Uncertainty Analyses Applied to the UAM/TMI-1 Lattice Calculations Using the DRAGON (Version 4.05) Code and Based on JENDL-4 and ENDF/B-VII.1 Covariance Data

The OECD/NEA Uncertainty Analysis in Modeling (UAM) expert group organized and launched the UAM benchmark. Its main objective is to perform uncertainty analysis in light water reactor (LWR) predictions at all modeling stages. In this paper, multigroup microscopic cross-sectional uncertainties are propagated through the DRAGON (version 4.05) lattice code in order to perform uncertainty analysis on and 2-group homogenized macroscopic cross-sections. The chosen test case corresponds to the Three Mile Island-1 (TMI-1) lattice, which is a 15 15 pressurized water reactor (PWR) fuel assembly segment with poison and at full power conditions. A statistical methodology is employed for the uncertainty assessment, where cross-sections of certain isotopes of various elements belonging to the 172-group DRAGLIB library format are considered as normal random variables. Two libraries were created for such purposes, one based on JENDL-4 data and the other one based on the recently released ENDF/B-VII.1 data. Therefore, multigroup uncertainties based on both nuclear data libraries needed to be computed for the different isotopic reactions by means of ERRORJ. The uncertainty assessment performed on and macroscopic cross-sections, that is based on JENDL-4 data, was much higher than the assessment based on ENDF/B-VII.1 data. It was found that the computed Uranium 235 fission covariance matrix based on JENDL-4 is much larger at the thermal and resonant regions than, for instance, the covariance matrix based on ENDF/B-VII.1 data. This can be the main cause of significant discrepancies between different uncertainty assessments.

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Uncertainty Analysis of Pressurized Water Reactor Core Cycle Depletion Calculation

10.13182/t31068 ◽

2019 ◽

Author(s):

J. Hou ◽

K. Ivanov ◽

K. Zeng

Keyword(s):

Uncertainty Analysis ◽

Reactor Core ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Water Reactor

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Prospect on the LOCA Analysis Method on the Advanced Pressurized Water Reactor in China

18th International Conference on Nuclear Engineering: Volume 4, Parts A and B ◽

10.1115/icone18-29345 ◽

2010 ◽

Cited By ~ 1

Author(s):

Yan Wang ◽

Xie Heng

Keyword(s):

Uncertainty Analysis ◽

System Analysis ◽

High Efficiency ◽

Current System ◽

National Laboratory ◽

Pressurized Water Reactor ◽

Economic Competition ◽

Pressurized Water ◽

Water Reactor ◽

Best Estimate

The LOCA analysis for the advanced pressurized water reactor (PWR) is very important and the methods on it are developing. There are two basic approaches for LOCA (loss of coolant accident) licensing at current. One is based on the conservative requirement of Appendix K of 10CFR50.46 of USNRC, and another is the best estimate (BE) analysis methodology which needs strict sensitivity and uncertainty analysis. The results achieved by the best estimate analysis are closer to the reality than those achieved by the conservative methodology, and the realistic BELOCA analysis in nuclear realm becomes an international trend currently although its development still meet lots of challenges. The research and design on AP1000 to be built in China and larger advanced pressurized water reactor (CAP1400 or CAP1700) as one of Chinese national science & technology major project is in progress. The reliable licensing LOCA analysis as one of the most important accident safety analysis is absolutely necessary. There are three ways to get the code applied in licensing accident analysis: the first way is developing code based on the best estimated methodology with strict uncertainty analysis, the second way is to develop new analysis code based on the conservative Appendix K, and the third way is improving the current system analysis code, which had been verified and validated by many cases, to satisfy the requirements of Appendix K. The last one may be the most feasible way for the AP1000 design with high efficiency and economic competition. Some code like RELAP5 has been used for LOCA analysis, and its results showed good agreement with the test data. RELAP is the transient thermal-hydraulic system analysis code developed by Idaho National Laboratory, in which some model and correlations are not consistent with the conservative requirements of Appendix K, so it can not be applied for licensing LOCA analysis and evaluation directly. In this paper the way to develop analysis code for LOCA license is discussed, and some areas in RELAP code needed to be modified for according with Appendix K are also described, which will be helpful for the advanced PWR design and development in China.

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Uncertainty analysis of a large break loss of coolant accident in a pressurized water reactor using non-parametric methods

Reliability Engineering & System Safety ◽

10.1016/j.ress.2018.02.005 ◽

2018 ◽

Vol 174 ◽

pp. 19-28 ◽

Cited By ~ 21

Author(s):

F. Sanchez-Saez ◽

A.I. Sánchez ◽

J.F. Villanueva ◽

S. Carlos ◽

S. Martorell

Keyword(s):

Uncertainty Analysis ◽

Parametric Methods ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Water Reactor ◽

Loss Of Coolant Accident ◽

Loss Of Coolant ◽

Non Parametric

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The Dilution Dependency of Multigroup Uncertainties

Science and Technology of Nuclear Installations ◽

10.1155/2014/306406 ◽

2014 ◽

Vol 2014 ◽

pp. 1-14

Author(s):

M. R. Ball ◽

C. McEwan ◽

D. R. Novog ◽

J. C. Luxat

Keyword(s):

Uncertainty Analysis ◽

Cross Sections ◽

Infinite Dilution ◽

Nuclear Data ◽

Monte Carlo Sampling ◽

Reactor Physics ◽

Cross Sectional ◽

Rigorous Approach ◽

Common Strategy ◽

Potential Methods

The propagation of nuclear data uncertainties through reactor physics calculation has received attention through the Organization for Economic Cooperation and Development—Nuclear Energy Agency’s Uncertainty Analysis in Modelling (UAM) benchmark. A common strategy for performing lattice physics uncertainty analysis involves starting with nuclear data and covariance matrix which is typically available at infinite dilution. To describe the uncertainty of all multigroup physics parameters—including those at finite dilution—additional calculations must be performed that relate uncertainties in an infinite dilution cross-section to those at the problem dilution. Two potential methods for propagating dilution-related uncertainties were studied in this work. The first assumed a correlation between continuous-energy and multigroup cross-sectional data and uncertainties, which is convenient for direct implementation in lattice physics codes. The second is based on a more rigorous approach involving the Monte Carlo sampling of resonance parameters in evaluated nuclear data using the TALYS software. When applied to a light water fuel cell, the two approaches show significant differences, indicating that the assumption of the first method did not capture the complexity of physics parameter data uncertainties. It was found that the covariance of problem-dilution multigroup parameters for selected neutron cross-sections can vary significantly from their infinite-dilution counterparts.

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A Flooding Induced Station Blackout Analysis for a Pressurized Water Reactor Using the RISMC Toolkit

Science and Technology of Nuclear Installations ◽

10.1155/2015/308163 ◽

2015 ◽

Vol 2015 ◽

pp. 1-14 ◽

Cited By ~ 5

Author(s):

Diego Mandelli ◽

Steven Prescott ◽

Curtis Smith ◽

Andrea Alfonsi ◽

Cristian Rabiti ◽

...

Keyword(s):

System Analysis ◽

Test Case ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Water Reactor ◽

Safety Margins ◽

Particle Hydrodynamics ◽

Station Blackout ◽

New Generation ◽

The Impact

In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation) and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code called NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. In addition, the impact of power uprate is determined in terms of both core damage probability and safety margins.

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