Reactivity Worth Research Articles

Analysis of Reactivity Measurements in Core Physics Experiments on Full-MOX BWR

A core physics experimental program FUBILA has been performed to study core physics characteristics of full-MOX BWR cores consisting of high Pu-enriched MOX assemblies for high burnups. The program includes the measurement of reactivity worth, which is essential in operating BWR cores. The reactivity worth is due to the reactivity caused by (1) changes in the in-channel void fraction of assemblies, (2) the insertion of a B4C control blade, (3) Gd2O3-UO2 rods and UO2 rods in assemblies, and (4) the mixing of boron in a moderator related to a stand-by liquid control system. The reactivity worth was measured by the modified neutron source multiplication method using the reactivity worth of a pilot rod as a reference worth. The measured reactivity worth was determined by processing count rates of neutron detectors taking into account the detector efficiency and effective neutron source intensity analyzed by three-dimensional transport calculations. Comparing the measured reactivity worth with the results obtained by deterministic calculation methods and a Monte Carlo calculation method, the accuracy of the calculations was evaluated. The calculated results generally well reproduce the measurements, except for the boron reactivity worth in the moderator, for which the calculations overestimate the reactivity worth.

Measurement and Analysis of Reactivity Worth of 237Np Sample in Cores of TCA and FCA

Measurement and Analysis of Reactivity Worth of 237Np Sample in Cores of TCA and FCA

Benchmark Calculation of WIMS/RFSP against Physics Measurement Data of Wolsong Nuclear Power Plants

Benchmark calculations of Canada deuterium uranium (CANDU) reactor physics design and analysis codes have been performed for a lattice code WIMS-AECL, a supercell code DRAGON, and a core analysis code RFSP by using the physics measurement data of Wolsong nuclear power plants. In this study, the lattice and reactivity device models were examined based on Wolsong-2 measurement data for the criticality and reactivity device worth. Sensitivity calculations were also performed for the number of energy groups and the cross-section library. Using the lattice and reactivity device models obtained from the Wolsong-2 calculation, the benchmark calculations were extended to the Wolsong-3 and Wolsong-4 plants. Compared to a previous study, this study showed that the results of the criticality and reactivity device worth calculations were improved when the material data were updated and the exact two-group cross sections were used. For the three nuclear power plants, the calculated core reactivity was within 0.2% Δk of criticality. The zone controller unit reactivity worth was estimated to have a maximum error of ~8%. The total reactivity worth of other reactivity control devices was consistent with the measurement data within 13%. The root-mean-square error of the flux distribution calculation was <12% when compared with flux scans performed during Phase B physics tests. In conclusion, the CANDU physics design and analysis codes used in this benchmark study predicted the physics parameters within the allowed uncertainty level of the measurement data.

Advanced neutronics tools for BWR design calculations

This paper summarizes the developments implemented in the new APOLLO2.8 neutronics tool to meet the required target accuracy in LWR applications, particularly void effects and pin-by-pin power map in BWRs. The Method of Characteristics was developed to allow efficient LWR assembly calculations in 2D-exact heterogeneous geometry; resonant reaction calculation was improved by the optimized SHEM-281 group mesh, which avoids resonance self-shielding approximation below 23 eV, and the new space-dependent method for resonant mixture that accounts for resonance overlapping. Furthermore, a new library CEA2005, processed from JEFF3.1 evaluations involving feedback from Critical Experiments and LWR P.I.E, is used. The specific “2005–2007 BWR Plan” settled to demonstrate the validation/qualification of this neutronics tool is described. Some results from the validation process are presented: the comparison of APOLLO2.8 results to reference Monte Carlo TRIPOLI4 results on specific BWR benchmarks emphasizes the ability of the deterministic tool to calculate BWR assembly multiplication factor within 200 pcm accuracy for void fraction varying from 0 to 100%. The qualification process against the BASALA mock-up experiment stresses APOLLO2.8/CEA2005 performances: pin-by-pin power is always predicted within 2% accuracy, reactivity worth of B4C or Hf cruciform control blade, as well as Gd pins, is predicted within 1.2% accuracy.

Sodium Void Reactivity Worth Calculations for Fast Critical Assemblies without Whole-Lattice Homogenization

In the present paper, we calculate the sodium void reactivity worth of fast critical assemblies without whole-lattice homogenization in order to reduce errors associated with lattice homogenization. Firstly, we solve a neutron transport benchmark problem simulating fast critical assemblies composed of thin material plates with a discrete ordinates transport solver. The discrete ordinates transport solutions agree well with the Monte Carlo reference solutions; hence, we confirm the validity of the deterministic transport calculations for the sodium void reactivity worth of lattice-heterogeneous critical assemblies. Thereafter, the existing experimental data are calculated without whole-lattice homogenization. The result suggests that the lattice homogenization results in the overestimation of the leakage component of sodium void reactivity worth when the leakage component parallel to plate boundaries is dominant. Utilizing the numerical method without whole-lattice homogenization and the nuclear data JENDL-3.3, numerical solutions agree with the experimental data within 3σ of the experimental uncertainties.

Diagnostic analysis of pin-removal reactivity worth experiments in a SVEA-96+ test lattice

The present paper concerns a novel type of integral database generated in the course of the LWR–PROTEUS Phase I experiments, viz. the relative reactivity effects of removing individual fuel pins from a highly heterogeneous BWR assembly. The measurements reported were conducted in the central assembly of a 3 × 3 test-zone configuration of SVEA-96+ assemblies in PROTEUS, under full-density water moderation conditions. Calculations of the pin-removal reactivity worths have been carried out using two different LWR assembly codes, viz. CASMO-4 and BOXER. The discrepancies observed between experiment and calculation have been analysed in detail, on the basis of an extended reactivity decomposition methodology. This has allowed quantification of the different phase-space contributions (in terms of reaction rate types, energy groups and spatial regions) of a given calculated pin-removal reactivity worth, thus providing useful insights regarding the most important sources of error in each of the assembly codes investigated.

Fertile-Free Annular Fuel for Plutonium Recycling

Up to 50% increase in the power density of the existing pressurized water reactor (PWR)-type reactors can be achieved by the use of internally and externally cooled annular fuel geometry. As a result, the accumulated stockpiles of Pu, especially if incorporated in fertile-free inert matrix, can be burnt at a substantially higher rate as compared with the conventional mixed oxide-fueled reactors operating at standard power density. In this work, we explore the basic feasibility of a PWR core fully loaded with Pu incorporated in fertile-free fuel of annular internally and externally cooled geometry and operating at 150% of nominal power density. We evaluate basic burnable poison designs, fuel management strategies, and reactivity feedback coefficients. The three-dimensional full core neutronic analysis performed with Studsvik Core Management System showed that the design of such a Pu-loaded annular fuel core is feasible but significantly more challenging than the Pu fertile-free core with solid fuel pins operating at nominal power density. The main difficulty arises from the fact that the annular fuel core requires at least 50% higher initial Pu loading in order to maintain the standard fuel cycle length of 18 months. Such a high Pu loading results in hardening of the neutron spectrum and consequent reduction in reactivity worth of all reactivity control mechanisms and, in some cases, positive moderator temperature coefficient (MTC). The use of isotopically enriched Gd and Er burnable poisons was found to be beneficial with respect to maximizing Pu burnup and reducing power peaking factors. Overall, the annular fertile-free Pu-loaded high-power-density core appears to be feasible, although it still has relatively high power peaking and potential for slightly positive MTC at beginning of cycle. However, we estimate that limiting the power density to 140% of the nominal case would assure acceptable core power peaking and negative MTC at all times during the cycle.

Lithium-6-Based Passive Reactivity Control Devices for a Gas-Cooled Fast Reactor

In this paper passive reactivity control devices for a Generation IV gas-cooled fast reactor (GCFR) are discussed. The proposed devices use liquid 6Li as absorber. The device is triggered by a freeze seal, and upon activation the 6Li is irreversibly introduced into the core region by pressure differences. The device is dubbed the lithium injection module (LIM). Transient thermohydraulic calculations were done using the CATHARE2 code on a simplified thermohydraulic model of GFR600, a 600-MW(thermal) GCFR investigated in the scope of the European GCFR-STREP. The thermohydraulic model uses an accurate model of the ceramic fuel plates and includes natural convection decay heat removal circuits. To properly account for power production during the transient, a synthetic decay power curve was made based on the ANSI/ANS-5.1-1994 law. Loss-of-flow and control rod withdrawal/ejection transients are presented. Neutronic calculations show that the LIMs have a low reactivity worth between -2.1 and -1.5 $. In spite of their low worth, the LIMs are capable of keeping the reactor power bounded during all calculated transients. Shutdown is not always achieved, depending on the kind of transient under consideration. For pressurized loss of flow, recriticality due to Doppler feedback may become problematic in the natural-circulation phase. For rapid control rod ejections, the resulting very fast power transients cause concern for material degradation. One LIM would be enough to control reactor power, but redundancy may call for more than one LIM in the core.

Evaluation of the 80-group KAFAX library using JEFF3.0, ENDF/B-VI.8 and JENDL3.3 for minor actinide and fission product transmutation

For a core design work of a nuclear transmutation reactor, nuclear data were added to the KAFAX-F22 library based on three master libraries, JEFF3.0, ENDF/B-VI.8 and JENDL3.3. Evaluation of the modified library was done by comparing criticality and the reaction ratio with experimental data. Evaluation of Cm-244 and Np-237 was calculated by TRANSX and DANTSYS for critical assemblies, JEZEBEL and FLATTOP. For the evaluation of minor actinide isotopes for which there was no experiment, calculation results of benchmark problems were compared with MCNP4c2 results. The geometry and composition of these problems are those of the reference model for the Pb–Bi cooled fast spectrum transmutation reactor. There were good agreements among the evaluation data for most of the isotopes except Am-241, Am-243, Cs-135 and Cs-137. In case of Cm-242, results of the KAFAX library from ENDF/B-VI.8 showed a large discrepancy in reactivity worth compared with other libraries.

Critical experiments to determine amount of U-235 in research reactor fuel assemblies

Four different critical core configurations of the IRR1 [Bettan, M., Hirshfeld H., Levine, S.H., 2002. Method to determine the burnup of the IRR1 fuel assemblies. In: Proceedings of the 21st Conference of the Nuclear Societies in Israel, p. 38–41], Cores 1, 2, 3, and 4, have been used to determine experimentally the reactivity changes caused by interchanging different fuel assemblies (FAs) in the same core position [Bettan, M., Hirshfeld H., Levine, S.H., 2002. Method to determine the burnup of the IRR1 fuel assemblies. In: Proceedings of the 21st Conference of the Nuclear Societies in Israel, p. 38–41; Levine, S.H., Kim, S.S., 1985. Development of an asymmetric multiple-position neutron source (AMPNS) method to monitor the criticality of a degraded reactor core. Ann. Nucl. Energy 12, 517; IAEA-TECDOC-633, 1992. Determination of research fuel burnup, January]. The different fuel assemblies varied in their burnup or loss in U-235 causing the core reactivity to change systematically with each FA interchange. The control blade height was altered to compensate for this change in the core reactivity to maintain the core critical. The control blade reactivity worth as a function of its height was measured by period measurements so that the reactivity change due to the interchange of an FA was determined. The WIMS/EXT2 [WIMS-D4: Winfrith Improved Multigroup Scheme (WIMS) code system, 1969. Computer Code Package CCC-576, RSICC/ORNL, Oak Ridge, TN; Fowler, T.B. et al., 1967. EXTERMINATOR-II: A FORTRAN IV Code for solving multigroup diffusion equations in two dimensions, ORNL-4078, ORNL, April] codes are used to evaluate the change in core reactivity as a function of FA U-235 or burnup. By relating the calculation of the reactivity change with the measured reactivity change, the burnup or U-235 mass in each FA is determined. The U-235 mass of the FAs gave good agreement between measurements made of the same FA in different cores and appears to bring better agreement between the keff calculated for the different cores when each are just critical. The results of these measurements and calculations are presented in this paper.

ICONE15-10371 Application of Bias Factor Method with Use of Virtual Experimental Value to Prediction Uncertainty Reduction in Void Reactivity Worth of Breeding Light Water Reactor

ICONE15-10371 Application of Bias Factor Method with Use of Virtual Experimental Value to Prediction Uncertainty Reduction in Void Reactivity Worth of Breeding Light Water Reactor

Sensitivity Analysis based on Transport Theory

To estimate the uncertainty of the neutronic parameters it is required to calculate sensitivity coefficients to cross section changes. The sensitivity coefficients are usually calculated based on the diffusion theory. However, the accuracy of the sensitivity coefficients becomes doubtful because of the use of diffusion theory. The purpose of this paper is to derive the generalized perturbation theory using the transport theory, and to evaluate transport effect on the sensitivity coefficients. Sensitivity calculations were performed in 70 energy groups. The calculated sensitivity coefficients were compared with those obtained from direct transport calculations to check the accuracy of the present method, and those obtained from the generalized perturbation calculation based on the diffusion theory to investigate the transport effect on sensitivity coefficients. From the comparison, the difference was remarkably large for the neutronic parameters such as reaction rate distribution, reaction rate ratio, and reactivity worth. For example, the sensitivity coefficient of the 238U capture to 239Pu fission rate ratio in the lower core region of the under-moderated LWR core was large in the energy range 17.6–22.6 eV, and the sensitivity to 238U capture cross section calculated by the transport theory was 20% larger than that by the diffusion theory.

06/00642 Sustainable and economic hydrogen cogeneration from nuclear energy in competitive power markets: Miller, A. I. and Duffey, R. B. Energy, 2005, 30, (14), 2690–2702.

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 600 MWe European Lead-cooled SYstem (ELSY), under development within the 6th and 7th EURATOM Framework Programmes. Calculations have been carried out on a 1500 MWth 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 5000 pcm) is due to the predominant contributions of even isotopes – among which 238U plays a major role, being responsible for roughly 4300 pcm – as a consequence of their fission cross-section high sensitivity to spectral hardening (threshold reactions), despite their modest contribution to the total fissions.

BWR核設計コードによるMOX臨界試験EPICUREおよびMISTRALの解析

For the purpose of verification of the nuclear analysis method of BWR for mixed oxide (MOX) cores, UO2 and MOX fuel critical experiments EPICURE and MISTRAL were analyzed using nuclear design codes HINES and CERES with ENDF/B nuclear data file. The critical keffs of the absorber worth experiments, the water hole worth experiments and the 2D void worth experiments agreed with those of the reference experiments within about 0.1%Δk. The root mean square differences of radial power distributions between calculation and measurement were almost less than 2.0%. The calculated reactivity worth values of the absorbers, the water hole and the 2D void agreed with the measured values within nearly experimental uncertainties. These results indicate that the nuclear analysis method of BWR in the present paper give the same accuracy for the UO2 cores and the MOX cores.

On the accuracy of reactor physics calculations for square HPLWR fuel assemblies

Although the supercritical-pressure or high-performance light water reactor (HPLWR) concept is largely based on the well-established technological experience available with conventional light water reactors, there is still no consensus on various key design features such as an optimal layout for the fuel assembly. This results mainly from the very large density variations of supercritical-pressure water in the core, which render it difficult to ensure reliable values for parameters such as power peaking factors and reactivity worths. The present paper describes studies carried out to compare deterministic and Monte Carlo codes for analysing a representative square HPLWR lattice with uniform 5%-enriched UO 2 fuel. The main purpose has been to assess the prediction accuracies achievable for integral parameters such as the multiplication factor, control absorber effectiveness, moderator/coolant density reactivity feedback and pin power distributions. The results show good agreement between the deterministic and stochastic calculations for the unperturbed lattice. However, for certain perturbed situations involving, for example, local coolant density changes in the assembly or control absorber insertion, the observed discrepancies are large enough to question the basic viability of the reactor physics design, e.g. with respect to the thermal performance.

Managing the reactivity excess of the gas turbine-modular helium reactor by burnable poison and control rods

The gas turbine-modular helium reactor coupled to the deep burn in-core fuel management strategy offers the extraordinary capability to incinerate over 50% of the initial inventory of fissile material. This extraordinary feature, coming from an advanced and well tested fuel element design, which takes advantage of the TRISO particles technology, is maintained while the reactor is loaded with the most different types of fuels. In the present work, we assumed the reactor operating at the equilibrium of the fuel composition, obtained by a 6 years irradiation of light water reactor waste, and we investigated the effects of the introduction of the burnable poison and the control rods; we equipped the core with all the three types of control rods: operational, startup and shutdown ones. We employed as burnable poison natural erbium, due to the 167Er increasing neutron capture microscopic cross-section in the energy range where the neutron spectrum exhibits the thermal peak; in addition, we utilized boron carbide, with 90% enrichment in 10B, as the absorption material of the control rods. Concerning the burnable poison studies, we focused on the k eff value, the 167Er mass during burnup, the influence of modifying the radius of the BISO particles kernel and the fuel and moderator coefficients of temperature. Concerning the control rods studies, we investigated the reactivity worth, the changes in the neutron flux profile due to a partial insertion, the influence of modifying the radius of the BISO particles kernel and the β eff, at the beginning of the operation.

Effect of different cross-section data sets on reflectors of a typical material test research reactor

Studies were carried out to investigate the effects of various neutron cross-section data sets on the effectiveness of different reflector materials used in a typical MTR type research reactor by three-dimensional modeling of the reactor core using reactor simulation codes WIMS-D4 and CITATION. The use of old and newly released WIMSD libraries based on different cross-section data sets resulted in almost same effective reflector thicknesses for the same reflector materials, however, significant differences were observed in reactivity worths of some reflectors.

Assessment of the MDNBR enhancement methodologies for the SMART control rods banks withdrawal event

For an electricity generation and seawater desalination, a 330 MW System-integrated Modular Advanced ReacTor (SMART) was developed by KAERI. The safety level of the SMART is enhanced when compared to that of the typical commercial reactors, with the aid of an elimination of a large break loss of coolant accident by placing the major components of the primary system in a reactor vessel and the adoption of a new technology and a passive design concept into the safety system. However, the events related to reactivity and power distribution anomalies have been evaluated as vulnerable points when compared to the other initiating events in the SMART, since the reactivity worth of the control rods (CR) banks is quite large due to the boron free core concept. Especially, safety margins, i.e., minimum departure from nucleate boiling ratio (MDNBR), are significantly threatened during the CR banks withdrawal event. Therefore, MDNBR enhancement methodology for the CR banks withdrawal event should be considered to further enhance the safety level of the SMART design. Two methodologies have been suggested to enhance the MDNBR during the CR banks withdrawal event: the application of a DNBR trip function into a core protection system and a turbine trip delay methodology. Sensitivity studies are performed to evaluate the two MDNBR enhancement methodologies and show that the suggested methodologies could enhance the MDNBR during the CR banks withdrawal event of the SMART.

Reactivity worth of the control rods at the ET-RR-2 reactor core

Abstract Safety analysis of research reactors include a determination of control rod worth and kinetic parameters upon which the dynamic performance of the reactor depends. The core control of the ET-RR-2 reactor is accomplished through six Ag-In-Cd alloy-absorbing plates, which can be inserted into the core with a high velocity. In the present work, the fully and partially inserted control rod configuration in 3D reactor geometry and the effect of reactivity worth of the control rods on the reactor core is studied. The reactivity worth were calculated for each control rod and also for combinations of more than one rod. The total worth of all control rods in the reactor core is about 16.95% k/k from the core reactivity. Other neutronic parameters, i. e. the 3D neutron flux distribution and the change of the power peaking factor as a results of inserted and withdrawing the control rods were calculated.

Development of a Fissile Materials Irradiation Capability for Advanced Fuel Testing at the MIT Research Reactor

A fissile materials irradiation capability has been developed at the Massachusetts Institute of Technology (MIT) Research Reactor (MITR) to support nuclear engineering studies in the area of advanced fuels. The focus of the expected research is to investigate the basic properties of advanced nuclear fuels using small aggregates of fissile material. As such, this program is intended to complement the ongoing fuel evaluation programs at test reactors. Candidates for study at the MITR include vibration-packed annular fuel for light water reactors and microparticle fuels for high-temperature gas reactors. Technical considerations that pertain to the design of the MITR facility are enumerated including those specified by 10 CFR 50 concerning the definition of a research reactor and those contained in a separate license amendment that was issued by the U.S. Nuclear Regulatory Commission to MIT for these types of experiments. The former includes limits on the cross-sectional area of the experiment, the physical form of the irradiated material, and the removal of heat. The latter addresses experiment reactivity worth, thermal-hydraulic considerations, avoidance of fission product release, and experiment specific temperature scrams.