gnssn.iaea.org Documents... · Web viewThe Kingdom of Morocco is a non-nuclear country situated in the Maghreb region of North Africa. Geographically, Morocco is characterized by

Draft Safety Case for the Management of Disused Sealed Radioactive

Sources in the Kingdom of Morocco

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 2 of 86DRAFT SAFETY CASE FOR THE MANAGEMENT OF DISUSED SEALED RADIOACTIVE SOURCES IN MOROCCO

1.0 CONTENTS

1.0 CONTENTS................................................................................................................2

2.0 INTRODUCTION........................................................................................................5

3.0 CONTEXT OF THE SAFETY CASE...........................................................................6

3.1 PURPOSE OF THE SAFETY CASE...........................................................................6

3.2 SCOPE OF THE SAFETY CASE................................................................................9

3.3 GRADED APPROACH...............................................................................................9

3.4 STRATEGY FOR SAFETY.......................................................................................10

3.5 DEMONSTRATION OF SAFETY..............................................................................11

4.0 SITE, FACILITY AND PROCESS DESCRIPTION....................................................14

4.1 SITE DESCRIPTION................................................................................................15

4.2 FACILITY DESCRIPTION.........................................................................................22

4.3 DSRS INVENTORY..................................................................................................34

5.0 SAFETY ASSESSMENT...........................................................................................36

5.1 SAFETY ASSESSMENT CONTEXT.........................................................................36

5.2 SAFETY ASSESSMENT ENDPOINTS.....................................................................39

5.3 DEVELOPMENT OF SCENARIOS...........................................................................42

5.4 DATA USED AND ASSUMPTIONS MADE FOR THE SAFETY ASSESSMENT.......44

6.0 SAFETY ASSESSMENT...........................................................................................45

6.1 BASIC ENGINEERING ANALYSES..........................................................................45

6.2 QUANTITATIVE DETERMINISTIC ASSESSMENT OF WORKER DOSE................48

6.3 QUANTITATIVE DETERMINISTIC ASSESSMENT OF WORKER AND PUBLIC DOSE FORANTICIPATED OPERATIONAL OCCURRENCE SCENARIOS:.......................53

6.4 DETERMINISTIC ASSESSMENT OF WORKER AND PUBLIC DOSE FOR OTHER ACCIDENT SCENARIOS.........................................................................................53

6.5 OPTIMIZATION OF PROTECTION: ASSESSMENT................................................55

6.6 NON-RADIOLOGICAL HAZARD ASSESSMENT.....................................................58

6.7 ASSESSMENT OF THE IMPLEMENTED WASTE MANAGEMENT PRACTICE.....58

6.8 MANAGEMENT SYSTEM ASSESSMENT...............................................................59

6.9 ASSESSMENT OF UNCERTAINTIES......................................................................60

7.0 IDENTIFICATION OF FACILITY SPECIFIC LIMITS AND CONDITIONS.................61

8.0 INTEGRATION OF SAFETY ARGUMENTS.............................................................62




8.1 FACILITY DESIGN AND ENGINEERING.................................................................62

8.2 FACILITY OPERATION............................................................................................62

8.3 OPTIMIZATION OF PROTECTION..........................................................................62

8.4 WASTE MANAGEMENT PRACTISE........................................................................63

8.5 INTEGRATED MANAGEMENT SYSTEM.................................................................63

8.6 UNCERTAINTIES.....................................................................................................63

9.0 COMPARISON WITH SAFETY CRITERIA AND CONCLUSIONS............................63

10.0 INTERACTING PROCESSES..................................................................................64

10.1 LEGISLATION AND REGULATIONS RELATING TO THE MANAGEMENT OF DSRS IN MOROCCO..........................................................................................................64

10.2 REGULATORY BODY..............................................................................................65

11.0 REFERENCES.........................................................................................................67

12.0 ASPECTS REQUIRING FURTHER CLARIFICATION AND ACTION PLAN.............71

13.0 APPENDIXES...........................................................................................................74

APPENDIX A: DATA USED AND ASSUMPTIONS MADE FOR THE SAFETY ASSESSMENT....74

APPENDIX B: HOT SPOT DOSE CALCULATION.......................................................................76

14.0 ANNEXES.................................................................................................................79

ANNEX A: STORAGE BUILDING DESIGN CALCULATIONS: RADIATION PROTECTION BARRIER.................................................................................................................. 79

ANNEX B: WASTE MANAGEMENT BUILDING MAINTENANCE SCHEDULE............................81




ABBREVIATIONSCENM –Nuclear Studies Centre of MâamoraCNRP – National Centre of Radiation ProtectionCNESTEN–National Centre of Nuclear Energy, Sciences and Techniques DSRS – Disused Sealed Radioactive SourcesCSF – Central Storage FacilityRPO – Radiation Protection OfficerIAEA – International Atomic Energy Agency




2.0 INTRODUCTION

The Kingdom of Morocco is a non-nuclear country situated in the Maghreb region of North Africa. Geographically, Morocco is characterized by a rugged mountainous interior and large portions of desert. It is one of only three countries (with Spain and France) to have both Atlantic and Mediterranean coastlines.

Morocco has a 2 Megawatts TRIGA research reactor located at the Centre d’Etudes Nucléaires de la Maâmora (CENM), which is situated 22 km North-East of Rabat-Salé, and is the only existing nuclear study centre in Morocco which was authorized for exploitation in May 2009 by the Ministry of Energy and Mines. Facilities to treat radioactive waste generated at the national level are located in the research reactor site. Until now all radioactive wastes and disused sealed radioactive sources (DSRS) generated in Morocco arise from the use of radioactive materials in the form of sealed and unsealed sources in industry, medical, education and research fields.

1500 sealed radioactive sources are currently used in 150 establishments, the main radio-elements being Cs-137 and Co-60. In Morocco, the majority of sealed radioactive sources are used in the industry as radiometric gauges, i.e. level gauges, density gauges, thickness gauges and moisture gauges and for radiography, in the medical field and in the mining and oil industry for logging. Morocco does not manufacture any sealed radioactive sources.

The main waste management facility in the country is located at CENM. The facility is operated by National Centre of Nuclear Energy, Sciences and Techniques (CNESTEN) who was established and assigned with the responsibility of managing radioactive waste including DSRS by Law No. 12-02 of 2005. All the waste treatment, conditioning, handling, storing and transport operations are carried out by CNESTEN in the radioactive waste management facilities at CENM (except the storage of spent fuel which takes place in research reactor building).

CNESTEN is also responsible for the collection, transport, receiving and conditioning of DSRS. Generators of radioactive waste are responsible for waste generation control (waste minimization), pre-treatment, treatment, conditioning and the characterization of waste according to the technical specification established by the central operating organization (CNESTEN). High activity (e.g., Category 1 and 2) DSRS 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 • Transferring the disused source to the central waste management facility (CNESTEN)

Regulatory function for non-nuclear facilities is conducted by the National Centre for Radiation Protection (CNRP), which is under the Ministry of Health by order of decree.

Morocco does not have any authorized disposal facilities planned or in use. All conditioned DSRS and radioactive waste are stored in the long term storage building of CNESTEN for 50 yr storage period. Morocco currently has no nuclear power programme but has plans to build a nuclear power plant at Sidi Boulbra and a desalination plant.

2.1 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.

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. A National Commission of Nuclear Safety (NCNS) was 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 International Commission of Radiation




Protection (ICRP) recommendations and the Basic Safety Standards of the IAEA. This decree also mentions that CNRP is the Regulatory Body dealing with non-nuclear facilities.

The Moroccan government provides funds to the CNESTEN to carry out all operations related to the radioactive waste management and to ensure the safety of the radioactive waste management facility during its operating life-time. The generator of radioactive waste pays for the collection, the treatment and the storage of waste. The Moroccan regulatory framework does not cover the financial aspects and responsibility for providing funds for decommissioning. Since the CENM is a public research centre, the government remains responsible to cover any cost of its future decommissioning including the decommissioning of the research reactor on CENM.

A new law, Law (No.142-12), related to nuclear and radiation safety, nuclear security and safeguards was promulgated and published in Moroccan official bulletin on 11 September 2014. This law will be in force after one year after promulgation and publication i.e. September 2015. The aim of this law is to upgrade the national legislative and regulatory framework to ensure consistency with the international approach to nuclear and radiation safety, nuclear security and safeguards and to establish an independent single regulatory authority (independent from departments having the role of promoters or users of nuclear technology and energy). The new law, Law (No.142-12), will after entering into force replace Law (No 005-71) of 12 October 1971 as well as decree number 2-97-30 of 28 October 1997.

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.

2.2 Regulatory Body

Morocco currently has two regulatory bodies:1. The National Centre of Radiation Protection (CNRP) which is under the Ministry of

health and covers all activities taking place in non-nuclear installations.2. The Ministry of Energy, Mines, Water and Environment (MEMEE) which covers activities

occurring in nuclear installations.




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.

The recently promulgated law (law 142-12) provides for the establishment of an independent and unified Regulatory Authority, the Nuclear Safety and Security Agency. The new Agency will take over the regulatory functions currently assigned respectively to the CNRP under the Ministry of Health and to the Ministry of Energy, Mines, Water and Environment (MEMEE).

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.

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 a new law that inter alia include radioactive waste management regulations that address a number of waste management aspects such as waste classification and transportation of radioactive waste including DSRS etc. Morocco takes into consideration the IAEA safety standards and other publications (e.g., TECDOCs) related to the radioactive waste management published by the IAEA.

Generators of radioactive waste are responsible for waste generation control (waste minimization), pre-treatment, treatment, conditioning and the characterization of waste according to the technical specification established by the central operating organization (CNESTEN). High activity (e.g., Category 1 and 2) DSRS 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 • 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 cover the cost of its management and send it to the CNESTEN. Orphan sources for which the owner can’t be identified are transferred to CNESTEN for its management. A special case is related to the orphan sources which are detected in metallic scrap, in which 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 published clearance criteria or levels but adopts the concept of clearance and the use of the applicable derived clearance levels as published by the IAEA [29].




3.0 CONTEXT OF THE SAFETY CASE

This section provides the context for the safety case, including a description of the legal and regulatory framework for the management of DSRS in the country, as well as the national policy and strategy.

3.1 Purpose of the Safety Case

The purpose of this safety case is to apply for authorization of DSRS dismantling and conditioning within the site of CENM. The operator of CENM, CNESTEN is authorized to conduct predisposal management and long-term (50 year) storage of radioactive waste and DSRS (authorization for construction was issued under Decree 2-99-111 on 26 Feb 1999, and authorization for operation was issued under Regulation 2004-08 on 19 Jan 2009).

CENM is authorized for predisposal management and long-term (50 year) storage of radioactive waste and DSRS (authorization for construction was issued under Decree 2-99-111 on 26 Feb 1999, and authorization for operation was issued under Regulation 2004-08 on19 Jan 2009). However, the existing authorization does not explicitly address the safety of dismantling and conditioning of DSRS (although in 2014 CNESTEN submitted an annual safety report which addressed the potential impact on safety of the dismantling of DSRS [27]).

Therefore, the purpose of this version of the safety case is to determine whether the existing facility is acceptable from a safety point of view and to determine corrective actions required to improve safety. A principal application of the Safety Case will be in the license application and approval process. This safety case will be submitted by CNESTEN as part of the application for authorization of DSRS dismantling and conditioning within the site of CENM.

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

Application to modify the facility or activity Demonstration of the safety of dismantling of dismantling, conditioning and storage

of Category 3 to 5 DSRS at the CENM Treatment Facility, located within the CENM Waste Management Facilities. The categorization of DSRS is based on the categorization system of the IAEA Safety Guide RS-G-1.9 [28].




Demonstration of the safety of collection at users facilities, transport of DSRS to the CENM Waste Management facility, receiving and characterization of the DSRS, clearance and discharge activities (e.g., iridium sources [wire] from hospitals), temporary storage, conditioning and longer term storage.

Optimization of the respective waste management activities described above. Assessment of the maximum inventory of DSRS for each campaign (the ‘radiological

capacity’ of the facility). Definition of Limits, Controls and Conditions that will be applicable to the facilities

and the respective activities described above. Assessment of the Management systems to ensure the safety of the respective

waste management activities described above. Input to the improvement of existing Radiation Protection (RP) programmes and

procedures.This draft safety case includes the information and identification of gaps in information required to demonstrate and ensure the safety of dismantling of DSRS as performed by CNESTEN. This includes amongst others the context for the evaluation of the safety case; a description of the legislation and regulations pertaining to the safe management of DSRS in Morocco; a description of the regulatory function as well as the appointed waste operator; a description of the site, facility and activity description, and inventory of DSRS; a description of the management systems and procedures necessary to ensure compliance to set safety criteria and to sustain an acceptable level of safety; an assessment of the results; identification of limiting conditions; and aspects that require further clarification.

This Safety Case will take into consideration the International Atomic Energy Agency (IAEA) General Safety Requirement (GSR) Part 5 on the Predisposal Management of Radioactive Waste [1] and the IAEA Safety Guide GSG-3 on the Safety Case and Safety Assessment for the Predisposal Management of Radioactive Waste [2]. Safety criteria will be taken from the Moroccan regulatory framework and GSR Part 3 on Radiation Protection and Safety of Radiation Sources: International Basic Safety Standards [4].




3.2 Scope of the Safety Case

The scope of the safety case will be focused on the dismantling, conditioning and storage of DSRS in the centralized radioactive waste management facility at CENM and the operational aspects of the facilities, which are defined below: Collection and transport of DSRS to the facility; Receiving, identification, characterization and handling of DSRS at the facility; Temporary storage of the DSRS that have not yet been dismantled; Dismantling of DSRS; Conditioning of the DSRS for long term storage; Handling and placement of the dismantled and conditioned DSRS into long term

storage.

This version of the safety case will 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 or selection of

possible sites. The construction and commissioning of such facilities. Decommissioning or decommissioning planning of facilities.

This Safety Case will consider interfaces with the existing Final Safety Analysis Report for the CNESTEN Radioactive Waste Facility [5].

3.3 Graded Approach

A graded approach is applied for defining the extent and depth of this safety case by the use of qualitative assessment of hazards and deterministic analysis of doses to potential receptors (e.g., workers and public). This takes into consideration the safety significance, complexity and maturity of the facility and the dismantling, conditioning and storage activities.

As stated in Section 3.0, CNESTEN is licensed and authorized for predisposal management and long-term (50 year) storage of radioactive waste and DSRS (including storage of Category 1 and 2 DSRS), the safety of which are addressed in the FSAR for the CENM waste management facility [5]. In 2013 CNESTEN submitted an annual safety




report which assessed the potential impact on safety of the dismantling of DSRS [27] and concluded that there was no impact on the safety of the facility.

The radiological hazard when undertaking the various management activities involving DSRS can be regarded as low. The scope of this Safety Case is limited to the dismantling, conditioning and storage of Category 3-5 DSRS at a facility that is licensed and authorized for conditioning and storage of DSRS (without dismantling). While Category 1 and 2 DSRS are stored in the storage building, they are stored in their working shields (they are not dismantled at the facility).

The simplicity of the activities involving the management of DSRS using simple but proven with techniques an inherent high level of passive safety and very limited reliance on active protection systems. Most of the activities involving DSRS entail handling of the DSRS inside robust working shields which limits external exposure potential. DSRS will only be handled inside their working shields. Any sources found to be leaking or contaminated will not be dismantled but will instead be conditioned without dismantling. The only time that bare DSRS will be handled during normal operations is during source conditioning operations using lead glass and lead brick for shielding (there is no hot cell at the facility). In such instances the risk is reduced by performing the work in accordance with specific work procedures and under work permit systems where there are 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.

3.4 Strategy for Safety

This section describes the strategy for safety, including the approach that was taken in the facility design and all the respective DSRS management activities to comply with the regulatory requirements and to ensure that good engineering practice has been adopted and that safety and protection are optimized.

In view of the scope of the safety case as defined in 3.2 the following strategies for demonstrating safety of the management of DSRS are adopted:

Defense in Depth – In this instance care is taken to ensure that multiple safety layers are available. This principle is considered to ensure that no important safety argument is based on a single level of protection.




Passive safety – the use of passive systems is regarded as contributing to the safety: Shielding – Shielding is 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 is considered.

Selection of implemented waste management practice – Approach to waste management with regards to the following is regarded as contributing to safety:

o 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.

o 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.

o 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.

o Interim storage of DSRS only occurs inside proper containment such as the original working shields or another type of suitable shield and containment.

o Conditioned DSRS are stored in a dedicated storage area with passive safety features and adequate access control.

o Separation of processes involved DSRS operations and radioactive waste operations.

o Sources are stored according to their radionuclide content (e.g., sources with long-lived radionuclides are stored separate from those with short-lived radionuclides). In this way, periodic inspection and continuous radiological monitoring of the storage building and of the waste drums/packages is facilitated.

3.5 Demonstration of Safety

The section describes the approach to demonstration of safety, e.g. the safety objectives and safety principles that must be applied and the regulatory requirements that must be met. Taking cognizance of the scope of the safety case and the application of the graded




approach as described above, the safety of the waste management facilities will be evaluated and demonstrated as described hereafter.

3.5.1 Elements related to the structure (engineering)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 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) Principles of safety and security 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 and occupancy of 400 h per year i.e. 2.5 µSv/h.

- The lighting system provides adequate lighting and permits the performance of operations in a safe manner.

- Physical delineation of areas designed for storage and for the main waste (DSRS) management operations are isolated, which ensures 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.

- 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.




The structure resists the maximum load of the sources that are intended to be stored.

- A storage system design that minimize worker’s exposure to sources of greater or unknown activity or that may have not been conditioned.

- Engineering systems ensuring safety during operational occurrences and potential accidents that could occur during dismantling and conditioning: Floor and wall finish allow easy decontamination. The segregation of the different areas that limits the potential

dispersion of any contamination. Liquid waste collection system inside the facility that collects liquids

resulting from surface decontamination activities thus preventing its release to the environment. The system includes a retention tank that permits environmental monitoring before releasing to the environment.

The facility has is provided with fire detection and firefighting equipment.

- The facility design makes provision for physical security features commensurate with the anticipated security threat [30]. 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 and motion sensor alarms. Vehicle access points to the buildings. A separate personnel door is

provided to segregate personnel from vehicle movements. Zones 2 (green zone) and 3 (yellow zone) of the buildings are

windowless so as to improve shielding and security performance.

3.5.2 Demonstration of Safety (assessment)Quantitative and qualitative assessment will be performed to assess the impact of the dismantling, conditioning and storage activities. 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;

For all other credible occurrences: a quantitative and qualitative assessment of the impact of other occurrences and the listing of specific preventative and mitigating measures.

The results from the quantitative and qualitative assessment as defined above will also be compared to the proposed target and objectives set for the optimization of protection.

Qualitative assessments will be performed of the following: Non-radiological hazards of the facilities and the listing of specific control

measures; Current waste management practice and approach to waste management as a

contributing factor to safety; Availability and level of implementation of an integrated management system to

ensure a sustained level of safety during the operational phase of the facilities. This assessment will focus on RP, work procedures, Quality Assurance (QA) aspects and processes for the management of operating limits and conditions.

3.5.3 Management of uncertainties 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.

4.0 SITE, FACILITY AND PROCESS DESCRIPTIONThis section provides description of the waste management facility and activities included within the scope of the safety case, including information and knowledge about the facility and the activities to be carried out, providing the basis on which safety assessment is carried out.

In addition to guidance in GSG-3, this safety case must also considers requirements contained in Article 7 of decree No. 2-94-666 of 7 December 1994, which states that authorization for the construction are required and that such an application should be based on a preliminary safety analyses report that covers the following:




A site plan showing the perimeter of the site to house the proposed nuclear facility; Calculations and assessments showing that it was taking into account the natural

events and characteristics of the site, accompanied by an impact study to demonstrate that the proposed installation can be built and operated without risk to personnel operator, the public and the environment;

The evaluation of the ability of the site to receive the system must take into account the following aspects:

o Effects of external events due to natural causes or man-made action and that would occur in the site area;

o Characteristic of the site or its environment, which may influence the transfer to living beings of radioactive material released;

o Density and distribution of the population, with a preliminary assessment of radiological effects on the population

The information on the basic design of the proposed facility and the safety systems, radiation protection and the results of the analysis safety may affect the structure, systems and components having related to nuclear safety;

Information on the control, handling and storage of radioactive waste from the nuclear facility;

Information on Quality Assurance programs of the applicant and its service providers and property;

The internal emergency plan for the installation; Arrangements for physical protection of the facility; The provisions on liability for nuclear damages Information on the operating organization and its staff qualifications

4.1 Site Description

4.1.1 General Description of the SiteThe land where the CENM is situated is attached to Haddada community and is located in Maâmora forest, 22 km North-East of Rabat-Salé, 15 km south-south-West Kénitra and 8 km west of the Atlantic coast. The largest cities in the region are Salé, Rabat and Kénitra; they are located in a radius around 20 km from the site, with populations of 339,000, 700,000 and 246,000 inhabitants, respectively. The nearest town to the site is Sidi Bouknadelis, located 8 km from the site.




Figure 1. Location of CENM

The site itself is mainly woody and extends over a rectangle of 550 x 450 meters that is an area of 24.75 ha. It has a flat terrain with a slope of about 2% of the southeast to the northwest, representing a difference of level of 10 meters along this axis, and does not prevent the drainage of rainwater. The site is located such that the connection to regional infrastructure such as road network, water and power supply is feasible.

The perimeter of the site is enclosed by a metal fence. This fence is erected on the CENM property line.A 50 m buffer zone, consisting of a 20 m cleared zone and a 30 m zone with isolated trees ismaintained between the fence and the buildings and facilities of the CENM site. This zone is mainly maintained for physical protection and fire prevention purposes. Figure 2 shows the buffer zone at the site.




Figure 2. Fire and Security Buffer Zone at CENM site

A Preliminary Safety Analyses Report [5] was submitted to the Regulatory Body for the approval to construct the waste management facilities on the Mâamora site. A summary of the site description that was assessed in [5] is covered below.

8.1.2 Demography Demographic data published in the report [5] is based on census studies performed during 1982 and 1989. Figure 3 illustrates the demographic distribution:




Figure 3. Demographic Distribution [5]

The main demographical findings as published in report [5] are the following: Distribution of population in a radius 20 km around the site:

o 0 - 1 Km : 0 Inhabitantso 1 - 2 km : 177 Inhabitants with 77 and 100 in the north-west and south-east

sectors respectively.o 2 – 5 km : 2107 Inhabitants all originally from rural areas and included in the

north-north-east and west sectors.o 5 – 20 km : 700898 Inhabitants.

Urban areas: Most of the inhabitants of the urban areas are in the 2-20 km zone and are located in the west-south-west, southwest, northeast and north-northeast sectors. The major population centers in the region are: o Salé: 339384 Inhabitants. It is located about 20 km south-west of the site. The

center of this town being located beyond the 20 km circle and only a fraction of




the inhabitants are considered. (Note that Rabat is located outside the circle of 20 km radius.)

o Kénitra: 246000 Inhabitants. It is located about 15 km north-northeast of the site. Rural areas: The total number of rural inhabitants is about 100000. The total

population in rural areas, in the radii (2-5 km), (5-10 km) and (10-20 km), is approximately 2107, 21500 and 89500, respectively.

Population growth: According to projections, the total population in rural areas in radii (2-5 km), (5-10 km) and (10-20 km), will be 3200, 32900 and 136700 respectively in the year 2037. This shows that the rural population is growing moderately by about 1.4% per year. Similarly, the transition from 589,960 inhabitants in 1989 to 1,506,000 people in 2037 for urban areas implies an average increase of about 4% per year.

8.1.3 Meteorology

Local meteorological conditions: The site is located in an area of high permanent barometric pressure, standing on the tropical Atlantic Ocean, called "Azores high". Isobaric individuals linked to it have a decisive influence. This is the area cis-Atlantic or Atlantic, very open to the west and much softened by moisture from the Atlantic Ocean (CNESTN 1996) [5].

Wind conditions: Data was collected from the weather station (10 m height) on the sitein 1992 and 1994,consisting of approximately, 17,000 hourly measurements.

The evaluation of the data shows that the most dominant winds are northwest (11%) and south-south-east(9%) with generally low speed winds. The annual average wind speed is of the order of 1.7 m/s, approximately, (6 km/h).

The wind rose is illustrated in Figure 4.




Figure 4. Wind Rose Data

Average monthly maximums and minimums are respectively (5.3 m / s) (0.1 m / s). The maximum wind speed recorded during these two years is (7.3 m / s). The table below shows the average speed (km / h) of seasonal winds, measured at a height of 20 m. these values are compared in the table, to those given by the Directorate of National Meteorology, related to the data of weather stations in Rabat and Kénitra:

Average speed in km / hWinter Spring Summer Autumn

KénitraRabatCEN (high. 10

m)CEN (high. 20

m)

7.99.45.59

11.912.66.89 .7

11.510.86.68.2

910.15.58.2

Temperatures

Data of national meteorology daily bulletins, for the nearest stations to CEN, namely Kénitra and Rabat-Salé stations, show that annual average temperatures, rarely




exceed (18 °C) and the maximum and minimum averages, are 22.4 °C and 12 °C respectively. By evaluation of the site meteorological data of only two years (1992, 1994)on the recorded temperatures, no frost day (at negative temperature) was detected. The table below shows the average temperatures at a height of 10 m for the various seasons:

Averagetemperature in height 10 m

Winter Spring Summer AutumnCEN 13.97 18.57 22.49 16.38

Humidity

The average annual humidity on the Maâmora is 75% and the maximum can reach 100%Humidity for each season is given in the following table that compares the values detected by the site station with those measured in the nearby stations:

Relative average of humidity (%)Winter Spring Summer Autumn

KénitraRabatCEN

77.971.578.07

65.46471.89

63 .764.571.93

626279.66

Cloud cover

The average annual number of hours of sunshine is 4439 hours of which 20% are measured during the winter time. The days of no insolation or cloud cover are, on average 3 days per year. The days of continuous cloud cover or insolation are, on average 56 days per year.

Evaporation

The evaporation rate was measured for the Maâmora site measuring the evaporation from an evaporation pan placed on the site. The measured annual evaporation was found to be 1142 mm of which 265 mm occurred during winter and spring and 877 mm occurred during summer and autumn.

Precipitation

The amounts of rain are measured with the rain gauge on the site. The annual amount of rain that is recorded is about 530 mm. The seasonal rainfall distribution is reflected in the table below:

Rainfall (mm) per season




Winter Spring Summer Autumn

CEN 281.7 78.82 52.46 117.6

By evaluating the precipitation and evaporation data of the Maâmorasite, it is clear that during winter the rainfall is high while the evaporation rate is low and during summer it is vice versa.

8.1.4 Site Geology and GeohydrologyReport [5] covers site geology and geohydrology and related information.The Maâmora site is located at the boundary of two major structural assemblies: the rigid primary Meseta, dipping south to north regularly with a 3 degree slope and the Rif area whose formations were advanced on previous north-east to south-west in thrust sheets. This area, attributed to the Rharb basin is affected by subsidence continued since the Middle Vindobonian, time corresponding to the formation of the layers. Still active today, the subsidence is underlined by a depression 3m of sewers of the Romanian city of Banasa located in the heart of the flat Rharb, a hundred kilometers east of the site.

The water table Maâmora is based on, the blue marl Mio-Pliocene, and flows into the sands and marine Pliocene sandstone, and in sandy or pebbly gréso-Villafranchien dune formations or continental. It rises in the south to the limit of my Pliocene outcrops, and flows range between the northwest and northeast. The morphology of this groundwater basin corresponds to a large bowl, 4/5 of which are at altitudes below 200 m, while the borders, have, very mild relief, the water table Maâmora covers an area 7500 km2. Its boundaries are formed by the Atlantic Ocean to the west; the Outita hills to the east and the landforms between Khémisset and Tifelt the south. The Maâmora site is part of the coastal zone which extends between Salé and Kénitra, covering an area of 390 km2 and flows north.

The average depth of the water table does not exceed 30m near the site, but in some places it exceeds 70 m. aquifer thickness is of the order of ten meters around the site; it hardly exceeds 30 m.Regarding the Hydrogeological structure of Maâmora, the water table, the Maâmora, based on blue sandy clays Mio-Pliocene and flows into the sands and marine sandstones of the Pliocene, and in sandy or pebbly gréso-Villafranchien dune formations or continental. It rises in the south to the limit, outcrops, Mio-Pliocene and flows range between north-west and south-east.




Three preferred flow directions should be noted:• The drainage area to the sea • The drainage area to the Gharb• The drainage area to the river Beht

8.1.5 Site Seismology Report [5] covers site seismology and related information aswell as the methodology for the derivation of a reference event.

The site of the Maâmora is located in an area of medium seismicity very close to the Mediterranean pit characterized by great tectonic instability. Consequently, all the civil engineering works of the Centre were designed, calculated and verified, to the earthquake: Reference intensity, MSK unit level VIII, leading to a safety level of earthquake IX MSK. This was determined using methods from the recommendations of the French Association for Earthquake Engineering (AFPS 1990) and the Unified Technical Document (DTU) "seismic Rules" (1969) (revised in 1982).

8.1.6 Aircraft Crash Probability

Report [4] covers the assessment of the aircraft crash probability and risk in detail. In summary, the magnitudes of the probability of aircraft crash per year per square meter for the different groups of aviation are given in the following table:

Maâmora Site Civil aviation General aviation Military aviationAnnualprobability per unit of surface (impact/an/m2)

6.87 x10-12 5.53 x10-11 1.88 x10-11

The annual probability of an aircraft crash into the radioactive Waste management Facility with approximate overall dimensions of 20 m x 25 m assuming the maximum crash probability of 5.53x10-11 is 1.7x10-8. Based on this low probability, no provision has been made in the design and construction of the waste management facilities to mitigate against air craft crash risks.

4.2 Facility Description

The waste treatment and storage facility at CENM 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 ofthree levels (4.00 m below ground level, ground level, and 3.50 meters above ground level). 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.

Long term storage facilityconsisting 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 walls is 0.40 m.

Figure 5: Waste Management Facilities at CENM

Figure 6: Vault in Storage Building

Waste Treatment Facility

Waste Storage Facility




4.2.1 Facility Design and Construction

This section provides basic information regarding design considerations, applied design and construction codes and standards. Applicable Regulations, Codes and Standards for Construction of CENM are provided in (Technicatome 1999) [11]. Design review and construction reports, and certificates of conformance are provided in (Technicatome) [26].

4.2.1.1 EarthquakeThe maximum considered earthquake corresponds to an earthquake of intensity VIII internationally.For the waste module, the reinforced concrete part of the treatment building DT is calculated in elastic mode, under the action of the maximum earthquake.Therefore, the confinement provided by the concrete structures of DT building will be preserved under the action of the earthquake. For the part of metal structures, mechanical strength will be retained.

The storage building waste DE was the object of simplified calculations at maximum earthquake.Under normal conditions, this building has no role containment of radioactive materials. In addition to its function as a radiation protection staff, it protects waste drums against the weather.Under the action of the maximum earthquake, the mechanical strength of the metal structure will be preserved.

The concrete walls ensuring the biological protection, may present some cracks, the extension will be limited and that should not affect their protective role. For this, the storage building is partitioned into four equally sized cells separated by concrete walls.The sealing under the invert of the barrels of storage area is secured, in normal conditions, by a diaphragm arranged for this purpose. This function will be preserved under the action of the maximum earthquake.

4.2.1.2 FireThe premises, within which there are objects with a high calorific value, are equipped with fire detectors, usually ionization smoke detector.The first response capacity is manually implemented. They consist of portable extinguishers, wall installed in




sufficient numbers in the building. Any fire alarm will result in the immediate intervention of firefighters installed on the site.

4.2.1.3 FloodingMitigative measures against internal flooding areas follows:• waterproof rafts• pits for receiving the effluent tank,• wells with the possibility of recovery pump




4.2.2 Main Safety and Security Related Design Features

4.2.2.1 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.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. The building is designed to withstand a floor load 20KN/m2 for the basement, and10KN/m2for the ground floor (zone 2 and 3)and 2.5 KN/m2for the non-controlled zone. The storage building is to withstand a floor load of 5KN/m2

(Technicatome 1999, Design Floor Load, Document TA-103377) [6]. The slab is sufficiently thick around the building perimeter to support the walls and

locally around all internal stanchions; Rain water is prevented from entering the buildings by surface contouring and

drainage channels around the buildings. Resistance to water penetration from the ground is provided by a polyethylene

damp proof membrane to the underside of the slab (CNESTN 2005) [7]. The interior construction of the building is such that the risk of any liquids being

released to the environment is minimized. 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 10m3. The floor is 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. The sump pit is provided with monitors for the detection of liquids and of gamma radiation which sends an alarm signal to the control room of the radioactive waste management unit and to the radiation protection unit (CNESTN 2005) [7].

The floor slab has a steel floated finish with an epoxy paint coating to provide a hard wearing and decontaminable surface (CNESTN 2005) [7].

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 (CNESTN 2005) [7].




4.2.2.2 Shielding

The waste treatment and storage building structures provide efficient shielding from radiation to limit exposures outside the building to less 2.5 µSv/h. Design inputs and outputs demonstrating sufficient shielding capacity are provided in [7] and [8].

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 are shielded by other packages, internal building structures or by concrete blocks.

4.2.2.3 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.

Zones 2 and 3 are windowless so as to improve its shielding and security performance.

4.2.2.4 Waste Treatment Facility Layout

The layout of the basement, ground floor and first floor of the waste treatment facility are illustrated in Figures 6a, 6b, and 6c below (Figures4.3.1, 4.3.2, and4.3.3 of Ref [7]).




Fig 6a. Waste Treatment Facility Basement

Fig 6a. Waste Treatment Facility Ground Floor




Fig 6c. Waste Treatment Facility First Floor

4.2.2.5 Waste Storage Facility Layout

The layout of the waste treatment facility is illustrated in Figure 7 below (Figure 4.3.3 of Ref [7]).




Fig 7. Waste Storage Facility

4.2.2.6 Fire protectionFire 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 [10]. Fire detection and firefighting equipment has been installed. Such equipment are tested and maintained; they are visually inspected weekly and are replaced [13]. High quality electrical equipment complying with national quality standards is installed in both buildings. The site is maintained clear of vegetation and combustible materials are not stored on the site. Fire detection equipment is installed, fighting equipment is provided and strict compliance is maintained with national and local fire regulations.

4.2.2.7 VentilationBelow is a summary of the ventilation system (FSAR) [7] and system performance levels that are relevant to the DSRS activities. Inspection and maintenance of the ventilation system are performed in accordance with CENM procedures [13]. Annex B provides the maintenance schedule for the waste management facility. [34]

The functions of ventilation and air conditioning system of the installation are the following:

the continued depression of some locals




limiting the possible contamination inside treatment building by renewal of air filtering the air of the installation before discharge Temperature control (heating and air conditioning)

The ventilation system consists of three distinct and separate networks ensuring respectively:

Blowing the fresh air, Transfer of air between certain locals, Extraction of the air and the discharge to the outside.

The air conditioning system comprises: Heat pump AIR / WATER ensuring the conditioning of blowing air in the treatment

building A set of reversible heat pumps with separate condensing unit (Split Systems)

reversible, providing air conditioning to certain locals in treatment building

The ventilation system was designed assuming the following environmental parameters:(i) outside temperatures

Summer: 32 ° C, relative humidity 48% Winter: 5 ° C, relative humidity 95%

(ii) internal temperatures (for treated locals) Summer: 25 ° C, relative humidity 50% Winter: 20 ° C

(iii) renewal rate of air in locals (Rn / h) Zone 1: 1 Rn / h blowing and extraction Zone 2: 2 Rn / h blowing 2.5 Rn / h extraction Zone 3: 2.5 Rn / h blowing 3Rn / h extraction Gloves Box: 10.5 Rn / h extraction.

The storage building is provided with natural ventilation; outlets are located high on the building walls and covered with grids to prevent the access of animals, birds and insects.

4.2.2.8 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 (Technicatome 1999) [11]. Good levels of lighting are provided throughout the treatment and storage facilities and high quality, long life components are used to reduce maintenance needs.

The system called "Electricity High Voltage/Low Voltage (HV/LV)". The electrical system HV/LV provides power to the entire site, and supplies LV (220 V) electric power to all of the buildings within CENM. Its functions are:

Transport, process and distribute through an internal network, electricity from the national Moroccan network from the delivery point to the different buildings;

In case of failure of the national network, ensure the emergency power of the site to allow operation of facilities, in terms of reliability, safety and best availability.

Design Parameters:In general, the HV/LV system meets French and Moroccan standards for electrical installations. These standards are provided in Ref [11]. Safety functions of the system are summarized below [33].

Nuclear hazard: Not applicable Seismic hazard: Not applicable Fire hazard: The substation HV/LV is classified as "fire class." As such, it is

equipped with a fire detection system. Fire protection is manual, with the network fire extinguishers and mobile.

Lightning hazard: The specific risks, with surges of atmospheric origin are included in buildings and modules, installation of lightning rods. The detailed study will determine whether or not to implement the installation of head arresters.

Explosion: No local or work area, being classified as hazardous, the installation does not require the implementation of increased safety equipment or anti explosion equipment.

External power failure: In case of loss of public power distribution network, the site diesel generator starts automatically. After stabilization of voltage and frequency, the electrical power is restored to the reactor building, the isotope




production building, and to the waste management facility (including the storage building).

Failure of the diesel generator: In the first 10 minutes after the failure of the external power supply, only electrical equipment which is powered by the UPS remains in operation. Periodic testing of the diesel generator is conducted to ensure proper operation.

4.2.2.9 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(Technicatome 1999) [11].

4.2.2.10 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 to optimize safety and protection and to ensure that no individual dose constraints or limits are exceeded(CNESTN Operating Procedures) [14 – 23].

Procedures for operational radiation protection [12] and maintenance and inspection [13] are formally documented and approved. An incident reporting system [24] and CENM Emergency Preparedness and Response plan [25] are drawn up and approved. These procedures 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.




4.2.3 Waste Treatment Facility Operation

A batch of standard Category 3 sources are collected from the Waste Storage Facility in their working shields and transferred to the Waste Treatment Facility. Individual sources are placed on a working area equipped with a shield (Shield and relevant tool description as well as shielding performance/requirements to be defined). The sources are removed from their working shields inspected, recorded and placed into a shielded storage container. After completion of the campaign, the storage container is closed and returned to the Waste Storage Facility.. 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. (Details about operational procedures to be added)

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 before it is transferred to the Waste Storage Facility. Two such campaigns are conducted per annum.

Approximately 8 hours per week is spent on general cleaning, inspection and maintenance purposes.

4.2.4 Waste Storage Facility Operation

The transported DSRS are received at the Waste Storage 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 storage location in the Storage Facility at CNESTEN. The storage location is inspected and surveyed monthly by operators and an RPO

Used and 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 Approximately 2 hours per week is spent on general cleaning, inspection and

maintenance purposes.

4.2.5 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. Reference the RPP

4.2.6 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, quality assurance (QA), operational, maintenance and inspection programmes to ensure protection and safety are optimized and that no personal dose limits or constraints are exceeded (CNESTN 2014) [9]. 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 [9].




4.3 DSRS Inventory

Morocco does not manufacture any sealed radioactive sources. 1500 sealed radioactive sources are currently used in 150 establishments, the main radio-elements being Cs-137 and Co-60. In Morocco, the majority of sealed radioactive sources are used in the industry as radiometric gauges, i.e. level gauges, density gauges, thickness gauges and moisture gauges and for radiography; in the medical field and in the mining and oil industry for logging.

The current inventory of the inventory of DSRS and sealed sources in use are reflected by the figures below: Figure 5 - DSRS Inventory at CNESTEN [30] Figure 6 - DSRS Inventory at User facilities [31] Figure 7 - Sealed Sources in use in Morocco [31] Figure 8 - Radionuclides in use in Morocco [31]

Figure 5: DSRS Inventory at CNESTEN




Figure 6: National DSRS Inventory

Figure 7: Sealed Sources in Use in Morocco




Figure 7: Sealed Sources (Radionuclides) in Use in Morocco

5.0 SAFETY ASSESSMENT

5.1 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.

5.1.1 Strategy for Safety Assessment

5.1.1.1 Basic Engineering AnalysesThe 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.

5.1.1.2 Demonstration of safety of 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; 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 storage location in the Storage Facility at CNESTEN. The storage location is inspected and surveyed monthly by operators and an RPO.

- A batch of standard Category 3 sources are collected from the Waste Storage Facility in their working shields and transferred to the Waste Treatment Facility. Individual sources are 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. After completion of the campaign, the storage container is closed and returned to the Waste Storage Facility. 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. These non-standard category sources units and equipment are typically those that are difficulty to open or remove the source. Due to the difficulty with the equipment (source still shielded inside the equipment) the equipment cannot be handled behind a shielded work area. However as soon as the source can be removed from the equipment, the source is directly moved behind the shielded work area, where the source can be inspected, identified, characterized, recorded and transferred into the 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 curebefore it is transferred to the Waste Storage Facility. 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.

• 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 spreadsheets 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.




5.1.1.3 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.

5.1.1.4 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.

5.1.1.5 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:o 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.

o 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.

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




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

5.1.1.6 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.

5.1.1.7 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.

5.2 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.

This section provides the safety objectives and safety principles that are applied and the regulatory requirements that must be met.

5.2.1 Radiation Protection

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.




To protect against the dangers of ionizing radiation, two categories of dose receptors are considered:

Workers Members of the public

The two important principles considered are: The dose limitation system, in accordance with national regulations, Optimizing the dose according to the principle of optimization, to maintain any

exposure as low as possible taking into account economic and social factors.

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

Protection of Workers

All 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 limits. Exposure of the pregnant woman must be optimized. 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.

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.

IAEA Safety CriteriaThe safety criteria as stated in GSR Part 3 [4] are summarized below and are also used to assess exposure levels:

Protection of Workers An effective dose of 20 mSv per year averaged over five consecutive years (100

mSv in 5 years) and of 50 mSv in any single year; An equivalent dose to the lens of the eye of 20 mSv per year averaged over five

consecutive years (100 mSv in 5 years) and of 50 mSv in any single year; An equivalent dose to the extremities (hands and feet) or to the skin of 500 mSv

in a year.

Additional restrictions apply to occupational exposure for a female worker who has notified pregnancy or is breast-feeding (para. 3.114 of [4].

Protection of Public An effective dose of 1 mSv in a year; In special circumstances, a higher value of effective dose in a single year could

apply, provided that the average effective dose over five consecutive years does not exceed 1 mSv per year;

An equivalent dose to the lens of the eye of 15 mSv in a year; An equivalent dose to the skin of 50 mSv in a year.

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.

5.3 Development of Scenarios5.3.1 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.)

5.3.2 Accident Scenarios5.3.3 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 complies with the applicable requirements of the IAEA transport regulations, thus having the applicable signs on the truck and Tremcard available inside the vehicle. The vehicle capsizes, the driver/s cannot take emergency response action due to injury. 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.

5.3.3.1 Other Accident ScenariosThe following other accident scenarios will be considered in the Moroccan Safety Assessment:

- Accident Scenario 1: A DSRS waste container which is not filled to capacity and not conditioned are collected from the temporary storage location in the Waste Storage Facility. The container contains thirty 50 mCi Cs-137 sources. During liftingof the container with the overhead crane, the container slipped and fell to the floor of the reception area of the Storage Facility. On impact the container lost its lid and all the sources. The sources were in close approximation from each other. This happened close to the operator who spends 30 seconds within 1 m from the sources before the area is evacuated. After emergency intervention and planning the operatorspend 30 minutes at an average distance 1.5 m from the sources collecting them with tongs and placing them into a shielded container. The RPO supervised this operation at an average distance of 3 m from the sources. The container was closed and transferred back to the storage area.

- Accident Scenario 2: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 mCiCs-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 3: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.

5.4 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 are 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. Further justification for the applied dose rate data is provided in Annex A.The assumptions made aregenerally conservative.

6.0 SAFETY ASSESSMENT

6.1 Basic Engineering Analyses

Table 1: Basic Engineering Analysis

Item Requirement Compliance Ref Comments1. General: Facility Design, Construction and Maintenance1.1 Basic site characteristics

and credible external events have been considered in the design

yes Reference to design information –lacking (FSAR)

See comments in section 8.2.1

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

yes

References to e.g.




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)

design information and CoCs and are lackingRegulation code applicable

Reference to plans/ procedures are lacking

References to Systems/ procedures are lacking

2. Safety and security aspects were considered in the design of the facility2.1 The characteristics of the

walls ensuring a level of dose rate that complies with the restriction for public exposure (1 mSv/a or 2.5 µSv/h) even for the maximum anticipated inventory.

Reference to design input report lacking(FSAR)

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

Lighting system is according to French standards

Reference to lighting surveillance report lackingRegulation code applicable

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

Yes it was taken in consideration

Reference to section 8.2.2 (section and sub-sections need to be updated)FSAR

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.

Yes each area has sufficient space see the layout

Reference to section 8.2.2 (section and sub-sections need to be updated)

2.5 Storage areas were designed under the principle of labyrinth, which contributes to optimize the exposure of workers. (Stored DSRS

The storage area is designated in the form of vault with a sufficient thickness to protect worker


Assessment of doses to worker and public show that doses are well below limits (Annex )




and waste operations are not in taking place in the same area)

and people (see storage building description)

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.

The drum where the dismantled sources are conditioned will be staked not directly on the floor.

Reference to section 8.2.2 (section and sub-sections need to be updated) also reference to procedure to manage limits and conditions

2.7 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.

Yes

Yes

Yes

Yes

Reference to section 8.2.2 (section and sub-sections need to be updated)Reference to relevant procedures and plans are lacking


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.

No there is no vault with special shielding. Current storage system has a robust design and shielding capacity.


3. Engineering systems ensuring safety for situations of occurrences and accidents3.1 Floor and wall finish allow

easy decontamination.Yes Reference to

section 8.2.2 (section and sub-sections




need to be updated)

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

Yes Reference to section 8.2.2 (section and sub-sections need to be updated)

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.

Yes there is two tanks of 5 cubic meter


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

Facilities are equipped with fire detection and fighting equipment e.g. smoke detector and fire extinguishers (different types)


4. Facility design provides physical security features commensurate with the security threat4.1 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.The locking systems is digital each member should place his finger before putting the code when he wants to open the main doors of any building(treatment or storage)


4.2 Buildings are equipped with intrusion alarms.

Yes and infra redcamera placed outside the building at the entrance to each building


4.3 The buildings have vehicle access points. A separate personnel door

Yes. Reference to section 8.2.2 (section and




is provided to segregate personnel from vehicle movements.

sub-sections need to be updated)

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

Process and storage areas are not equipped with windows


6.2 Quantitative Deterministic Assessment of Worker Dose6.2.1 Activity 1:Collection of DSRS at Interim Stores

Table 2: Collection of DSRS at Interim StoresDose rate [µSv/h] Justification/Notes Time per action [h] Actions per year

Whole 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

6.2.2 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 5 6 10 300


1

Drivers (2) Driving TI or measured DR on consignments-50 % reduction in DR due to installed shield


6.2.3 Activity 3: Receiving of DSRS at CENM

Table 4: Receiving at CSF

Dose rate [µSv/h] Justification/Notes Time per action [h] Actions per yearWhole 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 60Whole 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


6.2.4 Activity 4:Temporary Storage of Category 3 Sources.




Table 5: Temporary StorageDose rate [µSv/h] Justification/Notes Time per action [h] Actions per year

Whole Body 10 Doserate at 1 m (1) 0.1 20 20Extremity 30 Contact doserate (2) 0.1 12 36

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


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

Table 6: Conditioning Campaign 1

Operator Groups Operator Actions

Exposure Type Exposure Data Exposure time Annual dose [µSv/a]

Dose rate [µSv/h] Justification/Notes

Time per action [h]

Actions per year

1Transporter(2

)Transport

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

Extremity 30Contact dose rate(2) 0.017 100 51

2

Operators(2)

Handling

Whole Body 10 Dose rate at 1 m (1)0.017 100 17


Dismantling

Whole Body 10Dose rate behind

shield (3) 0.1 100 100

Extremity 30Contact dose rate (2) 0.1 100 300

Source Transfer

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

Extremity 10000 Unshielded DR (5)0.005 100 5000

Inspection and maintenance



3 RPO (1)Supervision &surveying

Whole Body 10Ambient 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




6.2.6 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 Annual dose [µSv/a]

Dose rate [µSv/h] Justification/Notes

Time per action [h]

Actions per year

1Transporter(2

)Transport

Whole Body 15 Dose rate at 1 m (1) 0.25 3 11


2

Operators(2)

Handling



Dismantling

Whole Body 25 DR in shield (3)0.17 3 13


Source Transfer

Whole Body 5000DR on open source(4) 0.01 3 150

Extremity 10000DR on open source(5) 0.01 3 300

Inspection and Maintenance

Whole Body 10Ambient dose rate(6) 0.5 10 50

Extremity N/A3

RPO (1) Supervision & surveying

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

Extremity NA

Facility Surveillance

Whole Body 10Ambient 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 ratewithout 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 tongs)(6) - Maximum measured ambient dose rate in area during conditioning of sources(7) - Maximum measured ambient dose rate in facility




6.2.7 Activity 7: Transfer of Conditioned Waste Packages to the Waste Store including its Surveillance, Inspection and Maintenance

Table 8: Transfer to Waste StorageDose rate [µSv/h] Justification/Notes Time per action [h] Actions per year



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

6.2.8 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 240Extremity 600 120

Transport Whole Body 300 100 11 120Extremity 51 3 200

Operators

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

Dismantling Whole Body 100 13Extremity 300 26

Source Transfer Whole Body 1000 150Extremity 5000 300


Whole Body 40 120 50 1000Extremity 1200

RPO

Supervision and Surveying

Whole Body 40 12 500 15Extremity 60

Facility Surveillance

Whole Body 135 60Extremity

The maximum total dose to the Operator/Loader/Transporter is therefore: Whole body: 3.6mSv/a Extremity: 8.1mSv/a

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




6.3 Quantitative Deterministic Assessment of Worker and Public Dose ForAnticipated 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 µSvif simultaneouslyirradiated by 2 sources. The maximum additional dose to the worker would therefore be in the order of 25µSvif the same argument is used.

Occurrence Scenario2;-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 µSvand 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; Wholebody 1250 µSv and extremity 2500 µSv.

6.4 Deterministic Assessment of Worker and Public Dose for Other Accident Scenarios

Accident Scenario 1;- The maximum Operator and RPO doses are projected by the following:

The maximum Operator and RPO 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

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

(1) - Dose rate at 1 m calculated for an unshielded 50 mCi Cs-137 source using a specific gamma ray constant of 0.33 R.m2/Ci.h. The dose rate was adjusted for 30 sources by assuming a linear relationship between dose rate and activity (Simplified point source geometry)

(2) - Dose rate at 1.5 m calculated for 30 unshielded 50 mCi Cs-137 sources using the dose rate at 1 m and the inverse square law. (3) - Dose rate at 3 m calculated for 30 unshielded 50 mCi Cs-137 sources using the dose rate at 1 m and the inverse square law.

1111

3RPO Supervision Whole Body 556 [µSv/a]

Calcutalted dose rate at 3 m (3) 0.5 278

Whole Body 2222 [µSv/a]Calcutalted dose rate at 1.5 m (2) 0.5source recovery

Receptors Actions

1Operator source handling

2Operator

Exposure Type Exposure Data Exposure Parameters Dose [µSv]

Whole Body 5000 [µSv/a]Calcutalted dose rate at 1 m (1) 0.0167 83

The maximum total doses to the Operator and RPO is therefore: Whole body: 1.2mSv and 0.28 mSv respectively. Note that if Co-60 sources with the same activity are assumed that the approximate doses to the Operator and RPO would be 4.8mSv and 1.1mSv respectively.

Accident Scenario 2;- 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 11 below:

Table 11: Accident Scenario 2




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

Accident Scenario 3; 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 aboveare derived based on the assumptions, calculations and modelling indicated in Table 12below:

Table 12: Accident Scenario 2Units [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

6.5 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 13 below.




Table 13: 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 DoseOperator/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 transferaction,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 747µSv/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/ Transporter

25µSv - 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.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)


Operator/ 1250 µSv 2500µSv Possible to Evaluate the







Loader/ Transporter

prevent exposure by simple design and operational changes.

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 Operator and RPO Dose : Accident Scenario 1

Operator 1200 µSv - Dose mainly due to external radiation. Possible to prevent exposure by prevent falling and or loss of sources in the event of a fall event. Dose could also be reduced by using more operators during intervention.

Review design of source container

Review the source container handling procedure

Ensure that ALARA review is prescribed by intervention procedure in order to minimize individual dose.

RPO (Insignificant)

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

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) -Operator/Loader/ Transporter

3200µSv -

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

6.6 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

6.7 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 14: Assessment of Implemented Waste Management Practice

Item Requirement Compliance Comments Ref1. 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

Section 6.1




Item Requirement Compliance Comments Refthe waste management facilities demonstrate intent and commitment

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 of suitable containment.

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

Section 8.3

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

6.8 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 15: Management System Assessment

Item Requirement Compliance Comments Ref1. A written and approved

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

System Manual/s Procedures are referenced.

No documented reviews or audits.

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

Clear reference to System Manual/s Procedures. Review and audit reports are needed. Lists could also be attached- see email send on 2015-01-12.

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


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 limitationsand conditions associated with the facilities, their performance criteria and how and at what interval their performance will be





assessed and recorded, are listed

6.9 Assessment of Uncertainties

The outcome of a provisional quantitative assessment of uncertainties related to the safety case is presented in the table below:

Table 16: Assessment of Uncertainties

Item Uncertainty Comments/ Recommendations

Ref

1. Uncertainty in the source termused 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 thesafety 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.

7.0 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.

8.0 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;

8.1 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.




8.2 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.

8.3 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, personneldosimetry results and scheduled optimization review sessions, is recommended.

8.4 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.

8.5 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 below, addresses recommendations regarding the development of and integrated management system.

8.6 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.

9.0 COMPARISON WITH SAFETY CRITERIA AND CONCLUSIONS

The results of the quantitative safety assessment results as reflected in section Table 13 above, is well within the national and international safety criteria as listed in sections 6.3and 6.4 respectively for workers and 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.

Recommendations:o census datashould be updatedo Wind condition data should be updated and more and more recent data

10.0 INTERACTING PROCESSES

11.0 REFERENCES

Number Title1 GSR Part 5 International Atomic Energy Agency, Predisposal

Management of Radioactive Waste, IAEA Safety Standards Series No. GSR Part 5, IAEA, Vienna (2009).

2 GSG-3 International Atomic Energy Agency, 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 CNESTN 1996 Preliminary Safety Analysis Report from the Centre of Nuclear Studies of Mâamora, Volume I, Site and general description of the facilities, May 1996

6 Technicatome 1999 Technicatome, Design Floor Load, Document TA-103377, 1999

7 CNESTN 2005 CNESTN,Final Safety Analysis Report for the Radioactive Waste Facility (Rapport Définitif de Sûreté du CEN de la MAAMORA, MODULE DECHETS RADIOACTIFS), Volume IV, Module Déchetsradioactifs, CENM/RPS/01-05, Jan 2005

8 Technicatome 2002 Technicatome, CENM Design Criteria (Cen De La MaâmoraMethodologieD’examen Et Justification Du Choix Des Objectifs De Surete), TA-91048 D, Sep 2002.

9 CNESTN 2014 CNESTN,Quality Manual of the Exploitation of Radioactive Waste Unit, M/PEIN/UED/EX/001, Rev. 1, 28 April 2014.

10 Technicatome 2002 Technicatome, Fire Protection Design of the CENM (Cen De La Maamora Protection Incendie, TA-114703 D, Sep 2002.

11 Technicatome 1999 Technicatome, Applicable Regulations, Codes and Standards for the Construction of CENM, TA-103671, June 1999.

12 Technicatome 2003 Technicatome, CENM Radiation Protection Plan (Cen De La Maâmora Note Systeme Radioprotection Module De Système DE-A-03),




TA-106545 D, Sep 2003.13 CENM 2014 CENM Maintenance and Inspection Procedures,

DTL/SMT/2K7/167, 2014.14 CNESTN 2008 CNESTN, Procedure for the Collection and Receipt

of Solid Radioactive Waste and Sealed Sources, P./UED/EX/16, May 2008

15 CNESTN 2007 CNESTN, Procedure for the Decay Storage of Radioactive Waste and Sealed Sources contaminated by Short Lived Radionuclides, P/UED/IN/07, July 2007

16 CNESTN 2007 CNESTN, Emergency Response Plan for the Handling of Radioactive Waste and Sealed Sources using Overhead Crane, P/UED/IN/20, May 2007

17 CNESTN 2009 CNESTN, Procedure for Treatment (Evaporation) of Liquids contaminated by Radionuclides, P/UED/IN/17, Feb 2009

18 CNESTN 2006 CNESTN, Procedure for the Collection and Receipt of Liquids contaminated by Radionuclides, P/UED/EX/02, Feb 2006

19 CNESTN 2007 CNESTN, Procedure for the Handling of Containers and Heavy Items, P/UED/IN/18, May 2007

20 CNESTN 2007 CNESTN, Procedure for the Compaction of Solid Radioactive Waste, P/UED/IN/08, Apr 2007

21 CNESTN 2007 CNESTN, Procedure for the Conditioning of DSRS without Dismantling, P/UED/IN/10, June 2007

22 CNESTN 2007 CNESTN, Procedure for the Offsite Transport of Radioactive Waste and DSRS, P/UED/EX/04, Apr 2007

23 CNESTN 2007 CNESTN, Procedure for Laundry of non-contaminated PPE, P/UED/IN/09, Feb 2005

24 CNESTN CNESTN Safety Unit Procedures for Incident Reporting System

25 CNESTN 2005 CNESTN Final Safety Analysis Report: Emergency Response and Preparedness Plan




(CENTRE National De L’energie, Des Sciences Et Des Techniques Nucleaires, Centre D’etudesNucleaires De La Maamora, Rapport Definitif De Sûrete, Plan D’urgenceInterne)

26 Technicatome Technicatome Maintenance Unit Design review and construction reports,and certificates of conformance

27 CNESTN Annual safety report assessing the potential impact on safety of the dismantling of DSRS

28 IAEA 2005 INTERNATIONAL ATOMIC ENERGY AGENCY, Categorization of Radioactive Sources, IAEA Safety Standards Series No. RS-G-1.9, IAEA, Vienna (2005).

29 IAEA 1998 INTERNATIONAL ATOMIC ENERGY AGENCY, Clearance of Materials Resulting from the Use of Radionuclides in Medicine, Industry and Research, IAEA-TECDOC-1000, IAEA, Vienna (1998).

30 CNESTN 2005 CNESTN Final Safety Analysis Report: Physical Protection Programme (CENTRE National De L’energie, Des Sciences Et Des Techniques Nucleaires, Centre D’etudesNucleaires De La Maamora, Rapport Definitif De Sûrete, Plan D’urgenceInterne)

31 CNESTEN/CNRP 2014 CNESTEN/CNRP, Fifth National Report For the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management, Morocco, October 2004

31 CNRP 2014 Presentation by CNRP, meeting with Necsa Corp, Morocco, 2014.

32 GSR Part 4 International Atomic Energy Agency, Safety Assessment for Facilities and Activities, IAEA Safety Standards Series No. GSR Part 4, IAEA, Vienna (2009). Currently under revision (DS462)

33 Technicatome 1999 Technicatome, Design Parameters for the CENM Site HV/LV Electrical System, Electricity HV / LV, TA-104866, June 1999

34 SPIE 2014 SPIE, Maintenance Schedule for the CENM Waste




Management Building, , 2014




12.0 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 17: Aspects requiring further ClarificationItem Aspects Requiring Clarification Recommendation/Action1. Moroccan Legal and regulatory Framework1.1 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: 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.

The new law has been incorporated in principle in section 6.5. Sections of this document need to be revised once the new law is fully implemented

2. Basic Engineering Analyses2.1 A number of unresolved issued and

gapshave 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 Protection3.1 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 • Actions/audit to ensure/verify




exposure. 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.

Review the design and handling procedures of the source container with the objective to prevent falling and loss of shielding events.

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

Develop procedures (inspection and testing) to ensure housekeeping and storage practices relating to fire prevention and control are established and maintained.

4. Non-Radiological Hazards4.1 Comprehensive assessment of non-

radiological hazards. Plan, schedule and conduct a

comprehensive non radiological hazard assessment.

5. Implemented Waste Management Practice5.1 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 System6.1 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 Uncertainties7.1 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 Conditions8.1 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.




13.0 APPENDIXES

APPENDIX A: 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. The assumed values and justification for selection are reflected in Table 1 below.

Table 1: Assumptions and justifications for quantitative deterministic assessmentCondition Dose Rate Justification

1 Ambient dose rate in Storage room

15 µSv/h Typical ambient dose rate in a store full with DSRS units (units packed on various shelves and racks, racks about 1.5m apart) Dose rate measured between racks and in walkways)

2 Ambient dose rate in Receiving hall

10 µSv/h The DSRS inventory in the receiving area will typical be less than in the Storage room, thus assume a lower rate than in the storage room.

3 Average contact dose rate on DSRS units and equipment

70 µSv/h Typical average contact dose rate on DSRS units/equipment (as recorded on units received at Necsa (South Africa)- Can be changed based on Moroccan experience

4 Average dose rate 1m from DSRS units and equipment

5 µSv/h Typical average dose rate 1m from DSRS units/equipment (as recorded on units received at Necsa (South Africa)- Can be changed based on Moroccan experience – no measurement result on units at CNSTN available

5 Unshielded Cat 3 source dose rate at 0.5 m

12000 µSv/h No sources are handled by hand, thus always by tongs, assumed to be 0.5m long. This value is conservative average dose rate :0.5m dose rate for max Co-60 0.368Ci is 16700 µSv/h.1.5m dose rate for max Co-60 0.368Ci is 1800 µSv/h.

6 Unshielded Cat 3 source dose rate at 1.5 m

1200 µSv/h

7 Whole body dose due to shielded Cat 3 source at

50 µSv/h Removing sources behind 10 cm lead brick wall on the working bench. Person stands directly behind lead bricks with hands not shielded. Refer also to




Condition Dose Rate Justification

0.5m behind 10cm lead working shield

assumptions for 5 and 6 above

8 Average dose rate 1m from final package containing conditioned DSRS

50 µSv/h Typical average dose rates on a final package (e.g. 210L metal drum lined with concrete and cavity in center for a lead shield containing several capsules. These capsules each contain several DSRS).

9 Average contact dose rate on final package containing conditioned DSRS

1000 µSv/h




APPENDIX B: HOT SPOT DOSE CALCULATION

HotSpot Version 3.0 General Fire (Aug 06, 2014 09:03 AM)

Source Material : Cs-137 F 30.0yMaterial-at-Risk (MAR) : 3.7000E+07 BqDamage Ratio (DR) : 1.00Airborne Fraction (ARF) : 1.00E-02Respirable Fraction (RF) : 1.00E+00Leakpath Factor (LPF) : 1.000Respirable Source Term : 3.70E+05 BqNon-respirable Source Term : 0.00E+00 BqRelease Radius : 1 mCloud Top : 10 mPhysical Height of Fire : 5 mEffective Release Height : 8.52 mWind Speed (h=10 m) : 2.00 m/sAvg Wind Speed (h=H-eff) : 1.95 m/sStability Class : DRespirable Dep. Vel. : 0.30 cm/sNon-respirable Dep. Vel. : 8.00 cm/sReceptor Height : 1.5 mInversion Layer Height : NoneSample Time : 10.000 minBreathing Rate : 3.33E-04 m3/secDistance Coordinates : All distances are on the Plume Centerline

Maximum Dose Distance : 0.091 km Maximum TED : 9.67E-10 SvInner Contour Dose : 1.00E-10 SvMiddle Contour Dose : 1.00E-11 SvOuter Contour Dose : 1.00E-12 SvExceeds 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 SubmersionInclude 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 shownGround Roughness Correction Factor: 1.000

DISTANCE T E D RESPIRABLETIME-

INTEGRATED AIR CONCENTRATION

GROUND SURFACE

DEPOSITION

GROUND SHINE DOSE

RATE

ARRIVALTIME

(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 <00:010.200 5.2E-10 2.4E+02 7.3E-04 1.5E-12 00:010.300 2.9E-10 1.4E+02 4.1E-04 8.1E-13 00:020.400 1.9E-10 8.6E+01 2.6E-04 5.1E-13 00:030.500 1.3E-10 6.0E+01 1.8E-04 3.6E-13 00:041.000 4.1E-11 1.9E+01 5.7E-05 1.1E-13 00:082.000 1.3E-11 6.3E+00 1.9E-05 3.7E-14 00:17










14.0 ANNEXES

ANNEX A: STORAGE BUILDING DESIGN CALCULATIONS: RADIATION PROTECTION BARRIER

Input Data

Source characteristics:Cuboid of 8.8 x 5.9 m, height equal to 4 barrels, or 2.8 m.Physical composition: 10% steel, 60% plastic and 30% concrete, Bulk density: 0.66Activity: The total activity is equal to 178 GBq

Protective shield:The thickness of the walls is 40 cm concrete.The dose rate at the boundary of the outer contact to the wall of the cell must be <10 µSv/h.The dose rate at 1.3 m from the wall (external public area) must be 1 µSv/h.

Calculations

The points of calculation are:• Point A: dose rate at the outer contacts the wall of the cell,• Point B: dose rate at 1.3 m from the wall of the cell,• Point C: dose rate level of the upper bridge, between two cells, sense of the length.• Point D: dose rate at the level of the upper bridge between two cells, sense of the width




Results

PointsA B C D *

Dose rate µSv/h

0,4 0,5 2,5 45

Conclusion

The values of dose rate in the vicinity of the storage building respect the limits of a public area.On the bridge, the value of the dose rate is within the limit of zone 2, although it exceeds the value of the target dose rate of such a zone, which is why this area was declared a zone 3 ; This classification is not an obstacle for the operation, because the location is considered in all cases subject to limited stays

References

Technicatome 2002. Technicatome, Design Calculations of Radiation Protection Barriers (Cen De La Maamora Calculs De Dimensionnement Des Barrieres De Radioprotection), TA-99505C, Sep 2002




ANNEX B: WASTE MANAGEMENT BUILDING MAINTENANCE SCHEDULE

SYNTHESE DE LA MAINTENANCE DES EQUIPEMENTS IMPORTANTS POUR LA SURETESYNTHESE DES INTEVENTIONS ANNUELLES 2014 (Maintenance préventive)

Centre Technique Catégorie Fréquence CR RemarqueENSEMBLE DES EXTRACTEURS MP T Entretien préventif des extracteurs: - avec test d'automatisme des extracteursVENTILATION NUCLEAIRE MP H Contrôle de bon fonctionnement des équipements de la ventilationINDICATEURS TRANSMETTEURS DE PRESSION MP A Contrôle des indicateurs transmetteurs de pression (calibration)

EFFLUENTS MP H Contrôle des équipements de traitement et de stockage des effluentsCENTRALE DE TRAITEMENT D'AIR MP T Entretien préventif de la CTA.RADIO PROTECTION MP H Contrôle préventif des équipements de radioprotection du module.MANUTENTION MP Q Essai et contrôle des ponts roulants.

PROTECTION CORP: BOITES A GANTS MP S Contrôle de l’état des gants et les valeurs des dépressions sur l'ensemble des boites à gants.

RADIO PROTECTION/ETALONAGE MP A Etalonnage et vérification des équipements de radioprotection. – avec test de basculement des pièges à iode.

TOUR DE REFROIDISSEMENT MP T Entretien préventif de la tour de refroidissement.

NB : La balise MA002 est à l’arrêt : défaut du détecteur et de la carte de mesure qui doivent été changer




SYNTHESE DES INTEVENTIONS ANNUELLES 2014 (Travaux/Demandes)

Date d'intervention Catégorie Centre Technique CR Remarque

26/12/2014 Maintenance corrective VENTILATION NUCLEAIRE

Apres vérification, on a constaté un défaut sur u fusible FB004 qui protége l'alimentation normal secoure. changement du fusible et réarmement de l'installation et contrôle de bon fonctionnement.

24/07/2014 Travaux Associés VENTILATION NUCLEAIRE Test d'automatisme de la ventilation.

19/06/2014 Travaux Associés RADIO PROTECTION Déplacement de l’ancien contrôleur mains et pieds des modules pour installer le nouveau.

- Maintenance corrective RADIOPROTECTION Changement des filtres à charbon actif des balises et contrôle de bon fonctionnement

17/03/2014 Maintenance corrective VENTILATION NUCLEAIRE Réarmement de la ventilation après une coupure de la tension normal

Remarque:Les interventions sont réalisées suivant les procédures d'interventions sur les équipements des effluents, ventilation, radioprotection.