3.4.1. This chapter of the safety analysis report should provide relevant information on the reactor to demonstrate its capability to fulfil relevant safety functions throughout the design life in all plant states. The reactor pressure vessel as a part of the reactor coolant system pressure boundary should be described separately in chapter 5 of the safety analysis report. The contents of chapter 4 of the safety analysis report should demonstrate compliance with Requirements 43–46 of SSR‑2/1 (Rev. 1) [3]. Recommendations on meeting the safety requirements applicable to this chapter of the safety analysis report are provided in IAEA Safety Standards Series No. SSG‑52, Design of the Reactor Core for Nuclear Power Plants [27]; the information included in this chapter should take account of those recommendations, as applicable.
Summary description
3.4.2. A summary description8 should be provided of the mechanical, neutronic and thermohydraulic behaviour of the various reactor components, including the fuel, the reactor vessel internals, the reactivity control systems, and related instrumentation and control systems.
3.4.3. For each of the reactor components, a more detailed description should be provided, in accordance with Appendix II.
Fuel design
3.4.4. A description should be provided of the main elements of the fuel9 (with account taken of Appendix II, as applicable), together with a justification of the selected design bases. The justification of the design bases of the fuel should include a description of the design limits for the fuel and the functional characteristics in terms of the desired performance under all plant states.
8 For this chapter and for other chapters of the safety analysis report, Appendix II provides guidance on describing the design of the nuclear power plant SSCs.
9 In this Safety Guide, the term ‘fuel’ means arrays (assemblies or bundles) of fuel rods, including fuel pellets, insulator pellets, springs, tubular cladding, end closures, hydrogen getters and fill gas; burnable poison rods, including components similar to those in fuel rods;
spacer grids and springs; end plates; channel boxes; and reactivity control rods.
Nuclear design
3.4.5. The following information should be provided in this section:
(a) The nuclear design bases, including nuclear design limits and reactivity control limits, such as limits on excess reactivity, fuel burnup, reactivity coefficients, neutron flux distribution, power distribution control and reactivity insertion rates;
(b) The nuclear characteristics of the lattice, including core physics parameters, fuel enrichment distributions in 235U (and plutonium vector contents, if applicable), distribution and concentrations of burnable poison rods, burnup distribution, boron reactivity coefficient and boron concentrations, type of control rods and their locations, shutdown margin specification, and refuelling schemes;
(c) The analytical tools, methods and computer codes (together with information on code verification and validation, including uncertainties) used to calculate the neutronic characteristics of the core, including reactivity control characteristics;
(d) The additional nuclear safety parameters of the reactor core, such as radial and axial power peaking factors and the maximum linear heat generation rate;
(e) The neutronic stability of the core, including xenon stability, throughout an operating cycle, with consideration given to possible anomalies in the different modes of normal operation covered by the design basis;
(f) Special core configurations, such as a mixed core or mixed modes of normal operation.
Thermohydraulic design
3.4.6. This section should provide the following information:
(a) The thermohydraulic design bases for the reactor core and attendant structures, and the interface requirements for the thermohydraulic design of the reactor coolant system;
(b) The analytical tools, methods and computer codes (including their verification and validation, together with consideration of the uncertainties) used to calculate thermohydraulic parameters;
(c) Flow, pressure and temperature distributions, with specification of the limiting values and their comparison with the design limits;
(d) A demonstration of the thermohydraulic stability of the core.
Design of the reactor control, shutdown and monitoring systems
3.4.7. The reactor control, shutdown and monitoring systems should be described in this section of the safety analysis report. It should be demonstrated that these systems, including any essential auxiliary equipment and hydraulic systems, are designed and installed to provide the required functional performance and are properly isolated from other equipment. In addition, the design limits and the design evaluation of the reactor control, shutdown and monitoring systems should be described.
Evaluation of the combined performance of reactivity control systems 3.4.8. This section should describe the relevant situations in which two or more reactivity control systems are used during accidents and should provide an evaluation of the combined functional performance.
3.4.9. This section should also include failure analyses that demonstrate that the reactivity control systems are not susceptible to common cause failures. These analyses should consider failures originating within any of the reactivity control systems as well as those originating from other plant equipment and should be accompanied by comprehensive and logical supporting discussions.
Core components
3.4.10. This section of the safety analysis report should provide descriptions of the following:
(a) The systems of core components, defined as the general external details of the fuel, the structures into which the fuel has been assembled (e.g. fuel rods assembled into a fuel assembly or fuel bundle), related components necessary for fuel positioning and all supporting elements internal to the reactor, including any separate provisions for moderation and fuel location.
Reference should be made to the other sections of the safety analysis report that cover related aspects of the reactor core as well as fuel handling and storage.
(b) The physical and chemical properties of the materials used for the core components, including the neutronic, thermohydraulic, structural and mechanical characteristics of the components.
(c) The expected response of core components to static and dynamic mechanical loads and the behaviour of these components with respect to design limits, together with a description of the effects of irradiation and corrosion on the
ability of the core components to fulfil their safety functions adequately over the lifetime of the plant.
(d) Any significant subsystem component, including any separate provision for moderation and fuel location, with corresponding design drawings.
(e) The conclusions from a consideration of the effects of in‑service maintenance programmes on the fulfilment of safety functions, including surveillance and inspection programmes to monitor the effects of irradiation and ageing on the core components.
CHAPTER 5: REACTOR COOLANT SYSTEM AND ASSOCIATED