Feedback on neutron capture cross sections of 238Pu and 241Am from analyses of measured isotopic compositions of irradiated LWR fuels and MOX core physics experiments

The atomic fractions of ²³⁸Pu and ²⁴¹Am in MOX fuels recycled in light water reactors are 1% to 2% and not significant compared with those of major Pu isotopes. On the other hand, recent evaluated nuclear data libraries, such as JENDL-4.0 and JEFF-3.2, give noticeably different thermal and epithermal neutron capture cross sections for ²³⁸Pu and ²⁴¹Am. The thermal neutron capture cross sections of ²³⁸Pu and ²⁴¹Am in JEFF-3.2 are 31% and 9% larger than those of JENDL-4.0, respectively. This paper shows the effect of the differences in the neutron cross sections on analysis results of two different integral experiments. The first is the isotopic compositions of ²³⁸Pu on UO₂ and MOX fuels irradiated in BWR and PWR, and the second is the critical experiments of the water moderated cores fully loaded with MOX fuels. The former was analyzed by using the continuous energy Monte Carlo burnup calculation code MVP-BURN and the latter by the continuous energy Monte Carlo calculation code MVP. The comparisons between the calculated and measured results indicate that the most likely thermal and epithermal neutron capture cross sections of ²³⁸Pu and ²⁴¹Am should be around at the middle between those of JEFF-3.2 and JENDL-4.0.     

Validation of the sodium void reactivity effect prediction using JEFF-3.1 nuclear data

Near 60 Na void experiments performed in the zero power reactors MASURCA (CEA-Cadarache) and ZPPR (Argonne West – Idaho) have been analyzed using JEFF-3.1 nuclear data and the ERANOS-2.1 (deterministic) and TRIPOLI-4 (Monte-Carlo) codes. Some comparative calculations have been performed also using either JEFF-3.1, ENDF/B-VII.0 or JENDL-3.3 nuclear data for ²³Na, as these three ²³Na evaluations show marked differences. The Na void experiments have been selected to cover spectral conditions ranging from the relatively hard flux in the outer zone of a small fast reactor to the relatively soft flux in the inner zone of a large fast reactor. For in-fuel Na void patterns, there is a good agreement between ERANOS and TRIPOLI computations, while the deterministic calculations significantly underestimate the leakage component for Na void patterns in fertile regions. The agreement between ERANOS-2.1 + JEFF-3.1 predictions and experimental values is excellent for in-fuel Na void patterns in MASURCA experiments, but a significant underestimation of the leakage component occurs for in-fuel Na void patterns in ZPPR. For fertile Na void patterns, there is a clear underestimation of the leakage component, quantitatively different for MASURCA and ZPPR experiments. Variations in ²³Na cross-section data also result in significant differences: ENDF/B-VII.0 and JENDL-3.3 nuclear data for ²³Na increase noticeably the predicted Na void worth values with respect to JEFF-3.1 data. The three ²³Na evaluations differ at high energy (>500 keV, and especially >2 MeV), and this stresses the need for accurate additional measurements in this energy range.     

Performance of a Monte-Carlo solution for the H.B. Robinson-2 Pressure Vessel Dosimetry Benchmark

At the Paul Scherrer Institut, a methodology for PWR fast neutron fluence estimations, based on the Monte-Carlo particle transport code MCNPX with general-purpose neutron data libraries and using neutron source data from deterministic 3-D core-follow calculations, has been developed. The methodology has been validated on the basis of experimental data related to the fluence at the inner surface of a Swiss PWR Reactor Pressure Vessel. In this technical note, a first objective is to enlarge the validation basis of the methodology as well as to extend it for applications to RPV outer-surfaces. To that aim, a preliminary analysis with the MCNPX-2.4.0 code, along with the JEFF-3.1 continuous-energy neutron data library, of the “H.B. Robinson-2 Pressure Vessel Benchmark”, providing in-vessel and out-vessel experimental dosimetry data, is presented. In addition, considering that the available original solutions of the benchmark employed deterministic transport methods with associated libraries, a second objective of this technical note is to assess the progress achieved for this type of problems when applying modern Monte-Carlo based methods. The results show that for the ²³⁷Np and ²³⁸U fission dosimeters, which were of primary interest in the given study, a non-negligible improvement is seen in the MCNPX solution, indicating thus a rather good performance of the employed Monte-Carlo method and providing thereby, additional confidence for the overall PSI fluence methodology. For other high-energy dosimeters, the presented new results do not show yet any significant accuracy improvement versus previously reported results. This can however not be confirmed before additional studies, e.g., with focus on improvements of the involved modelling approximations and the statistical precision of the MCNPX calculations, be carried out. Similarly, investigations of neutron cross-section and dosimetry libraries effects remain to be addressed. These further studies are however not included here since at this stage, the principal aim was mainly to model and analyse this benchmark at most consistent manner with the previous solutions using a continuous-energy Monte-Carlo based method.     

Safety performance comparation of MOX, nitride and metallic fuel based 25–100 MWe Pb–Bi cooled long life fast reactors without on-site refuelling

In this paper the safety performance of 25–100 MWe Pb–Bi cooled long life fast reactors based on three types of fuels: MOX, nitride and metal is compared and discussed. In the fourth generation NPP paradigm, especially for Pb–Bi cooled fast reactors, inherent safety capability is necessary against some standard accidents such as unprotected loss of flow (ULOF), unprotected rod run-out transient over power (UTOP), unprotected loss of heat sink (ULOHS). Selection of fuel type will have important impact on the overall system safety performance.

The results of safety analysis of long life Pb–Bi cooled fast reactors without on-site fuelling using nitride, MOX and metal fuel have been performed. The reactors show the inherent safety pattern with enough safety margins during ULOF and UTOP accidents. For MOX fuelled reactors, ULOF accident is more severe than UTOP accident while for nitride fuelled cores UTOP accident may push power much higher than that comparable MOX fuelled cores.     

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 Measured Isotopic Compositions of High-Burnup BWR MOX Fuel

Analysis of measured isotopic compositions of four high-burnup BWR MOX fuel samples was performed by using a general-purpose neutronic calculation code SRAC and a continuous-energy Monte Carlo burnup code MVP-BURN. The initial Pu fissile content of the samples was 5.52 wt%, and the burnups ranged from 50 to 80 GWd/t. It is confirmed that a geometrical model including the effect of UO₂ assemblies adjacent to the MOX assembly is necessary in the burnup calculations to obtain accurate calculated isotopic compositions. The calculated results of MVP-BURN with JENDL-3.3 taking such effect into account show more accurate results for major actinides (U, Pu, and Am isotopes) and most fission products than those of infinite assembly calculations. The paper also shows the results calculated using SRAC with JENDL-3.3, ENDF/B-VII, and JEFF-3.1.     

Problems of lead nuclear data in fusion blanket design

In an irradiation experiment using a LiAl/Pb assembly, we found out that the neutron flux inside the assembly calculated with JENDL-3.3 underestimates an experimental value in the 10–16 MeV region by around 30% and that in the 0.5–5 MeV region by around 15%, while the calculated flux with JEFF-3.1 overestimates the measurement in the 5–10 MeV region by around 20%. In order to reveal a reason of the discrepancy, problems of the nuclear data libraries for lead were investigated. As a result, the following problems of the evaluated libraries were pointed out: the cross-sections of the (n,2n) reaction in JENDL-3.3 for lead isotopes are too large and cause a significant underestimation of the neutron flux above 10 MeV, which appeared in the analysis of the above experiment. Inelastic scattering data for ²⁰⁸Pb in JENDL-3.3 reproduce previous experimental double-differential cross-section data most well. However, those for the other lead isotopes have some problems and cause a large underestimation of the neutron flux from 0.5 to 5 MeV. The reason of the overestimation in the energy region of 5–10 MeV with JEFF-3.1 is still unclear.     

Accuracy of Cell Calculation Methods Used for Analysis of High Conversion Light Water Reactor Lattice

Validation tests were made for the accuracy of cell calculation methods used in analyses of tight lattices of a mixed-oxide (MOX) fuel core in a high conversion light water reactor (HCLWR). A series of cell calculations was carried out for the lattices referred from an international HCLWR benchmark comparison, with emphasis placed on the resonance calculation methods; the NR, IR approximations, the collision probability method with ultra-fine energy group. Verification was also performed for the geometrical modelling; a hexagonal/cylindrical cell, and the boundary condition; mirror/white reflection. In the calculations, important reactor physics parameters, such as the neutron multiplication factor, the conversion ratio and the void coefficient, were evaluated using the above methods for various HCLWR lattices with different moderator to fuel volume ratios, fuel materials and fissile plutonium enrichments.

The calculated results were compared with each other, and the accuracy and applicability of each method were clarified by comparison with continuous energy Monte Carlo calculations. It was verified that the accuracy of the IR approximation became worse when the neutron spectrum became harder. It was also concluded that the cylindrical cell model with the white boundary condition was not so suitable for MOX fuelled lattices, as for UO₂ fuelled lattices.     

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.     

JEFF-3.1.1 Nuclear Data Validation for Sodium Fast Reactors

The JEFF-3.1.1 Nuclear Data Library is the latest version of the Joint Evaluated Fission and Fusion Library. We present the status of the validation of this library using the Monte Carlo Code TRIPOLI 4.5 and the deterministic code package ERANOS 2.2 for fast reactor calculations. For this purpose, we reanalyze a selected set of integral experiments performed in the MASURCA mock-up (CEA/CADARACHE), in the ZPPR mock-up at ANL (USA), and in the SUPERPHENIX Power Reactor. Furthermore, we also present the analysis of pure sample irradiation experiments PROFIL and PROFIL-2 performed in the PHENIX reactor, as this kind of experiment provides a direct feedback on nuclear capture data. We observe good performances of these calculation tools for criticality calculations and fuel inventory prediction. From this validation work, some required improvements on nuclear data are highlighted, as well as the need for new specific integral experiments. The main trends observed are the following:

—Reactivity of clean and fresh cores: the results obtained with JEFF-3.1.1 are consistent with those obtained using JEFF-3.1 within 80 pcm, but there is an overestimation of the calculated reactivity of all the experiments between 40 and 800 pcm depending on the spectrum hardness (Pu content) and fuel composition, the discrepancy being larger in hard spectra. Additional investigations are in progress to understand this behaviour.

—Integral capture cross sections (PROFIL and PROFIL-2):/C/E - 1/≤ 3%, except for ²⁴¹Pu (C/E ≈ 1:08), ²⁴²Pu (C/E ≈ 1:18), ²³⁷Np (C/E ≈ 0:92), ²⁴³Am (C/E ≈ 0:93), ²⁴⁴Cm (C/E ≈ 1:35); impact of the trends observed on the individual fission products put in PROFIL and PROFIL-2 is ≈ 4% on the part of Δ_ρburnup due to fission products.     

Analyses with latest major nuclear data libraries of the fission rate ratios for several TRU nuclides in the FCA-IX experiments

In the 1980s, a series of integral experiments was conducted in FCA-IX assemblies with systematically changed neutron spectra covering from the intermediate to fast ones. The experiments provide systematic data of central fission rates for TRU nuclides containing minor actinides, ²³⁷Np, ²³⁸Pu, ²³⁹Pu, ²⁴²Pu, ²⁴¹Am, ²⁴³Am, and ²⁴⁴Cm. Regarding the fission rate ratios relative to ²³⁹Pu, benchmark models had been recently developed for validation of nuclear data for the TRU's fission cross sections. In this paper, the latest major nuclear data libraries, JENDL-4.0, ENDF/B-VII.1, and JEFF-3.2, are compared on the benchmark models. For the libraries, the analyses by a Monte Carlo calculation code show obvious overestimations particularly for the fission rate ratios of ²⁴⁴Cm to ²³⁹Pu. Additionally, a large discrepancy about by 20% between the libraries is revealed for the fission rate ratio of ²³⁸Pu to ²³⁹Pu measured in the intermediate neutron spectrum. The causes of discrepancies are furthermore clarified by sensitivity analyses.     

Comparison of calculated and measured reaction rates obtained through foil activation in the subcritical dual spectrum facility YALINA-Booster

Reaction rates were measured by the foil activation technique to obtain neutron spectrum information in a subcritical core driven by an external neutron source. The experimental results are compared with Monte Carlo calculations in order to examine the capability of the Monte Carlo code MCNP together with ENDFB-6.8, JEFF-3.1.1 and CENDL-3.1 neutron cross section libraries to predict the neutron spectrum dependent reaction rates correctly in a subcritical core. The focus lies on fast neutrons. A discrepancy is found in the calculated-to-experimental values of the reaction rates and an inaccurate cross section is identified in CENDL-3.1.     

A feasibility study of ferro-boron as in-core shield material in fast breeder reactors

Shields around core and blankets form major part of reactor assembly in fast reactors as the incident neutron spectrum is hard with negligible thermal component and has anisotropic angular distribution. In this paper, a study is presented on the use of ferro-boron as neutron shield material in pool type fast reactors. The reference case chosen is the Prototype Fast Breeder Reactor (PFBR), a 500 MWe which is sodium cooled, pool type, mixed oxide (MOX) fuelled reactor, which is under construction at Kalpakkam, India. It is shown through 2D transport calculations, carried out using 175 neutron multigroup cross-sections, that this low cost material as shield is capable of satisfying the radiological safety criteria as well as the shields in the reference case. The secondary sodium activity and dose in steam generator building are marginally lower than the reference case. The total shield material weight will be lower by about 50 tonnes and the material cost lower by a factor 5 as compared to PFBR shields comprising of stainless steel and B₄C.     

Neutronic performance comparation of MOX, nitride and metallic fuel based 25–100 MWe Pb–Bi cooled long life fast reactors without on-site refuelling

In this paper the performance of 25–100 MWe Pb–Bi cooled long life fast reactors based on three type of fuels: MOX, Nitride and Metal are compared and discussed. In general MOX fuel (UO₂–PuO₂) has lower atomic density compared to the nitride or metal fuel, but MOX fuel has some advantages such as higher Doppler coefficient, high melting point and availability. Nitride fuel has advantages such as high density, high thermal conductivity, and high melting point, but need N-15 to avoid C-14 problems.

The results show that nitride fuel as well as MOX fuel can be used to develop 25–100 MWe (75–300 MWth) Pb–Bi cooled long life reactors without on-site fuelling. The results show that nitride fuels have more superior neutronic characteristics compared to that of MOX fuel due to higher density. However, in the large power level both fuels can be easily applied. In lower power level the MOX fuel need higher fuel volume fraction to reach the comparable target of nitride fuelled cores.     

Evaluation of neutron cross sections for a complete set of Dy isotopes

Neutron cross sections for a complete set of Dy isotopes, ^{156,158,160,161,162,163,164}Dy, were evaluated in the incident energy range from 10⁻⁵ eV to 20 MeV. In the low energy region, including thermal and resolved resonances, our evaluations are based on the latest data published in the Atlas of Neutron Resonances. In the unresolved resonance region we performed additional evaluation by using the averages of the resolved resonances and adjusting them to the experimental data. In the fast neutron region, we used the nuclear reaction model code EMPIRE-2.19 with the model parameters adjusted to the experimental data. The results are compared with the available experimental data and with the existing nuclear data libraries, including ENDF/B-VI.8 and JEFF-3.1. The new evaluations are suitable for neutron transport calculations and they were adopted by the new US evaluated nuclear data library, ENDF/B-VII.0, released in December 2006.     

Analysis of core physics experiments on irradiated BWR MOX fuel in the REBUS program

Critical experiments were performed in the REBUS program on a core loaded with a test bundle including 16 irradiated BWR-type MOX rods of average burnup of 61 GWd/t. The experimental data were analyzed using diffusion, transport, and continuous-energy Monte Carlo calculation codes coupled with nuclear data libraries based on JENDL-3.2 or JENDL-3.3. Biases in effective multiplication factors of the critical cores were ?1.0%Δk for the diffusion calculations (JENDL-3.2), ?0.3%Δk for the transport calculations (JENDL-3.3), and 0.2%Δk for the Monte Carlo calculations (JENDL-3.2). The measured core fission rate and co-activation rate distributions were generally well reproduced using the three types of calculations. The burnup reactivity determined using the measured water level reactivity coefficients was ?2.41 ± 0.08%Δk/kk’, which also agreed with the results of the three type of calculations within the measurement and calculation errors. The most probable isotopic inventories in the irradiated MOX rods was tentatively obtained by using the ratios of the calculation to chemical assay data on a pellet sample, and the burnup reactivity was reanalyzed to split the calculation error into those due to the inventory and reactivity calculations. This approach showed that the inventory calculation error compensated the reactivity calculation error.     

板状Bi样品泄漏中子和γ谱实验测量与模拟计算

通过飞行时间法,测量了氘氘脉冲中子与不同厚度209Bi样品作用后61°和119°方向的泄漏中子飞行时间谱和泄漏γ能谱,样品尺寸分别为30 cm×30 cm×5 cm、30 cm×30 cm×10 cm和30 cm×30 cm×15 cm。采用BC501A液体闪烁体探测器测量0.8～3.2 MeV能区的泄漏中子飞行时间谱,钾冰晶石探测器（CLYC）测量0.2～0.8 MeV的泄漏中子飞行时间谱和泄漏γ能谱。用MCNP-4C程序对泄漏中子飞行时间谱和泄漏γ能谱进行了模拟计算,²⁰⁹Bi的评价中子核数据分别采用了CENDL-3.1库、ENDF/B-Ⅷ.0库、JENDL-4.0库以及JEFF-3.3库中的数据,模拟结果分别与实验结果进行比较分析,研究结果表明,泄漏中子谱CENDL-3.1库的模拟结果在119°方向弹性峰位置有较严重的低估现象,JENDL-4.0库在1.5 MeV附近（第二非弹能区）有一定高估,而在低能区有明显低估;泄漏γ能谱JENDL-4.0库和JEFF-3.3库的模拟结果与实验结果偏差明显,而CENDL-3.1库符合较好。     

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 ...