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In this study, the characteristics of changes in reactivity due to increasing turnup of uranium-fueled fast reactors are analyzed. A new classification system for nuclear reactor cores based on their uncontrolled tendency for reactivity changes during turnup was introduced and the design-optimization strategy for any fast reactor core aimed at a minimized reactivity swing is outlined. The 235U feed-fuel enrichment level that minimizes the burnup reactivity swing of a sodium-cooled metallic-fueled core is 10% to 12.5% for an average target fuel burnup of 1% to 20% FIMA (fission of initial metal atom). The higher the target burnup of the system, the lower the feed-fuel enrichment level that minimizes swing. The minimum attainable swing for a 125-MW (thermal) metallic-fueled sodium-cooled core is found to be ~200 pcm for 5% FIMA burnup and increases to ~ 800 pcm for a system aiming at 10% FIMA. In general, if the target discharge burnup is doubled, the minimum attainable burnup reactivity swing quadruples. Any optimized minimum reactivity swing core will form a positive parabolic uncontrolled reactivity trajectory with burnup, where the beginning of cycle and end of cycle reactivities are equal. Uranium-fueled fast cores with minimized burnup reactivity swing are net consumers of fissile material, with a fissile conversion ratio in the range of 0.7 to 0.9. [ABSTRACT FROM AUTHOR]

To consolidate the safety assessment for liquid-metal fast breeder reactors (LMFBRs), hypothetical core disruptive accident (HCDA) sequences have been extensively studied over the past decades. Numerous analyses of the so-called initiating phase (or primaly phase) of a HCDA have been made with the safety analysis system code SAS4A. The SAS4A accident analysis code requires that subassemblies or groups of subassemblies be represented together as independent channels. For simulating a severe accident sequence, a subassembly-to-channel assignment procedure has to be implemented to produce the consistent SAS4A input decks. Generally, one uses imposed criteria over relevant reactor parameters to determine the subassembly-to-channel arrangement. The multiple-assembly-per-channel approach introduces corewide coherency effects, which can affect the reactivity balance and therefore the overall accident development. In this paper, a subassembly-to-channel assignment procedure based on the subassembly power-to-flow ratio is presented and implemented to generate the SAS4A input decks over a range of parameter values. The corresponding SAS4A calculations have been performed on a large LMFBR. The purpose of the present series of calculations is to investigate the magnitude of errors encountered in the analysis of the initiating phase related to the subassembly-to-channel arrangement selection, by comparison with a one-subassembly-per-channel reference solution. It appears that a refinement in the channel arrangement substantially reduces corewide coherency effects. Analysis of the calculations also suggests that an accurate representation of the scenario requires the number of channels to be on approximately the same order of magnitude as the total number of subassemblies. Numerical results are examined to provide the reader with quantitative measurements of bias related to subassembly-to-channel arrangement. [ABSTRACT FROM AUTHOR]

The Electric Power Research Institute (EPRI) has sponsored the development of a set of benchmarks that can be used to quantify the bias and uncertainly in computed reactivity decrements due to burnup. The bias and uncertainly covers imprecision in both the nuclide inventory and cross sections. The EPRI benchmarks are a function of enrichment, operating conditions (such as soluble boron concentration, burnable absorbers, and specific power), and storage rack conditions. The benchmarks are analyzed using SCALE 6.1 with both ENDF/BV and ENDF/B-VII cross-section libraries. The depletion analyses are performed using the TRITON module, and the criticality calculations are performed with KENO-V.a and MCNP. The analysis shows that SCALE 6.1 with the ENDF/ B-VII 238-group cross-section library supports the use of a depletion bias of only 0.0015 in Δk, where k represents the neutron multiplication factor, at peak reactivity after discharge jS"om the core. This peak reactivity occurs after 100 h of cooling. If credit is taken for more cooling, the bias should be increased to 0.0025. The depletion uncertainty is 0.0064. Using MCNP for the criticality calculations rather than KENO-V.a produces essentially the same results if the same ENDF/B cross-section library is used. Reliance on the ENDF/B-V cross-section library produces much larger disagreement with the benchmarks. The analysis covers numerous combinations of depletion and criticality options. In all cases, the historical uncertainty of 5% of the Δk of depletion ("Kopp memo") was shown to be conservative for fuel with >30 GWd/T burnup. However, the Kopp memo's uncertainty may be exceeded at low burnups where the absolute magnitude of the uncertainty is small. [ABSTRACT FROM AUTHOR]

Pressurized water reactor (PWR) assembly reactivity distributions are inferred from ~600 in-core flux maps taken during 44 cycles of operation of the Catawba and McGuire nuclear power plants. The reactivity distribution for each .flux map is determined by systematically searching for fuel subbatch reactivities that minimize differences bem,een measured and computed 235U fission rates. More than eight million core calculations are used to reduce one million measured signals to a set of ~2500 experimental fuel reactivities for fuel with up to 55 GWd/T burnup. These measured reactivity changes with depletion can be used to validate computer code systems used for burnup credit. To reduce the effort required to quantify computer code system biases and uncertainties, the measured changes in fuel depletion reactivity have been reduced to a set of experimental PWR lattice benchmarks for the change in reactivity as a function of fuel burnup. Results demonstrate that the uncertainty of hot-full-power (HFP) depletion reactivity of the benchmarks is <250 pcm up to 55 GWd/T burnup. Oak Ridge National Laboratory's TSUNAMI tools are used to extend HFP results to cold conditions, and reactivity decrement uncertainties increase to ~600 pcm. These experimental benchmarks provide a basis for quantification of combined nuclide inventory and cross-section uncertainties in computed reactivity decrements. It is demonstrated that flux map data reduction is not sensitive to the analytical tools (CASMO/SIMULATE) employed here, and experimental fuel depletion reactivity decrements and uncertainties are anticipated to be independent of fuel management code system use for the data reduction. For CASMO-based analysis, the HFP reactivity burnup decrement biases are shown to be <250 pcm up to 55 GWd/T burnup, and results show that the historical "Kopp memo" 5% reactivity decrement uncertainty assumption is conservative. INSET: 3. [ABSTRACT FROM AUTHOR]

The use of highly-enriched fuels is an effective method for reducing the number of spent fuel assemblies and improving fuel cycle economics, e.g., with >5 wt%235 U. However, from a criticality safety point of view, such high enrichment levels require a significant investment for the considerable modification of most facilities and equipment. Erbia-credit super-high-burnup fuel offers an effective solution that can solve the problem: Small amounts of erbia added to the entire amount of UO2 powder can reduce the reactivity level to less than that observed at a 5 wt% enrichment level, thus eliminating the need for the modifications mentioned above. A series of criticality safety analyses has been performed to determine the minimum and sufficient content of erbia that can guarantee a suitable erbia credit. As a noteworthy result, the erbia content required was determined for corresponding values of uranium enrichment in a range >5 wt%, as indicated in our ECOS (Erbia COntent for Sub-criticality judgment) diagram. This paper outlines a series of criticality safety analyses and explains how the minimum erbia content can be determined to ensure sub-criticality for a >5 wt% enrichment fuel to ensure that the fuel obtained is equivalent to that whose enrichment is <5 wt%. [ABSTRACT FROM AUTHOR]

Spent nuclear fuel transported in large casks must remain subcritical in all credible configurations for normal operation and hypothetical accident conditions. The effects on spent nuclear fuel reactivity from "worst-case" accident scenarios were surveyed in NUREG/CR-6835, "Effects of Fuel Failure on Criticality Safety and Radiation Dose for Spent Fuel Casks." The survey used scenarios that were postulated to provide theoretical upper limits for reactivity effects of fuel relocation, although they were described as going "beyond credible conditions." These scenarios involved physical changes either to fuel assembly rod arrays or to collections of fuel pellets with the fuel skeleton removed. To provide more credible estimates of the probability and maximum reactivity changes, a process is presented that deconstructs each scenario into a set of subscenarios and identifies the physical phenomena required to create the subscenario. The boundary between credible but unlikely scenarios and incredible scenarios is more easily discernible with this process. For marginally credible worst-case scenarios, it is concluded that the maximum reasonable reactivity increase either is less than the mandated administrative nuclear criticality safety margin for scenarios involving physical changes to fuel assembly rod arrays or is a substantial reactivity decrease for scenarios involving collections of fuel pellets. A cask designer could apply scenario deconstruction to evaluate the physical limits that apply to a particular transportation cask, and perform calculations specific to a particular cask design to show that criticality safety requirements are met. [ABSTRACT FROM AUTHOR]

The present paper describes the evaluation of the control rod worth data obtained in the Monju restart core. The best-estimate value and its uncertainty are evaluated in detail. As in the criticality evaluation, data obtained in the previous test are evaluated in the same level of detail. Evaluated results of control rod worth are consistent among control rods at symmetrical positions for a random uncertainty. The experimental uncertainty is ±2%, where the dominant uncertainty originates from the delayed neutron parameters used in the period method. The correlation in the uncertainties is also evaluated among different control rods and tests. One can discuss a difference in worth among different control rods and cores without a detailed knowledge of the original uncertainty evaluation. Based on the evaluated data, calculation accuracy is investigated with JENDL-3.3 and JENDL-4.0. It is confirmed that the calculation accuracy is within an experimental uncertainty of ±2% for each layer and 10B content. [ABSTRACT FROM AUTHOR]

The present technical note describes the evaluation of the isothermal temperature coefficient data obtained in the Monju restart core. As in the preceding evaluations on the criticality and the control rod worth, the best-estimate value and its uncertainty are evaluated as accurately as possible. Data obtained in the previous test are evaluated in the same level of detail. The measured data show that the fuel composition change from the previous test decreases the magnitude of the temperature coefficient by ~8%. Through a sensitivity analysis, it is confirmed that the decrease is mainly brought by the composition of 241Pu and 241Am. The best calculation accuracy within the experimental uncertainty of 2% is attained with JENDL-4.0 for the previous core; however, such a good accuracy is not achieved for the restart core. A further experimental investigation is required to solve the dependence of calculation accuracy on the core configurations. [ABSTRACT FROM AUTHOR]

The present work describes the preliminary design of a 25-MW(thermal) nuclear reactor capable of providing safe and reliable heating and electricity to any Canadian Forces Bases, especially in the Arctic, as well as in comparable civilian applications. The aim of the project is to provide a nuclear reactor system with sufficient inherent safety characteristics as it is intended to run in automatic mode and be monitored by operators with limited experience and training. For the neutronics calculations, the design work of the reactor's core is carried out using the probabilistic simulation code MCNP 5 along with the Win frith Improved Multigroup Scheme-Atomic Energy of Canada Limited (WIMS-AECL) deterministic code, Version 3.1, thus permitting a code-to-code comparison of the numerical results. Several design constraints related to coolant temperature and pressure, reactivity control, fuel enrichment, and time between refueling have been considered. The final reactor concept, named the Super Near Boiling 25 reactor (SNB25), provides heat energy dedicated to building and domestic water heating and supplies electricity through an organic Rankine cycle energy conversion plant. SNB25 employs TRISO fuel particles, contained in zirconium-sheathed fuel rods, and is light water cooled and moderated. Complete reactivity control is achieved through simple and reliable mechanical means consisting of 133 control rods and six adjustable radial reflector plates. The optimized reactor core configuration, along with its intrinsic control system, allows for the power plant to operate safely for more than a decade between refuelings from a typical central heating plant or the basement of a multilevel office building. The work also included a preliminary investigation of the nonnuclear part of the energy supply system including heat exchangers and the turbine-driven, electricity-generating system. [ABSTRACT FROM AUTHOR]

The aim of this study was to validate a Monte Carlo-based model of the Greek Research Reactor-1 (GRR-1) developed with the MCNP5 code. The GRR-1 core was modeled in detail using the exact geometry without approximations. The inventory, of the core was derived using the WIMS-ANL code, taking into account the different 235U burnup of each fuel assembly. The model was validated against experimentally determined control rod reactivity worth and neutron flux measurements performed in various irradiation positions. The ratio of the calculated-to-measured integral reactivity of each of the five control rods was found to be 0,972 ± 0.151, 1.083 ± 0.168, 1.156 ± 0.179, 0.874 5: 0.137, and 1.097 ± 0.170. The calculated-to-measured thermal neutron flux ratios ranged from 0.83 ± 0.04 to 1.22 ± 0.07. Therefore, good agreement between MCNP calculated and experimental values was observed. The GRR-1 core model will be fully implemented in the design of material irradiation experiments along with reactor safety and fuel management studies. [ABSTRACT FROM AUTHOR]

A conceptual design study has been performed on upgrading the core performance of the Japanese fast breeder reactor (FBR) Monju. The main aim of this study is to investigate and demonstrate the feasibility of an upgraded core with an extended refueling interval of 365 equivalent full-power days and increased average fuel burnup of 150 GWd/t, which are expected in future commercial FBRs. Two main design measures have been taken to accommodate the largely increased burnup reactivity loss and the reactivity control characteristics for the 1-yr cycle operation: (a) A modified fuel pin specification with increased pin diameter, pellet density, and fissile height has been chosen to improve the burnup reactivity loss per extended cycle, and (b) the control rod specification has been modified to enhance the reactivity worth by increasing the 10B content to ensure sufficient shutdown margin. The major core characteristics that have been evaluated are the core power distribution, safety-related reactivities such as Doppler and sodium void effect, thermal hydraulics, and reactivity control characteristics. The results show that even a medium-sized upgraded core with a volume of ~2.5 m³ could achieve the primary targeted performance of 1-yr cycle operation, without causing significant drawbacks to the core characteristics and safety aspects. The feasibility of the upgraded core concept has thus been demonstrated. [ABSTRACT FROM AUTHOR]

The Encapsulated Nuclear Heat Source (ENHS) is a small lead-bismuth-cooled Generation IV reactor designed to have a once-for-life core with a nearly zero burnup reactivity swing and heat removal by natural circulation. All the ENHS cores designed so far have positive coolant void reactivity. This work searches for ENHS core designs having negative coolant void reactivity feedback and quantifies the penalty associated with a design for negative void reactivity. The approaches tried for turning the positive void coefficient negative are as follows: (a) enhancing the neutron leakage probability by reducing the fuel length, introducing neutron absorbers at the core boundary, using a gas-lift pump to introduce gas bubbles throughout the coolant in the core and fission gas plenum regions, and incorporating neutron-streaming channels in and adjacent to the core; (b) introducing into the core materials, such as Ca3N2, having enhanced absorption cross section at high energy; and (c) introducing into the core materials, such as CaH2, that will keep the neutron spectrum softer in the case of coolant voiding. The preferred negative void reactivity core design consists of 100-cm-long fuel rods, a 20-cm-thick B4C layer below the core, and a voided channel around the core radial boundary. The reactivity effect is negative when the coolant is voided from the entire core and even from only the central region of the core. In order to maintain a nearly zero burnup reactivity swing over at least 20 effective full-power years (EFPY), the core pitch-to-diameter ratio (P/D) has to be reduced from the reference value of 1.36 to 1.20. Correspondingly, for a given ENHS module dimensions the power level that can be removed by natural circulation from this core is ~75% of the reference core. Allowing a burnup reactivity swing of ~0.1% over 20 EFPY enables attaining a negative void reactivity core having P/D of 1.27 that can deliver ~93% of the P/D = 1.36 core power. [ABSTRACT FROM AUTHOR]

This paper investigates the basic feasibility of using reactor-grade Pu infertile-free fuel (FFF) matrix in pressurized water reactors (PWRs). Several important issues were investigated in this work: the Pu loading required to achieve a specific interrefueling interval, the impact of inert matrix composition on reactivity constrained length of cycle, and the potential of utilizing burnable poisons (BPs) to alleviate degradation of the reactivity control mechanism and temperature coefficients Although the subject was addressed in the past, no systematic approach for assessment of BP utilization in FFF cores was published. In this work, we examine all commercially available BP materials in all geometrical arrangements currently used by the nuclear industry with regards to their potential to alleviate the problems associated with the use of FFF in PWRs. The recently proposed MgO-ZrO2 solid-state solution fuel matrix, which appears to be very promising in terms of thermal properties and radiation damage resistance, was used as a reference matrix material in this work. The neutronic impact of the relative amounts of MgO and ZrOP2 in the matrix were also studied. The analysis was performed with a neutron transport and fuel assembly burnup code BOXER. A modified linear reactivity model was applied to the two-dimensional single fuel assembly results to approximate the full core characteristics. Based on the results of the performed analyses, the Pu-loaded FFF core demonstrated potential feasibility to be used in existing PWRs. Major FFF core design problems may be significantly mitigated through the correct choice of BP design. It was found that a combination of BP materials and geometries may be required to meet all FFF design goals. The use of enriched (in most effective isotope) BPs, such as 167Er and 157Gd, may further improve the BP effectiveness and reduce the fuel cycle length penalty associated with their use. [ABSTRACT FROM AUTHOR]

In late 1993 and early 1994, tests in France and Japan showed that cladding damage in fuel rods with burnups above 50 GWd/ton occurs at much lower energies than in unirradiated fuel rods when exposed to large power pulses. During the last decade, significant additional test results have become available to permit an interim assessment of potential cladding failure in reactivity-initiated accidents (RIAs) in reactors with fuel burnups above 40 GWd/ton, which is generally regarded as high-burnup fuel. These data are summarized, and systematic biases due to atypical test conditions are identified. The magnitude of biases in the fuel enthalpy for failure are estimated to range from -19 to +27 cal/g for the cases analyzed. With these adjustments, a lower bound of the enthalpies for experimentally observed cladding failure is compared with potential enthalpies in pressurized water reactors and boiling water reactors. Based on available information on control rod worths, it is concluded that current operating reactors in the United States are not likely to experience cladding failure during the worst postulated RIAs. [ABSTRACT FROM AUTHOR]

The Belgian Material Test Reactor BR2 is a strongly heterogeneous high-flux engineering test reactor at SCKCEN (Centre d'Etude de l'Energie Nucléaire) in Mol with a thermal power of 60 to 100 MW. It deploys highly enriched uranium, water-cooled concentric plate fuel elements, positioned inside a beryllium reflector with a complex hyperboloid arrangement of test holes. The objective of this paper is to validate the MCNP&ORIGEN-S three-dimensional (3-D) model for reactivity predictions of the entire BR2 core during reactor operation. We employ the Monte Carlo code MCNP-4C to evaluate the effective multiplication factor keff and 3-D space-dependent specific power distribution. The one-dimensional code ORIGEN-S is used to calculate the isotopic fuel depletion versus burnup and to prepare a database with depleted fuel compositions. The approach taken is to evaluate the 3-D power distribution at each time step and along with the database to evaluate the 3-D isotopic fuel depletion at the next step and to deduce the corresponding shim rod positions of the reactor operation. The capabilities of both codes are fully exploited without constraints on the number of involved isotope depletion chains or an increase of the computational time. The reactor has a complex operation, with important shutdowns between cycles, and its reactivity is strongly influenced by poisons, mainly ³He and 6Li from the beryllium reflector, and the burnable absorbers 149Sm and 10B in the fresh UAlx fuel. The computational predictions for the shim rod positions at various restarts are within 0.5 $ (β eff = 0.0072). [ABSTRACT FROM AUTHOR]

NFCSim is an event-driven, time-dependent simulation code modeling the flow of materials through the nuclear fuel cycle. NFCSim tracks mass flow at the level of discrete reactor fuel charges/discharges and logs the history of nuclear material as it progresses through a detailed series of processes and facilities, generating life-cycle material balances for any number of reactors. NFCSim is an ideal tool for analysis--of the economics, sustainability, or proliferation resistance--of nonequilibrium, interacting, or evolving reactor fleets. The software couples with a criticality and burnup engine, LACE (Los Alamos Criticality Engine). LACE implements a piecewise-linear, reactor-specific reactivity model for its criticality calculations. This model constructs fluence-dependent reactivity traces for any facility; it is designed to address nuclear economies in which either a steady state is never obtained or is a poor approximation. LACE operates in transient and equilibrium fuel management regimes at the refueling batch level, derives reactor- and cycle-dependent initial fuel compositions, and invokes ORIGEN2.x to carry out burnup calculations. [ABSTRACT FROM AUTHOR]

An advanced analysis code, ARKAS_cellule, has been developed to determine the core distortion and the mechanical behavior of fast reactors. In this code, each hexagonal subassembly duct is represented by a folded thin plate structure divided into a user-specified number of shell elements so that the interduct contact forms and the cross-sectional distortion effect of each duct are properly taken into account. In this paper, the numerical model of the ARKAS_cellule code is introduced, and the analytical results for two validation problems are presented. From a single duct compaction analysis, the first validation problem, it is clarified that the new analytical model is applicable to simulating the change of duct compaction stiffness that depends on the loading conditions such as the loading pad forms and the number of contact faces. The second validation analysis has been conducted by comparison with the experimental values obtained by the National Nuclear Corporation Limited in the United Kingdom using the core restraint uniplanar experimental rig (CRUPER), an ex-reactor rig in which a cluster of 91 short ducts is compressed by 30 movable peripheral rams toward the center of the cluster in seven stages. The analysis clarified that the predictions obtained using ARKAS_cellule agree well with the measured ram loads and interwrapper gap widths during the compaction sequence. One may conclude that ARKAS_cellule is valid for quantitative analysis of the core mechanical behavior and will be particularly useful for the evaluation of transient deformation of core assemblies during accidents in which the distortion of loading pads have important effects on obtaining favorable reactivity feedback. [ABSTRACT FROM AUTHOR]