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

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.

scholarly journals Understanding the likelihood and consequences of post-closure criticality in a geological disposal facility

2015 ◽  
Vol 79 (6) ◽  
pp. 1551-1561 ◽  
Author(s):  
R. J. Winsley ◽  
T. D. Baldwin ◽  
T. W. Hicks ◽  
R. M. Mason ◽  
P. N. Smith
Keyword(s):  
Spent Fuel ◽  
Geological Disposal ◽  
Safe Disposal ◽  
Waste Package ◽  
Disposal Facility ◽  
Safety Case ◽  
Operational Phase ◽  
Strength Rock ◽  
Criticality Safety

AbstractA geological disposal facility (GDF) will include fissile materials that could, under certain conditions, lead to criticality. Demonstration of criticality safety therefore forms an important part of a GDF's safety case.Containment provided by the waste package will contribute to criticality safety during package transport and the GDF operational phase. The GDF multiple-barrier system will ensure that criticality is prevented for some time after facility closure. However, on longer post-closure timescales, conditions in the GDF will evolve and it is necessary to demonstrate: an understanding of the conditions under which criticality could occur; the likelihood of such conditions occurring; and the consequences of criticality should it occur.Work has addressed disposal of all of the UK's higher-activity wastes in three illustrative geologies. This paper, however, focuses on presenting results to support safe disposal of spent fuel, plutonium and highlyenriched uranium in higher-strength rock.The results support a safety case assertion that post-closure criticality is of low likelihood and, if it was to occur, the consequences would be tolerable.

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.

Sign in / Sign up

Export Citation Format

Share Document