevent the access of animals, birds and
insects. No ventilation is provided for in the storage room. A local extraction hood will be installed at
the conditioning area. The extracted air will be filtrated through a High-Efficiency Particulate Filter
(HEPA) to prevent the potential release of radioactive materials. The extracted air will be released
outside the facility into the atmosphere. These filters have a high efficiency to capture any possible
contamination which could be released when a damaged or leaking source is handled and
conditioned.
8.2.2.7 Electrical power and Lighting
Electrical power is provided for lighting, small power tools and security detection/warning equipment.
All installations and equipment are of high quality and will comply with national standards. Good
levels of lighting is provided throughout the treatment and storage facilities and quality, long life
components are used to reduce maintenance needs. (Provide a reference to an installation report or
Electrical Certification of Compliance)
8.2.2.8 Mechanical handling equipment
Readily available and good quality manually operated mechanical handling equipment need to be
available once the facility becomes operational. Such equipment is subject to national
regulation/requirements as applicable to statutory equipment (e.g. maintenance, inspection and load
test records) and is used/operated by trained/licensed operators.
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8.3
Proposed Facility Operation
Due to the fact that the proposed Central DSRS Management Facility is currently still under
construction at CNSTN, the Facility Operations that will take place in the future will rely on typical
data collected during similar type operations elsewhere taking cognizance of the activities and types
of DSRS mostly handled.
Proposed operational activities within the waste management facility involve reception, treatment,
conditioning and emplacement of packages in the store, inspection and characterisation of DSRS,
equipment and the stored packages and maintenance of the building and equipment.
At this stage however no neutron sources will be conditioned. On review of the current Tunisian
inventory there are not many neutron sources, and for that reason the possible dose impact of
these will not be included in the safety case.
It is possible that some minor repairs may be carried out from time to time to the source housings,
packaging or containers. The facility design is such as to make these operations simple and easy to
undertake in the least time possible. Written operational procedures will be drawn up to ensure the
activities are carried out safely and in the least time reasonably possible and to optimize safety and
protection and to ensure that no individual dose constraints or limits are exceeded.
Operational radiation protection, maintenance and inspection programmes will be formally
documented and approved, an incident reporting system and emergency plans will be drawn up and
approved. These programmes will be drafted and implemented based on and justified by this safety
case. Reference the applicable procedures.
Records will be maintained of all operational activities, packages and equipment will be clearly
marked and labeled and an inventory maintained of all equipment, DSRS and waste placed in the
store.
8.3.1
Proposed Waste Management Facility Operation

DSRS will be collected from the current owners by the Central Waste Management Facility
personnel.
These will be checked, loaded and then transported in accordance with the
applicable IAEA transports requirements (IAEA SSR-6: Regulations for the safe transport of
radioactive material)
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
The transported DSRS will be received at the Central Waste Management Facility at CNSTN.
The sources will be surveyed, off loaded, inspected and segregated. Approximately 10
consignments with a total of 20 sources will be received annually.

The DSRS will be transferred to the storage room in the Central Waste Management Facility
at CNSTN.

The received units will be placed in the receiving hall for temporary storage.

Conditioning of the standard Category 3 sources will be done on a batch basis. During this
operation the sources in their working shields will be removed from their storage location
and placed on a working area equipped with a lead shield (this shield allows the operator to
remove a source from the unit, while his body is being shielded). Standard Category 3
sources are those that are typically in a good condition, e.g. not corroded, source shutter
and positioning mechanism still in a good working condition. Thus to remove the source will
be quite easy and quick. The sources will be removed from their working shields, inspected,
source and unit information recorded, consolidated and seal welded in a stainless steel
capsule, capsule/s placed into a lead shield and lead shield placed into a retrievable concrete
shielded storage container. The storage container is also closed and prepared for long term
storage. It is assumed that 30 sources are conditioned in a year. Due to the radiation risk
during the handling of unshielded sources, this conditioning operation is done under the
supervision and oversight of the facility RPO.

Non-Standard Category 3 sources will be collected from their storage location in their
working shields and placed on a working area without shielding. These non-standard
category sources units and equipment are typically those that are difficulty to open or
remove the source. Due to the difficulty with the equipment (source still shielded inside the
equipment) the equipment cannot be handled behind a shielded work area. However as
soon as the source can be removed from the equipment, the source is directly moved behind
the shielded work area, where the source can be identified, characterised and transferred
into the capsule. The further steps followed are the same as done with standard Category 3
sources, described above.

Once the lead shield with the encapsulated DSRS has reached its filling capacity, the lead
shield in placed in a retrievable concrete shielded storage container. The concrete container
is closed with steel bars to prevent easy accessibility to the lead shield, where after the drum
is closed with a lid and clamp.
The sources inside the shielded container are retrievable.
The final waste package is thereafter transferred to an interim storage area where the waste
package it left to cure. Two such campaigns are conducted per annum.
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
The facility is visited by the facility operators for approximately 8 hours per month for
general cleaning, inspection and maintenance purposes.

Final storage waste packages are transferred to the waste storage facility and emplaced in
the storage room. Two such campaigns are conducted per annum.

The waste storage room is inspected and monitored on a monthly basis

The facility is visited by the facility RPO for approximately 4 hours per month for the routine
radiological inspection and surveillance purposes.
8.3.2
Operational Radiation Protection
The waste treatment and storage facilities are designated and operated as a radiological controlled
areas and people working in the facility are designated as occupationally exposed persons with the
necessary training, dosimetry and medical control.
A radiation protection programme will be implemented and will cover routine monitoring of the
facility and its environment, monitoring of specific operations such as treatment and emplacement
activities and any special monitoring that may be required from time to time. The programme will
make provision for the monitoring of external radiation levels and surface contamination and also
include the access and egress control measures applicable during the different operations.
8.3.3
Management System
The establishment and implementation of an integrated management system is important for the
proper management of DSRS. A management system for the processing, handling and storage of
Radioactive Waste compliant with international safety standards needs to be demonstrated by
CNSTN.
Written operational procedures will be drawn up to ensure the activities are carried out safely and in
the least time reasonably possible to optimize safety and protection and to ensure that no individual
dose constraints or limits are exceeded.
A formally documented and approved management system that integrates radiation protection QA,
operational, maintenance and inspection programmes to ensure protection and safety are optimized
and that no personal dose limits or constraints are exceeded needs to be implemented. The
management system will inter alia include an incident reporting system, emergency plans and
document and record management. The integrated management system must be continuously
updated and will reflect the recommendations from this safety case. Refer here to system manuals
and procedures.
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Records must be maintained of all operational activities, packages and equipment must be clearly
marked and labelled and an inventory must be maintained of all equipment and waste placed in the
store.
8.4
DSRS Inventory
CNSTN maintains a DSRS inventory of the sources used at their facilities by using RAIS 3.0. See
Appendix D.
Similarly CNRP maintains a national waste inventory on RAIS 3.0 but are in the process of moving
inventory to RAIS WEB. A copy of the current inventory is included as Appendix C.
CNRP are
however in the process of updating the National Inventory. This safety case needs to be updated
with the latest national inventory once it becomes available.
CNRP has indicated that there currently are ± 100 DSRS (inside their working shields) that has been
conditioned and placed in a cement matrix and is currently controlled by the CNRP and stored at the
respective user facilities.
Conditioned Ra-226 sources from former medical activities are currently stored at a local storage
facility which belongs to the ministry of health and controlled by CNRP.
Low Activity Check sources are kept at the premises of the CNRP laboratories for training purposes.
9.0
SAFETY ASSESSMENT
9.1
Safety Assessment Context
The purpose and philosophy for the safety assessment have been defined in section 7 of this report
for the scope of this safety case as defined in 7.1 specifically. Section 7.0 covers some information
related to the strategy for safety assessment which will be expanded in this section.
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9.1.1
Strategy for Safety Assessment
9.1.1.1 Basic Engineering Analyses
The list of the required engineering aspects and design features as listed in section 7.3.1 will be used
as a checklist to qualitatively assess and comment on the compliance of the waste management
facilities to the specific requirements.
9.1.1.2 Demonstration of the safety of the radioactive waste management activities performed by CNSTN.
•
For normal operation; quantitative deterministic assessment of worker dose due to the range of
activities by various occupational groups of CNSTN; the breakdown of normal operational
activities are the following:
-
Collection of DSRS at User facilities: Two Loaders from CNSTN inspects and load
consignments into a vehicle that is dedicated for transportation of sealed sources.
-
Collected DSRS are transported in a dedicated vehicle by 2 drivers from Users and interim
storage locations to the Central Waste Management facility at Sidi Thabet.
-
The transported DSRS are received at the Central Waste Management Facility at CNSTN.
The sources are surveyed, off loaded, inspected and segregated. Approximately 10
consignments with a total of 20 sources are received annually.
-
The DSRS are transferred to the storage room in the Central Waste Management Facility at
CNSTN.
-
Standard Category 3 sources in their working shields are removed from their storage
location and placed on a working area equipped with a lead shield. The sources are
removed from their working shields inspected, characterised, source and unit information
recorded, consolidated and seal welded in a stainless steel capsule, capsule/s placed into a
lead shield and lead shield placed into a retrievable concrete shielded storage container.
The storage container is also closed and prepared for final storage.
30 sources are
conditioned per year.
-
Non Standard Category 3 sources are collected from their storage location in their working
shields and placed on a working area without shielding. The sources are removed from
their working shields inspected, recorded and placed into a shielded storage container. One
source is dismantled per campaign and 3 campaigns are conducted annually.
-
Once the lead shield with the encapsulated DSRS has reached its filling capacity, the lead
shield is placed in a retrievable concrete shielded storage container. The container is closed
with steel bars and closed with a lid and clamp.
The waste package is thereafter
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transferred to an interim waiting area inside the main hall of the waste management facility
where the waste package is left to cure. Two such campaigns are foreseen to be conducted
per annum.
-
The waste operational areas of the central waste management facility is visited by the
facility operators for approximately 8 hours per month for general cleaning, inspection and
maintenance purposes.
-
Cured waste packages are transferred to the in the storage room. Two such campaigns are
conducted per annum.
-
The waste storage room is inspected and monitored by the facility RPO on a monthly basis
for a total of 4 hours.
•
For anticipated operational occurrences: quantitative deterministic assessment of worker and
public dose as applicable. Specific credible and enveloping scenarios will be developed and
doses to workers and public as applicable will be calculated with the use of simple models and
the use of conservative assumptions.
•
All other credible occurrences; Qualitative assessment of the impact of other occurrences and
the listing of specific preventative and mitigating measures. Other design basis and beyond
design basis events will be considered and enveloping scenarios will be developed. The
anticipated
consequences
associated
with
such
events
will
be
listed
with
comments/recommendation for further analyses and/or proposed preventative and mitigating
measures.
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9.1.1.3 The results from the quantitative and qualitative assessment as defined in 9.1.1.2 above will also
be compared to the proposed target and objectives set for the optimization of protection. No
specific optimization comments and recommendations will be made in the case of doses below 1
mSv/a.
9.1.1.4 A qualitative assessment of the non-radiological hazards of the facilities and the listing of specific
control measures will be performed. Non-radiological hazards will be listed and categorized in
terms of its hazard potential. Comments and recommendation will be made per hazard as
applicable.
9.1.1.5 A qualitative assessment of the implemented waste management practice; – The approach to
waste management with regards to the following will regarded as contributing to the inherent level
of safety:
Clearly defined responsibilities for waste management.
Implementation of the principles of waste minimization and avoidance, namely, re-use or reprocessing of waste, return to supplier, safe and secure storage and conditioning and final
disposal of waste.
Hazards and the generation of secondary waste, associated with all waste management
operations (routine and ad hoc) are known, monitored, projected and managed by due
management processes.
Interdependencies between the various steps of waste management are known and managed.
Waste acceptance criteria are defined, waste management activities and the outputs of such
activities are aligned with set waste acceptance criteria.
Interim storage of DSRS will only take place inside proper containment such as the original
working shields or another type of suitable containment.
Conditioned DSRS will be stored in a dedicated storage area with passive safety features and
adequate access control.
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9.1.1.6 A qualitative assessment of the availability, level of implementation of an integrated management
system to ensure a sustained level of safety during the operational phase of the facilities will be
conducted. This assessment will focus on RP, work procedures, QA aspects (mainly recordkeeping
and change management) and processes for the management of limits and conditions.
9.1.1.7 Uncertainties inherent to the assumptions made in the quantitative assessments or any other
uncertainties identified during the safety assessment will be evaluated to determine its impact on
safety. Uncertainties with a significant impact on safety will be listed with recommendation for its
management.
9.2
Safety Assessment Endpoints
The following quantitative assessment endpoints will be applicable:
-
Radiation dose to workers performing the various normal DSRS management activities at CNSTN
radiation doses to worker and the public as applicable due to anticipated operational occurrences.
It should be noted that it is expected and assumed that the same CNSTN personnel will be
performing all the respective DSRS management activities at CNSTN. Doses received during the
various activities are therefore accumulated for these workers. Doses will be evaluated against
the safety criteria as listed in section 6.4 and will also be compared with latest IAEA
recommended annual dose limits for occupationally exposed persons as described in [4].
-
9.3
9.3.1
The assessments will cover activities taking place over a 1 year period.
Development of Scenarios
Normal Operations
The normal operations scenarios for which worker doses are quantified are listed in 9.1.1.2 above. A
separate spread sheet is developed for each activity and all relevant assumptions are listed below
each spreadsheet. (See Section 10.0.)
9.3.2
Accident Scenarios
9.3.2.1 Anticipated Operational Occurrence Scenarios
The consequence of following postulated initiating events will be evaluated:
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-
Operational Occurrence 1
The CNSTN transport vehicle carrying three working shields with DSRS during the transport of
these units from the end user to the CNSTN waste treatment facility is involved in an accident.
The vehicle complies with the applicable requirements of the IAEA transport regulations, thus
having the applicable signs on the truck and Tremcard available inside the vehicle. The vehicle
capsizes, the driver/s cannot take emergency response action due to injury, and the working
shields with DSRS are flung from the vehicle and end up next to the road. The working shields
were all packaged inside one secondary container which could not withstand the impact which
led to the three units being separated from each other. The working shields are, however, still
intact with the DSRS inside and no loss of containment takes place. The tree units contained
two Co-60 sources, each with an activity of 25 mCi and one Cs-137 source with an activity of 50
mCi. First responders and other members of the public arrive at the scene of the accident and
spent one hour in close proximity (1 m) from the sources, before other CNSTN arrive on the
scene and ensure the public is kept at a safe distance from the vehicle. The sources are
recovered and surveyed by CNSTN RPOs and operators (30 min in close proximity) who then
continue with loading and transportation of the sources.
-
Operational Occurrence 2
The operator left a bare Cat 3 Co-60 source on the shielded workbench during the conditioning
campaign and removal of the source from its working shield in the waste treatment facility at
CNSTN. The operator did not wear his EPD and was under the impression that the source was
placed inside the shielded waste storage container and continued to work on another working
shield to remove the source. No alarm was made and the RPO invigilation was interrupted.
When the RPO returned after 45 minutes the elevated dose rate in the area was detected. The
RPO immediately evacuated the working area after which the misplaced source was detected
and placed inside the shielded container.
-
Operational Occurrence 3
During the conditioning campaign of non-standard Cat 3 sources, the operators dismantled an
unknown/non-standard source without the aid of the shielded work bench. After the primary
shield has been removed the dose rate in the area increased to above expected levels. Since
the source was unknown to the operators they did not know how to remove the source. The
operators panicked, did not evacuate the area and continued to try to remove the source and
spend 15 minutes in close proximity of the source before they managed to remove the source
and place it in the shielded storage container.
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9.3.2.2 Other Accident Scenarios
The following other accident scenarios will be considered:
-
Accident Scenario 1
The electrical wiring in the waste treatment facility creates a short circuit that results in a fire.
The fire spreads and causes the smoke detectors to activate an alarm. Some of the working
shields are being damaged by the fire before any fire-fighting personnel could arrive. A 50 mCi
Cs-137 source is ruptured in the process and starts leaking. Fire-fighting personnel arrive and
by using powder based fire-fighting equipment managed to quench the fire. With an assumed
release fraction of 10 % (5 mCi), contamination spread by the fire into the facility while 20 %
(10 mCi) of the released activity escaped from the building through the natural ventilation
system and through the opened doors to the environment. Fire-fighting personnel used
respirators and spend 20 min in the contaminated zones. After the fire was put out, the
remaining activity settled in the areas. Workers used protective suits and respirators to cleanup the contaminated zones.
-
Accident Scenario 2
During transport of two 10 mCi Cs-137 sources inside their working shields the CNSTN
transport vehicle is in involved in an accident and caught fire. The operators/drivers are not
in a position to remove the units from the vehicle. Due to the extreme heat from burning
vehicle fuel the sources are damaged to the extent that it starts leaking. The fire causes the
contamination to disperse to the immediate environment. Members of the public (residence
in the area close to the accident) are in close approximation of the burning vehicle and
exposed to the dispersed contamination. A release fraction of 10 % (2 x 1 mCi = 2 mCi) and
conservative metrological conditions are assumed.
-
Accident Scenario 3:
A DSRS waste container which is not filled to capacity and not conditioned are collected from
the receiving and treatment hall in the Waste Treatment and Storage Facility. The container
contains thirty 50 mCi Cs-137 sources. During lifting of the container with the overhead
crane, the container slipped and fell to the floor of the reception area of the Storage Room.
On impact the container lost its lid and all the sources. The sources were in close
approximation from each other. This happened close to the operator who spends 30 seconds
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within 1 m from the sources before the area is evacuated. After emergency intervention and
planning the operator spend 30 minutes at an average distance 1.5 m from the sources
collecting them with tongs and placing them into a shielded container. The RPO supervised
this operation at an average distance of 3 m from the sources. The container was closed and
transferred back to the Storage Room.
9.4
Data Used and Assumptions Made for the Safety Assessment
In order to perform the calculations for the safety assessment for the DSRS management activities in
Tunisia certain measured and calculated data will be used. In some instances, however, real-time
data is not available resulting in making certain assumptions. These assumptions are based on
experience performing similar types of activities elsewhere in the world. The assumptions made are
generally conservative. The current inventory of CNSTN (Appendix C) and the national inventory kept
by the CNRP (Appendix B) were reviewed. The assumed values and justification for selection are
reflected in Table 2 below. Some of these values are based on non-Tunisian experience since no
actual data is available. These assumed values should be changed as soon as local data is available.
Table 2: Assumptions and justifications for quantitative deterministic assessment
Condition
1
Dose Rate
Justification
Ambient dose rate 15 µSv/h
Typical ambient dose rate in a store full with DSRS
in Storage room
units (units packed on various shelves and racks, raks
about 1.5m apart)
Dose rate measured
between
racks and in walkways)
2
Ambient dose rate 10 µSv/h
The DSRS inventory in the receiving area will typical
in Receiving hall
be less than in the Storage room, thus assume a lower
rate than in the storage room.
3
Average
contact 70 µSv/h
Typical
average
contact
dose
rate
on
DSRS
dose rate on DSRS
units/equipment (as recorded on units received at
units
Necsa (South Africa)-
and
equipment
Can be changed based on
CNSTN experience – no dose rate results on units at
CNSTN available
4
Average dose rate 5 µSv/h
Typical
1m
DSRS
units/equipment (as recorded on units received at
and
Necsa (South Africa)- Can be changed based on
from
units
equipment
average
dose
rate
1m
from
DSRS
CNSTN experience – no measurement result on units
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Condition
Dose Rate
Justification
at CNSTN available
5
6
Unshielded Cat 3 12000 µSv/h
No sources are handled by hand, thus always by
source dose rate at
tongs, assumed to be 0.5m long.
0.5 m
conservative average dose rate (70% of maximum,
Unshielded Cat 3 1200 µSv/h
see below) of the Tunisian Cat 3 inventory.
source dose rate at
The inventories were reviewed and Co-60 and Cs-137
1.5 m
were found as the nuclides with high energy gammas
and which are in the majority.
This value is
Maximum current
activity of Cat 3 Co-60 and Cs-137 DSRS in the CNSTN
inventory is 14.8 mCi and 0.223 mCi respectively.
Maximum current activity of Cat 3 Co-60 and Cs-137
DSRS in the national inventory is 0.368 Ci and 40 mCi
respectively.
The average activity of both inventories is significantly
lower.
0.5m dose rate for max Co-60 0.368Ci is 16700 µSv/h.
1.5m dose rate for max Co-60 0.368Ci is 1800 µSv/h.
7
Whole body dose 50 µSv/h
Removing sources behind 10 cm lead brick wall on the
due to shielded Cat
working bench.
3 source at 0.5m
bricks with hands not shielded.
behind 10cm lead
assumptions for 5 and 6 above
Person stands directly behind lead
Refer also to
working shield
8
9
Average dose rate 50 µSv/h
Typical average dose rates on a final package (e.g.
1m
final
210L metal drum lined with concrete and cavity in
package containing
centre for a lead shield containing several capsules.
conditioned DSRS
These capsules each contain several DSRS).
from
Average
contact 1000 µSv/h
dose rate on final
package containing
conditioned DSRS
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10.0 SAFETY ASSESSMENT
10.1 Basic Engineering Analyses
Table 3: Basic Engineering Analyses
Item
Requirement (Note 1).
Compliance
Ref
1. General: Facility Design, Construction and Maintenance
1.1
Basic site characteristics and
To be confirmed
credible external events have
been considered in the design
1.2
Quality assurance has been
No design information has
considered in the design,
been supplied.
construction, maintenance and
modification the waste
management facilities:
Design approval and
 The facilities have been
designed and constructed in
certificates to be supplied.
accordance with acceptable
national construction codes
and standards.
Plans to be developed or
- Inspection and maintenance
plans exist and are
supplied.
implemented
To be supplied
- Formal processes are defined
and implemented for the
evaluation, approval and
implementation of
modifications (Change
management)
2. Safety and security aspects were considered in the design of the facility
2.1
The characteristics of the walls
To be modelled or to be
ensuring a level of dose rate
included as a facility limits
that complies with the restriction
and included in the
for public exposure (1 mSv/a)
procedure for
even for the maximum
management of the
anticipated inventory.
facility limits and
conditions.
2.2
The lighting system will be
To be demonstrated by
adequate and permits the
facility lighting
performance of operations in a
measurement (lumens) to
safe manner.
be confirmed on an
annual basis.
2.3
Physical delineation of areas
. Main storage area is
designed for storage and for the
isolated from the waste
main waste management
operations areas.
operations are isolated, this way
- separate
it is ensured the appropriated
decontamination area
segregation of materials
within main storage
optimizing worker’s exposure
facility will be demarcated
Comments
Storage room
has walls 270
mm thick to
provide
shielding
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Item
2.4
2.5
2.6
2.7
Requirement (Note 1).
during operations
Each delineated area has a
sufficient physical space that
ensures a minimal probability of
accident occurrence during
waste management operations
and package handling.
Storage areas were designed
under the principle of labyrinth,
which contributes to optimize
the exposure of workers.
(Stored DSRS and waste
operations are not in taking
place in the same area)
Waste packages with sources
are stored in a manner such that
packages are not in contact the
floor or interior surface of the
building walls. This allows for
inspection and control
operations and the potential
corrosion of
packaging/containers is limited.
Unconditioned radioactive
sources are stored in storage
systems ensuring normal
operation and minimizing
probability of accidents. Their
main characteristics are:
Compliance
and controlled as
controlled area
Confirmed. Receiving,
treatment.
Decontamination and
conditioning area 160 m2
and storage room 46 m2.
Ref
Comments
No labyrinth system. Only
a wall placed in front of
entrance to storage room
to reduce radiation shine
from storage area.
To be demonstrated once
facility is commissioned.
Also to be included in
work procedures
To be assessed.
 Storage capacity is greater
than current and foreseen
needs of management.
 It ensures source
segregation. In this way,
periodic inspection and
radiological monitoring of the
storage building and of the
waste drums/packages is
facilitated.
Total inventory limits to
be developed.
 Its structure resists the
maximum load of the sources
that are intended to be
stored.
Maximum load capacities
to be demonstrated.
Segregation of sources is
provided for as part of the
receipt procedure. Clear
procedures need to be
developed for the
assessment and handling
of unknown sources.

3. Engineering systems ensuring safety for situations of occurrences and accidents
3.1
Floor and wall finish allow easy
To be ensured during
decontamination.
completion of facility and
confirmed.
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Item
3.2
3.3
Requirement (Note 1).
The segregation of the different
areas limits the potential
dispersion of any contamination.
Compliance
No dynamic or static
containment systems
needed during normal
DSRS operations.
Evaluation to determine
the need, system and
procedure to handle and
process contaminated
DSRS. Extraction hood to
be fitted where sources
will be removed from their
working shields (at the
conditioning area)
Collection sump for
potentially contaminated
liquids at the
decontamination bay. The
sump drains to the
collection tanks.
Prepare procedure for
sampling and release of
effluent
To be confirmed.
Ref
Comments
In case of a potential surface
decontamination using liquids
there is a collection system
inside the facility that prevents
its release to the environment.
The system has a retention tank
that permits environmental
monitoring before releasing to
the environment.
3.4
The facility has its own fire
detection and fire-fighting
equipment.
4. Facility design provides physical security features commensurate with the security threat
4.1
Robust building construction
Facility inspection showed
with high integrity doors and
robust building
locks to the treatment and
construction with high
storage areas.
integrity doors and locking
systems. To be confirmed
during final installation
4.2
Buildings are equipped with
To be confirmed
intrusion alarms.
4.3
The buildings have vehicle
The building has one
access points. A separate
vehicle access point.
personnel door is provided to
Personnel entrance is via
segregate personnel from
the change rooms
vehicle movements.
between the
administrative section and
the radiologically
controlled areas.
4.4
No windows are provided so as
Main Hall is equipped with
to improve its shielding and
windows at a 6.5 m
security performances.
height. The windows
cannot open and will
provide only sunlight.
Security enhancements
are, however, required
specifically on the
windows close to second
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Item
Requirement (Note 1).
Compliance
floor patio.
Ref
Comments
Note 1: Requirements based on IAEA safety standard: WS-G-6.1: Storage of Radioactive waste
10.2 Quantitative Deterministic Assessment of Worker Dose
Refer to section 8.2.1 for description and assumed number of actions per year.
10.2.1
Activity 1: Collection of DSRS at Interim Stores
Table 4: Collection of DSRS at Interim Stores
Operator Groups
Operator Actions
Exposure Type
Exposure Data
Dose rate
[µSv/h]
1
Loaders(3)
Inspection and ID
Loading
10.2.2
Whole Body
5
Extremity
70
Whole Body
5
Extremity
70
Justification/Notes
Dose rate at 1m
from DSRS, see
section 9.4 (4)
Contact dose rate
on DSRS, see section
9.4 (3)
Dose rate at 1m
from DSRS, see
section 9.4 (4)
Contact dose rate
on DSRS, see section
9.4 (3)
Exposure time
Time per
action [h]
Actions per year
Annual dose
[µSv/a]
0.25
20
25
0.25
20
350
1.75
20
175
1.75
20
2450
Activity 2: Collected DSRS are transported in a dedicated vehicle by 2 drivers from Users and interim
storage locations to CNSTN.
Table 5: Transport of DSRS
Operator Groups
Operator Actions
Exposure Type
Exposure Data
Dose rate
[µSv/h]
1
Drivers (2)
Driving
Whole Body
5
Justification/Notes
Dose rate in vehicle
cab. Based on dose
rate at 1m from
DSRS, see section
9.4 (4) and
assuming 2 units on
truck and having
increased distance
and vehicle body
between drivers
and sources
Exposure time
Time per
action [h]
Actions per year
3
10
Annual dose
[µSv/a]
150
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10.2.3
Activity 3: Receipt of DSRS at CNSTN
Table 6: Receipt of DSRS at CNSTN
Operator Groups
Operator Actions
Exposure Type
Exposure Data
Dose rate
[µSv/h]
1
Operators (2)
Off loading
Inspection and
segregation
Whole Body
5
Extremity
70
Whole Body
5
Extremity
70
Whole Body
5
Extremity
70
2
RPO
Surveying
10.2.4
Justification/Notes
Dose rate at 1m
from DSRS, see
section 9.4 (4)
Contact dose rate
on DSRS, see section
9.4 (3)
Dose rate at 1m
from DSRS, see
section 9.4 (4)
Contact dose rate
on DSRS, see section
9.4 (3)
Dose rate at 1m
from DSRS, see
section 9.4 (4)
Contact dose rate
on DSRS, see section
9.4 (3)
Exposure time
Time per
action [h]
Actions per year
Annual dose
[µSv/a]
1
20
100
0.1
20
140
0.2
20
20
0.1
20
140
0.2
20
20
0.1
20
140
Activity 4: Temporary Storage of Category 3 Sources and general activities in the CSF.
Table 7: Temporary Storage of Cat 3 Sources and general activities in the CSF
Operator Groups
Operator
Actions
Exposure Type
Exposure Data
Dose rate
[µSv/h]
1
Operators(2)
2
RPO (1)
10.2.5
Placement of
Cat 3 in
temporary
storage
General
Inspection,
maintenance
and cleaning
of whole CSF
Radiological
Inspection &
surveying of
CSF
Whole Body
5
Extremity
70
Whole Body
10
Extremity
NA
Whole Body
10
Extremity
NA
Justification/Notes
Dose rate at 1m from
DSRS, see section 9.4
(4)
Contact dose rate on
DSRS, see section 9.4
(3)
Ambient dose rate,
see section 9.4 (2)
Ambient dose rate
see section 9.4 (2)
Exposure time
Time per
action [h]
Actions per year
Annual dose
[µSv/a]
0.1
20
10
0.1
20
140
8
12
960
4
12
480
Activity 5: Conditioning Campaign 1: Standard Cat 3 Sources
Table 8: Conditioning Campaign 1
Operator Groups
Operator
Actions
Exposure Type
Exposure Data
Dose rate
[µSv/h]
1
Operators(2)
Handling
Whole Body
5
Justification/Notes
Dose rate at 1m
from DSRS, see
section 9.4 (4)
Exposure time
Time per action
[h]
Actions per year
0.017
30
Annual dose
[µSv/a]
2.55
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SOURCES IN TUNISIA
Operator Groups
2
RPO (1)
10.2.6
Operator
Actions
Dismantling
Source
Transfer, refer
to section
8.2.1
Loading
source and
shield in
container and
closing of
Storage
Container
refer to
section 8.2.1
Supervision &
surveying
during
conditioning
Exposure Type
Exposure Data
Dose rate
[µSv/h]
Extremity
70
Whole Body
50
Extremity
70
Whole Body
50
Extremity
Whole Body
Extremity
Justification/Notes
Contact dose rate
on DSRS, see section
9.4 (3)
DR behind shield,
see section 9.4 (7)
Unshielded
doserate see section
9.4 (3) (2)
DR behind shield
see section 9.4 (7)
Unshielded DR see
section 9.4 5 (5)
Doserate at 1 m on
a typical final waste
package see section
9.4 (8)
12000
50
Contact doserate on
a typical final waste
package see section
9.4 (8)
1000
Whole Body
15
Extremity
NA
Ambient doserate
see section 9.4 (1)
Exposure time
Time per action
[h]
Actions per year
Annual dose
[µSv/a]
0.017
30
35.7
0.1
30
150
0.1
30
210
0.01
30
15
0.005
30
1800
0.3
2
33
0.3
2
660
0.5
30
225
Activity 6: Conditioning Campaign 2: Non-Standard & Linear Cat 3 Sources
Table 9: Conditioning Campaign 2
Operator Groups
Operator Actions
Exposure Type
Exposure Data
Dose rate [µSv/h]
Extremity
12000
Whole Body
1200
Extremity
12000
Justification/Notes
Dose rate at 1m
from DSRS, see
section 9.4 (4)
Contact dose rate
on DSRS, see
section 9.4 (3)
DR behind shield,
see section 9.4 (7)
Unshielded DR see
section 9.4 (5)
Unshielded DR see
section 9.4 (6)
Unshielded DR see
section 9.4 (5)
Whole Body
10
Ambient doserate
see section 9.4 (2)
Extremity
NA
1
Handling
Operators(2)
Dismantling
Source Transfer
Sources are
combined with
those in
Campaign 1,
refer to
Whole Body
5
Extremity
70
Whole Body
50
2
RPO (1)
Supervision &
surveying
Annual dose
[µSv/a]
Exposure time
Time per action [h]
Actions per year
0.017
3
0.3
0.017
3
3.6
0.17
3
25.5
0.01
3
360
0.01
3
36
0.01
3
300
0.5
3
15
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SOURCES IN TUNISIA
10.2.7
Activity 7: Transfer of Conditioned Waste Packages to the Waste Store including its Surveillance,
Inspection and Maintenance
Table 10: Transfer of Conditioned Waste Packages to the Waste Store
Operator Groups
Operator
Actions
Exposure Type
1
Whole Body
Operators(2)
Handling and
placement into
storage room
10.2.8
Extremity
Exposure Data
Dose rate
Justification/Note
[µSv/h]
s
Dose rate at 1m
from final
package, see
50
section 9.4 (8)
Contact dose rate
on final package,
1000
see section 9.4 (9)
Exposure time
Time per action
[h]
Actions per year
Annual dose
[µSv/a]
0.3
2
30
0.05
2
100
Worker Dose Summary
The maximum worker dose is summarised in the table below. The maximum dose has been
obtained reflecting the assumptions that the same individuals conduct the transporter/loader and
operator functions and the same RPO conducts the RPO function.
Table 11: Worker Dose Summary
Operator
Groups
Loaders/
Transporters
Operator
Actions
Exposure
Type
Worker Dose Per Activity [uSv/a]
3
4
5
6
7
Whole Body
100
10
3
1
30
Extremity
140
140
36
4
100
Whole Body
150
26
Extremity
210
360
Source
Transfer
Whole Body
15
36
Extremity
1800
360
Inspection
and
Maintenance
Whole Body
20
Extremity
140
Storage
container
preparation
Whole Body
33
Extremity
660
Supervision
and Surveying
Whole Body
20
Extremity
140
Inspection &
Loading
Transport
1
Whole Body
200
Extremity
2800
Whole Body
2
150
Extremity
Handling & off
loading
Dismantling
Operators
RPO
960
225
15
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SOURCES IN TUNISIA
Operator
Groups
Operator
Actions
Facility
Surveillance
Exposure
Type
Worker Dose Per Activity [uSv/a]
1
2
Whole Body
3
4
5
6
7
480
Extremity
The maximum total dose to the Operator/Loader/Transporter is therefore:
-
Whole body:
Extremity:
1.734 mSv/a
6.750 mSv/a
The maximum total dose to the Operator (excluding loading/transport) is therefore:
-
Whole body:
Extremity:
1.384 mSv/a
3.950 mSv/a
The maximum total dose to the RPO is therefore:
-
Whole body:
Extremity:
0.74 mSv/a
0.14 mSv/a.
10.3 Deterministic Assessment of Worker and Public Dose for Anticipated Operational
Occurrences:
The scenarios as defined in Section 9.3.2.1 above is assessed by simply calculation.
Occurrence scenario 1
The maximum public and additional worker dose is calculated by multiplying the maximum
anticipated dose rate of 5 µSv/h from a shielded cat 3 source as used in Table 2 with the exposure
times of 1 hour and 30 min for the public and workers respectively:
The maximum public dose would therefore be 7.5 µSv or even 22.5 µSv if simultaneously
irradiated by 3 sources. The maximum additional dose to the worker would therefore be in the order
of 22.5 µSv if the same argument is used.
Occurrence scenario 2
The Maximum additional dose to the worker due to the occurrence is calculated by increasing the
exposure time of the operator’s source transfer activity as tabled in Table 8 above to 45 minutes.
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SOURCES IN TUNISIA
The maximum additional dose to the worker would therefore be; whole body 38 µSv and, assuming
that the hands are the whole time behind the work shield, extremity 9000 µSv.
Occurrence scenario 3
The Maximum additional dose to the worker due to the occurrence is calculated by increasing the
exposure time of the operator’s source transfer activity as tabled Table 9 above from 0.6 minutes to
15 minutes.
The maximum additional dose to the worker would therefore be; whole body 264 µSv and
extremity 2640 µSv.
10.4 The following other accident scenarios will be considered in the Tunisian Safety
Assessment:
Accident Scenario 1
The maximum public and additional worker doses projected for the scenario as defined in Section
9.3.2.2 above are derived based on the assumptions, calculations and modelling indicated in the
table below:
Table 12: Accident scenario 1
Receptors
Actions
Firefighting
Personnel
Firefighting
Exposure type
Whole body
Internal dose
Public
Living close
to area of
accident
Operators
Clean-up
Internal radiation
and exposure
from cloud &
ground shine
Whole body
Internal dose
Exposure data
Units
Justification/
Notes
500 µSv/h
Ambient dose
rate [1]
1E6 Bq/m3
Activity
concentration [2]
Exposure parameters
Time per
Other/Units [x]
action [h]
0.3
150
0.3
Respirator. Eff. 0%
Breathing rate: 1.2 m3/h
AMAD: 1um
DCF: 4.8E-9 Sv/Bq
Dispersion and dose modelled with Hotspot assuming a ground level
release, see Appendix A for Hot spot inputs and results [3]
1728
50 µSv/h
800
2E7 Bq/m2
Ambient dose
rate [4]
Surface
contamination
[5]
16
16
Respirator. Eff. 0%
Breathing rate 1.2 m3/h
AMAD 1um
DCF: 4.8E-9 Sv/Bq
Dose
[µSv]
0.967
1843
Notes:
[1]: Elevated ambient dose rate due to 50mCi Cs-137 damaged and exposed source and activity release into the air.
Assume conservatively ambient dose rate of 500 µSv/h based on dose rate from exposed source and activity released
into the air (dose rate 0.5m from 50mCi Cs-137 source is 560 µSv/h)
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SOURCES IN TUNISIA
[2]: Projected airborne activity concentration levels calculated on the assumption that 10% of 50mCi Cs-137 source is
dispersed homogeneously inside the facility with a total volume of 200m3
[3]: Hotspot dispersion model assuming 10mCi (20% of 50mCi source) Cs-137 is released into the atmosphere, long
term exposure (4 days), conservative metrological conditions and using highest exposure in dispersion plume
(maximum TED).
[4]: Elevated ambient dose rate due to 5mCi Cs-137 spread as contamination over the whole area to be cleaned.
Assume conservatively ambient dose rate of 50 µSv/h based on dose rate from activity spread in area (dose rate 0.5m
from 5mCi Cs-137 source is 56 µSv/h)
[5]: Maximum projected surface contamination levels calculated on the assumption that the 5mCi (10%) which was
released into the facility, has settled homogeneously on a 10m2 area.
Accident scenario 2
The maximum public dose for this scenario as defined in Section 9.3.2.2 above (second scenario)
are derived based on the assumptions, calculations and modelling indicated in the table below:
Table 13: Accident scenario 2
Receptors
Actions
Exposure type
Public
Living close
to area of
accident
Internal radiation
and exposure
from cloud &
ground shine
Exposure data
Exposure parameters
Units
Justification/
Time per
Other/Units [x]
Notes
action [h]
Dispersion and dose modelled with Hotspot assuming a ground level release
see Appendix B for Hot spot inputs and results [1]
Dose
[µSv]
0.193
Notes:
[1]: Hotspot dispersion model assuming 2mCi (10% of two 10mCi sources) Cs-137 is released into the atmosphere,
long term exposure (4 days), conservative metrological conditions and using highest exposure in dispersion plume
(maximum TED).
Accident scenario 3
The maximum Operator and RPO doses projected for the occurrence and scenario as defined in
section 9.3.2.2 above are derived based on the assumptions, calculations and modelling indicated in
the table below:
Table 14: Accident scenario 2
Receptors
Operator
Operator
RPO
Actions
Source
handling
Sources
recovery
Supervision
Exposure type
Exposure data
Units
Whole body
5000 µSv/h
Whole body
2222 µSv/h
Whole body
556 µSv/h
Justification/
Notes
Calculated dose
rate at 1 m [1]
Calculated dose
rate at 1.5 m [2]
Calculated dose
rate at 3 m [3]
Exposure parameters
Time per
Other/Units [x]
action [h]
0.0167
Dose
[µSv]
0.5
1111
0.5
278
83
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DRAFT SAFETY CASE FOR THE MANAGEMENT OF DISUSED SEALED RADIOACTIVE
SOURCES IN TUNISIA
Notes:
[1]: Dose rate at 1 m calculated for an unshielded 50mCi Cs-137 source using a specific gamma ray constant of 0.33
R.m2/Ci.h. The dose rate was adjusted for 30 sources by assuming a linear relationship between dose rate and
activity (Simplified point source geometry)
[2]: Dose rate at 1.5 m calculated for 30 unshielded 50mCi Cs-137 sources using the dose rate at 1 m, and inverse
square law.
[3]: Dose rate at 3 m calculated for 30 unshielded 50mCi Cs-137 sources using the dose rate at 1 m, and inverse
square law.
The maximum total doses to the Operator and RPO is therefore: Whole body: 1.2 mSv and 0.28
mSv respectively.
Note that if Co-60 sources with the same activity are assumed that the
approximate doses to the Operator and RPO would be 4.8 mSv and 1.1 mSv respectively.
10.5 Comparative Dose Assessment: SAFRAN
A dose assessment for the DSRS activities at the CNSTN CSF was also performed using the
SAFRAN version 2.1.4.0 dose assessment tool. The purposes of this assessment was to enable a
comparison between the simple spreadsheet assessment as performed in Section 10.2 above and
the SAFRAN tool.
The SAFRAN dose assessment was performed only for the Receiving, Interim storage,
Conditioning and Longer term storage activities. The results of the SAFRAN assessment are
provided in table attached as Appendix D.
The results of the SAFRAN assessment for the respective activities assessed are the same as
calculated in the tables above in Section 10.2.
10.6 Optimization of Protection: Assessment
The summary of the outcome of the quantitative assessment of the radiological consequence of
normal operations, anticipated operational and other occurrences as well as comments and
recommendations regarding the optimization of protection, are covered in the table below.
Table 15: Optimization of Protection: Assessment
Occupatio
nal Group/
Receptor
Dose /Dose Rate
[µSv/µSv/a]
Whole
ExtreBody/
mities
ED
Comments
Recommendations
1. Normal Operation: Quantitative Deterministic Assessment of Worker Dose
Operator/
1734
This assumes that the
 Ensure that the applicable
6750 µSv/a
Loader/
µSv/a
operator is experienced
personnel is aware of the
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SOURCES IN TUNISIA
Occupatio
nal Group/
Transporter
Receptor
Dose /Dose Rate
[µSv/µSv/a]
Comments
and trained in performing
the tasks. If the level of
conservatism associated
with the dose assessment
is considered, the annual
exposure to workers is low
which limits the margin for
further optimization of
protection. Most of the
exposure is due to the
source transfer, which is a
needed and justified
action.
Recommendations
associated risks and is
experienced in the tasks. If
not such personnel is
available, the conditioning of
sources (which has a higher
risk of exposure) should be
performed with expert
support or supervision.
 Implementation of a formal
operational optimization
programme where actual
doses are measured and
specific reduction strategies
are considered and
implemented
 Define source transfer as a
safety critical action and
consider design and
procedures to reduce
exposure potential
 None
RPO invigilation is justified
i.t.o. dose limitation and
740
RPO
140 µSv/a
control. RPO dose is below
µSv/a
current optimization trigger
level.
2. Anticipated Operational Occurrences: Quantitative Deterministic Assessment of Worker and Public
Dose : Occurrence 1
Operator/L
Exposure levels are below
 Actions to ensure compliance
oader/
22.5 µSv
optimization trigger levels
to the transport regulations.
Transporter
and sufficient control is
inherent to the compliance
Public
22.5 µSv
to the transport regulations
3. Anticipated Operational Occurrences: Quantitative Deterministic Assessment of Worker and Public
Dose : Occurrence 2
 Evaluate the possibility to
install a radiation alarm
Operator/
Expose levels are low but
system with an alarm set
Loader/
38 µSv
9000 µSv
possible to prevent by
point of about 150 µSv/h
Transporter
simple design changes.
(response, testing and
maintenance procedures)
4. Anticipated Operational Occurrences: Quantitative Deterministic Assessment of Worker and Public
Dose : Occurrence 3
 Evaluate the possibility to
install a radiation alarm
system with an alarm set
point of about 150 µSv/h
Operator/
Possible to prevent
(response, testing and
Loader/
264 µSv
2640 µSv
exposure by simple design
maintenance procedures)
Transporter
and operational changes.
 Formalised procedure to
ensure the prior evaluation
of unknown/ nonstandard
sources and planning of its
dismantling
5. Other Occurrences: Quantitative Deterministic Assessment of Worker and Public Dose : Accident
Scenario 1
Fire
Dose mainly due to
 Initiate a fire and fire
fighting
2028 µSv
external radiation. Dose
protection system evaluation
Personnel
due to contamination and
of the areas.
dispersion of beta gamma
 Assess the possibility to store
Public
0.967 µSv
emitters is low. Possible to
unconditioned sources in
which is
prevent exposure by simple
vaults or other fire proof
insignificant
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DRAFT SAFETY CASE FOR THE MANAGEMENT OF DISUSED SEALED RADIOACTIVE
SOURCES IN TUNISIA
Occupatio
nal Group/
Receptor
Dose /Dose Rate
[µSv/µSv/a]
Comments
Recommendations
design and operational
changes to prevent fires
and to mitigate the
consequences of fires.
system.
 Review procedures to ensure
Operator/
housekeeping and storage
Loader/
800 µSv
practices that are aligned with
Transporter
fire prevention and control
measures.
6. Other Occurrences: Quantitative Deterministic Assessment of Worker and Public Dose : Accident
Scenario 2
Exposure levels are below
• Actions to ensure compliance
0.193 µSv
optimization trigger levels
to the transport regulations
Public
which is
and sufficient control is
insignificant
inherent to the compliance
to the transport regulations
7. Other Occurrences: Quantitative Deterministic Assessment of Worker and RPO Dose : Accident
Scenario 3
Dose mainly due to
• Review design of source
external radiation. Possible
container
Operator
1200 µSv
to prevent exposure by
• Review the source container
prevent falling and or loss
handling procedure
of sources in the event of a
• Ensure that ALARA review is
fall event. Dose could also
prescribed by intervention
RPO
280 µSv
be reduced by using more
procedure in order to
operators during
minimize individual dose.
intervention
10.7 Non-radiological Hazard Assessment
The following non-radiological hazards are relevant to the operation of the Waste management
facilities at CNSTN:

Conventional Hazards: Manual handling of heavy objects, overhead loads, using of driven and
manual tools, working on elevated heights. These hazards are managed by a general
awareness of the hazards, training and appointment and the compulsory use of personal
protective equipment while performing specific activities.

Hazardous chemical substances: May include flammable and toxic chemical stored and used in
the waste treatment facility or the presence of other hazardous/irritant substances such as
cement, dust, lead, asbestos, etc. Hazardous chemical substances are controlled by
maintaining inventories of such materials, proper storage practices, work procedures that
prescribe the requirements for the safe handling of such substance e.g. personal protective
equipment requirements.
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10.8 Assessment of the Implemented Waste Management Practice
The outcome of the quantitative assessment of the current and future waste management
practice to be implemented by CNSTN is tabled below:
Table 16: Quantitative Assessment: Current and Future Waste Management Practices
Item
Requirement
Compliance Comments
Ref
1.
Clearly defined responsibilities for
waste management.
2.
Implementation of the principles
of waste minimization and
avoidance, namely, re-use or reprocessing of waste, return to
supplier, safe and secure storage
and conditioning and final
disposal of waste.
3.
Hazards and the generation of
secondary waste, associated with
all waste management operations
(routine and ad hoc) are known,
monitored, projected and
managed by due management
processes.
4.
Interdependencies between the
various steps of waste
management are known and
managed. Waste acceptance
The existing Legal Framework of
Tunisia is outdated. Tunisia is
however in the process of drafting a
new National Nuclear Law that will
specify the responsibilities for the
generation and management of
radioactive waste. The proposed CSF
currently under construction at
CNSTN and planned operation of the
waste management facility
demonstrates intent and
commitment.
Principles will be defined in the new
Legal Framework and implemented in
the case of DSRS to the point of
conditioning. No final disposal option
is available. A Radioactive Waste
Management Agency will be
constituted under the new Nuclear
Law
CNRP is currently responsible for the
National Source Inventory and
CNSTN is responsible for their own
inventory. RP requirements in terms
of this requirement are being
implemented. There is no treatment
of standard DSRS as the proposed
CSF is still under construction. CNRP
does however condition DSRS at the
end-users where DSRS is stored
inside their working shields in a
cement matrix. Conditioning
exercises are planned to mitigate
exposure. Facilities to treat nonstandard sources or deviating e.g.
contaminated sources do not yet
exist. No procedures to assess and
plan the handling of non-standard
exist.
Future waste management practices
will be performed as described in
Section 9.1.1.2. Compliance to be
demonstrated once new facility
becomes operational.
No written conditioning specification
or a WAC for the proposed storage
facility exists as yet. It was also not
indicated how the current CNRP
Section
6.1.3
Section
6
Section
8.2
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Item
5.
6.
Requirement
Compliance Comments
criteria are defined, waste
management activities and the
outputs of such activities are
aligned with set waste acceptance
criteria.
Interim storage of DSRS will only
take place inside proper
containment such as the original
working shields or another type of
suitable containment.
conditioning actions and specification
are aligned with future disposal
options (allows for recovery of DSRS
from waste packages).
Conditioned DSRS will be stored
in a dedicated storage area with
passive safety features and
adequate access control.
DSRS in their original working shields
(as per the national inventory) is
stored at the end-users as single
units or in a cement matrix waste
container. DSRS stored in future in
centralised waste management
facility should be stored inside
working shields or other suitable
containment. Need to be
demonstrated.
. To be confirmed.
Ref
Section
8.3
Section
8.3 and
6.4
10.9 Management System Assessment
The outcome of the quantitative assessment of only the main requirements of an integrated
management system as implemented by CNSTN is tabled below:
Table 17: Quantitative Assessment: Integrated Management System
Item
Requirement
Compliance Comments
1.
2.
3.
A written and approved integrated
management system is maintained
to ensure a sustained level of
safety during the operational phase
of the facilities.
The Quality Assurance part of the
integrated management system
inter alia covers:
 Quality policy and objectives
 Organisation and responsibilities
 Documentation, waste tracking
and record keeping
 Product realisation and work
procedures
 Worker training and
appointment
 Change control of procedure
and facilities
 Non-conformance and event
management
 Auditing and system review
An RP programme exist and inter
alia covers:
 RP organisation, training and
No written and approved
management system has been
developed as yet.
Not yet existing
CNSTN has an existing RP
Programme for the proposed CSF.
Needs to be updated.
Ref
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Item
4.
10.10
Requirement
appointment
 Zone classification, criteria and
access control
 Workplace monitoring and
surveillance
 Personnel monitoring and
medical surveillance
 Environmental monitoring
 RP instrumentation control
 Clearance/exemption
surveillance and control
The integrated management
system inter alia covers:
 An approved WAC for receipt of
DSRS at the waste treatment
facility
 An approved WAC for receipt of
DSRS waste packages at the
waste storage facility
 Procedure in which all
operational limitations and
conditions associated with the
facilities, their performance
criteria and how and at what
interval their performance will
be assessed and recorded, are
listed
Compliance Comments
Ref
Proposed CSF still under
construction; therefore no existing
WAC .
Assessment of Uncertainties
The outcome of a provisional quantitative assessment of uncertainties related to the safety case is
presented in the table below:
Table 18: Provisional Quantitative Assessment: Safety Case Uncertainties
Comments/
Ref
Item
Uncertainty
Recommendations
1.
2.
Uncertainty in the source term used
in the safety assessment. The
source term is defined for cat 3
sources and specifically for beta/
gamma emitters such as Cs-137
and Co-60. The impact of normal
operations and occurrences could
be significantly higher if higher
activity sources or alpha emitting
sources have been considered. The
critical pathway in the case of alpha
emitting radionuclides for
contamination scenarios is internal
radiation.
Uncertainty regarding the dose rate
information used in the safety
assessment. Although it was aimed
The operational limits and
conditions of the operational waste
treatment facility should limit the
range of source that could be
received under the current
authorization. The facility WAC
should state the limits and
conditions as mentioned above and
include a process and authorization
requirements for the receipt of any
unknown sources of sources
outside the facility WAC.
Confirmatory monitoring should be
performed and used to verify the
dose rate assumptions or be used
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to use conservative data, the
exposure data used for the various
exposure scenarios is not based on
scientific arguments, measurement
or modelling results.
No certainty as to whether the
CNSTN DSRS Management Facility
will in fact become the national
centralised facility although CNRP
and CNSTN both support that
principle.
Although the main methodologies
for management of DSRS do not
differ much and is based on
standard practice, the specific
activities (procedural steps) within
the methodologies may be different
than what was assumed for the
safety assessment.
3.
4.
10.11
Comments/
Recommendations
Uncertainty
Item
Ref
as bases to update exposure
scenarios and data.
Confirm this proposal. Safety Case
will have to be updated if not the
case.
The specific procedures should be
developed and tested to ensure its
suitability for CNSTN. The safety
assessment should be updated to
reflect this. Procedures should also
align with best international
practice.
Assessment of possible Public exposures
Exposure to the public is only possible due to the following:
-
Elevated radiation levels outside the facility security fence. Radiation levels will be routinely
monitored by the facility RPO to confirm that radiation levels remains less than 10 x natural
background levels (or any limit prescribed by the regulator), refer to section 8.2.2.2.
Thus
public will not receive elevated exposures.
-
Severe accidents at the facility or during the transport of DSRS to the facility. These have been
evaluated in section 10.3. The resulting possible public exposure during any of these events is
insignificant.
10.12
Assessment of possible environmental pathways
Liquid
If contaminated liquid is released into the environment it could contaminate the area and have
possible later impact on the public being exposed to this soil.
The release of effluent from the facility is for the following reasons highly unlikely:
-
No contaminated effluent is generated during the routine operation of the facility. The facility is
developed for the storages of sealed radioactive sources. The probability of a source to start
leaking during storage and conditioning is very small.
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-
Before any waste package is accepted in the storage facility, it will be checked by the facility
RPO for any sign of contamination.
Should it be contaminated, the waste package will be
decontaminated and sealed.
-
If a contaminated item or leaking source is received (very unlikely) and have to be
decontaminated, these will be handled in the decontamination bay. This bay is located inside
the receiving and treatment hall. This area is installed with a drainage system, which will divert
any water from this area into dedicated effluent tanks. In addition standard practice during
decontamination and cleaning is to use as little as possible water (minimisation of secondary
waste)
-
Routine radiation and contamination surveys are done in the facility by the facility RPO. Should
presence of any contamination be detected, the applicable area or DSRS will be decontaminated
Gaseous
Gaseous release into the environment and exposing the public is only possible during an accident
scenario. These have been evaluated in section 10.3. The resulting possible public exposure during
any of these events is insignificant.
10.13 Waste Management
Should any solid waste be generated during the conditioning of DSRS (e.g. protective clothing,
gloves, overshoes, cleaning papers) this shall be kept separate and accumulated in a waste
drum, which will be sealed when not in use.
Any waste generated or accumulated during the clean-up of a possible event where a source was
leaking or contamination was spread, the waste shall be accumulated in waste drums.
Record shall be kept of the waste drums and their applicable content.
The drum shall be regarded as containing radiation and kept in the CSF.
11.0
HUMAN RESOURCES
The management team of the CNSTN is responsible to ensure that all the personnel involved in
any of the transport, operation, inspection and control activities at the Centralized Storage
Facility are properly trained and experienced in the applicable activities and associated risks.
Evidence of the applicable training shall be kept as a QA record.
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All the activities shall be done by the applicable personnel in accordance with documented and
approved work procedures. All the personnel shall be registered, trained and controlled as
radiation workers.
The following trained personnel would be required:
-
Radiation Protection Officer
-
Facility operators
12.0 IDENTIFICATION OF FACILITY SPECIFIC LIMITS AND CONDITIONS
Based on the safety assessment, the following facility limitation and conditions are derived:
 No neutron sources will be conditioned. Current Tunisian inventory does not include many neutron
sources. These source units shall only be safely stored.
 The radiation outside the facility depends on various factors, which include: external wall thickness
and density, external radiation on packages, nuclides, package design, total facility inventory and
storage configuration. Due to the number of factors, it is not possible to determine the maximum
nuclide and activity inventory of the facility which will ensure that the radiation outside the facility
will be acceptable for possible public exposure. For this reason routine RPO surveys outside the
facility are required.
These then needs to confirm that radiation levels remains less than 10 x
natural background levels (or any limit prescribed by the regulator).
Should the surveys show
elevated levels, applicable corrective actions should be implemented inside the CSF.
 The assessment assumed 30 standard and 3 non-standard DSRS to be conditioned per year. These
activities contributes the most to personnel exposure. The current foreseen exposure is relatively
low. After actual operation, these could be reconsidered and the number of cappains increased and
specified as limiting condition..
 Specify Sources i.t.o. radionuclides and activity limits that may be received and processed as
standard and non-standard campaigns. A process that includes evaluation and authorization of
receipt, handling and treatment of sources other than the specified sources.
 The storage location and maximum inventory of DSRS in such locations in the waste treatment
facility should be specified and controlled.
 The maximum localized and ambient dose rates inside the waste treatment and facility should be
specified and should not be in excess 250 and 25 µSv/h respectively (general limits applied for
controlled areas).
 The maximum inventory for the storage facility needs to be derived and specified
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 The maximum localized and ambient dose rates inside the waste storage facility, in operator zones
should be specified and should not be in excess 250 and 25 µSv/h respectively.
 The maximum dose rate outside any of the waste management facilities should not be in excess of
2.5 µSv/h.
 Annual reporting of facility operations and RP surveillance data to the regulatory body.
13.0 INTEGRATION OF SAFETY ARGUMENTS
The provisional synthesis of safety arguments below should be considered within the scope of the
safety case i.e. constructed and operational stage facilities;
13.1 Facility Design and Engineering
Although a range of facility design, engineering and construction related aspects have been identified
as relevant to safety, it still needs to be obtained/demonstrated, the proposed CSF currently under
construction seems robust with features that indicate that safety and security have been considered.
Unresolved issues related to facility design and engineering including management systems to ensure a
sustained level of safety (e.g. maintenance and change management) are covered in Section 15.0
below.
13.2 Facility Operation
The safety assessment indicates that the facility can be operated well within safety criteria as far as
DSRS activities are considered. The safety assessment may also be used as basis to increase the extent
and range of operations related to high activity DSRS taking cognisance of an acceptable margin that
needs to be maintained. The assessment of occurrences also indicates consequences well within safety
and risk criteria. (The equivalent risks of the occurrences could be demonstrated as low and below 10-5
per year even at frequencies of 10-1 to 10-2 per year). Uncertainties exist mainly regarding source term
assumptions and some scientific data. Unresolved issues (Section 15.0) included continued action the
verify assumptions and scientific data. Some facility specific limits and conditions have also been
recommended in order to mitigate some uncertainties.
13.3 Optimization of protection
The margin for optimization of protection associated with the DSRS activities is limited in view of the
relative low consequences and conservatism of assumptions made. Some facility design and procedural
changes could however reconsider for further optimization of protection. An operational optimization of
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protection program, that is based on activity specific RP surveillance, personnel dosimetry results and
scheduled optimization review sessions, is recommended.
13.4 Waste Management Practise
Good waste management practice is generally evident from the intent of the legal framework,
organisational arrangements and defined responsibilities, to establish waste management facilities and
the waste management facility operations. The interdependencies amongst the various waste
management steps seem to be considered to the point of waste treatment. The alignment between
conditioning, conditioning specification, storage and disposal is not clear nor has any written and
approved WAC been made available. Recommendations regarding unresolved issues are covered in
Section 15.0 below.
13.5 Integrated Management System
Although some management systems and procedures have been implemented no evidence of such
written and approved system were supplied. Management of unresolved issues as covered in Section
15.0 below, addresses recommendations regarding the development of an integrated management
system.
13.6 Uncertainties
The provisionally identified uncertainties is neither of such a nature nor extent that the associated
detriment in confidence in the safety case would result in the recommendation of drastic measures.
Uncertainties are manageable by setting specific facility limits and conditions, preparing WAC and by
implementation of some confirmatory monitoring plans. The management of unresolved issues as
covered in Section 15.0 below, covers management of uncertainties.
14.0 COMPARISON WITH SAFETY CRITERIA AND CONCLUSIONS
The Quantitative safety assessment results as reflected in Section 10.5 above, is well within the safety
criteria as listed in Section 6.3.1 for workers and Section 6.3.2 for the public. The safety case for the
DSRS operations in the waste management facilities at CNSTN is supported subject to a formal plan
and schedule to address the identified unresolved issues as covered in Section 15.0 below.
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15.0 ASPECTS REQUIRING CLARIFICATION AND RECOMMENDATION
The
identified
aspects
requiring
further
clarification
with
commensurate
management
recommendations are tabled below:
Table 19: Aspects Requiring Clarification/Recommendations
Item
Aspects Requiring Clarification
Recommendation
1. Tunisian Legal and regulatory Framework
During the preparation of this case, a draft National
To update section 6.1.3 of the report to
1.1
Nuclear Law for Tunisia was in the process of being
reviewed for signature and to be promulgated by
government.
2. Basic Engineering Analyses
A number of unresolved issued and gaps have been
2.1
identified in the basic engineering analyses as listed
in section 10.1. that needs to be resolved or
managed.
3. Optimization of Protection
Optimization Normal Operation related exposure
3.1
3.2
Optimization of occurrence related exposure.
include the provisions of the new
legislation
Develop a strategy and plan to obtain
relevant information and documentation.
If it is not possible to obtain certain
information, further justification should be
considered. The plan should make
provision for the revision of the safety
case.
 Development and implementation of a
formal operational optimization (ALARA)
programme where actual doses are
measured and specific reduction
strategies are considered and
implemented.
 Define source transfer as a safety
critical action and review design and
procedures to reduce exposure
potential.
•
•
•



Actions/audit to ensure compliance to
the transport regulations.
Evaluate the possibility to install a
radiation alarm system with an alarm
set point of about 150 µSv/h
(response, testing and maintenance
procedures).
Develop and implement a procedure to
ensure the prior evaluation of
unknown/ nonstandard sources and
planning of its storage and treatment.
Initiate a fire and fire protection
system evaluation of the areas if not
installed before commissioning.
Assess the possibility to store
unconditioned sources in vaults or
other fire proof systems.
Develop procedures (inspection and
testing) to ensure housekeeping and
storage practices relating to fire
prevention and control are established
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Item
Aspects Requiring Clarification
Recommendation
and maintained.
4. Non-Radiological Hazards
Comprehensive assessment of non-radiological
4.1

hazards.
5. Implemented Waste Management Practice
WAC.
5.1
5.2
Interdependencies related to disposal

6. Integrated Management System
No written and approved management system
6.1
documents have been provided.
•
7. Management of Uncertainties
Uncertainties related to source term.
7.1
7.2
Uncertainties regarding dose rate assumption.
7.3
Uncertainties regarding the specific procedural steps
in the DSRS management process.
8. Facility Specific Limits and Conditions
Procedure for defining and management of facility
8.1
specific limits and conditions
16.0
Plan, schedule and conduct a
comprehensive non radiological hazard
assessment.
(Covered in integrated management
system 6. below)
National waste management plan to make
provision for disposal- could be a longer
term action but commitments related to
disposal are necessary.
Plan and schedule an integrated
management system review that is
focussed the main requirements as
listed in the table in 10.8 above.
•
(Covered by Facility limits and
condition in 8. below and be actions to
develop WAC in as covered in 6.
Above)
 Develop and implement a confirmatory
monitoring plan to verify the dose rate
assumptions once facility becomes
operational. This could be used as
bases to update exposure scenarios
and data.
 Develop procedures for each
management step and test suitability
for the CNSTN facility..
 Development of a procedure that lists
the agreed limits and conditions as
applicable to the various facilities and
activities as recommended in section
11. Above. The procedure should
include the specified limits and
conditions, how and when and by
whom compliance/ performance will be
verified as well as the related
recording and reporting requirements.
APPENDIX A: HOTSPOT OUTPUT FOR ACCIDENT SCENARIO 1
HotSpot Version 3.0.2 General Fire Nov 21, 2014 07:53 AM
Source Material
Material-at-Risk (MAR)
Damage Ratio
(DR)
Airborne Fraction (ARF)
Respirable Fraction (RF)
Leakpath Factor (LPF)
Respirable Source Term
:
:
:
:
:
:
:
Cs-137 F
30.0y
1.8500E+09 Bq
1.00
0.200
1.000
1.000
3.70E+08 Bq
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Non-respirable Source Term
Release Radius
Cloud Top
Physical Height of Fire
Effective Release Height
Wind Speed (h=10 m)
Avg Wind Speed (h=H-eff)
Stability Class
Respirable Dep. Vel.
Non-respirable Dep. Vel.
Receptor Height
Inversion Layer Height
Sample Time
Breathing Rate
Distance Coordinates
:
:
:
:
:
:
:
:
:
:
:
:
:
:
:
0.00E+00 Bq
1m
10 m
5m
8.52 m
2.00 m/s
1.95 m/s
D
0.30 cm/s
8.00 cm/s
1.5 m
None
10.000 min
3.33E-04 m3/sec
All distances are on the Plume Centerline
Maximum Dose Distance
Maximum TED
Inner Contour Dose
Middle Contour Dose
Outer Contour Dose
Exceeds Inner Dose Out To
Exceeds Middle Dose Out To
Exceeds Outer Dose Out To
:
:
:
:
:
:
:
:
0.091 km
9.67E-07 Sv
1.00E-10 Sv
1.00E-11 Sv
1.00E-12 Sv
71 km
> 200 km
> 200 km
FGR-13 Dose Conversion Data - Total Effective Dose (TED)
Include Plume Passage Inhalation and Submersion
Include Ground Shine (Weathering Correction Factor : None)
Include Resuspension (Resuspension Factor : NCRP Report No. 129)
Exposure Window:(Start: 0.00 days; Duration: 4.00 days) [100% stay time].
Initial Deposition and Dose Rate shown
Ground Roughness Correction Factor: 1.000
Distance
TED
km
0.100
0.200
0.300
0.400
0.500
1.000
2.000
(Sv)
9.6E-07
5.2E-07
2.9E-07
1.9E-07
1.3E-07
4.1E-08
1.3E-08
RESPIRABLE
TIME-INTEGRATED
AIR CONCENTRATION
(Bq-sec)/m3
4.5E+05
2.4E+05
1.4E+05
8.6E+04
6.0E+04
1.9E+04
6.3E+03
GROUND SURFACE
DEPOSITION
GROUND SHINE
DOSE RATE
(kBq/m2
1.3E+00
7.3E-01
4.1E-01
2.6E-01
1.8E-01
5.7E-02
1.9E-02
(Sv/hr)
2.6E-09
1.5E-09
8.1E-10
5.1E-10
3.6E-10
1.1E-10
3.7E-11
ARRIVAL
TIME
(hour:min)
<00:01
00:01
00:02
00:03
00:04
00:08
00:10
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17.0
APPENDIX B: HOTSPOT OUTPUT FOR ACCIDENT SCENARIO 2
HotSpot Version 3.0.2 General Fire Nov 21, 2014 07:19 AM
Source Material
Material-at-Risk (MAR)
Damage Ratio
(DR)
Airborne Fraction (ARF)
Respirable Fraction (RF)
Leakpath Factor (LPF)
Respirable Source Term
Non-respirable Source Term
Release Radius
Cloud Top
Physical Height of Fire
Effective Release Height
Wind Speed (h=10 m)
Avg Wind Speed (h=H-eff)
Stability Class
Respirable Dep. Vel.
Non-respirable Dep. Vel.
Receptor Height
Inversion Layer Height
Sample Time
Breathing Rate
Distance Coordinates
: Cs-137 F
30.0y
: 7.4000E+08 Bq
: 1.00
: 0.100
: 1.000
: 1.000
: 7.40E+07 Bq
: 0.00E+00 Bq
:1m
: 10 m
:5m
: 8.52 m
: 2.00 m/s
: 1.95 m/s
:D
: 0.30 cm/s
: 8.00 cm/s
: 1.5 m
: None
: 10.000 min
: 3.33E-04 m3/sec
: All distances are on the Plume Centerline
Maximum Dose Distance
Maximum TED
Inner Contour Dose
Middle Contour Dose
Outer Contour Dose
Exceeds Inner Dose Out To
Exceeds Middle Dose Out To
Exceeds Outer Dose Out To
:
:
:
:
:
:
:
:
0.091 km
1.93E-07 Sv
1.00E-10 Sv
1.00E-11 Sv
1.00E-12 Sv
20 km
122 km
> 200 km
FGR-13 Dose Conversion Data - Total Effective Dose (TED)
Include Plume Passage Inhalation and Submersion
Include Ground Shine (Weathering Correction Factor : None)
Include Resuspension (Resuspension Factor : NCRP Report No. 129)
Exposure Window:(Start: 0.00 days; Duration: 4.00 days) [100% stay time].
Initial Deposition and Dose Rate shown
Ground Roughness Correction Factor: 1.000
Distance
TED
km
0.100
0.200
0.300
0.400
0.500
1.000
2.000
(Sv)
1.9E-07
1.0E-07
5.9E-08
3.7E-08
2.6E-08
8.2E-09
2.7E-09
RESPIRABLE
TIME-INTEGRATED
AIR CONCENTRATION
(Bq-sec)/m3
9.0E+04
4.9E+04
2.7E+04
1.7E+04
1.2E+04
3.8E+03
1.3E+03
GROUND SURFACE
DEPOSITION
GROUND SHINE
DOSE RATE
(kBq/m2
2.6E-01
1.5E-01
8.2E-02
5.2E-02
3.6E-02
1.1E-02
3.8E-03
(Sv/hr)
5.2E-10
2.9E-10
1.6E-10
1.0E-10
7.1E-11
2.3E-11
7.4E-12
ARRIVAL
TIME
(hour:min)
<00:01
00:01
00:02
00:03
00:04
00:08
00:17
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18.0
APPENDIX C: NATIONAL INVENTORY (CNRP)
Nature du
radioelement
Cs137
Cs137
Cs137
CS137
Cs137
Cs137
Co 60
Co 60
Co 60
Co 60
Co 60
Co 60
Co 60
Co 60
yb169
yb169
Ra/Be
Ra/Be
Ra/Be
Cs137
Irt 92(2 sources)
Ir192
in 92
Ir192
Co60
Co60
Co60
Co60
Co60
Co60
Co60
Co60
Co60
AmBe
AmBe
Cs+AmBe
Cs+AmBe
Co60
Cs137
Co60
Co60
Pm147
Pm147
CS137
Cs137
Cs137
Activite
N° de la serie / type de la source
119
221
234
505
I 7906
9333
36
269
AS537
AS538
NH012
Date de mesure
de I'activite
9,9uCi
7,51 Ci
15uCi
3uCi
2,6Ci
3mCi
4,55mCi
1,13mCi
20,15 mCi
9,85mCi
30Ci
7,33Ci
54,6mCi
2,1 mCi
3Ci
3Ci
5mCi
30mCi
1mCi
8,73 Ci
1971
23/02/1981
13/09/1963
12/09/1963
01/03/1981
01/03/1981
03/02/1963
26/02/1963
15/02/1963
23/01/1963
20/04/1978
01/02/1992
15/02/1963
28/02/1963
17/08/1976
17/09/1976
30/07/1962
01/01/1973
01/01/1973
04/10/1973
?
provenance de la Russie
AG30-01
AG30-02
AG30-03
AG30-04
AG30-05
AG30-06
AG30-07
AG30-08
AG100-01
Solo20
Solo25
CPN
Solo25
Irradiateur
QG20
QG100
92Ci
=0Ci
=0Ci
20mCi
20mCi
20mCi
20mCi
20mCi
20mCi
20mCi
20mCi
60mCi
10mCi
40mCi
10mCi+50
mCi
10mCi+50
mCi
4500Ci
?
?
250nCi
9,25MBq
27mCi
12,3Ci
10mCi
16/05/1979
01/01/1982
Jan-82
1976
1976
1976
1976
1976
1976
1976
1976
1976
07/02/1981
12/12/1968
?
?
1984
1989
1982
1981
1993
Periode
30ans
30ans
30ans
30ans
30ans
30ans
5ans
5ans
5ans
5ans
5ans
5ans
5ans
5ans
30ans
This document is the property of NECSA
and shall not be used, reproduced,
transmitted or disclosed without prior
written permission.
Doc. No.:
NLM-REP-14/197 Rev 2
Page No.:
Page 68 of 76
DRAFT SAFETY CASE FOR THE MANAGEMENT OF DISUSED SEALED RADIOACTIVE
SOURCES IN TUNISIA
Cs137
CS137
Cs137
Cs137Am241Be
Am241
Am241
C06O
Cs137
Cs137Am241Be
P32
Ra226
Ra226
Ra226
C06O
Cs137
Cs137
Cs137
Cs137
?
Am241
Am241
Cs137
Cs137
Cs137
Cs137
Cs137
Cs137
Cs137
Cs137
Cs137
C06O
Cs137
?
Cs137
Ca45
Cs137
Pm147
Na22
Cs134
Nitrate de
thorium
Gd153
Thalium170
Zn65
Cs137
Eu152
Ag110
Cerium 139
Sr90
TROXLER
10mCi
10mCi
10mCi
7,3mCi-40mCi
199
3
199
199
3
1981-19803
TROXLER
100mCi
100mCi
76,3mCi
2,95Ci
8,8mCi-40mCi
198
198
5
5
03/12/197
03/12/197
9
16/9/819
16/10/81
05/10/199
? 8
?
?
?
Nov85
Jan86
Nov85
NovOct85
80
Paratonnerre
Paratonnerre
Paratonnerre
?
bascule nucleaire-2016GG
bascule nucleaire-2012GG
bascule nucleaire-1916GG
bascule nucleaire-2006GG
jauge de niveau-AG30-3170
Paratonnerre-HELITA
AMH5 335/63
Paratonnerre-HELITA
AMH5 335/64
bascule nucleaire-2041GG
bascule nucleaire-2032GG
densimetre -AF37
bascule nucleaire- B360
bascule nucleaire 2040GG
jauge de
niveau -1805
jauge de niveau -1800
densimetre -DG5
densimetre - DG5
jauge de niveau - 48
densimetre - DG5
densimetre
densimetre
?
?
370MBq/ml
1,64mCi
1,64mCi
1,64mCi
0,750mCi
0,750mCi
109mCi
200mC
i
789mCi
202mC
i
20mCi
289mC
?i
0,0374
GBq
0,037G
Bq
0,037G
Bq
1mCi
?
1mCi
1mCi
Jan86
Jan86
Sep80
Jan86
Jan74
Jan08/11/197
74
7
08/11/197
7
08/11/199
0
31/10/199
0
01/10/198
8
?
Jun88
Jun88
source etalon
2sources
etalon
0,1 mCi
05/06/196
9
This document is the property of NECSA
and shall not be used, reproduced,
transmitted or disclosed without prior written
permission.
Doc. No.:
NLM-REP-14/197 Rev
1
Page No.:
Page 69 of 76
DRAFT SAFETY CASE FOR THE MANAGEMENT OF DISUSED SEALED RADIOACTIVE
SOURCES IN TUNISIA
19.0
APPENDIX D: CNSTN INVENTORY
A0
At
RN
Q
A0
(KBq)
(Ci)
At
(KBq)
(Ci)
Origin
Nature
Ray
T (an)
Da
N.R
Co-60
1
3,7 E+12
0,99E+05
6,90E+11
1,86E+04
CBI
Solid
γ
5,27
mars-99
1001
Co-60
1
6,66E+11
1,80E+04
1,98E+11
5,34E+03
Hongrie
Solid
γ
5,27
oct.-01
1002
Co-60-cs137
1
1,48E+01
4,00E-07
1,01E+01
2,74E-07
ST
Solid
γ
30,16
juin-95
1003
Cf-252
1
8,70E+02
2,35E-05
1,36E+01
3,68E-07
Canberra
Solid
η
2,64
Mar-96
1004
C-14/H-3
1
1,70E+02
4,60E-06
7,54E-21
2,04E-28
CMI
Solid
β
1,266
Sep-97
1005
Am-241
1
5,03E+02
1,36E-05
6,65E-02
1,80E-09
CMI
Solid
α
432,2
Sep-97
1006
Co-60
1
5,06E+02
1,37E-05
4,92E+02
1,33E-05
CMI
Solid
γ
5,27
Sep-97
1007
Cs-137
1
4,68E+02
1,27E-05
7,57E+01
2,05E-06
CMI
Solid
γ
30,2
Sep-97
1008
Na-22
1
5,22E+02
1,41E-05
3,38E+02
9,13E-06
CMI
Solid
γ
2,6
Sep-97
1009
Pu-236
1
4,21E-02
1,14E-09
1,17E+01
3,15E-07
AIEA
Solid
α
2,85
Sep-97
1010
U-232
1
3,20E-02
8,65E-10
1,31E-03
3,55E-11
AIEA
Solid
α
Sep-97
1011
Y-80
1
6,17E+02
1,67E-05
0,00E+00
0,00E+00
CMI
Solid
γ
15,41
Sep-97
1012
H-3
20
< 7,40E+00
2,00E-07
2,77E-02
7,49E-10
AIEA
Solid
β
12,3
avr.-98
1013
C-14
20
< 3,70E+00
1,00E-07
3,25E+02
8,79E-06
AIEA
Solid
β
5715
avr.-98
1014
Source mixte
Calibration
Plate
68,9
Form No.: NLM-DIV-FORM-00-002 Rev: 03
This document is the property of NECSA
and shall not be used, reproduced,
transmitted or disclosed without prior
written permission.
Doc. No.:
NLM-REP-14/197 Rev 2
Page No.:
Page 70 of 76
DRAFT SAFETY CASE FOR THE MANAGEMENT OF DISUSED SEALED RADIOACTIVE
SOURCES IN TUNISIA
A0
(Ci)
At
(KBq)
At
(Ci)
Origin
Nature
Ray
T (an)
Da
N.R
RN
Q
A0
(KBq)
Co-57
1
7,16E+02
1,94E-05
3,41E+00
9,22E-08
CMI
Solid
γ
0,745
juil-98
1015
Cd-109
1
3,46E+02
9,34E-06
3,69E+00
9,98E-08
CMI
Solid
γ
1,26
juil-98
1016
Mn-54
1
3,76E+02
1,02E-05
6,56E-03
1,77E-10
CMI
Solid
γ
0,854
juil-98
1017
Th-228
1
2,65E+03
7,17E-05
2,87E+01
7,74E-07
Canada
Solid
α
1,914
juin-99
1018
Ra-226
1
4,71E+03
1,27E-04
4,68E+03
1,27E-04
Canada
Solid
α
1600
juin-99
1019
Th-229
1
5,30E-02
1,43E-09
5,29E-02
1,43E-09
AIEA
Solid
α
7340
mai-00
1020
Am-241
3
7,40E+02
2,00E-05
7,28E+02
1,97E-05
AIEA
Solid
α
432,2
02-Jan
1021
Co-60
1
1,85E+06
5,00E-02
5,48E+05
1,48E-02
AIEA
Solid
γ
5,27
02-Oct
1022
Co-60
1
1,11E+06
3,00E-02
3,29E+05
8,89E-03
AIEA
Solid
γ
5,27
02-Oct
1023
Co-60
1
5,55E+05
1,50E-02
1,64E+05
4,44E-03
AIEA
Solid
γ
5,27
02-Oct
1024
Tl-204
1
3,70E+00
1,00E-07
3,35E+00
9,05E-08
ST
Solid
β
63
déc-02
1025
Sr-90
2
3,70E+00
1,00E-07
2,97E+00
8,04E-08
ST
Solid
β
28,6
03-Jan
1026
Co-57
1
3,70E+01
1,00E-06
3,62E+01
9,77E-07
ST
Solid
γ
271,79
03-Jan
1027
Na-22
1
3,70E+01
1,00E-06
3,36E+00
9,08E-08
ST
Solid
γ
2,6
03-Jan
1028
Co-60
2
3,70E+01
1,00E-06
1,11E+01
3,01E-07
ST
Solid
γ
5,27
03-Jan
1029
Cd-109
1
3,70E+01
1,00E-06
2,69E-01
7,27E-09
ST
Solid
γ
1,267
03-Jan
1030
Mn-54
1
3,70E+01
1,00E-06
2,49E-02
6,72E-10
ST
Solid
γ
0,854
03-Jan
1031
Br-133
Cs-137
1
3,70E+01
1,00E-06
2,08E+01
5,61E-07
ST
Solid
γ
10,8
03-Jan
1032
(Unknown)
1
3,70E+01
1,00E-06
3,01E+01
8,13E-07
ST
Solid
γ
30,2
03-Jan
1033
This document is the property of NECSA
and shall not be used, reproduced,
transmitted or disclosed without prior
written permission.
Doc. No.:
NLM-REP-14/197 Rev 2
Page No.:
Page 71 of 76
DRAFT SAFETY CASE FOR THE MANAGEMENT OF DISUSED SEALED RADIOACTIVE
SOURCES IN TUNISIA
A0
(Ci)
At
(KBq)
At
(Ci)
Origin
Nature
Ray
T (an)
Da
N.R
RN
Q
A0
(KBq)
Ra-226
1
2,81E+02
7,59E-06
2,79E+02
7,55E-06
CMI
Solid
γ
1600
août-98
1034
Am-241 /Be
1
3,70E+07
1,00E+00
3,68E+07
9,96E-01
AIEA/ NESCA
Solid
η
432,2
déc-07
1035
Am-241
1
3,54E+01
9,55E-07
3,49E+01
9,43E-07
IPL
Solid
α
432,2
sept.-03
2001
Co-60
1
3,52E+00
9,51E-08
1,18E+00
3,18E-08
IPL
Solid
γ
5,27
sept.-03
2002
Cs-137
1
3,79E+01
1,02E-06
3,32E+01
8,96E-07
IPL
Solid
γ
30,2
sept.-03
2003
Zn-65
1
3,60E+01
9,73E-07
9,19E-02
2,48E-09
AREVA
Solid
X-γ
0,668
mars-06
2004
Fe-55
1
7,20E+01
1,95E-06
1,63E+01
4,40E-07
AREVA
Solid
X-γ
2,68
mars-06
2005
Cd-109
1
2,40E+01
6,49E-07
5,79E-02
1,56E-09
AREVA
Solid
X-γ
1,265
mars-06
2006
Am-241
1
3,70E+06
1,00E-01
3,62E+06
9,78E-02
AREVA
Solid
X-γ
432,2
janv.-01
2007
Fer-55
1
3,70E+05
1,00E-02
1,37E+04
3,70E-04
AREVA
Solid
X-γ
2,68
Avr 98
2008
Cd-109
1
3,70E+05
1,00E-02
6,42E+01
1,74E-06
AREVA
Solid
X-γ
1,265
Avr 99
2009
Am-241
1
29,6
8,00E-07
2,89E+01
7,80E-07
CEA
Solid
α
432,2
mars-96
2010
Co-60
1
4,41E+01
1,19E-06
5,52E+00
1,49E-07
CEA
Solid
γ
5,27
mars-96
2011
Cs-137
1
4,10E+01
1,11E-06
2,85E+01
7,71E-07
CEA
Solid
γ
30,2
mars-96
2012
Na-22
1
3,00E+01
8,11E-07
4,45E-01
1,20E-08
CEA
Solid
γ
2,602
mars-96
2013
Am-241
1
34,4
9,30E-07
3,35E+01
9,06E-07
CEA
Solid
γ
432,2
mars-96
2014
Co-60
1
43,2
1,17E-06
5,40E+00
1,46E-07
CEA
Solid
γ
5,27
mars-96
2015
Cs-137
1
45,6
1,23E-06
3,17E+01
8,58E-07
CEA
Solid
γ
30,2
mars-96
2016
Na-22
1
30,4
8,22E-07
4,51E-01
1,22E-08
CEA
Solid
γ
2,602
mars-96
2017
Ba-133
1
3,92E+01
1,06E-06
2,28E+01
6,17E-07
CEA
Solid
γ
10,7
sept.-03
2018
Cd-109
1
3,67E+02
9,92E-06
3,84E+00
1,04E-07
CEA
Solid
γ
1,267
sept.-03
2019
This document is the property of NECSA
and shall not be used, reproduced,
transmitted or disclosed without prior
written permission.
Doc. No.:
NLM-REP-14/197 Rev 2
Page No.:
Page 72 of 76
DRAFT SAFETY CASE FOR THE MANAGEMENT OF DISUSED SEALED RADIOACTIVE
SOURCES IN TUNISIA
A0
(Ci)
At
(KBq)
At
(Ci)
Origin
Nature
Ray
T (an)
Da
N.R
RN
Q
A0
(KBq)
Eu-152
1
3,67E+00
9,92E-08
1,77E+00
4,77E-08
CEA
Solid
γ
13,51
sept.-03
2020
Pm-147
1
3,60E-01
9,74E-09
8,31E-03
2,25E-10
CEA
Solid
β
2.62
oct.-97
2021
Cl-36
1
3,84E-01
1,04E-08
3,84E-01
1,04E-08
CEA
Solid
β
3 E+5
oct.-97
2022
Sr-90
3
3,66E-01
9,89E-09
2,59E-01
6,99E-09
CEA
Solid
β
28,5
oct.-97
2023
Tc-99
1
3,23E-01
8,72E-09
3,23E-01
8,72E-09
CEA
Solid
β
2,1 E+5
oct.-97
2024
C-14
1
3,94E+00
1,07E-07
3,94E+00
1,06E-07
CEA
Solid
β
5730
oct.-97
2025
1
2,78E-03
7,51E-11
2,67E-03
7,23E-11
NIST
Solid
α
72
sept.-97
2026
γ
432,2
janv.-08
2027
source mixte
(Am-241,U238,
U-234, Pu239)
source mixte
(Am241,Mn54)
MB
1
3,70E+00
1,00E-07
3,68E+00
9,94E-08
ANALYTICS
(0,5 L)
This document is the property of NECSA
and shall not be used, reproduced,
transmitted or disclosed without prior
written permission.
Doc. No.:
NLM-REP-14/197 Rev 2
Page No.:
Page 73 of 76
DRAFT SAFETY CASE FOR THE MANAGEMENT OF DISUSED SEALED RADIOACTIVE
SOURCES IN TUNISIA
RN
Q
A0
(KBq)
A0
(Ci)
At
(KBq)
At
(Ci)
Origin
source mixte
(Am241,Mn54)
Nature
Ray
T (an)
Da
N.R
γ
432,2
janv.-08
2028
mai-95
2029
MB
1
3,70E+00
1,00E-07
3,60E+00
9,74E-08
ANALYTICS
(1 L)
Tc-99
1
1,91E-04
5,16E-12
0,00E+00
0,00E+00
ANALYTICS
Solid
β
2,111
E+5
Cs-137
1
3,26E+01
8,80E-07
2,24E+01
6,06E-07
ANALYTICS
Solid
β
30,2
sept.-95
2030
Th-230
1
1,64E-04
4,43E-12
1,64E-04
4,43E-12
ANALYTICS
Solid
β
7,703
E+4
sept.-95
2031
Na-22
1
2,63E+00
7,11E-08
7,27E-02
1,96E-09
CMI
Solid
γ
2,6
juil.-98
2032
Ra-226
1
3,16E+00
8,55E-08
6,08E-14
1,64E-21
CMI
Solid
γ
1600
juil.-98
2033
Cs-137
1
2,94E+00
7,93E-08
4,75E-05
1,28E-12
CMI
Solid
γ
30,2
juil.-98
2034
Am-241
1
8,91E+00
2,41E-07
8,72E+00
2,36E-07
CMI
Solid
γ
432,2
juil.-98
2035
Co-60
1
3,51E+00
9,48E-08
5,97E-01
1,61E-08
CMI
Solid
γ
5,27
juil.-98
2036
Ra-226
1
2,96E+00
7,99E-08
2,94E+00
7,94E-08
CMI
Solid
γ
1600
sept.-97
2037
Cd-109
1
1,76E+01
4,75E-07
1,13E-02
3,06E-10
CMI
Solid
γ
1,267
juil.-98
2038
Sr-85
1
3,86E+00
1,04E-07
3,52E+00
9,50E-08
CMI
Solid
γ
64,78
juil.-98
2039
This document is the property of NECSA
and shall not be used, reproduced,
transmitted or disclosed without prior
written permission.
Doc. No.:
NLM-REP-14/197 Rev 2
Page No.:
Page 74 of 76
DRAFT SAFETY CASE FOR THE MANAGEMENT OF DISUSED SEALED RADIOACTIVE
SOURCES IN TUNISIA
A0
(Ci)
At
(KBq)
At
(Ci)
Origin
Nature
Ray
T (an)
Da
N.R
RN
Q
A0
(KBq)
source mixte
gamma +
Am-241
1
9,15E+01
2,47E-06
9,03E+01
2,44E-06
CMI
Solid
γ
432,2
avr.-03
2040
Ba-133
1
4,88E+00
1,32E-07
2,78E+00
7,51E-08
CMI
Solid
γ
10,7
avr.-03
2041
Eu-152
1
5,11E+00
1,38E-07
3,27E+00
8,83E-08
CMI
Solid
γ
13,51
avr.-03
2042
Ra-226
1
5,18E+00
1,40E-07
5,17E+00
1,40E-07
CMI
Solid
γ
1600
avr.-03
2043
Am-241
1
5,28E-01
1,43E-08
5,24E-01
1,42E-08
E&Z
Solid
γ
432,2
Mars.-07
2044
Sr-90/ Y-90
1
4,99E-01
1,35E-08
4,49E-01
1,21E-08
E&Z
Solid
γ
28,5
févr.-07
2045
Sr-85
1
1,608
4,35E-08
1,51E+00
4,08E-08
AREVA
Solid
X
64,78
Mars.-06
2046
Co-57
1
40,2
1,09E-06
1,65E-05
4,47E-13
CEA
Solid
γ
0,745
Mars.-96
2047
Cr-51
1
60,6
1,64E-06
2,25E-62
6,09E-70
CEA
Solid
γ
0,075
Mars.-96
2048
Mn-54
1
35,9
9,70E-07
9,65E-05
2,61E-12
CEA
Solid
γ
0,854
Mars.-96
2049
Sr-85
1
41,7
1,13E-06
3,52E+01
9,52E-07
CEA
Solid
γ
64,78
Mars.-96
2050
Y-88
1
34,7
9,38E-07
2,01E-15
5,44E-23
CEA
Solid
γ
0,293
Mars.-96
2051
Co-57
1
57,8
1,56E-06
2,38E-05
6,43E-13
CEA
Solid
γ
0,745
Mars.-96
2052
Cr-51
1
58,4
1,58E-06
2,17E-62
5,87E-70
CEA
Solid
γ
0,075
Mars.-96
2053
Mn-54
1
42,9
1,16E-06
1,15E-04
3,12E-12
CEA
Solid
γ
0,854
Mars.-96
2054
This document is the property of NECSA
and shall not be used, reproduced,
transmitted or disclosed without prior
written permission.
Doc. No.:
NLM-REP-14/197 Rev 2
Page No.:
Page 75 of 76
DRAFT SAFETY CASE FOR THE MANAGEMENT OF DISUSED SEALED RADIOACTIVE
SOURCES IN TUNISIA
A0
(Ci)
At
(KBq)
At
(Ci)
Origin
Nature
Ray
T (an)
Da
N.R
RN
Q
A0
(KBq)
Sr-85
1
45,1
1,22E-06
3,81E+01
1,03E-06
CEA
Solid
γ
64,78
Mars.-96
2055
Y-88
1
38,2
1,03E-06
1,05E-07
2,83E-15
CEA
Solid
γ
0,293
Mars.-96
2056
Mn-54
1
2,58
7,00E-08
4,75E-05
1,28E-12
CEA
Solid
γ
0,854
Juillet.-98
2057
Y-88
1
4,533
1,23E-07
6,08E-14
1,64E-21
CEA
Solid
γ
0,293
juillet.-98
2058
Co-57
1
0,783
2,12E-08
1,27E-06
3,45E-14
CEA
Solid
γ
0,745
juillet.-98
2059
Mixed
Gamma
1
Sans activité
-
-
-
CEA
Solid
γ
-
juillet.-98
2060
Cs-137
1
9,17E+03
2,48E-04
8,24E+03
2,23E-04
AIEA
Solid
γ
30,2
sept.-05
3001
I-131
1
1.11E+06
3,00E-02
2,66E-59
7,19E-67
Pharmacie
centrale
Solid
γ
0,022
Avr.07
3002
This document is the property of NECSA
and shall not be used, reproduced,
transmitted or disclosed without prior written
permission.
Doc. No.:
Page No.:
NLM-REP-14/197 Rev
1
Page 76 of 76
DRAFT SAFETY CASE FOR THE MANAGEMENT OF DISUSED SEALED RADIOACTIVE
SOURCES IN TUNISIA
20.0
APPENDIX E: SAFRAN ASSESSMENT
Form No.: NLM-DIV-FORM-00-002 Rev: 03