Criticality Safety Evaluation Research Articles

Criticality evaluation considering nonuniformity effect using Monte Carlo perturbation method

ABSTRACT The objective of this study is to evaluate nonuniformity effect relevant to criticality safety evaluation by means of the flattened principle of fuel importance distribution. We proposed a novel approach using an MVP3.0 perturbation function with JENDL-4.0 and then built a tool to calculate the nonuniformity effect. The tool estimated an optimum fuel distribution that corresponding to the highest multiplication factor in a homogeneity model of UO2–H2O slurry. Consequently, nonuniformity effects specified by the reactivity of a multiplication factor were as much as 0.5 to 5.6%dk/k. Further, preliminary evaluations of heterogeneity effects on optimum fuel distributions were computed using statistical geometry model incorporated in MVP3.0 and then heterogeneity effects observed in optimum fuel distributions also reached 2.3%dk/k at the maximum in multiplication factor. Thus, the effects of both nonuniformity and heterogeneity should be properly taken into account in criticality evaluations.

Nuclear Data Uncertainty Quantification in Criticality Safety Evaluations for Spent Nuclear Fuel Geological Disposal

Presently, a criticality safety evaluation methodology for the final geological disposal of Swiss spent nuclear fuel is under development at the Paul Scherrer Institute in collaboration with the Swiss National Technical Competence Centre in the field of deep geological disposal of radioactive waste. This method in essence pursues a best estimate plus uncertainty approach and includes burnup credit. Burnup credit is applied by means of a computational scheme called BUCSS-R (Burnup Credit System for the Swiss Reactors–Repository case) which is complemented by the quantification of uncertainties from various sources. BUCSS-R consists in depletion, decay and criticality calculations with CASMO5, SERPENT2 and MCNP6, respectively, determining the keff eigenvalues of the disposal canister loaded with the Swiss spent nuclear fuel assemblies. However, the depletion calculation in the first and the criticality calculation in the third step, in particular, are subject to uncertainties in the nuclear data input. In previous studies, the effects of these nuclear data-related uncertainties on obtained keff values, stemming from each of the two steps, have been quantified independently. Both contributions to the overall uncertainty in the calculated keff values have, therefore, been considered as fully correlated leading to an overly conservative estimation of total uncertainties. This study presents a consistent approach eliminating the need to assume and take into account unrealistically strong correlations in the keff results. The nuclear data uncertainty quantification for both depletion and criticality calculation is now performed at once using one and the same set of perturbation factors for uncertainty propagation through the corresponding calculation steps of the evaluation method. The present results reveal the overestimation of nuclear data-related uncertainties by the previous approach, in particular for spent nuclear fuel with a high burn-up, and underline the importance of consistent nuclear data uncertainty quantification methods. However, only canister loadings with UO2 fuel assemblies are considered, not offering insights into potentially different trends in nuclear data-related uncertainties for mixed oxide fuel assemblies.

Analyses of used BWR fuel assay data with the integrated burnup code system SWAT4.0

ABSTRACT Criticality safety of the fuel debris from the Fukushima Daiichi Nuclear Power Plant is one of the most important issues, and the adoption of burnup credit is desired for criticality safety evaluation. To adopt the burnup credit, validation of the burnup calculation codes is required. Assay data of the used nuclear fuel irradiated by the Fukushima Daini Nuclear Power Plant Unit 2 are evaluated to validate the SWAT4.0 code for solving the BWR fuel burnup problem. The calculation results revealed that the number densities of many heavy nuclides and fission products show good agreement with the experimental data, except for those of 237Np, 238Pu, and samarium isotopes. These differences were considered to originate from inappropriate assumption of void fraction. Our results implied overestimation of the (n, γ) cross-section of 237Np in JENDL-4.0. The Calculation/Experiment – 1 (C/E–1) value did not depend on the type of fuel rod (UO2 or UO2–Gd2O3), which was similar to the case of PWR fuel. The differences in the number densities of 235U, 239Pu, 240Pu, 241Pu, 149Sm, and 151Sm have a large impact on k eff. However, the reactivity uncertainty related to the burnup analysis was less than 3%. These results indicate that SWAT4.0 appropriately analyzes the isotopic composition of BWR fuel, and it has sufficient accuracy to be adopted in the burnup credit evaluation of fuel debris.

Criticality Safety Evaluation of a Swiss wet storage pool using a global uncertainty analysis methodology

Uncertainty quantification is a key component in the Criticality Safety Evaluation (CSE) of spent nuclear fuel systems. An important source of uncertainties is caused by manufacturing and technological parameter tolerances. In this work, such class of uncertainties are evaluated for a Swiss wet storage pool. The selected configuration corresponds to a case where the target criticality eigenvalue is close to the upper criticality safety limits. Although current PSI CSE safety criteria are fulfilled, it is reasonable to apply uncertainty quantification methodologies in order to provide the regulatory authorities with additional information relevant for safety evaluations.The MTUQ (Manufacturing and Technological Uncertainty Quantification) methodology, based on global stochastic sampling was the selected tool for the analysis. Such tool is specifically designed for the treatment of geometrical/material uncertainties for any target system. In particular the MTUQ advanced modelling capability allows the implementation of realistic boundary condition, with a resulting detailed evaluation of statistical quantities of interest in CSE. Therein, the computational code implemented is the MCNP Monte Carlo based neutron transport code.The analysis showed the benefits in using realistic modelling compared to the traditional one-factor-at-time methodology applied to system modelled using repeated structures. A detailed comparison between the 2 approaches is also presented. Finally, it is discussed the role of asymmetrical probability distribution functions to feed the global statistical methods.

Application of the PSI-NUSS Tool for the Estimation of Nuclear Data Related [formula omitted] Uncertainties for the OECD/NEA WPNCS UACSA Phase I Benchmark

At the Paul Scherrer Institute (PSI), a methodology titled PSI-NUSS is under development for the propagation of nuclear data uncertainties into Criticality Safety Evaluation (CSE) with the Monte Carlo code MCNPX. The primary purpose is to provide a complementary option for the uncertainty assessment related to nuclear data, versus the traditional approach which relies on estimating biases/uncertainties based on validation studies against representative critical benchmark experiments. In the present paper, the PSI-NUSS methodology is applied to quantify nuclear data uncertainties for the OECD/NEA UACSA Exercise Phase I benchmark. One underlying reason is that PSI's CSE methodology developed so far and previously applied for this benchmark was based on using a more conventional approach, involving engineering guesses in order to estimate uncertainties in the calculated effective multiplication factor (keff). Therefore, as the PSI-NUSS methodology aims precisely at integrating a more rigorous treatment of the specific type of uncertainties from nuclear data for CSE, its application to the UACSA is conducted here: nuclear data related uncertainty component is estimated and compared to results obtained by other participants using different codes/libraries and methodologies.

Review of studies on criticality safety evaluation and criticality experiment methods

Since the early 1960s, many studies on criticality safety evaluation have been conducted in Japan. Computer code systems were developed initially by employing finite difference methods, and more recently by using Monte Carlo methods. Criticality experiments have also been carried out in many laboratories in Japan as well as overseas. By effectively using these study results, the Japanese Criticality Safety Handbook was published in 1988, almost the intermediate point of the last 50 years. An increased interest has been shown in criticality safety studies, and a Working Party on Nuclear Criticality Safety (WPNCS) was set up by the Nuclear Science Committee of Organisation Economic Co-operation and Development in 1997. WPNCS has several task forces in charge of each of the International Criticality Safety Benchmark Evaluation Program (ICSBEP), Subcritical Measurement, Experimental Needs, Burn-up Credit Studies and Minimum Critical Values. Criticality safety studies in Japan have been carried out in cooperation with WPNCS. This paper describes criticality safety study activities in Japan along with the contents of the Japanese Criticality Safety Handbook and the tasks of WPNCS.

Criticality safety analysis for a core catcher designed in Korea

Corium is a molten mixture of portions of a reactor core generated by a core melting accident. Corium includes fissionable materials; therefore, a criticality safety analysis must be performed for the core catcher design. This study analyzes the criticality safety of corium arranged in a core catcher developed in Korea. The corium composition was calculated for a 1400MWe nuclear power plant. There are several variables involved in the criticality evaluation of corium, thus conservative assumptions were used to reduce the number of variables. A criticality evaluation procedure was employed to assess the operational failure of the core catcher under different accident scenarios. Four kinds of scenarios were selected, and criticality evaluations were pursued for each case. The multiplication factors in each condition were calculated with MCNP5 code. Also, the code bias was calculated with the benchmark problems of 262 LEU experiments to account for the uncertainty of MCNP code. All evaluation results for the assumed scenarios showed that the core catcher satisfies the regulatory guidelines for criticality safety. The calculation results will be used in the design of a core catcher being developed in Korea. It is expected that the data calculated in this study can be used as reference data for criticality safety evaluations of core melting accidents. Also, the procedure for criticality safety evaluation proposed in this study can be utilized to establish regulatory guidelines in Korea.

Identification of Integral Benchmarks for Nuclear Data Testing Using DICE (Database for the International Handbook of Evaluated Criticality Safety Benchmark Experiments)

Typical users of the International Criticality Safety Evaluation Project (ICSBEP) Handbook have specific criteria to which they desire to find matching experiments. Depending on the application, those criteria may consist of any combination of physical or chemical characteristics and/or various neutronic parameters. The ICSBEP Handbook contains a structured format helping the user narrow the search for experiments of interest. However, with nearly 4300 different experimental configurations and the ever increasing addition of experimental data, the necessity to perform multiple criteria searches have rendered these features insufficient. As a result, a relational database was created with information extracted from the ICSBEP Handbook. A users’ interface was designed by OECD and DOE to allow the interrogation of this database. The database and the corresponding users’ interface are referred to as DICE. DICE currently offers the capability to perform multiple criteria searches that go beyond simple fuel, physical form and spectra and includes expanded general information, fuel form, moderator/coolant, neutron-absorbing material, cladding, reflector, separator, geometry, benchmark results, spectra, and neutron balance parameters. DICE also includes the capability to display graphical representations of neutron spectra, detailed neutron balance, sensitivity coefficients for capture, fission, elastic scattering, inelastic scattering, nu-bar and mu-bar, as well as severalmore » other features.« less

Criticality Safety Evaluation of Materials Concerning Pyroprocessing

Criticality evaluation of materials concerning pyroprocessing has been performed employing Monte Carlo techniques. Fresh UO2 fuel and spent PWR fuel have been employed in an electroreducer and, according to the pyroprocessing material flow, their metallic products are also evaluated correspondingly. TRU and Pu metals are evaluated concerning the electrowinner products. Totally, 6 kinds of reflectors and 4 stuffing materials are introduced to simulate the complexity during the pyroprocessing. With the optimized water intrusion models, certain subcritical data have been determined: 33.6 kg for UO2 fuel in electroreducer, 617.5 kg for spent fuel in electroreducer, 613.6 kg for metallic spent fuel in electroreducer, and 40.0 kg for U in electrorefiner. The conservative cases for electrowinner indicate the subcritical masses of 6.8 kg for Pu and 7.9 kg for TRU. However, with regard to the dilution effect of Cd, the Pu mass needed to reach the subcriticality changes from 11.1 kg with 1.2 kg of Cd to 500.0 kg with 2000.0 kg of Cd. The TRU mass needed to reach subcriticality changes from 13.0 kg with 1.4 kg Cd to 714.0 kg with 2856.0 kg of Cd. TSUNAMI has been employed to clarify the contributions of each nuclide quantitatively. The data obtained would therefore be used in the conceptual design study of pyroprocessing facilities.

Criticality Safety Evaluation of Materials Concerning Pyroprocessing

Criticality Safety Evaluation of Materials Concerning Pyroprocessing

The effect of modern thermal neutron scattering sublibraries on criticality safety evaluations of wet storage pools

Within the framework of a recent criticality safety evaluation (CSE) performed for the licensing of a new commercial wet storage pool using MCNPX-2.5.0 along with the ENDF/B-VII.0 and JEFF-3.1 continuous energy cross-section libraries, the maximum permissible initial fuel-enrichment limit for water reflected configurations was found to be dependant upon the applied neutron cross section library. More detailed investigations showed that the difference is mainly caused by different sublibraries for thermal neutron scattering based on parameterizations of the S ( α , β ) scattering matrix. This paper addresses the main findings of these investigations and emphasizes which configurations are particularly sensitive to the parameterization of S ( α , β ) applied in the CSE analysis. The results are presented in an attempt, on the one hand, to assess more precisely the sensitivity upon thermal scattering modeling data in CSE of wet storage pools, and on the other hand, to assess whether the data employed in the most recent neutron data libraries can be considered as an improvement for this type of analyses.

HTC Experimental Program: Validation and Calculational Analysis

In the 1980s a series of the Haut Taux de Combustion (HTC) critical experiments with fuel pins in a water-moderated lattice was conducted at the Apparatus B experimental facility in Valduc (Commissariat á l’Energie Atomique, France) with the support of the Institut de Radioprotection et de Sûreté Nucléaire and AREVA NC. Four series of experiments were designed to assess profit associated with actinide-only burnup credit in the criticality safety evaluation for fuel handling, pool storage, and spent-fuel cask conditions. The HTC rods, specifically fabricated for the experiments, simulated typical pressurized water reactor uranium oxide spent fuel that had an initial enrichment of 4.5 wt% 235U and was burned to 37.5 GWd/tonne U.The configurations have been modeled with the CRISTAL criticality package and SCALE 5.1 code system. Sensitivity/uncertainty analysis has been employed to evaluate the HTC experiments and to study their applicability for validation of burnup credit calculations. This paper presents the experimental program, the principal results of the experiment evaluation, and modeling. The HTC data applicability to burnup credit validation is demonstrated with an example of spent-fuel storage models.

Nuclear Criticality Safety Evaluation of a Mixture of MOX, UO2 and Additive in the Most Conservative Concentration Distribution

The nuclear criticality safety evaluation of blenders that are used at a mixed uranium-plutonium oxide (MOX) fuel plant must take into account the nonuniform distribution of powders in three principal components, i.e., MOX, uranium dioxide (UO2) and zinc stearate, which is a fuel additive. The model blender considered in this article contained a mixture of 33 wt% PUO2-enriched MOX, depleted UO2 and zinc stearate in the form of an upside-down truncated cone that was surrounded by 30-cm-thick polyethylene. To limit the number of calculation cases, the fissile plutonium mass of the mixture was fixed at 98 kg, and the total concentration of MOX and UO2 was fixed at 4.0 g/cm3. The most conservative fuel distribution with respect to nuclear criticality safety under these constraints was calculated with a two-dimensional optimum fuel distribution code OPT-TWO, so that the importance distribution of MOX and that of zinc stearate could be individually flattened by conserving the mass of each component. The OPT-TWO calculation was followed by a criticality calculation performed with the MCNP code to obtain the neutron multiplication factor of the fuel system in the optimum fuel distribution. The most conservative fuel distribution obtained in this research was typically depicted as a layer of zinc stearate embedded within the central MOX region that was surrounded by the peripheral UO2 region. An increase of up to 25% in the neutron multiplication factor was found; two factors with comparable but independent contributions were the nonuniform concentrations of plutonium enrichment and zinc stearate.

Nuclear Criticality Safety Evaluation of a Mixture of MOX, UO2 and Additive in the Most Conservative Concentration Distribution

The nuclear criticality safety evaluation of blenders that are used at a mixed uranium-plutonium oxide (MOX) fuel plant must take into account the nonuniform distribution of powders in three principal components, i.e., MOX, uranium dioxide (UO2) and zinc stearate, which is a fuel additive. The model blender considered in this article contained a mixture of 33 wt% PUO2-enriched MOX, depleted UO2 and zinc stearate in the form of an upside-down truncated cone that was surrounded by 30-cm-thick polyethylene. To limit the number of calculation cases, the fissile plutonium mass of the mixture was fixed at 98 kg, and the total concentration of MOX and UO2 was fixed at 4.0 g/cm3. The most conservative fuel distribution with respect to nuclear criticality safety under these constraints was calculated with a two-dimensional optimum fuel distribution code OPT-TWO, so that the importance distribution of MOX and that of zinc stearate could be individually flattened by conserving the mass of each component. The OPT-TWO calculation was followed by a criticality calculation performed with the MCNP code to obtain the neutron multiplication factor of the fuel system in the optimum fuel distribution. The most conservative fuel distribution obtained in this research was typically depicted as a layer of zinc stearate embedded within the central MOX region that was surrounded by the peripheral UO2 region. An increase of up to 25% in the neutron multiplication factor was found; two factors with comparable but independent contributions were the nonuniform concentrations of plutonium enrichment and zinc stearate.

Calculation of Criticality Condition Data for Single-unit Homogeneous Uranium Materials in Six Chemical Forms

Single-unit criticality condition data were calculated for homogeneous uranium materials in six chemical forms for revision of the Data Collection appendix to the Nuclear Criticality Safety Handbook. The calculated criticality condition data were the estimated critical and the estimated lower-limit critical masses and volumes of spheres, diameters of infinitely-long cylinders and thicknesses of infinite-area slabs of uranium materials in six chemical forms encountered in criticality safety evaluation of nuclear fuel cycle facilities. The chemical forms were U-H2O, UO2-H2O, UO2F2, UO2(NO3)2, ADU-H2O and UF6-HF, the first and last ones of which were not dealt with the Data Collection. The calculations were made with the continuous-energy Monte Carlo criticality calculation code MVP and the Japanese Evaluated Nuclear Data Library JENDL-3.2. The values and precision of the present calculations are discussed in comparison with previous results.

Calculation of Criticality Condition Data for Single-unit Homogeneous Uranium Materials in Six Chemical Forms

Single-unit criticality condition data were calculated for homogeneous uranium materials in six chemical forms for revision of the Data Collection appendix to the Nuclear Criticality Safety Handbook. The calculated criticality condition data were the estimated critical and the estimated lower-limit critical masses and volumes of spheres, diameters of infinitely-long cylinders and thicknesses of infinite-area slabs of uranium materials in six chemical forms encountered in criticality safety evaluation of nuclear fuel cycle facilities. The chemical forms were U-H2O, UO2-H2O, UO2F2, UO2(NO3)2, ADU-H2O and UF6-HF, the first and last ones of which were not dealt with the Data Collection. The calculations were made with the continuous-energy Monte Carlo criticality calculation code MVP and the Japanese Evaluated Nuclear Data Library JENDL-3.2. The values and precision of the present calculations are discussed in comparison with previous results.

Calculation of Nuclear Characteristic Parameters and Drawing Subcriticality Judgment Graphs of Infinite Fuel Systems for Typical Nuclear Fuels

Nuclear characteristic parameters were calculated and subcriticality judgment graphs were drawn for revision of the Data Collection for the Nuclear Criticality Safety Handbook. The calculated nuclear characteristic parameters were the neutron multiplication factor in infinite media, migration area and diffusion constant for 11 kinds of typical fuels encountered in criticality safety evaluation of nuclear fuel cycle facilities. These fuels included ADU-H2O, UF6-HF and Pu(NO3)4-UO2(NO3)2 solution, of which data were not cited in the Data Collection. The calculation was made with the Japanese Evaluated Nuclear Data Library JENDL-3.2 and a sequence of criticality calculation codes, SRAC, POST and SIMCRI. The subcriticality judgment graphs that depict the region of the two variables (a) uranium enrichment, 239Pu/Pu ratio or plutonium enrichment and (b) H/(Pu+U) ratio where the neutron multiplication factor in infinite media k∞≤0.98 were drawn for the same kinds of fuels except UF6-HF. The limitations of the subcriticality judgment graphs were also discussed.

Validation of the Continuous-Energy Monte Carlo Criticality-Safety Analysis System MVP and JENDL-3.2 Using the Internationally Evaluated Criticality Benchmarks

Validation of the continuous-energy Monte Carlo criticality-safety analysis system, comprising the MVP code and neutron cross sections based on JENDL-3.2, was examined using benchmarks evaluated in the “International Handbook of Evaluated Criticality Safety Benchmark Experiments.” Eight experiments (116 configurations) for the plutonium solution and plutonium-uranium mixture systems performed at Valduc, Battelle Pacific Northwest Laboratories, and other facilities were selected and used in the studies. The averaged multiplication factors calculated with MVP and MCNP-4B using the same neutron cross-section libraries based on JENDL-3.2 were in good agreement. Based on methods provided in the Japanese nuclear criticality-safety handbook, the estimated criticality lower-limit multiplication factors to be used as a subcriticality criterion for the criticality-safety evaluation of nuclear facilities were obtained. The analysis proved the applicability of the MVP code to the criticality-safety analysis of nuclear fuel facilities, particularly to the analysis of systems fueled with plutonium and in homogeneous and thermal-energy conditions.

Burnup Importance Function Introduced to Give an Insight into the End Effect

In order to facilitate discussions based on quantitative analysis about the end effect, which was often talked about in connection to burnup credit in criticality safety evaluation of spent fuel, a burnup importance function was introduced. This function showed the burnup effect on the reactivity as a function of the fuel position; an explicit expression of this function was derived by considering a change in reactivity with respect to a slight variation in fuel burnup. The burnup importance function was applied to the Phase IIA benchmark model that was adopted by the Organization for Economic Cooperation and Development/Nuclear Energy Agency Expert Group on Burnup Credit Criticality Safety. The function clearly displayed that burnup importance of the end regions increased (a) as burnup, (b) as cooling time, (c) in consideration of burnup profile, and (d) in consideration of fission products. Comparison of the burnup importance for different initial enrichments was also shown.

Effects of Organic Solvent on Infinite Neutron Multiplication Factor of Homogeneous Plutonium Nitrate Solution System

The validity of replacing organic solvent with water in making models for criticality safety evaluation analysis of fuel solution systems was examined by the calculation of the infinite neutron multiplication factor(k inf) for the homogeneous plutonium(IV) nitrate in aqueous and organic solvent solutions. The atomic number densities of fuel solutions were determined by the density formula based on measurements. It was confirmed that k inf of plutonium(IV) nitrate-30 vol% tributylphosphate (TBP)-dodecane organic solvent system is nearly equal to that of plutonium nitrate-aqueous solution system in general. However, more strictly, the values of k inf of organic solvent system became slightly larger than the corresponding values of aqueous solution in the Pu concentration ranged up to 70g/l, which gives the opposite result to Weber's for plutonium nitrate30 vol% TBP-kerosene organic system. Consequently, the estimated lower limit critical concentration for 239Pu(NO3)4 organic solvent system was found 1% smaller than the corresponding value for aqueous solution system. It was also found that k inf tends to increase by increasing the concentration of TBP in organic solvent.