Fast Critical Assembly Research Articles

Experiments on Criticality and Reactivity Worths in the FCA-XXII-1 Assembly Simulating Highly Enriched MOX Fueled Tight Lattice LWR Cores

A series of integral experiments were conducted at the fast critical assembly (FCA) of the Japan Atomic Energy Agency, simulating light water reactor cores with a tight lattice cell of highly enriched mixed-oxide (MOX) fuel containing >15% fissile plutonium (Pu). The three experimental configurations of the FCA-XXII-1 assembly were constructed using foamed polystyrene with different void fractions (45%, 65%, and 95%) to clarify the prediction accuracy of neutronics calculation codes and nuclear data libraries among various neutron spectra. The hydrogen-to–nuclear fuel atomic ratio varied from 0.1 to 0.8. The nuclear characteristics measured in the experiments were criticality (keff), moderator void reactivity worths, and sample reactivity worths using boron carbide (20%, 60%, and 90% 10B enrichment) and Pu (92%, 81%, and 75% fissile Pu ratio). Preliminary analyses on experiments were conducted using a deterministic calculation code system conventionally used for fast reactors and the Japanese evaluated nuclear data library of JENDL-4.0. The calculated keff values overestimated the experiments beyond the experimental uncertainties. However, most reactivity worth calculations agreed well with the experimental values. Even beyond the experimental uncertainties, discrepancies between the calculation and the experiment were <13%. Specifically in the reactivity worth analyses of the softer neutron spectra configurations, the treatment of ultrafine energy groups obviously improved the prediction accuracy of the deterministic calculations. Furthermore, reference calculations for criticality and large reactivity worths were performed with the Monte Carlo calculation code MVP3 by modeling the experimental configurations in detail, confirming that the deterministic calculations closely agreed with the reference values.

Shedding light on the 239Pu fission source term with new high-precision experiments and advanced fission modeling

In the last decade, there has been a renaissance of fission research resulting in new high-precision experiments and advanced fission modeling. For instance, the Chi-Nu and CEA teams supplied, for the first time, the 239Pu prompt fission neutron spectrum (PFNS) for broad ranges of incident and outgoing neutron energies. The CEA team also measured 239Pu average prompt neutron multiplicities, ν̄p, with lower statistical uncertainties and a technique significantly different than the one used in the past. The NIFFTE collaboration provided 239Pu(n,f)/235U(n,f) cross section shape ratios with uncertainties below 1% utilizing a novel detector type. Advanced fission event generators were developed, among them CGMF, FIFRELIN, FREYA, and GEF, which calculate post-scission fission observables in a correlated manner. These new experimental data and more consistent fission models change the evaluated PFNS, ν̄p, and (n,f) cross sections, some only modestly, compared to ENDF/B-VIII.0. In turn, the individual new nuclear data distinctly change simulated effective neutron multiplication factors of fast critical assemblies, but their combined impact is small, while affecting the prediction of LLNL pulsed sphere neutron leakage spectra and reaction rates only within experimental uncertainties. Also, the parameters obtained from fitting to ν̄p reproduce various post-scission fission observables within the uncertainties of experimental data. This indicates that new differential experiments and consistent fission modeling reduce compensating errors present in ENDF/B-VIII.0.

Low Enrichment Nuclear Fuel Based on Uranium-Zirconium Carbonitride: Reactor Tests and Preparation for Studies at Critical Assemblies

Uranium-zirconium carbonitride has been developed at the Federal State Unitarian Enterprise Scientific Research Institute the Scientific Industrial Association "Luch" (LUCH FSUE) and is a high-density high-temperature fuel with high heat conductivity capable of being used in various types of reactors, including fast reactors. The main problem hindering wide application of this fuel is insufficient knowledge of its behavior under irradiation, especially at high burnup. In the USSR, HEU UZrCN (96% by U-235) fuel underwent reactor testing to a low burnup of approximately 1%. However, to confirm practical applicability of this fuel, the fuel needs to be reactor tested at a high burnup. In the framework of a joint Belarussian-American-Russian effort, LEU (19.75% by U-235) UZrCN fuel will undergo reactor testing to a burnup of approximately 40% in the SM-3 reactor at the Joint Stock Company "State Scientific Center Research Institute of Atomic Reactors" (JSC "SSC RIAR"). To conduct the prolonged reactor experiment, an irradiating device with an experimental capsule containing UZrCN pellets has been made, neutronic analysis and thermophysical analysis have been carried out and a programme of pre-irradiation experiments has been implemented. In June 2019 a methodical experiment was carried out in the SM-3 reactor at the JSC "SSC RIAR" in order to confirm operability of the irradiating device developed. Testing of the irradiating device and refinement of the prolonged irradiation experiment procedure were carried out at cell 11 of the reflector of the SM-3 reactor at the JSC "SSC RIAR". The period of irradiation equaled 23.3 effective days. The mean power density in the tested pellets throughout the methodical reactor testing was 516 W=cm3. Burnup achieved for the pellets studied was 0.63% FIFA. The "Giacint" and "Kristal", critical facilities of the State Scientific Institution "The Joint Institute for Power and Nuclear Research - Sosny" (the Scientific Institution "JIPNR - Sosny") will be used for studying neutronic characteristics of critical and subcritical fast assemblies simulating the physical particulars of the cores of advanced gas- or liquid metal cooled fast reactor systems and accelerator driven systems. This paper provides a detailed description of the results of preparatory works for conducting the reactor testing and the results of the methodical reactor experiment. Also the experimental programme and the descriptions of the design and composition of the fast critical assemblies with UZrCN fuel are presented.

Time sequence method for prompt neutron decay constant determination

Aiming at measuring the prompt neutron decay constant α of fast critical assemblies, Rossi-α method was improved by means of the acquisition of neutron and gamma time sequence. The application of the time sequence method in delayed critical & subcritical experiments at a fast critical assembly was here in described, focus in particular on the gamma analysis and the exploration of A/C ratio which directly affects the measurement precision of α. The reliability of the time sequence method is proved by the agreement of the α values obtained based on neutron and gamma. The A/C ratio based on neutron and gamma shows an obvious difference and the corresponding conclusion guides to a larger A/C ratio. The time sequence method shows advantages of higher experimental efficiency and flexible data analysis which contribute to more accurate experimental result.

Population-based variance reduction for dynamic Monte Carlo

Dynamic Monte Carlo (DMC) simulation of realistic nuclear reactors requires powerful variance reduction methods for even a few seconds of real time calculations. State-of-the-art numerical methods deal with the dynamic nature of the problem via successive Monte Carlo transport and TH (thermal-hydraulic) runs in a time step by time step manner. Such halting of the sample population at the beginning of time steps also allows for a joint handling of samples in a variance reduction effort. A theoretical framework is given for the connection of weight distribution and tally variance by factorizing it into a population variance accumulated by previous time steps and the variance caused by the transport process in the last interval. A long term importance function is proposed for decreasing the main contribution of a power release tally variance. Novel techniques are shown and compared when using Russian Roulette and Splitting. A simple fast critical assembly and a detailed thermal reactor geometry are used for testing showing that a factor of at least two orders of magnitude is to be gained by a simple population comb targeting the average weight. Further improvement using importance and variance functions is less than a factor two.

Nuclear data uncertainty propagation for spectral reaction ratios

This paper describes several approaches to propagating nuclear data uncertainty in integral capture and fission reaction rates, focusing on the spectral hardness indices of 193Ir(n,n′)193mIr/191Ir(n,g)192Ir and 238U(n,f)/235U(n,f) in fast critical assemblies. Uncertainties in the nuclear data were propagated through the following methods: a sensitivity method (utilizing first order perturbation and covariance matrices), a BFMC method (random cross sections generated through covariance matrices), and Total Monte Carlo method (random cross sections generated from varying nuclear parameters). The uncertainty estimates from the three approaches are evaluated against the associated experimental uncertainties of a ZEUS/COMET critical assembly irradiation. The uncertainty estimates of the three approaches agree fairly well, and contribute a significant portion to the overall uncertainty of the spectral hardness indices.

Activation product interpretation of structural material for fast critical assemblies

This work evaluates whether activation products in stainless steel components can be used to infer initial conditions (i.e. alloy composition, position within the fluence, etc.) and the associated uncertainties in a fast neutron irradiation. Activation measurements of stainless steel and elemental components were obtained through a Flattop critical assembly irradiation experiment. A forward model was developed, utilizing MCNP6.1 and FISPACT ii, to estimate activation product concentrations from several input parameters: alloy composition, source specifications, and foil positioning. While some of the calculated activation product concentrations agreed fairly well with measurements, some isotopics (i.e. 57Co) had large discrepancies. It became clear that uncertainties in both measurements and nuclear data drove this divergence for select isotopics, and the quantification of this impact was needed. By treating both experimental measurement and nuclear data uncertainty in a Bayesian framework, it was possible to infer unknowns from given activation ratios. The developed approach showed promise in constraining mass parameters, but degeneracy existed when considering position within the neutron flux. Higher threshold reactions could potentially break the degeneracy, such as 57Co, but uncertainty estimation needs to be better resolved for those isotopes.

A new cross section adjustment method of removing systematic errors in fast reactors

A new cross section adjustment method has been derived in which systematic errors in measured data and calculated results of neutronics characteristics are estimated and removed in the adjustment. Bias factors which are the ratio between measured data and calculated results are used to estimate systematic errors. The difference of the bias factors from unity is caused generally by systematic errors and stochastic errors. Therefore by determining whether the difference is within the total stochastic errors of measurements and calculations, systematic errors are estimated. Since stochastic errors are determined for individual confidence levels, systematic errors are also dependent to the confidence levels. The method has been applied to cross section adjustments using 589 measured data obtained from fast critical assemblies and fast reactors. The adjustments results are compared with those of the conventional adjustment method. Also the effect of the confidence level to the adjusted cross sections is discussed.

Fission Rate Ratios of FCA-IX Assemblies as Integral Experiment for Assessment of TRU’s Fission Cross Sections

At the fast critical assembly (FCA) of JAEA, central fission rate ratios for TRU such as 237 Np, 238 Pu, 239 Pu, 242 Pu, 241 Am, 243 Am, and 244 Cm were measured in the seven uraniumfueled assemblies (FCA-IX assemblies) with systematically changed neutron spectra. The FCA-IX assemblies were constructed with simplicity both in geometry and composition. By virtue of these FCA-IX assemblies where the simple combinations of uranium fuel and diluent (graphite and stainless steel) in their core regions were systematically varied, the neutron spectra of them cover from the intermediate to fast one. Taking their advantages, benchmark models with respect to the central fission rate ratios had been recently developed for the evaluation of the TRU’s fission cross sections. As an application of these benchmark models, the Japanese Evaluated Nuclear Data Library JENDL-4.0 was utilized by a Monte Carlo calculation code. Several results show large discrepancies between the calculation and experimental values. The benchmark models would be well suited for the evaluation and modification of the nuclear data for the TRU’s fission cross sections.

A Re-Analysis of Historical Los Alamos Critical Assembly Reaction Rate Measurements

Starting in the 1950s and continuing into the early 1970s, a number of foil irradiations and fission chamber measurements were made in a variety of Fast critical assemblies at Los Alamos National Laboratory. These include (i) Godiva, a bare HEU spherical assembly; (ii) Flattop-25, a spherical assembly consisting of an HEU core and a natural uranium reflector; (iii) Jezebel, a bare 239 Pu assembly; and (iv) Flattop-Pu, a spherical assembly consisting of a 239 Pu core and a natural uranium reflector. In most instances the irradiations occur at or near the center of the assembly, but in selected instances data were obtained for a radial traverse extending into the Flattop reflector region. Measurements were made for a number of threshold reactions, including 45 Sc(n,2n)44m Sc, 51 V(n ,α)48 Sc, 75 As(n,2n)74 As, 89 Y(n,2n)88 Y, 90 Zr(n,2n)89 Zr, 103 Rh(n,2n)102g Rh, 107 Ag(n,2n)106m Ag, 169 Tm(n,2n)168 Tm, 175 Lu(n,2n)174 Lu, 191 Ir(n,2n)190 Ir, 197 Au(n,2n)196 Au, 203 Tl(n,2n)202 Tl, 204 Pb(n,2n)203 Pb and 238 U(n,2n)237 U. Fission ratio data for 238 U(n,f)/235 U(n,f) and 239 Pu(n,f)/235 U(n,f) were also obtained. We report C/E values from MCNP6 calculations using ENDF/B-VII.1 and IRDFF-v1.03 cross section data.

Benchmark models for criticalities of FCA-IX assemblies with systematically changed neutron spectra

New benchmark models with respect to criticality data are established on the basis of seven uranium-fueled assemblies constructed in the ninth experimental series at the fast critical assembly (FCA) facility. By virtue of these FCA-IX assemblies, where the simple combinations of uranium fuel and diluent (graphite and stainless steel) in their core regions were systematically varied, the neutron spectra of these benchmark models cover those of various reactor types, from fast to sub-moderated reactors. The sample calculations of the benchmark models by a continuous-energy Monte Carlo (MC) code showed obvious differences between even the latest versions of two major nuclear data libraries, JENDL-4.0 and ENDF/B-VII.1. The present benchmark models would be well suited for the assessment and improvement of the nuclear data for 235U, 238U, graphite, and stainless steel. In addition, the verification of the deterministic method was performed on the benchmark models by comparison with the MC calculations. The present benchmark models are also available to users of deterministic calculation codes for the assessment and improvement of nuclear data.

Thoughts on Sensitivity Analysis and Uncertainty Propagation Methods with Respect to the Prompt Fission Neutron Spectrum Impact on Critical Assemblies

The prompt fission neutron spectrum (PFNS) uncertainties in the n+239Pu fission reaction are used to study the impact on several fast critical assemblies modeled in the MCNP6.1 code. The newly developed sensitivity capability in MCNP6.1 is used to compute the keff sensitivity coefficients with respect to the PFNS. In comparison, the covariance matrix given in the ENDF/B-VII.1 library is decomposed and randomly sampled realizations of the PFNS are propagated through the criticality calculation, preserving the PFNS covariance matrix. The information gathered from both approaches, including the overall keff uncertainty, is statistically analyzed. Overall, the forward and backward approaches agree as expected. The results from a new method appear to be limited by the process used to evaluate the PFNS and is not necessarily a flaw of the method itself. Final thoughts and directions for future work are suggested.

Experiments on small-size fast critical fuel assemblies at the AKSAMIT facility and their use for development of computational models

Small-size fast critical assemblies with highly enriched fuel at the AKSAMIT facility are described in detail. Computational models of the critical assemblies at room temperature are given. The calculation results for the critical parameters are compared with the experimental data. A good agreement between the calculations and the experimental data is shown. The physical models developed for the critical assemblies, as well as the experimental results, can be applied to verify various codes intended for calculation of the neutronic characteristics of small-size fast nuclear reactors. For these experiments, the results computed using the codes of the MCU family show a high quality of the neutron data and of the physical models used.

Benchmarking ENDF/B-VII.1, JENDL-4.0 and JEFF-3.1.1 with MCNP6

Recent releases of three major world nuclear reaction data libraries, ENDF/B-VII.1, JENDL-4.0, and JEFF-3.1.1, have been tested extensively using benchmark calculations. The calculations were performed with the latest release of the continuous energy Monte Carlo neutronics code MCNP, i.e. MCNP6. Three types of benchmarks were used, viz. criticality safety benchmarks, (fusion) shielding benchmarks, and reference systems for which the effective delayed neutron fraction is reported.For criticality safety, more than 2000 benchmarks from the International Handbook of Criticality Safety Benchmark Experiments were used. Benchmarks from all categories were used, ranging from low-enriched uranium, compound fuel, thermal spectrum ones (LEU-COMP-THERM), to mixed uranium-plutonium, metallic fuel, fast spectrum ones (MIX-MET-FAST). For fusion shielding many benchmarks were based on IAEA specifications for the Oktavian experiments (for Al, Co, Cr, Cu, LiF, Mn, Mo, Si, Ti, W, Zr), Fusion Neutronics Source in Japan (for Be, C, N, O, Fe, Pb), and Pulsed Sphere experiments at Lawrence Livermore National Laboratory (for 6Li, 7Li, Be, C, N, O, Mg, Al, Ti, Fe, Pb, D2O, H2O, concrete, polyethylene and teflon). The new functionality in MCNP6 to calculate the effective delayed neutron fraction was tested by comparison with more than thirty measurements in widely varying systems. Among these were measurements in the Tank Critical Assembly (TCA in Japan) and IPEN/MB-01 (Brazil), both with a thermal spectrum, two cores in Masurca (France) and three cores in the Fast Critical Assembly (FCA, Japan), all with fast spectra.The performance of the three libraries, in combination with MCNP6, is shown to be good. The results for the LEU-COMP-THERM category are on average very close to the benchmark value. Also for most other categories the results are satisfactory. Deviations from the benchmark values do occur in certain benchmark series, or in isolated cases within benchmark series. Such instances can often be related to nuclear data for specific non-fissile elements, such as C, Fe, or Gd. Indications are that the intermediate and mixed spectrum cases are less well described.The results for the shielding benchmarks are generally good, with very similar results for the three libraries in the majority of cases. Nevertheless there are, in certain cases, strong deviations between calculated and benchmark values, such as for Co and Mg. Also, the results show discrepancies at certain energies or angles for e.g. C, N, O, Mo, and W.The functionality of MCNP6 to calculate the effective delayed neutron fraction yields very good results for all three libraries.

Measurement and analysis of reflector reactivity worth by replacing stainless steel with zirconium at the fast critical assembly (FCA)

A reflector reactivity worth was measured by replacing stainless steel with zirconium at the FCA. The experimental result of the positive reflector reactivity worth demonstrates the effectiveness of the zirconium reflector compared with the SS reflector in the fast reactor core. This paper also focuses on the validation of standard calculation methods used for fast reactors with JENDL-4.0. As a result, it is confirmed that the standard calculation methods for the reflector reactivity worth show agreement within the experimental error.

Benchmark Calculations of Sodium-Void Experiments with Uranium Fuels at the Fast Critical Assembly FCA

The benchmark specification for the sodium-void reactivity experiments is given. The detailed benchmark calcu- lations for the new integral data were performed by using a continuous-energy Monte Carlo code (MVP) with use of the evaluated nuclear data libraries JENDL-3.2, -3.3, -4.0, ENDF/B-VII.0 and JEFF-3.1. The MVP calculations were performed with 2,000,000,000 particles. The statistical uncertainties of the MVP calculations were within several percents of the measured sodium-void reactivities. The ratios of calculated to experimental (C/E) values of sodi- um-void reactivities with respect to JENDL-3.3, ENDF/B-VII.0 and JEFF-3.1 are less than those with respect to JENDL-3.2 and -4.0. The analysis results are similar to those of sodium-void reactivities previously obtained at the BFS facility which is another fast critical assembly in Russia. The benchmark calculations demonstrate the improve- ment of the reliability of the integral data such as the new integral data obtained at the FCA and the previously obtained data in the BFS and the usefulness of the new integral data for the validation of the re-evaluated cross sec- tion of 235 U. In addition, the benchmark tests with various correction factors by a deterministic calculation system are summarized for the help to validate the re-evaluated capture cross section of 235 U.

Fission Product Yields from Fission Spectrum n+ 239Pu for ENDF/B-VII.1

We describe a new cumulated fission product yield (FPY) evaluation for fission spectrum neutrons on plutonium that updates the ENDF/B-VI evaluation by England and Rider, for the forthcoming ENDF/B-VII.1 database release. 1 1 We intend to release the ENDF/B-VII.1 database in December 2011, and all released data are subject to CSEWG approval. It is possible that the released evaluated data will differ from those presented in this paper; the evaluated date presented here can be referred to as ENDF/B-VII.1 beta 0. We focus on FPs that are needed for high accuracy burnup assessments; that is, for inferring the number of fissions in a neutron environment. Los Alamos conducted an experiment in the 1970s in the Bigten fast critical assembly to determine fission product yields as part of the Interlaboratory Reaction Rate (ILRR) collaboration, and this has defined the Laboratory's fission standard to this day. Our evaluation includes use of the LANL-ILRR measurements (not previously available to evaluators) as well as other Laboratory FPY measurements published in the literature, especially the high-accuracy mass spectrometry data from Maeck and others. Because the measurement database for some of the FPs is small — especially for 99Mo — we use a meta-analysis that incorporates insights from other accurately-measured benchmark FP data, using R-value ratio measurements. The meta-analysis supports the FP measurements from the LANL-ILRR experiment. Differences between our new evaluations and ENDF/B-VI are small for some FPs (less than 1–2%-relative for 95Zr, 140Ba, 144Ce), but are larger for 99Mo (4%-relative) and 147Nd (5%-relative, at 1.5 MeV) respectively. We present evidence for an incident neutron energy dependence to the 147Nd fission product yield that accounts for observed differences in the FPY at a few-hundred keV average energy in fast reactors versus measurements made at higher average neutron energies in Los Alamos' fast critical assemblies. Accounting for such FPY neutron energy dependencies is important if one wants to reach a goal of determining the number of fissions to accuracies of 1–2%. An evaluation of the energy-dependence of fission product yields is given for all A values based on systematical trends in the measured data, with a focus on the energy dependence over the fast neutron energy range from 0.2–2 MeV. Based on these trends, we present an evaluation of the FPY data at 0.5 and 2.0 MeV average incident neutron energies. This new set of ENDF/B-VII data will enable users to linearly interpolate between the pooled FPY data at ∼0.5 MeV and our new data at 2 MeV to obtain FPYs at other energies.

Accuracy of criticality calculations made using current nuclear data libraries for MOX fuelled fast critical assemblies

Criticality calculations have been made for a set of ten mixed plutonium–uranium oxide (MOX) fuelled fast critical assemblies using the current nuclear data libraries, JEFF-3.1, JEFF-3.1.1, JENDL-3.3 and ENDF/B-VII.0. The results obtained using the different libraries are compared and conclusions drawn concerning the accuracy of criticality calculations made for MOX fuelled fast reactors.

Development of Hydride Absorber for Fast Reactor—Evaluation of Design Method for Hydride Absorber Rod—

A hydride control rod is being developed to improve the economy of fast reactor plants because it has a longer lifetime than the currently used B4C control rod. A hydride burnable poison rod is also under development to reduce the number of control rods by decreasing core excess reactivity. Hydrogen in the hydride control rod causes neutron spectrum interference between the fuel and control rod regions. Thus, the study on core design was performed with the continuous-energy Monte Carlo code MVP using the nuclear data library JENDL-3.3 to deal with this phenomenon precisely. To evaluate the applicability of MVP to hydride absorber rod design, two benchmark calculations were carried out. One of them is a hydrogen-contained metal fuel fast core constructed in Fast Critical Assembly (FCA) and the other is the Nuclear Safety Research Reactor (NSRR) core where zirconium-hydride fuel (U-ZrH1.6) rods are loaded. These benchmark calculations and the design study on a fast reactor core with hafnium-hydride control rods have revealed that MVP is a reliable tool for hydride absorber rod design.

Development of Hydride Absorber for Fast Reactor —Evaluation of Design Method for Hydride Absorber Rod—

A hydride control rod is being developed to improve the economy of fast reactor plants because it has a longer lifetime than the currently used B4C control rod. A hydride burnable poison rod is also under development to reduce the number of control rods by decreasing core excess reactivity. Hydrogen in the hydride control rod causes neutron spectrum interference between the fuel and control rod regions. Thus, the study on core design was performed with the continuous-energy Monte Carlo code MVP using the nuclear data library JENDL-3.3 to deal with this phenomenon precisely. To evaluate the applicability of MVP to hydride absorber rod design, two benchmark calculations were carried out. One of them is a hydrogen-contained metal fuel fast core constructed in Fast Critical Assembly (FCA) and the other is the Nuclear Safety Research Reactor (NSRR) core where zirconium-hydride fuel (U-ZrH1.6) rods are loaded. These benchmark calculations and the design study on a fast reactor core with hafnium-hydride control rods have revealed that MVP is a reliable tool for hydride absorber rod design.