A study of burnup credit in criticality safety analysis for PBR spent fuel pebbles

2019 ◽  
Vol 132 ◽  
pp. 347-356 ◽  
Author(s):  
Shang-Chien Wu ◽  
Der-Sheng Chao ◽  
Jenq-Horng Liang
Keyword(s):  
Safety Analysis ◽  
Spent Fuel ◽  
Criticality Safety

scholarly journals CRITICALITY SAFETY ANALYSIS OF GBC-32 SPENT FUEL CASK WITH IMPROVED NEUTRON ABSORBER CONCEPT

2021 ◽  
Vol 247 ◽  
pp. 17003
Author(s):  
Martin Lovecký ◽  
Jiří Závorka ◽  
Jana Jiřičková ◽  
Radek Škoda
Keyword(s):  
Fuel Assembly ◽  
Boron Content ◽  
High Efficiency ◽  
Safety Analysis ◽  
Spent Fuel ◽  
Neutron Absorber ◽  
Safety Margins ◽  
Absorber Material ◽  
Inner Diameter ◽  
Criticality Safety

Higher enrichment of nuclear fuel along the manufacturing limit of boron content in steel and aluminum alloys represents a significant challenge in designing spent fuel transport and storage facilities. One possible solution for spent fuel pools and casks is the burnup credit method that allows for decreasing very high safety margins associated with fresh fuel assumption in spent fuel facilities. An alternative solution based on placing neutron absorber material directly into the fuel assembly is proposed here. A neutron absorber permanently fixed in guide tubes decreases system reactivity more efficiently than absorber sheets between the assemblies. The efficiency of the newly proposed concept is demonstrated on the criticality safety analysis of the GBC-32 spent fuel cask. Absorber rods from 8 different elements are placed within Westinghouse OFA 17x17 guide tubes. Currently used boron is a good option because of high absorption cross section, low atomic mass and chemical compatibility with various alloys. Alternative options (e.g., Sm, Eu, Gd, Dy, Hf, Re, Ir) are based on very good absorbers that do not require alloy compatibility since the absorbers can be placed inside zirconium or steel cladding. Because of high efficiency of the newly proposed absorber concept, boron content in BORAL sheets can be decreased to more competitive economics. Moreover, fuel assembly pitch is investigated in order to change cask wall inner diameter that will result in lower material consumption for the cask wall with the same shielding thickness.

Shielding calculation and criticality safety analysis of spent fuel transportation cask in research reactors

2016 ◽  
Vol 108 ◽  
pp. 129-132 ◽  
Author(s):  
A. Mohammadi ◽  
M. Hassanzadeh ◽  
M. Gharib
Keyword(s):  
Safety Analysis ◽  
Spent Fuel ◽  
Research Reactors ◽  
Criticality Safety

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

Impact of control rod insertion during burnup on PWR fuel assembly isotopic composition

Kerntechnik ◽  
2021 ◽  
Vol 86 (2) ◽  
pp. 173-181
Author(s):  
R. M. Refeat ◽  
H. K. Louis
Keyword(s):  
Safety Analysis ◽  
Second Phase ◽  
Spent Fuel ◽  
Reactor Operation ◽  
Control Rod ◽  
Two Phases ◽  
And Control ◽  
Criticality Analysis ◽  
Control Rods ◽  
Criticality Safety

Abstract Criticality analysis of spent fuel assumes that the fuel material is unburned which means that it is in its most reactive condition. In fact, this is not the real situation for fuel as it is burned during reactor operation causing reduction in the reactivity. Considering the reduction in reactivity during spent fuel calculations is the Burn-up Credit concept (BUC). In addition, the control rods radial and axial positions have an effect on the reactivity which can be considered in the criticality safety analysis. This paper studies the effect of burnup and control rods (CRs) movement on reactivity and isotopes inventory. Calculations are carried out in two phases, first kinf is calculated for different burnup profiles with control rods are either fully withdrawn or fully inserted. In the second phase keff is calculated for different control rods insertion levels. For both phases, burnup calculations are performed for a UO2 assembly then multiplication factor calculations of burned UO2 assemblies in cold state are done. The burnup calculations are performed using MCNP6 code and ENDF/B-VII library for different burnup levels up to 45 GWd/tU. The results obtained can be taken in consideration in criticality safety analysis performed for the spent fuel to improve the economic efficiency for manufacture, storage and transportation of fissile materials.

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.

scholarly journals Nuclear criticality safety analysis of a spent fuel waste package in a tuff repository

10.2172/59297 ◽  
1983 ◽  
Author(s):  
B.H. Weren ◽  
M.A. Capo ◽  
W.C. O`Neal
Keyword(s):  
Safety Analysis ◽  
Spent Fuel ◽  
Waste Package ◽  
Nuclear Criticality Safety ◽  
Criticality Safety

scholarly journals Criticality safety analysis of spent fuel pool for TRIGA Puspati Reactor using MCNP5

2019 ◽  
Vol 555 ◽  
pp. 012001
Author(s):  
Nurlaila Syamsul Bahri ◽  
Jasman Zainal ◽  
Khaidzir Hamzah ◽  
Muneer Aziz Mohammad Saleh ◽  
Mohamad Hairie Rabir
Keyword(s):  
Safety Analysis ◽  
Spent Fuel ◽  
Spent Fuel Pool ◽  
Criticality Safety

scholarly journals CRITICALITY SAFETY ANALYSIS OF THE DRY CASK DESIGN WITH AIR GAPS FOR RDNK SPENT PEBBLE FUELS STORAGECriticality Safety Analysis of the Dry Cask Design with Air Gaps for RDNK Spent Pebble Fuels Storage

2021 ◽  
Vol 23 (3) ◽  
pp. 123
Author(s):  
Pungky Ayu Artiani ◽  
Yuli Purwanto ◽  
Aisyah Aisyah ◽  
Ratiko Ratiko ◽  
Jaka Rachmadetin ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Thermal Power ◽  
Heat Removal ◽  
Threshold Value ◽  
Safety Analysis ◽  
Spent Fuel ◽  
Air Gaps ◽  
Fuel Elements ◽  
Submerged Conditions ◽  
Criticality Safety

Reaktor Daya Non-Komersial (RDNK) with a 10 MW thermal power has been proposed as one of the technology options for the first nuclear power plant program in Indonesia. The reactor is a High Temperature Gas-Cooled Reactor-type with spherical fuel elements called pebbles. To support this program, it is necessary to prepare dry cask to safely store the spent pebble fuels that will be generated by the RDNK. The dry cask design has been proposed based on the Castor THTR/AVR but modified with air gaps to facilitate decay heat removal. The objective of this study is to evaluate criticality safety through keff  value of the proposed dry cask design for the RDNK spent fuel. The keff  values were calculated using MCNP5 program for the dry cask with 25, 50, 75, and 100% of canister capacity. The values were calculated for dry casks with and without air gaps in normal, submerged, tumbled, and both tumbled and submerged conditions. The results of calculated keff  values for the dry cask with air gaps at 100% of canister capacity from the former to the latter conditions were 0.127, 0.539, 0.123, and 0.539, respectively. These keff values were smaller than the criticality threshold value of 0.95. Therefore, it can be concluded that the dry cask with air gaps design comply the criticality safety criteria in the aforementioned conditions.

Sensitivity study on criticality safety analysis of multiple misloading within the spent fuel storage cask

2020 ◽  
Vol 144 ◽  
pp. 107516
Author(s):  
Saed Alrawash ◽  
Muth Boravy ◽  
Seung Uk Yoo ◽  
Hyuk Han ◽  
Soon Young Kim ◽  
...  
Keyword(s):  
Safety Analysis ◽  
Sensitivity Study ◽  
Spent Fuel ◽  
Spent Fuel Storage ◽  
Fuel Storage ◽  
Criticality Safety
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