Retention of Molten Corium in Calandria Vessel of PHWR With Moderator Drain Pipe

Retention of molten corium inside calandria vessel is crucial for arresting accident progression in pressurized heavy water reactors (PHWRs) during severe accidents. Our earlier tests have demonstrated corium retention and its cooling inside the calandria vessel of PHWRs through external cooling by vault water. However, the presence of nozzles and moderator drain pipe at the bottom of calandria vessel has not been considered in these studies. These nozzles and drain pipes used for moderator circulation can make the viability of corium retention even more challenging. Once the moderator has evaporated, debris reheating, compacting, and finally melting can cause the release of molten corium into the moderator recirculation system. This can lead to the relocation of corium beyond calandria vessel. The corium might reach the pump room or calandria vault after the failure of moderator drain pipe and/or moderator pump seals. This has severe consequences on containment integrity due to molten corium concrete interaction (MCCI). The risks posed by MCCI can be avoided if corium can be contained inside calandria vessel even with the presence of nozzles (at the bottom of the vessel) or if at all it enters into the drain line, does not cause its failure. Thus, it becomes crucial to evaluate the challenges faced by “in-vessel retention” (IVR) as a severe accident management strategy due to the presence of openings in the calandria vessel. Relatively colder debris present near the bottom of calandria vessel might help in obstructing the nozzles of the moderator drain line and can prevent the entry of hot molten corium into the moderator cooling line. The role of debris, therefore, becomes important under such scenarios for not just insulation of calandria vessel from hot corium but also for retention of corium within the vessel. In this article, these issues are addressed by conducting two sets of experiments for assessment of retention capability (IVR) of calandria vessel: (i) with the presence of debris and (ii) without debris at the bottom of calandria vessel. The moderator recirculation line was scaled to simulate the heat transfer from corium to vault water and solidification of corium simulant while flowing through the moderator drain pipe. It was observed that debris bed present at the bottom of the vessel helps in arresting the molten corium front and thus prevents corium from entering into moderator drain pipe. When experiments were conducted without debris, molten corium was found to be relocating in the moderator drain pipe. The drain pipe, however, did not fail under the thermal load.

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 the Effect of Severe Accident Scenario on Debris Properties in Lower Plenum of Nordic BWR Using Different Versions of MELCOR Code

Nordic Boiling Water Reactors (BWRs) employ ex-vessel debris coolability as a severe accident management strategy (SAM). Core melt is released into a deep pool of water where formation of noncoolable debris bed and ex-vessel steam explosion can pose credible threats to containment integrity. Success of the strategy depends on the scenario of melt release from the vessel that determines the melt-coolant interaction phenomena. The melt release conditions are determined by the in-vessel phase of severe accident progression. Specifically, properties of debris relocated into the lower plenum have influence on the vessel failure and melt release mode. In this work we use MELCOR code for prediction of the relocated debris. Over the years, many code modifications have been made to improve prediction of severe accident progression in light-water reactors. The main objective of this work is to evaluate the effect of models and best practices in different versions of MELCOR code on the in-vessel phase of different accident progression scenarios in Nordic BWR. The results of the analysis show that the MELCOR code versions 1.86 and 2.1 generate qualitatively similar results. Significant discrepancy in the timing of the core support failure and relocated debris mass in the MELCOR 2.2 compared to the MELCOR 1.86 and 2.1 has been found for a domain of scenarios with delayed time of depressurization. The discrepancies in the results can be explained by the changes in the modeling of degradation of the core components and changes in the Lipinski dryout model in MELCOR 2.2.

Comparative Calculations of TROI TS-2 and TS-3 Steam Explosion Experiments With TEXAS-V

During a severe accident, contact of the molten corium with the coolant water may cause an energetic steam explosion which is a rapid increase of explosive vaporization by transfer to the water of a significant part of the energy in the corium melt. This steam explosion has been considered as an adverse effect when the water is used to cool the molten corium and could threaten reactor vessel, reactor cavity, containment integrity. In this study, TROI TS-2 and TS-3 experiments as part of the OECD/SERENA-2 project were analyzed with TEXAS-V. Input parameters were based on actual TROI experiment data. In mixing simulations, calculated results were compared to melt front behavior, void fraction in trigger time and other parameters in experiment results. In explosion simulations, corresponding to TROI experiments an external triggering was employed at the moment that melt front reached heights of 0.4 m. Calculated results of peak pressure and impulse at the bottom were compared with TROI experiment results. Melt front behaviors of the melt was different from the experimental results in both TS-2 and TS-3. Void fraction in triggering time in TS-2 was in good agreement with the experiment results and in TS-3 was slightly overestimated. The peak pressure and impulse at bottom were successfully predicted by TEXAS-V. These calculations will allow establishing whether the limitations and differences observed in the simulations of the experiments are important for the reactor case.

Effects of Filtered Containment Venting on Fission Product Releases During CANDU Reactor Severe Accidents

Severe accidents are of increasing concern in the nuclear industry worldwide since the accidents at Fukushima Daiichi (March 2011). These events have significant consequences that must be mitigated to ensure public and employee safety. Filtered containment venting (FCV) systems are beneficial in this context as they would help to maintain containment integrity while also reducing radionuclide releases to the environment. This paper explores the degree to which filtered containment venting would reduce fission product releases during two Canada Deuterium Uranium (CANDU) 6 severe accident scenarios, namely a station blackout (SBO) and a large loss of coolant accident (LLOCA) (with limited emergency cooling). The effects on the progression of the severe accident and radionuclide releases to the environment are explored using the Modular Accident Analysis Program (MAAP)–CANDU integrated severe accident analysis code. The stylized filtered containment venting system model employed in this study avoids containment failure and significantly reduces radionuclide releases by 95–97% for non-noble gas fission products. Filtered containment venting is shown to be a suitable technology for the mitigation of severe accidents in CANDU, maintaining containment integrity and reducing radionuclide releases to the environment.

Strategies to Ensure Safety Measures for ABWRs in Kashiwazaki-Kariwa Nuclear Power Station

This paper describes strategies to ensure safety measures for ABWRs in Kashiwazaki-Kariwa Nuclear Power Station reflecting the lessons learned from the Fukushima Daiichi accident. The accident and response actions were analysed to extract lessons. Based on the lessons policies to enhance safety and containment integrity were derived. Firstly with considerations of multiple failures of the Fukushima Daiichi accident, defense in depth (DID) was enhanced by applying more diverse safety measures. For this purpose, in addition to refurbishing safety measures for beyond design base events, safety measures for each DID layer was enhanced not only by strengthen robustness to single failure but also by strengthen diversity and by physical separation. Especially capacities for ensuring containment integrity have been drastically improved by installing diverse safety measures for the third layer and the fourth layer of DID including the alternative coolant circulation system. Secondly phased approach was introduced in choosing mitigation measures considering timing of the response actions and required reliabilities. The basic concept of phased approach is that we have to select safety measures based on the assumption that possible and available measures and their required reliability are different depending on time to spare. Lastly performance requirements was clarified for containment vessel and its auxiliary systems after core damage. The basic policy was to define clearly twice design pressure of PCV as its upper limit pressure and to define 200 degree C as PCV upper limit temperature, and then to define function requirements for other auxiliary systems in order to contain fission products in the PCV and to make them decay as long as possible not only by enhancing PCV capacity but by utilizing auxiliary systems to ensure PCV integrity after core damage. Various safety measures were implemented based on these policies and applied at ABWRs in Kashiwazaki-Kariwa Nuclear Power Station.

Development of New MHI LOCA Analysis Code: MCOBRA/RELAP5-GOTHIC

Mitsubishi Heavy Industries, Ltd. (MHI) has developed the safety analysis code system, MCOBRA/RELAP5-GOTHIC, for large break LOCA, small break LOCA and containment integrity analyses of PWR. The code system contains three codes, MCOBRA code, M-RELAP5 code and GOTHIC code. MHI implemented some models and correlations to MCOBRA and M-RELAP5 code to improve prediction of important phenomena of LOCA event. MHI conducted a lot of analyses for separate effect tests and integral effect tests to confirm the applicability of the MHI’s safety analysis code to LOCA key phenomena, like core heat transfer, DNB, entrainment, break flow, counter-current flow limitation (CCFL), core froth level, and so on. The code model uncertainties are quantified to calculate core heat transfer break flow rate, loop seal phenomena; heat transfer in steam generators, and so on. The results show that the code system can predict the key phenomena well and can calculate LOCA event scenario. Consequently, it is confirmed that the new code system is applicable to the LOCA licensing analysis.

The Assessment of Containment Functional Failure Frequency in the Revised Level 2 PRA Standard in Japan

In this paper, we introduce the overview of the requirements and the complementary information on the evaluation of containment functional failure frequency (CFF) in the revised version of “A Standard for Procedures of Probabilistic Risk Assessment of Nuclear Power Plants during Power Operation (Level 2 PRA) “[1] in Japan, which was developed and revised at the Level 2 PRA Subcommittee under the Atomic Energy Society of Japan (AESJ). Although the Level 2 PRA standard includes the evaluation of CFF and radiological source terms, we explain only the evaluation of CFF in this paper. In the evaluation of CFF, the physical response analysis and the probabilistic analysis are included as follows. The accident progression analysis is performed for each of the plant damage states, considering the operation status of mitigation systems, thermal-hydraulic behavior and core damage progression, and occurrences of some key events such as reactor pressure vessel failure. The containment event tree (CET) is developed classifying the accident progress in tree diagram. In the CET, some headings are arranged sequentially considering the accident progression. The headings correspond to the phenomena occurrence and the systems operation status, and a branch probability is assigned at each branch of heading. The branch probabilities of the phenomena are evaluated by either the Risk Oriented Accident Analysis Methodology (ROAAM) or the Decomposition Event Tree (DET) analysis considering the containment threats. The branch probabilities on the phenomena are set as the probability distributions, because the phenomena and the analysis have uncertainties. The branch probabilities on the systems operation are evaluated using the fault tree analysis and human error analysis. The containment functional failure modes are assigned at the end state of the CET considering the type of load against containment integrity. For the evaluation of the non-energetic load, the integral codes such as MELCOR [2], THALES-2 [3], and MAAP4 [4] etc. are used. On the other hand, various mechanistic codes are used for the evaluation of energetic phenomena such as steam explosion. The containment functional failure is judged by comparing the ultimate strength or the fragility of containment structure and the generated loads. After all, the CFF can be obtained by summing the frequency of containment functional failure mode. In the Level 2 PRA standard in Japan, the requirements in each evaluation process above are described. In addition, the technical background and the examples as the complementary information on each requirement are described in the Annex of the standard to help the application of the standard. In this revision, the body is revised to clarify the requirements on the quantification of the CET. The Annex is revised to incorporate the up-to-date information on severe accident research and severe accident management (SAM) measures. The updated information includes the melt stratification (OECD/MASCA project [5]), the steam explosion (SERENA project [6] and PULiMS/SES experiments [7]), the ex-vessel debris coolability (OECD/MCCI project [8]), debris jet breakup, the melt spreading, the coolability of the particulate bed, and the containment vessel (CV) fragility evaluation. Some future challenges are extracted from the lessons learned from the Fukushima Daiichi accident, such as development of the Level 2 PRA for the external hazard as earthquake and tsunami, quantification of impact on the containment integrity of hydrogen detonation in the adjacent buildings, and human error evaluation in the external hazard.