Monte Carlo Code MVP Research Articles

Experimental Analyses of Capture Reaction Rates for Epithermal and Resonance Neutrons in Source-Driven Subcritical Cores

Neutron irradiation experiments are carried out in source-driven subcritical cores with high-energy neutrons generated by spallation reactions between a 100-MeV proton beam and a lead-bismuth target at the Kyoto University Critical Assembly. The main objective of the experiments is to investigate the effect of epithermal and resonance neutrons on the accuracy of capture reaction rates with respect to a subcriticality variation. Activation foils of copper, indium, tantalum, and tungsten are employed to obtain capture reaction rates for epithermal and resonance neutrons by applying the cadmium difference method. Also, the applicability of the foils for the measurement of the reaction rates for epithermal and resonance neutrons is substantiated in the critical irradiation experiments performed prior to the subcritical experiments. The subcritical experiments are conducted with three different subcriticalities by changing the control rod insertion pattern. The measured reaction rates are compared with the calculated values obtained by the Monte Carlo code MVP with JENDL-4.0, and the ratio of the calculation and experiment values of the reaction rates shows equivalent values within the 1σ errors regardless of a difference in the subcriticality. The compared results indicate that the numerical analyses have a consistent accuracy of reaction rates in epithermal and resonance energy regions for a subcriticality variation in source-driven subcritical cores.

Source multiplication measurements and neutron correlation analyses for a highly enriched uranium subcritical core driven by an inherent source in Kyoto University Critical Assembly

ABSTRACT For a highly enriched uranium subcritical core driven by an inherent or an external Am-Be source in Kyoto University Critical Assembly (KUCA), a series of the source multiplication measurements, the Feynman-α and the Rossi-α neutron correlation analyses was carried out to demonstrate quantitatively that the use of the former inherent source had some advantages over that of the latter external source. Under a reactor drive by the external source, the subcriticality obtained by the neutron source multiplication method (NSM) significantly depended on neutron counter position and was consequently in very poor agreement with that calculated by the continuous-energy Monte Carlo code MVP (version 3) with the nuclear data library JENDL-4.0. In contrast to the above drive, the other drive by the inherent source spectacularly reduced the counter-position dependence and lead to a good agreement with the calculation by the MVP. The prompt-neutron decay constant determined from the Feynman-α analysis under the external source was also significantly dependent on the counter position. The Rossi-α analysis had a relatively slight counter-position dependence of the decay constant to the Feynman-α analysis but had a very low S/N ratio, which indicated an inferior applicability of the Rossi-α analysis to the reactor drive and enlarged a statistical uncertainty of the decay constant. On the other hand, the reactor drive by the inherent source considerably reduced the counter-position dependence of the Feynman-α analysis and spectacularly enhanced the S/N ratio of the Rossi-α analysis. An effective strength of the inherent source could be also determined from the S/N ratio. When the prompt-neutron decay constant obtained under this reactor drive was transformed into each subcriticality by the Simmons and King method, where a subcritical dependence of the generation time and the effective delayed-neutron fraction was neglected, the subcriticality transformed seemed to agree well with that calculated by the MVP. When the decay constant was transformed by the inhour method, where the subcritical dependence of these kinetic parameters was considered, a systematic disagreement between the transformed and the calculated subcriticality appeared.

Measurement of a very large negative reactivity inserted by rapid withdrawal of a partial fuel loading in Kyoto University Critical Assembly

ABSTRACT Solid-moderated cores of the Kyoto University Critical Assembly (KUCA) have an emergency shutdown system to avoid any reactivity-initiated accidents. When a reactor protection system of KUCA detects some abnormal conditions, the emergency system withdraws downward several fuel and reflector assemblies assigned as a central fuel loading from the core in 10 and several seconds. The withdrawal of the fuel loading inserts a very large negative reactivity into the core. The negative reactivity has been measured by the integral count technique; however, the application of the integral count technique to this measurement must have been impossible in principle. The integral count technique assumes an instantaneous insertion of a large negative reactivity but the assumption must fail for this measurement. An accurate measurement of the negative reactivity is of extreme importance from the viewpoint of reactor safety. In this study, the least-squares inverse kinetics method was employed to determine the negative reactivity from time-sequence counts data of a neutron counter. The negative reactivity determined by this method was in good agreement with that calculated by the continuous-energy Monte Carlo code MVP (version 3) with the nuclear data library JENDL-4.0.

Conceptual Study on Recriticality Prevention Core Having Duplex Pellets with Neutron Absorber in Outer Core in a Fast Reactor

In a fast reactor, we evaluated a new core concept that prevents severe recriticality after whole-scale molten formation in a severe accident. A core concept in which Duplex pellets including neutron absorber are loaded in the outer core has been proposed. Analysis by the continuous energy model Monte Carlo code MVP using the JENDL-4.0 nuclear data library revealed that this fast reactor core has large negative reactivity due to fuel melting at the time of a severe accident, so that the core prevents recriticality. Regarding the core nuclear and thermal characteristics, the loading of Duplex pellets including neutron absorber in the outer core caused no significant differences from the normal core without Duplex pellets.

An impact of inherent neutron source on subcriticality measurement in a highly enriched uranium core of Kyoto University Critical Assembly

ABSTRACT In a variety of highly enriched uranium cores of Kyoto University Critical Assembly, many different subcriticality measurements have been strenuously performed. However, any influence of neutron source inherent in the highly enriched uranium fuels on these measurements has hardly been studied. This is because the influence has been expected to be negligible in the fuels. In this study, we revaluated the influence on pulsed neutron, accelerator-beam trip and rod drop measurements to reveal an unexpected impact of the weak inherent source. Especially, the inherent source was injurious to most of the beam trip and the rod drop measurements based on the integral count method. The least-squares inverse kinetics analysis also had a significant influence on the inherent source. In the area ratio analysis for a pulsed neutron measurement, a considerable number of neutrons from the inherent source was mixed into delayed-neutron area. When the influence was considered in these data analyses, the subcritical reactivity of the above measurements was in good agreement with that calculated by the continuous-energy Monte Carlo code MVP.

Resonance self-shielding treatment and analysis of resonance integral tables for Fully Ceramic Micro-encapsulated fuels with the Embedded Self-Shielding Method

A resonance calculation method for Fully Ceramic Micro-encapsulated (FCM) fuels containing stochastically dispersed tri-structural isotropic (TRISO) coated fuel particles has been proposed in this research. The purpose is to obtain effective homogenized resonance cross sections representing the effect of double heterogeneity (DH). The Monte Carlo code MVP is used to solve the heterogeneity in the isolated FCM fuels pin, and to couple with the method of characteristic (MOC) to generate heterogeneous resonance integral (RI) tables for the following Embedded Self-Shielding Method (ESSM) based resonance calculation. The ESSM method is employed to treat the heterogeneity in the lattice (Dancoff effect). Heterogeneous RI tables based on different packing fraction may differ in relationship between background and self-shielded cross sections. Analysis of this distinction’s impact on ESSM calculation has been conducted. Furthermore, resonance interference effect is neglected in ESSM because heterogeneous RI tables are developed for isolated resonant nuclide system. Therefore, Resonance Interference Factor (RIF) Method was applied to correct cross sections from ESSM. To obtain RIF, slowing-down equation need be solved twice for different homogeneous systems. Whereas, the cross section neglecting interference effect could be received by interpolating in homogeneous RI tables to improve efficiency. Evaluation of errors caused by interpolation in RIF method was performed with reference solutions from MVP. Numerical results show that the new method is proved effective in DH capability for resonance self-shielding calculation.

Study of the neutron multiplication effect in an active neutron method

ABSTRACTThe neutron multiplication effect appears when an item contains large amounts of nuclear material. The neutron multiplication effect in this paper means the effect of subsequent fission reactions which are caused by fission neutrons produced by interrogation neutrons from a neutron generator. The previous active neutron method could not distinguish between first-fission and subsequent-fission neutrons and might overestimate the amount of nuclear material. However, the neutron multiplication effect in the active neutron method has not been adequately investigated. We discuss the evaluation method of the multiplication effect in the fast neutron direct interrogation method, one of the active neutron methods, using simulations with the Monte Carlo code MVP and experiments involving uranium waste drums. The first-generation neutrons from an external neutron source generate fission neutrons called second-generation neutrons, the second-generation neutrons generate third-generation neutrons, and so on. This study supposes that the neutron multiplication effect is mainly caused by the third-generation neutrons under the condition that the fourth-generation neutrons are much fewer. This paper proposes a correction method for the neutron multiplication effect in the measured data.

Design and burn-up analyses of new type holder for silicon neutron transmutation doping

We have developed a new silicon irradiation holder with a neutron filter to increase the irradiation efficiency. The neutron filter is made of an alloy of aluminum and B4C particles. We fabricated a new holder based on the results of design analyses. This filter has limited use in applications requiring prolonged use due to a decrease in the amount of 10B in B4C particles. We investigated the influence of 10B reduction on doping distribution in a silicon ingot by using the Monte Carlo Code MVP.

Validation of Burnup Calculation Code SWAT4 by Evaluation of Isotopic Composition Data of Mixed Oxide Fuel Irradiated in Pressurized Water Reactor

Abstract This paper provides validation results of SWAT4, the revised version of SWAT3.1, by the analysis of PIE data of MOX fuel irradiated in a pressurized water reactor. SWAT4 is a system that combines the point burnup system ORIGEN2 and neutron transport calculation solvers, the continuous energy Monte Carlo code MVP or MCNP5, and the deterministic neutronics calculation code SRAC2006. The calculation results of SWAT4 have generally same trends with the case of UO2 fuel analyses. For major uranium and plutonium isotopes, deviations less than 5% were obtained. This means that SWAT4 has the same accuracy to predict isotopic compositions of irradiated MOX fuel with the case of UO2 fuel. The radial distribution of isotopes in a pellet was also analyzed, whose results were compared with that measured by SIMS. SWAT4 predicted well the isotope and burnup distributions in an irradiated MOX pellet.

Recent developments of JAEA’s Monte Carlo code MVP for reactor physics applications

This paper describes the recent development status of a Monte Carlo code MVP developed at Japan Atomic Energy Agency. The basic features and capabilities of MVP are overviewed. In addition, new capabilities useful for reactor analysis are also described.

Azimuthal flux distribution measurements around fuel rods in reduced-moderation LWR lattices

An advanced analysis method named “micro reactor physics approach” was proposed, and the approach is needed for future reactor design considering the neutron behavior in fuel pellets. In order to validate the approach, neutron flux distribution measurements in a fuel pellet should be required. We have measured azimuthal flux distribution of fuel rods in Toshiba Nuclear Critical Assembly (NCA). A foil activation method with metallic foils was used for the measurement. Measured values were analyzed by a continuous energy Monte Carlo code MVP with the JENDL-3.3 library. The measurements are useful for the validation of an advanced fuel design method considering the neutron behavior in fuel pellets.

Monte Carlo calculations for dose rate at indoors exposed by distributed Cs-137 activities

The correlation of dose rates inside of the typical Japanese house to different distributions of Cs-137 activities outside are calculated by using the Monte Carlo code MVP with its attached gamma cross section library for aiming at the planning of decontamination strategies and setting the priorities in order to decrease the amounts of waste soil effectively, which will be disposed and stored for a long time with shield. The study is summarized that (1) the decontamination of the roof is not effective in decreasing the dose rate of indoors when surrounding field is contaminated, (2) the plowing of wide area will decrease the dose rate by one tenth even if the depth is only 5 cm, (3) the dose rate of the 2nd floor is higher than that of the 1st floor in some decontamination cases.

Benchmark analysis of SPERT-IV reactor with Monte Carlo code MVP

The benchmark experiment of the SPERT-IV D-12/25 reactor core has been analyzed with the Monte Carlo code MVP using the cross-section libraries based on JENDL-3.3. The MVP simulation was performed for the clean and cold core. The estimated values of Keff at the experimental critical rod height and the core excess reactivity were within 5% with the experimental data. Thermal neutron flux profiles at different vertical and horizontal positions of the core were also estimated. Cadmium Ratio at different point of the core was also estimated. All estimated results have been compared with the experimental results. Generally good agreement has been found between experimentally determined and the calculated results.

Improvement of Core Performance by Introduction of Moderators in a Blanket Region of Fast Reactors

An application of deuteride moderator for fast reactor cores is proposed for power flattening that can mitigate thermal spikes and alleviate the decrease in breeding ratio, which sometimes occurs when hydrogen moderator is applied as a moderator. Zirconium deuteride is employed in a form of pin arrays at the inner most rows of radial blanket fuel assemblies, which works as a reflector in order to flatten the radial power distribution in the outer core region of MONJU. The power flattening can be utilized to increase core average burn-up by increasing operational time. The core characteristics have been evaluated with a continuous-energy model Monte Carlo code MVP and the JENDL-3.3 cross-section library. The result indicates that the discharged fuel burn-up can be increased by about 7% relative to that of no moderator in the blanket region due to the power flattening when the number of deuteride moderator pins is 61. The core characteristics and core safety such as void reactivity, Doppler coefficient, and reactivity insertion that occurred at dissolution of deuteron were evaluated. It was clear that the serious drawback did not appear from the viewpoints of the core characteristics and core safety.

Conceptual design for a small modular district heating reactor for Mongolia

The energy demand in Mongolia is expected to increase in the near future. The purpose of the present study was to design a small, modular district heating reactor with passive safety features for long core life to meet the country’s expected energy demand. The design concept was based on the advanced and well-proven technology of the high temperature gas-cooled reactor (HTGR) idea. The reactor size was roughly estimated without any active cooling system to remove the decay heat, and the optimal core dimensions chosen were 8m diameter and 8m height with 330MWth of power. Fundamental analyses of the reactor for criticality and burnup were performed by using continuous energy Monte Carlo code MVP 2.0, and the proposed reactor with 20wt% of uranium enrichment can be critical around 16years. The initial effective neutron multiplication factor can be suppressed to 1.1351 when there are 9 burnable poison (BP) rods with 12wt% of boron carbide in B4C/C composite and 10wt% of 10B per assembly.

A study on TRIGA core reconfiguration with new irradiation channels

This study is being carried on as a provision of achieving enhanced irradiation facilities in the TRIGA MARK II research reactor core. Two reconfigured cores, each one having three irradiation channels are being analyzed with the Monte Carlo code MVP and the nuclear data library JENDL-3.3. The existing core has only one irradiation channel at the center. The graphite dummies are rearranged to surround the new irradiation channels in the new configurations. The results for each reconfigured core are compared with the existing core as reference. From this study important information are obtained for enhanced as well as economic utilization of TRIGA core. The results show positively the possibility to introduce new irradiation channels without disturbing the core excess reactivity and neutron flux significantly as well as keeping the shutdown margin and peaking factor close to the reference.

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.

Analysis of Core Physics Experiments on Fresh and Irradiated BR3 MOX Fuel in REBUS Program

As part of an international experimental program REBUS, core physics experiments have been implemented on a UO2 core, which consists of 3.3 and 4.0 wt% UO2 fuel rods in a square pitch of 1.26 cm, and two partial MOX cores, which replace 7 × 7 UO2 fuel rods in the center of the UO2 core by fuel bundles made of fresh BR3 MOX fuel or irradiated BR3 MOX fuel with an average burnup of 20GWd/t. Burnup calculations of the BR3 MOX fuel were performed using a general-purpose neutronic calculation code SRAC, and core calculations of the three critical cores were carried out using SRAC, a transport calculation code THREEDANT, and a continuous-energy Monte Carlo code MVP. The measured inventories of major U and Pu isotopes on a sample taken from the BR3 MOX fuel agree with the results of the burnup calculations within 3% deviation. The k eff's of the three cores are from 0.985 to 1.002. The measured burnup reactivity of the irradiated BR3 MOX fuel was well reproduced by the three types of core calculations. The influence of the accuracy of the inventory calculations on burnup reactivity was studied by comparing between the calculated and measured inventories. The result indicates that the biases in the inventory and reactivity calculations compensate each other, and it makes the total biases of the burnup reactivity small.

Analysis of Core Physics Experiments on Fresh and Irradiated BR3 MOX Fuel in REBUS Program

As part of an international experimental program REBUS, core physics experiments have been implemented on a UO2 core, which consists of 3.3 and 4.0 wt% UO2 fuel rods in a square pitch of 1.26 cm, and two partial MOX cores, which replace 7 × 7 UO2 fuel rods in the center of the UO2 core by fuel bundles made of fresh BR3 MOX fuel or irradiated BR3 MOX fuel with an average burnup of 20GWd/t. Burnup calculations of the BR3 MOX fuel were performed using a general-purpose neutronic calculation code SRAC, and core calculations of the three critical cores were carried out using SRAC, a transport calculation code THREEDANT, and a continuous-energy Monte Carlo code MVP. The measured inventories of major U and Pu isotopes on a sample taken from the BR3 MOX fuel agree with the results of the burnup calculations within 3% deviation. The k eff's of the three cores are from 0.985 to 1.002. The measured burnup reactivity of the irradiated BR3 MOX fuel was well reproduced by the three types of core calculations. The influence of the accuracy of the inventory calculations on burnup reactivity was studied by comparing between the calculated and measured inventories. The result indicates that the biases in the inventory and reactivity calculations compensate each other, and it makes the total biases of the burnup reactivity small.