Criticality Safety Assessment of a TRIGA Reactor Spent-Fuel Pool Under Accident Conditions

1997 ◽  
Vol 117 (2) ◽  
pp. 248-254 ◽  
Author(s):  
Bogdan Glumac ◽  
Matjaž Ravnik ◽  
Marjan Logar
Keyword(s):  
Safety Assessment ◽  
Spent Fuel ◽  
Triga Reactor ◽  
Spent Fuel Pool ◽  
Accident Conditions ◽  
Criticality Safety

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 

Evaluation of Alternate Water Injection Cooling for Accident Conditions in Spent Fuel Pool Using MAAP Code

2019 ◽  
Vol 2019 (0) ◽  
pp. S08106
Author(s):  
Satoshi NISHIMURA ◽  
Masaaki SATAKE ◽  
Yoshihisa NISHI ◽  
Yoshiyuki KAJI ◽  
Yoshiyuki NEMOTO
Keyword(s):  
Water Injection ◽  
Spent Fuel ◽  
Spent Fuel Pool ◽  
Accident Conditions

The Effect of Pitch, Burnup, and Absorbers on a TRIGA Spent-Fuel Pool Criticality Safety

2003 ◽  
Vol 141 (2) ◽  
pp. 211-219 ◽  
Author(s):  
Marjan Logar ◽  
Robert Jeraj ◽  
Bogdan Glumac
Keyword(s):  
Spent Fuel ◽  
Spent Fuel Pool ◽  
Criticality Safety

TRIGA reactor spent fuel pool under severe earthquake conditions

1998 ◽  
Vol 183 (3) ◽  
pp. 261-267 ◽  
Author(s):  
Marjan Logar ◽  
Bogdan Glumac ◽  
Marko Maučec
Keyword(s):  
Spent Fuel ◽  
Triga Reactor ◽  
Spent Fuel Pool

The Importance of Reliable Indications for Accident Management

2016 ◽  
Author(s):  
Robert J. Lutz ◽  
Bill T. Williamson
Keyword(s):  
Severe Accident ◽  
Spent Fuel ◽  
Critical Importance ◽  
Dc Power ◽  
The Status ◽  
Spent Fuel Pool ◽  
Accident Conditions ◽  
And Control ◽  
Parameter Values ◽  
Support Guidelines

The accidents at the Three Mile Island Unit 2 (TMI-2) and Fukushima Daiichi Units 1, 2, and 3 demonstrate the critical importance of accurate, relevant, and timely information on the status of reactor systems during a severe accident. There is evidence that the failure of key instrumentation to provide reliable information to the control room licensed operators contributed to the severity of the accident at both TMI and Fuskushima Daiichi. These events also highlight the critical importance of understanding and focusing on the key elements of system status information in an environment where operators may be overwhelmed with superfluous and sometimes conflicting data and yet have to make urgent decisions. While progress in these areas has been made since TMI-2, the accident at Fukushima suggests there may still be some potential for further improvement in critical plant instrumentation. As a result, several approaches are being employed to provide better information to emergency response personnel during a severe accident. The first approach being taken by the PWROG and BWROG is the identification of methods to obtain information related to key plant parameters when there is a loss of dc power for instrumentation and control. The FLEX guidance in NEI 12-06 requires that reliable instrumentation be available to ensure core, containment and spent fuel pool cooling is maintained for the beyond design basis events for which FLEX was intended. For the most part, this instrumentation that is important for FLEX is the same instrumentation that is used for diagnosis of severe accident conditions and challenges to fission product barriers. Generic FLEX Support Guidelines have been developed to provide a uniform basis for plants to meet the NEI 12-06 requirements that includes methods to obtain key parameter values in the event of a loss of all dc instrument power. The PWROG and the BWROG have also taken a complimentary approach to provide Technical Support Guidance (TSG) for instrumentation during a severe accident. This approach identifies the primary instrumentation as well as alternate instrumentation and other tools to validate the indications from the primary instrumentation. The validation consists of: a) comparing the primary instrument indications to the alternate instrumentation, b) comparing instrument indications to related instrumentation, c) comparing instrument indications and trends to expected trends based on the accident progression and actions already implemented, and d) comparing instrument indications to information in calculational aids.

Validation of UNIST Monte Carlo code MCS for criticality safety analysis of PWR spent fuel pool and storage cask

2018 ◽  
Vol 114 ◽  
pp. 495-509 ◽  
Author(s):  
Jaerim Jang ◽  
Wonkyeong Kim ◽  
Sanggeol Jeong ◽  
Eun Jeong ◽  
Jinsu Park ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Safety Analysis ◽  
Monte Carlo Code ◽  
Spent Fuel ◽  
Spent Fuel Pool ◽  
And Storage ◽  
Criticality Safety

Probabilistic Safety Assessment for CAP Nuclear Power Plant Spent Fuel Pool

2017 ◽  
Author(s):  
Xu Yiquan ◽  
Zhuo Yucheng ◽  
Yang Yajun ◽  
Fu Hao
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Safety Assessment ◽  
Spent Fuel ◽  
Pressurized Water ◽  
Probabilistic Safety Assessment ◽  
Core Damage ◽  
Spent Fuel Pool ◽  
Probabilistic Safety

For china advanced passive (CAP) pressurized water reactor (PWR) nuclear power plant, spent fuel damage risk potentially induced by internal events is assessed. Spent fuel damage frequency (FDF) is quantified by event tree and fault tree model using probabilistic safety assessment (PSA) software RiskSpectrum. For different operation conditions total FDF is 2.05×10−9 per reactor-year which is much lower than core damage frequency (about 2.41×10−7 per reactor-year). By assumption of completely radioactivity releasing, large radioactive release frequency caused by FDF is one order of magnitude lower than that caused by core damage (about 2.38×10−8 per reactor-year). Since the multiple prevention and mitigation measures in CAP PWR nuclear power plant responding to spent fuel pool accidents, the risk of spent fuel pool is much lower than that of reactor core, and the safety goals of nuclear safety guide can be satisfied.

Axial Burnup Profile Influence on Criticality Safety of WWER Spent Fuel

2006 ◽  
Author(s):  
Surik Bznuni ◽  
Armen Amirjanyan ◽  
Shahen Poghosyan
Keyword(s):  
Safety Assessment ◽  
Safety Analysis ◽  
Spent Fuel ◽  
End Effect ◽  
Oak Ridge ◽  
Fuel Assemblies ◽  
Spent Fuel Storage ◽  
Fuel Storage ◽  
Criticality Analysis ◽  
Criticality Safety

Criticality safety assessment for WWER-440 NUHOMS® cask with spent nuclear fuel from Armenian NPP has been performed. The cask was designed in a such way that the neutron multiplication factor keff must be below 0,95 for all operational modes and accident conditions. Usually for criticality analysis, fresh fuel approach with the highest enrichment is taken as conservative assumption as it was done for ANPP. Nuclear and Radiation Safety Centre of Armenian Nuclear Regulatory Authority (NRSC ANRA) in order to improve future fuel storage efficiency, initiated research with taking into account burn up credit in the criticality safety assessment. Axial burn up profile (end effect) has essential impact on criticality safety justification analysis. However this phenomenon wasn’t taken into account in the Safety Analysis Report of NUHOMS® spent fuel storage constructed on the site of ANPP. Although ANRA doesn’t yet accept burn up credit approach for ANPP spent fuel storage, assessment of impact of axial burn up profile on criticality of spent fuel assemblies has important value for future activities of ANRA. This paper presents results of criticality safety analysis of spent fuel assemblies with axial burn up profile. Horizontal burn up profile isn’t taken account since influence of the horizontal variation of the burn up is much less than the axial variation. The Actinides and Actinides + Fission Products approach are discussed. The calculations were carried out with STARBUCS module of SCALE 5.0 code package developed at Oak Ridge National laboratory. SCALE5.0 sequence CSAS26 (KENO-VI) was used for evaluation the keff for 3-D problems. Obtained results showed that criticality of ANPP spent fuel cask is very sensitive to the end effect. Using Burn up profiles of Control Assemblies in both approaches leads to much more increasing than in case of Working Assemblies. Usually increasing burn up leads to decreasing Δkeff, hence decreasing end effect. However for WWER-440 Control Assemblies that worked only within 6th (operative) group increasing burn up leads to increasing of the end effect.

Risk Analysis of Spent Fuel Pool Caused by Seismic Events

2018 ◽  
Vol 4 (4) ◽  
Author(s):  
Ming Wang ◽  
Modi Lin ◽  
Jinkai Wang
Keyword(s):  
Quantitative Analysis ◽  
Safety Assessment ◽  
Mitigation Measures ◽  
Seismic Events ◽  
Spent Fuel ◽  
Seismic Capacity ◽  
Fuel Assemblies ◽  
Spent Fuel Pool ◽  
The Common ◽  
External Events

Spent fuel pool (SFP) stores fuel assemblies removed from the reactor over the years. SFP and its accident mitigation measures may fail simultaneously at the time of the earthquake, which may cause serious accident consequences. This paper uses probabilistic safety assessment (PSA) method to quantitatively evaluate the risk of SFP for a CPR1000 unit caused by seismic events. Quantitative analysis results show that seismic events' risk is the highest in all internal events and external events for SFP. In order to reduce the risk of SFP, more attention should be paid to improve seismic capacity or reduce the common failure for systems and components associated with SFP under the earthquake situation.

Behavior of the Spent Fuel Pool and Dose Rate Calculations for the CPR1000 Reactor Under Severe Accident Conditions Using the Integral Code MELCOR

2013 ◽  
Author(s):  
Klaus Mueller ◽  
Moses Yeung ◽  
Justin Byard ◽  
Zhen Xun Peng ◽  
Jun Tao ◽  
...  
Keyword(s):  
Fission Product ◽  
Dose Rate ◽  
Flow Rate ◽  
Hydrogen Combustion ◽  
Severe Accident ◽  
Pure Hydrogen ◽  
Spent Fuel ◽  
Product Release ◽  
Spent Fuel Pool ◽  
Accident Conditions

The behavior of the spent fuel pool and the fission product release and transport for the CPR1000 reactor under severe accident conditions was analyzed using the integral severe accident code MELCOR. In the investigated accident scenario a total failure of the pump of the spent fuel cooling system was assumed. Furthermore, it is assumed that accident management fails to bring water into the spent fuel pool using mobile pumps or due to the non-recovery of the cooling pump. The grace time available for measures in order to avoid significant fission product release to the environment is determined. The calculated hydrogen mass flow rate due to clad oxidation and the steam flow rate from the spent fuel pool to the compartment above the spent fuel pool serve as boundary conditions for the three dimensional fluid dynamics code GASFLOW to assess possible hydrogen combustion or detonation in the compartment. Using this spent fuel pool MELCOR model the dose submerged in air or water can be determined. The calculated gamma dose rate in a specific compartment can be used for equipment qualification and compartment accessibility assessment. It was found that after four days the fuel assemblies are significantly heated-up and ten hours later the fission products are released as well as a significant amount of hydrogen is produced. A preliminary GASFLOW analysis shows by assuming an air atmosphere in the fuel building, that the risk of a hydrogen combustion or detonation is high. In late state of the accident a convection flow of pure hydrogen is established in spent fuel pool region. It was shown, that the flow conditions strongly influence the fission product transport behavior and consequently the dose rates in the compartment above the spent fuel pool.

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