Dosimetry Reaction Rates Research Articles

Investigation on A-FNS neutron spectrum monitor system

We proposed the neutron spectrum measurement system for Advanced Fusion Neutron Source (A-FNS). It is configured with fission chambers, Self-Powered Neutron Detectors, neutron activation systems using water flow, and activation monitors. In this study, we mainly investigated the multiple activation foils which are sensitive to high energy neutrons. We selected 14 main dosimetry threshold reactions based on specific A-FNS usage conditions. The reactions are as follows: 55Mn(n,2n)54Mn, 59Co(n,p)59Fe, 59Co(n,2n)58Co, 59Co(n,3n)57Co, 75As(n,2n)74As, 89Y(n,2n)88Y, 93Nb(n,2n)92mNb, 169Tm(n,2n)168Tm, 169Tm(n,3n)167Tm, 197Au(n,2n)196Au, 197Au(n,3n)195Au, 209Bi(n,3n)207Bi, 209Bi(n,4n)206Bi, and 209Bi(n,5n)205Bi. The dosimetry reaction rates were calculated using Monte Carlo code MCNP-5.15 with McDeLicious-11 and nuclear data libraries FENDL-3.1c, IRDFF-1.05 and EAF-2010. As for neutron-induced cross section data of the reactions in the high energy region, we clarified that there were limited experimental data, and proposed new experiment for their evaluations.

Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Boiling Water Reactor Benchmark Problem

This paper describes a comparison of contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a boiling water reactor calculational benchmark problem. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission. The benchmark problem was originally evaluated by Brookhaven National Laboratory using the Oak Ridge National Laboratory discrete ordinates code DORT and the BUGLE-93 cross-section library. In this paper, the Westinghouse RAPTOR-M3G three-dimensional discrete ordinates code was used. A variety of cross-section libraries were used with RAPTOR-M3G including the BUGLE93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory and ALPAN-VII.0 developed at Westinghouse. In comparing the calculated fast reaction rates using the four aforementioned cross-section libraries in the pressure vessel capsule, for six dosimetry reaction rates, a maximum relative difference of 8% was observed. As such, it is concluded that the results calculated by RAPTOR-M3G are consistent with the benchmark and further that the different vintage BUGLE cross-section libraries investigated are largely self-consistent.

Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Pressurized Water Reactor Standard Core Loading Benchmark Problem

This paper compares contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a pressurized water reactor calculational benchmark problem with a standard out-in core loading. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission and used the Oak Ridge National Laboratory two-dimensional discrete ordinates code DORT and the BUGLE-93 cross-section library for the calculations. In this paper, a Westinghouse three-dimensional discrete ordinates code with parallel processing, the RAPTOR-M3G code was used. A variety of cross section libraries were used with RAPTOR-M3G including the BUGLE-93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory, and the broad-group ALPAN-VII.0 cross-section library developed at Westinghouse. In comparing the calculation-to-calculation reaction rates using the BUGLE-93 cross-section library at the thermal shield, pressure vessel, and cavity capsules, for eleven dosimetry reaction rates, a maximum relative difference of 5% was observed, with the exception of 65 Cu(n,2n) in the pressure vessel capsule that had a 90% relative difference with respect to the reference results. It is thought that the 65 Cu(n,2n) reaction rate reported in the reference for the pressure vessel capsule is not correct. In considering the libraries developed after BUGLE-93, a maximum relative difference of 12% was observed in reaction rates, with respect to the reference results, for 237 Np(n,f) in the cavity capsule using BUGLE-B7.

Integral test of International Reactor Dosimetry and Fusion File with Li2O assembly and DT neutron source at JAEA/FNS

In order to validate a new library of dosimetry cross section data, International Reactor Dosimetry and Fusion File release 1.0 (IRDFF 1.0), not only for DT neutrons but also for neutrons with energy of less than 14MeV, we perform an integral test with a Li2O rectangular assembly of 60.7cm in thickness and a DT neutron source at JAEA/FNS. We place a lot of activation foils at depths of 10.1cm and 30.4cm for measurements of dosimetry reaction rates in small space along the central axis in the assembly, measure decay gamma-rays from the activation foils with high-purity Ge detectors after the DT neutron irradiation by the foil activation technique, and deduce a variety of dosimetry reaction rates. We calculate the reaction rates by using a Monte Carlo code MCNP5-1.40 and the nuclear data library ENDF/B-VII.1 with the IRDFF-v.1.05 as the response functions for the dosimetry reactions. The calculation results generally show good agreements with the measured ones, and it can be confirmed that most of the data in IRDFF-v.1.05 are valid for the neutron field in the Li2O assembly with the DT neutrons.

Integral test of International Reactor Dosimetry and Fusion File on graphite assembly with DT neutron at JAEA/FNS

We perform an integral test of a new library of dosimetry cross section data, International Reactor Dosimetry and Fusion File release 1.0 (IRDFF 1.0), which has been released from the International Atomic Energy Agency recently. A pseudo-cylindrical graphite assembly is used with a DT neutron source at the fusion neutronics source facility of Japan Atomic Energy Agency. A lot of foils for measurements of dosimetry reaction rates are placed in the small space along the central axis in the assembly. After the DT neutron irradiation, forty dosimetry reaction rates are measured and are compared with calculated values, which are obtained by using a Monte Carlo neutron transport code MCNP5-1.40 and a nuclear data library ENDF/B-VII.1 with the IRDFF-v.1.03 as the response functions for the dosimetry reactions. The calculated results generally show good agreements with the experimental ones. The cross section data of the 68Zn(n,np)67Cu and 68Zn(n,d)67Cu reactions and the 113In(n,γ)114mIn reaction, which are not included in IRDFF-v1.03, are needed for the analysis of the 67Zn(n,p)67Cu reaction and the 115In(n,2n)114mIn reaction as the competitive reactions, respectively.

Benchmark experiment on titanium with DT neutron at JAEA/FNS

Titanium is contained in lithium titanate which is a tritium breeding material candidate. In the nuclear design, accurate nuclear data are needed. However, few benchmark experiments had been performed for titanium. We performed a benchmark experiment with a titanium assembly and a DT neutron source at JAEA/FNS. The titanium assembly was covered with Li2O blocks in order to reduce background neutrons. Dosimetry reaction rates were measured with niobium, indium and gold foils inside the assembly. And fission rates of 235U were measured by using micro fission chambers. This experiment was analyzed by using the Monte Carlo neutron transport code MCNP5-1.40 with recent nuclear data libraries of ENDF/B-VII.0, ENDF/B-VII.1, JEFF-3.1.2, JENDL-4.0 and JENDL-4.0u1. The calculation results were compared with the measured one in order to validate the nuclear data libraries of titanium. The calculated results with ENDF/B-VII.1 agreed with the measured one the best because the (n,2n) and (n,n′cont) reaction cross section data and resonance parameters were improved.

ENDF/B-VII.0, JEFF-3.1 and JENDL-4 Iron and Water Cross Sections Analysis Using the PCA-REPLICA Shielding

An accurate calculation of neutrons penetration through the side-shield of a Pressurized Water Reactor (PWR) is important for the estimation of nuclear heating in the thermal shields and in the Reactor Pressure Vessel (RPV), also for estimate the leakage of the particles in the reactor cavity. The cross section data and the calculation method have an important role in this kind of problems. For this object the PCA-REPLICA benchmark where the neutron transports through typical PWR were performed at AEA technology Winfrith, UK (United Kingdom). This work aims to contribute validation of the method of calculation and analyse the cross sections, for Iron ( Fe 56 ) and water ( H 1 and O 16 ), presented by: ENDF/B-VII.0, JEFF-3.1 and JENDL-4 libraries, for the PCA-REPICA benchmark, using the Monte Carlo approach by the mean of MCNPX code. The benchmark analysis was made based on the calculated-to-measured (C/M) dosimetry reaction rates of In 103 monitors, at different depth in a water/iron shield reproducing the ex-core radial geometry of a PWR. The calculations of PCA-REPLICA experiment showed us that calculation method is effective for the protection study of the REP. And generally, the average C/M ratios obtained for the three detectors are reasonably good when the uncertainties of the measurements are taken into account.

Benchmark on the 3-D VENUS-2 MOX-Fueled Reactor Dosimetry Calculations by DANTSYS Code System

Abstract The benchmark calculation for the 3-D VENUS-2 MOX-fueled reactor dosimetry has been performed by the TRANSX/DANTSYS code system. The nuclear data from the ENDF/B-VI.8, JENDL-3.3, and JEFF-3.0 have been processed for the MATXS-format 199-group neutron libraries by NJOY99.90. The equivalent fission fluxes at the 34 measurement positions on the core mid-plane have been calculated using the dosimetry reaction rates for the 58Ni(n,p), 115In(n,n'), 103Rh(n,n'), 64Zn(n,p), 237Np(n,f), and 27Al(n,α) detectors. The estimated results by the JENDL-3.3 and JEFF-3.0 have been compared with those by the ENDF/B-VI.8.

Benchmark experiment on silicon carbide with D–T neutrons and validation of nuclear data libraries

Silicon carbide (SiC) is one of the low activation structural materials for fusion reactors. A benchmark experiment on SiC was conducted at the deuterium–tritium (D–T) neutron source facility FNS in JAERI, and the validity of the cross section data for SiC was investigated. An experimental assembly made of sintered SiC (457×457 mm, and 711 mm in thickness) was bombarded by D–T neutrons, and neutron spectra, gamma-ray spectra, dosimetry reaction rates and gamma-ray heating rates were measured in the assembly. Computational analyses of the experiment were performed by Monte Carlo transport calculations with the MCNP-4B code and several recent evaluated nuclear data files. The following findings were pointed out. The calculations with JENDL Fusion File and JENDL-3.2 predicted adequately the neutron fluxes from 14 MeV to 0.3 eV as well as the gamma-ray fluxes at any positions up to ∼600 mm depth in the SiC assembly. Results for the calculation with FENDL/E-2.0 were similar to those with the two JENDL calculations, but the neutron fluxes in the energy range from 1 to 14 MeV were calculated slightly smaller. The cross section data for silicon in FENDL/E-1.0 were found inadequate for nuclear design calculations.

Benchmark Experiment on Vanadium with D-T Neutrons and Validation of Evaluated Nuclear Data Libraries by Analysis of the Experiment

A neutronics benchmark experiment on vanadium, which is a low activation fusion reactor material, was conducted by using the D-T neutron source facility of FNS/JAERI. Neutron spectra, dosimetry reaction rates, γ-ray spectra and γ-ray heating rates were measured in a vanadium experimental assembly. Benchmark tests for four evaluated nuclear data files were performed by analyzing the experiment. As a result, the following problems were pointed out in view of accuracy of fusion reactor designs. (1) JENDL-FF and JENDL-3.2: Total cross section should be reexamined especially at ∼2keV. (2) ENDF/B-VI: Double differential cross sections for 14 MeV neutrons should be revised because of the isotropic angular distribution for continuum neutron emission. Gamma-ray production cross sections are too small and discrete γ-ray peaks are not represented clearly. (3) EFF-3: Gamma-ray production cross sections are too large.

Measurement of Gamma-Ray Spectra and Heating Rates in Iron and Stainless Steel Shields Bombarded by Deuterium-Tritium Neutrons and Validation of Secondary-Gamma-Ray Data in Evaluated Nuclear Data Libraries

Benchmark experiments for the validation of secondary-gamma-ray data are conducted for iron and Type 316 stainless steel (SS316) shield assemblies bombarded by deuterium-tritium neutrons. Gamma-ray spectra and heating rates for both threshold and capture gamma rays are measured. With the present experimental data for gamma rays, a set of benchmark data for iron and SS316, including neutron energy spectra in entire energies and various dosimetry reaction rates, is completed for the first time. Secondary-gamma-ray data in JENDL-3.1, JENDL-3.2, JENDL-Fusion File, and FENDL/E-1.0 are tested by benchmark calculation of the experiments. As a result, larger gamma-ray-production cross sections for threshold reactions in JENDL-3.1 and JENDL-3.2 and an inconsistent energy balance of the (n, γ) reactions in JENDL-3.1 are found. From the viewpoint of fusion engineering, the first priority in evaluating secondary-gamma-ray data should be conserving the energy balance. A rigid energy balance in both the JENDL Fusion File and FENDL/E-1.0 is confirmed for both threshold and capture gamma rays. The JENDL Fusion File and FENDL/E-1.0 provide the highly accurate secondary-gamma-ray data for iron and SS316 needed for fusion reactor nuclear design.

Cadmium Cutoff Energy for MCNP Modeling of Dosimetry Reactions

Reactor pressure vessel neutron fluence calculations are verified by comparing calculated reaction rates of dosimetry foils assumed to be placed in the cavity surrounding the pressure vessel with corresponding experimentally determined dosimetry reaction rates at the same location. The irradiated foils used are bare and cadmium covered. The cadmium-covered foils see a different neutron flux than the uncovered foils. Computer models computing reaction rates for the foils must approximate the effect of the cadmium foil cover. The MCNP code modeling of reactor cavity dosimetry reactions to validate the computed neutron flux hitting the pressure vessel of a pressurized water reactor (PWR) requires the choice of an appropriate cadmium cutoff energy. Since the neutron spectrum is harder in the cavity than in the core, the same cadmium cutoff energy may not be universally applied. The correct cutoff energy for the cavity of a PWR was computed by using MCNP4a and a representative cavity spectrum. Four cadmium-covered foil reactions were analyzed, and the appropriate energy cutoff was determined to be [approximately] 0.6 eV for [approximately]0.5-mm-thick cadmium covers.