Control Rod Worth Research Articles

Neutronic Analysis of the PSBR Using a Burnup-Coupled MCNP Simulation with MURE

The main goal of this study is to verify the accuracy of burnup-coupled neutronic calculations when employing the Monte Carlo Utility for Reactor Evolutions (MURE) and MCNP5 codes for modeling TRIGA-type reactors, in this case the Penn State Breazeale Reactor (PSBR) core. Research and educational requirements mainly direct the PSBR operating schedule. With such operating schedules, one particular area of concern, specifically relating to nuclear analytical applications, is time-dependent changes in the neutronic characteristics of the reactor, specifically within the irradiation positions. Particular concern exists among scientists performing neutron activation analysis measurements as to whether continuous variations in reactor operations would cause significant fluctuations in the neutronic characterization parameters of the irradiation positions. A secondary objective of this study is to analyze fluctuations in the neutronic characterization parameters and their dependence on various core conditions as examined by detailed burnup-coupled neutronic simulations. In this study, a burnup-coupled neutronic simulation model of the PSBR is developed using the MURE and MCNP5 codes. The simulation results are verified by a series of experiments including measurements of the core excess reactivity starting from the first core loading in 1965 to 2012, control rod worth, fission product buildup, temperature-dependent reactivity loss, integral control rod worth curves, individual fuel element worth, and neutron flux. Local neutronic calculations of the simulation are confirmed by measuring neutronic characterization parameters for one of the irradiation positions within the PSBR core, namely, dry irradiation tube 1. Analyzing time-dependent data predicted by the simulation, the neutron temperature and the measure of the nonideal epithermal neutron flux distribution are found to be reasonably static. Conversely, the thermal-to-epithermal neutron flux ratio and spectral index are found to be relatively responsive to alterations in the core.

Improvement of discontinuity factor for strong absorber region

At Institute of Nuclear and New Energy Technology (INET) the discontinuity factor corrected diffusion method with the homogenization technology was developed and applied in the control rod worth calculation of the pebble bed high temperature gas cooled reactor. But the result with the normal procedure is not accurate enough for a strong absorber. The numerical analysis shows that the strong absorber still has great influence on the flux distribution in the nearby graphite region, so that the flux distribution obtained by the normal diffusion method does not agree with the transport result. Thus, two improvements were proposed in this paper. First, instead of the neutron flux in the middle of the fine mesh, the surface flux of the absorber region was calculated through the net current in the boundary of the region; and then, while the discontinuity factor of the homogenized absorber region should be calculated, the discontinuity factor of the neighboring graphite region on the other side of the interface should also be calculated to eliminate the influence of the strong absorber. The numerical results demonstrate that, based on the improved method, the accuracy of heterogeneous transport calculation can be achieved by a diffusion calculation.

Steady State and Accident Transient Analysis Burning Weapons-Grade Plutonium in Thorium and Uranium with Silicon Carbide Cladding

A comparison of burning weapons-grade plutonium in a standard pressurized water reactor (PWR) using thoria or urania as a fuel matrix has been performed. Two cladding options were considered: a silicon carbide (SiC) matrix of 0.76-mm thickness and Zircaloy of 0.57-mm thickness. As expected, in terms of percentage and total plutonium mass burned, there was a large benefit in using thoria as a matrix compared to urania. Additionally, a smaller amount of plutonium is required in a core when SiC is the cladding because of lower neutron absorption in SiC. The thorium system was also better from a plutonium-burning viewpoint. It resulted in less weapons-useable U and Pu at discharge and more burned over an assembly’s lifetime. At discharge, the fuel was shown to have lower multiples of minimum amounts needed for weapons, even when 233U breeding was taken into account. Thoria-plutonia fuel has different kinetic characteristics from urania-plutonia or enriched urania fuel, so a limited safety comparison of such fuels was made for two reactivity insertion accidents: (1) the highest worth rod ejection and (2) main-steam-line break (MSLB). The accident analyses were performed at both beginning and end of cycle. While the control rod worths are higher in the simulated thoria-plutonia and urania-plutonia cores than in conventional urania-loaded cores, the enthalpy added during the accident was lower than current safety limits for conventional cores. During the MSLB accident, all cases showed acceptable behavior, indicating that the less negative moderator temperature coefficients of thoria-plutonia and urania-plutonia fuel were not limiting.

Assembly Design of Pressurized Water Reactors with Fully Ceramic Microencapsulated Fuel

Fully ceramic microencapsulated (FCM) fuel consists of TRISO (tristructural-isotropic) fuel particles embedded in a ceramic matrix (SiC) to form fuel pellets and rods and offers improved fission product retention and lower operating temperature with expected superior performance in normal and off-normal conditions compared to conventional fuel. When coupled with SiC cladding, FCM fuel eliminates zirconium altogether and is expected to drastically reduce hydrogen generation during a beyond-design-basis accident. In order to be deployed in current or future pressurized water reactors (PWRs), FCM fuel must meet or exceed the neutronic performance of conventional fuel. Limited by low heavy metal loading, an FCM fuel assembly requires increased enrichment and large fuel rods to match the cycle length of a conventional fuel assembly.This study investigated the core design, neutronics, and thermal hydraulics of a PWR loaded with FCM fuel and sought to optimize the assembly design to minimize the enrichment required to reach fuel performance similar to that of conventional fuel. It was found that the implementation of FCM fuel in a 17 × 17 assembly requires close to 20% enrichment and large fuel rods. Such design performs comparably to conventional fuel (4.5% enrichment) in terms of cycle length, reactivity coefficients, intra-assembly power peaking factor, burnable poison penalty, and control rod worth but requires an increase of pumping power. A parametric analysis spanned a large design space varying fuel outer diameter and pitch-to-diameter ratio (P/D) and downselected two alternate assembly designs: 11 × 11 (1.65-cm outer diameter and 1.18 P/D) and 9 × 9 (2.12-cm outer diameter and 1.12 P/D). These designs meet the cycle length requirement with 18.6% and 16.2% enrichments, respectively, but feature a smaller minimum departure from nucleate boiling ratio (MDNBR) compared to a reference assembly. It was estimated that a slight increase in rod outer diameter increases MDNBR to the desired level and implies a pressure drop increase of 10%.

Investigation of different burnable absorbers effects on the neutronic characteristics of PWR assembly

There is an urgent need to find compact burnable BA materials as an alternative to the traditional materials (B4C and Ag–In–Cd). These materials must possess higher density, good thermophysical properties, radiation resistance, and swelling resistance. In this article, the most interesting rare earth nuclides for neutron absorption (gadolinium, erbium, dysprosium, hafanium and dysprosium) are studied and compared from the point of view of their possibilities for the control of core reactivity in PWRs to increase the cycle length. MCNPX code is used to design a three dimensional model for PWR assembly. The PWR assembly is loaded in different operation process with 2% Gd2O3 and 2% Er2O3 as integral burnable absorbers (IBAs). The effects of gadolinia and erbia on the thermal neutron flux and normalized power are investigated. The study presents that erbium is more effective IBA than gadolinium for a reactor with a very long cycle. The dysprosium titanate, dysprosium hafnate and hafnium diboride are investigated as material for a control rod in PWR assembly. The effects of these control rod materials on the multiplication factor and the concentrations of fission products are calculated. From the viewpoint of the achievable fuel discharge burnup, it was found that the suggested control rod materials achieve a satisfactory control rod worth (CRW), i.e., the CRW of the presented materials is comparable with the CRW of the standard AIC control rods.

Bias factor method using random sampling technique

ABSTRACTToward the practical use of the bias factor method for actual light water reactor core analyses, the bias factor method using the random sampling technique is newly proposed. The bias factor method is one of the correction methods using information of E/C values in existing measurable systems, to reduce biases and uncertainties of predicted core characteristics parameters. By the aid of the random sampling technique, our proposed bias factor method can be carried out using only forward calculations without any adjoint calculations, and can easily take into account burnup and thermal-hydraulic feedback effects, which are difficult points in the practical application to actual core analyses. Although the statistical error due to the random sampling technique is inevitable in the proposed method, the statistical error can be simply quantified by the resampling technique such as the bootstrap method. As one of the feasibility studies, effectiveness of the proposed method is verified through a numerical experiment which virtually simulates a typical equilibrium pressurized water reactor core. In this verification problem, it is clarified that E/C values of control rod worth at the beginning of cycle under the hot zero power condition are useful information to reduce biases and uncertainties of predicted assembly-wise power distributions during operation of hot full power.

Characterization of control rod worths and fuel rod power peaking factors in the university of Utah TRIGA Mark I reactor

Control rod reactivity (worths) for the three control rods and fuel rod power peaking factors in the University of Utah research reactor (100 kW TRIGA Mark I) are characterized using the AGENT code system and the results described in this paper. These values are compared to the MCNP6 and existing experimental measurements. In addition, the eigenvalue, neutron spatial flux distributions and reaction rates are analyzed and discussed. The AGENT code system is widely benchmarked for various reactor types and complexities in their geometric arrangements of the assemblies and reactor core material distributions. Thus, it is used as a base methodology to evaluate neutronics variables of the research reactor at the University of Utah. With its much shorter computation time than MCNP6, AGENT provides agreement with the MCNP6 within a 0.5 % difference for the eigenvalue and a maximum difference of 10% in the power peaking factor values. Differential and integral control rod worths obtained by AGENT show well agreement with MCNP6 and the theoretical model. However, regulating the control rod worth is somewhat overestimated by both MCNP6 and AGENT models when compared to the experimental/theoretical values. In comparison to MCNP6, the total control rod worths and shutdown margin obtained with AGENT show better agreement to the experimental values.

New burnable absorber for long-cycle low boron operation of PWRs

This paper presents a new high performance burnable absorber (BA) design for advanced Pressurized Water Reactors (PWRs) aiming for a long-cycle operation with a low soluble boron concentration. The new BA consists of a UO2–157Gd2O3 rod covered with a thin layer of Zr167Er2. A key feature of this new BA is that enriched isotopes, 157Gd and 167Er, are used as absorber materials. Since the high absorption cross section of 157Gd can reduce the mass fraction of Gd2O3 in UO2–Gd2O3, the thermal margin of fuel rods will increase with higher heat conductivity. Also, the 157Gd transmutes into 158Gd by neutron absorption and therefore the residual penalty at the end of cycle (EOC) will decrease. Since 167Er has a resonance near the thermal neutron energy region, the moderator temperature coefficient (MTC) will become more negative and the control rod worth will increase. These advantages of the new BA are demonstrated with three verification cases: a 17×17 Westinghouse (WH) type fuel assembly, a 16×16 Combustion Engineering (CE) type fuel assembly, and an OPR-1000 equilibrium core.

Analysis of fuel subassembly innerduct configurational effects on the core characteristics and power distribution of a sodium-cooled fast breeder reactor

The power distribution and core characteristics in various configurations of fuel subassemblies with an innerduct structure in the Japan sodium-cooled fast reactor were evaluated using a Monte Carlo code for neutron transport and burnup calculation. The correlation between the fraction of fuel subassemblies facing outward and the degree of power increase at the core center was observed regardless of the compositions. This indicated that the spatial fissile distribution caused by innerduct configurations was the major factor of the difference in the power distribution. A power increase was also found in an off-center region, and it tended to be greater than that at the core center because of the steep gradient of neutron flux intensity. The differences in the worth of control rods caused by the innerduct configurations were confirmed.

Monte Carlo modeling of the Pakistan Research Reactor-1 (PARR-1)

Pakistan Research Reactor-1 (PARR-1) is a swimming pool type research reactor established in 1965. Originally, it was designed to operate at 5MW utilizing 93% Highly Enriched Uranium (HEU) fuel. Due to proliferation concerns, the reactor core was converted to LEU core in 1992, utilizing Uranium Silicide (U3Si2-Al) fuel having 235U enrichment of 19.99% and with an up graded operating power of 9MW. The first LEU core was operated at full reactor power (9MW) composed of only fresh fuel and is termed as first high power core. The objective of this article is to develop the detailed model of this first high power core of PARR-1 employing Monte Carlo N-Particle (MCNP5) computer code. The MCNP5 is equipped with the cross-section library ENDF B-VI. To validate the MCNP model, the simulated values of initial criticality with control rod positions, excess reactivity, shutdown margin and combined control rod worth are compared with the published work. The relative error in criticality is 2.6% (with control rod positions 51.6% out of the core), 2.7% in excess reactivity, 6% in the shutdown margin and 3.3% for combined control rod worth has been observed. This model shall be applied to predict the power profile, peaking factors and feedback reactivity coefficients of the reactor in future.

Burnable absorber-integrated Guide Thimble (BigT) – I: design concepts and neutronic characterization on the fuel assembly benchmarks

This paper presents the conceptual designs of a new burnable absorber (BA) for the pressurized water reactor (PWR), which is named ‘Burnable absorber-integrated Guide Thimble’ (BigT). The BigT integrates BA materials into standard guide thimble in a PWR fuel assembly. Neutronic sensitivities and practical design considerations of the BigT concept are points of highlight in the first half of the paper. Specifically, the BigT concepts are characterized in view of its BA material and spatial self-shielding variations. In addition, the BigT replaceability requirement, bottom-end design specifications and thermal–hydraulic considerations are also deliberated. Meanwhile, much of the second half of the paper is devoted to demonstrate practical viability of the BigT absorbers via comparative evaluations against the conventional BA technologies in representative 17×17 and 16×16 fuel assembly lattices. For the 17×17 lattice evaluations, all three BigT variants are benchmarked against Westinghouse's existing BA technologies, while in the 16×16 assembly analyses, the BigT designs are compared against traditional integral gadolinia-urania rod design. All analyses clearly show that the BigT absorbers perform as well as the commercial BA technologies in terms of reactivity and power peaking management. In addition, it has been shown that sufficiently high control rod worth can be obtained with the BigT absorbers in place. All neutronic simulations were completed using the Monte Carlo Serpent code with ENDF/B-VII.0 library.

Directional diffusion coefficients and leakage-corrected discontinuity factors: Implementation in Serpent and tests

A recently proposed method has been extended to treat the generation of radial directional diffusion coefficients with the Monte Carlo code Serpent. In addition to a previous implementation for the axial direction, the new method was applied to the study of a sodium-cooled fast reactor by combined Monte Carlo and diffusion theory. The performance of anisotropic diffusion coefficients was found to be superior to that one of standard diffusion coefficients already available in Serpent. A diffusion solver was also implemented in Serpent in order to compute leakage-corrected discontinuity factors in hexagonal-Z geometry for multiplicative and non-multiplicative multi-group problems. The use of face discontinuity factors improved the agreement among results achieved by diffusion theory and by full-core Monte Carlo calculations. The performance of anisotropic diffusion and discontinuity factors is compared in terms of sensitivity to the micro-energy group structure used in Serpent, core eigenvalue, power distribution, sodium void reactivity worth and control rod worth. Anisotropic diffusion coefficients furnished a better overall agreement. Means of improving the performance of both modeling approaches are proposed.

TRIPOLI-4 Simulation of Core Physics Parameters for Two 3600-MW(thermal) Sodium-Cooled Fast Reactors

To support the development of Sodium-cooled Fast Reactors (SFRs) of Generation IV nuclear energy systems and to study the use of the TRIPOLI-4® Monte Carlo code and the JEFF-3.1.1 nuclear data library on the core neutronics of large fast neutron reactors, in this work two recent Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) computational benchmarks of two 3600-MW(thermal) SFRs were analyzed with the continuous-energy TRIPOLI-4 code. Both a mixed oxide [(U,Pu)O2] core and a carbide [(U,Pu)C] core were investigated. Under two different fast neutron spectra, the reactor physics parameters—Keff, βeff (effective delayed neutron fraction), sodium void worth, Doppler constant, control rod worth, and core power distribution—were calculated for the beginning of equilibrium cycle condition. Both the pin-by-pin heterogeneous and fuel assembly–level homogeneous calculation models were applied in the whole-core simulation in order to evaluate their impact on the calculation results of SFR reactor physics parameters. The ENDF/B-VII.0 data library from the evaluation was also used with TRIPOLI-4 to study its impact on the SFR core reactivity and the boron carbide control rod worth. Using the mesh tally option, the energy deposition tally, and the upgraded display tool of TRIPOLI-4, radial power distribution and core power maps of the two cores were calculated and compared.

Heavy Water Zero Power Reactor (HWZPR) mixed core first criticality, calculation and experiments

HWZPR original fuel is natural U metal fuel but other kinds of fuel can also be utilized. In a research work on UO2 fuel, reactor fuel was partially replaced by natural UO2 fuel and physical parameters of the new core were compared with original core. Thirty six natural U metal fuel rods were substituted by natural UO2 fuel assemblies. Prior to the first criticality operation with the new core, it was simulated by stochastic and deterministic calculation methods i.e. MCNP-4C and WIMS-CITATON codes, respectively. In order to investigate criticality and safety of the mixed core, important reactor physics parameters such as effective multiplication factor (Keff) at different water levels, critical water level, reactivity worth of D2O and reactivity worth of safety and control rods were calculated.The calculated results ensured reactor criticality and satisfied reactor safety criteria. Therefore, with the permission of the reactor safety committee, the first criticality operation was performed successfully. Later, during a series of reactor operations important physical parameters were measured experimentally. There is good consistency between the theoretical and experimental results.

An improved assembly homogenization approach for plate-type research reactor

We have developed an improved assembly homogenization approach for plate-type research reactor. Compared with the conventional cell and assembly homogenization way, three advanced homogenization methods are used in our improved approach, including the direct single and multi-assembly calculations, treatment of the assembly environmental effect by multi-assembly calculations and assembly classification, and the superhomogenization (SPH) method. The approach is carried out with DRAGON v4 code. The neutronic analysis for fresh fueled core of JRR-3M with UAlx–Al dispersion type fuel is carried out by diffusion code CITATION with 3-group homogenized cross sections which are generated by the improved approach. Results by Monte Carlo code are taken as references, and results by the conventional three-step method are given for comparison purposes. The calculation results show that the improved assembly homogenization approach is effective to improve the accuracy of keff eigenvalue, control rod worth and neutron flux distributions for whole core calculations. Especially, the control rod worth can be calculated accurately by the improved approach.

Analysis of safety parameters for the MINERVE reactor

Abstract The present study aims to show the effect of different cross section libraries on MINERVE reactor safety parameters. The MCNP5 calculation model of the MINERVE reactor facility is used to determine core and safety parameters such as axial and radial fission rate distributions, control rod worth and spectral indices. Different neutron spectra were achieved by changing the experimental lattice within the MINERVE reactor. MINERVE provides a large experimental basis for the improvement of the cross section databases. The current study calculates these parameters by MCNP5 code using three different cross section libraries (the continuous energy cross sections of the ENDFB-VI, T16-2003, and ENDF/B-VII.1 libraries). Neutronic calculations were performed for R1UO2, R1MOX, R2UO2, and R2MOX core configurations, representative of a LWR loaded with UO2, mixed oxide matrix, over moderated UO2 and over moderated MOX, respectively. The study aims to determine the most suitable cross section library to be used with MCNP5 reactor calculation code for light water reactor fuel lattice. The MCNP5 results were compared with the experimental results.

Neutronic analysis for core conversion (HEU–LEU) of the low power research reactor using the MCNP4C code

Abstract Comparative studies for conversion of the fuel from HEU to LEU in the miniature neutron source reactor (MNSR) have been performed using the MCNP4C code. The HEU fuel (UAl4-Al, 90% enriched with Al clad) and LEU (UO2 12.6% enriched with zircaloy-4 alloy clad) cores have been analyzed in this study. The existing HEU core of MNSR was analyzed to validate the neutronic model of reactor, while the LEU core was studied to prove the possibility of fuel conversion of the existing HEU core. The proposed LEU core contained the same number of fuel pins as the HEU core. All other structure materials and dimensions of HEU and LEU cores were the same except the increase in the radius of control rod material from 0.195 to 0.205 cm and keeping the outer diameter of the control rod unchanged in the LEU core. The effective multiplication factor (keff ), excess reactivity (ρ ex ), control rod worth (CRW), shutdown margin (SDM), safety reactivity factor (SRF), delayed neutron fraction (β eff ) and the neutron fluxes in the irradiation tubes for the existing and the potential LEU fuel were investigated. The results showed that the safety parameters and the neutron fluxes in the irradiation tubes of the LEU fuels were in good agreements with the HEU results. Therefore, the LEU fuel was validated to be a suitable choice for fuel conversion of the MNSR in the future.

Adjustment method of deterministic control rods worth computation based on measurements and auxiliary Monte Carlo runs

The control rods (CRs) worth is key parameter for the research reactors (RRs) operation and utilization. Control rods worth computation is a challenge for the full deterministic calculation methodology, including the few group cross section generation, and the core analysis. The purpose of this work is to interpret our codes system, and their applicability of obtaining reliable CRs worth by some engineering adjustments. Cross sections collapsing in three energy groups is made by WIMS and SN2 codes, while the core analysis is performed by CITATION. We use these codes for the design, construction, and operation of our research reactor CMRR (China Mianyang Research Reactor). However, due to the intrinsic deficiency of the diffusion theory and homogenizing approximation, the directly obtained results, such as CRs worth and neutron flux distributions are not satisfactory. So two points of simple adjustments are made to generate the few group cross-sections with the assistance of measurements and auxiliary Monte Carlo runs. The first step is to adjust the fuel cross sections by changing properly the mass of a non-fissile material, such as the mass of the two 0.4mm Cd wires existing at both sides of each uranium plate, so that the core model of CITATION can get good eigenvalue when all CRs are completely extracted. The second step is to revise the shim absorber cross section of CRs by adjusting the hafnium mass, so that the CITATION model can get correct critical rods position. In this manuscript, the JRR-3M (Japan Research Reactor No. 3 Modified) reactor is employed as a demonstration. Final revised results are validated with the stochastic simulation and experimental measurement values, including the critical rods position, the differential/integral CR curves, and the thermal neutron flux distributions.

Full Core modeling techniques for research reactors with irregular geometries using Serpent and PARCS applied to the CROCUS reactor

This paper summarizes the results of modeling methodologies developed for the zero-power (100W) teaching and research reactor CROCUS located in the Laboratory for Reactor Physics and Systems Behavior (LRS) at the Swiss Federal Institute of Technology in Lausanne (EPFL). The study gives evidence that the Monte Carlo code Serpent can be used effectively as a lattice physics tool for small reactors. CROCUS’ core has an irregular geometry with two fuel zones of different lattice pitches. This and the reactor’s small size necessitate the use of nonstandard cross-section homogenization techniques when modeling the full core with a 3D nodal diffusion code (e.g. PARCS). The primary goal of this work is the development of these techniques for steady-state neutronics and future transient neutronics analyses of not only CROCUS, but research reactors in general. In addition, the modeling methods can provide useful insight for analyzing small modular reactor concepts based on light water technology. Static computational models of CROCUS with the codes Serpent and MCNP5 are presented and methodologies are analyzed for using Serpent and SerpentXS to prepare macroscopic homogenized group cross-sections for a pin-by-pin model of CROCUS with PARCS. The most accurate homogenization scheme lead to a difference in terms of keff of 385pcm between the Serpent and PARCS model, while the MCNP5 and Serpent models differed in terms of keff by 13pcm (within the statistical error of each simulation). Comparisons of the axial power profiles between the Serpent model as a reference and a set of PARCS models using different homogenization techniques showed a consistent root-mean-square deviation of ∼8%, indicating that the differences are not due to the homogenization technique but rather arise from the definition of the diffusion coefficients produced by Serpent. A comparison of the radial power profiles between the best PARCS model and full-core Serpent model showed largest relative differences in terms of power prediction at the core periphery, which is believed to be the product of the geometry simplifications made, the diffusion coefficients produced by Serpent, and the two-group energy structure used. The worth of a single control rod reproduced in PARCS showed a difference of −33pcm from its 169pcm worth simulated in Serpent.

Solution of the BEAVRS benchmark using the nTRACER direct whole core calculation code

The BEAVRS (Benchmark for Evaluation and Validation of Reactor Simulation) benchmark is solved by the nTRACER direct whole core calculation code to assess its accuracy and to examine the solution dependence on modeling parameters. A sophisticated nTRACER core model representing the BEAVRS core is prepared after a series of sensitivity study to ensure solution accuracy. The resulting solutions for several hot-zero-power (HZP) states are compared first with the corresponding Monte Carlo solutions, which consist of the McCARD solutions for the assembly problems and the OpenMC solutions for the core problems, and then with the measured data which include the control rod worths (CRWs) and incore detector signals as well as the critical boron concentrations (CBC). The core depletion calculation is performed for the initial and second cycles with a set of approximated power histories and the calculated CBCs are compared with the measured data. The comparison results show that the criticality, control rod bank worths at HZP and the boron let-down curves of two cycles agree well with the measurements within 180 pcm and 25 ppm, respectively.