Engineering › Aerospace Engineering

Nuclear reactor physics and engineering

Works Count: 102163

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Description

This cluster of papers covers advanced nuclear reactor technology, including topics such as molten salt reactors, neutron transport, Generation IV reactors, the thorium fuel cycle, and small modular reactors. It also discusses the development and application of Monte Carlo codes, lead-cooled fast reactors, transmutation processes, and nuclear data libraries.

Keywords

Nuclear Reactor; Molten Salt; Neutron Transport; Generation IV; Thorium Fuel Cycle; Monte Carlo Code; Lead-Cooled Fast Reactor; Transmutation; Small Modular Reactors; Nuclear Data Library

Monte Carlo Particle Transport Methods: Neutron and Photon Calculations

2018-05-04

Iván Lux , L. Koblinger

With this book we try to reach several more-or-less unattainable goals namely: To compromise in a single book all the most important achievements of Monte Carlo calculations for solving neutron … With this book we try to reach several more-or-less unattainable goals namely: To compromise in a single book all the most important achievements of Monte Carlo calculations for solving neutron and photon transport problems. To present a book which discusses the same topics in the three levels known from the literature and gives us useful information for both beginners and experienced readers. It lists both well-established old techniques and also newest findings.

Empirical Equations to Calculate 16 of the Transport Collision Integrals Ω(<i>l, s</i>)* for the Lennard-Jones (12–6) Potential

1972-08-01

Philip D. Neufeld , A. Janzen , R. A. Aziz

We have calculated 16 of the reduced transport collision integrals Ω(l, s)* as a function of reduced temperature T* for the Lennard-Jones (12–6) potential. These calculations are more accurate than … We have calculated 16 of the reduced transport collision integrals Ω(l, s)* as a function of reduced temperature T* for the Lennard-Jones (12–6) potential. These calculations are more accurate than those of Hirschfelder, Curtiss, and Bird, which are frequently used. Empirical equations are presented which allow the calculation of the collision integrals for any reduced temperature in the range 0.3≤ T*≤ 100 without interpolation from tables. The error in the values so obtained is probably less than 0.1%.

ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology

2006-12-01

M. B. Chadwick , P. Obložinský , M. Herman +7 more

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Dynamic Models for Steam and Hydro Turbines in Power System Studies

1973-11-01

IEEE Report

Basic models for speed-governing systems and turbines in power system stability studies are presented. These models provide adequate representation for hydro, fossil-fired, and pressurized water reactor nuclear units in most … Basic models for speed-governing systems and turbines in power system stability studies are presented. These models provide adequate representation for hydro, fossil-fired, and pressurized water reactor nuclear units in most stability analyses. Models for boiling water reactor nuclear units are to be presented at a later date. Typical parameters are given.

Initial MCNP6 Release Overview

2012-12-01

Tim Goorley , Michael R. James , Tom Booth +7 more

MCNP6 is simply and accurately described as the merger of MCNP5 and MCNPX capabilities, but it is much more than the sum of those two computer codes. MCNP6 is the … MCNP6 is simply and accurately described as the merger of MCNP5 and MCNPX capabilities, but it is much more than the sum of those two computer codes. MCNP6 is the result of five years of effort by the MCNP5 and MCNPX code development teams. These groups of people, residing in Los Alamos National Laboratory's (LANL) X Computational Physics Division, Monte Carlo Codes Group (XCP-3), and Decision Applications Division, Radiation Transport and Applications Team (D-5), respectively, have combined their code development efforts to produce the next evolution of MCNP. While maintenance and bug fixes will continue for MCNP5 1.60 and MCNPX 2.7.0 for upcoming years, new code development capabilities only will be developed and released in MCNP6. In fact, the initial release of MCNP6 contains 16 new features not previously found in either code. These new features include the abilities to import unstructured mesh geometries from the finite element code Abaqus, to transport photons down to 1.0 eV, to transport electrons down to 10.0 eV, to model complete atomic relaxation emissions, and to generate or read mesh geometries for use with the LANL discrete ordinates code Partisn. The first release of MCNP6, MCNP6 Beta 2, is now available through the Radiation Safety Information Computational Center, and the first production release is expected in calendar year 2012. High confidence in the MCNP6 code is based on its performance with the verification and validation test suites, comparisons to its predecessor codes, the regression test suite, its code development process, and the underlying high-quality nuclear and atomic databases.

The Serpent Monte Carlo code: Status, development and applications in 2013

2014-09-08

Jaakko Leppänen , Maria Pusa , Tuomas Viitanen +2 more

Modern Nuclear Data Evaluation with the TALYS Code System

2012-11-30

A. J. Koning , D. Rochman

JENDL-4.0: A New Library for Nuclear Science and Engineering

2011-01-01

Keiichi SHIBATA , Osamu Iwamoto , Tsuneo NAKAGAWA +7 more

Abstract The fourth version of the Japanese Evaluated Nuclear Data Library has been produced in cooperation with the Japanese Nuclear Data Committee. In the new library, much emphasis is placed … Abstract The fourth version of the Japanese Evaluated Nuclear Data Library has been produced in cooperation with the Japanese Nuclear Data Committee. In the new library, much emphasis is placed on the improvements of fission product and minor actinoid data. Two nuclear model codes were developed in order to evaluate the cross sections of fission products and minor actinoids. Coupled-channel optical model parameters, which can be applied to wide mass and energy regions, were obtained for nuclear model calculations. Thermal cross sections of actinoids were carefully examined by considering experimental data or by the systematics of neighboring nuclei. Most of the fission cross sections were derived from experimental data. A simultaneous evaluation was performed for the fission cross sections of important uranium and plutonium isotopes above 10 keV. New evaluations were performed for the thirty fissionproduct nuclides that had not been contained in the previous library JENDL-3.3. The data for light elements and structural materials were partly reevaluated. Moreover, covariances were estimated mainly for actinoids. The new library was released as JENDL-4.0, and the data can be retrieved from the Web site of the JAEA Nuclear Data Center. KEYWORDS: JENDL-4.0nuclear dataevaluationcross sectionnuclear model calculationexperimental dataactinoidfission productlight elementstructural material

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Statistical multifragmentation of nuclei

1995-06-01

J.P. Bondorf , A. S. Botvina , A. S. Iljinov +2 more

Variational Methods in Nuclear Collision Problems

1948-12-15

W. Kohn

Variational methods, similar to the Rayleigh-Ritz method for bound state calculations, are developed for the phase shifts and elements of the scattering matrix in nuclear collisions. Numerical applications to neutron-proton … Variational methods, similar to the Rayleigh-Ritz method for bound state calculations, are developed for the phase shifts and elements of the scattering matrix in nuclear collisions. Numerical applications to neutron-proton and neutron-deuteron scattering involving trial functions with undetermined coefficients are described. Another variational principle, for scattering amplitudes, is shown to lead to the Born approximations and a formula recently derived by Schwinger. It may also be used in conjunction with the method of undetermined coefficients.

Particle and Heavy Ion Transport code System, PHITS, version 2.52

2013-07-31

Tatsuhiko Sato , Koji Niita , Norihiro Matsuda +7 more

An upgraded version of the Particle and Heavy Ion Transport code System, PHITS2.52, was developed and released to the public. The new version has been greatly improved from the previously … An upgraded version of the Particle and Heavy Ion Transport code System, PHITS2.52, was developed and released to the public. The new version has been greatly improved from the previously released version, PHITS2.24, in terms of not only the code itself but also the contents of its package, such as the attached data libraries. In the new version, a higher accuracy of simulation was achieved by implementing several latest nuclear reaction models. The reliability of the simulation was improved by modifying both the algorithms for the electron-, positron-, and photon-transport simulations and the procedure for calculating the statistical uncertainties of the tally results. Estimation of the time evolution of radioactivity became feasible by incorporating the activation calculation program DCHAIN-SP into the new package. The efficiency of the simulation was also improved as a result of the implementation of shared-memory parallelization and the optimization of several time-consuming algorithms. Furthermore, a number of new user-support tools and functions that help users to intuitively and effectively perform PHITS simulations were developed and incorporated. Due to these improvements, PHITS is now a more powerful tool for particle transport simulation applicable to various research and development fields, such as nuclear technology, accelerator design, medical physics, and cosmic-ray research.

Monte Carlo Calculations of Nuclear Evaporation Processes. III. Applications to Low-Energy Reactions

1959-11-01

I. Dostrovsky , Z. Fraenkel , Gérard Friedlander

Monte Carlo calculations of nuclear reactions in the low-energy ($E<50$ Mev) region are described. The calculations are based on the nuclear evaporation model of Weisskopf. Continuum theory was used for … Monte Carlo calculations of nuclear reactions in the low-energy ($E<50$ Mev) region are described. The calculations are based on the nuclear evaporation model of Weisskopf. Continuum theory was used for the calculation of inverse reaction cross sections. In the calculation of the level densities of excited nuclei, pairing and shell energy corrections were used in terms of characteristic level displacements. The accurate equation rather than the approximate Maxwell distribution was used for the selection of the kinetic energy of the evaporated particle. Experimentally determined $Q$-values for the various reactions were used. The calculations are compared with experimental measurements for about 60 excitation functions of nuclear reactions in the mass range ${\mathrm{Cr}}^{50}$-${\mathrm{Se}}^{74}$. Cameron's values for pairing energies were used at the outset; but a new set of pairing and shell energy correction values, which leads to substantially improved agreement with the experimental curves, is presented. The procedure which was used to arrive at this set is described and several features of the set are discussed. The need for a further downward correction of the level density of symmetrical ($A=2Z$) nuclei is indicated. Computed excitation functions are shown for all the reactions studied as well as for several reactions for which experimental data are not yet available. Further experiments on reaction cross sections are suggested which would allow a unique determination of the pairing and shell energy corrections of level densities for any value of $Z$ and $N$ in the region under discussion. The existence of a unique set of these correction terms would provide strong evidence for the validity of evaporation theory for the reactions considered.

Fitting using finite Monte Carlo samples

1993-10-01

R. J. Barlow , C. Beeston

Statistics and Nuclear Reactions

1937-08-15

V. F. Weisskopf

It is possible to apply statistical methods to the calculation of nuclear processes provided that the energies involved are large in comparison with the lowest excitation energies of nuclei. Expressions … It is possible to apply statistical methods to the calculation of nuclear processes provided that the energies involved are large in comparison with the lowest excitation energies of nuclei. Expressions are obtained for the emission probability of neutrons or charged particles by highly excited heavy nuclei. These expressions are built up in a similar way to the formula for the probability of evaporation of a particle from a body at low temperatures. In applying it to the impact of high energy neutrons on heavy nuclei, the mean energy loss per impact turns out to be $E[1\ensuremath{-}2{(\frac{a}{E})}^{\frac{1}{2}}]$ where $E$ is the energy of the incident neutrons and $a$ is dependent on the nuclear structure; we can put approximately $a\ensuremath{\sim}0.05\ensuremath{-}0.2$ MV. The energy distribution of the scattered neutrons is approximately a Maxwellian one with a mean energy of $2{(\mathrm{aE})}^{\frac{1}{2}}$.

Elementary solutions of the transport equation and their applications

1960-01-01

K. M. Case

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Neutron Transport Theory

2018-02-09

Prof. Weston M. Stacey

The differential and integral neutron transport equations that describe the neutron distribution in space, energy, and direction are derived and methods for solving them are described. The application of analog … The differential and integral neutron transport equations that describe the neutron distribution in space, energy, and direction are derived and methods for solving them are described. The application of analog simulation (Monte Carlo) methods to determine the neutron distribution is described.

Generation-IV nuclear power: A review of the state of the science

2008-11-10

T. Abram , S.E. Ion

ORIGEN2: A Versatile Computer Code for Calculating the Nuclide Compositions and Characteristics of Nuclear Materials

1983-09-01

A.G. Croff

ORIGEN2 is a versatile point-depletion and radioactive-decay computer code for use in simulating nuclear fuel cycles and calculating the nuclide compositions and characteristics of materials contained therein. It represents a … ORIGEN2 is a versatile point-depletion and radioactive-decay computer code for use in simulating nuclear fuel cycles and calculating the nuclide compositions and characteristics of materials contained therein. It represents a revision and update of the original ORIGEN computer code, which was developed at the Oak Ridge National Laboratory (ORNL) and distributed worldwide beginning in the early 1970s. Included in ORIGEN2 are provisions for incorporating data generated by more sophisticated reactor physics codes, a free-format input, and a highly flexible and controllable output; with these features, ORIGEN2 has the capability for simulating a wide variety of fuel cycle flow sheets.The decay, cross-section, fission product yield, and photon emission data bases employed by ORIGEN2 have been extensively updated, and the list of reactors that can be simulated includes pressurized water reactors, boiling water reactors, liquid-metal fast breeder reactors, and Canada deuterium uranium reactors. A number of verification activities have been undertaken, including (a) comparison of ORIGEN2 decay heat results with both calculated and experimental values, and (b) comparison of predicted spent fuel compositions with measured values. The agreement between ORIGEN2 and the comparison bases is generally very good. Future work concerning ORIGEN2 will involve continued maintenance and user support along with additional verification studies and limited modifications to enhance its flexibility and usability. ORIGEN2 can be obtained, free of charge, from the ORNL Radiation Shielding Information Center.

The molten salt reactor (MSR) in generation IV: Overview and perspectives

2014-03-09

J. Serp , M. Allibert , O. Beneš +7 more

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On the Yield of Nuclear Reactions with Heavy Elements

1940-03-15

V. F. Weisskopf , Douglas H. Ewing

The cross sections for different kinds of nuclear reactions are calculated as functions of the energy of the bombarding particles by means of statistical methods. Their application is restricted to … The cross sections for different kinds of nuclear reactions are calculated as functions of the energy of the bombarding particles by means of statistical methods. Their application is restricted to heavy elements ($A>50$) and to bombarding energies greater than 1 Mev. The excitation curves of several ($p,n$)-reactions have been measured for elements with $A$ between 60 and 115; it is found that the measured cross sections and their dependence on the energy suggests a nuclear radius of $R=1.3\ifmmode\times\else\texttimes\fi{}{10}^{\ensuremath{-}13}\ifmmode\times\else\texttimes\fi{}{A}^{\frac{1}{3}}$ cm for these elements. Section I gives a complete discussion of the calculated cross sections. Section II and III contain the derivations of these expressions. Section IV describes the new experimental material and its implications for the theory.

Fast iterative methods for discrete-ordinates particle transport calculations

2002-01-01

Marvin L. Adams , Edward W. Larsen

Calculation of complication probability factors for non-uniform normal tissue irradiation: The effective volume method gerald

1989-06-01

Gerald J. Kutcher , Chandra Burman

ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data

2011-12-01

M. B. Chadwick , M. Herman , P. Obložinský +7 more

MOOSE: A parallel computational framework for coupled systems of nonlinear equations

2009-07-20

Derek Gaston , Chris Newman , Glen Hansen +1 more

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Design and development of the AHWR—the Indian thorium fuelled innovative nuclear reactor

2006-02-20

R.K. Sinha , A. Kakodkar

Effective Field Theory

1993-12-01

Howard Georgi

Following a major shortage of 99Mo in the 2009–2010 period, concern grew that the aging reactor production facilities needed to be replaced. Most producers were using highly enriched 235U (HEU) … Following a major shortage of 99Mo in the 2009–2010 period, concern grew that the aging reactor production facilities needed to be replaced. Most producers were using highly enriched 235U (HEU) as the target material. The Organisation for Economic Co-...Read More

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A generalized thermodynamic correlation based on three‐parameter corresponding states

1975-05-01

Byung Ik Lee , Michael G. Kesler

Abstract The volumetric and thermodynamic functions correlated by Pitzer and co‐workers analytically represented with improved accuracy by a modified BWR equation of state. The representation provides a smooth transition between … Abstract The volumetric and thermodynamic functions correlated by Pitzer and co‐workers analytically represented with improved accuracy by a modified BWR equation of state. The representation provides a smooth transition between the original tables of Pitzer et al. and more recent extensions to lower temperatures. It is in a form particularly convenient for computer use.

A multidimensional unfolding method based on Bayes' theorem

1995-08-01

G. D’Agostini

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Japanese Evaluated Nuclear Data Library Version 3 Revision-2: JENDL-3.2.

1995-01-01

Tsuneo NAKAGAWA , Keiichi SHIBATA , Satoshi Chiba +7 more

The revision work of JENDL-3 has been made by considering feedback information of various benchmark tests. The main revised quantities are the resonance parameters, capture and inelastic scattering cross sections, … The revision work of JENDL-3 has been made by considering feedback information of various benchmark tests. The main revised quantities are the resonance parameters, capture and inelastic scattering cross sections, and fission spectra of main actinide nuclides, the total and inelastic scattering cross sections of structural materials, the resonance parameters the capture and inelastic scattering cross sections of fission products, and the γr-ray production data. The revised data were released as JENDL-3.2 in June 1994. The preliminary benchmark tests indicate that JENDL-3.2 predicts various reactor characteristics more successfully than the previous version of JENDL-3.1.

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OpenMC: A state-of-the-art Monte Carlo code for research and development

2014-08-14

Paul Romano , Nicholas E. Horelik , Bryan Herman +3 more

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Linear transport theory

1968-04-01

Alvin M. Weinberg

Computational methods of neutron transport

1985-01-01

M.M.R. Williams

Isospin in Nuclear Physics

1970-10-01

L. W. Nordheim

"Isospin in Nuclear Physics." Nuclear Science and Engineering, 42(1), pp. 116–117 Additional informationNotes on contributorsLothar W. NordheimAbout the Reviewer: Lothar Nordheim has made substantial contributions to both nuclear and reactor … "Isospin in Nuclear Physics." Nuclear Science and Engineering, 42(1), pp. 116–117 Additional informationNotes on contributorsLothar W. NordheimAbout the Reviewer: Lothar Nordheim has made substantial contributions to both nuclear and reactor physics. He has held positions at universities, national laboratories, and in industry. He is at present a consultant for Gulf General Atomic Incorporated. He is a Fellow of the American Nuclear and Physical Societies and a member of the Editorial Advisory Committee of Nuclear Science and Engineering.

Monte Carlo Principles and Neutron Transport Problems

1971-12-01

R. R. Coveyou

"Monte Carlo Principles and Neutron Transport Problems." Nuclear Science and Engineering, 46(3), pp. 439–440 "Monte Carlo Principles and Neutron Transport Problems." Nuclear Science and Engineering, 46(3), pp. 439–440

Shell-model approach to nuclear reactions

1970-08-01

K. Harada

TENDL: Complete Nuclear Data Library for Innovative Nuclear Science and Technology

2019-01-01

A. J. Koning , D. Rochman , Jean-Christophe Sublet +3 more

The NUBASE2016 evaluation of nuclear properties

2017-03-01

G. Audi , F. G. Kondev , M. Wang +2 more

This paper presents the NUBASE2016 evaluation that contains the recommended values for nuclear and decay properties of 3437 nuclides in their ground and excited isomeric (T1/2 ⩾ 100 ns) states. … This paper presents the NUBASE2016 evaluation that contains the recommended values for nuclear and decay properties of 3437 nuclides in their ground and excited isomeric (T1/2 ⩾ 100 ns) states. All nuclides for which any experimental information is known were considered. NUBASE2016 covers all data published by October 2016 in primary (journal articles) and secondary (mainly laboratory reports and conference proceedings) references, together with the corresponding bibliographical information. During the development of NUBASE2016, the data available in the 'Evaluated Nuclear Structure Data File' (ENSDF) database were consulted and critically assessed for their validity and completeness. Furthermore, a large amount of new data and some older experimental results that were missing from ENSDF were compiled, evaluated and included in NUBASE2016. The atomic mass values were taken from the 'Atomic Mass Evaluation' (AME2016, second and third parts of the present issue). In cases where no experimental data were available for a particular nuclide, trends in the behavior of specific properties in neighboring nuclides (TNN) were examined. This approach allowed to estimate values for a range of properties that are labeled in NUBASE2016 as 'non-experimental' (flagged '#'). Evaluation procedures and policies used during the development of this database are presented, together with a detailed table of recommended values and their uncertainties.

The joint evaluated fission and fusion nuclear data library, JEFF-3.3

2020-07-01

Arjan Plompen , Ó. Cabellos , C. De Saint Jean +7 more

Abstract The joint evaluated fission and fusion nuclear data library 3.3 is described. New evaluations for neutron-induced interactions with the major actinides $$^{235}\hbox {U}$$ <mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML"><mml:mrow><mml:msup><mml:mrow /><mml:mn>235</mml:mn></mml:msup><mml:mtext>U</mml:mtext></mml:mrow></mml:math> , $$^{238}\hbox {U}$$ … Abstract The joint evaluated fission and fusion nuclear data library 3.3 is described. New evaluations for neutron-induced interactions with the major actinides $$^{235}\hbox {U}$$ <mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML"><mml:mrow><mml:msup><mml:mrow /><mml:mn>235</mml:mn></mml:msup><mml:mtext>U</mml:mtext></mml:mrow></mml:math> , $$^{238}\hbox {U}$$ <mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML"><mml:mrow><mml:msup><mml:mrow /><mml:mn>238</mml:mn></mml:msup><mml:mtext>U</mml:mtext></mml:mrow></mml:math> and $$^{239}\hbox {Pu}$$ <mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML"><mml:mrow><mml:msup><mml:mrow /><mml:mn>239</mml:mn></mml:msup><mml:mtext>Pu</mml:mtext></mml:mrow></mml:math> , on $$^{241}\hbox {Am}$$ <mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML"><mml:mrow><mml:msup><mml:mrow /><mml:mn>241</mml:mn></mml:msup><mml:mtext>Am</mml:mtext></mml:mrow></mml:math> and $$^{23}\hbox {Na}$$ <mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML"><mml:mrow><mml:msup><mml:mrow /><mml:mn>23</mml:mn></mml:msup><mml:mtext>Na</mml:mtext></mml:mrow></mml:math> , $$^{59}\hbox {Ni}$$ <mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML"><mml:mrow><mml:msup><mml:mrow /><mml:mn>59</mml:mn></mml:msup><mml:mtext>Ni</mml:mtext></mml:mrow></mml:math> , Cr, Cu, Zr, Cd, Hf, W, Au, Pb and Bi are presented. It includes new fission yields, prompt fission neutron spectra and average number of neutrons per fission. In addition, new data for radioactive decay, thermal neutron scattering, gamma-ray emission, neutron activation, delayed neutrons and displacement damage are presented. JEFF-3.3 was complemented by files from the TENDL project. The libraries for photon, proton, deuteron, triton, helion and alpha-particle induced reactions are from TENDL-2017. The demands for uncertainty quantification in modeling led to many new covariance data for the evaluations. A comparison between results from model calculations using the JEFF-3.3 library and those from benchmark experiments for criticality, delayed neutron yields, shielding and decay heat, reveals that JEFF-3.3 performes very well for a wide range of nuclear technology applications, in particular nuclear energy.

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Techniques for Nuclear and Particle Physics Experiments

1994-01-01

William R. Leo

Japanese Evaluated Nuclear Data Library Version 3 Revision-3: JENDL-3.3

2002-11-01

Keiichi SHIBATA , Toshihiko Kawano , Tsuneo NAKAGAWA +7 more

Abstract Evaluation for JENDL-3.3 has been performed by considering the accumulated feedback information and various benchmark tests of the previous library JENDL-3.2. The major problems of the JENDL-3.2 data were … Abstract Evaluation for JENDL-3.3 has been performed by considering the accumulated feedback information and various benchmark tests of the previous library JENDL-3.2. The major problems of the JENDL-3.2 data were solved by the new library: overestimation of criticality values for thermal fission reactors was improved by the modifications of fission cross sections and fission neutron spectra for 235U; incorrect energy distributions of secondary neutrons from important heavy nuclides were replaced with statistical model calculations; the inconsistency between elemental and isotopic evaluations was removed for medium-heavy nuclides. Moreover, covariance data were provided for 20 nuclides. The reliability of JENDL-3.3 was investigated by the benchmark analyses on reactor and shielding performances. The results of the analyses indicate that JENDL-3.3 predicts various reactor and shielding characteristics better than JENDL- 3.2. KEYWORDS: JENDL-3.3nuclear dataevaluationreliabilityresonancecross sectionspectrumcovariancebenchmarkcriticalityshielding

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Estimation of Levelized Cost of Energy for Small Modular Reactors in Colombia: A Monte Carlo Simulation Approach

2025-06-25

Prieto Camilo , Patiño Diego , V. A. Jose | International Journal of Energy Economics and Policy

Small modular reactors (SMRs) offer significant prospects for Colombia to diversify and decarbonize its energy mix by 2038, as specified in the National Energy Plan (PEN) 2022-2052. However, current assessments … Small modular reactors (SMRs) offer significant prospects for Colombia to diversify and decarbonize its energy mix by 2038, as specified in the National Energy Plan (PEN) 2022-2052. However, current assessments primarily focus on capital expenditure (CAPEX) indicators, such as the overnight capital cost (OCC), while overlooking the Levelized cost of energy (LCOE), which provides a more comprehensive measure of long-term economic viability. This study employs a Monte Carlo simulation to calculate the LCOE of a 300 MW SMRs in Colombia for the 2038-2042 period, including probabilistic distributions for OCC, operational expenditures (OPEX), fuel cost, and capacity factor. The results indicate a median LCOE of $77.71/MWh, with a range from $68.26/MWh in optimistic scenarios to $117.80/MWh in pessimistic ones. These findings suggest that SMRs could serve as a cost-competitive alternative to coal-fired power plants, particularly when externalities such as carbon emissions are considered. Sensitivity analysis identifies OCC and the weighted average cost of capital (WACC) as the most influential cost drivers. Additionally, fuel procurement strategies, including reprocessed fuel and long-term contracts, can further reduce operational costs. This study underscores the importance of integrating LCOE into energy planning and calls for regulatory and financial mechanisms to support SMRs deployment in Colombia.

The applications and prospects of BEPU methodology in Generation III nuclear reactors

2025-06-25

Mengyan Hu , Chengcheng Deng , Shizhen Yu +2 more | Nuclear Engineering and Design

Research and development of high temperature gas-cooled reactor nuclear power plants for combined production of electricity, heat and hydrogen

2025-06-24

Vladimir Kindra , Andrey Rogalev , D.S. Kovalev +2 more | International Journal of Hydrogen Energy

Development of a Relative Rate Technique to Measure Criegee Intermediate Reactivity

2025-06-24

Pengcheng Wang , Yue Wang , Jinhe Wang +4 more | Environmental Science & Technology Letters

Characterization of MHD pressure losses in a mock-up of the WCLL Test Blanket Module

2025-06-24

Cyril Courtessole , H.-J. Brinkmann , L. Bühler +2 more | Fusion Engineering and Design

Refueling decision method in the PB-HTR based on the reactivity equilibrium equation

2025-06-24

Zhongkai Sheng , Bing Xia , Ziwen Zhang +3 more | Progress in Nuclear Energy

Development of a steady-state single-channel model of the BAEC 3MWt TRIGA mark-II research reactor for thermal and hydrodynamic analysis

2025-06-24

Saad Islam , Ali Akbar Mahdi , Mohammad Wahiduzzaman Khan +1 more | Nuclear Engineering and Design

A neutron transport calculation approach via the coupling of 2D/1D and 3D SN nodal methods

2025-06-24

Hongchun Wu , Shuaizheng Li , Jun Chen +5 more | Annals of Nuclear Energy

Experimental and numerical investigation on the flow and heat transfer characteristics of lead-bismuth eutectic through a helical single rod and four-rod bundle

2025-06-23

Dianqiang Jiang , Yi‐Wen Chen , Dalin Zhang +6 more | International Journal of Heat and Mass Transfer

Demonstration of TOFFEE: A Response Uncertainty Quantification Tool

2025-06-23

Austin Williams , Lance Drouet , Sandra Bogetic | Nuclear Science and Engineering

Sensitivity Analysis of an Overcooling Transient in the Generic Fluoride-Salt-Cooled High-Temperature Reactor

2025-06-23

Jonathan L. Barthle , Isabelle O. Lindsay , Nader Satvat +1 more | Nuclear Science and Engineering

Equivalence of the Peierls and Serber Wartime Expressions for the Critical Radius of a Fissile Isotope

2025-06-23

B. Cameron Reed | Nuclear Technology

Assessment of JEFF Nuclear Data Libraries for Reactor Physics and Spent Fuel Composition Applications

2025-06-23

Mathieu Hursin , Fan Xia , D. Rochman | Nuclear Science and Engineering

The implications of impurities on shutdown dose rates in the burning plasma experimental tokamak

2025-06-21

Lin Kong , Xin Tong , Xiaokang Zhang +1 more | Fusion Engineering and Design

On-The-Fly multigroup cross section generation method with continuous energy accuracy

2025-06-20

Farzad Rahnema , Sheng Zhang | Annals of Nuclear Energy

An overview of molten salt reactor research experiments at NRG PALLAS

2025-06-20

Eva de Visser – Týnová , P.R. Hania , Mathilde Laot +5 more | Nuclear Engineering and Design

SPARTA: A flux adjustment methodology to interpret complex experiments

2025-06-19

Paul Ferney , Benjamin Baker , Mark D. DeHart | Annals of Nuclear Energy

The molten salt tritium transport experiment: A pumped fluoride salt loop for hydrogen isotope experimentation

2025-06-19

Thomas F. Fuerst , Anthony Bowers , H. Gietl +6 more | Annals of Nuclear Energy

Computational study into the experimental capabilities of the MTRR SCW reactor

2025-06-19

Anton S. Lapin , V. Yu. Blandinsky , V. A. Nevinitsa +4 more | Nuclear Energy and Technology

Two stages of the MTRR-SCW reactor operation are planned: a test stage and a research stage. This paper considers the research stage of the MTRR-SCW experimental reactor operation, the purpose … Two stages of the MTRR-SCW reactor operation are planned: a test stage and a research stage. This paper considers the research stage of the MTRR-SCW experimental reactor operation, the purpose of which is to investigate current and advanced light-water reactors. The MTRR-SCW driver-type core provides a fast neutron spectrum with the possibility for the local warmup in ampoule devices and independent loop channels. Irradiation channels will be installed in the core center and periphery, as well as instead of the reactor’s changeable reflector cartridges. The MTRR-SCW irradiation channels and independent loops will provide ample opportunities both for undertaking a research on effects from neutron irradiation of different materials, and for testing a variety of fuel assembly designs and operating conditions (temperature, pressure, neutron spectrum), as well as for investigating transients and emergency processes. The MTRR-SCW channels can be used to irradiate different types of fuel, and structural and absorbing materials with different coolant inlet temperatures (from 250 to 450 °C) and, consequently, its inlet density (from 800 to 100 kg/m 3 respectively), providing different neutron spectrum options for the experimental fuel assembly in a range from thermal to fast spectrum. The MTRR-SCW allows experiments to increase power and simulate emergency processes, including reactivity accidents (RIA). The strong primary and safeguard vessels of the independent loop channels also make it possible to simulate loss-of-pressure emergencies of the LB LOCA and SB LOCA types. The peripheral independent loop channel will allow undertaking experiments for simulation of alternative reactor concepts with reactors with SKD coolant parameters, such as a single-circuit concept with the pseudophase transition in the core (VVER-SKD-1700), and with natural coolant circulation in the core (SKDI). In addition, the peripheral channel allows accelerated irradiation of fuel rods used in current VVER reactors, taking into account the reproduction of the ratio between damaging dose and burnup rates.

Research of the few-group cross-sections calculation based on the software package MOSASAUR

2025-06-19

Chen Zhao , Lianjie Wang , Lei Lou +3 more | Progress in Nuclear Energy

Design of a High-Assay Low-Enriched Uranium Tri-Structural Isotropic Critical Experiment for Advanced Reactor Validation

2025-06-18

Holly Trellue , Theresa Cutler , Erik Luther +7 more | Nuclear Technology

Power Start-Up of the IVG.1M Reactor with Low-Enriched Uranium Fuel: Main Results

2025-06-18

Erlan Batyrbekov , Vladimir Vityuk , Viktor Baklanov +7 more | Energies

In support of global efforts to strengthen the nuclear non-proliferation regime, the IVG.1M research water-cooled thermal reactor at the National Nuclear Center of the Republic of Kazakhstan was successfully converted … In support of global efforts to strengthen the nuclear non-proliferation regime, the IVG.1M research water-cooled thermal reactor at the National Nuclear Center of the Republic of Kazakhstan was successfully converted to low-enriched uranium (LEU, 19.75% 235U) fuel in 2023. The reactor’s operability with innovative bimetallic, fiber-type, dual-blade LEU fuel rods was experimentally verified during power start-up experiments. The test program included investigations of power distribution in the core, evaluation of temperature, power, and hydrodynamic reactivity effects, and the measurement of fission product release to the coolant. The results were in good agreement with safety calculations, confirming that the enrichment reduction did not degrade reactor performance characteristics. It was shown that the power reactivity effect increased by more than 1.5 times at a power level of 9 MW. The measured temperature reactivity coefficient (≈0.021 βeff/°C) and the level of fission product release remained within acceptable and expected limits.

Accurate and scalable receiver-level flux prediction: A fully data-driven solution

2025-06-18

Mathias Kuhl , Max Pargmann , Daniel Maldonado Quinto +1 more | Solar Energy

Analytical technique of multi-energy groups point kinetics model for multidimensional homogeneous reactors

2025-06-18

Ahmed E. Aboanber , Abdallah A. Nahla , A.M. Khallaf | Indian Journal of Physics

Abstract The solution of the balancing equation of the neutron flux coupled with the delayed neutron precursor concentrations is the fundamental neutronic problem of any transient analysis. A finite difference … Abstract The solution of the balancing equation of the neutron flux coupled with the delayed neutron precursor concentrations is the fundamental neutronic problem of any transient analysis. A finite difference solution to a two- or three-dimensional dynamics issue necessitates an incredibly high number of calculations, highlighting the need for more straightforward approaches that may be used effectively in ordinary studies. The point reactor model, discussed and used in numerous literature, is the simplest of these techniques. This paper provides an analytical technique for the multi-energy groups point kinetics model, which is at the core of multidimensional homogeneous reactors. Furthermore, like with the quasi-static method, this analytical technique is based on variable separation, and a matrix approach is used to formulate the stiff coupled differential equations system. The exponential function of a coefficient matrix can be represented as a polynomial function with variable coefficients. These coefficients can be determined analytically by utilizing the eigenvalues of the coefficient matrix. The suggested method applies to 2D and 3D homogeneous reactors with various forms of linear, sinusoidal, and pulse reactivity, including nonlinear reactivity, through temperature feedback. The numerical data comparison using analytical methods enhanced accuracy and efficacy, showing strong agreement with conventional and benchmark techniques.

Efficient Acceleration in Solving the 2D Neutron Diffusion Equation with CUDA: Exploring the Collaborative Practicality of Colab<i></i>

2025-06-18

Paulo Igor de Oliveira Pessoa , Edson Henrice | Brazilian Journal of Radiation Sciences

This paper explores an approach to accelerate the finite difference method applied to solving the two-dimensional (2D) neutron diffusion equation for two energy groups (2G) independent of time. The main … This paper explores an approach to accelerate the finite difference method applied to solving the two-dimensional (2D) neutron diffusion equation for two energy groups (2G) independent of time. The main innovation lies in the implementation of a performance optimization method, emphasizing the practicality of development in Python using direct browser collaboration through Google Colaboratory (Colab). Utilizing CUDA (Compute Unified Device Architecture) for GPU acceleration, we achieve significant computational performance improvements. The study compares Python implementations using CuPy and NumPy libraries with traditional FORTRAN implementations utilizing the LAPACK library, highlighting the efficiency and precision of GPU-accelerated calculations. Results show that Python with CuPy significantly outperforms NumPy, both in a Colab environment and on a personal desktop computer. This demonstrates the practicality of cloud-based solutions for intensive computations, as the ability to run code directly in the browser through Colab eliminates the need for extensive local hardware resources. The results emphasize the convenience of executing complex simulations without relying on physical computers, promoting greater flexibility and accessibility in computational research. All computational codes are available on GitHub for transparency and reproducibility.

Investigation on coupling methods for heat transfer field between a reactor core and heat pipes

2025-06-17

Yihu Wang , Ziang Guo , Tenglong Cong +2 more | Annals of Nuclear Energy

Genetic Algorithm-Based Optimization of Online Diesel Fuel Upgrading Process for Nuclear Power Emergency

2025-06-17

Lanqi Zhang , Hao Li , Fengyi Liu +3 more | Applied Sciences

To enhance the oxidative stability of aging diesel fuel stored in nuclear power emergency systems, we propose a novel hybrid optimization framework that integrates a Genetic Algorithm (GA), State-Space Network … To enhance the oxidative stability of aging diesel fuel stored in nuclear power emergency systems, we propose a novel hybrid optimization framework that integrates a Genetic Algorithm (GA), State-Space Network (SSN) modeling, and Computational Fluid Dynamics (CFD) simulation. Unlike previous studies that address treatment efficiency, flow optimization, or simulation separately, our method achieves real-time, simulation-informed optimization by embedding CFD-based performance evaluation directly into the GA fitness function. The SSN is employed to construct a comprehensive superstructure of feasible conditioning paths, which are dynamically explored and optimized by the GA under flow and boundary constraints. The CFD model, implemented via Ansys Fluent, accurately simulates the antioxidant mixing process in the tank and provides feedback on concentration uniformity at key monitoring points. The results demonstrate that the proposed framework reduces the conditioning time by 5.38% and significantly enhances the additive distribution uniformity. This work offers a generalizable approach for intelligent diesel upgrading in high-reliability energy systems and contributes a structured pathway for integrating data-driven optimization with physical process simulation.

An integrated framework for the prediction of axial offset anomaly in pressurized water reactors

2025-06-17

Zhiqiang Zhang , Jinhe Wang , Lei Song +2 more | Deleted Journal

During the operation of pressurized water reactors, erosion products can be deposited on the surfaces of the fuel rods, and these deposits lead to enrichment of boric acid, which influences … During the operation of pressurized water reactors, erosion products can be deposited on the surfaces of the fuel rods, and these deposits lead to enrichment of boric acid, which influences the power distribution along the axial direction and results in axial offset anomaly (AOA)/Chalk River unidentified deposit (CRUD)-induced power shift (CIPS). In this paper, a fully integrated CIPS risk prediction framework is proposed. The framework comprises three coupled modules: CRUD deposition prediction, boron concentration calculation, and neutron flux calculation. These modules are dynamically coupled through CRUD deposition distribution and boric acid concentration data, ensuring comprehensive and up-to-date predictions. To demonstrate the framework’s capabilities, we provide a case study involving a 500-day reactor operation. Our results indicate a maximum CRUD thickness approaching 40 μm, primarily located in the upper half of the fuel rod. This deposition corresponds to a maximum boric acid concentration ratio exceeding 300, which significantly reduces heat flux in the affected region and precipitates a severe AOA/CIPS incident.

Criticality and sensitivity/uncertainty analysis of the DNRR with LEU fuel using MCNP6.3 and the latest data libraries

2025-06-17

D. Dau , Nguyen Nhi Dien , Ton-Nghiem Huynh +7 more | Nuclear Engineering and Design

Predictive density profile control with discrete pellets, applied to integrated simulations of ITER

2025-06-17

Christopher A. Orrico , M. van Berkel , T.O.S.J. Bosman +4 more | Nuclear Fusion

Abstract Reliable core density control with pellet fueling will be necessary to achieve required fusion power output in future fusion tokamaks such as ITER. The discrete nature of fuel pellets, … Abstract Reliable core density control with pellet fueling will be necessary to achieve required fusion power output in future fusion tokamaks such as ITER. The discrete nature of fuel pellets, however, complicates the density profile control problem significantly. As a solution, we propose a predictive density profile controller that considers fuel pellets as discrete actuators, while ensuring operation within prescribed density limits. The model predictive control (MPC) scheme we deploy combines the offset-free method to correct prediction model inaccuracies and our novel modified penalty term homotopy algorithm for real-time MPC (PTH-MPC). To demonstrate density profile control with discrete pellets, we couple the PTH-MPC density controller with JINTRAC integrated simulations of the ITER 15 MA/5.3 T scenario, using HPI2 to model discrete pellet ablation and deposition. We compare the density controller performance in integrated simulations using the Bohm/gyro-Bohm turbulent transport model against integrated simulations using the TGLF turbulent transport model. We highlight the necessity of treating pellets as discrete events for controller performance and for remaining within density limits. We conclude that PTH-MPC is a promising candidate for density profile control with pellets fueling in ITER and other future tokamaks and recommend further improvements using learning-based and robust MPC. We also note the limitations of quasi-linear turbulent transport models in simulations involving discrete pellets.

Neutronics calculation and analysis of a small lead-bismuth fast reactor core with control rods integrated within fuel assemblies

2025-06-13

Jiawei Gao , Haoyu Fan , Xianan Du +2 more | Annals of Nuclear Energy

Development, Validation, and Comparison of an OpenMC-Based Transport-Depletion Model for the Moroccan TRIGA Mark II with MCNP6.2

2025-06-13

Mohamed Mira , O. El Hajjaji , T. El Berdouni +5 more | Nuclear Technology

An experimental approach to the calculate of air-to-water dose ratio in Tehran Research Reactor (TRR) spent fuel using TL dosimeter

2025-06-12

Ruhollah Adeli | Nuclear Engineering and Technology

Predictive accuracy of neutronics of light-water-moderated and graphite-reflected core with low-enriched uranium fuel at UTR-KINKI

2025-06-12

Hiroshi Shiga , Kenichi Watanabe , Masato Tabuchi +3 more | Journal of Nuclear Science and Technology