scholarly journals VERIFICATION AND VALIDATION OF BACK-END CYCLE SOURCE TERM CALCULATION OF THE NODAL CODE RAST-K

2021 ◽  
Vol 247 ◽  
pp. 10025
Author(s):  
Jaerim Jang ◽  
Bamidele Ebiwonjumi ◽  
Wonkyeong Kim ◽  
Jinsu Park ◽  
Deokjung Lee
Keyword(s):  
Source Term ◽  
Spent Nuclear Fuel ◽  
Interpolation Method ◽  
Prediction Method ◽  
Computation Time ◽  
Verification And Validation ◽  
Fuel Temperature ◽  
Pressurized Water ◽  
Decay Heat ◽  
Fuel Assemblies

Verification and validation (V&V) results of source term calculation capability implemented in the nodal diffusion code RAST-K are presented in this paper. An isotope inventory prediction method is presented in this work which is implemented with RAST-K and the lattice code STREAM. STREAM generates cross-section and provides number density information by history branch calculations. RAST-K supplies the power history and three history indexes (boron concentration, moderator temperature and fuel temperature). The main feature of the newly implemented spent nuclear fuel (SNF) characterization is the direct consideration of three-dimensional (3D) core simulation conditions by using operation history information. As a result of this, it could reduce the computation time. The implemented SNF analysis capability have two main functions. The first is to predict isotope inventory by Lagrange non-linear interpolation method, using power history correction factors. The second is to calculate the radiological response activity, decay heat, and neutron/gamma source strengths. The V&V of these two functions are thus presented herein. The isotope inventory prediction is validated with measured data from ten SNF samples of Takahama-3 and six samples of Calvert Cliffs-1 pressurized water reactors (PWR). Eighteen decay heat measurements of Ringhals Unit 3 PWR fuel assemblies are then employed to validate the decay heat calculation results. In addition, STREAM is employed in a code-to-code comparison for verification. The fuel assemblies cover the burnup range 14.3 - 47.25 GWd/tU, initial enrichment of 2.1 - 4.11 235U w/o and cooling time of 3.96 to 20.01 years. The comparison to STREAM shows the accuracy of the RAST-K SNF and prediction of the decay heat is within 4%. Overall, this paper demonstrates that RAST-K SNF calculation can be applied to the back-end cycle source term analysis.

scholarly journals FEATURES OF THE SOURCE TERM CALCULATION CAPABILITY OF THE NECP-X CODE

2021 ◽  
Vol 247 ◽  
pp. 10024
Author(s):  
Xingjian Wen ◽  
Zhouyu Liu ◽  
Kai Huang ◽  
Liangzhi Cao
Keyword(s):  
Source Term ◽  
Spent Nuclear Fuel ◽  
High Fidelity ◽  
Memory Requirement ◽  
Compression Method ◽  
Heat Measurement ◽  
Neutron Spectra ◽  
Decay Heat ◽  
Heat Calculation ◽  
Full Activation

The source term calculation capability is developed for the high-fidelity neutronics code NECP-X. Generally, a full activation library is used, but the memory requirement is unacceptable for the high-fidelity calculation. In order to minimize the memory requirement during the calculation with very strict conditions, a new generalized activation chain compressed method is proposed based on the influence qualification. One basic compression element is a reaction channel or an isotope, and the influence of every compression element to the final results are qualified. To enlarge the range of application of the new compressed library, an effective method to determine representative problems, which utilizes the neutron spectra and neutron flux, is developed and analyzed. Based on the ENDF-VII.0, EAF-2010 evaluated nuclear library and the influence qualification-based activation library compression method, a new compressed activation library is generated. The VERA-3A problem and the KAIST problem are used to assess the accuracy and the efficiency of the new activation library. 85 measurements of decay heat from decay heat measurement facilities GE-Morris and CLAB are used to validate the decay heat calculation in NECP-X. The results show good accuracy of NECP-X in predicting radiation source term of the spent nuclear fuel and significant memory saving when using new compressed activation library.

scholarly journals Preliminary Assessment of Criticality Safety Constraints for Swiss Spent Nuclear Fuel Loading in Disposal Canisters

Materials ◽  
2019 ◽  
Vol 12 (3) ◽  
pp. 494
Author(s):  
Alexander Vasiliev ◽  
Jose Herrero ◽  
Marco Pecchia ◽  
Dimitri Rochman ◽  
Hakim Ferroukhi ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Operating Conditions ◽  
Design Parameters ◽  
Fuel Loading ◽  
Pressurized Water ◽  
Fuel Assemblies ◽  
National Cooperative ◽  
Criticality Safety

This paper presents preliminary criticality safety assessments performed by the Paul Scherrer Institute (PSI) in cooperation with the Swiss National Cooperative for the Disposal of Radioactive Waste (Nagra) for spent nuclear fuel disposal canisters loaded with Swiss Pressurized Water Reactor (PWR) UO2 spent fuel assemblies. The burnup credit application is examined with respect to both existing concepts: taking into account actinides only and taking into account actinides plus fission products. The criticality safety calculations are integrated with uncertainty quantifications that are as detailed as possible, accounting for the uncertainties in the nuclear data used, fuel assembly and disposal canister design parameters and operating conditions, as well as the radiation-induced changes in the fuel assembly geometry. Furthermore, the most penalising axial and radial burnup profiles and the most reactive fuel loading configuration for the canisters were taken into account accordingly. The results of the study are presented with the help of loading curves showing what minimum average fuel assembly burnup is required for the given initial fuel enrichment of fresh fuel assemblies to ensure that the effective neutron multiplication factor, keff, of the canister would comply with the imposed criticality safety criterion.

scholarly journals Simulation of the Thermal Conditions of Cask with Fuel Assemblies of BN-350 Reactor for Dry Storage

2019 ◽  
Vol 2019 ◽  
pp. 1-5
Author(s):  
Ye. T. Koyanbayev ◽  
M. K. Skakov ◽  
D. A. Ganovichev ◽  
Ye. A. Martynenko ◽  
A. A. Sitnikov
Keyword(s):  
Thermal Condition ◽  
Spent Nuclear Fuel ◽  
Thermal Conditions ◽  
Structural Elements ◽  
Fuel Temperature ◽  
Ansys Software ◽  
Dry Storage ◽  
Fuel Assemblies ◽  
Fuel Storage ◽  
Thermal Calculations

The analysis of the thermal condition of spent FA (fuel assembly) of BN-350 reactor in a six-place cask for dry storage is presented. Simulation of the thermal condition of the cask is conducted with finite elements method using ANSYS software. Calculations of fuel temperature, fuel cladding, and assembly structural elements are the part of the safety analysis for storage of spent FA. In conclusion, the results of the thermal calculations in the cases of filling cask with argon and atmospheric air are given when the thickness of the insulation cask with concrete is 0.5 and 1 m. As a result of the calculated studies, the parameters of SNF (spent nuclear fuel) storage are determined, under which the fuel temperatures will have minimum and maximum values.

The Effect of Serpent 2 Calculation Parameters on Evaluated Spent Nuclear Fuel Source Term

2021 ◽  
Author(s):  
Riku Tuominen ◽  
Ville Valtavirta
Keyword(s):  
Source Term ◽  
Spent Nuclear Fuel ◽  
Photon Emission ◽  
Relative Difference ◽  
Fission Rate ◽  
Decay Heat ◽  
Probability Table ◽  
Zone Division ◽  
Energy Dependent ◽  
Fuel Source

Abstract The estimation of spent nuclear fuel source term (decay heat, reactivity, nuclide inventory etc.) has several sources of uncertainty such as uncertainties in nuclear data, uncertainties in the operation history, choice of calculation parameters etc. In this work the effect of calculation parameters is studied by estimating the source term with the built-in burnup capability of Serpent. The effect of the following parameters is considered: depletion zone division, burnup steps, unresolved resonance probability table sampling, Doppler-Broadening Rejection Correction (DBRC) and energy dependent branching ratios. As a test case a 2D BWR fuel assembly was modelled by first running a burnup calculation followed by a decay calculation. The following source term components were considered when investigating the effect of the studied parameters: total decay heat, photon emission rate and spontaneous fission rate. In general the differences resulting from the use of different parameter variations were small for all three studied source term components. For the decay heat largest absolute relative difference was approximately 0.6 % and for the photon emission rate approximately 1.1 %. For the spontaneous fission rate maximum absolute relative difference of nearly 8 % was observed. For all three components the variation of the depletion zone division resulted in the largest relative differences. Clear differences were also observed for burnup step length and DBRC variations. The use of unresolved resonance probability table sampling and energy dependent branching ratios had an insignificant effect on the studied source term components.

Spent Fuel Pool Accidents in a Nordic BWR

2013 ◽  
Author(s):  
Niina E. Könönen
Keyword(s):  
Natural Circulation ◽  
Water Injection ◽  
Heat Power ◽  
Spent Fuel ◽  
Fuel Temperature ◽  
Fuel Rods ◽  
Decay Heat ◽  
Fuel Assemblies ◽  
Loss Of Coolant ◽  
Spent Fuel Pool

Loss-of-pool-cooling accidents at the spent fuel pools in the reactor hall of a Nordic BWR have been studied using the MELCOR 1.8.6 code. Studies were made with several different calculation nodalizations. Other investigated variables were the total decay heat power of fuel assemblies in the pool, the initiator of the accident (loss of pool cooling or loss of coolant from the pool), the LOCA leak elevation, the alignment of the re-flooding injection and the use of a lid on top of the pool. From the results it was observed that ensuring natural circulation of air in the fuel is essential in preventing fuel damages. In cases where the air flow is prevented (loss of pool cooling, LOCA break elevation above the bottom of fuel assemblies or a lid on top of the pool) the fuel is damaged in nearly all cases. Only with the pool decay power being sufficiently low (2227 kW) the melting of the fuel was prevented. If water injection to the SFP can be restarted before the fuel temperature reaches the cladding failure criterion (1173 K), the rewetting of fuel assemblies is successful and the fuel temperatures will quickly decrease and the assemblies will remain intact. The re-flooding of already damaged fuel assemblies will also cool the fuel rods and reduce the radioactive releases to the environment. However, it may result in greater hydrogen releases.

Advanced First Core Design for the Westinghouse AP1000

2009 ◽  
Author(s):  
Robert J. Fetterman
Keyword(s):  
Fuel Cycle ◽  
The United States ◽  
Pressurized Water ◽  
Loading Pattern ◽  
The Core ◽  
Low Leakage ◽  
Fuel Assemblies ◽  
Cycle Loading ◽  
Water Reactors ◽  
Under Construction

As the nuclear renaissance is now upon us and new plants are either under construction or being ordered, a considerable amount of attention has also turned to the design of the first fuel cycle. Requirements for core designs originate in the Utilities Requirements Document (URD) for the United States and the European Utilities Requirements (EUR) for Europe. First core designs created during the development of these documents were based on core design technology dating back to the 1970’s, where the first cycle core loading pattern placed the highest enrichment fuel on the core periphery and two other lower enrichments in the core interior. While this sort of core design provided acceptable performance, it underutilized the higher enriched fuel assemblies and tended to make transition to the first reload cycle challenging, especially considering that reload core designs are now almost entirely of the Low Leakage Loading Pattern (LLLP) design. The demands placed on today’s existing fleet of pressurized water reactors for improved fuel performance and economy are also desired for the upcoming Generation III+ fleet of plants. As a result of these demands, Westinghouse has developed an Advanced First Core (AFCPP) design for the initial cycle loading pattern. This loading pattern design simulates the reactivity distribution of an 18 month low leakage reload cycle design by placing the higher enriched assemblies in the core interior which results in improved uranium utilization for those fuel assemblies carried through the first and second reload cycles. Another feature of the advanced first core design is radial zoning of the high enriched assemblies, which allows these assemblies to be located in the core interior while still maintaining margin to peaking factor limits throughout the cycle. Finally, the advanced first core loading pattern also employs a variety of burnable absorber designs and lengths to yield radial and axial power distributions very similar to those found in typical low leakage reload cycle designs. This paper will describe each of these key features and demonstrate the operating margins of the AFC design and the ability of the AFC design to allow easy transition into 18 month low leakage reload cycles. The fuel economics of the AFC design will also be compared to those of a more traditional first core loading pattern.

scholarly journals Nuclear Data Uncertainty Quantification in Criticality Safety Evaluations for Spent Nuclear Fuel Geological Disposal

2021 ◽  
Vol 11 (14) ◽  
pp. 6499
Author(s):  
Matthias Frankl ◽  
Mathieu Hursin ◽  
Dimitri Rochman ◽  
Alexander Vasiliev ◽  
Hakim Ferroukhi
Keyword(s):  
Uncertainty Quantification ◽  
Nuclear Fuel ◽  
Evaluation Method ◽  
Spent Nuclear Fuel ◽  
Nuclear Data ◽  
Data Uncertainty ◽  
Evaluation Methodology ◽  
Geological Disposal ◽  
Fuel Assemblies ◽  
Criticality Safety

Presently, a criticality safety evaluation methodology for the final geological disposal of Swiss spent nuclear fuel is under development at the Paul Scherrer Institute in collaboration with the Swiss National Technical Competence Centre in the field of deep geological disposal of radioactive waste. This method in essence pursues a best estimate plus uncertainty approach and includes burnup credit. Burnup credit is applied by means of a computational scheme called BUCSS-R (Burnup Credit System for the Swiss Reactors–Repository case) which is complemented by the quantification of uncertainties from various sources. BUCSS-R consists in depletion, decay and criticality calculations with CASMO5, SERPENT2 and MCNP6, respectively, determining the keff eigenvalues of the disposal canister loaded with the Swiss spent nuclear fuel assemblies. However, the depletion calculation in the first and the criticality calculation in the third step, in particular, are subject to uncertainties in the nuclear data input. In previous studies, the effects of these nuclear data-related uncertainties on obtained keff values, stemming from each of the two steps, have been quantified independently. Both contributions to the overall uncertainty in the calculated keff values have, therefore, been considered as fully correlated leading to an overly conservative estimation of total uncertainties. This study presents a consistent approach eliminating the need to assume and take into account unrealistically strong correlations in the keff results. The nuclear data uncertainty quantification for both depletion and criticality calculation is now performed at once using one and the same set of perturbation factors for uncertainty propagation through the corresponding calculation steps of the evaluation method. The present results reveal the overestimation of nuclear data-related uncertainties by the previous approach, in particular for spent nuclear fuel with a high burn-up, and underline the importance of consistent nuclear data uncertainty quantification methods. However, only canister loadings with UO2 fuel assemblies are considered, not offering insights into potentially different trends in nuclear data-related uncertainties for mixed oxide fuel assemblies.

Development of a Thermal Analysis Model for a Nuclear Spent Fuel Storage Cask and Experimental Verification With Prototype Testing

1989 ◽  
Vol 111 (4) ◽  
pp. 647-651 ◽  
Author(s):  
J. Y. Hwang ◽  
L. E. Efferding
Keyword(s):  
Thermal Analysis ◽  
Storage Period ◽  
Department Of Energy ◽  
Pressurized Water Reactor ◽  
Spent Fuel ◽  
Analysis Model ◽  
Pressurized Water ◽  
Fuel Assemblies ◽  
Spent Fuel Storage ◽  
Fuel Storage

A thermal analysis evaluation is presented of a nuclear spent fuel dry storage cask designed by the Westinghouse Nuclear Components Division. The cask is designed to provide passive cooling of 24 Pressurized Water Reactor (PWR) spent fuel assemblies for a storage period of at least 20 years at a nuclear utility site (Independent Spent Fuel Storage Installation). A comparison is presented between analytical predictions and experimental results for a demonstration cask built by Westinghouse and tested under a joint program with the Department of Energy and Virginia Power Company. Demonstration testing with nuclear spent fuel assemblies was performed on a cask configuration designed to store 24 intact spent fuel assemblies or canisters containing fuel consolidated from 48 assemblies.

A Small PWR-Core Physical Calculation Based on PWR-Core Analysis Code CORAL

2021 ◽  
Author(s):  
Wen Yang ◽  
Lun Zhou ◽  
Junrong Qiu ◽  
Yun Tai
Keyword(s):  
Density Distribution ◽  
Power Distribution ◽  
Channel Model ◽  
Three Dimensional ◽  
Least Square ◽  
Object Oriented Programming ◽  
Core Analysis ◽  
Fuel Temperature ◽  
Fuel Assemblies ◽  
Calculation Results

Abstract Three dimensional PWR-core analysis code CORAL is developed by Wuhan Second Ship Design and Research Institute. This code provides basic functions including three-dimensional power distribution, fine power reconstruction, fuel temperature distribution, critical search, control rod worth, reactivity coefficients, burnup and nuclide density distribution, etc. CORAL employ nodal expansion method to solve neutron diffusion equation, and the least square method is used to achieve few group constants, and sub-channel model and one-dimensional heat transfer is used to calculate fuel temperature and coolant density distribution, and burnup distribution and nuclide nuclear density could be obtained by solving macro-depletion and micro-depletion equation. The CORAL code is convenient to update and maintain in consider of modular, object-oriented programming technology. In order to analyze the computational accuracy of the CORAL code in small PWR-core and its capability to deal with heterogeneous, calculation analysis are carried out based on the material and geometry parameters of the SMART core. The core has 57 fuel assemblies, with 8, 20 or 24 gadolinium rods arranged in the fuel assemblies. In this paper, a quantitative comparison and analysis of the small PWR problem calculation results are carried out. Numerical results, including effective multiplication factor, assembly power distribution and pin power distribution, all agree well with the calculation results of OpenMC or Bamboo at both hot zero-power (HZP) and hot full-power (HFP) conditions.

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