Systematic Source Term Analyses for Level 3 PSA of a BWR With Mark-II Type Containment With THALES-2 Code

Author(s):  
Jun Ishikawa ◽  
Ken Muramatsu ◽  
Toru Sakamoto

The THALES-2 code is an integrated severe accident analysis code developed at the Japan Atomic Energy Research Institute in order to simulate the accident progression and transport of radioactive material for probabilistic safety assessment (PSA) of a nuclear power plant. As part of a level 3 PSA being performed at JAERI for a 1,100MWe BWR-5 with a Mark-II containment, a series of calculations were performed by THALES-2 to evaluate the source terms for extensive accident scenarios. For some of the containment failure modes not modeled in THALES-2, such as steam explosion, simple models were coupled with the analysis results of THALES-2 to estimate the source terms. This paper presents the methods and insights from the analyses. An insight from the analyses was that the source terms depend more strongly on the differences in the containment function failure scenarios, such as overpressure failure, controlled containment venting, and small leakage to the reactor building, than those core damage sequences.

1985 ◽  
Vol 28 (6) ◽  
pp. 17-23
Author(s):  
John Graham

The nuclear source term, defined as the quantity, timing, and characteristic of the release of radioactive material to the environment following a core-melt accident, was thoroughly debated in 1985. This debate, summarized here, turns on the Nuclear Regulatory Commission's (NRC) source term for radioactive iodine, which is postulated as potentially the most life-threatening radionuclide that might escape in a nuclear power-plant accident. A generic radioiodine source term has been used by NRC as the surrogate for all others; thus, it has become one of the bases on which nuclear-safety regulations are founded. Following the Three Mile Island (TMI) accident, from which only traces of radioiodine escaped, scientists began arguing that nuclear regulations based on source-term calculations are erroneous and should be modified. The American Nuclear Society (ANS) and industry researchers have concluded that warranted reductions in the NRC source terms could range from a factor of ten to several factors of ten in most accident scenarios. The American Physical Society (APS), after agreeing with a large body of the conclusions from the other research groups, has told NRC that its source-term data base is still inadequate because of the existence of a number of uncertainties it found therein. Although APS presented no such conclusion, its findings made clear to NRC that an early reduction of all source terms is not warranted. The anti-nuclear lobby agrees with APS. The NRC has taken a cautious, conservative approach to the revision of its regulations based on new source-term data, although it too concedes that its old methodologies and conclusions must be revised and ultimately superceded.


Author(s):  
Hsoung-Wei Chou ◽  
Chin-Cheng Huang

After the Fukushima nuclear accident occurred in Japan on March 11, 2011, the compound disaster beyond design basis which may severely impact the nuclear safety has been noticed and paid much more attention. In addition to the original emergency operating procedures (EOPs) and severe accident management guidance (SAMG) of nuclear power plant, the licensee in Taiwan developed the ultimate response guideline (URG) when EOPs and SAMG cannot be performed effectively due to loss of power and water supply by the Fukushima-like compound disaster. Once the URG procedures are initiated, the operators will conduct reactor depressurization, low pressure water injection and containment venting to strictly prevent the core damage and the release of radioactive material. In the paper, the fracture probabilities of boiling water reactor (BWR) pressure vessels with incremental levels of radiation embrittlement under URG operation are evaluated by probabilistic fracture mechanics (PFM) analysis. First, the models of PFM FAVOR code concerning the beltline shell welds of reactor pressure vessels (RPVs) associated with a very conservative flaw distribution are built. Then, the hypothetical transients of URG operation obtained from the thermal hydraulic analyses for Taiwan domestic BWRs are applied as the loading condition. The analysis results demonstrate that performing URG operation will not cause significant fracture probability for RPV, even at an extremely embrittled condition. The URG procedures can ensure the prevention of core damage as well as maintenance of structural integrity of RPV in the situation of long-term loss of electric power when suffering from the Fukushima-like accidents.


Author(s):  
Naoto Kasahara ◽  
Izumi Nakamura ◽  
Hideo Machida ◽  
Hitoshi Nakamura ◽  
Koji Okamoto

As the important lessons learned from the Fukushima-nuclear power plant accident, mitigation of failure consequences and prevention of catastrophic failure became essential against severe accident and excessive earthquake conditions. To improve mitigation measures and accident management, clarification of failure behaviors with locations is premise under design extension conditions such as severe accidents and earthquakes. Design extension conditions induce some different failure modes from design conditions. Furthermore, best estimation for these failure modes are required for preparing countermeasures and management. Therefore, this study focused on identification of failure modes under design extension conditions. To observe ultimate failure behaviors of structures under extreme loadings, new experimental techniques were adopted with simulation materials such as lead and lead-antimony alloy, which has very small yield stress. Postulated failure modes of main components under design extension conditions were investigated according three categories of loading modes. The first loading mode is high temperature and internal pressure. Under this mode, ductile fracture and local failure were investigated. At the structural discontinuities, local failure may become dominant. The second is high temperature and external pressure loading mode. Buckling and fracture were investigated. Buckling occurs however hardly break without additional loads or constraints. The last loading is excessive earthquake. Ratchet deformation, collapse, and fatigue were investigated. Among them, low-cycle fatigue is dominant.


Author(s):  
Kampanart Silva ◽  
Yuki Ishiwatari ◽  
Shogo Takahara

Risk evaluation is an important assessment tool of nuclear safety, and a common index of direct/indirect influences of severe accidents as a compound of risk is necessary then. In this research, various influences of severe accidents are converted to monetary value and integrated. The integrated influence is calculated in a unit of “cost per severe accident” and “cost per kWh”. The authors must emphasize that the aim is not to estimate the accident cost itself but to extend the scope of “risk-informed decision making” for continuous safety improvements of nuclear energy. To calculate the “cost per severe accident” and the “cost per kWh”, typical sequences of severe accidents are picked-up first. Containment failure frequency (CFF) and source terms of each sequence are taken from the results of level 2 probabilistic risk assessment (PRA). The source terms of each sequence is input into the level 3 PRA code OSCAAR which was developed by Japan Atomic Energy Agency (JAEA). The calculations have been made for 248 meteorological sequences, and the results presented in this study are given as expectation values for various meteorological conditions. Using these outputs, the cost per severe accident is calculated. It consists of various costs and other influences converted into monetary values. This methodology is applied to a virtual 1,100 MWe BWR-5 plant. Seismic events are considered as the initiating events. The data obtained from the open documents on the Fukushima Accident are utilized as much as possible. Sensitivity analyses are carried out to identify the dominant influences, sensitive assumptions/parameters to the cost per accident or per kWh. Based on these findings, optimization of radiation protection countermeasures is recommended. Also, the effects of sever accident management are investigated.


Author(s):  
Kwang-Il Ahn ◽  
Jae-Uk Shin

The primary purpose of this study is to assess the release of source terms into the environment for representative spent fuel pool (SFP) severe accident scenarios in a reference pressurized water reactor (PWR). For this, two typical accident scenarios (loss-of-cooling and loss-of-pool-inventory accidents) and two different reactor operating modes (normal and refueling modes) are considered in the analysis. The secondary purpose of this study is to assess the impact of an emergency makeup water injection strategy, which is one of representative SFP severe accident mitigation (SAM) strategies being employed after the Fukushima accident, upon the release of the radiological source terms. A total of 16 cases, consisting of four base cases and three injection cases for each base case were simulated using the MELCOR1.8.6 SFP version. The, analysis results are given in terms of (a) the key thermal-hydraulic behaviors during an accident progression and (b) releases of radiological fission products (such as Cesium and Iodine) into the environment. In terms of a release of Cesium and Iodine into the environment, the present study show that the two cases subject to a loss of pool inventory (i.e., LOPI-N-03 and LOPI-R-00) lead to the worst results with the respective release fractions of 77.5% and 59.4%.


Author(s):  
Zibin Liu ◽  
Dingqing Guo ◽  
Bing Zhang ◽  
Jinkai Wang

The phenomenon of temperature-induced steam generator tube rupture (TI-SGTR) is a typical phenomenon in the severe accident process of nuclear power plants. The occurrence of the phenomenon may result in the radioactive material bypass the containment, causing a large radioactive release. This paper investigates modeling methods of the phenomenon of temperature-induced SGTR in level 2 PSA and presents an optimizing modeling method to calculate the probability of branching probability of TI-SGTR, aiming at improving the rationality and veracity of level 2 PSA.


2011 ◽  
Vol 2011 ◽  
pp. 1-13 ◽  
Author(s):  
Vadim E. Seleznev ◽  
Vladimir V. Aleshin ◽  
Sergey N. Pryalov

The paper describes one of the variants of mathematical models of a fluid dynamics process inside the containment, which occurs in the conditions of operation of spray systems in severe accidents at nuclear power plant. The source of emergency emissions in this case is the leak of the coolant or rupture at full cross-section of the main circulating pipeline in a reactor building. Leak or rupture characteristics define the localization and the temporal law of functioning of a source of emergency emission (or accrued operating) of warmed up hydrogen and steam in the containment. Operation of this source at the course of analyzed accident models should be described by the assignment of the relevant Dirichlet boundary conditions. Functioning of the passive autocatalytic recombiners of hydrogen is described in the form of the complex Newton boundary conditions.


Author(s):  
Hiroto Itoh ◽  
Xiaoyu Zheng ◽  
Hitoshi Tamaki ◽  
Yu Maruyama

The influence of the in-vessel melt progression on the uncertainty of source terms was examined in the uncertainty analysis with integral severe accident analysis code MELCOR (Ver. 1.8.5), taking the accident at Unit 2 of the Fukushima Daiichi nuclear power plant as an example. The 32 parameters selected from the rough screening analysis were sampled by Latin hypercube sampling technique in accordance with the uncertainty distributions specified for each parameter. The uncertainty distributions of the outputs, including the source terms of the representative radioactive materials (Cs, CsI, Te and Ba), the total mass of in-vessel H2 generation and the total debris mass released from the reactor pressure vessel to the drywell, were obtained through the uncertainty analysis with an assumption of the failure of drywell. Based on various types of correlation coefficient for each parameter, 9 significant uncertain parameters potentially dominating the source terms were identified. These 9 parameters were transferred to the subsequent sensitivity and uncertainty analyses, in which the influence of the transportation of radioactive materials was taken into account.


Author(s):  
Robert J. Lutz ◽  
James H. Scobel ◽  
Richard G. Anderson ◽  
Terry Schulz

Probabilistic Risk Assessment (PRA) has been an integral part of the Westinghouse AP1000, and the former AP600, development programs from its inception. The design of the AP1000 plant is based on engineering solutions to reduce or eliminate many of the dominant risk contributors found in the existing generation of Pressurized Water Reactors (PWRs). Additional risk reduction features were identified from insights gained from the AP1000 PRA as it evolved with the design of the plant. These engineered solutions include severe accident prevention features that resulted in a significant reduction in the predicted core damage frequency. Examples include the removal of dependencies on electric power (both offsite power and diesel generators) and cooling water (service water and component cooling water), removal of common cause dependencies by using diverse components on parallel trains and reducing dependence on operator actions for key accident scenarios. Engineered solutions to severe accident consequence mitigation were also used in the AP1000 design based on PRA insights. Examples include in-vessel retention of molten core debris to eliminate the potential for ex-vessel phenomena challenges to containment integrity and passive containment heat removal through the containment shell to eliminate the potential for containment failure due to steam overpressure. Additionally, because the accident prevention and mitigation features of the AP1000 are engineered solutions, the traditional uncertainties associated with the core damage and release frequency are directly addressed.


2016 ◽  
Author(s):  
Kenji Iino ◽  
Ritsuo Yoshioka ◽  
Masao Fuchigami ◽  
Masayuki Nakao

The Great East Japan Earthquake on March 11, 2011 triggered huge tsunami waves that devastated the northeast region of Japan along the Pacific coastline. The Tokyo Electric Power Company (TEPCO) owned Fukushima Daiichi Nuclear Power Plant (Fukushima-1) survived the earthquake, however, not the tsunami that followed. Four of the 6 reactor units underwent Station Blackout. Unit 5 lost all its own AC power, however, it shared AC power with Unit 6. Units 1, 3, and 4 had hydrogen explosions that destroyed their reactor buildings, and even worse, 1, 2, and 3 had core meltdowns to release a large amount of radioactive material to their surroundings. The accident was rated Level 7 on the International Nuclear Event Scale, the worst level defined by International Atomic Energy Agency (IAEA). Reports and papers have been published by a number of entities including the Japanese Diet, Government, TEPCO, IAEA, and more. They give detail explanation of how the accident developed into a nuclear disaster explaining the direct and background causes and faults made after the accident broke out. Finding the accident process, i.e., how it happened, and its causes of why it happened, are the most important first steps in accident analysis. Figuring out how to prevent similar events in the future, or even if it is possible to do so, however, is equally important for our future. We started our study in 2014 to find what actions TEPCO could have taken before the accident for preventing it from growing into a catastrophe. Then in February 2015, we set the goal of our study group to find answers to the following two questions: A. Was the huge tsunami, induced by a huge earthquake, predictable at Fukushima-1? B. If it was predictable, what preparations at Fukushima-1 could have reduced the severity of the accident? In response to our invitation to experts in the nuclear field, active and retired people gathered from academia, manufacturers, utility companies, and even regulators. After a series of tense discussions, we reached the conclusions that: Aa. Tsunami of the level that hit Fukushima-1 in 2011 was well predictable, and, Ba. The accident would have been much less severe if the plant had prepared a set of equipment, and most of all, had exercised actions against such tsunami. Preparation at the plant to prevent the severe accident consisted of the following items 1 through 7, and drills in 8: 1. A number of 125Vdc and 250Vdc batteries, 2. Portable underwater pumps, 3. Portable AC generators with sufficient gasoline supply to run the pumps, and 4. High voltage AC power truck This set applied only to this specific accident. For preparing against many other situations that could have taken place at Fukushima-1, we recommend having, in addition, the following equipment and modifications. 5. Portable compressor to drive air-operated valves for venting, 6. Watertight modification to RCIC and HPCI control and instrumentation, 7. Fire engines for alternate low pressure water injection after vent (Fukushima-1 had three). Just making these preparations would not have been sufficient. Activating valves with DC batteries, for example, takes disengaging the regular power supply lines and hooking up the batteries. 8. Drills against extended loss of all electric power and seawater pump This item 8, on and off-site drills was the most important preparation that should had been made. All other necessary preparations to save the plant in such cases would have followed logically.


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