The Emergency Operating Strategy Study of Containment Spray Based on Radiation Level in LBLOCA Condition

Large break LOCA will cause an increase in containment pressure, containment temperature and containment radiation level in nuclear power plants (NPPs). Containment spray is one of the most effective ways to mitigate the consequences of large break LOCA for these following facts, first, with the large space containment design, the containment spray can decrease the pressure peak and keeps containment integrity. Secondly, the containment spray can decrease the aerosol radiation level in containment, iodine in particular, and reduce the risks of radioactive release. Above all, the common strategy of containment spray in NPPs generally includes automatic actuation with high spray flow, in order to achieve good results in relevant accident conditions. Meanwhile, the strategy to shutdown containment spray should be considered as a result of these facts that, a weakened effect in decreasing radiation will occur and negative containment pressure may cause containment integrity damage in post-accident long term operation. For the above considerations, the emergency operating strategy of containment spray based on radiation level in large break LOCA condition and the relevant best estimate work are studied based on one NPP in this paper, in order to achieve reasonable results in containment spray operating strategy, which are able to optimize containment spray and reduce the bad consequences.

Analysis of Processes in the Containment Using ATHLET-CD and COCOSYS Codes

A brief description of performed input deck modifications and results of stand-alone and coupled calculations of Dn 200 mm loss of coolant accident with simultaneous total station blackout accident scenario for Rivne Nuclear Power Plant Unit 1 (WWER‑440/V-213) with application of ATHLET-CD 3.1A and COCOSYS 2.4 codes are presented in the paper. ATHLET-CD stand-alone calculation was performed with constant containment pressure (a time dependent volume with constant pressure and temperature was used as a boundary volume for leakage). Further, mass and energy release and fission products from the primary system obtained during ATHLET‑CD stand-alone calculation were used to perform COCOSYS stand-alone calculation. In addition, coupled ATHLET-CD and COCOSYS calculation was performed. All the computer analyzes were performed until the lower head failure. ATHLET‑CD model was extended with core degradation module (ECORE), which allowed calculation of scenario until reactor pressure vessel failure. According to the results of comparative analysis, nearly the same behavior of the main parameters in the stand-alone and coupled calculation at an early phase of scenario was obtained. Some small differences occur due to distinction in behavior of water and steam mass flows released through the break and due to existence of heat transfer from the primary system structures to the containment compartments during coupled calculation of transient. As for middle and late phases of the accident, some differences between stand-alone and coupled calculation results for analyzed scenario are present. These differences are caused by different total fission products and aerosols release from the reactor coolant system to the containment compartments. The above information allows recommending application of coupled code/model versions for performing the computer severe accident analyses.

The Design and Analysis of a Containment Vacuum and Pressure Vessel System

A nested confinement (inner) and containment (outer) vessel system is under development to conduct small shock-physics experiments in a high-speed proton imaging facility at Los Alamos National Laboratory. The dual vessel system is necessary to serve as a qualified confinement system and containment buffer boundary between a high explosives experiment and the environment. The paper describes the preliminary engineering design and analyses that have been performed on the outer containment pressure vessel, following ASME BPVC Sect. VIII Div. 1, for both pressure and vacuum conditions. Other engineering attributes which will be presented include an internal support structure for a nested inner vessel, an external integrated support and alignment structure for the complete vessel system, and the vacuum and gas handling equipment.

Pool Scrubbing Efficiency of Hypothesized Filtered Containment Venting System for APR1400 Nuclear Power Plants

For a severe accident (a core melting accident) of nuclear power plants, a heat-up of the molten core might cause a overpressurizing of containment building to be damaged, if there couldn’t be given a proper cooling and/or a depressurizing strategy. In order to depressurize containment building and also to minimize the release of radioactive materials, filtered containment venting system (FCVS) might be used for a one of possible options. For a wet-type FCVS, radioactive aerosol released from molten core could be decontaminated by water pool, which is called pool scrubbing effect. The objective of this study is to find out regulatory insights for evaluating a wet-type FCVS for Korean nuclear power plant, APR1400. MELCOR, which is a severe accident analysis code developed by Sandia National Laboratories, was used for simulating postulated accidents. A full-plant scale calculation was performed considering the accident conditions such as temperature, pressure flow rate from containment to the pool of FCVS, behavior of radioactive materials and decontamination factors (DFs) for them. FCVS was operated with containment pressure set points. The decrease thermal margin between containment atmosphere and the pool of the FCVS influenced the DF, because the decreased amount of the steam due to the lowered thermal margin interrupted the radioactive aerosols and steam condensed.