The core averaged burn-up reactivity has been measured and calculated for the Joyo MK-II core. In order to evaluate the relationship between the calculational error of burn-up reactivity and the nuclear data or calculated neutron flux, the ...
National Technical Information Service (NTIS)
In order to evaluate core burnup characteristics of a high conversion light water reactor (HCLWR) with tight pitched lattice, core burnup calculation was made using two dimensional diffusion method. The volume ratio of moderator to fuel is about 0.8 in th...
The results of calculating the nuclear fuel burnup and transuranium isotope buildup in the RBMK type reactor core performed by means of the combined iterative method are given. The TRIFON and TERMIT programs have been used for solving the spatial and ener...
Calculations were made to determine the fuel burn-up of the U-Mo fueled Hallam reference core. A peak exposure of 7000 Mwd/t, corresponding to a radial- averaged core exposure of 4400 Mwd/t, or continuous full power operation for about 1.5 years, requires a DELTA k/k of 6.45%; 4.06% for ...
Energy Citations Database
A direct result of making extensive neutron flux measurements in the ETR was the possibility of calculating the time and position variation of the core power and the burnup of U/sup 235/ in the core. These values were calculated for each of the sixteen control rod fuel sections and for ...
In order to evaluate core performances of BWR type reactors, the three dimensional core burnup calculation code COREBN-BWR and the fuel management code HIST-BWR have been developed. In analyses of BWR type reactors, thermal hydraulics calculations must be...
This study was undertaken for the US Nuclear Regulatory Commission to determine the fuel enthalpy during a rod drop accident (RDA) for cores with high burnup fuel. The calculations were done with the RAMONA-4B code which models the core with 3-dimensional neutron kinetics and multiple parallel coolant channels. The ...
The neutron source to be used in calculations of the irradiation of nuclear reactor pressure vessels depends not only on the power distribution in the core but also on the burnup distribution. The burnup affects both the strength and the spectrum of the source, with each effect increasing the displacement rate in ...
The technique and KODRA computer code used for calculating homogeneous reactor core isotope composition change with fuel burnup are described. The calculation of changes in the neutron efficient multiplication factor, energy release and neutron flux spati...
Demonstration that advanced reactor designs satisfy safety and performance goals requires the analysis of uncertainties in calculated reactor characteristics. Two of the important performance characteristics of advanced liquid metal reactors (LMR) are the burnup reactivity swing and the local power peaking factor. Previous work reported a study of the ...
An analysis of metal-, oxide, and nitride-fueled advanced liquid metal reactor cores was performed to investigate the calculated differences in sodium void reactivity, and to determine the relationship between sodium void reactivity and burnup reactivity ...
In order to evaluate basic core performances of a high conversion boiling water reactor with an axially heterogeneous core, three dimensional core burnup calculations coupled with thermal-hydraulics calculations were made under the Haling strategy. The ef...
The new burnup strategy CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy production) is proposed. With this burnup strategy, distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed and without any change in their shapes. The excess reactivity is constant ...
The response of an ex-core neutron detector to fuel assembly vibrations in an 1150-MWe Westinghouse pressurized-water reactor (PWR) was determined by performing space-dependent reactor-kinetics calculations. The effect on the detector response of reducing the soluble-boron concentration in the coolant and fuel burnup over the first ...
DOE Information Bridge
An analytical method of calculating flux distributions and burnup of important uranium and plutonium isotopes has been developed for the Berkeley TRIGA III research reactor. The core is described by two group, two dimensional diffusion theory with cross sections averaged over a TRIGA cell region. Burnup is ...
This technical report presents burn-up and in-core fuel management calculations to determine a configuration for the equilibrium core for Tehran Research Reactor (TRR) at upgraded power level of 7.5 MW. Two different equilibrium core configurations have been concluded at this stage of design ...
The results of calculation of nuclear fuel burnup and independent fission product buildup from exp 235 U, exp 238 U, exp 239 Pu and exp 241 Pu in the WWER-440 reactor core are presented. For determining the physical performances of the WWER-440 reactor el...
High Burnup Fuel Behaviour has been growing in importance since middle 80's when pellet microstructure changes (rim effect) and cladding oxidation rates increase were observed. Later on, Cadarache reactivity tests revealed cladding integrity failures below safety limits. These phenomena, occurred at high burnup, stressed the necessity of having a ...
The results of a limited analysis of the extended SM-2 critical experiments are given. A review of the analytical models is presented to ascertain the accuracy of the reactivity calculations for the reference SMi-2 Core I. A new reference B-l0 loading was derived based on analysis and experimental measurements. Calculations ...
Aluminum-clad TRIGA fuel in the University of Arizona Mark I reactor was replaced with stainless steel-clad TRIGA fuel which had high burnup from previous use. Good knowledge of previous burnup was found to be important, in order that the fuel could be arranged to give enough excess reactivity for operation pulse mode with core ...
When calculating the burn-up one generally applies rough approximations, e.g., by calculating the time-dependent decrease of the reactivity of a fuel system under the provision of a neutron flux constant with regard to space and time and a space constant fuel distribution, disregarding the criticality conditions for the ...
A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE Burn-up has been performed. The objective of this research is to get optimal design parameters of such type reactors. The parameters of nuclear design including the critical condition, conversion ratio, and burn-up level were compared. These parameters are ...
NASA Astrophysics Data System (ADS)
Isotopic compositions and burnup of Core III fuel in the Saxton Reactor were determined. Measured isotopic buildups and depletions were compared with calculated values. (JWR)
In order to predict the behaviour of boiling water reactor (BWR) cores during longterm operation, reactor cycle calculations including control rods have to be used to analyze the fuel loadings with respect to their burnup behaviour and operation safety. S...
Fuel elements used in the ORR whole-core LEU fuel demonstration have been gamma-scanned to determine axial distributions of UZLa and TXCs fission product activities. This data has been analyzed to determine cycle-averaged fuel element powers, residual STVU masses, and burnups of discharged fuel elements. Methods used to analyze the data are discussed and ...
We present the results from experimental and calculated investigations of the effect the deformation of fuel assembly casings, fuel rod bundle, and single fuel rods in the cores of fast-neutron reactors has on their temperature conditions. It is shown that the distortion of a fuel assembly (FA) is determined to a considerable extent by temperature ...
A 300 kW small medical reactor was designed to be used for boron neutron capture therapy (BNCT) at KAIST in 1996. In this paper, analysis for the core life cycle of the original design of the BNCT facility and modifications of the fuel assembly configuration and enrichment to get a proper life cycle were performed and a criticality, neutron flux distribution and fuel ...
A 300 kW small medical reactor was designed to be used for boron neutron capture therapy (BNCT) at KAIST in 1996 [1]. In this paper, analysis for the core life cycle of the original design of the BNCT facility and modifications of the fuel assembly configuration and enrichment to get a proper life cycle were performed and a criticality, neutron flux distribution and fuel ...
Simplification of the pebble bed reactor by removing the unloading device from the system was peformed. For this reactor design, a suitable fuel-loading scheme is the Peu a Peu (little by little) fueling scheme. In the Peu a Peu modus, there is no unloading device; as such, the fuels are never discharged and remain at the bottom of the core during reactor operation. This means ...
Simplification of the pebble bed reactor by removing the unloading device from the system was peformed. For this reactor design, a suitable fuel-loading scheme is the Peu � Peu (little by little) fueling scheme. In the Peu � Peu modus, there is no unloading device; as such, the fuels are never discharged and remain at the bottom of the core during reactor operation. This ...
The progress in the Nuclear Engineering field leads to developing new generations of Nuclear Power Plants (NPP) with complex rector core designs, such as cores loaded partially with mixed-oxide (MOX) fuel, high burn-up loadings, and cores with advanced designs of fuel assemblies and control rods. Such heterogeneous ...
This report attempts to summarize and consolidate the existing knowledge on axial burnup distribution issues that are important to burnup credit criticality safety calculations. Recently released Nuclear Regulatory Commission (NRC) staff guidance permits ...
Data and functions of the cell burnup calculation of the SRAC system were revised to improve mainly the accuracy of the burnup calculation of high conversion light water reactors (HCLWRs). New burnup chain models were developed in order to treat fission p...
Waste packages are loaded with commercial spent nuclear fuel (SNF) that satisfies the minimum burnup requirements of a criticality loading curve. The burnup value assigned by the originating nuclear utility to each SNF assembly (assigned burnup) is used to load waste packages in compliance with a criticality loading curve. The ...
The experimental results of criticality benchmark experiments performed at the Jozef Stefan Institute TRIGA Mark II reactor are presented. The experiments were performed with partly burned fuel in two compact and uniform core configurations in the same arrangements as were used in the fresh fuel criticality benchmark experiment performed in 1991. In the experiments, both ...
This work describes a methodology to define the equilibrium core and a MTR (Materials Testing Reactors) type reactor's fuel management upon multiple boundary conditions, such as: end cycle and permitted maximum reactivities, burn-up extraction and maximun...
The RHOBURN program that determines various parameters for the Livermore Pool Type Reactor using a matrix of 840 cells, stored in two-dimensional form, is described. The code calculates the burnup of U/sup 235/on the basis of the megawatt days run during a fuel cycle and as a function of the neutron flux and fuel inventory in each of the 840 ...
The objective of this topical report is to present to the NRC for review and acceptance a methodology for using burnup credit in the design of criticality control systems for PWR spent fuel transportation packages, while maintaining the criticality safety margins and related requirements of 10 CFR Part 71 and 72. The proposed methodology consists of five major steps as ...
The efficient use of nodal methods for three-dimensional two-group reactor calculations requires homogenization over large volumes of nodes. This homogenization removes the internal structure of the nodes. On the other hand, accurate pinwise power distributions are indispensable for light water reactor design. A homogenization and dehomogenization procedure called the ...
The WIMS-D4 code has been modified (WIMS-D4m) to generate burn-up dependent microscopic cross sections for use in full core depletion calculations. The calculation of neutron absorption by fission products can be obtained from a reduced fission-product-chain model that includes the {sup 135}Xe and {sup 149}Sm ...
SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. It enables us to analyze the burnup problem using neutron spectrum depending on environment of ...
The HEU and LEU fuel elements used in the ORR whole-core demonstration were gamma-scanned to determine the axial distribution of the {sup 140}La and {sup 137}Cs activities. Analysis of this data is now complete. From the {sup 140}La activity distributions cycle-averaged powers were determined while the {sup 137}Cs data provided a measure of the final {sup 235}U ...
The calculation of coolant gas flow and heat transfer in cores of pebble- bed reactors requires special treatment of the fluid dynamics equations. Errors of the approximations and influence of errors of the input data of the calculation are estimated. It can be seen, that power density distribution has to be ...
A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup ...
Neutron and gamma-ray flux spectra are calculated using the MCNP5 computer code and a one-sixth core model of a prismatic Very High Temperature Reactor based on the General Atomics Gas Turbine-Modular Helium Reactor. Spectra are calculated in the five inner reflector graphite block rings, three annular active core ...
The enrichment of the Enrico Fermi core was determined by calculations based on critical experiments and was established at 25.6% contained U/sup 235/ for a total uranium contest of 1734 kg. This enrichment should ensure criticality for a minimum of 91 subassemblies assuming an equllibrium core for a ...
The combined iterative method is used to determine dependences of the fuel burnup and the actinide nuclide buildup in the RBMK reactor core on the coolant density dsub(Hsub(2)O). The calculations are conducted for the values dsub(Hsub(2)O)=0.68, 0.041 and...
A two-step procedure is developed for the optimization of in-core nuclear fuel management using perturbation theory to predict the effects of various core configurations. The first procedure is a cycle cost minimization using linear programming with a zoned core and discrete burnup groups. The second ...
A study was undertaken of power excursions in high burnup cores. There were three objectives in this study. One was to identify boiling water reactor (BWR) and pressurized water reactor (PWR) transients in which there is significant energy deposition in the fuel. Another was to analyze the response of BWRs to the rod drop accident (RDA) and other ...
This paper summarizes current progress and developments with the variational nodal method(VNM) and its implementaion within the DIF3D code suite. After a brief development of the mathematical basis for the VNM, results from two three-dimensional benchmarks are presented for a variety of computers. Then current applications of the VNM are discussed including diffusion theory ...
The power research reactors utilized MTR-type fuel elements containing up to 20$ enriched uranium. Calculation of Ksub(eff) have been performed for various fuel enrichments and burn up levels of up to 20% for a 10 MW MTR type research reactor specified by...
The WIMS-D4 code has been modified (WIMS-D4m) to generate burn-up dependent microscopic cross sections for use in full core depletion calculations. The calculation of neutron absorption by fission products can be obtained from a reduced fission-product-ch...
Design studies are performed for a commercial liquid-metal fast breeder reactor core that can achieve a burnup of 200 GWd/t. A plutonium-type asymmetric parfait core with two different plutonium-enriched zones in the axial direction as well as in the radial direction is studied. This core concept solves ...
During discharge of Core 2 from the Peach Bottom High-Temperature Gas-Cooled Reactor (HTGR), 55 driver elements, 21 test elements, three reflector elements, and one control rod with sleeve were axially gamma scanned with a high-resolution Ge(Li) detector. The purpose of the exercise was to determine fission product distributions for use in burnup ...
Calculations for long-term-disposal criticality safety of spent nuclear fuel requires the application of burnup credit because of the large mass of fissile material that will be present in the repository. Burnup credit calculations are based on depletion ...
The ORR Whole-Core LEU Fuel Demonstration, conducted as part of the US Reduced Enrichment Research and Test Reactor Program, has been successfully completed. Using commercially-fabricated U{sub 3}Si{sub 2}-Al 20%-enriched fuel elements (4.8 g U/cc) and fuel followers (3.5 g U/cc), the 30-MW Oak Ridge Research Reactor was safely converted from an all-HEU ...
Original design-basis cycle average burnup referenced for fuel discharged from Duke Power Company's Catawba and McGuire units in safety analysis calculations is 11 450 MWd/ton U. Due to changes in the fuel cycle that have developed since the initial design of these units, significant economic benefits can be realized by extending cycle ...
A general method (MCQ) has been developed by introducing a microscopic burnup scheme that uses the Monte Carlo calculated fluxes and microscopic reaction rates of a complex system and a depletion code for burnup calculations as a basis for solving nuclide material balance equations for each spatial region in which ...
Pressurized Water Reactors (PWR) are the most common nuclear reactor used today. The core of a PWR is composed of approximately 200 assemblies immersed in pressurized light water, which can be Uranium Oxyde assemblies (UOX) or Mixed Oxyde assemblies (MOX) coming from the reprocessings of used UOX. Electro-nuclear industries want to calculate the neutron ...
As an important part to validate the LWR core neutronics analysis methods, Japan Nuclear Energy Safety Organization (JNES) has been participating in the REBUS international program and performing analyses and evaluations of the reactor physics experiment data including the irradiated fuels. In REBUS program, physics experiments were performed at the VENUS critical test ...
This report describes DANDE - a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the ...
Unlike the fresh fuel approach, which assumes the initial isotopic compositions for criticality analyses, any burnup credit methodology must address the proper treatment of axial burnup distributions. A straightforward way of treating a given axial burnup distribution is to segment the fuel assembly into multiple meshes and to model ...
The predictions of APPR-1 core performance at zero power and for its operating life are reported, with a description of tbe methods employed. The window shade model has been used to predict a temperature coefficient of -0.5 x 10/sup -4/ DELTA k/ deg F at 68 deg F and -3.2 x 10/sup -4/ DELTA / deg F at 450 deg F. The lifetime based on unifarm material ...
Modular Pebble Bed Reactor (MPBR) is a high temperature gas-cooled nuclear power reactor currently under study as a next generation reactor system. In addition to its inherently safe design, a unique feature of this reactor is its multi-pass fuel circulation in which the fuel pebbles are randomly loaded and continuously cycled through the core until they reach their prescribed ...
Reactivity and fuel burnup studies were performed for a 255 Mw(e) sodium- graphite reactor of the advanced calandria core type. This reactor is briefly described. Initial criticality calculations and flux distributions were obtained, using two-group theory for enrichments between 2.0 at.% U/sup 325/ and 4.0 at.% U235. A ...
Basic data are given for the MGCR in terms of neutron temperature, macroscopic cross sections, two-group constants, burn-up, and fuel cycle costs. The nuclear properties of the reference design were determined by a study which covered a range of steel concentrations, voidage, and core size. Some of the more pertinent results of the reactor ...
In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product rele...
To date, criticality analyses performed in support of the certification of spent fuel casks in the United States do not take credit for the reactivity reduction that results from burnup. By taking credit for the fuel burnup, commonly referred to as ''burn...
In the AUS neutronics code system, the burnup module CHAR solves the nuclide depletion equations by an analytic technique in a number of spatial zones. CHAR is usually used as one component of a lattice burnup calculation but contains features which also ...
The changes due to neutron irradiation which occurred in the highly enriched uranium fuel slugs of the first core of the EBR-I were related to burnup (0 to 0.3 at.%) and to temperature (285 to 395 deg C). Initially, the slugs increased in length with increase in burnup, but at higher burnups the direction ...
Work performed under United States-Euratom Fast Reactor Exchange Program. ASB 71 is a two-dimensional burnup and management program in R-Z geometry for fast reactors. With a few exceptions it is coded in IBM Fortran and designed for computers of the IBM 360 or IBM 370 series (e.g. IBM 360/65 or IBM 370/165). The program can, however, be used for slower computers, ...
In general, neglecting the large intranodal cross-section gradients induced by depletion and feedback causes a nodal method to lose its accuracy in predicting nodal unknowns acceptably. Recently, in context of the analytic function expansion nodal (AFEN) method, Noh and Cho developed a rehomogenization burnup correction model that homogenizes the burnt fuel assembly using ...
Decreasing coke burnup in a coke dry quenching device is one method of improving the technicoeconomic indices of its functioning. Unfortunately, at present there is no standard method of calculating coke burnup. This prevents a thorough analysis to be made of the efficiency of the functioning of either individual dry coke quenching ...
The formulas and constants used in calculating the critical enrichment for a hot, clean, equilibrium xenonand samarium-poisoned 31-cell SRE core are presented. The results of this caiculation may be considered somewhat pessimistic since, where uncertanties arose, they were resolved to favor a higher enrichment, This is not to say, however, that ...
The changes in fuel and poison distributions and the changes in flux shape which accompany the consumption of fuel are studied. The technique employed is a perturbation calculation based on a one-velocity group treatment of the neutrons. The geometry is a spherical core surrounded by an infinite reflector. The programming forms for the IBM-650 ...
Simplified models of EBR-II control and safety rods have been developed for core modeling under various operational and shutdown conditions. A parametric study was performed on normal worth, high worth, and safety rod type control rods. A summary of worth changes due to individual modeling approximations is tabulated. Worth effects due to structural modeling simplification are ...
A fuel cycle program is presented for an intermediate power research reactor utilizing fully enriched MTR-type fuel elements. The fuel cycle program is considered at equilibrium after many cycles have past. The program consists of shifting elements from positions of high importance outward to positions of low importance through several paths. The paths are staggered so that only the ...
The dynamics of reactor operation result in nonuniform axial-burnup profiles in fuel with any significant burnup. At the beginning of life in a pressurized water reactor (PWR), a near-cosine axial-shaped flux will begin depleting fuel near the axial center of a fuel assembly at a greater rate than at the ends. As the reactor continues to operate, the ...
Safety analyses of criticality control systems for transportation packages include an assumption that the spent nuclear fuel (SNF) loaded into the package is fresh or unirradiated. In other words, the spent fuel is assumed to have its original, as-manufactured U-235 isotopic content. The ``fresh fuel`` assumption is very conservative since the potential reactivity of the nuclear fuel is ...
A comparative analysis and comparison of results obtained between 2-D lattice calculations and 3-D full core nodal calculations, in the frame of MOX fuel design, was conducted. This study revealed a set of advantages and disadvantages, with respect to each method, which can be used to guide the level of accuracy desired for future fuel ...
Fuel management requires that mass, energy, and reactivity balance be satisfied in each reload cycle. Procedures for selection of alternatives, core-state models, and fuel cost calculations have been developed for both equilibrium and transition cycles. Effective cycle lengths and fuel cycle variables--namely, reload batch size, schedule of incore ...
The calculation of the long-term reactivity of reactors is ondy possible with great mathematical efforts; since the burn-up (including conversion and poisoning) is dependent on the neutron flux and on the concentration of the fuel components, which are both variable with regard to time and space, and the fuel charging required for an operation with ...
This method is based on all burn-up depending data, namely particle densities and neutron spectra, being available in a burn-up library. This one is created by means of a small number of cell burn-up calculations which can easily be carried out and in whi...
The purpose of this paper is to discuss the theories, techniques and computer codes that are frequently used in numerical reactor criticality and burnup calculations. It is a part of an integrated nuclear reactor calculation scheme conducted by the Reacto...
Fort St. Vrain (FSV) fuel element 1-0743 was irradiated in core location 17.04.F.06 from July 3, 1976 until February 1, 1979. The element experienced an average fast neutron exposure of about 0.95 x 10/sup 25/ n/m/sup 2/ (E > 29 fJ)/sub HTGR/, a time-and-volume-averaged fuel temperature in the vicinity of 680/sup 0/C, fissile and fertile particle ...
As a natural outgrowth of improved nodal calculation methods and the accessibility of detailed fuel assembly operating data from core monitoring systems, taking credit for burnup in the storage of light water reactor fuel represents a logical alternative to reracking for storing higher enriched fuel. The paper summarizes a number of ...
For the design of a core loaded with mixed-oxide (MOX) fuel, it is frequently necessary to determine the plutonium content in an MOX fuel assembly that is equivalent in burnup to a conventional UO{sub 2} fuel assembly. The equivalent plutonium content can be determined experimentally based on the linear reactivity model (LRM) through a number of assembly ...
The RINSC is a 2 mega-watt, light water and graphite moderated and cooled reactor that has a graphite thermal column built as a user facility for sample irradiation. Over the past decade, after the reactor conversion from a highly-enriched uranium core to a low-enriched one, flux and dose measurements and calculations had been performed in the thermal ...
Six groups of U-2 wt% Zr fuel slugs were irradiated in the first core of the EBR-I to burnups of 0.080 to 0.189 at.% at calculated temperatures of 307 to 353 deg C. Two groups of cast specimens were found to be more dimensionally stable than four groups of wrought slugs. Of the wrought slungs, the as quenched group showed less ...
The primary objective of this report is to provide the technical basis for development of recommendations for updates to the NUREG-1465 Source Term for BWRs that will extend its applicability to accidents involving high burnup (HBU) cores. However, a secondary objective is to re-examine the fundamental characteristics of the prescription for fission ...
A neutronic analysis is presented of three incinerator subcritical lattices, driven by accelerated protons and designed to transmute the minor actinides, the {sup 99}Tc and the {sup 129}I, of light water reactor (LWR) waste. A calculational methodology must first be established to enable a neutronic burnup analysis of fission cores ...
The GRR-1 is a 5-MW pool-type, light-water-moderated and-cooled reactor fueled with MTR-type fuel elements. Recently received Be reflector blocks will soon be added to the core to add additional reactivity until fresh LEU fuel arrives. REBUS-3 xy fuel cycle analyses, using burnup dependent cross sections, were performed to assist in fuel management ...
Results are presented of a study concerning EBWR core designs in which uniformly enriched UC/sub 2/ fuel rods clad with Zr, capable of a maximum exposure at about 10,000 Mwd/t were used. Three core designs are described and nuclear constants and burnups characteristics are given. The third design (C), embodying improvements ...
In the present paper, the current status of covariance applications to fast reactor analysis and design is summarized with actual examples. The covariance applications are classified into three fields. First, covariances are used to quantify the uncertainty of nuclear core parameters such as criticality, control rod worth, reaction rate ratio, power distribution, sodium void ...
Monte Carlo calculations of a criticality experiment with burned fuel on the TRIGA Mark II research reactor are presented. The main objective was to incorporate burned fuel composition calculated with the WIMSD4 deterministic code into the MCNP4B Monte Carlo code and compare the calculated k{sub eff} with the measurements. The ...
Some procedures now being used for burnup calculations make use of a neutnon balance equation as a criticality relation for exploring reactivity during reactor life. This balance makes use of the resonance escape probability calculated for the core composition as an infinite medium as the effective ...
A detailed evaluation of the fuel-burnup dependent power distribution and the scram reactivity for the PIUS reactor design has been performed. The analyses were carried out using the CPM lattice physics and NODE-P2 core neutronics/thermal-hydraulics codes, and are based on the information provided in the PIUS Preliminary Safety Information Document. Cycle ...
Operations and material control and accountancy requirements for the Fuel Conditioning Facility demand accurate prediction of the mass flow of spent EBR-II driver fuel into the facility. This requires validated calculational tools that can predict the burnup and isotopic distribution in irradiated Zr-alloy fueled driver assemblies. Detailed ...
Uncertainties in the predicted isotopic concentrations in spent nuclear fuel represent one of the largest sources of overall uncertainty in criticality calculations that use burnup credit. The methods used to propagate the uncertainties in the calculated ...
Burnup calculations were made on a lattice model for high conversion light water reactor (HCLWR) in order to assess the feasibility of HCLWR concept and to obtain reference data on considering its neutronic characteristics. In these calculations, lattice ...
A general formulation of loading calculations and their transformation to a system of partial differential equations are shown. A generalization is made for loading in transiently changing power. The neglect of burn-up dependence on radius is considered. Solution methods with an analog computer are given, and examples are presented. (J.S.R.)
The method for calculating the nuclear fuel burnup which partially takes into account a neutron energy spectrum variation in the course of the solution of temporal equations of nuclide production kinetics is suggested. By means of this method the various ...
The calculation of burnup in mixed-oxide fuel pins irradiated in EBR-II is shown to agree with burnup measured by postirradiation radiochemical analysis. The mean percent deviation is 0.12% with a variance of 2.49%. This gives a high level of confidence that HIST, which uses 91% of the nominal EBR-II power, provides an accurate ...
Perturbation theory is developed for the nonlinear burnup equations describing the time-dependent behavior of the neutron and nuclide fields in a reactor core. General aspects of adjoint equations for nonliner systems are first discussed and then various approximations to the burnup equations are rigorously derived and their areas for ...
The determination of nuclear fuel burnup by the non-destructive gamma spectroscopy method is studied. A MTR (Materials Testing Reactor) -type fuel element is used in the measurement. The fuel element was removed from the reactor core in 1958 and, because ...
A study of the influence of {sup 135}Xe on the prompt negative temperature coefficient of the 14-MW Romanian TRIGA reactor has been performed. Because of its large absorption cross section below 0.1 eV, we expected that {sup 135}Xe might make a positive contribution to the temperature coefficient because the higher-energy neutrons are less likely to be absorbed by the Xe. This effect would be ...
An equation for the computation of fuel burnup in the Materials Testing Reactor core is derived. This expression relates total loss (W) of U/sup 235/ by both the (n,f) and (n, gamma ) reaction to fission energy (E), relative cross section ( sigma c/of), and total energy dissipated to cooling water during the period concerned (P) as reflected by ...
This paper summarizes the operating cores, the /sup 235/U fuel element burnups, and the core physics measurements made in the ORR Whole-Core LEU Fuel Demonstration.
Calculations were made in connection with xenon and samarium transients in the HFIR, including associated variations in the power distribution. Of particular interest was the possibility of burning samarium out of a core that was shutdown for a prolonged period of time after eleven or more days of operation at full power. The results indicate ...
The excess reactivity designed to be built into a reactor core is a primary determinant of the fissile loadings of the fuel rods in the core. For the TREAT Upgrade (TU) reactor the considerations that enter into the determination of the excess reactivity are different from those of conventional power reactors. The reactor is designed to operate in an ...
The fast-neutron capture cross sections of Pu-240, 241, 242 and Am-241 are reviewed. These nuclides are important to core physics of reactors that contain Pu-239. There have been several significant measurements of these cross sections in recent years. These measurements were instigated by the need for these cross sections for reactor calculations ...
The reactor analysis of the critical experiment and thc APPR-I resulted in a loading of the APPR-1 of 2I.098 gm of B-I0 with 22.5 kg U-235. This loading will rcsult ln adequate reactlvity for a core life of I3 MWYR bascd on uniform burnup of U-235 and B-10. Calculations iindicate that five of the seven control rods provided ...
This paper reports on a data base of the TRIGAP code that is generated for the 3-MW TRIGA MARK-II research reactor in Bangladesh. The library is created using the WIMS-D/4 code. Cross sections are calculated from zero burnup to 37% of initial {sup 235}U in 20 burnup steps. The created TRIGAP library is tested through practical ...
One responsibility of the Supervisor of Reactor Operations is to follow the TRIGA core depletion and recommend core loading changes for refueling and special experiments. Calculations required to analyze such changes normally use digital computers and are extremely difficult to perform for one who is not familiar with computer language ...
Significant research is currently being performed whereby fast reactor cores have been designed to burn transuranic materials reducing the volume and long-term radiotoxicity of spent nuclear fuel. These core and depletion models depend on various computer codes. This research used MCNPX 2.6.0 and ERANOS 2.1 to model a standard 250MWt Advanced Burner Test ...
This report discusses research conducted to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B W) design (Oconee) and a Westinghouse (W) ...
A study of high-burnup pressurized water reactor (PWR) fuel management schemes extending to 100 GWd/tonne is presented. The Studsvik Scandpower code suite was used to model a Westinghouse three-loop PWR core, and realistic loading patterns were derived. The loading patterns were optimized for minimum power peaking and maximum cycle length using engineering ...
The MARIA reactor at the Institute of Atomic Energy (IAE) in Swierk (30 km SE of Warsaw) in the Republic of Poland is considering conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel assemblies (FA). The FA design in MARIA is rather unique; a suitable LEU FA has never been designed or tested. IAE has contracted with CERCA (the fuel supply portion of AREVA in France) to ...
In this paper the accuracy of the nuclear design code system for the High-Temperature Engineering Test Reactor (HTTR) is evaluated for the neutronic characteristics that depend on core temperature by analyzing the overall temperature coefficients of reactivity and the effective multiplication factors obtained by an experiment in which the Very High Temperature Reactor Critical ...
A methodology for calculating inventories for the NBSR has been developed using the MCNPX computer code with the BURN option. A major advantage of the present methodology over the previous methodology, where MONTEBURNS and MCNP5 were used, is that more materials can be included in the model. The NBSR has 30 fuel elements each with a 17.8 cm (7 in) gap in the middle of the ...
This paper concerns two independent safety investigations on critical and sub-critical heavy liquid metal cooled fast reactors using simple flow paths. The first investigation applies to locating the steam generators in the risers instead of the down-comers of a simple flow path designed sub-critical reactor of 600 MW{sub th} power. This was compared to a similar design, but with the steam ...
Depletion calculations have been performed for the McClellan reactor history from January 1990 through August 1996. A database has been generated for continuing use by operations personnel which contains the isotopic inventory for all fuel elements and fuel-followed control rods maintained at McClellan. The calculations are based on the three-dimensional ...
Recently, a number of low void worth core designs with non-conventional core geometries have been proposed. Since these designs lack a good experimental and computational database, benchmark calculations are useful for the identification of possible biases in performance characteristics predictions. In this paper, a simplified ...
By means of the combined iterative method the nuclear fuel burnup in the WWER-1000 reactor is calculated. The fission product buildup separately for exp 235 U, exp 238 U, exp 239 Pu and exp 241 Pu is determined. The burnup dependences for exp 235 U, exp 2...
The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor fuel. These transuranic nuclides are incorporated into TRISO coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of ...
The main subjects of the meeting were the discussion of regulatory background, integral tests and analysis, plant calculations, separate-effect test and analysis, concerning high burnup phenomena during RIA accidents in reactors, especially LWR, BWR and P...
A problem of reliability of information on burn-up distribution in a WWR-M nuclear core is considered. A method to specify this information of the minimum 12 reactor reloadings is made and, as a result, sufficient reliable data of burn-up values are obtai...
The physics analysis (including burnup effects and an approximation of the void-distribution effects) of the axial power distribution of a core, with the upper 5 feet having a higher U/sup 235/ enrichment than the lower plus or minus 1/2 feet, was completed. A physics analysis (including burnup and voiddistribution effects) ...
In order to address the energy needs of developing countries and remote communities, Oregon State University has proposed the Multi-Application Small Light Water Reactor (MASLWR) design. In order to achieve five years of operation without refueling, use of 8% enriched fuel is necessary. This dissertation is focused on core design issues related with increased fuel enrichment ...
Parametric burnup calculations are performed to estimate radionuclide isotopic mass and activity concentrations for four different Training, Research, and Isotope General Atomics (TRIGA) nuclear reactor fuel element types: (1) Aluminum-clad standard, (2) Stainless Steel-clad standard, (3) High-enrichment Fuel Life Improvement Program (FLIP), and (4) ...
Gross gamma-ray dose rates from six spent TRIGA fuel elements were measured and compared to calculated values as a means to validate the reported element burnups. A newly installed and functional gamma-ray detection subsystem of the In-Cell Examination System was used to perform the measurements and is described in some detail. The analytical methodology ...
The ANSI/ANS 8.1 criticality safety standard recommends validation and benchmarking of the calculational methods used in evaluating criticality safety limits for away-from-reactor applications. The lack of critical experiments with burned light-water reactor (LWR) fuel in racks or in casks necessitates the validation of burnup credit methods by comparison ...
Nuclear core analysis has been performed for thorium core as an activity of IAEA CRP. The reactivity change with burnup of thorium fuel strongly depends on the nuclear property and content of seed material in the fuel. Lower contents of plutonium in (TH+P...
The specific experimental measurements of interest here involve the determination of parameters related to the actinide and fission product composition of samples from five elements taken from fuel assemblies discharged from the Turkey Point Unit 3 PWR. Two fuel assemblies were obtained for the purposes of nondestructive and destructive assay. These assemblies were initially fueled with 448 kg of ...
Detailed core management arrangements are developed requiring four operating cycles for the transition from present three-batch loading to an extended burnup four-batch plan for Zion-1. The ARMP code EPRI-NODE-P was used for core modeling. Although this w...
The spectral history problem encountered in reconstructing local homogeneous power distributions is investigated. Because of difficulties in most nodal codes concerning spectral interactions between neighboring assemblies when rebuilding the local power distribution, nodal codes assume a constant spectrum or do not properly consider local spectrum variations within an assembly. A simple, ...
A perturbation formulation is developed for the space-energy dependent burnup equations describing depletion and transmutation of nuclide densities in a coupled neutron-nuclide field, such as a reactor core. The formulation is developed in a form consistent with the computational methods used for depletion analysis. The analysis technique currently ...
A perturbation formulation is developed for the space-energy-dependent burnup equations describing depletion and transmutation of nuclide densities in a coupled neutron-nuclide field, such as a reactor core. The formulation is developed in a form consistent with the computational methods used for depletion analysis. The analysis technique currently ...
To simulate fast-reactor spectra, a simple shape function f(u; u/sub m/ ,B) with spectral parameters u/sub m/ and B has been derived by expanding the gross shape of the spectrnm into a series of Hermite functions. The parameters are determined by the moment method. It is shown through some sample problems that the shape function is applicable to both core and blanket. ...
This paper reports on an expert system for generating control rod patterns that has been developed. The knowledge is transformed into IF-THEN rules. The inference engine uses the Rete pattern matching algorithm to match facts, and rule premises and conflict resolution strategies to make the system function intelligently. A forward-chaining mechanism is adopted in the inference engine. The system ...
However, because of the miniscule burnup of enriched uranium-235 during the Earth departure phase (~10 grams out of 33 kilograms in each NTR core), ...
NASA Website
Mar 1, 2011 ... Abstract: Burnup and poisoning in nuclear reactor core and validity ... noise generated on the fuselage skin by a turbulent boundary layer. ...
Response of high-burnup prototype Consolidated Edison power-reactor fuel pins to the Darex-Thorex and SulfexThorex head-end processes was investigated on a laboratory scale. Losses of uranium to 6 M H/sub 2/SO/sub 4/(Sulfex) decladding solutions were in the 0.05 to 0.15% range, for burnups of from about 1000 to 20000 Mwd/t of ...
A whole-core demonstration of LEU fuel in the ORR is expected to begin during November 1985. Fuel elements will contain U/sub 3/Si/sub 2/ at 4.8 Mg U/m/sup 3/ and shim rod fuel followers will contain U/sub 3/Si/sub 2/ at 3.5 Mg U/m/sup 3/. Fuel fabrication is underway at B and W, CERCA, and NUKEM, with shipments scheduled to commence in October. The primary objectives of the ...
The Babcock and Wilcox integrated nuclear design system is divided into three broad areas: (1) BASIC NUCLEAR DATA PROCESSING-Basic nuclear data are collected, evaluated, and processed into a specified fine-energy mesh multigroup data file called a Master Library. (2) APPLICATIONS DATA PROCESSING-Data for selected materials are retrieved from the Master Library and processed into an optimally ...
An innovative concept of PFPWR50 has been studied, which is a small PWR using carbon-coated particle fuels with cladding. The thermal power of PFPWR50 is 50 MW and it provides district-heating service as long period as possible. The particle fuel has been used in a HTGR that uses graphite as moderator, but PFPWR50 uses both pressurized light water and graphite as moderator. This concept takes ...
The Working Party on Evaluation Cooperation (WPEC) of the OECD Nuclear Energy Agency Nuclear Science Committee has established an International Subgroup to perform an activity in order to develop a systematic approach to define data needs for Gen-IV and, in general, for advanced reactor systems. A methodology, based on sensitivity analysis has been agreed and representative ...
The ANSI/ANS 8.1 criticality safety standard recommends validation and benchmarking of the calculational methods used in evaluating criticality safety limits for away-from-reactor applications. The lack of critical experiments with burned light-water-reactor (LWR) fuel in racks or in casks necessitates the validation of burnup credit methods by comparison ...
Consideration of the depletion phenomena and isotopic uncertainties in burnup-credit criticality analysis places an increasing reliance on computational tools and significantly increases the overall complexity of the calculations. An automated analysis and data management capability is essential for practical implementation of large-scale ...
monteburns, a burnup computer code that uses the Monte Carlo technique, was developed at Los Alamos National Laboratory to be applied to a variety of nuclear design calculations (see accompanying paper on the development of monteburns). It is a fully automated burnup code that incorporates multiple irradiation steps and many other ...
This paper addresses two of the measurement specifications used in analyzing spent fuel packages to gain burnup credit. The philosophy and calculation of rejection criteria and measurement accuracy are discussed. Any assembly for which the declared measured value and reactor record value deviate by more than 10% will be rejected. Measurement accuracy ...
Fourteen irradiated assemblies were analyzed using nondestructive high resolution gamma spectrometry (HRGS). Measured and calculated (on the basis of calorimetric data) axial burnup profiles and average burnup values were compared. The measurements of spe...
The CPM collision probability code, which is a part of EPRI's Advanced Recycle Methodology Program (ARMP), was used to calculate end-of-cycle isotopics for a ThO/sub 2/-UO/sub 2/ pressurized water reactor core. Calculated isotopic concentrations are compared with measured concentrations for 13 samples from the first ...
This study investigated the neutronic characteristics of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR), a novel nuclear reactor concept that combines liquid salt (7LiF-BeF2---flibe) cooling and TRISO coated-particle fuel technology. The use of flibe enables operation at high power density and atmospheric pressure and improves passive decay-heat removal capabilities, but flibe, unlike ...
An assessment is made of the potential for Th-based fuel to minimize Pu and minor actinide (MA) production in pressurized water reactors (PWRs). Destruction rates and residual amounts of Pu and MA in the fuel used for transmutation are examined. In particular, sensitivity of these two parameters to the fuel lattice hydrogen to heavy metal (H/HM) ratio and to the fuel composition was systematically ...
Several computer codes have been developed to perform nuclear burnup calculations over the past few decades. In addition, because of advances in computer technology, it recently has become more desirable to use Monte Carlo techniques for such problems. Monte Carlo techniques generally offer two distinct advantages over discrete ordinate methods: (1) the ...
In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular ...
The Pebble Bed Reactor (PBR) is a helium-cooled, graphite-moderated high temperature nuclear power reactor. In addition to its inherently safe design, a unique feature of this reactor is its multipass fuel cycle in which graphite fuel pebbles (of varying enrichment) are randomly loaded and continuously circulated through the core until they reach their prescribed end-of-life ...
The motivation for taking credit for fuel assembly burnup in spent-fuel storage rack design is obvious. Several pressurized water reactor facilities, beginning with the standardized nuclear unit power plant system plants (Callaway and Wolf Creek) have been licensed to install racks that take credit for burnup. Designing racks to take credit for ...
Actinide cross section data at thermonuclear neutron energies are needed for the calculation of ICF pellet center burnup of fission reactor waste, viz. 14 MeV neutron fission of the very long-lived actinides that pose storage problems. A major advantage of pellet center burnup is safety: only milligrams of highly toxic and active ...
To date, criticality analyses performed in support of the certification of spent fuel casks in the United States do not take credit for the reactivity reduction that results from burnup. By taking credit for the fuel burnup, commonly referred to as burnup credit,'' the fuel loading capacity of these casks can be ...
A new reactor burnup strategy CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) was proposed, where shapes of neutron flux, nuclide densities and power density distributions remain constant but move upward (or downward) along its core axis. This burnup strategy can ...
The RINSC is a 2 mega-watt, light water cooled and graphite moderated reactor that has a graphite thermal column built as a user facility for sample irradiation. Over the past decade, after the reactor conversion from a highly-enriched uranium core to a low-enriched one, flux and dose measurements and calculations had been performed in the thermal column ...
The method and the results are reported of numerical calculations of fuel burning in ordinary water-cooled and -moderated reactors with a homogeneous core which use fresh slightly enriched fuel. The point of departure of the method is the notion of stationary conditions at which the process is maintained in the reactor as a result of a regular ...
Thermal and biological shielding requirements are reviewed. A reevaluation of the delayed neutron fraction was made to include those neutrons produced in the blanket. Calcultions of the decay heating due to heavy isotopes have been completed. The maximum permissible concentration in air of a mixture of fission gases was calculated to be 6 x 10/sup -7/ mu c/cc. Ten ...
An oxide version of the DART code has been generated in order to assess the irradiation behavior of UO{sub 2}-Al dispersion fuel. The aluminum-fuel interaction models were developed based on U{sub 3}O{sub 8}-Al irradiation data. Deformation of the fuel element occurs due to fuel particle swelling driven by both solid and gaseous fission products, as well as a consequence of the interaction between ...
The high power density fuel assembly average burnup reached 4,500 Mwd/t during the period. In a study of the performance in VBWR and the Big Rock Reactor of promising fuel assemblies efforts were devoted to two-pass swaged fuel rods, fuel specimen testing in the Trail Cable facility, two-pass swaged sintered UO/sub 2/ from ADU, and isostatic pressing of ...
The 14-MW TRIGA steady state reactor (SSR) located in Pitesti, Romania, first went critical in the fall of 1979. Initially, the core configuration for full power operation used 29 fuel clusters each containing a 5 {times} 5 square array of HEU (10 wt%) -- ZrH -- Er (2.8 wt%) fuel-moderator rods (1.295 cm o.d.) clad in Incology. With a total inventory of 35 HEU fuel clusters, ...
The codes WIMSD/4 and BORGES--part of the MTR-PC code package--have been applied to prepare the microscopic cross-section library for the main elements of miniature neutron source reactor (MNSR) core for six neutron energy groups. The generated library has been utilized by the 3D code CITATION to perform the calculation of fuel burn-up ...
PubMed
A proliferation resistant hexagonal tight lattice BWR fueled core for increased burnup and reduced fuel storage requirements. Annual progress report: August, 1999 to July, 2000 (DOE NERI)
The Deep-Burn (DB) concept [ ] focuses on the destruction of transuranic nuclides from used light water reactor (LWR) fuel. These transuranic nuclides are incorporated into tri-isotopic (TRISO) coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected ...
The Advanced Test Reactor (ATR) is a high power density, high neutron flux research reactor operating in the United States. The ATR has large irradiation test volumes located in high flux areas. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth with a maximum unperturbed thermal neutron flux rating of 1.0 x 1015 n/cm2�s. As a result, the ATR is a ...
Noise measurements were performed during the first and second fuel cycles of the Sequoyah-1 pressurized water reactor (PWR) to observe long-term changes in the ex-core neutron signatures. Increases in the ex-core neutron noise amplitude were observed throughout the 0.1- to 50.0-Hz range. In-core noise measurements indicate that fuel ...
DOE has submitted a topical report to the NRC justifying burnup credit based only on actinide isotopes (U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241). When this topical report is approved, it will allow a great deal of the commercial spent nuclear fuel to be transported in significantly higher capacity casks. A cost savings estimate for ...
The reactor EWA-10 has been working at 8 MW power since 1967. In order to determine the maximum admissible reactor power and the corresponding safety margin, the measurements of fuel cladding temperatures in steady state operation of the reactor were performed with an instrumented fuel assembly inserted into various positions in the core. The hot spots were found and ...
Rhodium detectors are used in many commercial pressurized water reactors PWRs (pressurized water reactor) as in-core neutron detectors. The signals from the detectors are the result of neutron absorption in {sup 103}Rh and the subsequent beta decay of {sup 104}Rh to {sup 104}Pd. The rhodium depletes {approximately}1% per full-power month, so corrections are necessary to the ...
A comparative neutronic study has been conducted on several LMR cores using both lead-magnesium eutectic alloy (97.7% Pb -- 2.3% Mg) and sodium as coolant. In order to have a consistent comparison for these two coolants on a common basis, i.e. interchangeable designs, this study used exactly the same reactor core layout, assembly design parameters, and ...
An equilibrium cycle method, embodied in the REBUS-3[1] code system, has generally been used in conventional fast reactor design activities. The equilibrium cycle method provides an efficient approach for modeling reactor system, compared to the more traditional non-equilibrium cycle fuel management calculation approach. Recently, the equilibrium analysis method has been ...
The purpose of this study is to provide insights into the neutronic similarities that may exist between a generic cask containing typical spent nuclear fuel assemblies and commercial reactor critical (CRC) state-points. Forty CRC state-points from five pressurized-water reactors were selected for the study and the type of CRC state-points that may be applicable for validation of ...
Aiming of effective utilization of plutonium, an idea of a very high burnup light water reactor (VHBLWR) core was proposed, and new concept of the fuel cycle, in which the breeding of TRU nuclides will be restrained (TRU Enclosing Fuel Cycle), was discuss...
Representative accident source terms patterned after the NUREG-1465 Source Term have been developed for high burnup fuel in BWRs and PWRs and for MOX fuel in a PWR with an ice-condenser containment. These source terms have been derived using nonparametric order statistics to develop distributions for the timing of radionuclide release during four accident phases and for ...
Fuel depletion analysis is concerned with evaluating fuel composition changes as a result of the depletion and buildup of fuel isotopic materials, and decay of fission products during reactor operation. These composition changes lead to variation of reactivity, breeding, and power distribution. Nodal techniques are being developed to treat power distribution calculation ...
The requirements of ANSI/ANS-8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit is to be taken for the reduced reactivity of burned or spent fuel relative to its original ''fresh'' composition, it is necessary to benchmark computational methods ...
In this paper, two 600 MW{sub e} reactors are compared regarding safety relevant reactivity coefficients, waste-burning capabilities and reactivity swings during burn-up. Furthermore, comparisons of unprotected Loss-of-Flow and Loss-of-Heat Sink calculations are presented. In the first part of this paper, oxide fuels with an inert {sup 92}Mo matrix ...
This report discusses research conducted to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B&W) design (Oconee) and a Westinghouse (W) ...
Spent pressurized water reactor (PWR) fuel is directly reused in the Canada deuterium uranium (CANDU) reactor. The direct use of spent fuel in the DUPIC fuel cycle can be difficult because the spent PWR fuel composition depends on the enrichment and burnup history, so the DUPIC fuel composition will change accordingly. To reduce the composition heterogeneity effect on ...
Space-dependent reactor kinetics calculations were performed to interpret observed increases in the amplitude of pressurized water reactor (PWR), ex-core neutron detector noise with increasing fuel burnup and correspondingly decreasing soluble boron concentration. These noise amplitude increases have occurred at both low frequencies ...
A technique has been developed for verification of light-water reactor operating history from measurements of irradiated fuel assemblies. The Fork detector is used to measure neutron and gross gamma-ray emissions from fuel assemblies. The measurements can be performed a few days after discharge or up to several years later. The neutron and gamma-ray ratios are used to check the consistency of the ...
A method has been developed to estimate the irradiation history and burnup of a spent fuel by gamma-ray spectroscopy. The gamma-ray spectrum, measured by using a Ge(Li) detector, is analyzed by the standard spectrum method to obtain the activity of the fission product. The irradiation history is fitted by the least-squares method to reproduce the activity of each ...
The reactivity transient following a control rod step strongly depends on quantities that determine the thermal reactivity feedback. For the special case of a pressurized water reactor, these quantities are the reactivity temperature coefficients and the heat transfer between fuel and coolant. Therefore, it is possible to determine these quantities by fitting appropriate model ...
