Reactivity Worth Research Articles

Validation of a fuel management code MCODE-FM against fission product poisoning and flux wire measurements of the MIT reactor

The Massachusetts Institute of Technology Research Reactor (MITR), with thermal power of 6 MW, is one of the five high-performance research reactors operated in the U.S. An in-house program MCODE-FM (MCNP-ORIGEN Coupled Depletion Program – Fuel Management) as the reference code supporting routine operation is currently developed by the MITR staff. Adopting Monte Carlo methods enables exact modeling of the MITR core geometry with use of continuous-energy nuclear data. A criticality search algorithm to track control blade movement is implemented in MCODE-FM. The code also features automation of input file generation, data manipulation, and post-processing of output data for fuel cycle analysis. Some verification and validation runs have been carried out by comparisons with the multiplication factor and the reactivity worth induced by control blades. In this study, further validation runs are performed, with two sets of measured data: 1) reactivity effects of fission product poisoning and fuel depletion and 2) reaction rate based thermal and fast neutron flux. Fission product poisoning effects calculated by MCODE-FM are in good agreement with the measured data obtained during the reactor start-up process (within 10%). The trend of peak xenon reactivity after the reactor shutdown is predicted accurately as well. As compared to the neutron activation experiment, MCODE-FM is found to slightly over-estimate the thermal flux with less than 10% discrepancy and the fast neutron flux with less than 20% discrepancy.

Core Design, Spent-Fuel Characteristics Assessment, and Fuel Cycle Analysis for Thorium-Uranium Breeding Recycle in Pressurized Water Reactors

This work is focused on core design, spent-fuel characteristics assessment, and fuel cycle analysis for thorium-uranium breeding recycle in a typical pressurized water reactor (PWR), without any major change to the fuel lattice and the core internals but substituting the uranium oxide (UOX) pellet with a thorium-based fuel pellet. Two mixed cores are investigated, one loaded with mixed reactor-grade plutonium-thorium oxide (PuThOX) fuel assemblies and the other with mixed reactor-grade 233U-thorium oxide (U3ThOX) fuel assemblies. The high purity of reactor-grade 233U extracted from burnt PuThOX fuel is used as seeds of U3ThOX for starting thorium-uranium breeding recycle.The core design and analysis indicated that thorium-uranium breeding recycle is technically feasible in current PWRs. In the mixed core with U3ThOX loading, the well-designed U3ThOX assemblies were located on the periphery of the core as a “blanket” region, which remain in core for six cycles and get breeding with 232Th-233U. The feedback parameters and kinetic parameters are dominated by the UOX fuel in the inner core. For the UOX/PuThOX mixed core, the higher plutonium content leads to harder neutron spectrum, smaller reactivity worth of neutron absorbers, and smaller delayed neutron fraction and prompt neutron lifetime, which are similar to the current mixed cores partially loaded with the plutonium-uranium mixed-oxide (MOX) fuel.The fuel cycle analysis has shown that 233U monorecycling with U3ThOX fuel could save 13% of natural uranium resource compared with UOX once-through fuel cycle, slightly more than that of plutonium monorecycling with MOX fuel. If 233U multirecycling with U3ThOX fuel is implemented, more natural uranium resource would be saved.

Installation of permanent cadmium-lined channel as a means for increasing epithermal NAA capabilities of miniature neutron source reactors

High demand for epithermal neutrons by the clients of the Nigerian Research Reactor-1 (NIRR-1), a Miniature Neutron Source Reactor (MNSR) has necessitated the need to explore avenues for increasing epithermal Neutron Activation Analysis (NAA) capabilities of the reactor. Safety and flux stability simulations were done by our group using Monte Carlo Transport Code MCNP5 for permanent cadmium line inside the irradiation channel of NIRR-1 and compared with the ones reported by other MNSR groups. The results of all these simulations revealed that the effect of cadmium-line on safety and flux stability is very minimal in the outer channel than in the inner channel. We have reported here the design specifications, preliminary studies done, steps followed in installation and measurements done in the pre and post installation of the permanent cadmium-line in outer channel of the reactor. We measured the reactivity worth of the old and new channel and readjusted the reactor's core excess reactivity after the installation. Results obtained are: reactivity worth of the old channel (0.12mk), reactivity worth of the new channel=0.336mk, temperature coefficient=−0.1mk/°C, control rod worth coefficient=0.023mk/mm and the core excess reactivity=3.85mk. We have also measured the radial and axial flux distribution in the channels of the reactor after the installation. The installation of the permanent cadmium-lined channel reported here will not only boost the sample handling capabilities of NIRR-1 but will also provide useful data to the MNSR community for upgrading their reactors.

Accuracy of thorium cross section of JENDL-4.0 library in thorium based fuel core evaluation

Considering the importance of thorium data and concerning about the accuracy of Th232 cross section library, a series of experiments on thorium critical core with different neutron spectra has been implemented at Kyoto University Critical Assembly (KUCA). Reactivity worth of control rod and thorium sample was measured after the cores experimentally achieved critical state. In order to verify the accuracy of thorium cross section library, calculations of effective multiplication factor, control rod worth, reactivity worth of Th plates for the same core configurations were done by MVP code (Nagaya et al., 2005) using JENDL-4.0 library (Shibata et al., 2011). From the comparison between the measured and calculated results, the calculations are found to overestimate effective multiplication factor about (0.90±0.01–0.99±0.01)%. By comparing the measured Th sample worth with the calculated one, Th capture underestimation is reassured. Sensitive study on reactivity worth evaluation was conducted and it suggests that Th capture cross section is needed to be adjusted at thermal energy range to provide more reliable evaluation for thorium based fuel core design and safety calculation.

Development of a control rod depletion methodology for the Westinghouse NEXUS system

The NEXUS project is an effort to merge and modernize the methods employed in Westinghouse PWR and BWR steady-state reactor physics codes. The NEXUS system relies on a once-through nodal cross-section generation methodology with an innovative and efficient technique for pin power recovery. The pin power methodology overcomes a well-known limitation of existing methodologies, namely the incapacity to properly account for heterogeneity changes due to the depletion environment. The so-called control rod history problem where control rods are repeatedly inserted and withdrawn during core depletion is a good example of such a case. In addition to the control rod history impact on pin power distributions, the insertion of control rods during extended periods leads to significant control rod depletion that affects the reactivity worth of the control rods which in turn can have a significant impact on pin powers. The importance of accurately predicting pin powers, combined with the need to adequately estimate the reactivity worth and nuclear end of life of control rods in BWRs and in generation III+ PWRs, has motivated the development of a novel control rod depletion model. This methodology and its numerical qualification, initially for PWR application only, is the topic of this paper. The focus is on describing the salient features of the model and on illustrating its performance by means of numerical experiments. It is shown that together with the NEXUS pin power recovery model, the control rod depletion methodology accurately predicts the reactivity feedback from repeated control rod insertions in a PWR core.

Assessing the Effect of Fuel Burnup on Control Rod Worth for HEU and LEU Cores of Gharr-1

An important parameter in the design and analysis of a nuclear reactor is the reactivity worth of the control rod which is a measure of the efficiency of the control rod to absorb excess reactivity. During reactor operation, the control rod worth is affected by factors such as the fuel burnup, Xenon concentration, Samarium concentration and the position of the control rod in the core. This study investigates the effect of fuel burnup on the control rod worth by comparing results of a fresh and an irradiated core of Ghana's Miniature Neutron Source Reactor for both HEU and LEU cores. In this study, two codes have been utilized namely BURNPRO for fuel burnup calculation and MCNP5 which uses densities of actinides of the irradiated fuel obtained from BURNPRO. Results showed a decrease of the control rod worth with burnup for the LEU while rod worth increased with burnup for the HEU core. The average thermal flux in both inner and outer irradiation sites also decreased significantly with burnup for both cores.

3D core burnup studies in 500 MWe Indian prototype fast breeder reactor to attain enhanced core burnup

Fast breeder reactors are capable of producing high fuel burnup because of higher internal breeding of fissile material and lesser parasitic capture of neutrons in the core. As these reactors need high fissile enrichment, high fuel burnup is desirable to be cost effective and to reduce the load on fuel reprocessing and fabrication plants. A pool type, liquid sodium cooled, mixed (Pu–U) oxide fueled 500MWe prototype fast breeder reactor (PFBR), under construction at Kalpakkam is designed for a peak burnup of 100GWd/t. This limitation on burnup is purely due to metallurgical properties of structural materials like clad and hexcan to withstand high neutron fluence, and not by the limitation on the excess reactivity available in the core. The 3D core burnup studies performed earlier for approach to equilibrium core of PFBR is continued to demonstrate the burnup potential of existing PFBR core. To increase the fuel burnup of PFBR, plutonium oxide enrichment is increased from 20.7%/27.7% to 22.1%/29.4% of core-1/core-2 which resulted in cycle length increase from 180 to 250 effective full power days (efpd), so that the peak fuel burnup increases from 100 to 134GWd/t, keeping all the core parameters under allowed safety limits. Number of diluents subassemblies is increased from eight to twelve at beginning of life core to bring down the initial core excess reactivity. PFBR refueling is revised to accommodate twelve diluents. Increase of 10B enrichment in control safety rods (CSRs) and diverse safety rods (DSRs) is made along with a design change in CSRs to increase the reactivity worth of absorber rods for ensuring safe reactor operation. This new enhanced burnup core can be used in the future with advanced structural materials for clad and hexcan, which can withstand higher neutron fluence and gamma radiation.

Fast breeder reactor core concept for heterogeneous minor actinide loading

As an alternative method to the homogeneous minor actinide (MA) recycling in fast breeder reactors, a heterogeneous MA loading core concept using a highly concentrated americium (Am)-containing fuel (Am target) is proposed. By the use of an extraction process for Am and curium (Cm) in the reprocessing of the spent fuel, Am (and a small amount of Cm) can be recovered and then concentrated to produce the target. The Am content in the heavy metal is assumed to range from 10 to 20 wt% in accordance with the target development scope. A mixed oxide fuel that contains uranium, plutonium, and neptunium is chosen as the base material of the target, so that the targets can generate a level of power equivalent to that of the driver fuels. It was found that a ring-shaped arrangement of Am targets between the inner and outer core regions exhibits a favorable MA transmutation performance without any significant deterioration in the core neutronic characteristics, including increases of the burnup reactivity and sodium void reactivity worth, and decreases of the breeding ratio and absolute value of the Doppler coefficient, etc., in comparison with those of a reference homogeneous MA loading case. It should be noted that the Am targets in this loading arrangement can contribute to the suppression of the core power distribution change along with burnup. A series of core designs, including core neutronics, thermal hydraulics, and fuel integrity evaluations, was also carried out for a representative Am target loading case. The results indicate that it is possible to design an Am target subassembly that can cope with the issues presented by highly concentrated Am, i.e., the deterioration of thermophysical properties and the accumulation of helium gas inside the target fuel pins. Therefore, the design feasibility of the heterogeneous target loading core has been enhanced.

Needs of reliable nuclear data and covariance matrices for Burnup Credit in JEFF-3 library

Burnup Credit (BUC) is the concept which consists in taking into account credit for the reduction of nuclear spent fuel reactivity due to its burnup. In the case of PWR-MOx spent fuel, studies pointed out that the contribution of the 15 most absorbing, stable and non-volatile fission products selected to the credit is as important as the one of the actinides. In order to get a “best estimate” value of the keff , biases of their inventory calculation and individual reactivity worth should be considered in criticality safety studies. This paper enhances the most penalizing bias towards criticality and highlights possible improvements of nuclear data for the 15 FPs of PWRMOx BUC. Concerning the fuel inventory, trends in function of the burnup can be derived from experimental validation of the DARWIN-2.3 package (using the JEFF-3.1.1/SHEM library). Thanks to the BUC oscillation programme of separated FPs in the MINERVE reactor and fully validated scheme PIMS, calculation over experiment ratios can be accurately transposed to tendencies on the FPs integral cross sections.

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.

Moderator temperature effects on reactivity of HEU core of MNSR

In this article we report on analyses that were performed to investigate the influence of cross section differences among libraries released by various centers on reactivity of Miniature Neutron Source Reactors. The 3D model of the core was developed with WIMS-D and PRIDE codes and six cross section libraries were used including JENDL-3.2, JEF-2.2, JEFF-3.3, ENDF/B-VI and ENDF/B-VII, and IAEA library. It was observed that all the libraries predict the reactivity within 10%, with IAEA library giving minimum reactivity worth, and JEF-2.2 data library resulted in highest worth.

Nuclear Data and the Effect of Gadolinium in the Moderator

Recent cross-section measurements on gadolinium have raised concerns over the accuracy of moderator poison reactivity coefficient calculations. Measurements have been made at the ZED-2 (Zero Energy Deuterium) critical facility, Chalk River Laboratories, AECL, to study the reactivity effect of gadolinium in the moderator. Since the neutron capture cross-section of boron is well known, measurements were also made with boron to provide calibration data for measurements with gadolinium. The measurements have been used to quantify the bias of the reactivity effect in full-core simulations of ZED-2 using MCNP, a neutron transport code used extensively for simulations of nuclear systems, along with the ENDF/B-VII.0 cross-section data. The results showed a bias of -0.41 ± 0.07 mk/ ppm, or -2.1% ± 0.3%, given a reactivity worth of -20.1 mk/ppm for gadolinium. Additional simulations also show that the gadolinium neutron capture cross-section has been over-corrected, relative to previous evaluations, in a beta version of ENDF/B VII.1, which incorporates the Leinweber data.

Analysis of the MOX critical experiments BASALA by pin-by-pin core analysis method using three-dimensional direct response matrix

An analysis of the MOX critical experiments BASALA was performed to verify the pin-by-pin core analysis method using a three-dimensional direct response matrix. The BASALA experiments simulate full MOX BWR cores, and they were carried out in the EOLE critical facility of the French Atomic Energy Commission (CEA) by the Nuclear Power Engineering Corporation (NUPEC) in collaboration with CEA. The BASALA experimental cores are very heterogeneous because their size is much smaller than that of commercial power plants. The main features of the pin-by-pin core analysis method using the three-dimensional direct response matrix are that the response matrix can reflect the intra-assembly heterogeneous effect, the diffusion approximation is not involved, and the fuel rod fission rate can be directly evaluated. The maximum difference of the critical k-effective values among all nine cores analyzed was about 0.4% Δk. The root mean square differences between the calculated and measured radial fuel rod fission rate distributions in the test assembly of all cores were within 1.8% and nearly comparable to measurement error. The calculated results of the reactivity worth agreed with the measured results within 9%. These good agreements mean that the pin-by-pin core analysis method using the three-dimensional direct response matrix accurately reflects the effects of the intra- and inter-assembly heterogeneities in heterogeneous systems like the BASALA experimental cores.

A coarse mesh radiation transport method for 2-D hexagonal geometry

In this paper, a whole-core stochastic–deterministic hybrid coarse mesh transport method is extended to 2-D hexagonal geometry. This method may be used to calculate the eigenvalue and explicit pin fission density profile of hexagonal reactor cores. It models the exact detail within complex heterogeneous cores without homogenizing regions or materials, and neither block-level nor core-level asymmetry poses any limitations to the method. It solves eigenvalue problems by first splitting the core into a set of coarse meshes, and then using Monte Carlo methods to create a library of response expansion coefficients, found by expanding the angular current in phase-space distribution using a set of polynomials orthogonal on the angular half-space defined by mesh boundaries. The coarse meshes are coupled by the angular current at their interfaces. A deterministic sweeping procedure is then used to iteratively construct the solution.The method is evaluated using benchmark problems based on a gas-cooled, graphite-moderated high temperature reactor. The method quickly solves problems to any level of detail desired by the user. In this paper, it is used to explicitly calculate the fission density of individual fuel pins and determine the reactivity worth of individual control rods. In every case, results for the core multiplication factor and pin fission density distribution are found within several minutes. Results are highly accurate when compared to direct Monte Carlo reference solutions; errors in the eigenvalue calculations are on the order of 0.02%, and errors in the pin fission density average less than 0.1%.

Interpretation of Fission Product Oscillations in the MINERVE Reactor, from Thermal to Epithermal Spectra

To take into account the reactivity loss in spent fuels, an experimental program was set up in 1993 at CEA-Cadarache, France, oscillating separated fission products (FPs) in the MINERVE reactor. Reactivity worth measurements of small samples allow the extraction of information about nuclear data of the studied isotopes. A fully validated calculation scheme has been implemented to interpret pile-oscillation measurements. Therefore, calculation over experiment ratios can be accurately transposed to trends in the integral capture cross section of the FPs. With the European JEFF3.1.1 library, results show a discrepancy below 3% for several nuclides: 155Gd, 149, 152Sm, 143Nd, and 95Mo, but improvements may be needed for some others: 133Cs, 103Rh, 99Ru, and 153Eu. Based on the Integral Data Assimilation technique, we propose new thermal cross-section values, (348 ± 14) b and (42 478 ± 1793) b, for two of the most absorbing nuclides, 143Nd and 149Sm, respectively.

Feasibility of Reduced Boron Concentration Operation in Pressurized Water Reactor Plants

The negative impact of a boron dilution accident on the safety of a current pressurized water reactor (PWR) initiated investigations with the aim of checking the feasibility of reduced boron concentration operation. In addition, reduction of the maximum boron concentration in a PWR is a practical and feasible means to substantially reduce the radiation dose to operators and to minimize corrosion damage. Four types of integral burnable absorbers have been considered: gadolinium, integral fuel burnable absorber (IFBA), erbia, and alumina boron carbide. Under consideration of four different kinds of fuel assemblies (FA), four core design candidates were developed by applying current PWR OPR-1000 technology and by keeping major engineering design constraints and the equivalent fuel enrichment level used in the reference core (REF) design. However, an optimal design was targeted to achieve comparable discharge burnup as well as favorable design safety parameters. The comparative analysis between the REF and the optimal core designs is presented here. One of the designs is suggested as the most promising and favorable low boron core (LBC) design in this framework. The proper combination of axial and radial enrichment zoning patterns plus a mixture of fresh FAs with depleted assemblies in an LBC design candidate with an IFBA-bearing FA at equilibrium cycle could bring a two times narrower axial offset variation than that of the REF design, maintain an acceptable power peaking factor ˜23% lower than the design limit, and achieve higher fuel burnup. It was observed that this optimal LBC design could comply with current OPR-1000 reactor acceptance criteria associated with smooth reactivity swing, more flattened power distribution, and desired limiting safety parameters despite an 18% loss of shutdown reactivity worth at beginning of cycle when compared to the REF design.

Evaluation of Neutron Cross Sections for Hafnium in the Resolved Resonance Range

The international High Priority Request list notes: “In the nuclear industry hafnium is used as neutron absorbing material to regulate the fission process. Interpretation of critical experiments with UOx fuel conducted by CEA in the AZUR zero-power reactors has shown systematic underestimation of the reactivity worth that may be attributed to an overestimated natural hafnium capture cross section in the epi-thermal energy range” To service the request for improved resonance data a PhD project has:a) Improved REFIT R-matrix evaluation code. b) Obtained hafnium oxide samples enriched in Hf176, 177, 178, 179 isotopes. c) Gained support from NUDAME and EUFRAT projects. d) Prepared enriched and natural Hf samples. e) Performed capture and transmission Time of Flight measurements at the GELINA linear accelerator. f) Analysed the capture counts to generate yields using AGS and AGL codes. g) Used REFIT to perform least squares analysis of GELINA measurements. (Included previous ORNL, Harwell and RPI transmissions and capture yields.) h) Submitted results to EXFOR. i) Included resolved resonance parameters in JEFF evaluations taking the resolved range to over 1 keV. j) Tested evaluations with suitable benchmarks. k) Passed resolved resonance data to CEA Cadarache for unresolved analysis. Resultant Hf evaluations will be included in JEFF3.2.

Application of the Modified Source Multiplication (MSM) Technique to Subcritical Reactivity Worth Measurements in Thermal and Fast Reactor Systems

The Amplified Source Multiplication (ASM) method and its improved Modified Source Multiplication (MSM) method have been widely used in the CEA's EOLE and MASURCA critical facilities over the past decades for the determination of reactivity worths by using fission chambers in subcritical configurations. They have been successfully applied to absorber (single or clusters) worth measurement in both thermal and fast spectra, or for (sodium or water) void reactivity worths. The ASM methodology, which is the basic technique to estimate a reactivity worth, uses relatively simple relationships between count rates of efficient miniature fission chambers located in slightly subcritical reference and perturbed configurations. If this method works quite well for small reactivity variation (a few effective delayed neutron fraction), its raw results needs to be corrected to take into account the flux perturbation in the fission chamber. This is performed by applying to the measurement a correction factor called MSM. Its characteristics is to take into account the local space and energy variation of the spectrum in the fission chamber, through standard perturbation theory applied to neutron transport calculation in the perturbed configuration. The proposed paper describes in details both methodologies, with their associated uncertainties. Applications on absorber cluster worth in the MISTRAL-4 full MOX mock-up core and the last core loaded in MASURCA show the importance of the MSM correction on raw data.

Measurement and Analysis of Reactivity Worth of 241Am Sample in Water-Moderated Low-Enriched UO2 Fuel Lattices at TCA

The reactivity worths of 22.82 grams of 241Am oxide sample were measured and theoretically analyzed in water-moderated UO2 fuel lattices in seven cores of the Tank-Type Critical Assembly (TCA) at the Japan Atomic Energy Agency for an integral test of 241Am nuclear data. These cores provided a systematic variation in the neutron spectrum between the thermal and resonance energy regions. The sample reactivity worth was measured with an uncertainty of 2.1% or less. The theoretical analysis was performed using the JENDL-3.3 nuclear data by a Monte Carlo calculation method. Ratios of calculation to experiment (C/Es) of the reactivity worth were between 0.91 and 0.97, and showed no apparent dependence on the neutron spectrum. In addition, sensitivity analysis based on the deterministic calculation method was carried out to obtain the impact of changing the 241Am capture cross section on the sample reactivity worth. The result of this analysis showed that the C/E could be significantly improved by almost uniformly increasing the 241Am capture cross section of JENDL-3.3 by 25–30%.

Investigation of the void reactivity effect in large-size Lead Fast Reactors

An aspect of great relevance in Lead Fast Reactors (LFRs) is the actual void reactivity evaluation. The purpose of this work is double: to inquire into the physical problem, and to evaluate the impact of different approaches and numerical methods on the calculation of the critical reactor parameters involved. Thus, results concerning void effect contributors have been evaluated through a cross-checked analysis performed by means of both a deterministic and a stochastic code. The field of investigation that has been assumed consists in the reference configuration of the 600MWe European Lead-cooled SYstem (ELSY), under development within the 6th and 7th EURATOM Framework Programmes. Calculations have been carried out on a 1500MWth MOX-fuelled core, composed by wrapper-less square Fuel Assemblies (FAs) with pins on a square lattice. The ERANOS (European Reactor Analysis Optimized System) deterministic code ver. 2.1 and the MCNP Monte Carlo code ver. 4c have been employed in conjunction with the JEFF-3.1 nuclear data library to assess the void reactivity variation and its breakdown into the most relevant nuclides, using both the neutron balance equation method and perturbation theory. Results have shown a very good agreement between ERANOS and MCNP outcomes: the huge reactivity worth determined by the core active region voiding (approximately 5000pcm) is due to the predominant contributions of even isotopes – among which 238U plays a major role, being responsible for roughly 4300pcm – as a consequence of their fission cross-section high sensitivity to spectral hardening (threshold reactions), despite their modest contribution to the total fissions.