Transient Analysis of Dedicated Pressure Relief Valve Under High-Pressure Condition in Severe Accident for Tianwan 56 Nuclear Power Plant

In the management of severe accident of nuclear reactor, the pressure relief of reactor coolant system (RCS) is an important mitigation measure to prevent high pressure core melt (HPCM). In the safety system improvement of Tianwan56 nuclear power plant, the optimization measure of adding the dedicated pressure relief valve (DPRV) for severe accident were adopted. This improvement allows the reactor to release the pressure of RCS before the reactor vessel being damaged to mitigate the consequence of reactor melt accident under high-pressure condition. Based on the analysis of severe accident sequences, the total loss of feed water accident is confirmed to cover the various severe accident consequences which may lead to HPCM accident. This paper studied the transient characteristics of total loss of feed water accident sequences, and the factors such as valve opening delay on the operating temperature of the valve were researched. Finally, the representative and envelope operating condition of DPRV under severe accident was clarified. Besides, the temperature curve of fluid passing through the valve and the maximum temperature the valve experienced were obtained. This research provides the valuable and indispensable basis to the operability and integrity analysis of DPRV in severe accident.

Level 1 PRA for External Vessel Storage Tank of Japan Sodium-Cooled Fast Reactor in Scheduled Refueling

Spent fuels are transferred from a reactor core to a spent fuel pool through an external vessel storage tank (EVST) filled with sodium in sodium-cooled fast reactors in Japan. This paper describes identification of dominant accident sequences leading to fuel failure by conducting probabilistic risk assessment for EVST designed for a next sodium-cooled fast reactor plant system in Japan to improve the EVST design. Based on the design information, this study has carried out identification of initiating events, event and fault tree analyses, human error probability analysis, and quantification of accident sequences. Fuel damage frequency of the EVST was evaluated approx. 10−6 /year in this paper. By considering the secondary sodium freezing, the fuel damage frequency was twice increased. The dominant accident sequence resulted from the common cause failure of the damper opening and/or the human error for the switching from the stand-by to the operation mode in the three stand-by cooling circuits. The second dominant accident sequence following the secondary pump trip is sodium freezing caused by the failure of air blower trip in the air cooler due to the common cause failures of secondary sodium flowmeter failure or erroneous opening of the air cooler damper. The Fussell-Vesely importance and risk achievement worth analyses have indicated high risk contributions.

Source Term Analysis Considering B4C/Steel Interaction and Oxidation During Severe Accidents

The Japan Atomic Energy Agency (JAEA) is pursuing the development and application of the methodologies on fission product (FP) chemistry for source term analysis by using the integrated severe accident analysis code THALES2. In the present study, models for the eutectic interaction of boron carbide (B4C) with steel and the B4C oxidation were incorporated into THALES2 code and applied to the source term analyses for a boiling water reactor (BWR) with Mark-I containment vessel (CV). Two severe accident sequences with drywell (D/W) failure by overpressure initiated by loss of core coolant injection (TQUV sequence) and long-term station blackout (TB sequence) were selected as representative sequences. The analyses indicated that a much larger amount of species from the B4C oxidation was produced in TB sequence than TQUV sequence. More than a half of carbon dioxide (CO2) produced by the B4C oxidation was predicted to dissolve into the water pool of the suppression chamber (S/C), which could largely influence pH of the water pool and consequent formation and release of volatile iodine species.

Development of the Severe Accident Evaluation Method on Second Coolant Leakages From the PHTS in a Loop-Type Sodium-Cooled Fast Reactor

The loss-of-reactor-level (LORL) is a typical accident condition causing a severe accident (SA) in sodium-cooled fast reactors (SFRs). In the loop-type SFR Monju, important cause of the LORL is the second coolant leakage at the low elevation of the primary heat transport system (PHTS), which occurs in cold standby in a different loop from that of the first coolant leakage in rated power operation because of excessive declining of the sodium level. This study developed an evaluation method for the LORL with considering countermeasures to prevent LORL: i.e., pumping sodium up into reactor vessel (RV) and interruption of sodium flowing out by siphon effect. To evaluate the effectiveness of the countermeasures, the transient behavior in the RV sodium level was analytically studied in representative accident sequences. The representative sequences with lowest sodium level were selected by considering combinations of possible coolant leakage positions. The evaluation result clarified that the LORL can be prevented by conducting the above-mentioned countermeasures to maintain the RV sodium level sufficient for the operation of decay heat removal system, even after the second coolant leakage of PHTS.

Study of Equipment Survivability Under Severe Accident Conditions

To improve nuclear power plant safety, severe accident prevention and mitigation for both new development and existing plants are generally required by various nuclear safety authorities worldwide. Although great efforts have been made, how to ensure equipment survivability under severe accident conditions is still a concern. This paper depicts an approach to demonstrate the equipment survivability under severe accident conditions by taking passive pressurized water reactor CAP1400 as an instance, including screening of severe accident sequences, determination of bounding environment conditions within containment, equipments identification used for severe accident mitigation and proposed test plan.