DETERMINISTIC SAFETY DEMONSTRATION

The safety demonstration, whose premises have to be presented within the DOPF volume 3, covers [12]: Events considered occurring and consequences considered in the design; Events which have to be practically eliminated, as they would lead to large or early radioactive release.

5.1. Events considered to occur and consequences considered in the design

The safety analysis, as part of the safety assessment, should proceed in parallel with the design process, with iteration between the two activities. The scope and level of detail of the safety analysis should increase as the design process progresses, so that the final safety analysis reflects the final design of the reactor as constructed.

5.1.1. Analysis of postulated single initiating events

The postulated single initiating events are those corresponding to the DiD Level 3.a (cf.

Table1). The associated radiological consequences shall meet safety objective O2 of WENRA. The response of the research reactor to the postulated initiating event should be predicted by conservative deterministic safety analyses.

5.1.2. Analysis of postulated multiple failure events without core melt

The postulated multiple failure events addressed in this section are those corresponding to the DID Level 3.b (cf. Table1). The associated radiological consequences shall meet safety objective O2 of WENRA. The response of the research reactor to the postulated multiple failure events may be predicted by a best estimate plus uncertainty safety analyses.

5.1.3. Analysis of core melt accidents

The postulated “core melt” accidents addressed in this section are those corresponding to the DID Level 4 (cf. Figure 3). The associated radiological consequences shall meet safety objective O3 of WENRA. The response of the research reactor to the “core melt” accidents may be predicted by a best estimate plus uncertainty safety analyses.

5.2. Events which have to be practically eliminated, as would lead to large or early radioactive release

Initiators, consequential faults, fuel melt sequences challenging the confinement resulting in accidental situations with a large or early release could be rejected and excluded from further analysis of radiological consequences provided that an acceptable justification is given by the designer to the regulatory authority. An acceptable justification is to demonstrate

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that any accident sequence with a large or early release is practically eliminated, i.e. if it is physically impossible for the accident sequence to occur or if the accident sequence can be considered with a high degree of confidence to be extremely unlikely to arise.

In order to quantify the notion of “extremely unlikely” it is important to give insights concerning the order of magnitude of acceptable reliability for the different upstream DiD levels. The residual risk of having early and large releases should be pushed to very low probabilities being aware that the demonstration cannot be claimed solely based on compliance with a general cut-off probabilistic value. Because of the existence of unpredictable common mode failures there is a limit to the reliability that can be allocated to the layers of provisions which materialize the DID levels. Other criteria will be considered as for example the simplicity of the safety architecture or the demonstrated degree of knowledge for the phenomena involved in the accident sequence. The final assessment will be done on a case by case basis.

6. CONCLUSIONS

The Belgian Federal Agency for Nuclear Control (FANC) is engaged in a process of pre-licensing for the experimental reactor MYRRHA. The regulatory framework applicable for the implementation of this project is described by FANC in a Strategic Note. The steps required to design a basic nuclear installation and to ensure compliance with the requirements of nuclear safety, security and safeguards are described and commented within a complementary document which present the template of a Design Options and Provision File (DOPF) which shall be prepared by the designer to engage the exchanges with the regulator.

The DID represents a pillar for both Strategic Note and the DOPF template. The two documents focus more specifically on aspects relating to refinements in the concept of DID and on practices for their integration in the process of design / assessment of the MYRRHA reactor.

REFERENCES

[1] MYRRHA: Multi-purpose Hybrid Research Reactor for High-tech Applications, http://myrrha.sckcen.be/

[2] FANC, 2011-04-09-MSC-5-3-2-FR, Description d'un processus de pre-licensing pour un projet de construction d'une installation nucléaire nouvelle et complexe, Révision 1.

[3] HAKIMI, N., MYRRHA Strategic Note, AFCN 2011-10-13-NH-5-4-3-EN, Revision 2.

[4] FIORINI, G.L., Guidance for the format and content of the Design Options and Provisions File, AFCN 2011-10-12-NH-5-4-3-EN, Revision 1.

[5] INTERNATIONAL ATOMIC ENERGY AGENCY, Fundamental Safety Principles, IAEA Safety Standards Series, Safety Fundamentals No. SF-1, IAEA, Vienna (2006).

[6] INTERNATIONAL ATOMIC ENERGY AGENCY, Safety of Nuclear Power Plants:

Design, IAEA Safety Standards, Specific Safety Requirements No. SSR-2/1, IAEA, Vienna (2012).

[7] INTERNATIONAL ATOMIC ENERGY AGENCY, Safety of Research Reactors, IAEA Safety Standards, Safety Requirements No. NS-R-4, IAEA, Vienna (2005).

[8] WESTERN EUROPEAN NUCLEAR REGULATORS ASSOCIATION (WENRA), WENRA Reactor Safety Reference Levels, January (2008).

[9] WESTERN EUROPEAN NUCLEAR REGULATORS ASSOCIATION (WENRA), WENRA Statement on Safety Objectives for New Nuclear Power Plants, November (2010)

27 [10] US DEPARTMENT OF ENERGY (DOE), A Technology Roadmap for Generation IV

Nuclear Energy Systems, GIF-002-00, USDOE, Washington (2002).

[11] INTERNATIONAL ATOMIC ENERGY AGENCY, Nuclear Security Recommendations on Physical Protection of Nuclear Material and Facilities, IAEA Nuclear Security Series No. 13, INFCIRC/225/revision 5, IAEA, Vienna (2011).

[12] WESTERN EUROPEAN NUCLEAR REGULATORS ASSOCIATION (WENRA), WENRA Booklet: Safety of New NPP Designs, October (2012).

[13] GENERATION IV INTERNATIONAL FORUM (GIF), Basis for the Safety Approach for Design & Assessment of Generation IV Nuclear Systems, GIF/RSWG/2007/002 Revision 1, November (2008).

[14] EUROPEAN UTILITY REQUIREMENTS for LWR Nuclear Power Plants, Revision C, April (2001). http://www.europeanutilityrequirements.org/

[15] GENERATION IV INTERNATIONAL FORUM (GIF), An Integrated Safety Assessment Methodology (ISAM) for Generation IV Nuclear Systems, GIF - Risk and Safety Working Group (RSWG), Version 1.1, June (2011).

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REINFORCEMENT OF DEFENCE-IN-DEPTH: MODIFICATION PRACTICE