The holistic risk analysis of spent fuel pool considering the fuel route risk

2021 ◽  
pp. 1748006X2199851
Author(s):  
Zhixin Xu ◽  
Ming Wang ◽  
Binyan Song ◽  
WenYu Hou ◽  
Chao Wang
Keyword(s):  
Nuclear Power ◽  
Comprehensive Evaluation ◽  
Point Of View ◽  
Spent Fuel ◽  
Nuclear Disaster ◽  
Evaluation Approach ◽  
Fuel Assemblies ◽  
Spent Fuel Pool ◽  
Large Water ◽  
Fukushima Nuclear Disaster

The Fukushima nuclear disaster has raised the importance on the reliability and risk research of the spent fuel pool (SFP), including the risk of internal events, fire, external hazards and so on. From a safety point of view, the low decay heat of the spent fuel assemblies and large water inventory in the SFP has made the accident progress goes very slow, but a large number of fuel assemblies are stored inside the spent fuel pool and without containment above the SFP building, it still has an unignored risk to the safety of the nuclear power plant. In this paper, a standardized approach for performing a holistic and comprehensive evaluation approach of the SFP risk based on the probabilistic safety analysis (PSA) method has been developed, including the Level 1 SFP PSA and Level 2 SFP PSA and external hazard PSA. The research scope of SFP PSA covers internal events, internal flooding, internal fires, external hazards and new risk source-fuel route risk is also included. The research will provide the risk insight of Spent Fuel Pool operation, and can help to make recommendation for the prevention and mitigation of SFP accidents which will be applicable for the SFP configuration risk management.

Analyses of Severe Accident Sequences in the Spent Fuel Pool of the VVER-1000 Type of Reactor With MELCOR 1.8.6

2014 ◽  
Author(s):  
Miroslav Kotouč
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Complex Geometry ◽  
Steel Plates ◽  
Severe Accident ◽  
Spent Fuel ◽  
Concrete Walls ◽  
Nuclear Disaster ◽  
Spent Fuel Pool ◽  
Fukushima Nuclear Disaster

Since the Fukushima nuclear disaster in 2011, much attention has been paid to investigation of severe accidents (SA) progression in spent fuel pools (SFP) of various types of nuclear power plants (NPP). In Czech Republic, 4 VVER-440 and 2 VVER-1000 types of reactors (at the Dukovany and Temelin NPPs, respectively) are currently under operation. In order to enhance their safety, especially with respect to station black-out (SBO) events, numerical analyses have been carried out evaluating the risks associated with accidents occurring also in the SFP. The present paper deals with analyses of 2 postulated scenarios (loss of cooling and loss of coolant) and is mainly focused on the input deck preparation for the integral, lumped parameter (LP) code for SA analyses — MELCOR 1.8.6. Emphasis is put on description of correct implementation of the complex geometry of the SFP, consisting of 3 distinct pools separated by concrete walls (lined with steel plates) in which fuel assemblies (FA) are stored in an absorber grid (rack). In the description of the prepared numerical model, light is shed on the encountered modeling issues, and corresponding hints and solutions are proposed in order to provide guidance for preparing adequate models for various types of SFP designs. Finally, some of the most characteristic results are presented for both postulated scenarios.

scholarly journals Temperature conditions in the RBMK spent fuel pool in the event of disturbances in its cooling mode

2021 ◽  
Vol 7 (1) ◽  
pp. 9-13
Author(s):  
David A. Hakobyan ◽  
Victor I. Slobodchuk
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Spent Nuclear Fuel ◽  
Maximum Temperature ◽  
Spent Fuel ◽  
Cooling Mode ◽  
Temperature Conditions ◽  
Fuel Assemblies ◽  
Spent Fuel Pool

The problems of reprocessing and long-term storage of spent nuclear fuel (SNF) at nuclear power plants with RBMK reactors have not been fully resolved so far. For this reason, nuclear power plants are forced to search for new options for the disposal of spent fuel, which can provide at least temporary SNF storage. One of the possible solutions to this problem is to switch to compacted SNF storage in reactor spent fuel pools (SFPs). As the number of spent fuel assemblies (SFAs) in SFPs increases, a greater amount of heat is released. In addition, no less important is the fact that a place for emergency FA discharging should be provided in SFPs. The paper presents the results of a numerical simulation of the temperature conditions in SFPs both for compacted SNF storage and for emergency FA discharging. Several types of disturbances in normal SFP cooling mode are considered, including partial loss of cooling water and exposure of SFAs. The simulation was performed using the ANSYS CFX software tool. Estimates were made of the time for heating water to the boiling point, as well as the time for heating the cladding of the fuel elements to a temperature of 650 °С. The most critical conditions are observed in the emergency FA discharging compartment. The results obtained make it possible to estimate the time that the personnel have to restore normal cooling mode of the spent fuel pool until the maximum temperature for water and spent fuel assemblies is reached.

Analysis of the HI-TRAC VW Transfer Cask Dose Rates for Spent Fuel Assemblies Loaded in Nuclear Power Plant Krsko Storage Campaign One

2021 ◽  
Author(s):  
Davor Grgic ◽  
Mario Matijevic ◽  
Paulina Duckic ◽  
Radomir Jecmenica
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Variance Reduction ◽  
Spent Fuel ◽  
Dose Rates ◽  
Dry Storage ◽  
Fuel Assemblies ◽  
Spent Fuel Pool ◽  
The Impact

Abstract In this paper shielding analysis was performed to determine neutron and gamma dose rates around the transfer cask HI-TRAC VW loaded with Spent Fuel Assemblies (SFA) from Nuclear Power Plant (NPP) Krsko Spent Fuel Dry Storage (SFDS) Campaign one. The HI-TRAC VW is a multi-layered cylindrical vessel designed to accept a Multi Purpose Canister (MPC) during loading, unloading and transfer to dry storage building. The MPC can contain up to 37 spent fuel assemblies. The analysis was divided into two steps. The first step was the source term generation using ORIGEN-S module of the SCALE code package. The source was calculated based on the operating history of spent fuel assemblies currently located in the NPP Krsko spent fuel pool. The obtained particle intensities and source spectra of the SFA were used in the second step to calculate the dose rates around the transfer cask. A comprehensive hybrid shielding analysis included the calculation of dose rates resulting from fuel neutrons and gammas, neutron induced gammas (n-g reaction), and hardware activation gammas under normal conditions and during accident scenario. To obtain the dose rates within the acceptable uncertainties, FW-CADIS variance reduction scheme, as implemented in ADVANTG code, was adopted for accelerating final MCNP6 calculations. The dose rates around HI-TRAC VW cask were calculated using MCNP6 code for all 16 casks loading belonging to Campaign one in order to illustrate the impact of fuel assembly selection schemes proposed by company responsible for project realization (Holtec International).

scholarly journals INTRODUCING BURNUP CREDIT FOR EXPANSION STAGE 2 OF THE EXTERNAL SPENT FUEL POOL AT GÖSGEN NPP

2021 ◽  
Vol 247 ◽  
pp. 17007
Author(s):  
Axel Hoefer ◽  
Martin Basler ◽  
Oliver Buss ◽  
Gaëtan Girardin ◽  
Fabian Jatuff ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Multiplication Factor ◽  
Spent Fuel ◽  
Statistical Confidence ◽  
Fuel Assemblies ◽  
Neutron Multiplication ◽  
Expansion Stage ◽  
Spent Fuel Pool ◽  
Storage Racks ◽  
Criticality Safety

We present a summary of the actinide-plus-fission-product burnup credit criticality safety licensing analysis for Expansion Stage 2 (ES2) of the external spent fuel pool at Gösgen nuclear power plant. In ES2, the nine Expansion Stage 1 storage racks currently installed in the external spent fuel pool are going to be supplemented by nine ES2 storage racks with a significantly reduced fuel assembly pitch. They are designed for loadings with fuel assemblies above a well-defined minimum required burnup. The objective of the criticality safety analysis is to calculate the minimum required burnup values for the uranium and MOX fuel assemblies to be stored in the ES2 storage racks. We use a methodology that allows us to take into account the reactivity effects due to variabilities and uncertainties of all relevant parameters involved in a burnup credit criticality safety assessment in a bounding manner. These include manufacturing tolerances of the fuel assemblies and storage racks, the irradiation histories and burnup profiles of the spent fuel assemblies, the bias of the depletion code used to calculate the isotopic inventories of the irradiated fuel, and the bias of the criticality code used to calculate the neutron multiplication factor of the considered storage configuration. A combination of different statistical procedures is used to evaluate and propagate the uncertainty information on the input parameters and translate it into statistical confidence statements about the neutron multiplication factor. It should be noted that the presented analysis is related to the first implementation of a significant burnup credit for wet storage of PWR fuel in Switzerland.

FUEL BURNUP AND FUEL POOL SHIELDING ANALYSIS FOR BUSHEHR NUCLEAR REACTOR VVER-1000

2006 ◽  
Vol 15 (04) ◽  
pp. 925-938 ◽  
Author(s):  
KAMAL HADAD ◽  
NAVID AYOBIAN
Keyword(s):  
Nuclear Power ◽  
Nuclear Reactor ◽  
Exposure Dose ◽  
Spent Fuel ◽  
Concrete Wall ◽  
Transport Code ◽  
Fuel Assemblies ◽  
Spent Fuel Pool ◽  
Γ Rays ◽  
Under Construction

Bushehr Nuclear power plant (BNPP) is currently under construction. The VVER-1000 reactor will be loaded with 126 tons of about 4% enriched fuel having 3-years life cycle. The spent fuel (SF) will be transferred into the spent fuel pool (SPF), where it stays for 8 years before being transferred to Russia. The SPF plays a crucial role during 8 years when the SP resides in there. This paper investigates the shielding of this structure as it is designed to shield the SF radiation. In this study, the SF isotope inventory, for different cycles and with different burnups, was calculated using WIMS/4D transport code. Using MCNP4C nuclear code, the intensity of γ rays was obtained in different layers of SFP shields. These layers include the water above fuel assemblies (FA) in pool, concrete wall of the pool and water laid above transferring fuels. Results show that γ rays leakage from the shield in the mentioned layers are in agreement with the plant's PSAR data. Finally we analyzed an accident were the water height above the FA in the pool drops to 47 cm. In this case it was observed that exposure dose above pool, 10 and 30 days from the accident, are still high and in the levels of 1000 and 758 R/hr.

Healing Fukushima: Radiation hazards and disaster medicine in post-3.11 Japan

2019 ◽  
Vol 49 (3) ◽  
pp. 333-354 ◽  
Author(s):  
Shi Lin Loh ◽  
Sulfikar Amir
Keyword(s):  
Public Health ◽  
Nuclear Power ◽  
Radiation Risk ◽  
Disaster Medicine ◽  
Radiation Hazards ◽  
Medical Practitioners ◽  
Life And Death ◽  
Nuclear Disaster ◽  
Fukushima Nuclear Disaster

What happens when expertise is forced to face disasters of unprecedented scales? How is knowledge produced in critical moments when every action and decision is a matter of life and death? And how are local social networks mobilized to cope with unforeseen crisis? This paper addresses these questions by examining the emergence of disaster medicine expertise in the aftermath of Fukushima nuclear disaster that struck Japan in 2011. Studies on Fukushima’s impact have to date revolved around the suffering of Tōhoku citizens and the development of Japan’s nuclear energy industry. Acknowledging the gravity of such work, this paper offers an alternative, but equally crucial angle on the disaster: that of the medical caregiving and public health system built in response to radiation hazards resulting from the triple meltdowns at Fukushima Daiichi nuclear power station. Through detailed interviews conducted with eight medical practitioners in Fukushima Prefecture, this paper analyses the significance and impact of Japan’s most recent radiation disaster on its public health infrastructure. To describe the contingent nature of radiation disaster medicine developed in response to radiation risk in Fukushima, we draw on Jasanoff’s characterization of scientific knowledge as ‘serviceable truths’ with regards to public policy and the law, suggesting that expertise in relation to disasters is usefully understood in analogous terms.

scholarly journals Microbial enrichment and gene functional categories revealed on the walls of a spent fuel pool of a nuclear power plant

PLoS ONE ◽  
2018 ◽  
Vol 13 (10) ◽  
pp. e0205228 ◽  
Author(s):  
Rosane Silva ◽  
Darcy Muniz de Almeida ◽  
Bianca Catarina Azeredo Cabral ◽  
Victor Hugo Giordano Dias ◽  
Isadora Cristina de Toledo e Mello ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Spent Fuel ◽  
Functional Categories ◽  
Spent Fuel Pool

scholarly journals Criticality safety analysis of spent fuel pool for a PWR using UO2, MOX, (Th-U)O2 and (TRU-Th)O2 fuels

2019 ◽  
Vol 7 (3A) ◽  
Author(s):  
Claubia Pereira ◽  
Jéssica P. Achilles ◽  
Fabiano Cardoso ◽  
Victor F. Castro ◽  
Maria Auxiliadora F. Veloso
Keyword(s):  
Normal Operation ◽  
Pressurized Water Reactor ◽  
Spent Fuel ◽  
Pressurized Water ◽  
Operation Conditions ◽  
Transport Code ◽  
Uranium Fuel ◽  
Fuel Assemblies ◽  
Spent Fuel Pool ◽  
Criticality Safety

A spent fuel pool of a typical Pressurized Water Reactor (PWR) was evaluated for criticality studies when it uses reprocessed fuels. PWR nuclear fuel assemblies with four types of fuels were considered: standard PWR fuel, MOX fuel, thorium-uranium fuel and reprocessed transuranic fuel spiked with thorium. The MOX and UO2 benchmark model was evaluated using SCALE 6.0 code with KENO-V transport code and then, adopted as a reference for other fuels compositions. The four fuel assemblies were submitted to irradiation at normal operation conditions. The burnup calculations were obtained using the TRITON sequence in the SCALE 6.0 code package. The fuel assemblies modeled use a benchmark 17x17 PWR fuel assembly dimensions. After irradiation, the fuels were inserted in the pool. The criticality safety limits were performed using the KENO-V transport code in the CSAS5 sequence. It was shown that mixing a quarter of reprocessed fuel withUO2 fuel in the pool, it would not need to be resized 

Application of TRACE and FRAPTRAN in the Spent Fuel Pool of Chinshan Nuclear Power Plant

2013 ◽  
Vol 479-480 ◽  
pp. 543-547
Author(s):  
Jong Rong Wang ◽  
Hao Tzu Lin ◽  
Wan Yun Li ◽  
Shao Wen Chen ◽  
Chun Kuan Shih
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Safety Issue ◽  
Water Injection ◽  
Safety Analysis ◽  
Spent Fuel ◽  
Spent Fuel Pool ◽  
Study Case

In the nuclear power plant (NPP) safety, the safety analysis of the NPP is very important work. In Fukushima NPP event, due to the earthquake and tsunami, the cooling system of the spent fuel pool failed and the safety issue of the spent fuel pool generated. In this study, the safety analysis of the Chinshan NPP spent fuel pool was performed by using TRACE and FRAPTRAN, which also assumed the cooling system of the spent fuel pool failed. There are two cases considered in this study. Case 1 is the no fire water injection in the spent fuel pool. Case 2 is the fire water injection while the water level of the spent fuel pool uncover the length of fuel rods over 1/3 full length. The analysis results of the case 1 show that the failure of cladding occurs in about 3.6 day. However, the results of case 2 indicate that the integrity of cladding is kept after the fire water injection.

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