Validation of MVP-II with JENDL-4.0, ENDF/B-VII.1, and JEFF-3.2 through HTC critical experiments

In order to validate MVP-II, Haut Taux de Combustion (HTC) experiments were analyzed using a code with relatively new nuclear data libraries, JENDL-4.0, ENDF/B-VII.1, and JEFF-3.2. The effective neutron multiplication factor k_eff values were obtained through analyses of all phases of the HTC experiments. Consequently, the k_eff biases evaluated for each nuclear data library were within 300 pcm. Additionally, microscopic production and capture reaction rates of major actinide isotopes were analyzed to substantiate differences among the libraries for a representative case of Phase 1 of the HTC experiments. The analysis showed that microscopic cross sections of ²³⁸Pu and ²⁴¹Am in JEFF-3.2 were somewhat large compared to those of ENDF/B-VII.1 and JENDL-4.0 for the representative case of Phase 1.     

Analysis of Core Physics Experiments on Fresh and Irradiated PWR UO2 Fuels in the REBUS Program

Critical experiments of two cores each loaded with fresh 5 × 5 test PWR-type fuel rods of ²³⁵U enrichment of 3.8 wt% or irradiated 5 × 5 test rods of rod average burnup of 55 GWd/t in the REBUS program were analyzed using diffusion, transport, and continuous-energy Monte Carlo calculation codes coupled with nuclear data libraries based on JENDL-3.2 and JENDL-3.3. Biases in effective multiplication factors k _eff's of the critical cores were about ?1:2%Δk for the diffusion calculations (JENDL-3.2), ?0:5%Δk for the transport calculations (JENDL-3.3), and ?0:5 and 0.1%Δk for the Monte Carlo calculations (JENDL-3.3 and JENDL-3.2, respectively). The measured core fission rate and Sc- or Co-activation rate distributions were generally well reproduced using the three types of calculation. The burnup reactivity determined using the measured water level reactivity coefficients was ?2:35 ± 0:07Δk/kk′. The calculated result of the Monte Carlo calculations agreed with it; however, the diffusion and transport calculations overestimated the absolute value by about 7%, which would be mainly attributed to the errors in the calculation of the reactivity caused by changing the fuel compositions from fresh fuel to irradiated fuel.     

Neutron cross section sensitivity analysis of UO2 and MOX core physics experiments on light water reactors

The perturbation theory based on the transport calculation has been applied to study sensitivity of neutron multiplication factors (k_eff's) to neutron cross sections used for the reactivity analysis of UO₂ and MOX core physics experiments on light water reactors. The studied cross sections were neutron capture, fission and elastic scattering cross sections, and a number of fission neutrons, ν. The obtained sensitivities were multiplied to relative differences in the cross sections between JENDL-4.0 and JENDL-3.3 in order to estimate the reactivity effects. The results show that the increase in k_eff, 0.3%Δk/kk′, from JENDL-3.3 to JENDL-4.0 for the UO₂ core is mainly attributed to the decreases in the capture cross sections of ²³⁸U. On the other hand, there are various contributions from the differences in the cross sections of U, Pu, and Am isotopes for the MOX cores. The major contributions to increase in k_eff are decreases in the capture cross sections of ²³⁸U,²³⁸Pu, ²³⁹Pu, and those to decrease in k_eff are decreases in ν of ²³⁹Pu and increases in the capture cross sections of²⁴¹Am. They compensate each other, and the difference in k_eff between JENDL-3.3 and JENDL-4.0 is less than 0.1%Δk/kk′ and relatively small.     

Comparisons of Deterministic Neutronic Calculations with Monte Carlo Results for an Advanced BWR Fuel Assembly with Hafnium Control Blades

From the neutronic viewpoint, the optimization of BWR core designs is strongly related to the accurate determination of flux variations inside and around fuel assemblies. These fluctuations, which are mainly due to the high heterogeneity of the fuel and moderator regions, as additionally to the presence of cruciform absorber blades, have a direct impact on reactor safety and performance. Of particular importance is the pin power distribution, leading to the need of assessing the capabilities of design tools in a sufficiently rigorous manner. The basic configuration chosen for the code comparisons corresponds to a SVEA-96 fuel assembly under full-density water moderation conditions, with inserted hafnium absorber blades. The calculational schemes employed are the Monte Carlo code MCNPX2.5, in conjunction with various nuclear data libraries (ENDF/B-VI, JEF2.2, JEFF3.0, JENDL-3.2 and JENDL3.3), and the deterministic codes CASMO4 with JEF2.2, BOXER with JEF1.0 and HELIOS 1.6 with ENDF/B-VI based libraries, respectively. The significant discrepancies observed in k _∞ predictions (>500pcm) are found to be mainly nuclear data related. On the other hand, data library effects have been found to be quite small for the prediction of pin-wise distributions of total fissions (F_tot), ²³⁸U captures (C8), as also of the C₈=F_tot ratio. Significant differences in these reaction rate distributions (up to several percent) have, however, been observed between the Monte Carlo and deterministic calculations, particularly in the vicinity of the hafnium blades and in the gadolinium pins.     

Critical and Subcritical Mass Calculations of Curium-243 to -247 Based on JENDL-3.2 for Revision of ANSI/ANS-8.15

Critical and subcritical masses were calculated for a sphere of five curium isotopes from ²⁴³Cm to ²⁴⁷Cm in metal and in metal-water mixtures considering three reflector conditions: bare, with a water reflector or a stainless steel reflector. The calculation were made mainly with a combination of a continuous energy Monte Carlo neutron transport calculation code, MCNP, and the Japanese Evaluated Nuclear Data Library, JENDL-3.2. Other evaluated nuclear data files, ENDF/B-VI and JEF-2.2, were also applied to find differences in calculation results of the neutron multiplication factor originated from different nuclear data files. A large dependence on the evaluated nuclear data files was found in the calculation results: more than 10%Δk/k relative differences in the neutron multiplication factor for a homogeneous mixture of ²⁴³Cm metal and water when JENDL-3.2 was replaced with ENDF/B-VI and JEF-2.2, respectively; and a 44% reduction in the critical mass by changing from JENDL-3.2 to ENDF/B-VI for ²⁴⁶Cm metal. The present study supplied basic information to the ANSI/ANS-8.15 Working Group for revision of the standard for nuclear criticality control of special actinide elements. The new or revised values of the subcritical mass limits for curium isotopes accepted by the ANSI/ANS-8.15 Working Group were finally summarized.     

Experimental Validation of Reaction Rate Distributions for a SVEA-96+ BWR Assembly with Hafnium Control Blades

The accurate estimation of reactor physics parameters related to the presence of cruciform absorber blades in boiling water reactors (BWRs) is important for safety assessment and for achieving flexible operation during the cycle. Characteristics that are affected significantly include distributions of the total fission (F_tot) and ²³⁸Ucapture (C₈) rates for controlled regions. Representative experimental investigations have been performed in the framework of the LWR-PROTEUS programme. In particular, the LWRPROTEUS I-2A experiments concerned the neutronics characterisation of a SVEA-96+ BWR assembly controlled with a hafnium (Hf) blade under full-density water moderation conditions. The current paper presents the comparisons of the measured F_tot and C₈ pin-wise distributions with a variety of stochastic and deterministic calculations: (a) MCNPX2.5 using recent nuclear data libraries (JEFF-3.1, ENDF/BVI. 8, and JENDL-3.3), (b) PHOENIX4 using ENDF/B-VI.3, (c) BOXER using JEF-1, (d) CASMO₄ using JEF-2.2, and (e) HELIOS1.6 using ENDF/B-VI.1. The calculation/experiment comparisons show standard deviations from 1.2% (MCNPX2.5) up to 1.9% (BOXER) for the prediction of the F_tot distribution, the highest individual discrepancy (7.6% with BOXER) being seen close to the “Hf-vertex.” The C₈ comparisons show systematically better agreement than those of F_tot, the lowest standard deviations being 1.0% (BOXER) and the highest only 1.4% (HELIOS). In addition, sensitivity studies highlight the greater importance of modelling aspects, compared with that of nuclear data libraries, for the achievement of satisfactory and validated F_tot and C₈ predictions.     

Analysis of Mixed Oxide Fuel Critical Experiments with Neutronics ANalysis Codes for Boiling Water Reactors

Critical experiments of UO₂ and full mixed oxide (MOX) fuel cores conducted at the Tank-type Critical Assembly (TCA) were aNalyzed using BWR design-purpose codes HINES and CERES with ENDF/B files and Monte Carlo fine analysis codes VMONT and MVP with the JENDL-3.2 library.

The averaged values of the multiplication factors calculated with HINES/CERES, VMONT and MVP agreed with those of experiments within 0.3%ΔAk. The values by the design-purpose codes showed a small difference of 0.1%Δk between UO₂ and MOX cores. Monte Carlo code results showed that the JENDL-3.2 library had a tendency to overestimate the multiplication factors of UO₂ cores by about 0.3%Δk compared with those values of MOX cores. The root mean square errors of calculated power distributions were less than 1% for HINES/CERES and VMONT.

These results showed that (1) the accuracy of these codes when applied to full MOX cores was almost the same as their accuracy for UO₂ cores, which confirmed the accuracy of present core design codes for full MOX cores; and (2) the accuracy of the 190-energy-group Monte Carlo calculation code VMONT was almost the same as that of the continuous-energy Monte Carlo calculation code MVP.     

Validation of Pb nuclear data by Monte Carlo analyses of sample reactivity experiments at Kyoto University Critical Assembly

Sample reactivity experiments on the uncertainty analyses of Pb nuclear data are carried out by substituting Al plates for Pb ones at the Kyoto University Critical Assembly, as part of basic research on Pb–Bi for the coolant. Numerical simulations of sample reactivity experiments are performed with the Monte Carlo calculation code MCNP6.1 together with four nuclear data libraries JENDL-3.3, JENDL-4.0, ENDF/B-VII.0 and JEFF-3.1, to examine the accuracy of cross-section uncertainties of Pb isotopes by comparing measured and calculated sample reactivities. A library update from JENDL-3.3 to JENDL-4.0 is demonstrated by the fact that the difference between Pb isotopes of the two JENDL libraries is dominant in the comparative study, through the experimental analyses of sample reactivity by the MCNP approach. In addition, JENDL-4.0 reveals a slight difference from ENDF/B-VII.0 in all Pb isotopes and ²⁷Al, and from JEFF-3.1 in ²³⁸U and ²⁷Al. Based on these results, further experiments are needed to investigate the uncertainties of Bi isotopes with the use of the Pb–Bi and Bi plates.     

Computational analysis of Bangladesh 3 MW TRIGA research reactor using MCNP4C,JENDL-3.3 and ENDF/B-Vl data libraries

The three-dimensional continuous energy Monte Carlo code MCNP4C was used to develop a versatile and accurate full-core model of the 3 MW TRIGA MARK II research reactor at Atomic Energy Research Establishment, Savar, Dhaka, Bangladesh. The model represents in detail all components of the core with literally no physical approximation. All fresh fuel and control elements as well as the vicinity of the core were precisely described. Validation of the JENDL-3.3 and ENDF/BVI continuous energy cross-section data for MCNP4C was performed against some well-known benchmark lattices. For TRIGA analysis, data from JENDL-3.3 and ENDF/B-VI in combination with the JENDL-3.2 and ENDF/B-V data files (for ^natZr, ^natMo, ^natCr, ^natFe, ^natNi, ^natSi, and ^natMg) at 300 K evaluations were used. Full S(α, β) scattering functions from ENDF/B-V for Zr in ZrH, H in ZrH and water molecule, and for graphite were used in both cases. The validation of the model was performed against the criticality and reactivity benchmark experiments of the TRIGA reactor. There is ∼20.0% decrease of thermal neutron flux occurs when the thermal library is removed during the calculation. Effect of erbium isotope that is present in the TRIGA fuel was also studied. In addition to the effective multiplication values, the well-known integral parameters: δ²⁸, δ²⁵, ρ²⁵, and C¹ were calculated and compared for both JENDL3.3 and ENDF/B-VI libraries and were found to be in very good agreement. Results are also reported for most of the analyses performed by JENDL-3.2 and ENDF/B-V data libraries.     

Detail analysis of fusion neutronics benchmark experiment for iron

In order to specify the best nuclear data on iron, the fusion neutronics benchmark experiment on iron at Japan Atomic Energy Agency (JAEA)/Fusion Neutronics Source (FNS) was analyzed in detail with MCNP-4C and the latest nuclear data libraries, JENDL-3.3, FENDL-2.1, JEFF-3.1 and ENDF/B-VII.0. As a result, totally the calculation result with ENDF/B-VII.0 agreed with the measurement best, except that it underestimated the measured neutron flux above 10 MeV with the depth. It was noted that the calculation result with JENDL-3.3 overestimated the measured neutrons below a few keV. Through the DORT calculations based on the iron data in ENDF/B-VII.0, it was found out that the first inelastic scattering cross-section data of ⁵⁷Fe in JENDL-3.3 caused the overestimation.     

Validation of Monte Carlo predictions of LWR-PROTEUS safety parameters using an improved whole-reactor model

The recent experimental programme conducted in the PROTEUS research reactor at the Paul Scherrer Institute (PSI) has concerned detailed investigations of advanced light water reactor (LWR) fuels. More than fifteen different configurations of the multi-zone critical facility have been studied, each of them requiring accurate estimation of operational safety parameters, in particular the critical driver loadings, shutdown rod worths and the effective delayed neutron fraction β_eff. The current paper presents a full-scale 3D Monte Carlo model for the facility, set up using the MCNPX code, which has been employed for calculation of the operational characteristics for seven different LWR-PROTEUS configurations. Thereby, a variety of nuclear data libraries (viz. ENDF/B6v2, ENDF/B6v8, JEF2.2, JEFF3.0, JEFF3.1, JENDL3.2, and JENDL3.3) have been used, and predictions of k_eff and shutdown rod worths compared with experimental values. Even though certain library-specific trends have been observed, the k_eff predictions are generally very satisfactory, viz. with discrepancies of <0.5% between calculation (C) and experiment (E). The results also confirm the consistent determination of reactivity variations, the C/E values for the shutdown (safety) rod worths being always within 5% of unity. In addition, the MCNP modelling of the multi-zone reactor has yielded interesting results for the delayed neutron fraction (β_eff) in the different configurations, a breakdown being made possible in each case in terms of delayed neutron group, fissioning nuclide, and reactor region.     

Measurement of Leakage Neutron Spectra from a Spherical Pile of Zirconium Irradiated with 14MeV Neutrons and Validation of Its Nuclear Data

In order to make a benchmark validation of the nuclear data for Zr, the leakage neutron spectrum from a Zr sphere of a 61-cm diameter was measured between 0.1 and 16MeV using a time-of-flight technique with a 14MeV neutron source facility, OKTAVIAN. The result was compared with the calculation using the Monte Carlo code MCNP-4A. To investigate the spectrum dependence on the individual neutron reactions, test calculations were carried out with the MCNP-4A code using the JENDL-3.2-based libraries, in which partial cross section values were reduced from the original values. From the comparison between the measured and the calculated spectra, it was found that each of the results could predict well the experiment in general. However, in detail, both ENDF/B-VI and EFF-2.4 gave considerable overestimation above 1 MeV. The JENDL-3.2 predicts the spectrum almost satisfactorily except below 0.8 MeV and around 10 MeV. The discrepancy found in JENDL-3.2 calculation is considered due to the cross section values of the (n, 2n) reaction and its secondary energy distributions (SED). The modified JENDL-3.2 library with the reduced (n, 2n) reaction values and the lower SED below 1 MeV reproduced the experiment with better agreement over the whole energy range.     

New investigations of the criticality property of pure 232U

This paper presents results of new calculations of criticality of ²³²U performed using the basic evaluated nuclear data files JENDL-3.2 (Japan) and ENDF/B-VI.5 (USA) using the NJOY-MCNP code system. Comparisons of these two basic nuclear data files have been done using PREPRO2000 code system and are presented. The critical mass of ²³²U calculated using ENDF/B-VI.5 and JENDL-3.2 respectively are 3.73 and 13.6 kg. The paper presents a mention of formation routes of ²³²U. A few remarks on the role of ²³²U in providing resistance to proliferation of fissile material with respect to utilizing thorium in thermal, fast, fusion and accelerator driven systems are made.     

Performance of the CENDL-3.2 and other major neutron data libraries for criticality calculations

Nuclear data are the cornerstones of reactor physics and shielding calculations.Recently,China released CENDL-3.2 in 2020,and the US released ENDF/B-Ⅷ.0 in 2018.Therefore,it is necessary to comprehensively evaluate the criticality computing performance of these newly released evaluated nuclear libraries.In this study,we used the NJOY2016 code to generate ACE format libraries based on the latest neutron data libraries（including CENDL-3.2,JEFF3.3,ENDF/B-Ⅷ.0,and JENDL4.0）.The MCNP code was used to ...     

Validation of an MCNP4B whole-reactor model for LWR-PROTEUS using ENDF/B-V,ENDF/B-VI and JEF-2.2 cross-section libraries

A detailed three-dimensional, continuous-energy MCNP4B model of the LWR-PROTEUS critical facility has been developed for the analysis of whole-reactor characteristics using ENDF/B-V, ENDF/B-VI and JEF-2.2 cross-section sets. The model has been applied to the determination of the critical loading, as well as the evaluation of reactivity worths for safety/shutdown rods, control rods, and individual driver-region fuel rods. The initially obtained results for the first configuration investigated (Core 1B) indicated that, for the same geometrical and materials specifications, the ENDF/B-V data library yields the closest critical prediction (discrepancy of 640±40 pcm), followed by ENDF/B-VI (980±40 pcm) and JEF-2.2 (1340±40 pcm). The differences in results between the three data libraries were studied by considering the contributions of individual materials to the neutron balance. ²³⁵U and ²³⁸U cross-sections from JEF-2.2, for example, explain an effect of ∼400 pcm. Refinement of the materials specifications in the MCNP4B whole-reactor model, in particular the impurities assumed for the graphite driver of the driver and reflector regions, has been shown to reduce the final discrepancy of the ENDF/B-V based k_eff result to ∼0.2%. The MCNP4B results for relative reactivity effects, such as control rod worths, are found to agree within experimental errors with the measured values.     

Calculations of Maxwellian-averaged cross sections and astrophysical reaction rates using the ENDF/B-VII.0, JEFF-3.1, JENDL-3.3, and ENDF/B-VI.8 evaluated nuclear reaction data libraries

We have calculated the Maxwellian-averaged cross sections and astrophysical reaction rates of the stellar nucleosynthesis reactions (n, γ), (n, fission), (n, p), (n, α), and (n, 2n) using the ENDF/B-VII.0, JEFF-3.1, JENDL-3.3, and ENDF/B-VI.8 evaluated nuclear reaction data libraries. These four major nuclear reaction libraries were processed under the same conditions for Maxwellian temperatures (kT) ranging from 1 keV to 1 MeV. We compare our current calculations of the s-process nucleosynthesis nuclei with previous data sets and discuss the differences between them and the implications for nuclear astrophysics.     

Measurement and Analysis of241 Am Fission Rate Ratio Relative to235 U Fission Rate in Thermal Neutron Systems Using Kyoto University Critical Assembly

Integral measurements of²⁴¹ Am fission rate ratio relative to²³⁵ U fission rate were performed at Kyoto University Critical Assembly. The fission rates were measured using the back-to-back type double fission chamber at five thermal cores with different H/²³⁵ U ratio so that the neutron spectra of the cores were systematically varied. The measured fission rate ratios, normalized to number of atoms, were 0. 0144 to 0. 0522, with a typical uncertainty of 2%. The measured data were compared with the calculated results using MVP based on JENDL-3.2, which gave the averaged calculated-to-experimental ratio (C/E) of 0.88. Obtained results of C/E using ²⁴¹Am fission cross sections from JENDL-3/2, ENDF/B-VI and JEF2.2 showed that the latter two gave larger C/E values than those by JENDL-3.2 by about 2% and 7 to 9%, respectively. It has been found that the large difference between JENDL-3.2 and JEF2.2 arises mainly from the significant cross section difference at the vicinity of resonance at 0.576 eV, whereas the difference of thermal cross sections, especially in the range of 0.01 eV to 0.2 eV also has significant contribution for well-thermalized cores.     

A proposal of a benchmark for βeff, βeff/Λ, and Λ of thermal reactors fueled with slightly enriched uranium

A reactor noise approach has been successfully performed at the IPEN/MB-01 research reactor facility for the experimental determination of the delayed neutron parameters β_eff, β_eff/Λ, and Λ. In the measurement of the β_eff parameter, the reactor power, which is of fundamental importance, was obtained with a very high level of accuracy by a fuel rod scanning technique and a subsequent irradiation of a highly enriched ²³⁵U foil for the fission density normalization. The final measured values of β_eff and β_eff/Λ show very good agreement with independent measurements and can be recommended as benchmark values for thermal reactor applications because their uncertainties are much lower than the target accuracy recommended for β_eff calculations (|C-E|/E less than 3%). The theory/experiment comparisons reveal that only JENDL3.3 attends the target accuracy for β_eff calculations. This result fully supports the reduction of the ²³⁵U thermal yield as proposed by Okajima and Sakurai. The ENDF/B-VI.8 library and its revised version performed at LANL overpredict β_eff by as much as 7.2%. The newly released JEFF-3.1 library shows a discrepancy of 4.8% for β_eff. For β_eff/Λ, the deviations are relatively larger (more than 10%) for all libraries due to the underprediction of the prompt neutron generation time (Λ).     

Measurement of keV-Neutron Capture Cross Section of 96Zr

The neutron capture cross section of ⁹⁶Zr at incident neutron energies from 15 to 100 keV has been measured by the time-of-flight method. Capture γ-rays were detected with an anti-Compton NaI(Tl) spectrometer, and the pulse-height weighting technique was applied to derive the neutron capture cross section. The present measurement provided the capture cross section as a function of incident neutron energy in the keV region. The results were compared with previous measurements and cross section data in the evaluated nuclear data libraries, JENDL-4.0, JENDL-3.3, ENDF/B-VII.0, and ENDF/B-VI.8. The present results revealed considerable underestimation of the evaluated cross sections in the high-energy region of 35–100 keV.