James M. Adams

National Institute of Standards and Technology

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A.D. Carlson

Brookhaven National Laboratory

S. M. Grimes

Ohio University

Shane R. Gardner

Pennsylvania State University

T. N. Massey

Ohio University

Alireza Haghighat

Virginia Tech

Michael T. Wenner

University of Florida

Apisit Patchimpattapong

Electricity Generating Authority of Thailand

Vefa Kucukboyaci

Westinghouse Electric (Japan)

David M. Gilliam

National Institute of Standards and Technology

J. S. Nico

National Institute of Standards and Technology

Cooperative Institutions

Pennsylvania State University

Institute for Nuclear Research

National Institute of Standards and Technology

Westinghouse Electric (United States)

Russian Academy of Sciences

National Institute of Standards

University of Florida

Los Alamos National Laboratory

Louisiana State University

University of Pennsylvania

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Operating Margin of Soviet RBMK-1000 Nuclear Power Reactors

Nuclear Technology (1991)

James M. Adams L. E. Hochreiter G.E. Robinson

A coupled thermal-hydraulic analysis is performed for the Soviet-designed RBMK-1000 nuclear power reactor to assess the operating margin to critical heat flux (CHF); the Chernobyl-4 reactor serves as the principal model for this study. Calculations are performed using a simplified subchannel analysis.The overall analysis involves an iterative search to determine the individual subchannel flow rates, and a boiling transition analysis is performed to obtain a measure of the core operating margin. The operating margin is determined via two distinct methods. The first involves a calculation of the core critical power ratio (CPR) using an empirically derived correlation that the Soviets developed expressly for the RBMK-1000. Additionally, various subchannel CHF correlations typical of those used in the design of nuclear-powered reactors in the United States are also employed.When the Soviet critical power correlation is used, the calculations carried out for both normal operating and reference overpower conditions result in CPRs of 1.155 and 1.019, respectively. In most cases, the subchannel CHF correlations indicate that additional operating margin over that calculated by the Soviet critical power correlation exists for this design.

Margin (machine learning)

Boiling water reactor

RBMK

Thermal hydraulics

10.13182/nt91-a34596

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Final Report - Novel investigation of iron cross sections via spherical shell transmission measurements and particle transport calculations for material embrittlement studies.

S. M. Grimes T. N. Massey A.D. Carlson James M. Adams Alireza Haghighat

OAK B204 We have been pursuing a multi-year project, funded by the U.S. Department of Energy, to study neutron scattering interactions in iron. The principal objective of this work is to investigate the well-known deficiency that exists for reactor pressure vessel neutron fluence determinations. Specifically, we are using the spherical-shell transmission method, employing iron shells with different thicknesses, and neutron time-of-flight (TOF) measurements of the scattered neutrons, in an effort to precisely determine specific energy regions over which deficiencies in the non-elastic scattering cross section for neutron scattering in iron appear to exist.

Embrittlement

Spherical shell

Neutron Transport

Particle (ecology)

10.2172/810274

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2023 NIST Center for Neutron Research Accomplishments and Opportunities

James M. Adams Steven R. Kline

NIST

Center (category theory)

Research center

10.6028/nist.sp.1294

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New Particle Transport Methods for Design and Optimization of Spherical-Shell Transmission Measurements

Journal of Nuclear Science and Technology (2002)

Alireza Haghighat Shane R. Gardner Vefa Kucukboyaci Michael T. Wenner James M. Adams

This paper discusses new particle transport methods developed for accurate investigation of iron non-elastic scattering cross sections using the spherical-shell transmission method, employing iron shells with different thicknesses, and neutron time-of-flight spectroscopy of the scattered neutrons. New calculational techniques based on the deterministic and Monte Carlo methods have been developed for design and optimization of the experiment, and for identification and reduction of experimental uncertainties. The new methods include a new tallying option for the MCNP code and new quadrature sets for the 3-D parallel Sn code PENTRAN. The new tallying option is used to determine the optimum source energy versus target thickness, and among the new quadrature sets, the Pn-Tn with ordinate-splitting technique has resulted in accurate flux distributions as compared to Monte Carlo prediction.

Neutron Transport

10.1080/00223131.2002.10875274

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The thermal neutron response of miniature silicon carbide semiconductor detectors

Nuclear Instruments and Methods in Physics Research Section A Accelerators Spectrometers Detectors and Associated Equipment (2003)

A.R. Dulloo F.H. Ruddy J.G. Seidel James M. Adams J. S. Nico

Neutron Detection

NIST

Bonner sphere

Semiconductor detector

10.1016/s0168-9002(02)01987-3

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Citations (74)

IMPROVING NEUTRON SOURCE CALIBRATIONS AT THE NATIONAL INSTITUTE OF STANDARDS AND TECHNOLOGY

James M. Adams David M. Gilliam

10.1142/9789812705563_0085

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A New Investigation of Iron Cross Sections via Spherical-Shell Transmission Measurements and Particle Transport Calculations

Journal of Nuclear Science and Technology (2002)

James M. Adams Arzu Alpan A.D. Carlson Shane R. Gardner S. M. Grimes

We are engaged in a multi-year project to study neutron scattering interactions in iron, the principal objective of which is to investigate the well-known deficiency that exists in reactor pressure vessel neutron fluence determinations. Specifically, we are using the spherical-shell transmission method, employing iron shells with different thicknesses, and neutron time-of-flight measurements of the scattered neutrons, in an effort to precisely determine specific energy regions over which deficiencies in the non-elastic scattering cross section for neutron scattering in iron appear to exist. The analysis of the experimental data involves correlating the data with theoretical calculations of neutron transport through the iron spheres in order to evaluate the degree to which the calculated neutron spectra predict the measured spectra relative to different types of particle interactions. In doing so, we have developed new methodologies for performing neutron transport calculations that will be useful to a range of transport problems. Preliminary results show good agreement between the experimental data and the calculated distribution of neutron flight times over much of the data range, except for the contribution due to breakup neutrons.

Neutron Transport

10.1080/00223131.2002.10875242

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Citations (1)

Novel Investigation of Iron Cross Sections via Spherical Shell Transmission Measurements and Particle Transport Calculations for Material Embrittlement Studies

Nuclear Science and Engineering (2012)

Michael T. Wenner Alireza Haghighat James M. Adams A.D. Carlson S. M. Grimes

AbstractWe have carried out a multifaceted research project to improve our knowledge of the iron nonelastic scattering cross sections. Spherical shell transmission measurements were made using time-of-flight techniques with neutrons from the 15N(p,n)15O and D(d,n)3He source reactions. For the 15N(p,n)15O work, measurements were made with a proton energy of 5.1 MeV. Measurements were made from 3 to 7-MeV deuteron energy for the D(d,n)3He work. For both source reactions, the angular range was as large as 15 to 135 deg. Two shell thicknesses were used. Comparisons are given between Monte Carlo predictions and experimental data.Utilizing a new tallying option, the estimated total iron cross sections at energies corresponding to the peak of the spectra for the 0-deg experiments were calculated to within 1% of the data in the ENDF/B-VII library. A processing code was developed to adjust ENDF format files to obtain closer agreement between measurements and calculations. Sensitivity analyses were performed at energies corresponding to the 0-deg beam angle neutrons. Using cross sections where the nonelastic and elastic cross sections were adjusted while constraining the total cross section to be constant, differences between experiment and calculation were reduced by ~;40% for a pressure vessel calculation. Such fluence calculations with adjusted cross sections indicate possible underestimation of neutron fluence, and therefore more material damage.

Nuclear data

10.13182/nse09-30

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Citations (8)

A New Monte Carlo Tallying Methodology for Optimizing the Nuclear Energy Research Initiative Spherical-Shell Transmission Experiment

Shane R. Gardner A. Haghighat Apisit Patchimpattapong James M. Adams A.D. Carlson

The development of a new tallying methodology for use with the MCNP Monte Carlo particle transport code is reported. The objective is to be able to determine with accuracy the iron nonelastic scattering cross sections via a spherical-shell transmission experiment using the time-of-flight (TOF) method. Different source energy and shell thickness combinations are examined. The actual transmission experiment and the new tallying algorithm are described, and the verification of the new tallying algorithm and its utilization to optimize a neutron TOF experiment are discussed. The results of the work will be used to alleviate the well-known deficiency that exists in the iron nonelastic scattering cross section and thereby reduce the uncertainty in calculations involving deep penetration in iron, e.g. reactor pressure vessel neutron fluence determinations.

Neutron Transport

Source

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The role of neptunium-237 in reactor pressure vessel dosimetry

Radioactivity & radiochemistry (1995)

James M. Adams E.D. McGarry

The purpose of reactor pressure vessel (RPV) dosimetry is to monitor the neutron fluence responsible for material embrittlement brought about by damage induced by fast neutron irradiation. Assessment of the damage enables one to define appropriate operating limits necessary to help ensure the operational integrity and safety of nuclear power reactors. Although a variety of methods exist for assessing the dose delivered to an object exposed to a radiation field, the nuclear industry has traditionally employed integral dosimeters (activation foils) as a means of tracking the total dose delivered at various locations about the RPV. Currently, we are investigating the accuracy of the {sup 237}Np (n,f) nuclear reaction for assessing the fast-neutron induced damage to RPVs. This reaction is attractive for two important reasons. Firstly, the response function of {sup 237}Np dosimeters is operative over (nearly) the entire neutron energy range of interest in irradiation-damage studies. Secondly, this reaction results in the production of long-lived {sup 137}Cs fission product activity. Hence, the integration capacity of these dosimeters will not saturate during reactor exposure, which can last for several years. In conjunction with this work, the authors have developed a high-flux reference neutron field - the Materials Dosimetry Reference Facility -more » at the University of Michigan Ford Nuclear Reactor. This facility enables the authors to investigate the response of RPV dosimeters in a well-characterized neutron field that is comparatively intense and possesses a spectral distribution representative of that observed in power reactors. Current research will be discussed.« less

Embrittlement

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