scholarly journals Neutronic Characteristics of ENDF/B-VIII.0 Compared to ENDF/B-VII.1 for Light-Water Reactor Analysis

2021 ◽  
Vol 2 (4) ◽  
pp. 318-335
Author(s):  
Kang-Seog Kim ◽  
William A. Wieselquist
Keyword(s):  
Nuclear Data ◽  
Data File ◽  
Monte Carlo Code ◽  
Pressurized Water ◽  
Benchmark Assessments ◽  
Light Water ◽  
Water Reactor ◽  
Water Reactors ◽  
Data Library ◽  
Neutronic Characteristics

The Evaluated Nuclear Data File (ENDF)/B-VIII.0 data library was released in 2018. To assess the new data during development and shortly after release, many validation calculations were performed with static, room-temperature benchmarks. Recently, when performing validation of ENDF/B-VIII.0 for pressurized water reactor depletion calculations, a regression in performance compared to ENDF/B-VII.1 was observed. This paper documents extensive benchmark calculations for light-water reactors performed using continuous energy Monte Carlo code with ENDF/B-VII.1 and -VIII.0 and neutronic characteristics of ENDF/B-VIII.0 are discussed and compared to those of ENDF/B-VII.1. It is our recommendation that ENDF/B data library assessment should include reactor-specific benchmark assessments, including depletion cases, such that these types of regressions may be caught earlier in the data development cycle.

Pressurized Water Reactor Environment Effect on 316 Stainless Steel Stress Hardening/Softening: An Experimental Study

2015 ◽  
Author(s):  
Subhasish Mohanty ◽  
William K. Soppet ◽  
Saurindranath Majumdar ◽  
Krishnamurti Natesan
Keyword(s):  
Stainless Steel ◽  
Fatigue Life ◽  
Stress Strain ◽  
Modeling Tools ◽  
Pressurized Water ◽  
Light Water ◽  
Water Reactor ◽  
Life Estimation ◽  
316 Stainless Steel ◽  
Water Reactors

In USA there are approximately 100 operating light water reactors (LWR) consisting fleet of both pressurized water reactors (PWR) and boiling water reactors (BWR). Most of these reactors were built before 1970 and the design lives of most of these reactors are 40 years. It is expected that by 2030, even those reactors that have received 20 year life extension license from the US nuclear regulatory commission (NRC) will begin to reach the end of their licensed periods of operation. For economical reason it is be beneficial to extend the license beyond 60 to perhaps 80 years that would enable existing plants to continue providing safe, clean and economic electricity without significant green house gas emissions. However, environmental fatigue is one of the major aging related issues for these reactors, and may create hurdles in long term sustainability of these reactors. To address some of the environmental fatigue related issues, Argonne National Laboratory (ANL) with the sponsorship of Department of Energy’s Light Water Reactor Sustainability (LWRS) program trying to develop mechanistic approach for more accurate life estimation of LWR components. In this context ANL conducted many fatigue experiments under different test and environment conditions on 316 stainless steel (316SS) material that is or similar grade steels are widely used in US reactors. Contrary to the conventional S∼N curve based empirical fatigue life estimation approach, the aim of the present DOE sponsored work is to understand material ageing more mechanistically (e.g. time dependent hardening and softening) under different test and environmental conditions. Better mechanistic understanding will help to develop computer based advanced modeling tools to better extrapolate stress-strain evolution of reactor component under multi-axial stress states and hence to help predicting their fatigue life more accurately. In this paper (part-I) the fatigue experiments under different test and environment conditions and related stress-strain results for 316 SS are discussed. In another paper (part-II) the related evolutionary cyclic plasticity material modeling techniques and results are discussed.

scholarly journals Uncertainty quantification and sensitivity studies on Thorium-fueled reactors

2020 ◽  
Vol 239 ◽  
pp. 22008
Author(s):  
Eliot Party ◽  
Xavier Doligez ◽  
Philippe Dessagne ◽  
Maëlle Kerveno ◽  
Greg Henning
Keyword(s):  
Cross Sections ◽  
Nuclear Data ◽  
Pressurized Water ◽  
Correlation Matrices ◽  
Target Uncertainty ◽  
Nuclear Data Library ◽  
Sensitivity Studies ◽  
Water Reactors ◽  
Data Library ◽  
Good Agreement

This paper shows how Total Monte Carlo (TMC) method and Perturbation Theory (PT) can be applied to quantify uncertainty due to nuclear data on reactor static calculations of integral parameters such as keff and βeff. This work focuses on thorium fueled reactors and it aims to rank different cross sections uncertainty regarding criticality calculations. The consistency of the two methods are first studied. The cross sections set used for the TMC method is computed to build adequate correlation matrices. Those matrices are then multiplied by the sensitivity coefficients obtained thanks to the PT to obtain global uncertainties that are compared to the ones calculated by the TMC method. Results in good agreement allow us to use correlation matrix from the state of the art nuclear data library (JEFF 3-3) that provide insight of uncertainty on keff and βeff for thorium fueled Pressurized Water Reactors. Finally, maximum uncertainties on cross sections are estimated to reach a target uncertainty on integral parameters. It is shown that a strong reduction of the current uncertainty is needed and consequently, new measurements and evaluations have to be performed.

Synthesis of I-131 labelled iodine species relevant during severe nuclear accidents in light water reactors

2013 ◽  
Vol 101 (10) ◽  
pp. 675-680 ◽  
Author(s):  
S. Tietze ◽  
M. R. St. J. Foreman ◽  
C. Ekberg
Keyword(s):  
Light Water Reactor ◽  
Nuclear Accidents ◽  
Light Water Reactors ◽  
Light Water ◽  
Water Reactor ◽  
Iodine Species ◽  
Water Reactors ◽  
Paint Films ◽  
Elemental Iodine ◽  
Severe Nuclear Accident

Summary Methods for the small scale synthesis of I-131 labelled iodine species relevant to severe nuclear accidents in light water reactors have been developed. The introduced methods allow the synthesis of impurity free, volatile, inorganic elemental iodine and volatile, organic iodides such as methyl- and ethyl iodide, as well as butyl iodide, chloroiodomethane, allyl iodide and benzyl iodide with ease. The radioactive iodine containing products are sufficiently stable to allow their storage for later use. Due to their volatility the liquid species can be easily converted into gaseous species and thus can be used in research in liquid and gaseous phase. The primary motivation for the development of these synthesis methods is to study the behaviour of volatile iodine species under the conditions of a severe nuclear accident in a light water reactor. Thus, the chemicals involved in the synthesis are chosen in a way to not generate impurities (chlorine and organic solvents) in the products which interfere with competing reactions relevant during a severe nuclear accident. Teknopox Aqua VA epoxy paint, which is used in Swedish light water reactor containments, and its reactions with the produced iodine species are described. The synthesised iodine species undergo chemisorption on paint films. Different to elemental iodine, the organic iodides are non-reactive with copper surfaces. The sorbed iodine species are partly re-released mainly in form of organic iodides and not as elemental iodine when the exposed paint films are heat treated. The partitioning and hydrolysis behaviour of gaseous methyl- and ethyl iodide between containment gas phase and water pools is found to be similar. The methods have been designed to minimise the use of harmful materials and the production of radioactive waste.

scholarly journals Pressurized water reactor spent nuclear fuel data library produced with the Serpent2 code

Data in Brief ◽  
2020 ◽  
Vol 33 ◽  
pp. 106429
Author(s):  
Zsolt Elter ◽  
Li Pöder Balkeståhl ◽  
Erik Branger ◽  
Sophie Grape
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Data Library

scholarly journals An Experimental Programme optimized with Uncertainty Propagation: PETALE in the CROCUS Reactor

2019 ◽  
Vol 211 ◽  
pp. 03003 ◽  
Author(s):  
Vincent Lamirand ◽  
Axel Laureau ◽  
Dimitri Rochman ◽  
Gregory Perret ◽  
Adrien Gruel ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Energy Range ◽  
Present Article ◽  
Uncertainty Propagation ◽  
Nuclear Data ◽  
Pressurized Water Reactors ◽  
Pressurized Water ◽  
Experimental Programme ◽  
Iron Nickel ◽  
Water Reactors

The PETALE experimental programme in the CROCUS reactor at EPFL intends to contribute to the validation and improvement of neutron nuclear data in the MeV energy range for stainless steel, particularly in the prospect of heavy reflector elements of pressurized water reactors. It mainly consists of several transmission experiments: first, through metallic sheets of nuclear-grade stainless steel interleaved with dosimeter foils, and, successively, through its elemental components of interest – iron, nickel, and chromium. The present article describes the study for optimizing the response of the dosimetry experiments to the nuclear data of interest.

Sign in / Sign up

Export Citation Format

Share Document