The SAR for an NPP contains information regarding BDBAs, including SAs, according to NP-006-16. Rostechnadzor’s requirements are established for the scope and structure of the SAR for NPP with WWER-type reactors. The report is submitted along with a set of documents to support the application to receive a license for NPP construction and operation.
All BDBAs and SAs scenarios resulting in excessive radiation doses for the NPP personnel and local public are singled out based on the results of the analysis and the standards established for DBAs for radioactivity release and its content in the environment. The scenarios to be considered in the documents pertaining to SAR are summarized in a list for further analysis.
Upon completion of the development of the base calculational model, including validation, the specific models (applied to each scenario in question) are developed.
The qualification certificate [
14] for the SOCRAT CC has the following limitation: the calculation results at the initial stage of an accident (before exceeding the maximum design limit of fuel rod damage) must be justified using certified thermohydraulic codes.
One method to validate the mathematical model for the CC is cross-verification with results obtained using other certified codes. Currently, the domestic certified codes TECH-M-97 and KORSAR/GP are adopted to perform computational analysis of the initial stage of an SAs. In this regard, an important challenge is the process of coordinating the behavior of RP parameters determined using the certified thermohydraulic codes TECH-M-97 and KORSAR/GP and using the SAs code SOCRAT, before the transition to the severe stage of the accident (i.e., before exceeding the maximum design limit of fuel rod damage) [
7,
8,
10,
12,
13].
The spectrum of SAs scenarios is fairly wide and is continuously expanding because of the development, design, and implementation of new equipment, as well as preparation and revisions of new regulatory documents. All the SAs scenarios considered at present in safety computational analysis can be divided into three key types: blackout, loss-of-primary coolant, loss-of-secondary coolant. The aforementioned types of accidents are often combined by additional failures. To ensure that the results to be obtained are adequate, it is necessary to perform cross-verification of the governing scenarios of SAs using certified thermohydraulic codes (TECH-M-97 and KORSAR/GP). The lists of governing scenarios differ depending on the specific design of the RP (type of a plant, additional equipment, etc.).
The accidents considered [
7,
8,
10,
12,
13] for the AES-2006 design include “NPP blackout with failure of steam generators heat removal system (SG PHRS),” “large break (LB) loss of coolant accident (LOCA) with failure of an emergency core coolant system (ECCS) active part,” and “small break (SB) LOCA with failure of an ECCS active part.”
Comparisons of the results obtained using the thermohydraulic codes and CC SOCRAT/B1 for these scenarios are provided in Refs. [
7,
8,
10,
12,
13].
The analysis of the findings from these documents has indicated that insignificant differences between some parameters obtained using the CC SOCRAT and the certified thermohydraulic codes TECH-M-97 and KORSAR/GP do not affect general accident scenarios. The specified differences are caused by variations in the modeling of the thermohydraulic processes as well as in the nodalization topology.
Different values of coolant flow through the reactor core under the conditions of natural circulation are caused by differences in the calculational models for steam generators. SOCRAT and KORSAR/GP use different multi-element models of steam generators. The differences in the topology of these multi-element models lead to the differences in the heat exchange between the primary and secondary circuits under natural circulation. This causes the differences in some parameters of the primary and secondary circuits: coolant flow through the core, primary pressure, pressurizer level, steam generator levels, and pressures.
Insignificant differences in the flows from (PCFS) (HA-2) are caused by different approaches used for modeling PCFS (HA-2). The TECH-M-97 code utilizes a special calculation model for PCFS (HA-2). However, in SOCRAT and KORSAR/GP, PCFS (HA-2) is modeled based on boundary conditions.
The differences in the primary coolant mass values obtained using TECH-М-97, SOCRAT, and KORSAR/GP are explained mainly by the different coolant releases into a pipeline break, which occur because of the use of different critical flow models. The consequence is that a higher water level in the reactor core is obtained using KORSAR/GP, and accordingly, time of the heating-up of the core internals in calculation using the CC SOCRAT.
There are insignificant differences between the values of steam flow through steam dump valve to the atmosphere (BRU-A). Hence, the differences in the steam generator pressures are caused by the use of different models of BRU-A operation.
The KORSAR/GP version compiled to run in multicore processors is used for calculations, which significantly shortens the computational time. The steam-zirconium model for the aforementioned version is currently under testing and improvement. Thus, the calculations using KORSAR/GP are performed without considering steam-zirconium reactions during fuel rod heating, which causes deviations in the values presented in graphs (maximum fuel rod cladding temperatures).
Based on the results of pre- and post-test calculations during experiments at the PSB-WWER test bench [
22] using KORSAR/GP (considering the model simulating the steam-zirconium reaction) and TECH-М-97, as well as based on the results of post-test calculations using SOCRAT, the capability to model LOCA adequately is verified in principle for all the presented CCs. The difference in the times for heating up the fuel rod claddings among the used CCs is approximately 10% of the experimental values.
Based on the calculation results for the considered BDBA [
7,
8,
10,
12,
13], with the support from the certified thermohydraulic codes, the SOCRAT SA CC allows for modeling the main processes and phenomena involved BDBAs, including severe core damage. The obtained results are compatible to those derived from the certified thermohydraulic codes TECH-M-97 and KORSAR/GP.
If the developed mathematical model for SOCRAT is adjusted in accordance with the results of calculations performed using the TECH-M-97 and/or KORSAR/GP, it will be possible to perform calculations that agree well with those for the initial stages of accidents.
The calculations performed in OKB Gidropress JSC using the certified SOCRAT/В1 CC for safety verification pertain to the in-vessel phase (before meltdown of the reactor pressure vessel and release of the entire mass of the melt and solid fragments) of the SAs. The measures to manage the accident and mitigate its consequences are planned accordingly, such that the maximum design limit of fuel rod damage is not exceeded, if possible.
The objectives of the calculations performed using SOCRAT/В1 for the in-vessel stage of BDBAs include the evaluation of time for the following main events; the beginning of fuel heating; the melting of the core and internals and transport of the molten materials to the lower head of the reactor pressure vessel (formation of a melt pool); the damage to the lower head of the reactor vessel and release of the melt and solid fragments into the reactor concrete cavity; the determination of the parameters of the melt (mass, temperature, fractional composition) flowing outside the reactor pressure vessel during vessel damage and the melt flowing into the reactor concrete cavity; and the determination of the parameters of the coolant (steam and water) and hydrogen flowing from the RP into the containment space during the accident, beginning with an initiating event and ending with the melt being released into the reactor concrete cavity.
These processes are presented schematically in Fig. 7.
Operational safety objectives are established for each level of BDBA severity, i.e., the objectives that the operating personnel at the NPP shall strive to achieve to avoid or limit damage to safety-related equipment and/or systems or to limit the release of radioactive materials into the environment.
Based on the computational analyses of BDBAs, the NPP conditions are determined. Using these conditions, the criteria for BDBA occurrence are identified, and the source of the BDBA can be traced by referring to the level of severity.
The section containing the BDBA analysis in the SAR for NPP identifies all engineering systems at the NPP (including the non-safety-related systems) that can be involved possibly for their out-design purpose and under out-design conditions of operation, to achieve the operative safety objectives and mitigate the accident consequences at each level of severity. The redundancy of the systems fulfilling similar functions is considered as well. Possible usage of the materials and equipment located at neighboring units as well as beyond the NPP site is described, and facilities for their delivery are planned. Success criteria for personnel actions are established to achieve the operational safety objectives at each level of accident severity. These criteria are defined based on the NPP conditions.
For calculations to verify the safety (Preliminary SAR (PSAR), In-depth Safety Assessment Report (ISAR), Final SAR (FSAR)), the acceptance criteria must be met.
When performing computational analyses of BDBAs, two types of criteria are considered: one for versions without core melting and the other for versions with core melting.
The RP safety for versions without core melting is verified, as a rule, based on the acceptance criteria:
-the primary and secondary pressure should not exceed the design pressure by 15% (with consideration of transient dynamics and a time of safety device actuation);
-no local melting of fuel;
-the maximum design limit of fuel rod damage should not be exceeded, i.e., the fuel rod cladding temperature should not exceed 1200°С, the local fraction of fuel rod cladding oxidation should not exceed 18% of the initial wall thickness, and the fraction of zirconium reacted should not be less than 1% of its mass in the fuel rod claddings;
-the fuel matrix integrity should be maintained while considering the operating conditions of the fuel.
The results obtained using the certified thermohydraulic codes TRAP-97 or KORSAR/GP are accepted as a basis for verification of meeting these criteria.
The RP safety for versions with core melting is verified, as a rule, based on the following acceptance criteria:
-the primary and secondary pressure should not exceed 115% of the design pressure (with consideration of transient dynamics and a time for safety device actuation);
-if the core remains cannot be cooled inside the reactor vessel, the pressure in the primary coolant system at the point of melting should not exceed 1 MPa;
-the limiting pressures and temperatures for ensuring containment integrity should not be exceeded.
The following approaches are additionally considered:
-it is necessary to consider the capability to manage BDBAs, which will be required after an accident occurs and based on which the detailed instructions are prescribed;
-the systems whose functioning does not involve active components may be considered as factors mitigating the consequences of the accident or limiting the reactivity release.
The SAR for an NPP describes the sequences of events, actuations, failures of systems, and equipment for BDBA scenarios included in a list for a specific NPP. An accident scenario is presented in the form of a table containing the main stages and the respective durations.
Considering the aforementioned criteria, the safety verification calculations involve calculations for different accident scenarios:
-calculation with regard to the operating personnel’s actions aimed at accident management;
-calculation with regard to the operating personnel’s actions aimed at accident management to reduce the primary pressure when damage to the reactor pressure vessel reduces the pressure below 1.0 MPa (for example, a forced opening of pilot-operated relief valves (PRZ PORV) and/or emergency gas removal system (EGRS) by operating personnel);
-calculation with regard to the accident management actions of operating personnel to prevent core heating, and transition of the accident to a severe stage with uncontrolled release of radioactive products outside the NPP (for example, RP cooling through BRU-A with water supply into the primary or the secondary circuit).
A description of the thermohydraulic processes occurring in the primary and secondary circuits of the RP are provided for all BDBAs (including SAs). The scope of the information to be provided should cover the following parameters and initial conditions:
-heat flux characteristics;
-pressure variation in circuits during an emergency transient;
-temperature variation of fuel rod claddings and fuel in the core components;
-coolant flowrates in the reactor and loops;
-primary coolant parameters at the inlet and outlet of the hottest channels of the reactor core;
-heat-engineering fuel characteristics;
-secondary coolant parameters;
-coolant flowrate in different systems affecting the scenario of an emergency transient;
-mass (fraction) of zirconium reacted with water steam in the core;
-hydrogen release from the reactor core and primary circuit;
-flowrate and enthalpy of coolant flowing out of the circuit.
Calculations for the in-vessel stage of BDBAs (including SAs) to verify safety can be performed considering the mutual effects of accidents at the RP and within the containment area. For this purpose, the certified CC ANGAR [
21] or KUPOL [
20] is used.
The thermohydraulic processes occurring in the containment rooms are simulated for accidents involving the release of coolant and/or core materials from the primary circuit into the containment. The scope of the information to be provided shall cover the following parameters:
-heat flux characteristics.
The SAR for an NPP describes the strategy of the corrective personnel’s actions under BDBA conditions, which are aimed at achieving the safety objectives at all possible levels of accident severity. The calculations indicate that implementation of the planned strategy during a BDBA caused by manifestation of any of the identified vulnerabilities at all possible levels of accident severity ensures either termination of the accident processes or significantly mitigates the consequences of the accident.
The thermohydraulic processes occurring in the reactor concrete cavity or fuel catcher, if included in the design, are simulated for SAs involving core material melting and falling out of the reactor pressure vessel into the containment.
The results of the SAs in-vessel stage calculations performed using the certified CC SOCRAT/В1 serve as initial data for subsequent calculations.
The scope of the presented information should cover at least the following parameters:
-change in the aggregate state of the melt components;
-variation in the temperature of the melt and cavity concrete or the structural components of a catcher;
-characteristics of heat fluxes;
-operation of the cooling systems of a catcher;
-change in configuration of the core cavity because of concrete damage;
-change in the thickness of the reactor compartment base plate at fuel melt location;
-mass (fraction) of zirconium and other metals reacted with steam;
-characteristics of steam explosions (energy released, parameters of shock waves affecting the reactor vessel and other structures of the RP and the containment).
The SAR for an NPP shall contain the analysis of the release and spreading of radioactive materials. The results of the SAs in-vessel stage calculations performed using the certified CC SOCRAT/В1 serve as initial data for subsequent calculations. The results also serve as the initial data for calculations to verify hydrogen explosion safety.
Based on the information presented in the section containing the BDBA analysis in the SAR, conclusions are drawn regarding the efficiency of the measures planned to manage BDBAs.
Based on the information in the SAR, the Russian Rostechnadzor assesses the sufficiency of the justifications for siting, construction, commissioning, operation, and decommissioning of an NPP at a specific site. These objectives are to avoid exceeding the assigned irradiation doses for personnel and the local public, comply with the standards for the release of radioactive substances and their content in the environment during normal operation and under DBAs, and determine the capability to mitigate the consequences of BDBAs.