Sealed Radioactive Sources in The Kingdom of Morocco Documents/2014... · 2014. 10. 31. · Sealed Radioactive Sources in The Kingdom of ... (IAEA) requirements with regards to predisposal

Draft Safety Case for the Management of Disused

Sealed Radioactive Sources in The Kingdom of

Morocco

NLM-REP-14/191 Date: 2014-08-29

Nuclear Liabilities Management Necsa P.O. Box 582 Pretoria, 0001

South Africa

Prepared by: AL Visagie

REPORT No.: NLM-REP-14/191

DATE: 29 August 2014

TITLE: Draft Safety Case for the Management of Disused Sealed Radioactive Sources in Morocco

1.0 AUTHORIZATION

NAME SIGNED DATE

PREPARED AL Visagie

REVIEWED S Dhlomo

APPROVED GR Liebenberg

1.1 DISTRIBUTION

NO. NAME NO. NAME NO. NAME

1 NLM QA Records 8 15

2 IAEA 9 16

3 10 17

4 11 18

5 12 19

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7 14 21

* = Distributed via E-mail

This document is the property of NECSA and shall not be used, reproduced, transmitted or disclosed without prior written permission.

Doc. No.: NLM-REP-14/191

Page No.: Page 3 of 54

DRAFT SAFETY CASE FOR THE MANAGEMENT OF DISUSED SEALED RADIOACTIVE SOURCES IN MOROCCO

1.2 CONTENTS

1.0 AUTHORIZATION ....................................................................................................... 2

1.1 DISTRIBUTION ........................................................................................................... 2

1.2 CONTENTS ................................................................................................................. 3

2.0 PURPOSE ................................................................................................................... 5

3.0 SCOPE ........................................................................................................................ 5

4.0 REFERENCES ............................................................................................................ 5

5.0 ABBREVIATIONS ........................................................................................................ 6

6.0 DSRS MANAGEMENT DESCRIPTION IN MOROCCO ............................................. 6

6.1 LEGISLATION AND REGULATIONS RELATING TO THE MANAGEMENT OF DSRS IN MOROCCO ............................................................................................................. 6

6.2 REGULATORY BODY ................................................................................................. 7

6.3 NATIONAL SAFETY CRITERIA .................................................................................. 8

6.4 NATIONAL RADIOACTIVE WASTE MANAGEMENT POLICY AND STRATEGY ...... 9

6.5 WASTE OPERATOR .................................................................................................. 11

7.0 GENERIC ASSESSMENT CONTEXT ....................................................................... 11

7.1 PURPOSE OF THE SAFETY CASE .......................................................................... 11

7.2 SCOPE OF THE SAFETY CASE .............................................................................. 12

7.3 DEMONSTRATION OF SAFETY .............................................................................. 13

7.4 GRADED APPROACH .............................................................................................. 16

7.5 SAFETY STRATEGY ................................................................................................ 16

8.0 SITE, FACILITY AND PROCESS DESCRIPTION .................................................... 17

8.1 SITE DESCRIPTION ................................................................................................. 17

8.2 FACILITY DESCRIPTION ......................................................................................... 17

8.3 FACILITY OPERATION ............................................................................................. 22

8.4 DSRS INVENTORY ................................................................................................... 24

9.0 SAFETY ASSESSMENT ........................................................................................... 27

9.1 SAFETY ASSESSMENT CONTEXT ......................................................................... 27

9.2 SAFETY ASSESSMENT ENDPOINTS ..................................................................... 30

9.3 DEVELOPMENT OF SCENARIOS ........................................................................... 30

9.4 DATA USED AND ASSUMPTIONS MADE FOR THE SAFETY ASSESSMENT ...... 32

10.0 SAFETY ASSESSMENT ........................................................................................... 32

10.1 BASIC ENGINEERING ANALYSES .......................................................................... 32

10.2 QUANTITATIVE DETERMINISTIC ASSESSMENT OF WORKER DOSE................ 35





10.3 QUANTITATIVE DETERMINISTIC ASSESSMENT OF WORKER AND PUBLIC DOSE

FOR ANTICIPATED OPERATIONAL OCCURRENCE SCENARIOS: ...................... 39

10.4 DETERMINISTIC ASSESSMENT OF WORKER AND PUBLIC DOSE FOR ACCIDENT SCENARIOS .......................................................................................... 40

10.5 OPTIMIZATION OF PROTECTION: ASSESSMENT ................................................ 41

10.6 COMPARISON OF SPREADSHEET ASSESSMENT WITH SAFRAN ASSESSMENT ................................................................................................................................... 43

10.7 NON-RADIOLOGICAL HAZARD ASSESSMENT ..................................................... 43

10.8 ASSESSMENT OF THE IMPLEMENTED WASTE MANAGEMENT PRACTICE ..... 44

10.9 MANAGEMENT SYSTEM ASSESSMENT ............................................................... 45

10.10 ASSESSMENT OF UNCERTAINTIES ...................................................................... 46

11.0 IDENTIFICATION OF FACILITY SPECIFIC LIMITS AND CONDITIONS ................. 47

12.0 INTEGRATION OF SAFETY ARGUMENTS ............................................................. 47

12.1 FACILITY DESIGN AND ENGINEERING ................................................................. 47

12.2 FACILITY OPERATION ............................................................................................. 48

12.3 OPTIMIZATION OF PROTECTION .......................................................................... 48

12.4 WASTE MANAGEMENT PRACTISE ........................................................................ 48

12.5 INTEGRATED MANAGEMENT SYSTEM ................................................................. 49

12.6 UNCERTAINTIES ...................................................................................................... 49

13.0 COMPARISON WITH SAFETY CRITERIA AND CONCLUSIONS ........................... 49

14.0 ASPECTS REQUIRING FURTHER CLARIFICATION AND ACTION PLAN ............ 49

15.0 APPENDIX A – HOT SPOT DOSE CALCULATION ................................................. 52

16.0 APPENDIX B – SAFRAN DOSE ASSESSMENT ..................................................... 54





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 the Kingdom of Morocco (Morocco).

3.0 SCOPE

The scope of the draft safety case includes all the available information and identification of gaps in

information required to demonstrate the safety and ensuring the safety of all waste management

activities relating to Disused Sealed Radioactive Sources (DSRS) as performed by the National Centre

of Nuclear Energy, Science and Techniques (CNESTEN). This will include amongst others a description

of the legislation and regulations pertaining to the safe management of DSRS in Morocco, 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, limiting conditions, aspects that require

clarification, management systems and procedures required to ensure compliance to set safety criteria

and to sustain an acceptable level of safety.

The Safety Case and associated Safety Assessment for the management of DSRS in Morocco will take

the International Atomic Energy Agency (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 [2]. The safety criteria will be taken from

international safety standards and used as a basis for evaluation of safety and protection. [4]

4.0 REFERENCES

Number Title

1 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 Case and Safety Assessment for Predisposal

Management of Radioactive Waste, Safety Standards No.

GSG-3, IAEA, Vienna(2013).

3 NLM-REP-14/016 Mission Report – Safety Case Development in Morocco

4 GSR Part 3 IAEA, Radiation Protection and Safety of Radiation

Sources: International Basic Safety Standards (2014)





5.0 ABBREVIATIONS

CENM –Nuclear Studies Centre of Maamora

CNRP – National Centre of Radiation Protection

CNESTEN–National Centre of Nuclear Energy, Sciences and Techniques

DSRS – Disused Sealed Radioactive Sources

CSF – Central Storage Facility

RPO – Radiation Protection Officer

IAEA – International Atomic Energy Agency

6.0 DSRS MANAGEMENT DESCRIPTION IN MOROCCO

6.1 Legislation and Regulations Relating to the Management of DSRS in Morocco

The legal and regulatory framework of Morocco covering radiological protection and use of nuclear

energy is based on Law (No. 005-71) of 12 October 1971. This law establishes general principles as

basis for implementation of lower level regulations or decrees. The current regulations apply to the

importation, exportation, acquisition, production, transformation, detention, use, sale, transit,

transport, recycling and re-use of equipment or substances capable of emitting ionizing radiation.

They also apply to the treatment, handling, conditioning, storage, clearance and disposal of

radioactive substances or waste and to any other activity involving a risk arising from ionizing

radiation.

CNESTEN was established and assigned with the responsibility of managing radioactive waste

including DSRS by Law No. 12-02 of 2005. Morocco has draft radioactive waste management

regulations that address number of waste management aspects such a waste classification and

transportation of radioactive waste including DSRS etc. A National Commission of Nuclear Safety

(NCNS) was also created by decree number 2-94-666 of 7 December 1994.The Commission is

responsible for the regulation of nuclear installations. The Commission is overseen by the

Department of Energy and Mines. The decree number 2-97-30 of 28 October 1997 states the

general principles of protection against hazards resulting from the use of ionizing radiation which is

based on the ICRP recommendations (International Commission of Radiation Protection) and the

Basic Safety Standards of the IAEA. This decree mentions that the National Centre for Radiation

Protection (CNRP), which is under the Ministry of Health, is the Regulatory Body dealing with non-

nuclear facilities.





The government provides fund to the CNESTEN to finance safety of radioactive waste management

facility during its operating life-time and to carry out all operation related to the radioactive waste

management.

The producer of radioactive waste pays for the collection, the treatment and the storage of his

generated waste.

In Moroccan regulation there is no article where the financial aspect of decommissioning is specified

and who is the responsible party to provide funding, but as the Nuclear Centre of Maamora (CENM)

is a public research centre. The government assures the responsibility to cover any cost of its future

decommissioning including the reactor.

Morocco does not have a separate formalised National Radioactive Management Policy and Strategy

and therefore also not a formalised National Radioactive Waste Management Plan. It should

however be noted that a number of policy and strategy aspects related to DSRS are covered in

other regulations or decrees as addressed in succeeding sections.

6.2 Regulatory Body

As indicated in 6.1 above, Morocco currently has two regulatory bodies namely NCNS and CNRP.NCNS

is responsible for the regulation of nuclear installations while the responsibility of CNRP is as follows:

• To control all the activities relating to the use of the non-nuclear sources of ionizing radiation

(including Import, export, use, transport, storage, clearance and disposal);

• To proceed to radiological monitoring of the workers assigned to work with such sources;

• Radiological monitoring of the environment and the food stuffs;

• To be the centralized custodian for studies and information relating to the protection against

ionizing radiation;

• Implementation of the national regulations regarding the protection against ionizing radiation;

• To contribute to the follow-up of programs with radiological or nuclear application;

• To participate in the provision of information and training in the field of protection against ionizing

radiation.

• To take all the appropriate measures to avert radiological hazards in case of an incident involving

sources, devices, equipment and installations emitting ionizing radiation.





Financial and human resources are provided to CNRP by the Ministry of Health to execute its legal

responsibilities. The Ministry of Energy and Mines oversees nuclear activities through licensing and

inspecting procedures and can intervene extensively in the production and use of nuclear energy.

A draft law, currently at parliament for finalization, mentions the establishment of a single regulatory

body responsible for nuclear and radiological safety and security. Morocco is about to establish a single

regulatory body which will deal with nuclear safety and radiation protection in nuclear and non-nuclear

facilities. At this subject a main law was already drafted and discussed with all concerned ministries, it

is signed by the government council, the ministerial council and the counsellors’ room of the

parliament and now it is in discussion and finalization stage on the representative’s room of the

parliament. This law covers all the joint convention obligations and headings relating to the Safety of

Spent Fuel Management and the Safety of Radioactive Waste Management.

6.3 National Safety Criteria

The decree No. 2-97-30 of 28 October 1997 prescribes the general principles of radiation protection

of workers and the members of the public for the use of ionizing radiation. These principles are

applicable to facilities at which the spent fuel and radioactive waste are managed.

The regulation establishes justification, optimisation and limitation as the basic principles of

protection and specifies the general conditions and requirements applicable to the different groups

and situations.

At the nuclear installations where the spent fuel and radioactive waste are managed, the

operational radiation protection measures are:

• The classification and delineation of working areas

• The classification of the employees in different categories

• The individual monitoring and/or the monitoring of working area

• The application of dose limitation

• The establishment of procedures related to radiation protection

• The discharge of liquid or gaseous is surveyed and monitored

Discharge of radioactive gaseous and radioactive aqueous effluent are specified and quantified by a

ministerial order.





6.3.1 Protection of Workers

All the measures, to keep the exposure to radiation at the lowest level reasonably achievable are

adopted. Occupational exposure of workers, as it is mentioned in the decree of radiation protection,

should not exceed the following limits:

• Effective dose of 20 mSv per year on average over five consecutive years

• Effective dose of 50 mSv in only one year

• Equivalent dose to the lens of the eye of 150 mSv in only one year

• Equivalent dose to the extremities (hands, feet) or to the skin of 500 mSv in one year

Pregnant women are not allowed to work under the working condition ‘A’ where the annual

exposures under normal situation, can exceed the three tenth of the above mentioned limits. The

exposure of the pregnant woman must be as low as reasonably achievable. The decree does not

allow the appointment of any occupationally exposed worker under the age of 18 years.

Occupationally exposed workers are provided with dosimeters (TLD) which are read monthly by the

regulatory body (CNRP). In addition the workers use the individual direct reading dosimeters

allowing control of short term exposure. Occupational exposed workers are subject to annual whole-

body counting in order to assess radionuclide uptake and internal exposure.

6.3.2 Protection of Public

All the measures, to keep the exposure of the public to radiation at the lowest level reasonably

achievable are adopted. The CNRP has established the following public dose limits in the decree

related to radiation protection:

• An effective dose limit of 1 mSv per year.

• In particular circumstances, the effective dose limit may be authorised to reach 5 mSv in one

year on condition that the average over five consecutive years does not exceed 1 mSv per

year,

• Equivalent dose limit for the lens of the eye of 15 mSv per year,

• Equivalent dose limit for the skin of 50 mSv per year.

6.4 National Radioactive Waste Management Policy and Strategy

Although Morocco does not have a formalised separate National Radioactive Waste Management

Policy and Strategy an overall policy of radioactive waste management is the protection of humans

and their environment by collection, treatment and storage of radioactive waste. This policy is

implemented by adopting a centralised radioactive waste management approach where the





CNESTEN is the organisation responsible for the management of radioactive waste generated at

national level.

Radioactive waste is regulated under the regulation of radiation protection, decree of 28 October

1997, and the regulation applicable to nuclear installations, decree of 7 December 1994, under the

main law of 12 October 1971 which introduces the general principles that govern the use of

radioactive sources.

Morocco has draft radioactive waste management regulations that address number of waste

management aspects such a waste classification and transportation of radioactive waste including

DSRS etc.

Morocco recognizes and considers the safety standard and all documents related to the radioactive

waste management published by the IAEA.

The generators of radioactive waste should keep control on waste generation to the minimum

possible, segregate, collect and characterize waste according to the technical specification

established by the central operating organization (CNESTEN)

The high activity (Category 1and 2) disused radioactive sources like those used in therapy for

cancer are usually returned to their suppliers. Other sealed sources are also returned to their

suppliers, if the specific import license provides for it, if not, they are stored in the interim at the

holder’s/user’s facilities for collection by CNESTEN. The generators/users are required to pay for the

collect and the treatment of their waste including DSRS.

Therefore when sealed sources become disused there are only two options:

• Returning the disused source to the supplier

• Or transferring the disused source to the central waste management facility (CNESTEN)

Orphan sources are not frequently found in Morocco. In case of such event occurring, the

regulatory body takes control of the sources to ensure its safe storage and find the owner if

possible in order to pay the cost of its management and send it to the CNESTEN. Orphan sources,

which the owner can’t be identified, are transferred to CNESTEN for its management. A special

case related to the orphan sources which are detected in metallic scrap. The owner of metallic scrap

facility should inform the regulatory body and according to the law he becomes responsible for the

safety of the source until the source is transferred to the CNESTEN.





Morocco does not have any authorized disposal facilities and conditioned DSRS waste containers are

stored in a long term storage building of CNESTEN. Morocco does not have published clearance

criteria or levels but adopted the concept of clearance and the use of the applicable derived

clearance levels as published by the IAEA.

6.5 Waste Operator

The legislation in Morocco determines that CNESTEN carries the responsibility for the management of

radioactive waste in Morocco. This means that CNESTEN is also responsible for managing all

radioactive waste management facilities. Centralised waste management facilities was licensed and

established by CNESTEN which includes the DSRS treatment and storage facilities at CENM.

All the waste treatment, conditioning, handling, storing and transport operations are carried out by

CNESTEN in the radioactive waste management facilities except the management of spent fuel which

takes place in reactor building. The collection, transport, receiving and conditioning relating to DSRS,

forms part of the CNESTEN activities.

7.0 GENERIC ASSESSMENT CONTEXT

7.1 Purpose of the Safety Case

A safety case is a living document that should be developed already during the design stages of a

facility or the planning stages of an activity and be updated prior to the next stage e.g. construction,

commissioning, operation and decommissioning of such a facility. This will then form the basis for

phased regulatory decisions as well as operational decisions.

Ideally, predisposal waste management facilities or activities should be developed in a step by step

manner. The step by step approach adopted should inter alia enable:

• The systematic collection, analysis and interpretation of the necessary

scientific and technical data;

• The evaluation of possible sites, radioactive waste management options,

long term strategy and available technology;

• The development of plans for design and operation;

• Optimization; iterative studies for design, operation and safety assessment with

progressivelyimproving data and comments from technical and regulatory reviews.

In the case of the Moroccan facilities that are already established and activities relating to the

management of DSRS are already taking place. The purpose of the assessment will therefore be





retrospective; to determine whether the existing situation (facility lifecycle stage) is acceptable from

asafety and security point of view and proactive; to determine whether corrective action to upgrade

safety and/or security measures is necessary.

The following specific aspects will be addressed in this safety case:

• Demonstration of the safety of the CENM Waste Management Facilities

• Demonstration of the safety of various radioactive waste management activities performed by

CNESTEN. These activities include collection at users facilities, transport of DSRS to the JAEC

Waste Management facility, receiving and characterization of the DSRS, temporary storage,

conditioning and longer term storage.

• Optimization of the respective waste management activities described above.

• Management systems implemented in support and to ensure the safety of the respective waste

management activities described above.

• Definition of Limits, Controls and conditions that will be applicable to the facilities and the

respective activities described above.

• Input to the improvement of existing RP programmes and activity procedures.

7.2 Scope of the Safety Case

The scope of the safety case for Morocco is limited due to existing lifecycle stage of the facilities i.e.

operational and will therefore be focused on the as build facilities and the operational aspects of the

facilities which are defined as the:

• Collection and transport of DSRS to the centralized radioactive waste management facility at

CENM;

• Receiving, identification, characterization and handling of DSRS when it arrives at the

centralized facility at CENM;

• Temporary storage of the DSRS at the centralized facility at CENM;

• Conditioning of the DSRS and further long term storage.

• Handling and placement into final storage.

This 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 and operational activities.

• 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.

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 waste management facilities 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 (1mSv/a) even for the maximum anticipated

inventory.

- 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

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. (Stored DSRS and waste operations

are not in taking place in the same area)





- 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:

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 firefighting 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 so as to improve its shielding and security

performances.





7.3.2 Demonstration of the safety of various radioactive waste management activities performed by

CNESTEN

Quantitative and qualitative assessment 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 CNESTEN;

• 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 Morocco no management activities were taking place. Real time

measurements could therefore not be obtained for the activities. Radiological assessment will be based

using a realistic conservative 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

performed. This assessment will focus on Radiation Protection (RP), work procedures, Quality

Assurance (QA) aspects and processes for the management of operating 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

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. Besides the live

cycle stage of the Moroccan waste management facilities i.e operational (see point 7.1 above)

which limits the scope of this safety case, the main factors for justification of a limited approach 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

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 will be 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.

In view of the scope of the safety case for Morocco as defined in 7.1 above the following strategy for

demonstrating safety of the management of DSRS will be adopted:

• Defense in Depth – In this instance care is taken to ensure multiple safety layers. This principle

will be considered to ensure no important safety argument is based on a single level of protection.

• Passive safety– the use of passive systems that will be regarded as contributing to the safety:

• Shielding – Shielding will be used to ensure that doses to workers and also the public, are as low

as possible. The optimization of shielding usage during all waste management activities including

transportation and storage will be considered.





• Conditioning of DSRS– The shorter the time lapse from classifying sealed sources as disused until

conditioning in a well shielded retrievable form and accessible location, the higher the contribution

to the safety of the system.

• Implemented waste management practice – Approach to waste management with regards to the

following will be regarded as contributing to safety:

• Clearly defined responsibilities for waste management.

o Implementation of the principles of waste minimization and avoidance, namely, re-use

or re-processing of waste, return to supplier, safe and secure storage and conditioning

and final disposal of waste.

o 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.

8.0 SITE, FACILITY AND PROCESS DESCRIPTION

8.1 Site Description

CNESTEN manages the centralised waste management facilities at CENM. All DSRS that are collected

in Morocco and not returned to suppliers are brought to these facilities for conditioning and

storage.(No CENM site description and other site related information is available at this stage)

8.2 Facility Description

The only existing radioactive waste management facility in Morocco is waste treatment and storage

facility at CENM. It consists of an operational waste treatment building and waste storage building

for low and intermediate level radioactive waste:

− Waste treatment facility with a ground surface area of 472.5 m2 consisting of three levels (4.00,

0.00, +3.50). The underground level houses the storage tanks, the ground level houses the

waste receipt, interim storage and DSRS treatment areas, evaporation system, compaction

system and the radiochemical laboratory. The first floor houses the offices of the technical staff





and concrete laboratory (From here on the report will only refer to the DSRS provisions of the

facility)

− Long term storage facility consisting of four vaults (616 drums/vault), each one with a surface

area of 52 m2 (8.8m x5.9 m) and a height of 3.5 m. The thickness of the concrete wall is 0.40 m

Figure 1: Waste Management Facilities at CENM

Figure 2: Vault in Storage Building

8.2.1 Facility Design and Construction.

Basic information regarding design considerations, applied design and construction codes and

standards need to be obtained to justify the current facility designs and as build integrity and

stability. Design and construction documentation e.g. design review and certificates of construction

could be referenced here.

Waste Treatment Facility Waste Storage Facility





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 has been indicated that the seismic hazard of the region has been taken into account in the

design of the waste treatment facility, the long term storage facility and of the overhead cranes for

the handling of waste packages. Basic information regarding the ground accelerations level that the

facilities would be able to withstand and its justification should be provided.

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;- to be confirmed

• The floor slab is able to support the concentrated point loads of the waste containers 5 t/m2,

and an impact load of resulting from accidental dropping of waste container of 5 tons from a

height of 2 m, as well as live loads of vehicles/equipment used to load the packages;-to be

confirmed

• The slab is sufficiently thick around the building perimeter to support the walls and locally

around all internal stanchions; -to be confirmed

• Rain water is prevented from entering the buildings by surface contouring and drainage

channels around the buildings.-to be confirmed

• Resistance to water penetration from the ground is provided by a polyethylene damp proof

membrane to the underside of the slab; -to be confirmed

• The interior construction of the building is such that the risk of any liquids being released to

the environment is minimized;-to be confirmed

• The waste management buildings are 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.

Provision has been made for inspection of the sump and sampling of accumulated liquid.-to be

confirmed





• The floor slab has a steel floated finish with an epoxy paint coating to provide a hard wearing

and decontaminable surface;-to be confirmed

• Where ducts, pipes or cables that pass through walls or the floor, suitable means to

accommodate expansion and provide fire resistance is provided and is such that the structural

and fire integrity of the building is not impaired. Water proofing is applied at the entry point to

a building -to be confirmed

8.2.2.2 Shielding

• The waste treatment and storage building structures provide efficient shielding from

radiation to limit exposures outside the building to less than 10 x natural background levels.

The design and construction ensure the required shielding values are provided for (see dose

assessment assumptions) and that no major cracks or shine paths are present in the as

constructed building. Individual packages will be shielded by other packages, internal

building structures or by concrete blocks. A labyrinth type entrance is provided to the

storage areas.- to be confirmed

8.2.2.3 Access and Physical Security

• Physical security is provided primarily by a number of passive physical barriers including a

site perimeter fence, a site access point with security guards, strong building construction,

high integrity doors and locks to the treatment and storage areas. Buildings are equipped

with cameras, intrusion alarms and a biometric security access system.

• The buildings have vehicle access points. A separate personnel door is also 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. -to be

confirmed

• No windows are provided so as to improve its shielding and security performances.

8.2.2.4 Waste Treatment Facility Layout

The layout of the waste treatment facility is illustrated in Figure 3 below; (Diagram to be provided

by CNESTEN)

DSRS receipt and characterisation area





DSRS interim storage areas

DSRS dismantling area with dismantling table equipped with a shield (shielding capacity)

DSRS conditioning area

Conditioned DSRS package storage area

Supervised and restricted area

Areas in which DSRS are present are subject to radiation protection control measures. Interior

walls will separate the various areas and provide radiation shielding. Access to the storage areas

will be via a labyrinth type arrangement to provide easy access and at the same time reduce

radiation shine.

8.2.2.5 Waste Storage Facility Layout

The layout of the waste treatment facility is illustrated in Figure 4 below; (Diagram to be provided

by CNESTEN)

Waste receipt area

Waste storage vaults

• 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;

8.2.2.6 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 has been installed. Such equipment are tested and maintained. –to be

confirmed. 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.

8.2.2.7 Ventilation

(Details to be provided by CNESTEN)

The storage 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.





8.2.2.8 Electrical power and Lighting

Electrical power is provided for lighting, small power tools and detection/warning equipment. All

installations and equipment are of high quality and 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. (Electrical Certification of Compliance and

national standards could be referenced).

8.2.2.9 Mechanical handling equipment

Readily available and good quality manually operated mechanical handling equipment is available.

Such equipment is subject to national regulation/requirements as applicable to statutory

equipment and is used/operated by trained/licensed operators (national requirements could be

referenced).

8.3 Facility Operation

Operational activities within the waste management facilities involve reception, treatment and

emplacement of packages, inspection of DSRS, equipment and the stored packages and

maintenance of the building and equipment. 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

that it makes these operations simple and easy to undertake in the least time possible. Written

operational procedures are 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 are formally

documented and approved, an incident reporting system and emergency plans are drawn up and

approved. – To be confirmed. These programmes will be updated based on and justified by this

safety case.

Records are maintained of all operational activities, packages and equipment are clearly marked

and labelled and an inventory maintained of all equipment, DSRS and waste placed in the store.





8.3.1 Waste Treatment Facility Operation

• The transported DSRS are received at the Central Waste Management Facility at CNESTEN.

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 a temporary storage location in the Central Waste

Management Facility at CNESTEN. The storage location is inspected and surveyed monthly

by operators and an RPO

• Standard Category 3 sources are collected from a batch of sources in their working shields

and placed on a working area equipped with a shield. The sources are removed from their

working shields inspected, recorded and placed into a shielded storage container. The

storage container is also closed and prepared for final storage. 50 sources are conditioned

per campaign and 2 campaigns are performed per year

• Non Standard Category 3 sources are collected from a batch of sources 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.

• Once the container with the DSRS has reached its filling capacity, the container is

conditioned by filling it with concrete. The 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.

• The facility is visited for approximately 8 hours per week for general cleaning, inspection

and maintenance purposes.

8.3.2 Waste Storage Facility Operation

• Cured waste packages are transferred to the waste storage facility and emplaced in the

storage vaults. Two such campaigns are conducted per annum.

• The waste storage facility is inspected and monitored on a monthly basis

• The facility is visited for approximately 2 hours per week for general cleaning, inspection

and maintenance purposes.





8.3.3 Operational Radiation Protection

The waste treatment and storage facilities are designated as a radiologically 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 has been implemented and 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

makes provision to monitor external radiation levels and surface contamination

8.3.4 Management System

The establishment and implementation of an integrated management system is paramount to 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

CNESTEN.

Written operational procedures are 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.

The formally documented and approved management system 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. (to be confirmed) The

management system inter alia includes an incident reporting system, emergency plans and

document and record management. The integrated management system is continuously updated

and will be revised to reflect the recommendations from this safety case.

Records are maintained of all operational activities, packages and equipment are clearly marked

and labelled and an inventory is maintained of all equipment and waste placed in the store.

8.4 DSRS Inventory

No manufacturing of sealed sources takes place in Morocco. About 1500 radioactive sealed sources

are used in 150 establishments. The main radio-isotopes in use are Cs137 and Co 60 sources.





The analysis of the quantity of the radioactive sealed sources used in Morocco and their applications,

shows that the big part of these sources are used as radiometric gauges: level gauges, density

gauges, thickness gauges and moisture gauges, which are often incorporated into fixed installations,

applications in the industrial radiography, in medicine and in the logging, then finally, a small quantity

of sources are used in the laboratory analysis, quality control and in the calibration process.

The current inventory of the inventory of DSRS and sealed sources in use are reflected by the

figures below: (Note that the figures below are extracts from the presentation of NCRP).

• Figure 5- DSRS Inventory at CNESTEN

• Figure 6- DSRS Inventory at User facilities

• Figure 7- Sealed Sources in use in Morocco

Figure 5: DSRS Inventory at CNESTEN





Figure 6: National DSRS Inventory

Figure 7: Sealed Sources in Use in Morocco

Figure 6

Figure- 3





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 covers some information

related to the strategy for safety assessment which will be expanded in this section.

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. The table in Section 10 will also include

identified unresolved issues and recommended retrospective corrective action.

9.1.1.2 Demonstration of the safety of the radioactive waste management activities performed by

CNESTEN.

• For normal operation; quantitative deterministic assessment of worker dose due to the range

of activities by various occupational groups of CNESTEN using Excel spreadsheet calculations

and SAFRAN; the breakdown of normal operational activities are the following:

- Collection of DSRS at Interim Stores: Three Loaders from CNESTEN 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 CNESTEN. The drivers are CNESTEN employees. The

vehicle is equipped with a 6 mm lead shield between the sources and the driver

positions.

- The transported DSRS are received at the Central Waste Management Facility at

CNESTEN. 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 a temporary storage location in the Central Waste

Management Facility at CNESTEN. The storage location is inspected and surveyed

monthly by operators and an RPO.

- Standard Category 3 sources are collected from a batch of sources in their working

shields and placed on a working area equipped with a shield. The sources are

removed from their working shields inspected, recorded and placed into a shielded





storage container. The storage container is also closed and prepared for final

storage. 50 sources are conditioned per campaign and 2 campaigns are performed

per year.

- Non Standard Category 3 sources are collected from a batch of sources 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 container with the DSRS has reached its filling capacity, the container is

conditioned by filling it with concrete. The 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.

- The waste treatment facility is visited for approximately 4 hours per week for general

cleaning, inspection and maintenance purposes.

- Cured waste packages are transferred to the waste storage facility and emplaced in

the storage vaults. Two such campaigns are conducted per annum.

- The waste storage facility is inspected and monitored on a monthly basis

- The waste storage facility is visited for approximately 2 hours per week for general

cleaning, inspection and maintenance purposes.

The SAFRAN tool will be utilized to compare worker doses performing certain activities with

the Excel spreadsheet calculations.

• 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

such as Excel spread sheets and “Hot Spot” 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.





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. 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 withregard 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 re-

processing 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.

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

performed. 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 quantitate assessment endpoints will be applicable:

• Radiation dose to workers performing the various normal DSRS management activities at

CENM and radiation doses to worker and the public as applicable due to anticipated

operational occurrences. It should be noted that the same CNESTEN personnel is performing

all the respective DSRS management activities at CENM. 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].

• The assessments will cover activities taking place over a 1 year period.

9.3 Development of Scenarios

9.3.1 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.)

9.3.2 Accident Scenarios

9.3.3 Anticipated Operational Occurrence Scenarios

The consequence of following postulated initiating events will be evaluated in the Morocco Safety

Assessment:

- Occurrence 1 Scenario: The transport vehicle carrying three working shields with DSRS is

involved in an accident. The vehicle capsizes 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. The sources are recovered and surveyed by CNESTEN RPOs





and operators (30 min in close proximity) who then continue with loading and transportation of

the sources.

- Occurrence 2 Scenario: The operator left a Cat 3 Co 60 source on the workbench during the

removal of the source from its working shield in the waste treatment facility at CENM. The

operator did not wear his EPD and was under the impression that the source was placed inside

the shielded waste container and continued to work on another 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 evacuated the working area after which

the misplaced source was detected and placed in the shielded container.

- Occurrence 3 Scenario: 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 container.

9.3.3.1 Other Accident Scenarios

The following other accident scenarios will be considered in the Moroccan Safety Assessment:

- Accident Scenario 1: The electrical wiring in 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 firefighting personnel

could arrive. A 50 mCi Cs-137 source is ruptured in the process and starts leaking. Firefighting

personnel arrive and by using powder based fire-fighting equipment managed to quench the

fire. With an assumed release fraction of 10 %, contamination spread by the fire into the

facility while 20 % of the released activity escaped from the building through the natural

ventilation system and through the opened doors to the environment. Firefighting personnel

used respirators and spend 20 minutes in the contaminated zones. After the fire was put out,

the remaining activity settled in the areas. Workers used protective suits and respirators to

clean-up the contaminated zones.

• Accident Scenario 2: During transport of two 10 mCi Cs-137 sources inside their working

shields the transport vehicle is in involved in an accident and caught fire. The operators are not





in a position to remove the units from the vehicle. Due to the extreme heat from burning 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 are in close proximity of the

burning vehicle and exposed to the dispersed contamination. A release fraction of 10 % and

conservative (not good) metrological conditions are assumed.

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 Morocco 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 but also realistic.

10.0 SAFETY ASSESSMENT

10.1 Basic Engineering Analyses

Table 1: Basic Engineering Analysis

Item Requirement Compliance Ref Comments 1. General: Facility Design, Construction and Maintenance 1.1 Basic site characteristics and

credible external events have been considered in the design

To be confirmed

1.2 Quality assurance has been considered in the design, construction, maintenance and modification the waste management facilities: • 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)

No design information has been supplied. Design approval and certificates to be supplied. Plans to be developed or supplied. To be supplied





Item Requirement Compliance Ref Comments 2. Safety and security aspects were considered in the design of the facility 2.1 The characteristics of the walls

ensuring a level of dose rate that complies with the restriction for public exposure (1 mSv/a) even for the maximum anticipated inventory.

To be modelled or to be included as a facility limits and included in the procedure for management of the facility limits and conditions.

2.2 The lighting system will be adequate and permits the performance of operations in a safe manner.

To be demonstrated by facility lighting measurement (lumens) to be repeated on an annual basis.

2.3 Physical delineation of areas designed for storage and for the main waste management operations are isolated, this way it is ensured the appropriated segregation of materials optimizing worker’s exposure during operations

To be demonstrated for all interim storage areas. Main storage area is isolated from the waste operations areas.

2.4 Each delineated area has a sufficient physical space that ensures a minimal probability of accident occurrence during waste management operations and package handling.

To be assessed.

2.5 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)

To be assessed.

2.6 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.

To be demonstrated.

2.7 Unconditioned radioactive sources are stored in storage systems ensuring normal operation and minimizing probability of accidents. Their main characteristics are:

To be assessed.





Item Requirement Compliance Ref Comments

• 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.

Total inventory limits to be developed. 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. Maximum load capacities to be demonstrated.

2.8 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.

To be confirmed.

3. Engineering systems ensuring safety for situations of occurrences and accidents 3.1 Floor and wall finish allow easy

decontamination. To be confirmed.

3.2 The segregation of the different areas limits the potential dispersion of any contamination.

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

3.3 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.

See 3.2 above. To be confirmed

3.4 The facility has its own fire detection and firefighting equipment.

To be confirmed.

4. Facility design provides physical security features commensurate with the security threat





Item Requirement Compliance Ref Comments 4.1 Robust building construction

with high integrity doors and locks to the treatment and storage areas.

Facility inspection showed robust building construction with high integrity doors and locking systems.

4.2 Buildings are equipped with intrusion alarms.

To be confirmed.

4.3 The buildings have vehicle access points. A separate personnel door is provided to segregate personnel from vehicle movements.

To be confirmed.

4.4 No windows are provided so as to improve its shielding and security performances.

Process and storage areas are not equipped with windows

10.2 Quantitative Deterministic Assessment of Worker Dose

10.2.1 Activity 1: Collection of DSRS at Interim Stores

Table 2: Collection of DSRS at Interim Stores

10.2.2 Activity 2: Collected DSRS are transported in a dedicated vehicle by 2 drivers from Users and

interim storage locations to CNESTEN.

Table 3: Transport to CSF

Dose rate [µSv/h] Justification/Notes Time per action [h] Actions per yearWhole Body 10 0.25 10 25Extremity 30 0.25 10 75Whole Body 10 1.75 10 175Extremity 30 1.75 10 525

Annual dose [µSv/a]

1

Loaders(3) Inspection and ID TI or measured DR on consignments

Loading

Operator Groups Operator Actions Exposure Type Exposure Data Exposure time

Dose rate [µSv/h] Justification/Notes Time per action [h] Actions per yearWhole Body 5 6 10 300


1

Drivers (2) Driving TI or measured DR on

consignments-50 % reduction in DR due to installed shield






10.2.3 Activity 3: Receiving of DSRS at CENM

Table 4: Receiving at CSF

10.2.4 Activity 4: Temporary Storage of Category 3 Sources.

Table 5: Temporary Storage

10.2.5 Activity 5: Conditioning Campaign 1: Standard Cat 3 Sources

Table 6: Conditioning Campaign 1

Operator Groups Operator Actions

Exposure Type Exposure Data Exposure time


Dose rate [µSv/h] Justification/Notes Time per action [h] Actions per year

1

Transporter(2)

Transport

Whole Body 10 Dose rate at 1 m (1) 0.1 100 100

Extremity 30 Contact dose rate (2)

0.017 100 51 2

Operators(2) Handling

Whole Body 10 Dose rate at 1 m (1)

0.017 100 17


0.017 100 51

Dismantling Whole Body 10 Dose rate behind

shield (3) 0.1 100 100


Whole Body 10 Doserate at 1 m (1) 1 20 200Extremity 30 Contact doserate (2) 0.1 20 60

Whole Body 10 Doserate at 1 m (1) 0.2 20 40Extremity 30 Contact doserate (2) 0.1 20 60


(1) - TI or measured DR on consignments(2) - Maximum doserate measured on contact of a cat3 source assembly

2

RPO Surveying


1

Loaders (2) Off loading

Inspection and segregation




Whole Body 10 Ambient doserate (3) 0.2 20 40Extremity NA

Whole Body 10 Doserate at 1 m (1) 0.1 12 12Extremity NA

(1) - TI or measured DR on consignments/ reference to documents/attachments(2) - Maximum doserate measured on contact of a cat3 source assembly/reference to documents/attachments(3) - Average measured ambient dose rate in the interim storage location

2

RPO (1)

Inspection & surveying


1

Operators(2) Placement

Inspection






Operator Groups Operator

Actions Exposure Type

Exposure Data Exposure time Annual dose

[µSv/a] Dose rate [µSv/h] Justification/Notes Time per action [h] Actions per year


0.1 100 300

Source Transfer

Whole Body 1000 Dose rate behind shield (4) 0.01 100 1000

Extremity 10000 Unshielded DR (5)

0.005 100 5000

Inspection and maintenance

Whole Body 200 Dose rate at 1 m (6)

0.3 2 120

Extremity 2000 Contact dose rate (7) 0.3 2 1200 3 RPO (1)

Supervision & surveying

Whole Body 10 Ambient dose rate (8) 0.5 100 500

Extremity NA

(1) -

TI or measured DR on consignments/ reference to documents/attachments

(2) -

Maximum dose rate measured on contact of a cat3 source assembly/reference to documents/attachments

(3) -

Maximum measured dose rate behind shield with shielded cat 3 source

(4) -

Maximum measured dose rate behind shield with unshielded cat 3 source

(5) -

Calculated dose rate at 15 cm from an unshielded cat 3 source (Maximum activity)

(6) -

Maximum measured dose rate 1 m from a full source storage container

(7) -

Maximum dose rate measured on contact of a full source storage container assembly/reference to documents/attachments (8) -

Maximum measured ambient dose rate in area during conditioning of sources

10.2.6 Activity 6: Conditioning Campaign 2: Non-Standard & Linear Cat 3 Sources and Facility Surveillance,

Inspection and Maintenance

Table 7: Conditioning Campaign 2

Operator Groups Operator Actions

Exposure Type Exposure Data Exposure time



1

Transporter(2)

Transport



0.017 3 3 2

Operators(2)

Handling



0.017 3 3

Dismantling

Whole Body 25 DR in shield (3)

0.17 3 13


0.17 3 26

Source Transfer

Whole Body 5000 DR on open source (4)

0.01 3 150

Extremity 10000 DR on open source (5)

0.01 3 300


Whole Body 10 Ambient dose rate (6)

0.5 10 50

Extremity N/A 3

RPO (1)

Supervision & surveying

Whole Body 10 Ambient dose rate (7) 0.5 3 15

Extremity NA

Facility Surveillance

Whole Body 10 Ambient dose rate (6) 1 12 120

Extremity N/A





(1) -

TI or measured DR on consignments/ reference to documents/attachments

(2) -

Maximum dose rate measured on contact of a Cat 3 source assembly/reference to documents/attachments

(3) -

Maximum measured dose rate without a shield on a shielded Cat 3 source at a distance of 0.5 m from the source

(4) -

Maximum dose rate at 1.5 m from an unshielded Cat 3 source

(5) -

Calculated dose rate at 0.5 m from an unshielded Cat 3 source (Handling with tonguesy)

(6) -

Maximum measured ambient dose rate in area during conditioning of sources

(7) -

Maximum measured ambient dose rate in facility

10.2.7 Activity 7: Transfer of Conditioned Waste Packages to the Waste Store including its Surveillance,


Table 8: Transfer to Waste Storage

10.2.8 Worker Dose Summary

The maximum worker dose is summarised in the Table 9 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 functions in both facilities.

Table 9: Worker Dose Summary

Operator Groups

Operator Actions

Exposure Type

Worker Dose Per Activity [uSv/a] 1 2 3 4 5 6 7

Loaders/ Transporters

Inspection Loading/ off

Whole Body 200 240 Extremity 600 120

Transport Whole Body 300 100 11 120 Extremity 51 3 200

Operators

Handling Whole Body 20 17 1 120 Extremity 36 51 3 200

Dismantling Whole Body 100 13 Extremity 300 26

Source Transfer Whole Body 1000 150 Extremity 5000 300

Inspection and Whole Body 40 120 50 1000




Whole Body 10 Ambient doserate (3) 2 50 1000Extremity NA

RPO (1) Whole Body 10 Ambient doserate (3) 0.5 12 60Extremity NA

(1) - Maximum measured doserate 1 m from a full source storage container(2) - Maximum allowable contact doserate allowed on a full source storage container interms of the transport regulations(3) - Maximum measured ambient doserate in area with maximum inventory

Exposure Data Exposure time Annual dose [µSv/a]

1Transporter(2)

Transport

Operator Groups Operator Actions Exposure Type

2

Operators(2) Handling


3 Supervision & surveying





Operator Groups

Operator Actions

Exposure Type

Worker Dose Per Activity [uSv/a] 1 2 3 4 5 6 7

Maintenance Extremity 1200

RPO

Supervision and Surveying

Whole Body 40 12 500 15 Extremity 60

Facility Surveillance

Whole Body 135 60 Extremity

The maximum total dose to the Operator/Loader/Transporter is therefore: Whole body: 3.6 mSv/a

Extremity: 8.1mSv/a

The maximum total dose to the RPO is therefore: Whole body: 0.747 mSv/a Extremity:

Insignificant

10.3 Quantitative Deterministic Assessment of Worker and Public Dose For Anticipated

Operational Occurrence Scenarios:

The scenarios as defined in section 9.3.3 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 25 µSv/h from a shielded cat 3 source as used in

10.2.5 with the exposure times of 1 hour and 30 min for the public and workers respectively:

The maximum public dose would therefore be 25 µSv or even 50 µSv if simultaneously irradiated

by 2 sources. The maximum additional dose to the worker would therefore be in the order of

25µSvif 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 calculated in

Table 5 to 45 minutes.

The maximum additional dose to the worker would therefore be; Whole body 750 µSv and

extremity 7500 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 calculated in

Table 6 to 15 minutes.





The maximum additional dose to the worker would therefore be; Whole body 1250 µSv and

extremity 2500 µSv.

10.4 Deterministic Assessment of Worker and Public Dose for Accident Scenarios

Accident Scenario 1;- The maximum public and additional worker dose is projected by the

following:

The maximum public and additional worker doses projected for the occurrence and scenario as

defined in section 9.3.3.1 above are derived based on the assumptions, calculations and modelling

indicated in the Table 10 below:

Table 10: Accident Scenario 1

Accident Scenario 2; The maximum public dose is projected by the following:

The maximum public dose for this occurrence and scenario as defined in section 9.3.3.1 above are

derived based on the assumptions, calculations and modelling indicated in Table 11 below:

Units [x] Justification/Notes Time per action [h] Other/Units [x]

Whole Body 1000 [µSv/a] Ambient doserate (1) 0.3 300

Internal Radiation 1E6 [Bqm-3] Activity Conc. (2) 0.3

Respirator. eff. -0% Breathing. Rate 1.2 m3/h AMAD 1 µm DCF 4.8E-9 [SvBq-1]

1728

2

Public Working on site

Internal Radiation and exposure from cloud & ground shine

9.7E-4

Whole Body 200 [µSv/a] Ambient doserate (1) 16 3200

Internal Radiation 2E7 [Bqm-2]Surface Contamination (5) 16

Resuspension F. 1E-6 m-1

Respirator. eff. -0% Breathing. Rate 1.2 m3/h AMAD 1 µm DCF 4.8E-9 [SvBq-1]

1.843

(1) - Elevated ambient dose rate due to the maximum release of 5 mCi Cs-137 into the area- to be confirmed(2) - Projected airborne activity concentration levels calculated on the assumption that the total released fraction of 5 mCi (10%) becomes

homogeneously dispersed in a 200 m3 area (3) - Hot spot dispersion modelling assuming a 1 mCi release, long term exposure (4 days) conservative metrological conditions and

the distance from release with highest concentration (Appendix A)(4) - Average Elevated ambient dose rate due to the maximum release of 5 mCi Cs-137 into the area over clean-up period- to be confirmed(5) - Maximum Projected surface contamination level levels calculated on the assumption that the total released fraction of 5 mCi (10%) settles

homogeneously on a 10m2 area

Clean-up

Receptors Actions

1

Firefighting Personnel

Firefighting

Exposure Type Exposure Data Exposure Parameters Dose [µSv]

Dispersion and dose modelled with Hotspot assumming a ground level release (3) See Appendix A

3

Operators





Table 11: Accident Scenario 2

10.5 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 Table 12 below.

Table 12: Optimization of Protection: Assessment

Occupational Group/

Receptor

Dose /Dose Rate [µSv/µSv/a]

Comments Recommendations

Whole Body/ED

Extremities

1. Normal Operation: Quantitative Deterministic Assessment of Worker Dose Operator/Loader/Transporter

3600µSv/a 8100 µSv/a 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 action, which is a needed and justified action.

• 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

RPO 747 µSv/a (Insignificant) RPO invigilation is justified ito. dose limitation and control. RPO dose is below current optimization trigger level.

• None

2. Anticipated Operational Occurrences: Quantitative Deterministic Assessment of Worker and Public Dose : Occurrence Scenario 1

Operator/Loader 25µSv - Exposure levels are • Actions to ensure

Units [x] Justification/Notes Time per action [h] Other/Units [x]2

Public Working on site

Internal Radiation and exposure from cloud & ground shine

1.9E-3 (2)

(1) - Hot Spot dispesion modelling assuming a 1 mCi release, long term exposure (4 days) concervative metrological conditions and the distance from release with highest concentration (Appendix A)

(2) - Hot Spot dose was adapted to make provision for a 2 mCi release due to the 10 % releases fraction assumption (linear relationship)

Dose [µSv]

Dispersion and dose modelled with Hotspot assumming a ground level release (1) See Appendix A

ActionsReceptors Exposure Type Exposure Data Exposure Parameters





Occupational

Group/ Receptor



Whole Body/ED

Extremities

/ Transporter below optimization trigger levels and sufficient control is inherent to the compliance to the transport regulations

compliance to the transport regulations.

Public 50 µSv -


Operator/Loader/ Transporter

750 µSv 7500µSv Expose levels are low but possible to prevent by simple design changes.

• Evaluate the possibility to install a radiation alarm system with an alarm set point of about 150 µSv/h (response, testing and maintenance procedures)



1250 µSv 2500µSv Possible to prevent exposure by simple design and operational changes.

• Evaluate the possibility to install a radiation alarm system with an alarm set point of about 150 µSv/h (response, testing and maintenance procedures)

• 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

Firefighting Personnel/Public

2028µSv - Dose mainly due to external radiation. Dose due to contamination and dispersion of beta gamma emitters is low. Possible to prevent exposure by simple design and operational changes to prevent fires and to mitigate the consequences of fires.

• Initiate a fire and fire protection system evaluation of the areas.

• Assess the possibility to store unconditioned sources in vaults or other fire proof system.

• Review procedures to ensure housekeeping and storage practices that are aligned with fire prevention and control measures.

Public (Insignificant) -


3200µSv -





Occupational

Group/ Receptor



Whole Body/ED

Extremities

5. Other Occurrences: Quantitative Deterministic Assessment of Worker and Public Dose : Accident Scenario 2

Public (Insignificant)

- Exposure levels are below optimization trigger levels and sufficient control is inherent to the compliance to the transport regulations

• Actions to ensure compliance to the transport regulations

10.6 Comparison of Spreadsheet Assessment with SAFRAN Assessment

A dose assessment for the DSRS activities at the JAEC CSF was also performed using the SAFRAN

dose assessment tool. The purposes of this assessment was to enable a comparison between the

simple spreadsheet assessment as performed in Sections 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 B.

The results of the SAFRAN assessment for the respective activities assessed are the same as

calculated in the spreadsheets above in Section 10.2.

10.7 Non-radiological Hazard Assessment

The following non-radiological hazards are relevant to the operation of the Waste management

facilities at CENM:

• 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.

Needs to be expanded and confirmed

10.8 Assessment of the Implemented Waste Management Practice

The outcome of the quantitative assessment of the waste management practice as implemented by

CNESTEN is tabled below.

Table 13: Assessment of Implemented Waste Management Practice

Item Requirement Compliance Comments Ref 1. Clearly defined responsibilities

for waste management. The Legal Framework of Morocco specifies the responsibilities for the generation and management of radioactive waste. The construction and operation of the waste management facilities demonstrate intent and commitment

Section 6.1

2. Implementation of the principles of waste minimization and avoidance, namely, re-use or re-processing of waste, return to supplier, safe and secure storage and conditioning and final disposal of waste.

Principles defined in Legal Framework and implemented in the case of DSRS to the point of conditioning. No final disposal option is available.

Section 6

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.

The treatment of standard DSRS is well planned and executed in a facility which is designed to mitigate exposure. Facilities to treat non-standard sources or deviating e.g. contaminated sources do not exist. No procedures to assess and plan the handling of non-standard sources have been supplied.

4. 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.

Although no formal WAC document for the receipt of DSRS was supplied, consignments of DSRS are assessed at the generators facility and again at the CNESTEN treatment facility as part of collection and transport procedure. No written conditioning specification or a WAC for the storage facility was supplied. It was also not indicated how the current conditioning actions and specification are aligned with future disposal options

Section 8.3

5. Interim storage of DSRS will only take place inside proper containment such as the original working shields or another type

All sources are received in their working shield in compliance to the transport regulations. Sources are only stored in their working shield

Section 8.3





Item Requirement Compliance Comments Ref

of suitable containment. or in a waste container.

6. Conditioned DSRS will be stored in a dedicated storage area with passive safety features and adequate access control.

Only conditioned waste packages are transferred and emplaced in a dedicated long term storage facility

Section 8.3

10.9 Management System Assessment

The outcome of the quantitative assessment of only the main requirements of an integrated

management system as implemented by CNESTEN is tabled below in Table 14:

Table 14: Management System Assessment

Item Requirement Compliance Comments Ref 1. A written and approved integrated

management system is maintained to ensure a sustained level of safety during the operational phase of the facilities.

No written and approved management system documents have been supplied to date.

2. 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

3. An RP programme exist and inter alia covers: • RP organisation, training and

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





Item Requirement Compliance Comments Ref 4. 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

10.10 Assessment of Uncertainties

The outcome of a provisional quantitative assessment of uncertainties related to the safety case is

presented in the table below:

Table 15: Assessment of Uncertainties

Item Uncertainty Comments/ Recommendations

Ref

1. 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.

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 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.

2. Uncertainty regarding the dose rate information used in the safety assessment. Although it was aimed to use conservative data, the exposure data used for the various exposure scenarios is not based on scientific arguments, measurement or modelling results.

Confirmatory monitoring should be performed and used to verify the dose rate assumptions or be used as bases to update exposure scenarios and data.





11.0 IDENTIFICATION OF FACILITY SPECIFIC LIMITS AND CONDITIONS

Based on the safety assessment, the following facility operational limitations and conditions are

derived:

• The number of waste standard and non-standard (ad-hoc) waste treatment and conditioning

campaigns should be specified. Based on the current safety assessment, such campaigns could

be increased to 4, 6 and 4 standard and non-standard waste treatment and DSRS container

conditioning campaigns respectively.

• Specify Sources in terms of 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.

• The maximum inventory for the storage facility needs to be derived and specified

• 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 exceed

2.5µSv/h.

• Annual reporting of facility operations and RP surveillance data to the regulatory body.

12.0 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;

12.1 Facility Design and Engineering

Although a range of facility design, engineering and construction related aspects have been

identified as relevant to safety, still need to be obtained/demonstrated, the as build facilities seems

robust with features that indicates 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 14 below.

12.2 Facility Operation

The safety assessment indicates that the facilities 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 14)

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.

12.3 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 be considered for further optimization of protection. An

operational optimization of protection program, that is based on activity specific RP surveillance,

personnel dosimetry results and scheduled optimization review sessions, is recommended.

12.4 Waste Management Practise

Good waste management practice is generally evident from the intend of the legal framework,

organisational arrangements and defined responsibilities, 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 14 below.





12.5 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 14. below, addresses recommendations regarding the development of and integrated

management system.

12.6 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 aspects that need

clarification as covered in section 14 below, covers management of uncertainties.

13.0 COMPARISON WITH SAFETY CRITERIA AND CONCLUSIONS

The Quantitative safety assessment results as reflected in section Table 12 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 CENM is supported subject to a

formal plan and schedule to address the identified unresolved issues as covered in section 14

below.

14.0 ASPECTS REQUIRING FURTHER CLARIFICATION AND ACTION PLAN

The safety case performed above indicated some information gaps that need to be addressed

before it will be regarded as a document that can be submitted to the regulatory authority for

review and approval.

The identified aspects requiring further clarification with commensurate management

recommendations and actions are tabled below:

Table 16: Aspects requiring further Clarification

Item Aspects Requiring Clarification Recommendation/Action 1. Moroccan Legal and regulatory Framework 1.1 During the preparation of this case, a draft law of

morocco was promulgated (Law 142-12). The draft law is structured to cover the following topics:

To update section 6 of the report to include the provisions of the new legislation. Information to be provided





Item Aspects Requiring Clarification Recommendation/Action

• Nuclear and radiological safety and security: • Definitions; • General provisions; • Licensing and notification processes; • Common provisions to licensing and notification

processes; • Licensing of radioactive waste management

activities; • Protection against ionizing radiation sources; • The use of ionizing radiation sources for medical or

dental purposes; • Physical protection, security safeguards and non-

proliferation; • Emergency planning; • Accreditation of services providers.

by CNESTEN.

2. Basic Engineering Analyses

2.1 A number of unresolved issued and gaps have been identified in the basic engineering analyses as listed in section 10.1 above that need to be resolved or managed.

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.

3.Optimization of Protection

3.1 Optimization Normal Operation related exposure • 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.

3.2 Optimization of occurrence related exposure. • Actions/audit to ensure/verify 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/ non-standard sources and planning of its storage and treatment.

• Initiate a fire and fire protection system evaluation of the areas.

• Assess the possibility to store unconditioned sources in vaults or other fire proof systems.

• Develop procedures (inspection and testing) to ensure housekeeping and





Item Aspects Requiring Clarification Recommendation/Action

storage practices relating to fire prevention and control are established and maintained.

4. Non-Radiological Hazards

4.1 Comprehensive assessment of non-radiological hazards.

• Plan, schedule and conduct a comprehensive non radiological hazard assessment.

5. Implemented Waste Management Practice

5.1 WAC. (Covered in integrated management system 6. below)

5.2 Interdependencies related to disposal • National waste management plan to make provision for disposal- could be a longer term action but commitments related to disposal are necessary.

6.Integrated Management System

6.1 No written and approved management system documents have been provided.

• Plan and schedule an integrated management system review that is focussed the main requirements as listed in the table in 10.8 above.

7. Management of Uncertainties

7.1 Uncertainties related to source term. • (Covered by Facility limits and condition in 8. below and be actions to develop WAC in as covered in 6. above)

7.2 Uncertainties regarding dose rate assumption. • Develop and implement a confirmatory monitoring plan to verify the dose rate assumptions. This could be used as bases to update exposure scenarios and data.

8. Facility Specific Limits and Conditions

8.1 Procedure for defining and management of facility specific limits and conditions

• 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.





15.0 APPENDIX A – HOT SPOT DOSE CALCULATION HotSpot Version 3.0 General Fire (Aug 06, 2014 09:03 AM)

Source Material : Cs-137 F 30.0y Material-at-Risk (MAR) : 3.7000E+07 Bq Damage Ratio (DR) : 1.00 Airborne Fraction (ARF) : 1.00E-02 Respirable Fraction (RF) : 1.00E+00 Leakpath Factor (LPF) : 1.000 Respirable Source Term : 3.70E+05 Bq Non-respirable Source Term : 0.00E+00 Bq Release Radius : 1 m Cloud Top : 10 m Physical Height of Fire : 5 m Effective Release Height : 8.52 m Wind Speed (h=10 m) : 2.00 m/s Avg Wind Speed (h=H-eff) : 1.95 m/s Stability Class : D Respirable Dep. Vel. : 0.30 cm/s Non-respirable Dep. Vel. : 8.00 cm/s Receptor Height : 1.5 m Inversion Layer Height : None Sample Time : 10.000 min Breathing Rate : 3.33E-04 m3/sec Distance Coordinates : All distances are on the Plume Centerline Maximum Dose Distance : 0.091 km Maximum TED : 9.67E-10 Sv Inner Contour Dose : 1.00E-10 Sv Middle Contour Dose : 1.00E-11 Sv Outer Contour Dose : 1.00E-12 Sv Exceeds Inner Dose Out To : 0.58 km Exceeds Middle Dose Out To : 2.4 km Exceeds Outer Dose Out To : 12 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 (ResuspensionFactor : 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 T E D RESPIRABLE TIME-

INTEGRATED AIR CONCENTRATION

GROUND SURFACE

DEPOSITION

GROUND SHINE DOSE

RATE

ARRIVAL TIME

(km) (Sv) (Bq-sec)/m3 (kBq/m2) (Sv/hr) (hour:min) 0.100 9.6E-10 4.5E+02 1.3E-03 2.6E-12 <00:01 0.200 5.2E-10 2.4E+02 7.3E-04 1.5E-12 00:01 0.300 2.9E-10 1.4E+02 4.1E-04 8.1E-13 00:02 0.400 1.9E-10 8.6E+01 2.6E-04 5.1E-13 00:03 0.500 1.3E-10 6.0E+01 1.8E-04 3.6E-13 00:04 1.000 4.1E-11 1.9E+01 5.7E-05 1.1E-13 00:08 2.000 1.3E-11 6.3E+00 1.9E-05 3.7E-14 00:17









16.0 APPENDIX B – SAFRAN DOSE ASSESSMENT