1 Scope
This standard specifies the criteria, analysis methods and evaluation contents to be followed in the analysis of loss-of-coolant accident for the performance evaluation of emergency reactor core cooling system of pressurized water reactor nuclear power plants.
It is applicable to the analysis of loss-of-coolant accident of pressurized water reactor nuclear power plants to evaluate the performance of emergency reactor core cooling system.
2 Normative references
The following referenced documents are indispensable for the application of this document. For dated references, only the edition cited applies. For undated references, the latest edition (including any amendments) applies.
NB/T 20103 Accident analysis and safety criteria for pressurized water reactor nuclear power pant
3 Terms and definitions
For the purposes of this document, the following terms and definitions apply.
3.1
loss-of-coolant accident (LOCA)
assumed accident of pipe failure at the pressure boundary of the reactor coolant system, ranging from the break that causes the loss-of-coolant rate exceeding the compensation capacity of the chemical and volume control system to the break that causes double-ended shear fracture of the largest pipe of the reactor coolant system
3.2
LOCA conservative evaluation code
a procedure for simulating the LOCA phenomenon and process by using a conservative model
3.3
LOCA best-estimate code
a procedure for simulating the LOCA phenomenon and process by using a realistic model
4 Restriction criteria
For the uranium dioxide fuel pellet, its cladding is made of zirconium alloy. If the assumed LOCA occurs in the pressurized water reactor nuclear power plant using cylindrical fuel rods, the acceptance criteria are as follows according to NB/T 20103:
Foreword i
1 Scope
2 Normative references
3 Terms and definitions
4 Restriction criteria
5 Analysis methods
