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).

Model of a Dry Storage Facility for a Medium Nuclear Power Plant

2001 ◽  
Author(s):  
Mile Bace ◽  
Kresimir Trontl ◽  
Dubravko Pevec
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Radiation Dose ◽  
Dose Rate ◽  
Nuclear Power ◽  
Storage Facility ◽  
Spent Fuel ◽  
Radiation Dose Rate ◽  
Dry Storage ◽  
Fuel Assemblies

Abstract The intention was to model a dry storage facility that could satisfy the needs of a medium nuclear power plant similar to the NPP Krsko. The attention has been focused on radiation dose rate analyses and criticality calculations. Using the SCALE 4.4 code package and modified QAD-CGGP code, we modeled a facility that satisfies the basic criteria for public radiation protection. The capacity of the storage is 1,400 spent fuel assemblies which is adequate for a forty years medium NPP lifetime.

Nuclear Criticality Analyses of the Spent Fuel Pool Under Loss of Spent Fuel Pool Water and Neutron Absorbers in the Racks for Taipower’s Chinshan Nuclear Power Plant

2013 ◽  
Author(s):  
Weng-Sheng Kuo
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Neutron Transport ◽  
Spent Fuel ◽  
Pool Water ◽  
Regulatory Bodies ◽  
Spent Fuel Pool ◽  
The Impact ◽  
Criticality Accidents

The nuclear criticality analyses of the spent fuel pool under the postulated conditions of loss of spent fuel pool water and loss of neutron absorbers in the spent fuel racks, for Taipower’s Chinshan Nuclear Power Plant, were performed primarily using the Monte Carlo program MCNP5 in association with the deterministic neutron transport code CASMO-4. The results of these analyses can be used to help understand the impact of these beyond-design-basis accidents to the nuclear criticality, as well as facilitate nuclear utilities and regulatory bodies to develop the safety measures and regulations needed to prevent the criticality accidents.

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

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.

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

2011 ◽  
Vol 145 ◽  
pp. 78-82 ◽  
Author(s):  
Jong Rong Wang ◽  
Hao Tzu Lin ◽  
Yung Shin Tseng ◽  
Chun Kuan Shih
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Safety Issue ◽  
Safety Analysis ◽  
Spent Fuel ◽  
Energy Research ◽  
Important Work ◽  
Spent Fuel Pool

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, the cooling system of the spent fuel pool failed and the safety issue of the spent fuel pool generated. After Fukushima NPP event, INER (Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.) performed the safety analysis of the spent fuel pool for Chinshan NPP which also assumed the cooling system of the spent fuel pool failed. The geometry of the Chinshan NPP spent fuel pool is 12.17 m × 7.87 m × 11.61 m and the initial condition is 60 ¢J / 1.013 × 105 Pa. In general, the NPP safety analysis is performed by the thermal hydraulic codes. The advanced thermal hydraulic code named TRACE for the NPP safety analysis is developing by U.S. NRC. Therefore, the safety analysis of the spent fuel pool for Chinshan NPP is performed by TRACE. Besides, this safety analysis is also performed by CFD. The analysis result of TRACE and CFD are similar. The results show that the uncovered of the fuels occur in 2.7 days and the metal-water reaction of the fuels occur in 3.5 days after the cooling system failed.

scholarly journals Xenon-133 and caesium-137 releases into the atmosphere from the Fukushima Dai-ichi nuclear power plant: determination of the source term, atmospheric dispersion, and deposition

2011 ◽  
Vol 11 (10) ◽  
pp. 28319-28394 ◽  
Author(s):  
A. Stohl ◽  
P. Seibert ◽  
G. Wotawa ◽  
D. Arnold ◽  
J. F. Burkhart ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
North America ◽  
Northern Hemisphere ◽  
Nuclear Power ◽  
Noble Gas ◽  
Measurement Data ◽  
Spent Fuel ◽  
Spent Fuel Pool ◽  
Xenon 133

Abstract. On 11 March 2011, an earthquake occurred about 130 km off the Pacific coast of Japan's main island Honshu, followed by a large tsunami. The resulting loss of electric power at the Fukushima Dai-ichi nuclear power plant (FD-NPP) developed into a disaster causing massive release of radioactivity into the atmosphere. In this study, we determine the emissions of two isotopes, the noble gas xenon-133 (133Xe) and the aerosol-bound caesium-137 (137Cs), which have very different release characteristics as well as behavior in the atmosphere. To determine radionuclide emissions as a function of height and time until 20 April, we made a first guess of release rates based on fuel inventories and documented accident events at the site. This first guess was subsequently improved by inverse modeling, which combined the first guess with the results of an atmospheric transport model, FLEXPART, and measurement data from several dozen stations in Japan, North America and other regions. We used both atmospheric activity concentration measurements as well as, for 137Cs, measurements of bulk deposition. Regarding 133Xe, we find a total release of 16.7 (uncertainty range 13.4–20.0) EBq, which is the largest radioactive noble gas release in history not associated with nuclear bomb testing. There is strong evidence that the first strong 133Xe release started very early, possibly immediately after the earthquake and the emergency shutdown on 11 March at 06:00 UTC. The entire noble gas inventory of reactor units 1–3 was set free into the atmosphere between 11 and 15 March 2011. For 137Cs, the inversion results give a total emission of 35.8 (23.3–50.1) PBq, or about 42% of the estimated Chernobyl emission. Our results indicate that 137Cs emissions peaked on 14–15 March but were generally high from 12 until 19 March, when they suddenly dropped by orders of magnitude exactly when spraying of water on the spent-fuel pool of unit 4 started. This indicates that emissions were not only coming from the damaged reactor cores, but also from the spent-fuel pool of unit 4 and confirms that the spraying was an effective countermeasure. We also explore the main dispersion and deposition patterns of the radioactive cloud, both regionally for Japan as well as for the entire Northern Hemisphere. While at first sight it seemed fortunate that westerly winds prevailed most of the time during the accident, a different picture emerges from our detailed analysis. Exactly during and following the period of the strongest 137Cs emissions on 14 and 15 March as well as after another period with strong emissions on 19 March, the radioactive plume was advected over Eastern Honshu Island, where precipitation deposited a large fraction of 137Cs on land surfaces. The plume was also dispersed quickly over the entire Northern Hemisphere, first reaching North America on 15 March and Europe on 22 March. In general, simulated and observed concentrations of 133Xe and 137Cs both at Japanese as well as at remote sites were in good quantitative agreement with each other. Altogether, we estimate that 6.4 TBq of 137Cs, or 19% of the total fallout until 20 April, were deposited over Japanese land areas, while most of the rest fell over the North Pacific Ocean. Only 0.7 TBq, or 2% of the total fallout were deposited on land areas other than Japan.

Update on Jose´ Cabrera NPP Decommissioning

2011 ◽  
Author(s):  
Nieves Marti´n ◽  
Manuel Rodri´guez
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Storage Facility ◽  
Spent Fuel ◽  
Nuclear Facilities ◽  
Material Management ◽  
Primary Loop ◽  
Dry Storage

ENRESA is the National Spanish Agency responsible of the dismantling of Nuclear Facilities, previous Transfer of ownership of the facility from the Utility to ENRESA. On April 30th 2006, Jose´ Cabrera Nuclear Power Plant (Fig. 1) was definitively shutdown, and two years later, on April 30th 2008, ENRESA requested the transfer of the ownership of the Plant from the Ministry along with the corresponding authorization for performance of the Dismantling and Decommissioning Plan. On February 1st 2010, ENRESA was authorized to initiate the dismantling of Jose´ Cabrera NPP, once the spent fuel has been stored on-site at a dry storage facility (ISFSI). Currently, preparatory activities are underway, including the modification of systems and auxiliary facilities for waste and material management. Main challenges of the project include the removal of major components (vessel, steam generator, pressurizer, main pump and primary loop), and the use of large containers (CE-2b) to reduce segmentation of activated parts.

Method for identifying 17-place casings with spent fuel assemblies of AMB reactors at the Beloyarskaya nuclear power plant

Atomic Energy ◽  
2007 ◽  
Vol 102 (6) ◽  
pp. 452-457 ◽  
Author(s):  
N. V. Gorin ◽  
Ya. Z. Kandiev ◽  
E. N. Lipilina ◽  
G. V. Rukavishnikov ◽  
Yu. I. Churikov ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Spent Fuel ◽  
Fuel Assemblies

Study on Water Hammer in the Overpressure Discharge Pipe of Nuclear Power Plant

2017 ◽  
Author(s):  
Qingmu Xu ◽  
Kun Cai ◽  
Jie Qin ◽  
Junkai Yuan ◽  
Juan Li
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Water Hammer ◽  
Liquid Column ◽  
Normal Operation ◽  
Pipeline System ◽  
Spent Fuel ◽  
Vapor Cavity ◽  
Spent Fuel Pool

Water hammer phenomenon is a significant pressure wave in pipe system caused by momentum change when the moving fluid is forced to stop or change direction instantaneously. Common causes of water hammer are sudden valve closing at the end of a pipeline system, pump failure, check valve slam etc. The steam transportation pipeline system may also be vulnerable to water hammer when it confronts with the situation where liquid and steam co-exist. Water hammer often occurs when steam condenses into water in a horizontal section of steam piping. Then steam “picks up” water to form a high-velocity “slug” and create extra stress to pipe. When steam is trapped into sub-cooled water, the collapse of vapor cavity can lead to collision of two columns of liquid, resulting in a large rise in pressure which will damage pipes, supporting structures and hydraulic machinery. Nuclear power plant is composed of complex equipments and piping systems, lots of which contain both liquid and steam. Hence, there is a potential threat of occurrence of water hammer to the normal operation of systems. Thus, this phenomenon needs to be well investigated and prevented with some effective methods. For the purpose of overpressure relief under severe accidents, the spent fuel pool cooling system of CAP1000 series nuclear power plant provides a discharge passage from containment to spent fuel pool. When the containment pressure exceeds the control value, valve is opened to discharge high-temperature and high-pressure steam until the pressure drops to a safety value. During this process, serious water hammer happens, causing pressure rise beyond the design pressure and further leading to damages to pipes and structures. Therefore, water hammer of overpressure discharge pipeline in CAP1000 plant is studied in this work. On the basis of verification of the capabilities of computational code RELAP5/MOD3.3, hydraulic transient of water hammer is simulated under different conditions. It is indicated that after steam discharge stops, residual steam in pipe condenses because of contact with sub-cooled water in spent fuel pool. Subsequently, the rapid backflow and vapor cavity lead to a severe water hammer. The detailed analysis has shown that water temperature of spent fuel pool has a decisive influence on the mechanism of water hammer phenomenon, including collision of liquid column to valve disc and cavity collapse in the horizontal pipe. The collision and separation of liquid column result in relatively lower pressure amplitude.

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