GB/T 47511-2026Safety verification requirements for changes in the core fuel management mode of pressurized water reactor nuclear power plants (English PDF)

7.1 Core Fuel Management and Nuclear Design

7.1.1 When the fuel management mode changes, the actual configuration of the unit and external demand should be comprehensively considered, and the nuclear power plant's reaction should be determined based on the following conditions. Core loading schemes for each fuel cycle of the reactor.

a) Types and fuel enrichment of new fuel assemblies;

b) The number of new fuel assemblies loaded into the reactor core for each fuel cycle;

c) New fuel assembly loading methods (e.g., "low-leakage" core loading mode);

d) Types, placement, and quantities of flammable and toxic substances;

e) Core fuel cycle length;

f) Flexible operating mode;

g) Main operating parameters of the reactor.

7.1.2 Nuclear designs based on core fuel management models shall meet (but are not limited to) the following design limits.

a) Design limits for nuclear enthalpy rise thermal channel factor (F NDeltaH);

b) Design limits for peak heat flux density (FQ);

c) Limits on the temperature reactivity coefficient of moderators;

d) Minimum shutdown margin limit;

e) Maximum fuel consumption limits for fuel assemblies and/or fuel rods;

7.2 Parameters required for accident analysis, fuel and system validation

7.2.1 General Rules When the core fuel management mode changes, the parameters required for accident analysis, fuel and system validation should be adjusted accordingly. The characteristics and requirements determine the parameters, which typically include.

a) General critical safety parameters;

b) Specific key security parameters;

c) Parameters required for transient analysis during normal operation;

d) Parameters required for design verification of fuel rods and fuel assemblies;

e) Parameters required for system validation;

f) Parameters required for analyzing the long-term transient effects of boron concentration on Loss-of-Cooler Accidents (LOCA);

g) Parameters required for decay heat calculation;

h) Parameters required for fuel pellet-cladding interaction (PCI) analysis;

i) Parameters required for Expected Transient (ATWS) analysis in cases where emergency shutdown fails.

7.2.2 General Key Safety Parameters Typical common key safety parameters include.

a) Maximum and minimum values of the temperature (or density) reactivity coefficient of the moderator;

b) The maximum and minimum values of the Doppler temperature reactivity coefficient;

c) Maximum and minimum values of the Doppler power coefficient;

7.7 Accident Analysis and Assessment of Radioactive Consequences

7.7.1 General Rules Accident analysis and radiological consequences assessment should cover the accidents analyzed in FSAR, using the parameters provided in

7.2 and their... He designed the input data, taking full account of its conservative requirements.

7.7.2 Accident analysis

7.7.2.1 For specific accidents where key safety parameters change or exceed the original FSAR assumption boundaries after a change in fuel management mode. For typical accidents, conduct accident analysis and evaluation to confirm that the accident consequences meet safety limits. Analysis must be performed using specific accident-specific key safety parameters. Typical accidents include.

a) Analysis of boron dilution accidents;

b) Analysis of control rod drop accidents;

c) Accident analysis of control rod assembly runaway under subcritical or low-power start-up conditions;

d) Accident analysis of single-beam control rod malfunction under power operation conditions;

e) Analysis of control stick ejection incidents;

f) Analysis of a main steam pipeline rupture accident;

g) Accident analysis of rod assembly runaway during power operation;

h) Analysis of fuel assembly misloading accidents;

i) LOCA incident analysis;

j) Non-LOCA incident analysis;

k) Mass-energy release calculation and containment response analysis.

7.7.2.2 For other FSAR accidents that are not affected by changes in fuel management mode or are already covered by the accident analysis, the following should be included. Its operating conditions and assumptions are compared with those of FSAR, and the changes in accident analysis methods, core power distribution, and other assumptions are comprehensively considered. The current situation demonstrates that these accidents do not require further detailed analysis. General key safety parameters based on envelope analysis can be used to analyze these accidents. A comprehensive evaluation should be conducted to confirm that the FSAR conclusions remain valid. A full accident analysis should still be performed if necessary. Common key safety parameters should be used. Typical incidents for which the FSAR conclusions are evaluated and confirmed to remain valid include.

a) A malfunction in the water supply system caused a decrease in water supply temperature;