WorldWideScience

Sample records for burnup

  1. Calibration of burnup monitor installed in Rokkasho Reprocessing Plant

    Energy Technology Data Exchange (ETDEWEB)

    Oeda, Kaoru; Naito, Hirofumi; Hirota, Masanari [Japan Nuclear Fuel Co. Ltd., Rokkasho, Aomori (Japan); Natsume, Koichiro [Isogo Engineering Center, Toshiba Corporation, Yokohama, Kanagawa (Japan); Kumanomido, Hironori [Nuclear Engineering Laboratory, Toshiba Corporation, Kawasaki, Kanagawa (Japan)

    2000-06-01

    Rokkasho Reprocessing Plant uses burnup credit for criticality control at the Spent Fuel Storage Facility (SFSF) and the Dissolution Facility. A burnup monitor measures nondestructively burnup value of a spent fuel assembly and guarantees the credit for burnup. For practical reasons, a standard radiation source is not used in calibration of the burnup monitor, but the burnup values of many spent fuel assemblies are measured based on operator-declared burnup values. This paper describes the concept of burnup credit, the burnup monitor, and the calibration method. It is concluded, from the results of calibration tests, that the calibration method is valid. (author)

  2. Analysis of high burnup fuel safety issues

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bock; Kim, D. H.; Bang, J. G.; Kim, Y. M.; Yang, Y. S.; Jung, Y. H.; Jeong, Y. H.; Nam, C.; Baik, J. H.; Song, K. W.; Kim, K. S

    2000-12-01

    Safety issues in steady state and transient behavior of high burnup LWR fuel above 50 - 60 MWD/kgU were analyzed. Effects of burnup extension upon fuel performance parameters was reviewed, and validity of both the fuel safety criteria and the performance analysis models which were based upon the lower burnup fuel test results was analyzed. It was found that further tests would be necessary in such areas as fuel failure and dispersion for RIA, and high temperature cladding corrosion and mechanical deformation for LOCA. Since domestic fuels have been irradiated in PWR up to burnup higher than 55 MWD/kgU-rod. avg., it can be said that Korea is in the same situation as the other countries in the high burnup fuel safety issues. Therefore, necessary research areas to be performed in Korea were derived. Considering that post-irradiation examination(PIE) for the domestic fuel of burnup higher than 30 MWD/kgU has not been done so far at all, it is primarily necessary to perform PIE for high burnup fuel, and then simulation tests for RIA and LOCA could be performed by using high burnup fuel specimens. For the areas which can not be performed in Korea, international cooperation will be helpful to obtain the test results. With those data base, safety of high burnup domestic fuels will be confirmed, current fuel safety criteria will be re-evaluated, and finally transient high burnup fuel behavior analysis technology will be developed through the fuel performance analysis code development.

  3. Integrated burnup calculation code system SWAT

    Energy Technology Data Exchange (ETDEWEB)

    Suyama, Kenya; Hirakawa, Naohiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Iwasaki, Tomohiko

    1997-11-01

    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 irradiation, combining SRAC which is Japanese standard thermal reactor analysis code system and ORIGEN2 which is burnup code widely used all over the world. SWAT makes effective cross section library based on results by SRAC, and performs the burnup analysis with ORIGEN2 using that library. SRAC and ORIGEN2 can be called as external module. SWAT has original cross section library on based JENDL-3.2 and libraries of fission yield and decay data prepared from JNDC FP Library second version. Using these libraries, user can use latest data in the calculation of SWAT besides the effective cross section prepared by SRAC. Also, User can make original ORIGEN2 library using the output file of SWAT. This report presents concept and user`s manual of SWAT. (author)

  4. ISOTOPIC MODEL FOR COMMERCIAL SNF BURNUP CREDIT

    Energy Technology Data Exchange (ETDEWEB)

    A.H. Wells

    2004-11-17

    The purpose of this report is to demonstrate a process for selecting bounding depletion parameters, show that they are conservative for pressurized water reactor (PWR) and boiling water reactor (BWR) spent nuclear fuel (SNF), and establish the range of burnup for which the parameters are conservative. The general range of applicability is for commercial light water reactor (LWR) SNF with initial enrichments between 2.0 and 5.0 weight percent {sup 235}U and burnups between 10 and 50 gigawatt-day per metric ton of uranium (GWd/MTU).

  5. A PWR Thorium Pin Cell Burnup Benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, Kevan Dean; Zhao, X.; Pilat, E. E; Hejzlar, P.

    2000-05-01

    As part of work to evaluate the potential benefits of using thorium in LWR fuel, a thorium fueled benchmark comparison was made in this study between state-of-the-art codes, MOCUP (MCNP4B + ORIGEN2), and CASMO-4 for burnup calculations. The MOCUP runs were done individually at MIT and INEEL, using the same model but with some differences in techniques and cross section libraries. Eigenvalue and isotope concentrations were compared on a PWR pin cell model up to high burnup. The eigenvalue comparison as a function of burnup is good: the maximum difference is within 2% and the average absolute difference less than 1%. The isotope concentration comparisons are better than a set of MOX fuel benchmarks and comparable to a set of uranium fuel benchmarks reported in the literature. The actinide and fission product data sources used in the MOCUP burnup calculations for a typical thorium fuel are documented. Reasons for code vs code differences are analyzed and discussed.

  6. High burnup effects in WWER fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Smirnov, V.; Smirnov, A. [RRC Research Institute of Atomic Reactors, Dimitrovqrad (Russian Federation)

    1996-03-01

    Since 1987 at the Research Institute of Atomic Reactors, the examinations of the WWER spent fuel assemblies has been carried out. These investigations are aimed to gain information on WWER spent fuel conditions in order to validate the fuel assemblies use during the 3 and 4 year fuel cycle in the WWER-440 and WWER-1000 units. At present time, the aim is to reach an average fuel burnup of 55 MWd/kgU. According to this aim, a new investigation program on the WWER spent fuel elements is started. The main objectives of this program are to study the high burnup effects and their influence on the WWER fuel properties. This paper presented the main statistical values of the WWER-440 and WWER-1000 reactors` fuel assemblies and their fragment parameters. Average burnup of fuel in the investigated fuel assemblies was in the range of 13 to 49.7 MWd/kgU. In this case, the numer of fuel cycles was from 1 to 4 during operation of the fuel assemblies.

  7. Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, William BJ J [ORNL

    2016-01-01

    A technical basis for peak reactivity boiling water reactor (BWR) burnup credit (BUC) methods was recently generated, and the technical basis for extended BWR BUC is now being developed. In this paper, a number of effects related to extended BWR BUC are analyzed, including three major operational effects in BWRs: the coolant density axial distribution, the use of control blades during operation, and the axial burnup profile. Specifically, uniform axial moderator density profiles are analyzed and compared to previous results and an additional temporal fidelity study combing moderator density profiles for three different fuel assemblies is presented. Realistic control blade histories and cask criticality results are compared to previously generated constructed control blade histories. Finally, a preliminary study of the axial burnup profile is provided.

  8. Revised SWAT. The integrated burnup calculation code system

    Energy Technology Data Exchange (ETDEWEB)

    Suyama, Kenya; Mochizuki, Hiroki [Department of Fuel Cycle Safety Research, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Kiyosumi, Takehide [The Japan Research Institute, Ltd., Tokyo (Japan)

    2000-07-01

    SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. This report shows an outline and a user's manual of revised SWAT. This revised SWAT includes expansion of functions, increasing supported machines, and correction of several bugs reported from users of previous SWAT. (author)

  9. Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian J [ORNL; Marshall, William BJ J [ORNL; Martinez-Gonzalez, Jesus S [ORNL

    2015-05-01

    Oak Ridge National Laboratory (ORNL) and the US Nuclear Regulatory Commission (NRC) have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation systems (often referred to as casks) and spent fuel pools (SFPs). This work is divided into two main phases. The first phase investigated the applicability of peak reactivity methods currently used in SFPs to transportation and storage casks and the validation of reactivity calculations and spent fuel compositions within these methods. The second phase focuses on extending BUC beyond peak reactivity. This paper documents the analysis of the effects of control blade insertion history, and moderator density and burnup axial profiles for extended BWR BUC.

  10. A Simple Global View of Fuel Burnup

    Science.gov (United States)

    Sekimoto, Hiroshi

    2017-01-01

    Reactor physics and fuel burnup are discussed in order to obtain a simple global view of the effects of nuclear reactor characteristics to fuel cycle system performance. It may provide some idea of free thinking and overall vision, though it is still a small part of nuclear energy system. At the beginning of this lecture, governing equations for nuclear reactors are presented. Since the set of these equations is so big and complicated, it is simplified by imposing some extreme conditions and the nuclear equilibrium equation is derived. Some features of future nuclear equilibrium state are obtained by solving this equation. The contribution of a nucleus charged into reactor core to the system performance indexes such as criticality is worth for understanding the importance of each nuclide. It is called nuclide importance and can be evaluated by using the equations adjoint to the nuclear equilibrium equation. Examples of some importances and their application to criticalily search problem are presented.

  11. Triton burnup measurements in KSTAR using a neutron activation system

    Science.gov (United States)

    Jo, Jungmin; Cheon, MunSeong; Kim, Jun Young; Rhee, T.; Kim, Junghee; Shi, Yue-Jiang; Isobe, M.; Ogawa, K.; Chung, Kyoung-Jae; Hwang, Y. S.

    2016-11-01

    Measurements of the time-integrated triton burnup for deuterium plasma in Korea Superconducting Tokamak Advanced Research (KSTAR) have been performed following the simultaneous detection of the d-d and d-t neutrons. The d-d neutrons were measured using a 3He proportional counter, fission chamber, and activated indium sample, whereas the d-t neutrons were detected using activated silicon and copper samples. The triton burnup ratio from KSTAR discharges is found to be in the range 0.01%-0.50% depending on the plasma conditions. The measured burnup ratio is compared with the prompt loss fraction of tritons calculated with the Lorentz orbit code and the classical slowing-down time. The burnup ratio is found to increase as plasma current and classical slowing-down time increase.

  12. Calibration of burnup monitor in the Rokkasho reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Oheda, K.; Naito, H.; Hirota, M. [Japan Nuclear Fuel Ltd., Aomori (Japan); Natsume, K. [Toshiba Corp., Yokohama, Kawasaki, Kanagawa (Japan); Kumanomido, H. [Toshiba Corp., Kawasaki, Kanagawa (Japan)

    1998-07-01

    The Rokkasho Reprocessing Plant has adopted a credit for burnup in criticality control in the Spent Fuel Storage Facility (SFSF) and the Dissolution Facility. The burnup monitor system, prepared for BWR and PWR type fuel assemblies, nondestructively measures the burnup value and determines the residual U-235 enrichment in a spent fuel assembly, and criticality is controlled by the value of residual U-235 enrichment in SFSF and by the value of top 50 cm average burnup in the Dissolution Facility. The burnup monitor consists of three measurement systems; a Boss gamma-ray profile measurement system, a high resolution gamma-ray spectrometry system, and a passive neutron measurement system. The monitor sensitivity is calibrated against operator-declared burnup values through repetitive measurements of 100 spent fuel assemblies: BWR 8 X 8, PWR 14 X 14. and 17 X 17. The outline of the measurement methods, objectives of the calibration, actual calibration method, and an example of calibration performed in a demonstration experiment are presented. (author)

  13. Analysis of high burnup pressurized water reactor fuel using uranium, plutonium, neodymium, and cesium isotope correlations with burnup

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jung Suk; Jeon, Young Shin; Park, Soon Dal; Ha, Yeong Keong; Song, Kyu Seok [Nuclear Chemistry Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-12-15

    The correlation of the isotopic composition of uranium, plutonium, neodymium, and cesium with the burnup for high burnup pressurized water reactor fuels irradiated in nuclear power reactors has been experimentally investigated. The total burnup was determined by Nd-148 and the fractional {sup 235}U burnup was determined by U and Pu mass spectrometric methods. The isotopic compositions of U, Pu, Nd, and Cs after their separation from the irradiated fuel samples were measured using thermal ionization mass spectrometry. The contents of these elements in the irradiated fuel were determined through an isotope dilution mass spectrometric method using {sup 233}U, {sup 242}Pu, {sup 150}Nd, and {sup 133}Cs as spikes. The activity ratios of Cs isotopes in the fuel samples were determined using gamma-ray spectrometry. The content of each element and its isotopic compositions in the irradiated fuel were expressed by their correlation with the total and fractional burnup, burnup parameters, and the isotopic compositions of different elements. The results obtained from the experimental methods were compared with those calculated using the ORIGEN-S code.

  14. Effect of burn-up and high burn-up structure on spent nuclear fuel alteration

    Energy Technology Data Exchange (ETDEWEB)

    Clarens, F.; Gonzalez-Robles, E.; Gimenez, F. J.; Casas, I.; Pablo, J. de; Serrano, D.; Wegen, D.; Glatz, J. P.; Martinez-Esparza, A.

    2009-07-01

    In this report the results of the experimental work carried out within the collaboration project between ITU-ENRESA-UPC/CTM on spent fuel (SF) covering the period 2005-2007 were presented. Studies on both RN release (Fast Release Fraction and matrix dissolution rate) and secondary phase formation were carried out by static and flow through experiments. Experiments were focussed on the study of the effect of BU with two PWR SF irradiated in commercial reactors with mean burn-ups of 48 and 60 MWd/KgU and; the effect of High Burn-up Structure (HBS) using powdered samples prepared from different radial positions. Additionally, two synthetic leaching solutions, bicarbonate and granitic bentonite ground wa ter were used. Higher releases were determined for RN from SF samples prepared from the center in comparison with the fuel from the periphery. However, within the studied range, no BU effect was observed. After one year of contact time, secondary phases were observed in batch experiments, covering the SF surface. Part of the work was performed for the Project NF-PRO of the European Commission 6th Framework Programme under contract no 2389. (Author)

  15. Strategies for Application of Isotopic Uncertainties in Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I.C.

    2002-12-23

    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 nuclide concentrations to the uncertainty in the predicted neutron multiplication factor (k{sub eff}) of the system can have a significant effect on the uncertainty in the safety margin in criticality calculations and ultimately affect the potential capacity of spent fuel transport and storage casks employing burnup credit. Methods that can provide a more accurate and realistic estimate of the uncertainty may enable increased spent fuel cask capacity and fewer casks needing to be transported, thereby reducing regulatory burden on licensee while maintaining safety for transporting spent fuel. This report surveys several different best-estimate strategies for considering the effects of nuclide uncertainties in burnup-credit analyses. The potential benefits of these strategies are illustrated for a prototypical burnup-credit cask design. The subcritical margin estimated using best-estimate methods is discussed in comparison to the margin estimated using conventional bounding methods of uncertainty propagation. To quantify the comparison, each of the strategies for estimating uncertainty has been performed using a common database of spent fuel isotopic assay measurements for pressurized-light-water reactor fuels and predicted nuclide concentrations obtained using the current version of the SCALE code system. The experimental database applied in this study has been significantly expanded to include new high-enrichment and high-burnup spent fuel assay data recently published for a wide range of important burnup-credit actinides and fission products. Expanded rare earth fission-product measurements performed at the Khlopin Radium Institute in Russia that contain the only known publicly-available measurement for {sup 103

  16. Size Design of CdZnTe Detector Shield for Measuring Burnup of Spent Fuel

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    <正>It is important to measure the burnup of spent fuel for nuclear safeguards, burnup credit and critical safety in spent-fuel reprocessing process. The purpose of this work is designing a portable device to

  17. Power excursion analysis for BWR`s at high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Diamond, D.J.; Neymoith, L.; Kohut, P. [Brookhaven National Lab., Upton, NY (United States)

    1996-03-01

    A study has been undertaken to determine the fuel enthalpy during a rod drop accident and during two thermal-hydraulic transients. The objective was to understand the consequences to high burnup fuel and the sources of uncertainty in the calculations. The analysis was done with RAMONA-4B, a computer code that models the neutron kinetics throughout the core along with the thermal-hydraulics in the core, vessel, and steamline. The results showed that the maximum fuel enthalpy in high burnup fuel will be affected by core design, initial conditions, and modeling assumptions. The important parameters in each of these categories are discussed in the paper.

  18. Need for higher fuel burnup at the Hatch Plant

    Energy Technology Data Exchange (ETDEWEB)

    Beckhman, J.T. [Georgia Power Co., Birmingham, AL (United States)

    1996-03-01

    Hatch is a BWR 4 and has been in operation for some time. The first unit became commercial about 1975. Obtaining higher burnups, or higher average discharge exposures, is nothing new at Hatch. Since we have started, the discharge exposure of the plant has increased. Now, of course, we are not approaching the numbers currently being discussed but, the average discharge exposure has increased from around 20,000 MWD/MTU in the early to mid-1980s to 34,000 MWD/MTU in 1994, I am talking about batch average values. There are also peak bundle and peak rod values. You will have to make the conversions if you think in one way or the other because I am talking in batch averages. During Hatch`s operating history we have had some problems with fuel failure. Higher burnup fuel raises a concern about how much fuel failure you are going to have. Fuel failure is, of course, an economic issue with us. Back in the early 1980s, we had a problem with crud-induced localized corrosion, known as CILC. We have gotten over that, but we had some times when it was up around 27 fuel failures a year. That is not a pleasant time to live through because it is not what you want from an economic viewpoint or any other. We have gotten that down. We have had some fuel failures recently, but they have not been related to fuel burnup or to corrosion. In fact, the number of failures has decreased from the early 1980s to the 90s even though burnup increased during that time. The fuel failures are more debris-related-type failures. In addition to increasing burnups, utilities are actively evaluating or have already incorporated power uprate and longer fuel cycles (e.g., 2-year cycles). The goal is to balance out the higher power density, longer cycles, higher burnup, and to have no leakers. Why do we as an industry want to have higher burnup fuel? That is what I want to tell you a little bit about.

  19. OREST - The hammer-origen burnup program system

    Energy Technology Data Exchange (ETDEWEB)

    Hesse, U. (Gesellschaft fur Reaktorsicherheit mbH Forschungsgelande, 8046 Garching bei Munchen (DE))

    1988-08-01

    Reliable prediction of the characteristics of irradiated light water reactor fuels (e.g., afterheat power, neutron and gamma radiation sources, final uranium and plutonium contents) is needed for many aspects of the nuclear fuel cycle. Two main problems must be solved: the simulation of all isotopic nuclear reactions and the simulation of neutron fluxes setting the reactions in motion. In state-of-the-art computer techniques, a combination of specialized codes for lattice cell and burnup calculations is preferred to solve these cross-linked problems in time or burnup step approximation. In the program system OREST, developed for official and commercial tasks in the Federal Republic of Germany nuclear fuel cycle, the well-known codes HAMMER and ORIGEN and directly coupled with a fuel rod temperature module.

  20. ATR PDQ and MCWO Fuel Burnup Analysis Codes Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    G.S. Chang; P. A. Roth; M. A. Lillo

    2009-11-01

    The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is being studied to determine the feasibility of converting it from the highly enriched Uranium (HEU) fuel that is currently uses to low enriched Uranium (LEU) fuel. In order to achieve this goal, it would be best to qualify some different computational methods than those that have been used at ATR for the past 40 years. This paper discusses two methods of calculating the burnup of ATR fuel elements. The existing method, that uses the PDQ code, is compared to a modern method that uses A General Monte Carlo N-Particle Transport Code (MCNP) combined with the Origen2.2 code. This modern method, MCNP with ORIGEN2.2 (MCWO), is found to give excellent agreement with the existing method (PDQ). Both of MCWO and PDQ are also in a very good agreement to the 235U burnup data generated by an analytical method.

  1. New burnup calculation of TRIGA IPR-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Meireles, Sincler P. de; Campolina, Daniel de A.M.; Santos, Andre A. Campagnole dos; Menezes, Maria A.B.C.; Mesquita, Amir Z., E-mail: sinclercdtn@hotmail.com.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil, operates since 1960.The reactor is operating for more than fifty years and has a long history of operation. Determining the current composition of the fuel is very important to calculate various parameters. The reactor burnup calculation has been performed before, however, new techniques, methods, software and increase of the processing capacity of the new computers motivates new investigations to be performed. This work presents the evolution of effective multiplication constant and the results of burnup. This new model has a more detailed geometry with the introduction of the new devices, like the control rods and the samarium discs. This increase of materials in the simulation in burnup calculation was very important for results. For these series of simulations a more recently cross section library, ENDF/B-VII, was used. To perform the calculations two Monte Carlo particle transport code were used: Serpent and MCNPX. The results obtained from two codes are presented and compared with previous studies in the literature. (author)

  2. Model biases in high-burnup fast reactor simulations

    Energy Technology Data Exchange (ETDEWEB)

    Touran, N.; Cheatham, J.; Petroski, R. [TerraPower LLC, 11235 S.E. 6th St, Bellevue, WA 98004 (United States)

    2012-07-01

    A new code system called the Advanced Reactor Modeling Interface (ARMI) has been developed that loosely couples multiscale, multiphysics nuclear reactor simulations to provide rapid, user-friendly, high-fidelity full systems analysis. Incorporating neutronic, thermal-hydraulic, safety/transient, fuel performance, core mechanical, and economic analyses, ARMI provides 'one-click' assessments of many multi-disciplined performance metrics and constraints that historically require iterations between many diverse experts. The capabilities of ARMI are implemented in this study to quantify neutronic biases of various modeling approximations typically made in fast reactor analysis at an equilibrium condition, after many repetitive shuffles. Sensitivities at equilibrium that result in very high discharge burnup are considered ( and >20% FIMA), as motivated by the development of the Traveling Wave Reactor. Model approximations discussed include homogenization, neutronic and depletion mesh resolution, thermal-hydraulic coupling, explicit control rod insertion, burnup-dependent cross sections, fission product model, burn chain truncation, and dynamic fuel performance. The sensitivities of these approximations on equilibrium discharge burnup, k{sub eff}, power density, delayed neutron fraction, and coolant temperature coefficient are discussed. (authors)

  3. Fuel burnup calculation of a research reactor plate element

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Nadia Rodrigues dos; Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes, E-mail: nadiasam@gmail.com, E-mail: zrlima@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    This work consists in simulating the burnup of two different plate type fuel elements, where one is the benchmark MTR of the IAEA, which is made of an alloy of uranium and aluminum, while the other belonging to a typical multipurpose reactor is composed of an alloy of uranium and silicon. The simulation is performed using the WIMSD-5B computer code, which makes use of deterministic methods for solving neutron transport. In developing this task, fuel element equivalent cells were calculated representing each of the reactors to obtain the initial concentrations of each isotope constituent element of the fuel cell and the thicknesses corresponding to each region of the cell, since this information is part of the input data. The compared values of the k∞ showed a similar behavior for the case of the MTR calculated with the WIMSD-5B and EPRI-CELL codes. Relating the graphs of the concentrations in the burnup of both reactors, there are aspects very similar to each isotope selected. The application WIMSD-5B code to calculate isotopic concentrations and burnup of the fuel element, proved to be satisfactory for the fulfillment of the objective of this work. (author)

  4. Models for fuel rod behaviour at high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Jernkvist, Lars O.; Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park, Uppsala (Sweden)

    2004-12-01

    This report deals with release of fission product gases and irradiation-induced restructuring in uranium dioxide nuclear fuel. Waterside corrosion of zirconium alloy clad tubes to light water reactor fuel rods is also discussed. Computational models, suitable for implementation in the FRAPCON-3.2 computer code, are proposed for these potentially life-limiting phenomena. Hence, an integrated model for the calculation or thermal fission gas release by intragranular diffusion, gas trapping in grain boundaries, irradiation-induced re-solution, grain boundary saturation, and grain boundary sweeping in UO{sub 2} fuel, under time varying temperature loads, is formulated. After a brief review of the status of thermal fission gas release modelling, we delineate the governing equations for the aforementioned processes. Grain growth kinetic modelling is briefly reviewed and pertinent data on grain growth of high burnup fuel obtained during power ramps in the Third Risoe Fission Gas Release Project are evaluated. Sample computations are performed, which clearly show the connection between fission gas release and gram growth as a function of time at different isotherms. Models are also proposed for the restructuring of uranium dioxide fuel at high burnup, the so-called rim formation, and its effect on fuel porosity build-up, fuel thermal conductivity and fission gas release. These models are assessed by use of recent experimental data from the High Burnup Rim Project, as well as from post irradiation examinations of high-burnup fuel, irradiated in power reactors. Moreover, models for clad oxide growth and hydrogen pickup in PWRs, applicable to Zircaloy-4, ZIRLO or M5 cladding, are formulated, based on recent in-reactor corrosion data for high-burnup fuel rods. Our evaluation of these data indicates that the oxidation rate of ZIRLO-type materials is about 20% lower than for standard Zircaloy-4 cladding under typical PWR conditions. Likewise, the oxidation rate of M5 seems to be

  5. MONTE-CARLO BURNUP CALCULATION UNCERTAINTY QUANTIFICATION AND PROPAGATION DETERMINATION

    Energy Technology Data Exchange (ETDEWEB)

    Nichols, T.; Sternat, M.; Charlton, W.

    2011-05-08

    MONTEBURNS is a Monte-Carlo depletion routine utilizing MCNP and ORIGEN 2.2. Uncertainties exist in the MCNP transport calculation, but this information is not passed to the depletion calculation in ORIGEN or saved. To quantify this transport uncertainty and determine how it propagates between burnup steps, a statistical analysis of a multiple repeated depletion runs is performed. The reactor model chosen is the Oak Ridge Research Reactor (ORR) in a single assembly, infinite lattice configuration. This model was burned for a 25.5 day cycle broken down into three steps. The output isotopics as well as effective multiplication factor (k-effective) were tabulated and histograms were created at each burnup step using the Scott Method to determine the bin width. It was expected that the gram quantities and k-effective histograms would produce normally distributed results since they were produced from a Monte-Carlo routine, but some of results do not. The standard deviation at each burnup step was consistent between fission product isotopes as expected, while the uranium isotopes created some unique results. The variation in the quantity of uranium was small enough that, from the reaction rate MCNP tally, round off error occurred producing a set of repeated results with slight variation. Statistical analyses were performed using the {chi}{sup 2} test against a normal distribution for several isotopes and the k-effective results. While the isotopes failed to reject the null hypothesis of being normally distributed, the {chi}{sup 2} statistic grew through the steps in the k-effective test. The null hypothesis was rejected in the later steps. These results suggest, for a high accuracy solution, MCNP cell material quantities less than 100 grams and greater kcode parameters are needed to minimize uncertainty propagation and minimize round off effects.

  6. Thermodynamic analysis for high burn-up fuel internal chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Fuji, Kensho; Kyoh, Bunkei [Kinki Univ., Higashi-Osaka, Osaka (Japan). Faculty of Science and Technology

    1997-09-01

    Chemical states of fission products and actinide elements in high burn-up LWR fuel pellets have been analyzed thermodynamically using the computer program SOLGASMIX-PV. Calculations with this computer code have been performed for a complex multi-component system, which comprises 54 chemical species. The analysis shows that neither alkali nor alkaline-earth uranates are formed, but alkali and alkaline-earth molybdates exist in irradiated LWR fuel pellets in contrast with their post irradiation examinations. These molybdates tend to increase with increasing oxygen potential in the fuel under operating conditions, whereas the zirconates decrease. (author)

  7. Determination of IRT-2M fuel burnup by gamma spectrometry.

    Science.gov (United States)

    Koleška, Michal; Viererbl, Ladislav; Marek, Milan; Ernest, Jaroslav; Šunka, Michal; Vinš, Miroslav

    2016-01-01

    A spectrometric system was developed for evaluating spent fuel in the LVR-15 research reactor, which employs highly enriched (36%) IRT-2M-type fuel. Such system allows the measurement of detailed fission product profiles. Within these measurements, nuclides such as (137)Cs, (134)Cs, (144)Ce, (106)Ru and (154)Eu may be detected in fuel assemblies with different cooling times varying between 1.67 and 7.53 years. Burnup calculations using the MCNPX Monte Carlo code data showed good agreement with measurements, though some discrepancies were observed in certain regions. These discrepancies are attributed to the evaluation of irradiation history, reactor regulation pattern and buildup schemes.

  8. New results from the NSRR experiments with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Fuketa, Toyoshi; Ishijima, Kiyomi; Mori, Yukihide [Japan Atomic Research Institute, Toaki, Ibaraki (Japan)] [and others

    1996-03-01

    Results obtained in the NSRR power burst experiments with irradiated PWR fuel rods with fuel burnup up to 50 MWd/kgU are described and discussed in this paper. Data concerning test method, test fuel rod, pulse irradiation, transient records during the pulse and post irradiation examination are described, and interpretations and discussions on fission gas release and fuel pellet fragmentation are presented. During the pulse-irradiation experiment with 50 MWd/kgU PWR fuel rod, the fuel rod failed at considerably low energy deposition level, and large amount of fission gas release and fragmentation of fuel pellets were observed.

  9. Evaluation of Isotopic Measurements and Burn-up Value of Sample GU3 of ARIANE Project

    Energy Technology Data Exchange (ETDEWEB)

    Tore, C.; Rodriguez Rivada, A.

    2014-07-01

    Estimation of the burn-up value of irradiated fuel and its isotopic composition are important for criticality analysis, spent fuel management and source term estimation. The practical way to estimate the irradiated fuel composition and burn.up value is calculation with validated code and nuclear data. Such validation of the neutronic codes and nuclear data requires the benchmarking with measured values. (Author)

  10. MTR core loading pattern optimization using burnup dependent group constants

    Directory of Open Access Journals (Sweden)

    Iqbal Masood

    2008-01-01

    Full Text Available A diffusion theory based MTR fuel management methodology has been developed for finding superior core loading patterns at any stage for MTR systems, keeping track of burnup of individual fuel assemblies throughout their history. It is based on using burnup dependent group constants obtained by the WIMS-D/4 computer code for standard fuel elements and control fuel elements. This methodology has been implemented in a computer program named BFMTR, which carries out detailed five group diffusion theory calculations using the CITATION code as a subroutine. The core-wide spatial flux and power profiles thus obtained are used for calculating the peak-to-average power and flux-ratios along with the available excess reactivity of the system. The fuel manager can use the BFMTR code for loading pattern optimization for maximizing the excess reactivity, keeping the peak-to-average power as well as flux-ratio within constraints. The results obtained by the BFMTR code have been found to be in good agreement with the corresponding experimental values for the equilibrium core of the Pakistan Research Reactor-1.

  11. Calibration of burnup monitor of spent nuclear fuel installed at Rokkasho reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Oeda, Kaoru; Matoba, Masaru; Wakabayashi, Genichiro [Kyushu Univ., Fukuoka (Japan). Faculty of Engineering; Naito, Hirofumi; Hirota, Masanari [Nuclear Fuel Industries Ltd., Tokyo (Japan); Morizaki, Hidetoshi; Kumanomido, Hironori; Natsume, Koichiro [Toshiba Corp., Tokyo (Japan)

    2001-05-01

    The spent nuclear fuel storage pool of Rokkasho reprocessing plant adopts the burnup credit' conception. Spent fuel assemblies are measured every one by one, by burnup monitors, and stored to a storage rack which is designed with specified residual enrichment. For nuclear criticality control, it is necessary for the burnup monitor that the measured value includes a kind of margin, which consists of errors of the monitor. In this paper, we describe the error of the burnup monitors, and the way of taking of the margin. From the result of calibration of the burnup monitor carried out from July through November, 1999, we describe that the way of taking of the margin is validated. And comments about possibility of error reduction are remarked. (author)

  12. Detailed description and user`s manual of high burnup fuel analysis code EXBURN-I

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Saitou, Hiroaki

    1997-11-01

    EXBURN-I has been developed for the analysis of LWR high burnup fuel behavior in normal operation and power transient conditions. In the high burnup region, phenomena occur which are different in quality from those expected for the extension of behaviors in the mid-burnup region. To analyze these phenomena, EXBURN-I has been formed by the incorporation of such new models as pellet thermal conductivity change, burnup-dependent FP gas release rate, and cladding oxide layer growth to the basic structure of low- and mid-burnup fuel analysis code FEMAXI-IV. The present report describes in detail the whole structure of the code, models, and materials properties. Also, it includes a detailed input manual and sample output, etc. (author). 55 refs.

  13. Preparation of data relevant to ''Equivalent Uniform Burnup'' and Equivalent Initial Enrichment'' for burnup credit evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Nomura, Yasushi; Okuno, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Murazaki, Minoru [Tokyo Nuclear Service Inc., Tokyo (Japan)

    2001-11-01

    Based on the PWR spent fuel composition data measured at JAERI, two kinds of simplified methods such as ''Equivalent Uniform Burnup'' and ''Equivalent Initial Enrichment'' have been introduced. And relevant evaluation curves have been prepared for criticality safety evaluation of spent fuel storage pool and transport casks, taking burnup of spent fuel into consideration. These simplified methods can be used to obtain an effective neutron multiplication factor for a spent fuel storage/transportation system by using the ORIGEN2.1 burnup code and the KENO-Va criticality code without considering axial burnup profile in spent fuel and other various factors introducing calculated errors. ''Equivalent Uniform Burnup'' is set up for its criticality analysis to be reactivity equivalent with the detailed analysis, in which the experimentally obtained isotopic composition together with a typical axial burnup profile and various factors such as irradiation history are considered on the conservative side. On the other hand, Equivalent Initial Enrichment'' is set up for its criticality analysis to be reactivity equivalent with the detailed analysis such as above when it is used in the so called fresh fuel assumption. (author)

  14. RAPID program to predict radial power and burnup distribution of UO{sub 2} fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bock; Song, Jae Sung; Bang, Je Gun; Kim, Dae Ho [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-02-01

    Due to the radial variation of the neutron flux and its energy spectrum inside UO{sub 2} fuel, the fission density and fissile isotope production rates are varied radially in the pellet, and it becomes necessary to know the accurate radial power and burnup variation to predict the high burnup fuel behavior such as rim effects. Therefore, to predict the radial distribution of power, burnup and fissionable nuclide densities in the pellet with the burnup and U-235 enrichment, RAPID(RAdial power and burnup Prediction by following fissile Isotope Distribution in the pellet) program was developed. It considers the specific radial variation of the neutron reaction of the nuclides while the constant radial variation of neutron reaction except neutron absorption of U-238 regardless of the nuclides, the burnup and U-235 enrichment is assumed in TUBRNP model which is recognized as the one of the most reliable models. Therefore, it is expected that RAPID may be more accurate than TUBRNP, specially at high burnup region. RAPID is based upon and validated by the detailed reactor physics code, HELIOS which is one of few codes that can calculates the radial variations of the nuclides inside the pellet. Comparison of RAPID prediction with the measured data of the irradiated fuels showed very good agreement. RAPID can be used to calculate the local variations of the fissionable nuclide concentrations as well as the local power and burnup inside that pellet as a function of the burnup up to 10 w/o U-235 enrichment and 150 MWD/kgU burnup under the LWR environment. (author). 8 refs., 50 figs., 1 tab.

  15. Extension and validation of the TRANSURANUS burn-up model for helium production in high burn-up LWR fuels

    Science.gov (United States)

    Botazzoli, Pietro; Luzzi, Lelio; Brémier, Stephane; Schubert, Arndt; Van Uffelen, Paul; Walker, Clive T.; Haeck, Wim; Goll, Wolfgang

    2011-12-01

    The TRANSURANUS burn-up model (TUBRNP) calculates the local concentration of the actinides, the main fission products, and 4He as a function of the radial position across a fuel rod. In this paper, the improvements in the helium production model as well as the extensions in the simulation of 238-242Pu, 241Am, 243Am and 242-245Cm isotopes are described. Experimental data used for the extended validation include new EPMA measurements of the local concentrations of Nd and Pu and recent SIMS measurements of the radial distributions of Pu, Am and Cm isotopes, both in a 3.5% enriched commercial PWR UO 2 fuel with a burn-up of 80 and 65 MWd/kgHM, respectively. Good agreement has been found between TUBRNP and the experimental data. The analysis has been complemented by detailed neutron transport calculations (VESTA code), and also revealed the need to update the branching ratio for the 241Am(n,γ) 242mAm reaction in typical PWR conditions.

  16. The burnup dependence of light water reactor spent fuel oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, B.D.

    1998-07-01

    Over the temperature range of interest for dry storage or for placement of spent fuel in a permanent repository under the conditions now being considered, UO{sub 2} is thermodynamically unstable with respect to oxidation to higher oxides. The multiple valence states of uranium allow for the accommodation of interstitial oxygen atoms in the fuel matrix. A variety of stoichiometric and nonstoichiometric phases is therefore possible as the fuel oxidizers from UO{sub 2} to higher oxides. The oxidation of UO{sub 2} has been studied extensively for over 40 years. It has been shown that spent fuel and unirradiated UO{sub 2} oxidize via different mechanisms and at different rates. The oxidation of LWR spent fuel from UO{sub 2} to UO{sub 2.4} was studied previously and is reasonably well understood. The study presented here was initiated to determine the mechanism and rate of oxidation from UO{sub 2.4} to higher oxides. During the early stages of this work, a large variability in the oxidation behavior of samples oxidized under nearly identical conditions was found. Based on previous work on the effect of dopants on UO{sub 2} oxidation and this initial variability, it was hypothesized that the substitution of fission product and actinide impurities for uranium atoms in the spent fuel matrix was the cause of the variable oxidation behavior. Since the impurity concentration is roughly proportional to the burnup of a specimen, the oxidation behavior of spent fuel was expected to be a function of both temperature and burnup. This report (1) summarizes the previous oxidation work for both unirradiated UO{sub 2} and spent fuel (Section 2.2) and presents the theoretical basis for the burnup (i.e., impurity concentration) dependence of the rate of oxidation (Sections 2.3, 2.4, and 2.5), (2) describes the experimental approach (Section 3) and results (Section 4) for the current oxidation tests on spent fuel, and (3) establishes a simple model to determine the activation energies

  17. Burnup calculations for the HOMER-15 and SAFE-300 reactors

    Science.gov (United States)

    Amiri, Benjamin W.; Poston, David I.

    2002-01-01

    The Heatpipe Power System (HPS) is a near-term low-cost space fission power system. As the U-235 fuel of the HPS is burned, higher actinides and fission products will be produced. This will cause changes in system reactivity, radioactivity, and decay power. One potential concern is that gaseous fission products may exert excessive pressure on the fuel pin cladding. To evaluate these issues, simulations were run in MONTEBURNS. MONTEBURNS is an automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN2. This paper describes the results of these simulations, as well as how those results compare with the current experimental database of irradiated materials. .

  18. Mechanical Fatigue Testing of High Burnup Fuel for Transportation Applications

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-05-01

    This report describes testing designed to determine the ability of high burnup (HBU) (>45 GWd/MTU) spent fuel to maintain its integrity under normal conditions of transportation. An innovative system, Cyclic Integrated Reversible-bending Fatigue Tester (CIRFT), has been developed at Oak Ridge National Laboratory (ORNL) to test and evaluate the mechanical behavior of spent nuclear fuel (SNF) under conditions relevant to storage and transportation. The CIRFT system is composed of a U-frame equipped with load cells for imposing the pure bending loads on the SNF rod test specimen and measuring the in-situ curvature of the fuel rod during bending using a set up with three linear variable differential transformers (LVDTs).

  19. Manufacturing Data Uncertainties Propagation Method in Burn-Up Problems

    Directory of Open Access Journals (Sweden)

    Thomas Frosio

    2017-01-01

    Full Text Available A nuclear data-based uncertainty propagation methodology is extended to enable propagation of manufacturing/technological data (TD uncertainties in a burn-up calculation problem, taking into account correlation terms between Boltzmann and Bateman terms. The methodology is applied to reactivity and power distributions in a Material Testing Reactor benchmark. Due to the inherent statistical behavior of manufacturing tolerances, Monte Carlo sampling method is used for determining output perturbations on integral quantities. A global sensitivity analysis (GSA is performed for each manufacturing parameter and allows identifying and ranking the influential parameters whose tolerances need to be better controlled. We show that the overall impact of some TD uncertainties, such as uranium enrichment, or fuel plate thickness, on the reactivity is negligible because the different core areas induce compensating effects on the global quantity. However, local quantities, such as power distributions, are strongly impacted by TD uncertainty propagations. For isotopic concentrations, no clear trends appear on the results.

  20. Simulation of triton burn-up in JET plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Loughlin, M.J.; Balet, B.; Jarvis, O.N.; Stubberfield, P.M. [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking

    1994-07-01

    This paper presents the first triton burn-up calculations for JET plasmas using the transport code TRANSP. Four hot ion H-mode deuterium plasmas are studied. For these discharges, the 2.5 MeV emission rises rapidly and then collapses abruptly. This phenomenon is not fully understood but in each case the collapse phase is associated with a large impurity influx known as the ``carbon bloom``. The peak 14 MeV emission occurs at this time, somewhat later than that of the 2.5 MeV neutron peak. The present results give a clear indication that there are no significant departures from classical slowing down and spatial diffusion for tritons in JET plasmas. (authors). 7 refs., 3 figs., 1 tab.

  1. Assessment of reactivity transient experiments with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ozer, O.; Yang, R.L.; Rashid, Y.R.; Montgomery, R.O.

    1996-03-01

    A few recent experiments aimed at determining the response of high-burnup LWR fuel during a reactivity initiated accident (RIA) have raised concerns that existing failure criteria may be inappropriate for such fuel. In particular, three experiments (SPERT CDC-859, NSRR HBO-1 and CABRI REP Na-1) appear to have resulted in fuel failures at only a fraction of the anticipated enthalpy levels. In evaluating the results of such RIA simulation experiments, however, it is necessary that the following two key considerations be taken into account: (1) Are the experiments representative of conditions that LWR fuel would experience during an in-reactor RIA event? (2) Is the fuel that is being utilized in the tests representative of the present (or anticipated) population of LWR fuel? Conducting experiments under conditions that can not occur in-reactor can trigger response modes that could not take place during in-reactor operation. Similarly, using unrepresentative fuel samples for the tests will produce failure information that is of limited relevance to commercial LWR fuel. This is particularly important for high-burnup fuel since the manner under which the test samples are base-irradiated prior to the test will impact the mechanical properties of the cladding and will therefore affect the RIA response. A good example of this effect can be seen in the results of the SPERT CDC-859 test and in the NSRR JM-4 and JM-5 tests. The conditions under which the fuel used for these tests was fabricated and/or base-irradiated prior to the RIA pulse resulted in the formation of multiple cladding defects in the form of hydride blisters. When this fuel was subjected to the RIA power pulse, it failed by developing multiple cracks that were closely correlated with the locations of the pre-existing hydride blisters. In the case of the JM tests, many of the cracks formed within the blisters themselves and did not propagate beyond the heavily hydrided regions.

  2. Burnup measurements on spent fuel elements of the RP-10 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vela Mora, Mariano; Gallardo Padilla, Alberto; Palomino, Jose Luis Castro, E-mail: mvela@ipen.gob.p [Instituto Peruano de Energia Nuclear (IPEN/Peru), Lima (Peru). Grupo de Calculo, Analisis y Seguridad de Reactores; Terremoto, Luis Antonio Albiac, E-mail: laaterre@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    This work describes the measurement, using nondestructive gamma-ray spectroscopy, of the average burnup attained by Material Testing Reactor (MTR) fuel elements irradiated in the RP-10 research reactor. Measurements were performed at the reactor storage pool area using {sup 137}Cs as the only burnup monitor, even for spent fuel elements with cooling times much shorter than two years. The experimental apparatus was previously calibrated in efficiency to obtain absolute average burnup values, which were compared against corresponding ones furnished by reactor physics calculations. The mean deviation between both values amounts to 6%. (author)

  3. Investigation on using neutron counting techniques for online burnup monitoring of pebble bed reactor fuels

    Science.gov (United States)

    Zhao, Zhongxiang

    Modular Pebble Bed Reactor (MPBR) is a high temperature gas-cooled nuclear power reactor. This project investigated the feasibility of using the passive neutron counting and active neutron/gamma counting for the on line fuel burnup measurement for MPBR. To investigate whether there is a correlation between neutron emission and fuel burnup, the MPBR fuel depletion was simulated under different irradiation conditions by ORIGEN2. It was found that the neutron emission from an irradiated pebble increases with burnup super-linearly and reaches to 104 neutron/sec/pebble at the discharge burnup. The photon emission from an irradiated pebble was found to be in the order of 1013 photon/sec/pebble at all burnup levels. Analysis shows that the neutron emission rate of an irradiated pebble is sensitive to its burnup history and the spectral-averaged one-group cross sections used in the depletion calculations, which consequently leads to large uncertainty in the correlation between neutron emission and burnup. At low burnup levels, the uncertainty in the neutron emission/burnup correlation is too high and the neutron emission rate is too low so that it is impossible to determine a pebble's burnup by on-line neutron counting at low burnup levels. At high burnup levels, the uncertainty in the neutron emission rate becomes less but is still large in quantity. However, considering the super-linear feature of the correlation, the uncertainty in burnup determination was found to be ˜7% at the discharge burnup, which is acceptable. Therefore, total neutron emission rate of a pebble can be used as a burnup indicator to determine whether a pebble should be discharged or not. The feasibility of using passive neutron counting methods for the on-line burnup measurement was investigated by using a general Monte Carlo code, MCNP, to assess the detectability of the neutron emission and the capability to discriminate gamma noise by commonly used neutron detectors. It was found that both He-3

  4. Technical Development on Burn-up Credit for Spent LWR Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I.C.

    2001-12-26

    Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled ''Technical Development on Criticality Safety Management for Spent LWR Fuels.'' Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burnup and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report.

  5. ThO{sub 2}-UO{sub 2} annular pins for high burnup fuels

    Energy Technology Data Exchange (ETDEWEB)

    Caner, Marc; Dugan, Edward T

    2000-06-01

    The main purpose of this work is to investigate the use of annular fuel pins (particularly pins containing thorium dioxide) for high burnup fuel. The following parameters were evaluated and compared between postulated mixed thorium-uranium dioxide standard and annular (9% void fraction) type fuel assemblies, as a function of burnup: the infinite multiplication factor, the uranium and plutonium isotopic compositions, the fuel temperature coefficient of reactivity and the conversion ratio. We used the SCALE-4.3 code system. The calculation method consisted in obtaining actinide and fission product number densities as functions of assembly burnup, by means of a 1-D transport calculation combined with a 0-D burnup calculation. These number densities were then used in a 3-D Monte Carlo code for obtaining k{sub {infinity}} from two-dimensional-symmetry 'snapshots'.

  6. Fission-gas release at extended burnups: effect of two-dimensional heat transfer

    Energy Technology Data Exchange (ETDEWEB)

    Tayal, M. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Yu, S.D. [Ryerson Polytechnic Univ., Toronto, Ontario (Canada); Lau, J.H.K

    2000-09-01

    To better simulate the performance of high-burnup CANDU fuel, a two-dimensional model for heat transfer between the pellet and the sheath has been added to the computer code ELESTRES. The model covers four relative orientations of the pellet and the sheath and their impacts on heat transfer and fission-gas release. The predictions of the code were compared to a database of 27 experimental irradiations involving extended burnups and normal burnups. The calculated values of fission gas release matched the measurements to an average of 94%. Thus, the two-dimensional heat transfer model increases the versatility of the ELESTRES code to better simulate fuels at normal as well as at extended burnups. (author)

  7. Technical development on burn-up credit for spent LWR fuels

    Energy Technology Data Exchange (ETDEWEB)

    Nakahara, Yoshinori; Suyama, Kenya; Suzaki, Takenori [eds.] [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-10-01

    Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled 'Technical Development on Criticality Safety Management for Spent LWR Fuels'. Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burn-up and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report. (author)

  8. Topical Report on Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    1998-09-01

    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 summarized below: (1) Validate a computer code system to calculate isotopic concentrations in SNF created during burnup in the reactor core and subsequent decay. (2) Validate a computer code system to predict the subcritical multiplication factor, keff, of a spent nuclear fuel package. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). and (5) Verify that SNF assemblies meet the package loading criteria and confirm proper fuel assembly selection prior to loading. (This step is required but the details are outside the scope of this topical report.) When reviewed and accepted by the NRC, this topical report will serve as a criterion document for criticality control analysts and will provide steps for the use of actinide-only burnup credit in the design of criticality control systems. The NRC-accepted burnup credit methodology will be used by commercial SNF storage and transportation package designers. Design-specific burnup credit criticality analyses will be defined, developed, and documented in the Safety Analysis Report (SAR) for each specific storage or transportation package that uses burnup credit. These SARs will then be submitted to the NRC for review and approval. This topical report is expected to be referenced in a number of storage and transportation cask applications to be submitted by commercial cask and canister designers to the NRC. Therefore, NRC acceptance of this topical report will result in increased efficiency of the

  9. French investigations of high burnup effect on LOCA thermomecanical behavior. Part two. Oxidation and quenching experiments under simulated LOCA conditions with high burnup clad material

    Energy Technology Data Exchange (ETDEWEB)

    GrandJean, C. [IPSN, Cadarache (France); Cauvin, R.; Lebuffe, C. [EDF/SCMI, Chinon (France)] [and others

    1997-01-01

    In the frame of the high burnup fuel studies to support a possible extension of the current discharge burnup limit, experimental programs have been undertaken, jointly by EDF and IPSN in order to study the thermal-shock behavior of high burnup fuel claddings under typical LOCA conditions. The TAGUS program used unirradiated cladding samples, bare or bearing a pre-corrosion state simulating the end-of-life state of high burnup fuel claddings: the TAGCIR program used actually irradiated cladding samples taken from high burnup rods irradiated over 5 cycles in a commercial EDF PWR and having reached a rod burnup close to 60 GWd/tU. The thermal-shock failure tests consisted in oxidizing the cladding samples under steam flow, on both inner and outer faces or on the outer face alone, and subjecting them to a final water quench. The heating was provided by an inductive furnace the power of which being regulated through monitoring of the sample surface temperature with use of a single-wave optical pyrometer. Analysis of the irradiated tests (TAGCIR series) evidenced an increased oxidation rate as compared to similar tests on unirradiated samples. Results of the quenching tests series on unirradiated and irradiated samples are plotted under the usual presentation of failure maps relative to the oxidation parameters ECR (equivalent cladding reacted) or e{sub {beta}} (thickness of the remaining beta phase layer) as a function of the oxidation temperature. Comparison of the failure limits for irradiated specimens to those for unirradiated specimens indicates a lower brittleness under two side oxidation and possibly the opposite under one-side oxidation. The tentative analysis of the oxidation and quenching tests results on irradiated samples reveals the important role played by the hydrogen charged during in-reactor corrosion on the oxidation kinetics and the failure bearing capability of the cladding under LOCA transient conditions.

  10. Burnup analysis of the VVER-1000 reactor using thorium-based fuel

    Energy Technology Data Exchange (ETDEWEB)

    Korkmaz, Mehmet E.; Agar, Osman; Bueyueker, Eylem [Karamanoglu Mehmetbey Univ., Karaman (Turkey). Faculty of Kamil Ozdag Science

    2014-12-15

    This paper aims to investigate {sup 232}Th/{sup 233}U fuel cycles in a VVER-1000 reactor through calculation by computer. The 3D core geometry of VVER-1000 system was designed using the Serpent Monte Carlo 1.1.19 Code. The Serpent Code using parallel programming interface (Message Passing Interface-MPI), was run on a workstation with 12-core and 48 GB RAM. {sup 232}Th/{sup 235}U/{sup 238}U oxide mixture was considered as fuel in the core, when the mass fraction of {sup 232}Th was increased as 0.05-0.1-0.2-0.3-0.4 respectively, the mass fraction of {sup 238}U equally was decreased. In the system, the calculations were made for 3 000 MW thermal power. For the burnup analyses, the core is assumed to deplete from initial fresh core up to a burnup of 16 MWd/kgU without refuelling considerations. In the burnup calculations, a burnup interval of 360 effective full power days (EFPDs) was defined. According to burnup, the mass changes of the {sup 232}Th, {sup 233}U, {sup 238}U, {sup 237}Np, {sup 239}Pu, {sup 241}Am and {sup 244}Cm were evaluated, and also flux and criticality of the system were calculated in dependence of the burnup rate.

  11. Technical Data to Justify Full Burnup Credit in Criticality Safety Licensing Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Enercon Services, Inc.

    2011-03-14

    Enercon Services, Inc. (ENERCON) was requested under Task Order No.2 to identify scientific and technical data needed to benchmark and justify Full Burnup Credit, which adds 16 fission products and 4 minor actinides1 to Actinide-Only burnup credit. The historical perspective for Full Burnup Credit is discussed, and interviews of organizations participating in burnup credit activities are summarized as a basis for identifying additional data needs and making recommendation. Input from burnup credit participants representing two segments of the commercial nuclear industry is provided. First, the Electric Power Research Institute (EPRI) has been very active in the development of Full Burnup Credit, representing the interests of nuclear utilities in achieving capacity gains for storage and transport casks. EPRI and its utility customers are interested in a swift resolution of the validation issues that are delaying the implementation of Full Burnup Credit [EPRI 2010b]. Second, used nuclear fuel storage and transportation Cask Vendors favor improving burnup credit beyond Actinide-Only burnup credit, although their discussion of specific burnup credit achievements and data needs was limited citing business sensitive and technical proprietary concerns. While Cask Vendor proprietary items are not specifically identified in this report, the needs of all nuclear industry participants are reflected in the conclusions and recommendations of this report. In addition, Oak Ridge National Laboratory (ORNL) and Sandia National Laboratory (SNL) were interviewed for their input into additional data needs to achieve Full Burnup Credit. ORNL was very open to discussions of Full Burnup Credit, with several telecoms and a visit by ENERCON to ORNL. For many years, ORNL has provided extensive support to the NRC regarding burnup credit in all of its forms. Discussions with ORNL focused on potential resolutions to the validation issues for the use of fission products. SNL was helpful in

  12. High Burn-Up Spent Nuclear Fuel Vibration Integrity Study

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jiang, Hao [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Rob L [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John M [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    The Oak Ridge National Laboratory (ORNL) has developed the cyclic integrated reversible-bending fatigue tester (CIRFT) approach to successfully demonstrate the controllable fatigue fracture on high burnup (HBU) spent nuclear fuel (SNF) in a normal vibration mode. CIRFT enables examination of the underlying mechanisms of SNF system dynamic performance. Due to the inhomogeneous composite structure of the SNF system, the detailed mechanisms of the pellet-pellet and pellet-clad interactions and the stress concentration effects at the pellet-pellet interface cannot be readily obtained from a CIRFT system measurement. Therefore, finite element analyses (FEAs) are used to translate the global moment-curvature measurement into local stress-strain profiles for further investigation. The major findings of CIRFT on the HBU SNF are as follows: SNF system interface bonding plays an important role in SNF vibration performance. Fuel structure contributes to SNF system stiffness. There are significant variations in stress and curvature of SNF systems during vibration cycles resulting from segment pellets and clad interactions. SNF failure initiates at the pellet-pellet interface region and appears to be spontaneous.

  13. Topical report on actinide-only burnup credit for PWR spent nuclear fuel packages. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    1997-04-01

    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 credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, k{sub eff}, of a spent nuclear fuel package. Fifty-seven UO{sub 2}, UO{sub 2}/Gd{sub 2}O{sub 3}, and UO{sub 2}/PuO{sub 2} critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k{sub eff} (which can be a function of the trending parameters) such that the biased k{sub eff}, when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package

  14. Tritium release from EXOTIC-7 orthosilicate pebbles. Effect of burnup and contact with beryllium during irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Scaffidi-Argentina, F.; Werle, H. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik

    1998-03-01

    EXOTIC-7 was the first in-pile test with {sup 6}Li-enriched (50%) lithium orthosilicate (Li{sub 4}SiO{sub 4}) pebbles and with DEMO representative Li-burnup. Post irradiation examinations of the Li{sub 4}SiO{sub 4} have been performed at the Forschungszentrum Karlsruhe (FZK), mainly to investigate the tritium release kinetics as well as the effect of Li-burnup and/or contact with beryllium during irradiation. The release rate of Li{sub 4}SiO{sub 4} from pure Li{sub 4}SiO{sub 4} bed of capsule 28.1-1 is characterized by a broad main peak at about 400degC and by a smaller peak at about 800degC, and that from the mixed beds of capsule 28.2 and 26.2-1 shows again these two peaks, but most of the tritium is now released from the 800degC peak. This shift of release from low to high temperature may be due to the higher Li-burnup and/or due to contact with Be during irradiation. Due to the very difficult interpretation of the in-situ tritium release data, residence times have been estimated on the basis of the out-of-pile tests. The residence time for Li{sub 4}SiO{sub 4} from caps. 28.1-1 irradiated at 10% Li-burnup agrees quite well with that of the same material irradiated at Li-burnup lower than 3% in the EXOTIC-6 experiment. In spite of the observed shift in the release peaks from low to high temperature, also the residence time for Li{sub 4}SiO{sub 4} from caps. 26.2-1 irradiated at 13% Li-burnup agrees quite well with the data from EXOTIC-6 experiment. On the other hand, the residence time for Li{sub 4}SiO{sub 4} from caps. 28.2 (Li-burnup 18%) is about a factor 1.7-3.8 higher than that for caps. 26.2-1. Based on these data on can conclude that up to 13% Li-burnup neither the contact with beryllium nor the Li-burnup have a detrimental effect on the tritium release of Li{sub 4}SiO{sub 4} pebbles, but at 18% Li-burnup the residence time is increased by about a factor three. (J.P.N.)

  15. Assessment of the uncertainties of MULTICELL calculations by the OECD NEA UAM PWR pin cell burnup benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Kereszturi, Andras [Hungarian Academy of Sciences, Budapest (Hungary). Centre for Energy Research; Panka, Istvan

    2015-09-15

    Defining precisely the burnup of the nuclear fuel is important from the point of view of core design calculations, safety analyses, criticality calculations (e.g. burnup credit calculations), etc. This paper deals with the uncertainties of MULTICELL calculations obtained by the solution of the OECD NEA UAM PWR pin cell burnup benchmark. In this assessment Monte-Carlo type statistical analyses are applied and the energy dependent covariance matrices of the cross-sections are taken into account. Additionally, the impact of the uncertainties of the fission yields is also considered. The target quantities are the burnup dependent uncertainties of the infinite multiplication factor, the two-group cross-sections, the reaction rates and the number densities of some isotopes up to the burnup of 60 MWd/kgU. In the paper the burnup dependent tendencies of the corresponding uncertainties and their sources are analyzed.

  16. Summary of high burnup fuel issues and NRC`s plan of action

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, R.O.

    1997-01-01

    For the past two years the Office of Nuclear Regulatory Research has concentrated mostly on the so-called reactivity-initiated accidents -- the RIAs -- in this session of the Water Reactor Safety Information Meeting, but this year there is a more varied agenda. RIAs are, of course, not the only events of interest for reactor safety that are affected by extended burnup operation. Their has now been enough time to consider a range of technical issues that arise at high burnup, and a list of such issues being addressed in their research program is given here. (1) High burnup capability of the steady-state code (FRAPCON) used for licensing audit calculations. (2) General capability (including high burnup) of the transient code (FRAPTRAN) used for special studies. (3) Adequacy at high burnup of fuel damage criteria used in regulation for reactivity accidents. (4) Adequacy at high burnup of models and fuel related criteria used in regulation for loss-of-coolant accidents (LOCAs). (5) Effect of high burnup on fuel system damage during normal operation, including control rod insertion problems. A distinction is made between technical issues, which may or may not have direct licensing impacts, and licensing issues. The RIAs became a licensing issue when the French test in CABRI showed that cladding failures could occur at fuel enthalpies much lower than a value currently used in licensing. Fuel assembly distortion became a licensing issue when control rod insertion was affected in some operating plants. In this presentation, these technical issues will be described and the NRC`s plan of action to address them will be discussed.

  17. New high burnup fuel models for NRC`s licensing audit code, FRAPCON

    Energy Technology Data Exchange (ETDEWEB)

    Lanning, D.D.; Beyer, C.E.; Painter, C.L. [Pacific Northwest Laboratory, Richland, WA (United States)

    1996-03-01

    Fuel behavior models have recently been updated within the U.S. Nuclear Regulatory Commission steady-state FRAPCON code used for auditing of fuel vendor/utility-codes and analyses. These modeling updates have concentrated on providing a best estimate prediction of steady-state fuel behavior up to the maximum burnup level s of current data (60 to 65 GWd/MTU rod-average). A decade has passed since these models were last updated. Currently, some U.S. utilities and fuel vendors are requesting approval for rod-average burnups greater than 60 GWd/MTU; however, until these recent updates the NRC did not have valid fuel performance models at these higher burnup levels. Pacific Northwest Laboratory (PNL) has reviewed 15 separate effects models within the FRAPCON fuel performance code (References 1 and 2) and identified nine models that needed updating for improved prediction of fuel behavior at high burnup levels. The six separate effects models not updated were the cladding thermal properties, cladding thermal expansion, cladding creepdown, fuel specific heat, fuel thermal expansion and open gap conductance. Comparison of these models to the currently available data indicates that these models still adequately predict the data within data uncertainties. The nine models identified as needing improvement for predicting high-burnup behavior are fission gas release (FGR), fuel thermal conductivity (accounting for both high burnup effects and burnable poison additions), fuel swelling, fuel relocation, radial power distribution, fuel-cladding contact gap conductance, cladding corrosion, cladding mechanical properties and cladding axial growth. Each of the updated models will be described in the following sections and the model predictions will be compared to currently available high burnup data.

  18. Model for evolution of grain size in the rim region of high burnup UO2 fuel

    Science.gov (United States)

    Xiao, Hongxing; Long, Chongsheng; Chen, Hongsheng

    2016-04-01

    The restructuring process of the high burnup structure (HBS) formation in UO2 fuel results in sub-micron size grains that accelerate the fission gas swelling, which will raise some concern over the safety of extended the nuclear fuel operation life in the reactor. A mechanistic and engineering model for evolution of grain size in the rim region of high burnup UO2 fuel based on the experimental observations of the HBS in the literature is presented. The model takes into account dislocations evolution under irradiation and the grain subdivision occur successively at increasing local burnup. It is assumed that the original driving force for subdivision of grain in the HBS of UO2 fuel is the production and accumulation of dislocation loops during irradiation. The dislocation loops can also be annealed through thermal diffusion when the temperature is high enough. The capability of this model is validated by the comparison with the experimental data of temperature threshold of subdivision, dislocation density and sub-grain size as a function of local burnup. It is shown that the calculated results of the dislocation density and subdivided grain size as a function of local burnup are in good agreement with the experimental results.

  19. Plutonium and Minor Actinides Recycling in Standard BWR using Equilibrium Burnup Model

    Directory of Open Access Journals (Sweden)

    Abdul Waris

    2008-03-01

    Full Text Available Plutonium (Pu and minor actinides (MA recycling in standard BWR with equilibrium burnup model has been studied. We considered the equilibrium burnup model as a simple time independent burnup method, which can manage all possible produced nuclides in any nuclear system. The equilibrium burnup code was bundled with a SRAC cell-calculation code to become a coupled cell-burnup calculation code system. The results show that the uranium enrichment for the criticality of the reactor, the amount of loaded fuel and the required natural uranium supply per year decrease for the Pu recycling and even much lower for the Pu & MA recycling case compared to those of the standard once-through BWR case. The neutron spectra become harder with the increasing number of recycled heavy nuclides in the reactor core. The total fissile rises from 4.77% of the total nuclides number density in the reactor core for the standard once-through BWR case to 6.64% and 6.72% for the Plutonium recycling case and the Pu & MA recycling case, respectively. The two later data may become the main basis why the required uranium enrichment declines and consequently diminishes the annual loaded fuel and the required natural uranium supply. All these facts demonstrate the advantage of plutonium and minor actinides recycling in BWR.

  20. Impact investigation of reactor fuel operating parameters on reactivity for use in burnup credit applications

    Science.gov (United States)

    Sloma, Tanya Noel

    When representing the behavior of commercial spent nuclear fuel (SNF), credit is sought for the reduced reactivity associated with the net depletion of fissile isotopes and the creation of neutron-absorbing isotopes, a process that begins when a commercial nuclear reactor is first operated at power. Burnup credit accounts for the reduced reactivity potential of a fuel assembly and varies with the fuel burnup, cooling time, and the initial enrichment of fissile material in the fuel. With regard to long-term SNF disposal and transportation, tremendous benefits, such as increased capacity, flexibility of design and system operations, and reduced overall costs, provide an incentive to seek burnup credit for criticality safety evaluations. The Nuclear Regulatory Commission issued Interim Staff Guidance 8, Revision 2 in 2002, endorsing burnup credit of actinide composition changes only; credit due to actinides encompasses approximately 30% of exiting pressurized water reactor SNF inventory and could potentially be increased to 90% if fission product credit were accepted. However, one significant issue for utilizing full burnup credit, compensating for actinide and fission product composition changes, is establishing a set of depletion parameters that produce an adequately conservative representation of the fuel's isotopic inventory. Depletion parameters can have a significant effect on the isotopic inventory of the fuel, and thus the residual reactivity. This research seeks to quantify the reactivity impact on a system from dominant depletion parameters (i.e., fuel temperature, moderator density, burnable poison rod, burnable poison rod history, and soluble boron concentration). Bounding depletion parameters were developed by statistical evaluation of a database containing reactor operating histories. The database was generated from summary reports of commercial reactor criticality data. Through depletion calculations, utilizing the SCALE 6 code package, several light

  1. Preliminary Neutronic Design of High Burnup OTTO Cycle Pebble Bed Reactor

    Directory of Open Access Journals (Sweden)

    T. Setiadipura

    2015-04-01

    Full Text Available The pebble bed type High Temperature Gas-cooled Reactor (HTGR is among the interesting nuclear reactor designs in terms of safety and flexibility for co-generation applications. In addition, the strong inherent safety characteristics of the pebble bed reactor (PBR which is based on natural mechanisms improve the simplicity of the PBR design, in particular for the Once-Through-Then-Out (OTTO cycle PBR design. One of the important challenges of the OTTO cycle PBR design, and nuclear reactor design in general, is improving the nuclear fuel utilization which is shown by attaining a higher burnup value. This study performed a preliminary neutronic design study of a 200 MWt OTTO cycle PBR with high burnup while fulfilling the safety criteria of the PBR design.The safety criteria of the design was represented by the per-fuel-pebble maximum power generation of 4.5 kW/pebble. The maximum burnup value was also limited by the tested maximum burnup value which maintained the integrity of the pebble fuel. Parametric surveys were performed to obtain the optimized parameters used in this study, which are the fuel enrichment, per-pebble heavy metal (HM loading, and the average axial speed of the fuel. An optimum design with burnup value of 131.1 MWd/Kg-HM was achieved in this study which is much higher compare to the burnup of the reference design HTR-MODUL and a previously proposed OTTO-cycle PBR design. This optimum design uses 17% U-235 enrichment with 4 g HM-loading per fuel pebble

  2. Sensitivity and parametric evaluations of significant aspects of burnup credit for PWR spent fuel packages

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    1996-05-01

    Spent fuel transportation and storage cask designs based on a burnup credit approach must consider issues that are not relevant in casks designed under a fresh-fuel loading assumption. For example, the spent fuel composition must be adequately characterized and the criticality analysis model can be complicated by the need to consider axial burnup variations. Parametric analyses are needed to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models must be evaluated to determine the sensitivity of criticality safety calculations to modeling assumptions. The purpose of this report is to describe analyses and evaluations performed in order to demonstrate the effect physical parameters and modeling assumptions have on the criticality analysis of spent fuel. The analyses in this report include determination and ranking of the most important actinides and fission products; study of the effect of various depletion scenarios on subsequent criticality calculations; establishment of trends in neutron multiplication as a function of fuel enrichment, burnup, cooling time- and a parametric and modeling evaluation of three-dimensional effects (e.g., axially varying burnup and temperature/density effects) in a conceptual cask design. The sensitivity and parametric evaluations were performed with the consideration of two different burnup credit approaches: (1) only actinides in the fuel are considered in the criticality analysis, and (2) both actinides and fission products are considered. Calculations described in this report were performed using the criticality and depletion sequences available in the SCALE code system and the SCALE 27-group burnup library. Although the results described herein do not constitute a validation of SCALE for use in spent fuel analysis, independent validation efforts have been completed and are described in other reports.

  3. Review of Halden Reactor Project high burnup fuel data that can be used in safety analyses

    Energy Technology Data Exchange (ETDEWEB)

    Wiesenack, W. [OECD Halden Reactor Project (Norway)

    1996-03-01

    The fuels and materials testing programmes carried out at the OECD Halden Reactor Project are aimed at providing data in support of a mechanistic understanding of phenomena, especially as related to high burnup fuel. The investigations are focused on identifying long term property changes, and irradiation techniques and instrumentation have been developed over the years which enable to assess fuel behaviour and properties in-pile. The fuel-cladding gap has an influence on both thermal and mechanical behaviour. Improved gap conductance due to gap closure at high exposure is observed even in the case of a strong contamination with released fission gas. On the other hand, pellet-cladding mechanical interaction, which is measured with cladding elongation detectors and diameter gauges, is re-established after a phase with less interaction and is increasing. These developments are exemplified with data showing changes of fuel temperature, hydraulic diameter and cladding elongation with burnup. Fuel swelling and cladding primary and secondary creep have been successfully measured in-pile. They provide data for, e.g., the possible cladding lift-off to be accounted for at high burnup. Fuel conductivity degradation is observed as a gradual temperature increase with burnup. This affects stored heat, fission gas release and temperature dependent fuel behaviour in general. The Halden Project`s data base on fission gas release shows that the phenomenon is associated with an accumulation of gas atoms at the grain boundaries to a critical concentration before appreciable release occurs. This is accompanied by an increase of the surface-to-volume ratio measured in-pile in gas flow experiments. A typical observation at high burnup is also that a burst release of fission gas may occur during a power decrease. Gas flow and pressure equilibration experiments have shown that axial communication is severely restricted at high burnup.

  4. Post Irradiation Examination Plan for High-Burnup Demonstration Project Sister Rods

    Energy Technology Data Exchange (ETDEWEB)

    Scaglione, John M [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Montgomery, Rose [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-04-01

    This test plan describes the experimental work to be implemented by the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) to characterize high burnup (HBU) spent nuclear fuel (SNF) in conjunction with the High Burnup Dry Storage Cask Research and Development Project and serves to coordinate and integrate the multi-year experimental program to collect and develop data regarding the continued storage and eventual transport of HBU (i.e., >45 GWd/MTU) SNF. The work scope involves the development, performance, technical integration, and oversight of measurements and collection of relevant data, guided by analyses and demonstration of need.

  5. Fuel burnup calculation of Ghana MNSR using ORIGEN2 and REBUS3 codes.

    Science.gov (United States)

    Abrefah, R G; Nyarko, B J B; Fletcher, J J; Akaho, E H K

    2013-10-01

    Ghana Research Reactor-1 core is to be converted from HEU fuel to LEU fuel in the near future and managing the spent nuclear fuel is very important. A fuel depletion analysis of the GHARR-1 core was performed using ORIGEN2 and REBUS3 codes to estimate the isotopic inventory at end-of-cycle in order to help in the design of an appropriate spent fuel cask. The results obtained for both codes were consistent for U-235 burnup weight percent and Pu-239 build up as a result of burnup.

  6. Progress of the RIA experiments with high burnup fuels and their evaluation in JAERI

    Energy Technology Data Exchange (ETDEWEB)

    Ishijima, Kiyomi; Fuketa, Toyoshi [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)

    1997-01-01

    Recent results obtained in the NSRR power burst experiments with high burnup PWR fuel rods are described and discussed in this paper. Data concerning test condition, transient records during pulse irradiation and post irradiation examination are described. Another high burnup PWR fuel rod failed in the test HBO-5 at the slightly higher energy deposition than that in the test HBO-1. The failure mechanism of the test HBO-5 is the same as that of the test HBO-1, that is, hydride-assisted PCMI. Some influence of the thermocouples welding on the failure behavior of the HBO-5 rod was observed.

  7. Comparison between SERPENT and MONTEBURNS codes applied to burnup calculations of a GFR-like configuration

    Energy Technology Data Exchange (ETDEWEB)

    Chersola, Davide [GeNERG – DIME/TEC, University of Genova, via all’Opera Pia 15/a, 16145 Genova (Italy); INFN, via Dodecaneso 33, 16146 Genova (Italy); Lomonaco, Guglielmo, E-mail: guglielmo.lomonaco@unige.it [GeNERG – DIME/TEC, University of Genova, via all’Opera Pia 15/a, 16145 Genova (Italy); INFN, via Dodecaneso 33, 16146 Genova (Italy); Marotta, Riccardo [GeNERG – DIME/TEC, University of Genova, via all’Opera Pia 15/a, 16145 Genova (Italy); INFN, via Dodecaneso 33, 16146 Genova (Italy); Mazzini, Guido [Centrum výzkumu Řež (Research Centre Rez), Husinec-Rez, cp. 130, 25068 Rez (Czech Republic)

    2014-07-01

    Highlights: • MC codes are widely adopted to analyze nuclear facilities, including GEN-IV reactors. • Burnup calculations are an efficient tool to test neutronic Monte Carlo codes. • In this comparison the used codes show some differences but a good agreement exists. - Abstract: This paper presents the comparison between two Monte Carlo based burnup codes: SERPENT and MONTEBURNS. Monte Carlo codes are fully and worldwide adopted to perform analyses on nuclear facilities, also in the frame of Generation IV advanced reactors simulations. Thus, faster and most powerful calculation codes are needed with the aim to analyze complex geometries and specific neutronic behaviors. Burnup calculations are an efficient tool to test neutronic Monte Carlo codes: indeed these calculations couple transport and depletion procedures, so that neutronic reactor behavior can be simulated in its totality. Comparisons have been performed on a configuration representing the Allegro MOX 75 MW{sub th} reactor proposed by the European GoFastR (Gas-cooled Fast Reactor) Project in the frame of the 7th Euratom Framework Program. Although in burnup and criticality comparisons the codes used in simulations show different calculation times and some differences in amounts and in precision (in term of statistical errors), a reasonably good agreement between them exists.

  8. Computational simulation of fuel burnup estimation for research reactors plate type

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Nadia Rodrigues dos, E-mail: nadiasam@gmail.com [Instituto Federal de Educacao, Ciencia e Tecnologia do Rio de Janeiro (IFRJ), Paracambi, RJ (Brazil); Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes, E-mail: zrlima@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    The aim of this study is to estimate the spatial fuel burnup, through computational simulation, in two research reactors plate type, loaded with dispersion fuel: the benchmark Material Test Research - International Atomic Energy Agency (MTR-IAEA) and a typical multipurpose reactor (MR). The first composed of plates with uranium oxide dispersed in aluminum (UAlx-Al) and a second composed with uranium silicide (U{sub 3}Si{sub 2}) dispersed in aluminum. To develop this work we used the deterministic code, WIMSD-5B, which performs the cell calculation solving the neutron transport equation, and the DF3DQ code, written in FORTRAN, which solves the three-dimensional neutron diffusion equation using the finite difference method. The methodology used was adequate to estimate the spatial fuel burnup , as the results was in accordance with chosen benchmark, given satisfactorily to the proposal presented in this work, even showing the possibility to be applied to other research reactors. For future work are suggested simulations with other WIMS libraries, other settings core and fuel types. Comparisons the WIMSD-5B results with programs often employed in fuel burnup calculations and also others commercial programs, are suggested too. Another proposal is to estimate the fuel burnup, taking into account the thermohydraulics parameters and the Xenon production. (author)

  9. Fission Gas Release in LWR Fuel Rods Exhibiting Very High Burn-Up

    DEFF Research Database (Denmark)

    Carlsen, H.

    1980-01-01

    Two UO2Zr BWR type test fuel rods were irradiated to a burn-up of about 38000 MWd/tUO2. After non-destructive characterization, the fission gas released to the internal free volume was extracted and analysed. The irradiation was simulated by means of the Danish fuel performance code WAFER-2, which...

  10. Reactivity effect of spent fuel depending on burn-up history

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, Takafumi [Nagoya Univ., Nagoya, Aichi (Japan); Suyama, Kenya; Nomura, Yasushi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Mochizuki, Hiroki [The Japan Research Institute, Ltd., Tokyo (Japan)

    2001-06-01

    It is well known that a composition of spent fuel depends on various parameter changes throughout a burn-up period. In this study we aimed at the boron concentration and its change, the coolant temperature and its spatial distribution, the specific power, the operation mode, and the duration of inspection, because the effects due to these parameters have not been analyzed in detail. The composition changes of spent fuel were calculated by using the burn-up code SWAT, when the parameters mentioned above varied in the range of actual variations. Moreover, to estimate the reactivity effect caused by the composition changes, the criticality calculations for an infinite array of spent fuel were carried out with computer codes SRAC95 or MVP. In this report the reactivity effects were arranged from the viewpoint of what parameters gave more positive reactivity effect. The results obtained through this study are useful to choose the burn-up calculation model when we take account of the burn-up credit in the spent fuel management. (author)

  11. Development of an MCNP-tally based burnup code and validation through PWR benchmark exercises

    Energy Technology Data Exchange (ETDEWEB)

    El Bakkari, B. [ERSN-LMR, Department of physics, Faculty of Sciences P.O.Box 2121, Tetuan (Morocco)], E-mail: bakkari@gmail.com; El Bardouni, T.; Merroun, O.; El Younoussi, Ch.; Boulaich, Y. [ERSN-LMR, Department of physics, Faculty of Sciences P.O.Box 2121, Tetuan (Morocco); Chakir, E. [EPTN-LPMR, Faculty of Sciences Kenitra (Morocco)

    2009-05-15

    The aim of this study is to evaluate the capabilities of a newly developed burnup code called BUCAL1. The code provides the full capabilities of the Monte Carlo code MCNP5, through the use of the MCNP tally information. BUCAL1 uses the fourth order Runge Kutta method with the predictor-corrector approach as the integration method to determine the fuel composition at a desired burnup step. Validation of BUCAL1 was done by code vs. code comparison. Results of two different kinds of codes are employed. The first one is CASMO-4, a deterministic multi-group two-dimensional transport code. The second kind is MCODE and MOCUP, a link MCNP-ORIGEN codes. These codes use different burnup algorithms to solve the depletion equations system. Eigenvalue and isotope concentrations were compared for two PWR uranium and thorium benchmark exercises at cold (300 K) and hot (900 K) conditions, respectively. The eigenvalue comparison between BUCAL1 and the aforementioned two kinds of codes shows a good prediction of the systems'k-inf values during the entire burnup history, and the maximum difference is within 2%. The differences between the BUCAL1 isotope concentrations and the predictions of CASMO-4, MCODE and MOCUP are generally better, and only for a few sets of isotopes these differences exceed 10%.

  12. Dependence of heavy metal burnup on nuclear data libraries for fast reactors

    CERN Document Server

    Ohki, S

    2003-01-01

    Japan Nuclear Cycle Development Institute (JNC) is considering the highly burnt fuel as well as the recycling of minor actinide (MA) in the development of commercialized fast reactor cycle systems. Higher accuracy in burnup calculation is going to be required for higher mass plutonium isotopes ( sup 2 sup 4 sup 0 Pu, etc.) and MA nuclides. In the framework of research and development aiming at the validation and necessary improvements of fast reactor burnup calculation, we investigated the differences among the burnup calculation results with the major nuclear data libraries: JEF-2.2, ENDF/B-VI Release 5, JENDL-3.2, and JENDL-3.3. We focused on the heavy metal nuclides such as plutonium and MA in the central core region of a conventional sodium-cooled fast reactor. For main heavy metal nuclides ( sup 2 sup 3 sup 5 U, sup 2 sup 3 sup 8 U, sup 2 sup 3 sup 9 Pu, sup 2 sup 4 sup 0 Pu, and sup 2 sup 4 sup 1 Pu), number densities after 1-cycle burnup did not change over one or two percent. Library dependence was re...

  13. Oxygen potential measurements in high burnup LWR U0 2 fuel

    Science.gov (United States)

    Matzke, Hj.

    1995-05-01

    A miniature solid state galvanic cell was used to measure the oxygen potential Δ overlineG( O2) of reactor irradiated U0 2 fuel at different burnups in the range of 28 to ⩾ 150 GWd d/t M. This very high burnup was achieved in the rim region of a fuel with a cross section average burnup of 75 GWd d/t M. The fuels had different enrichments and therefore different contributions of fission of 235U and 239Pu. The temperature range covered was 900 to 1350 K. None of the fuels showed a significant oxidation. Rather, if allowance is made for the dissolved rare earth fission products and the Pu formed during irradiation, some of the fuels were very slightly substoichiometric and the highest possible degree of oxidation corresponded to U0 2.001. In general, the Δ overlineG( O2) at 750°C was about -400 kJ/mol, corresponding to the Δ overlineG( O2) of the reaction Mo + O 2 → MoO 2. The implication of these results which are in contrast to commonly assumed ideas that U0 2 fuel oxidizes due to burnup, are discussed and the importance of the fission product Mo and of the zircaloy clad as oxygen buffers is outlined.

  14. Thermal properties of U–Mo alloys irradiated to moderate burnup and power

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas E., E-mail: Douglas.Burkes@pnnl.gov; Casella, Andrew M.; Casella, Amanda J.; Buck, Edgar C.; Pool, Karl N.; MacFarlan, Paul J.; Edwards, Matthew K.; Smith, Frances N.

    2015-09-15

    Highlights: • Thermal properties of irradiated U–Mo alloy monolithic fuel samples were measured. • Density, thermal diffusivity, and thermal conductivity are influenced by increasing burnup. • U–Mo chemistry and specific heat capacity was not as sensitive to increasing burnup. • Thermal conductivity decreased approximately 45% for a fission density of 4.52 × 10{sup 21} fissions cm{sup −3} at 200 °C. • An empirical model developed previously agrees well with the experimental measurements. - Abstract: A variety of physical and thermal property measurements as a function of temperature and fission density were performed on irradiated U–Mo alloy monolithic fuel samples with a Zr diffusion barrier and clad in aluminum alloy 6061. The U–Mo alloy density, thermal diffusivity, and thermal conductivity are strongly influenced by increasing burnup, mainly as the result of irradiation induced recrystallization and fission gas bubble formation and coalescence. U–Mo chemistry, specifically Mo content, and specific heat capacity was not as sensitive to increasing burnup. Measurements indicated that thermal conductivity of the U–Mo alloy decreased approximately 30% for a fission density of 3.30 × 10{sup 21} fissions cm{sup −3} and approximately 45% for a fission density of 4.52 × 10{sup 21} fissions cm{sup −3} from unirradiated values at 200 °C. An empirical thermal conductivity degradation model developed previously and summarized here agrees well with the experimental measurements.

  15. Radiochemical Assays of Irradiated VVER-440 Fuel for Use in Spent Fuel Burnup Credit Activities

    Energy Technology Data Exchange (ETDEWEB)

    Jardine, L J

    2005-04-25

    The objective of this spent fuel burnup credit work was to study and describe a VVER-440 reactor spent fuel assembly (FA) initial state before irradiation, its operational irradiation history and the resulting radionuclide distribution in the fuel assembly after irradiation. This work includes the following stages: (1) to pick out and select a specific spent (irradiated) FA for examination; (2) to describe the FA initial state before irradiation; (3) to describe the irradiation history, including thermal calculations; (4) to examine the burnup distribution of select radionuclides along the FA height and cross-section; (5) to examine the radionuclide distributions; (6) to determine the Kr-85 release into the plenum; (7) to select and prepare FA rod specimens for destructive examinations; (8) to determine the radionuclide compositions, isotope masses and burnup in the rod specimens; and (9) to analyze, document and process the results. The specific workscope included the destructive assay (DA) of spent fuel assembly rod segments with an {approx}38.5 MWd/KgU burnup from a single VVER-440 fuel assembly from the Novovorenezh reactor in Russia. Based on irradiation history criteria, four rods from the fuel assembly were selected and removed from the assembly for examination. Next, 8 sections were cut from the four rods and sent for destructive analysis of radionuclides by radiochemical analyses. The results were documented in a series of seven reports over a period of {approx}1 1/2 years.

  16. A simple gamma spectrometry method for evaluating the burnup of MTR-type HEU fuel elements

    Science.gov (United States)

    Makmal, T.; Aviv, O.; Gilad, E.

    2016-10-01

    A simple method for the evaluation of the burnup of a materials testing reactor (MTR) fuel element by gamma spectrometry is presented. The method was applied to a highly enriched uranium MTR nuclear fuel element that was irradiated in a 5 MW pool-type research reactor for a total period of 34 years. The experimental approach is based on in-situ measurements of the MTR fuel element in the reactor pool by a portable high-purity germanium detector located in a gamma cell. To corroborate the method, analytical calculations (based on the irradiation history of the fuel element) and computer simulations using a dedicated fuel cycle burnup code ORIGEN2 were performed. The burnup of the MTR fuel element was found to be 52.4±8.8%, which is in good agreement with the analytical calculations and the computer simulations. The method presented here is suitable for research reactors with either a regular or an irregular irradiation regime and for reactors with limited infrastructure and/or resources. In addition, its simplicity and the enhanced safety it confers may render this method suitable for IAEA inspectors in fuel element burnup assessments during on-site inspections.

  17. Parametric Study of the Effect of Burnable Poison Rods for PWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J.C.

    2001-09-28

    The Interim Staff Guidance on burnup credit (ISG-8) issued by the United States Nuclear Regulatory Commission's (U.S. NRC) Spent Fuel Project Office recommends restricting the use of burnup credit to assemblies that have not used burnable absorbers. This recommended restriction eliminates a large portion of the currently discharged spent fuel assemblies from cask loading, and thus severely limits the practical usefulness of burnup credit. In the absence of readily available information on burnable poison rod (BPR) design specifications and usage in U.S. pressurized-water-reactors (PWRs), and the subsequent reactivity effect of BPR exposure on discharged spent nuclear fuel (SNF), NRC staff has indicated a need for additional information in these areas. In response, this report presents a parametric study of the effect of BPR exposure on the reactivity of SNF for various BPR designs, fuel enrichments, and exposure conditions, and documents BPR design specifications. Trends in the reactivity effects of BPRs are established with infinite pin-cell and assembly array calculations with the SCALE and HELIOS code packages, respectively. Subsequently, the reactivity effects of BPRs for typical initial enrichment and burnup combinations are quantified based on three-dimensional (3-D) KENO V.a Monte Carlo calculations with a realistic rail-type cask designed for burnup credit. The calculations demonstrate that the positive reactivity effect due to BPR exposure increases nearly linearly with burnup and is dependent on the number, poison loading, and design of the BPRs and the initial fuel enrichment. Expected typical reactivity increases, based on one-cycle BPR exposure, were found to be less than 1% {Delta}k. Based on the presented analysis, guidance is offered on an appropriate approach for calculating bounding SNF isotopic data for assemblies exposed to BPRs. Although the analyses do not address the issue of validation of depletion methods for assembly designs with BPRs

  18. Application of Genetic Algorithm methodologies in fuel bundle burnup optimization of Pressurized Heavy Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jayalal, M.L., E-mail: jayalal@igcar.gov.in [Electronics, Instrumentation and Radiological Safety Group (EIRSG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India); Ramachandran, Suja [Electronics, Instrumentation and Radiological Safety Group (EIRSG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India); Rathakrishnan, S. [Reactor Physics Section, Madras Atomic Power Station (MAPS), Kalpakkam, Tamil Nadu (India); Satya Murty, S.A.V. [Electronics, Instrumentation and Radiological Safety Group (EIRSG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India); Sai Baba, M. [Resources Management Group (RMG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India)

    2015-01-15

    Highlights: • We study and compare Genetic Algorithms (GA) in the fuel bundle burnup optimization of an Indian Pressurized Heavy Water Reactor (PHWR) of 220 MWe. • Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are considered. • For the selected problem, Multi Objective GA performs better than Penalty Functions based GA. • In the present study, Multi Objective GA outperforms Penalty Functions based GA in convergence speed and better diversity in solutions. - Abstract: The work carried out as a part of application and comparison of GA techniques in nuclear reactor environment is presented in the study. The nuclear fuel management optimization problem selected for the study aims at arriving appropriate reference discharge burnup values for the two burnup zones of 220 MWe Pressurized Heavy Water Reactor (PHWR) core. Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are applied in this study. The study reveals, for the selected problem of PHWR fuel bundle burnup optimization, Multi Objective GA is more suitable than Penalty Functions based GA in the two aspects considered: by way of producing diverse feasible solutions and the convergence speed being better, i.e. it is capable of generating more number of feasible solutions, from earlier generations. It is observed that for the selected problem, the Multi Objective GA is 25.0% faster than Penalty Functions based GA with respect to CPU time, for generating 80% of the population with feasible solutions. When average computational time of fixed generations are considered, Penalty Functions based GA is 44.5% faster than Multi Objective GA. In the overall performance, the convergence speed of Multi Objective GA surpasses the computational time advantage of Penalty Functions based GA. The ability of Multi Objective GA in producing more diverse feasible solutions is a desired feature of the problem selected, that helps the

  19. OECD/NEA Burnup Credit Calculational Criticality Benchmark Phase I-B Results

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    1993-01-01

    Burnup credit is an ongoing technical concern for many countries that operate commercial nuclear power reactors. In a multinational cooperative effort to resolve burnup credit issues, a Burnup Credit Working Group has been formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development. This working group has established a set of well-defined calculational benchmarks designed to study significant aspects of burnup credit computational methods. These benchmarks are intended to provide a means for the intercomparison of computer codes, methods, and data applied in spent fuel analysis. The benchmarks have been divided into multiple phases, each phase focusing on a particular feature of burnup credit analysis. This report summarizes the results and findings of the Phase I-B benchmark, which was proposed to provide a comparison of the ability of different code systems and data libraries to perform depletion analysis for the prediction of spent fuel isotopic concentrations. Results included here represent 21 different sets of calculations submitted by 16 different organizations worldwide, and are based on a limited set of nuclides determined to have the most important effect on the neutron multiplication factor of light-water-reactor spent fuel. A comparison of all sets of results demonstrates that most methods are in agreement to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods are within 11% agreement about the average for all fission products studied. Furthermore, most deviations are less than 10%, and many are less than 5%. The exceptions are {sup 149}Sm, {sup 151}Sm, and {sup 155}Gd.

  20. OECD/NEA burnup credit calculational criticality benchmark Phase I-B results

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.; Parks, C.V. [Oak Ridge National Lab., TN (United States); Brady, M.C. [Sandia National Labs., Las Vegas, NV (United States)

    1996-06-01

    In most countries, criticality analysis of LWR fuel stored in racks and casks has assumed that the fuel is fresh with the maximum allowable initial enrichment. This assumption has led to the design of widely spaced and/or highly poisoned storage and transport arrays. If credit is assumed for fuel burnup, initial enrichment limitations can be raised in existing systems, and more compact and economical arrays can be designed. Such reliance on the reduced reactivity of spent fuel for criticality control is referred to as burnup credit. The Burnup Credit Working Group, formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development, has established a set of well-defined calculational benchmarks designed to study significant aspects of burnup credit computational methods. These benchmarks are intended to provide a means for the intercomparison of computer codes, methods, and data applied in spent fuel analysis. The benchmarks have been divided into multiple phases, each phase focusing on a particular feature of burnup credit analysis. This report summarizes the results and findings of the Phase I-B benchmark, which was proposed to provide a comparison of the ability of different code systems and data libraries to perform depletion analysis for the prediction of spent fuel isotopic concentrations. Results included here represent 21 different sets of calculations submitted by 16 different organizations worldwide and are based on a limited set of nuclides determined to have the most important effect on the neutron multiplication factor of light-water-reactor spent fuel. A comparison of all sets of results demonstrates that most methods agree to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods agree within 11% about the average for all fission products studied. Most deviations are less than 10%, and many are less than 5%. The exceptions are Sm 149, Sm 151, and Gd 155.

  1. Propagation of statistical and nuclear data uncertainties in Monte Carlo burn-up calculations

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Herranz, Nuria [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid, UPM (Spain)], E-mail: nuria@din.upm.es; Cabellos, Oscar [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid, UPM (Spain); Sanz, Javier [Departamento de Ingenieria Energetica, Universidad Nacional de Educacion a Distancia, UNED (Spain); Juan, Jesus [Laboratorio de Estadistica, Universidad Politecnica de Madrid, UPM (Spain); Kuijper, Jim C. [NRG - Fuels, Actinides and Isotopes Group, Petten (Netherlands)

    2008-04-15

    Two methodologies to propagate the uncertainties on the nuclide inventory in combined Monte Carlo-spectrum and burn-up calculations are presented, based on sensitivity/uncertainty and random sampling techniques (uncertainty Monte Carlo method). Both enable the assessment of the impact of uncertainties in the nuclear data as well as uncertainties due to the statistical nature of the Monte Carlo neutron transport calculation. The methodologies are implemented in our MCNP-ACAB system, which combines the neutron transport code MCNP-4C and the inventory code ACAB. A high burn-up benchmark problem is used to test the MCNP-ACAB performance in inventory predictions, with no uncertainties. A good agreement is found with the results of other participants. This benchmark problem is also used to assess the impact of nuclear data uncertainties and statistical flux errors in high burn-up applications. A detailed calculation is performed to evaluate the effect of cross-section uncertainties in the inventory prediction, taking into account the temporal evolution of the neutron flux level and spectrum. Very large uncertainties are found at the unusually high burn-up of this exercise (800 MWd/kgHM). To compare the impact of the statistical errors in the calculated flux with respect to the cross uncertainties, a simplified problem is considered, taking a constant neutron flux level and spectrum. It is shown that, provided that the flux statistical deviations in the Monte Carlo transport calculation do not exceed a given value, the effect of the flux errors in the calculated isotopic inventory are negligible (even at very high burn-up) compared to the effect of the large cross-section uncertainties available at present in the data files.

  2. High Burnup Dry Storage Cask Research and Development Project, Final Test Plan

    Energy Technology Data Exchange (ETDEWEB)

    None

    2014-02-27

    EPRI is leading a project team to develop and implement the first five years of a Test Plan to collect data from a SNF dry storage system containing high burnup fuel.12 The Test Plan defined in this document outlines the data to be collected, and the storage system design, procedures, and licensing necessary to implement the Test Plan.13 The main goals of the proposed test are to provide confirmatory data14 for models, future SNF dry storage cask design, and to support license renewals and new licenses for ISFSIs. To provide data that is most relevant to high burnup fuel in dry storage, the design of the test storage system must mimic real conditions that high burnup SNF experiences during all stages of dry storage: loading, cask drying, inert gas backfilling, and transfer to the ISFSI for multi-year storage.15 Along with other optional modeling, SETs, and SSTs, the data collected in this Test Plan can be used to evaluate the integrity of dry storage systems and the high burnup fuel contained therein over many decades. It should be noted that the Test Plan described in this document discusses essential activities that go beyond the first five years of Test Plan implementation.16 The first five years of the Test Plan include activities up through loading the cask, initiating the data collection, and beginning the long-term storage period at the ISFSI. The Test Plan encompasses the overall project that includes activities that may not be completed until 15 or more years from now, including continued data collection, shipment of the Research Project Cask to a Fuel Examination Facility, opening the cask at the Fuel Examination Facility, and examining the high burnup fuel after the initial storage period.

  3. Behaviour of fission gas in the rim region of high burn-up UO 2 fuel pellets with particular reference to results from an XRF investigation

    Science.gov (United States)

    Mogensen, M.; Pearce, J. H.; Walker, C. T.

    1999-01-01

    XRF and EPMA results for retained xenon from Battelle's high burn-up effects program are re-evaluated. The data reviewed are from commercial low enriched BWR fuel with burn-ups of 44.8-54.9 GWd/tU and high enriched PWR fuel with burn-ups from 62.5 to 83.1 GWd/tU. It is found that the high burn-up structure penetrated much deeper than initially reported. The local burn-up threshold for the formation of the high burn-up structure in those fuels with grain sizes in the normal range lay between 60 and 75 GWd/tU. The high burn-up structure was not detected by EPMA in a fuel that had a grain size of 78 μm although the local burn-up at the pellet rim had exceeded 80 GWd/tU. It is concluded that fission gas had been released from the high burn-up structure in three PWR fuel sections with burn-ups of 70.4, 72.2 and 83.1 GWd/tU. In the rim region of the last two sections at the locations where XRF indicated gas release the local burn-up was higher than 75 GWd/tU.

  4. Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, William BJ J [ORNL; Ade, Brian J [ORNL; Bowman, Stephen M [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL; Mertyurek, Ugur [ORNL; Radulescu, Georgeta [ORNL

    2015-01-01

    Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (keff) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of lattice design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15–20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup

  5. Burnup calculation by the method of first-flight collision probabilities using average chords prior to the first collision

    Science.gov (United States)

    Karpushkin, T. Yu.

    2012-12-01

    A technique to calculate the burnup of materials of cells and fuel assemblies using the matrices of first-flight neutron collision probabilities rebuilt at a given burnup step is presented. A method to rebuild and correct first collision probability matrices using average chords prior to the first neutron collision, which are calculated with the help of geometric modules of constructed stochastic neutron trajectories, is described. Results of calculation of the infinite multiplication factor for elementary cells with a modified material composition compared to the reference one as well as calculation of material burnup in the cells and fuel assemblies of a VVER-1000 are presented.

  6. Development of a Burnup Module DECBURN Based on the Krylov Subspace Method

    Energy Technology Data Exchange (ETDEWEB)

    Cho, J. Y.; Kim, K. S.; Shim, H. J.; Song, J. S

    2008-05-15

    This report is to develop a burnup module DECBURN that is essential for the reactor analysis and the assembly homogenization codes to trace the fuel composition change during the core burnup. The developed burnup module solves the burnup equation by the matrix exponential method based on the Krylov Subspace method. The final solution of the matrix exponential is obtained by the matrix scaling and squaring method. To develop DECBURN module, this report includes the followings as: (1) Krylov Subspace Method for Burnup Equation, (2) Manufacturing of the DECBURN module, (3) Library Structure Setup and Library Manufacturing, (4) Examination of the DECBURN module, (5) Implementation to the DeCART code and Verification. DECBURN library includes the decay constants, one-group cross section and the fission yields. Examination of the DECBURN module is performed by manufacturing a driver program, and the results of the DECBURN module is compared with those of the ORIGEN program. Also, the implemented DECBURN module to the DeCART code is applied to the LWR depletion benchmark and a OPR-1000 pin cell problem, and the solutions are compared with the HELIOS code to verify the computational soundness and accuracy. In this process, the criticality calculation method and the predictor-corrector scheme are introduced to the DeCART code for a function of the homogenization code. The examination by a driver program shows that the DECBURN module produces exactly the same solution with the ORIGEN program. DeCART code that equips the DECBURN module produces a compatible solution to the other codes for the LWR depletion benchmark. Also the multiplication factors of the DeCART code for the OPR-1000 pin cell problem agree to the HELIOS code within 100 pcm over the whole burnup steps. The multiplication factors with the criticality calculation are also compatible with the HELIOS code. These results mean that the developed DECBURN module works soundly and produces an accurate solution

  7. FRAPCON-3: Modifications to fuel rod material properties and performance models for high-burnup application

    Energy Technology Data Exchange (ETDEWEB)

    Lanning, D.D.; Beyer, C.E.; Painter, C.L.

    1997-12-01

    This volume describes the fuel rod material and performance models that were updated for the FRAPCON-3 steady-state fuel rod performance code. The property and performance models were changed to account for behavior at extended burnup levels up to 65 Gwd/MTU. The property and performance models updated were the fission gas release, fuel thermal conductivity, fuel swelling, fuel relocation, radial power distribution, solid-solid contact gap conductance, cladding corrosion and hydriding, cladding mechanical properties, and cladding axial growth. Each updated property and model was compared to well characterized data up to high burnup levels. The installation of these properties and models in the FRAPCON-3 code along with input instructions are provided in Volume 2 of this report and Volume 3 provides a code assessment based on comparison to integral performance data. The updated FRAPCON-3 code is intended to replace the earlier codes FRAPCON-2 and GAPCON-THERMAL-2. 94 refs., 61 figs., 9 tabs.

  8. Draft evaluation of the frequency for gas sampling for the high burnup confirmatory data project

    Energy Technology Data Exchange (ETDEWEB)

    Stockman, Christine T. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Alsaed, Halim A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-03-26

    This report fulfills the M3 milestone M3FT-15SN0802041, “Draft Evaluation of the Frequency for Gas Sampling for the High Burn-up Storage Demonstration Project” under Work Package FT-15SN080204, “ST Field Demonstration Support – SNL”. This report provides a technically based gas sampling frequency strategy for the High Burnup (HBU) Confirmatory Data Project. The evaluation of: 1) the types and magnitudes of gases that could be present in the project cask and, 2) the degradation mechanisms that could change gas compositions culminates in an adaptive gas sampling frequency strategy. This adaptive strategy is compared against the sampling frequency that has been developed based on operational considerations. Gas sampling will provide information on the presence of residual water (and byproducts associated with its reactions and decomposition) and breach of cladding, which could inform the decision of when to open the project cask.

  9. Determination of deuterium–tritium critical burn-up parameter by four temperature theory

    Energy Technology Data Exchange (ETDEWEB)

    Nazirzadeh, M.; Ghasemizad, A. [Department of Physics, University of Guilan, 41335-1914 Rasht (Iran, Islamic Republic of); Khanbabei, B. [School of Physics, Damghan University, 36716-41167 Damghan (Iran, Islamic Republic of)

    2015-12-15

    Conditions for thermonuclear burn-up of an equimolar mixture of deuterium-tritium in non-equilibrium plasma have been investigated by four temperature theory. The photon distribution shape significantly affects the nature of thermonuclear burn. In three temperature model, the photon distribution is Planckian but in four temperature theory the photon distribution has a pure Planck form below a certain cut-off energy and then for photon energy above this cut-off energy makes a transition to Bose-Einstein distribution with a finite chemical potential. The objective was to develop four temperature theory in a plasma to calculate the critical burn up parameter which depends upon initial density, the plasma components initial temperatures, and hot spot size. All the obtained results from four temperature theory model are compared with 3 temperature model. It is shown that the values of critical burn-up parameter calculated by four temperature theory are smaller than those of three temperature model.

  10. Experience with incomplete control rod insertion in fuel with burnup exceeding approximately 40 GWD/MTU

    Energy Technology Data Exchange (ETDEWEB)

    Kee, E. [Houston Lighting & Power Co., Wadworth, TX (United States)

    1997-01-01

    Analysis and measurement experience with fuel assemblies having incomplete control rod insertion at burnups of approximately 40 GWD/MTU is presented. Control rod motion dynamics and simplified structural analyses are presented and compared to measurement data. Fuel assembly growth measurements taken with the plant Refueling Machine Z-Tape are described and presented. Bow measurements (including plug gauging) are described and potential improvements are suggested. The measurements described and analysis performed show that sufficient guide tube bow (either from creep or yield buckling) is present in some high burnup assemblies to stop the control rods before they reach their full limit of travel. Recommendations are made that, if implemented, could improve cost performance related to testing and analysis activities.

  11. High burnup fuel behavior related to fission gas effects under reactivity initiated accidents (RIA) conditions

    Science.gov (United States)

    Lemoine, F.

    1997-09-01

    Specific aspects of irradiated fuel result from the increasing retention of gaseous and volatile fission products with burnup, which, under overpower conditions, can lead to solid fuel pressurization and swelling causing severe PCMI (pellet clad mechanical interaction). In order to assess the reliability of high burnup fuel under RIAs, experimental programs have been initiated which have provided important data concerning the transient fission gas behavior and the clad loading mechanisms. The importance of the rim zone is demonstrated based on three experiments resulting in clad failure at low enthalpy, which are explained by energetic considerations. High gas release in non-failure tests with low energy deposition underlines the importance of grain boundary and porosity gas. Measured final releases are strongly correlated to the microstructure evolution, depending on energy deposition, pulse width, initial and refabricated fuel rod design. Observed helium release can also increase internal pressure and gives hints to the gas behavior understanding.

  12. Thermal properties of U–Mo alloys irradiated to moderate burnup and power

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas E.; Casella, Andrew M.; Casella, Amanda J.; Buck, Edgar C.; Pool, Karl N.; MacFarlan, Paul J.; Edwards, Matthew K.; Smith, Frances N.

    2015-09-01

    A variety of physical and thermal property measurements as a function of temperature and fission density were performed on irradiated U-Mo alloy monolithic fuel samples with a Zr diffusion barrier and clad in aluminum alloy 6061. The U-Mo alloy density, thermal diffusivity, and thermal conductivity are strongly influenced by increasing burnup, mainly as the result of irradiation induced recrystallization and fission gas bubble formation and coalescence. U-Mo chemistry, specifically Mo content, and specific heat capacity was not as sensitive to increasing burnup. Measurements indicated that thermal conductivity of the U-Mo alloy decreased approximately 30% for a fission density of 2.88 × 1021 fissions cm-3 and approximately 45% for a fission density of 4.08 × 1021 fissions cm-3 from unirradiated values at 200 oC. An empirical thermal conductivity degradation model developed previously and summarized here agrees well with the experimental measurements.

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

    Directory of Open Access Journals (Sweden)

    E.K. Boafo

    2013-02-01

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

  14. Burn-up credit in criticality safety of PWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Mahmoud, Rowayda F., E-mail: Rowayda_mahmoud@yahoo.com [Metallurgy Department, Nuclear Research Center, Atomic Energy Authority (Egypt); Shaat, Mohamed K. [Nuclear Engineering, Reactors Department, Nuclear Research Center, Atomic Energy Authority (Egypt); Nagy, M.E.; Agamy, S.A. [Professor of Nuclear Engineering, Nuclear and Radiation Department, Alexandria University (Egypt); Abdelrahman, Adel A. [Metallurgy Department, Nuclear Research Center, Atomic Energy Authority (Egypt)

    2014-12-15

    Highlights: • Designing spent fuel wet storage using WIMS-5D and MCNP-5 code. • Studying fresh and burned fuel with/out absorber like “B{sub 4}C and Ag–In–Cd” in racks. • Sub-criticality was confirmed for fresh and burned fuel under specific cases. • Studies for BU credit recommend increasing fuel burn-up to 60.0 GWD/MTU. • Those studies require new core structure materials, fuel composition and cladding. - Abstract: The criticality safety calculations were performed for a proposed design of a wet spent fuel storage pool. This pool will be used for the storage of spent fuel discharged from a typical pressurized water reactor (PWR). The mathematical model based on the international validated codes, WIMS-5 and MCNP-5 were used for calculating the effective multiplication factor, k{sub eff}, for the spent fuel stored in the pool. The data library for the multi-group neutron microscopic cross-sections was used for the cell calculations. The k{sub eff} was calculated for several changes in water density, water level, assembly pitch and burn-up with different initial fuel enrichment and new types and amounts of fixed absorbers. Also, k{sub eff} was calculated for the conservative fresh fuel case. The results of the calculations confirmed that the effective multiplication factor for the spent fuel storage is sub-critical for all normal and abnormal states. The future strategy for the burn-up credit recommends increasing the fuel burn-up to a value >60.0 GWD/MTU, which requires new fuel composition and new fuel cladding material with the assessment of the effects of negative reactivity build up.

  15. S∧4 Reactor: Operating Lifetime and Estimates of Temperature and Burnup Reactivity Coefficients

    Science.gov (United States)

    King, Jeffrey C.; El-Genk, Mohamed S.

    2006-01-01

    The S∧4 reactor has a sectored, Mo-14%Re solid core for avoidance of single point failures in reactor cooling and Closed Brayton Cycle (CBC) energy conversion. The reactor is loaded with UN fuel, cooled with a He-Xe gas mixture at ~1200 K and operates at steady thermal power of 550 kW. Following a launch abort accident, the axial and radial BeO reflectors easily disassemble upon impact so that the bare reactor is subcriticial when submerged in wet sand or seawater and the core voids are filled with seawater. Spectral Shift Absorber (SSA) additives have been shown to increase the UN fuel enrichment and significantly reduce the total mass of the reactor. This paper investigates the effects of SSA additions on the temperature and burnup reactivity coefficients and the operational lifetime of the S∧4 reactor. SSAs slightly decrease the temperature reactivity feedback coefficient, but significantly increase the operating lifetime by decreasing the burnup reactivity coefficient. With no SSAs, fuel enrichment is only 58.5 wt% and the estimated operating lifetime is the shortest (7.6 years) with the highest temperature and burnup reactivity feedback coefficients (-0.2709 ¢/K and -1.3470 $/atom%). With europium-151 and gadolinium-155 additions, the enrichment (91.5 and 94 wt%) and operating lifetime (9.9 and 9.8 years) of the S∧4 reactor are the highest while the temperature and burnup reactivity coefficients (-0.2382 and -0.2447 ¢/K -0.9073 and 0.8502 $/atom%) are the lowest.

  16. Analysis of the effect of UO{sub 2} high burnup microstructure on fission gas release

    Energy Technology Data Exchange (ETDEWEB)

    Jernkvist, Lars Olof; Massih, Ali [Quantum Technologies AB, Uppsala Science Park (Sweden)

    2002-10-01

    This report deals with high-burnup phenomena with relevance to fission gas release from UO{sub 2} nuclear fuel. In particular, we study how the fission gas release is affected by local buildup of fissile plutonium isotopes and fission products at the fuel pellet periphery, with subsequent formation of a characteristic high-burnup rim zone micro-structure. An important aspect of these high-burnup effects is the degradation of fuel thermal conductivity, for which prevalent models are analysed and compared with respect to their theoretical bases and supporting experimental data. Moreover, the Halden IFA-429/519.9 high-burnup experiment is analysed by use of the FRAPCON3 computer code, into which modified and extended models for fission gas release are introduced. These models account for the change in Xe/Kr-ratio of produced and released fission gas with respect to time and space. In addition, several alternative correlations for fuel thermal conductivity are implemented, and their impact on calculated fission gas release is studied. The calculated fission gas release fraction in IFA-429/519.9 strongly depends on what correlation is used for the fuel thermal conductivity, since thermal release dominates over athermal release in this particular experiment. The conducted calculations show that athermal release processes account for less than 10% of the total gas release. However, athermal release from the fuel pellet rim zone is presumably underestimated by our models. This conclusion is corroborated by comparisons between measured and calculated Xe/Kr-ratios of the released fission gas.

  17. Advanced Corrosion-Resistant Zr Alloys for High Burnup and Generation IV Applications

    Energy Technology Data Exchange (ETDEWEB)

    Arthur Motta; Yong Hwan Jeong; R.J. Comstock; G.S. Was; Y.S. Kim

    2006-10-31

    The objective of this collaboration between four institutions in the US and Korea is to demonstrate a technical basis for the improvement of the corrosion resistance of zirconium-based alloys in more extreme operating environments (such as those present in severe fuel duty,cycles (high burnup, boiling, aggressive chemistry) andto investigate the feasibility (from the point of view of corrosion rate) of using advanced zirconium-based alloys in a supercritical water environment.

  18. Angra 1 high burnup fuel behaviour under reactivity initiated accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Daniel de Souza; Silva, Antonio Teixeira e, E-mail: dsgomes@ipen.b, E-mail: teixeira@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    The 16x16 NGF (Next Generation Fuel) fuel assembly, comprising of highly corrosive-resistant ZIRLO clad fuel rods, been replacing the current 16x16 Standard (16STD) fuel assembly in the Angra 1, a pressurized water reactor, with a net output of 626 MWe. The 16x16 NGF fuel assemblies are designed for a peak rod average burnup of up to 75 GWd/MTU, thus improving fuel utilization and reducing spent fuel storage issues. A design basis accident, the Reactivity Initiated Accident (RIA), became a concern for a further increase in burnup as the simulated RIA tests revealed a lower enthalpy threshold for fuel failure. Two fuel performance codes, FRAPCON and FRAPTRAN, were used to predict high burnup behavior of Angra 1, during an RIA. The maximum average linear fuel rating used was 17.62 KW/m. The FRAPCON 3.4 code was applied to simulate the steady-state performance of the 16 NGF fuel rods up to a burnup of 55 GWd/MTU. With FRAPTRAN-1.4 the fuel behavior was simulated for an RIA power pulse of 4.5 ms (FHWH), and enthalpy peak of 130 Cal/g. With FRAPCON-3.4, the corrosion and hydrogen pickup characteristics of the advanced ZIRLO clad fuel rods were added to the code by modifying the actual corrosion model for Zircaloy-4 through the multiplication of empirical factors, which were appropriate to each alloy, and by means of reducing the current hydrogen pickup fraction. (author)

  19. Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation - Vandellos II Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Gauld, Ian C [ORNL

    2011-01-01

    This report is one of the several recent NUREG/CR reports documenting benchmark-quality radiochemical assay data and the use of the data to validate computer code predictions of isotopic composition for spent nuclear fuel, to establish the uncertainty and bias associated with code predictions. The experimental data analyzed in the current report were acquired from a high-burnup fuel program coordinated by Spanish organizations. The measurements included extensive actinide and fission product data of importance to spent fuel safety applications, including burnup credit, decay heat, and radiation source terms. Six unique spent fuel samples from three uranium oxide fuel rods were analyzed. The fuel rods had a 4.5 wt % {sup 235}U initial enrichment and were irradiated in the Vandellos II pressurized water reactor operated in Spain. The burnups of the fuel samples range from 42 to 78 GWd/MTU. The measurements were used to validate the two-dimensional depletion sequence TRITON in the SCALE computer code system.

  20. Irradiation performance of PFBR MOX fuel after 112 GWd/t burn-up

    Science.gov (United States)

    Venkiteswaran, C. N.; Jayaraj, V. V.; Ojha, B. K.; Anandaraj, V.; Padalakshmi, M.; Vinodkumar, S.; Karthik, V.; Vijaykumar, Ran; Vijayaraghavan, A.; Divakar, R.; Johny, T.; Joseph, Jojo; Thirunavakkarasu, S.; Saravanan, T.; Philip, John; Rao, B. P. C.; Kasiviswanathan, K. V.; Jayakumar, T.

    2014-06-01

    The 500 MWe Prototype Fast Breeder Reactor (PFBR) which is in advanced stage of construction at Kalpakkam, India, will use mixed oxide (MOX) fuel with a target burnup of 100 GWd/t. The fuel pellet is of annular design to enable operation at a peak linear power of 450 W/cm with the requirement of minimum duration of pre-conditioning. The performance of the MOX fuel and the D9 clad and wrapper material was assessed through Post Irradiation Examinations (PIE) after test irradiation of 37 fuel pin subassembly in Fast Breeder Test Reactor (FBTR) to a burn-up of 112 GWd/t. Fission product distribution, swelling and fuel-clad gap evolution, central hole diameter variation, restructuring, fission gas release and clad wastage due to fuel-clad chemical interaction were evaluated through non-destructive and destructive examinations. The examinations have indicated that the MOX fuel can safely attain the desired target burn-up in PFBR.

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

    Directory of Open Access Journals (Sweden)

    Lecarpentier D.

    2013-03-01

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

  2. Burnup simulations of different fuel grades using the MCNPX Monte Carlo code

    Directory of Open Access Journals (Sweden)

    Asah-Opoku Fiifi

    2014-01-01

    Full Text Available Global energy problems range from the increasing cost of fuel to the unequal distribution of energy resources and the potential climate change resulting from the burning of fossil fuels. A sustainable nuclear energy would augment the current world energy supply and serve as a reliable future energy source. This research focuses on Monte Carlo simulations of pressurized water reactor systems. Three different fuel grades - mixed oxide fuel (MOX, uranium oxide fuel (UOX, and commercially enriched uranium or uranium metal (CEU - are used in this simulation and their impact on the effective multiplication factor (Keff and, hence, criticality and total radioactivity of the reactor core after fuel burnup analyzed. The effect of different clad materials on Keff is also studied. Burnup calculation results indicate a buildup of plutonium isotopes in UOX and CEU, as opposed to a decline in plutonium radioisotopes for MOX fuel burnup time. For MOX fuel, a decrease of 31.9% of the fissile plutonium isotope is observed, while for UOX and CEU, fissile plutonium isotopes increased by 82.3% and 83.8%, respectively. Keff results show zircaloy as a much more effective clad material in comparison to zirconium and stainless steel.

  3. Underestimation of nuclear fuel burnup – theory, demonstration and solution in numerical models

    Directory of Open Access Journals (Sweden)

    Gajda Paweł

    2016-01-01

    Full Text Available Monte Carlo methodology provides reference statistical solution of neutron transport criticality problems of nuclear systems. Estimated reaction rates can be applied as an input to Bateman equations that govern isotopic evolution of reactor materials. Because statistical solution of Boltzmann equation is computationally expensive, it is in practice applied to time steps of limited length. In this paper we show that simple staircase step model leads to underprediction of numerical fuel burnup (Fissions per Initial Metal Atom – FIMA. Theoretical considerations indicates that this error is inversely proportional to the length of the time step and origins from the variation of heating per source neutron. The bias can be diminished by application of predictor-corrector step model. A set of burnup simulations with various step length and coupling schemes has been performed. SERPENT code version 1.17 has been applied to the model of a typical fuel assembly from Pressurized Water Reactor. In reference case FIMA reaches 6.24% that is equivalent to about 60 GWD/tHM of industrial burnup. The discrepancies up to 1% have been observed depending on time step model and theoretical predictions are consistent with numerical results. Conclusions presented in this paper are important for research and development concerning nuclear fuel cycle also in the context of Gen4 systems.

  4. Burn-up characteristics of ADS system utilizing the fuel composition from MOX PWRs spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Marsodi E-mail: marsodi@batan.go.id; Lasman, K.A.S.; Nishihara, K. E-mail: nishi@omega.tokai.jaeri.go.jp; Osugi, T.; Tsujimoto, K.; Marsongkohadi; Su' ud, Z. E-mail: szaki@fi.itb.ac.id

    2002-12-01

    Burn-up characteristics of accelerator-driven system, ADS has been evaluated utilizing the fuel composition from MOX PWRs spent fuel. The system consists of a high intensity proton beam accelerator, spallation target, and sub-critical reactor core. The liquid lead-bismuth, Pb-Bi, as spallation target, was put in the center of the core region. The general approach was conducted throughout the nitride fuel that allows the utilities to choose the strategy for destroying or minimizing the most dangerous high level wastes in a fast neutron spectrum. The fuel introduced surrounding the target region was the same with the composition of MOX from 33 GWd/t PWRs spent-fuel with 5 year cooling and has been compared with the fuel composition from 45 and 60 GWd/t PWRs spent-fuel with the same cooling time. The basic characteristics of the system such as burn-up reactivity swing, power density, neutron fluxes distribution, and nuclides densities were obtained from the results of the neutronics and burn-up analyses using ATRAS computer code of the Japan Atomic Energy research Institute, JAERI.

  5. Thermal Hydraulic Analysis of 3 MW TRIGA Research Reactor of Bangladesh Considering Different Cycles of Burnup

    Directory of Open Access Journals (Sweden)

    M.H. Altaf

    2014-12-01

    Full Text Available Burnup dependent steady state thermal hydraulic analysis of TRIGA Mark-II research reactor has been carried out utilizing coupled point kinetics, neutronics and thermal hydraulics code EUREKA-2/RR. From the previous calculations of neutronics parameters including percentage burnup of individual fuel elements performed so far for 700 MWD burnt core of TRIGA reactor showed that the fuel rod predicted as hottest at the beginning of cycle (fresh core was found to remain as the hottest until 200 MWD of burn, but, with the progress of core burn, the hottest rod was found to be shifted and another rod in the core became the hottest. The present study intends to evaluate the thermal hydraulic parameters of these hottest fuel rods at different cycles of burnup, from beginning to 700 MWD core burnt considering reactor operates under steady state condition. Peak fuel centerline temperature, maximum cladding and coolant temperatures of the hottest channels were calculated. It revealed that maximum temperature reported for fuel clad and fuel centerline found to lie below their melting points which indicate that there is no chance of burnout on the fuel cladding surface and no blister in the fuel meat throughout the considered cycles of core burnt.

  6. Oxygen potential in the rim region of high burnup UO 2 fuel

    Science.gov (United States)

    Matzke, Hj.

    1994-01-01

    Small specimens from the rim region (fuel surface) of a UO 2 fuel rod with an average burnup of 7.6% FIMA were analysed in a miniaturized galvanic cell to determine their oxygen potential Δ Ḡ(O 2) . These fuel pieces represented the porous rim structure with very small grains known to be formed near the periphery of high burnup UO 2 fuel pellets. The oxygen potential of the rim material was very low, corresponding to that of unirradiated stoichiometric UO 2, or to that of slightly substoichiometric UO 2 containing rare earth fission products. No indication of oxidation due to fission was found, though most fission was that of Pu. Measurements on pieces from the inner, unrestructured fuel showed somewhat higher oxygen potentials corresponding to those of very slightly substoichiometric fuel if allowance is made for the incorporation of rare earths. These results are in contrast to some generally accepted ideas of burnup effects, and the possible reasons and implications are discussed.

  7. An empirical formulation to describe the evolution of the high burnup structure

    Science.gov (United States)

    Lemes, Martín; Soba, Alejandro; Denis, Alicia

    2015-01-01

    In the present work the behavior of fuel pellets for LWR power reactors in the high burnup range (average burnup higher than about 45 MWd/kgU) is analyzed. For extended irradiation periods, a considerable Pu concentration is reached in the pellet periphery (rim zone), that contributes to local burnup. Gradually, a new microstructure develops in that ring, characterized by small grains and large pores as compared with those of the original material. In this region Xe is absent from the solid lattice (although it continues to be dissolved in the rest of the pellet). The porous microstructure in the pellet edge causes local changes in the mechanical and thermal properties, thus affecting the overall fuel behavior. It is generally accepted that the evolution of porosity in the high burnup structure (HBS) is determinant of the retention capacity of the fission gases rejected from the fuel matrix. This is the reason why, during the latest years a considerable effort has been devoted to characterizing the parameters that influence porosity. Although the mechanisms governing the microstructural transformation have not been completely elucidated yet, some empirical expressions can be given, and this is the intention of the present work, for representing the main physical parameters. Starting from several works published in the open literature, some mathematical expressions were developed to describe the behavior and progress of porosity at local burnup values ranging from 60 to 300 MWd/kgU. The analysis includes the interactions of different orders between pores, the growth of the pore radius by capturing vacancies, the evolution of porosity, pore number density and overpressure within the closed pores, the inventory of fission gas dissolved in the matrix and retained in the pores. The model is mathematically expressed by a system of non-linear differential equations. In the present work, results of this calculation scheme are compared with experimental data available in

  8. Review of Technical Issues Related to Predicting Isotopic Compositions and Source Terms for High-Burnup LWR Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I. C.; Parks, C. V.

    2000-12-11

    This report has been prepared to review the technical issues important to the prediction of isotopic compositions and source terms for high-burnup, light-water-reactor (LWR) fuel as utilized in the licensing of spent fuel transport and storage systems. The current trend towards higher initial 235U enrichments, more complex assembly designs, and more efficient fuel management schemes has resulted in higher spent fuel burnups than seen in the past. This trend has led to a situation where high-burnup assemblies from operating LWRs now extend beyond the area where available experimental data can be used to validate the computational methods employed to calculate spent fuel inventories and source terms. This report provides a brief review of currently available validation data, including isotopic assays, decay heat measurements, and shielded dose-rate measurements. Potential new sources of experimental data available in the near term are identified. A review of the background issues important to isotopic predictions and some of the perceived technical challenges that high-burnup fuel presents to the current computational methods are discussed. Based on the review, the phenomena that need to be investigated further and the technical issues that require resolution are presented. The methods and data development that may be required to address the possible shortcomings of physics and depletion methods in the high-burnup and high-enrichment regime are also discussed. Finally, a sensitivity analysis methodology is presented. This methodology is currently being investigated at the Oak Ridge National Laboratory as a computational tool to better understand the changing relative significance of the underlying nuclear data in the different enrichment and burnup regimes and to identify the processes that are dominant in the high-burnup regime. The potential application of the sensitivity analysis methodology to help establish a range of applicability for experimental

  9. A criticality analysis of the GBC-32 dry storage cask with Hanbit nuclear power plant unit 3 fuel assemblies from the viewpoint of burnup credit

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Hyung Ju; Kim, Do Yeon; Park, Kwang Heon; Hong, Ser Gi [Dept. of Nuclear Engineering, Kyung Hee University, Seoul (Korea, Republic of)

    2016-06-15

    Nuclear criticality safety analyses (NCSAs) considering burnup credit were performed for the GBC-32 cask. The used nuclear fuel assemblies (UNFAs) discharged from Hanbit Nuclear Power Plant Unit 3 Cycle 6 were loaded into the cask. Their axial burnup distributions and average discharge burnups were evaluated using the DeCART and Multi-purpose Analyzer for Static and Transient Effects of Reactors (MASTER) codes, and NCSAs were performed using SCALE 6.1/STandardized Analysis of Reactivity for Burnup Credit using SCALE (STARBUCS) and Monte Carlo N-Particle transport code, version 6 (MCNP 6). The axial burnup distributions were determined for 20 UNFAs with various initial enrichments and burnups, which were applied to the criticality analysis for the cask system. The UNFAs for 20- and 30-year cooling times were assumed to be stored in the cask. The criticality analyses indicated that keff values for UNFAs with nonuniform axial burnup distributions were larger than those with a uniform distribution, that is, the end effects were positive but much smaller than those with the reference distribution. The axial burnup distributions for 20 UNFAs had shapes that were more symmetrical with a less steep gradient in the upper region than the reference ones of the United States Department of Energy. These differences in the axial burnup distributions resulted in a significant reduction in end effects compared with the reference.

  10. Burnup concept for a long-life fast reactor core using MCNPX.

    Energy Technology Data Exchange (ETDEWEB)

    Holschuh, Thomas Vernon,; Lewis, Tom Goslee,; Parma, Edward J.,

    2013-02-01

    This report describes a reactor design with a burnup concept for a long-life fast reactor core that was evaluated using Monte Carlo N-Particle eXtended (MCNPX). The current trend in advanced reactor design is the concept of a small modular reactor (SMR). However, very few of the SMR designs attempt to substantially increase the lifetime of a reactor core, especially without zone loading, fuel reshuffling, or other artificial mechanisms in the core that %E2%80%9Cflatten%E2%80%9D the power profile, including non-uniform cooling, non-uniform moderation, or strategic poison placement. Historically, the limitations of computing capabilities have prevented acceptable margins in the temporal component of the spatial excess reactivity in a reactor design, due primarily to the error in burnup calculations. This research was performed as an initial scoping analysis into the concept of a long-life fast reactor. It can be shown that a long-life fast reactor concept can be modeled using MCNPX to predict burnup and neutronics behavior. The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional Light Water Reactors (LWRs) or other SMR designs. For the purpose of this study, a single core design was investigated: a relatively small reactor core, yielding a medium amount of power (~200 to 400 MWth). The results of this scoping analysis were successful in providing a preliminary reactor design involving metal U-235/U-238 fuel with HT-9 fuel cladding and sodium coolant at a 20% volume fraction.

  11. Spent fuel dissolution rates as a function of burnup and water chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Gray, W.J.

    1998-06-01

    To help provide a source term for performance-assessment calculations, dissolution studies on light-water-reactor (LWR) spent fuel have been conducted over the past few years at Pacific Northwest National Laboratory in support of the Yucca Mountain Site Characterization Project. This report describes that work for fiscal years 1996 through mid-1998 and includes summaries of some results from previous years for completeness. The following conclusions were based on the results of various flowthrough dissolution rate tests and on tests designed to measure the inventories of {sup 129}I located within the fuel/cladding gap region of different spent fuels: (1) Spent fuels with burnups in the range 30 to 50 MWd/kgM all dissolved at about the same rate over the conditions tested. To help determine whether the lack of burnup dependence extends to higher and lower values, tests are in progress or planned for spent fuels with burnups of 13 and {approximately} 65 MWd/kgM. (2) Oxidation of spent fuel up to the U{sub 4}O{sub 9+x} stage does not have a large effect on intrinsic dissolution rates. However, this degree of oxidation could increase the dissolution rates of relatively intact fuel by opening the grain boundaries, thereby increasing the effective surface area that is available for contact by water. From a disposal viewpoint, this is a potentially more important consideration than the effect on intrinsic rates. (3) The gap inventories of {sup 129}I were found to be smaller than the fission gas release (FGR) for the same fuel rod with the exception of the rod with the highest FGR. Several additional fuels would have to be tested to determine whether a generalized relationship exists between FGR and {sup 129}I gap inventory for US LWR fuels.

  12. Criticality Analysis of Assembly Misload in a PWR Burnup Credit Cask

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J. C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2008-01-31

    The Interim Staff Guidance on bumup credit (ISG-8) for spent fuel in storage and transportation casks, issued by the Nuclear Regulatory Commission's Spent Fuel Project Office, recommends a bumup measurement for each assembly to confirm the reactor record and compliance with the assembly bumup value used for loading acceptance. This recommendation is intended to prevent unauthorized loading (misloading) of assemblies due to inaccuracies in reactor burnup records and/or improper assembly identification, thereby ensuring that the appropriate subcritical margin is maintained. This report presents a computational criticality safety analysis of the consequences of misloading fuel assemblies in a highcapacity cask that relies on burnup credit for criticality safety. The purpose of this report is to provide a quantitative understanding of the effects of fuel misloading events on safety margins. A wide variety of fuel-misloading configurations are investigated and results are provided for informational purposes. This report does not address the likelihood of occurrence for any of the misload configurations considered. For representative, qualified bumup-enrichment combinations, with and without fission products included, misloading two assemblies that are underburned by 75% results in an increase in keff of 0.025-0.045, while misloading four assemblies that are underburned by 50% also results in an increase in keff of 0.025-0.045. For the cask and conditions considered, a reduction in bumup of 20% in all assemblies results in an increase in kff of less than 0.035. Misloading a single fresh assembly with 3, 4, or 5 wt% 235U enrichment results in an increase in keffof--0.02, 0.04, or 0.06, respectively. The report concludes with a summary of these and other important findings, as well as a discussion of relevant issues that should be considered when assessing the appropriate role of burnup measurements.

  13. Thermodynamic analysis for high burn-up fuel internal chemistry. 2

    Energy Technology Data Exchange (ETDEWEB)

    Fuji, Kensho; Kyoh, Bunkei [Kinki Univ., Higashi-Osaka, Osaka (Japan)

    1998-09-01

    Thermodynamic calculations with the computer program SOLGASMIX-PV have been performed for the chemical states expected in irradiated fast breeder reactor (FBR) fuels containing transuranium (TRU) elements. The analysis shows that A (alkali and alkaline-earth)-molybdates exist, but neither A-uranates nor A-zirconates are formed in FBR fuel pellets irradiated to high burn-up. And increase of oxygen potential in irradiated FBR fuel is ascribed to growing amount of rare earth, noble metal and TRU elements. (author)

  14. Effects of Lower Drying-Storage Temperature on the Ductility of High-Burnup PWR Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Burtseva, T. A. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-08-30

    The purpose of this research effort is to determine the effects of canister and/or cask drying and storage on radial hydride precipitation in, and potential embrittlement of, high-burnup (HBU) pressurized water reactor (PWR) cladding alloys during cooling for a range of peak drying-storage temperatures (PCT) and hoop stresses. Extensive precipitation of radial hydrides could lower the failure hoop stresses and strains, relative to limits established for as-irradiated cladding from discharged fuel rods stored in pools, at temperatures below the ductile-to-brittle transition temperature (DBTT).

  15. A semi-empirical model for the formation and depletion of the high burnup structure in UO2

    Science.gov (United States)

    Pizzocri, D.; Cappia, F.; Luzzi, L.; Pastore, G.; Rondinella, V. V.; Van Uffelen, P.

    2017-04-01

    In the rim zone of UO2 nuclear fuel pellets, the combination of high burnup and low temperature drives a microstructural change, leading to the formation of the high burnup structure (HBS). In this work, we propose a semi-empirical model to describe the formation of the HBS, which embraces the polygonisation/recrystallization process and the depletion of intra-granular fission gas, describing them as inherently related. For this purpose, we performed grain-size measurements on samples at radial positions in which the restructuring was incomplete. Based on these new experimental data, we infer an exponential reduction of the average grain size with local effective burnup, paired with a simultaneous depletion of intra-granular fission gas driven by diffusion. The comparison with currently used models indicates the applicability of the herein developed model within integral fuel performance codes.

  16. Assesment of advanced step models for steady state Monte Carlo burnup calculations in application to prismatic HTGR

    Directory of Open Access Journals (Sweden)

    Kępisty Grzegorz

    2015-09-01

    Full Text Available In this paper, we compare the methodology of different time-step models in the context of Monte Carlo burnup calculations for nuclear reactors. We discuss the differences between staircase step model, slope model, bridge scheme and stochastic implicit Euler method proposed in literature. We focus on the spatial stability of depletion procedure and put additional emphasis on the problem of normalization of neutron source strength. Considered methodology has been implemented in our continuous energy Monte Carlo burnup code (MCB5. The burnup simulations have been performed using the simplified high temperature gas-cooled reactor (HTGR system with and without modeling of control rod withdrawal. Useful conclusions have been formulated on the basis of results.

  17. A multi-platform linking code for fuel burnup and radiotoxicity analysis

    Science.gov (United States)

    Cunha, R.; Pereira, C.; Veloso, M. A. F.; Cardoso, F.; Costa, A. L.

    2014-02-01

    A linking code between ORIGEN2.1 and MCNP has been developed at the Departamento de Engenharia Nuclear/UFMG to calculate coupled neutronic/isotopic results for nuclear systems and to produce a large number of criticality, burnup and radiotoxicity results. In its previous version, it evaluated the isotopic composition evolution in a Heat Pipe Power System model as well as the radiotoxicity and radioactivity during lifetime cycles. In the new version, the code presents features such as multi-platform execution and automatic results analysis. Improvements made in the code allow it to perform simulations in a simpler and faster way without compromising accuracy. Initially, the code generates a new input for MCNP based on the decisions of the user. After that, MCNP is run and data, such as recoverable energy per prompt fission neutron, reaction rates and keff, are automatically extracted from the output and used to calculate neutron flux and cross sections. These data are then used to construct new ORIGEN inputs, one for each cell in the core. Each new input is run on ORIGEN and generates outputs that represent the complete isotopic composition of the core on that time step. The results show good agreement between GB (Coupled Neutronic/Isotopic code) and Monteburns (Automated, Multi-Step Monte Carlo Burnup Code System), developed by the Los Alamos National Laboratory.

  18. Tritium Burn-up Depth and Tritium Break-Even Time

    Institute of Scientific and Technical Information of China (English)

    LI Cheng-Yue; DENG Bai-Quan; HUANG Jin-Hua; YAN Jian-Cheng

    2006-01-01

    @@ Similarly to but quite different from the xenon poisoning effects resulting from fission-produced iodine during the restart-up process of a fission reactor, we introduce a completely new concept of the tritium burn-up depth and tritium break-even time in the fusion energy research area. To show what the least required amount of tritium storage is used to start up a fusion reactor and how long a time the fusion reactor needs to be operated for achieving the tritium break-even during the initial start-up phase due to the finite tritium breeding time that is dependent on the tritium breeder, specific structure of breeding zone, layout of coolant flow pipe, tritium recovery scheme, extraction process, the tritium retention of reactor components, unrecoverable tritium fraction in breeder, leakage to the inertial gas container, and the natural decay etc., we describe this new phenomenon and answer this problem by setting up and by solving a set of equations, which express a dynamic subsystem model of the tritium inventory evolution in a fusion experimental breeder (FEB). It is found that the tritium burn-up depth is 317g and the tritium break-even time is approximately 240 full power days for FEB designed detail configuration and it is also found that after one-year operation, the tritium storage reaches 1.18kg that is more than theleast required amount of tritium storage to start up three of FEB-like fusion reactors.

  19. Fuel burnup analysis for IRIS reactor using MCNPX and WIMS-D5 codes

    Science.gov (United States)

    Amin, E. A.; Bashter, I. I.; Hassan, Nabil M.; Mustafa, S. S.

    2017-02-01

    International Reactor Innovative and Secure (IRIS) reactor is a compact power reactor designed with especial features. It contains Integral Fuel Burnable Absorber (IFBA). The core is heterogeneous both axially and radially. This work provides the full core burn up analysis for IRIS reactor using MCNPX and WIMDS-D5 codes. Criticality calculations, radial and axial power distributions and nuclear peaking factor at the different stages of burnup were studied. Effective multiplication factor values for the core were estimated by coupling MCNPX code with WIMS-D5 code and compared with SAS2H/KENO-V code values at different stages of burnup. The two calculation codes show good agreement and correlation. The values of radial and axial powers for the full core were also compared with published results given by SAS2H/KENO-V code (at the beginning and end of reactor operation). The behavior of both radial and axial power distribution is quiet similar to the other data published by SAS2H/KENO-V code. The peaking factor values estimated in the present work are close to its values calculated by SAS2H/KENO-V code.

  20. Comparison of neutron cross sections for selected fission products and isotopic composition analyses with burnup

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. H.; Gil, C. S.; Kim, J. D.; Jang, J. H.; Lee, Y. D. [KAERI, Taejon (Korea)

    2003-10-01

    The neutron absorption cross sections for 18 fission products evaluated within the framework of the KAERI-BNL international collaboration have been compared with the ENDF/B-VI release 7. Also, the influence of the new evaluations on isotopic compositions of the fission products as a function of burnup has been analyzed through the OECD/NEA burnup credit criticality benchmarks (Phase 1B) and the LWR/Pu recycling benchmarks. These calculations were performed by WIMSD-5B with the 69 group libraries prepared from three evaluated nuclear data libraries: ENDF/B-VI.7, ENDF/B-VI.8 including new evaluations in resonance region covering thermal region, and ENDF/B-VII expected including those in upper resonance region up to 20 MeV. For Xe-131, the composition calculated with ENDF/B-VI.8 shows maximum difference of 4.78% compared to ENDF/B-VI.7. However, the isotopic compositions of all fission products calculated with ENDF/B-VII shows no differences compared to ENDF/B-VI.7.

  1. EBSD and TEM characterization of high burn-up mixed oxide fuel

    Science.gov (United States)

    Teague, Melissa; Gorman, Brian; Miller, Brandon; King, Jeffrey

    2014-01-01

    Understanding and studying the irradiation behavior of high burn-up oxide fuel is critical to licensing of future fast breeder reactors. Advancements in experimental techniques and equipment are allowing for new insights into previously irradiated samples. In this work dual column focused ion beam (FIB)/scanning electron microscope (SEM) was utilized to prepared transmission electron microscope samples from mixed oxide fuel with a burn-up of 6.7% FIMA. Utilizing the FIB/SEM for preparation resulted in samples with a dose rate of <0.5 mRem/h compared to ∼1.1 R/h for a traditionally prepared TEM sample. The TEM analysis showed that the sample taken from the cooler rim region of the fuel pellet had ∼2.5× higher dislocation density than that of the sample taken from the mid-radius due to the lower irradiation temperature of the rim. The dual column FIB/SEM was additionally used to prepared and serially slice ∼25 μm cubes. High quality electron back scatter diffraction (EBSD) were collected from the face at each step, showing, for the first time, the ability to obtain EBSD data from high activity irradiated fuel.

  2. Evaluation of the HTTR criticality and burnup calculations with continuous-energy and multigroup cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Chiang, Min-Han; Wang, Jui-Yu [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Sheu, Rong-Jiun, E-mail: rjsheu@mx.nthu.edu.tw [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Liu, Yen-Wan Hsueh [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China)

    2014-05-01

    The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects.

  3. Instant release fraction and matrix release of high burn-up UO2 spent nuclear fuel: Effect of high burn-up structure and leaching solution composition

    Science.gov (United States)

    Serrano-Purroy, D.; Clarens, F.; González-Robles, E.; Glatz, J. P.; Wegen, D. H.; de Pablo, J.; Casas, I.; Giménez, J.; Martínez-Esparza, A.

    2012-08-01

    Two weak points in Performance Assessment (PA) exercises regarding the alteration of Spent Nuclear Fuel (SNF) are the contribution of the so-called Instant Release Fraction (IRF) and the effect of High Burn-Up Structure (HBS). This manuscript focuses on the effect of HBS in matrix (long term) and instant release of a Pressurised Water Reactor (PWR) SNF irradiated in a commercial reactor with a mean Burn-Up (BU) of 60 GWd/tU. In order to study the HBS contribution, two samples from different radial positions have been prepared. One from the centre of the SNF, labelled CORE, and one from the periphery, enriched with HBS and labelled OUT. Static leaching experiments have been carried out with two synthetic leaching solutions: bicarbonate (BIC) and Bentonitic Granitic Groundwater (BGW), and in all cases under oxidising conditions. IRF values have been calculated from the determined Fraction of Inventory in Aqueous Phase (FIAP). In all studied cases, some radionuclides (RN): Rb, Sr and Cs, have shown higher release rates than uranium, especially at the beginning of the experiment, and have been considered as IRF. Redox sensitive RN like Mo and Tc have been found to dissolve slightly faster than uranium and further studies might be needed to confirm if they can also be considered part of the IRF. Most of the remaining studied RN, mainly actinides and lanthanides, have been found to dissolve congruently with the uranium matrix. Finally, Zr, Ru and Rh presented lower release rates than the matrix. Higher matrix release has been determined for CORE than for OUT samples showing that the formation of HBS might have a protective effect against the oxidative corrosion of the SNF. On the contrary, no significant differences have been observed between the two studied leaching solutions (BIC and BGW). Two different IRF contributions have been determined. One corresponding to the fraction of inventory segregated in the external open grain boundaries, directly available to water and

  4. Conceptual design study on an upgraded future Monju core (2). Core concept with extended refueling interval and increased fuel burnup

    Energy Technology Data Exchange (ETDEWEB)

    Kinjo, Hidehito; Ishibashi, Jun-ichi; Nishi, Hiroshi [Japan Nuclear Cycle Development Inst., Tsuruga Head Office, International Cooperation and Technology Development Center, Tsuruga, Fukui (Japan); Kageyama, Takeshi [Nuclear Energy System Inc., Tokyo (Japan)

    2003-03-01

    A conceptual design study has been performed at the International Cooperation and Technology Development Center to investigate the feasibility of upgraded future Monju cores with extended refueling intervals of 365efpd/cycle and increased fuel burnup of 150 GWd/t. The goal of this study is to demonstrate the possible contribution of Monju to the improved economy and to efficient utilization, as one of the major facilities for fast neutron irradiation. Two design measures have been mainly taken to improve the core fuel burnup and reactivity control characteristics for the extended operating cycle length of 1 year: (1) The driver fuel pin specification with both increased pin diameter of 7.7mm and increased active core height of about 100cm has been chosen to reduce the burnup reactivity swing, (2) The absorber control rod specification has also been changed to enhance the control rod reactivity worth by increasing {sup 10}B-enrichment and absorber length, and to adequately secure the shutdown reactivity margin. The major core characteristics have been evaluated on the core power distribution, safety parameters such as sodium void reactivity and Doppler effect, thermal hydraulics and reactivity control characteristics. The results show that this core could achieve the targeted core performances of 1-year operating cycle as well as 150GWd/t discharged burnup, without causing any significant drawback on the core characteristics and safety aspects. The upgraded core concepts have, therefore, been confirmed as feasible. (author)

  5. 78 FR 67348 - Invitation for Public Comment on Draft Test Plan for the High Burnup Dry Storage Cask Research...

    Science.gov (United States)

    2013-11-12

    ...: U.S. Department of Energy, C/O Melissa Bates, 1955 Freemont Ave., MS 1235, Idaho Falls, ID 83415..., 1955 Fremont Ave., Attn: Melissa Bates, Idaho Falls, ID, between 8 a.m. and 3:30 p.m. MT, Monday.... Melissa Bates, Contracting Officers Representative, High Burnup Dry Storage Cask Research and...

  6. Kinetic parameters study based on burn-up for improving the performance of research reactor equilibrium core

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2014-01-01

    Full Text Available In this study kinetic parameters, effective delayed neutron fraction and prompt neutron generation time have been investigated at different burn-up stages for research reactor's equilibrium core utilizing low enriched uranium high density fuel (U3Si2-Al fuel with 4.8 g/cm3 of uranium. Results have been compared with reference operating core of Pakistan research Reactor-1. It was observed that by increasing fuel burn-up, effective delayed neutron fraction is decreased while prompt neutron generation time is increased. However, over all ratio beff/L is decreased with increasing burn-up. Prompt neutron generation time L in the understudy core is lower than reference operating core of reactor at all burn-up steps due to hard spectrum. It is observed that beff is larger in the understudy core than reference operating core of due to smaller size. Calculations were performed with the help of computer codes WIMSD/4 and CITATION.

  7. Extended calculations of OECD/NEA phase II-C burnup credit criticality benchmark problem for PWR spent fuel transport cask by using MCNP-4B2 code and JENDL-3.2 library

    Energy Technology Data Exchange (ETDEWEB)

    Kuroishi, Takeshi; Hoang, Anh Tuan; Nomura, Yasushi; Okuno, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    The reactivity effect of the asymmetry of axial burnup profile in burnup credit criticality safety is studied for a realistic PWR spent fuel transport cask proposed in the current OECD/NEA Phase II-C benchmark problem. The axial burnup profiles are simulated in 21 material zones based on in-core flux measurements varying from strong asymmetry to more or less no asymmetry. Criticality calculations in a 3-D model have been performed using the continuous energy Monte Carlo code MCNP-4B2 and the nuclear data library JENDL-3.2. Calculation conditions are determined with consideration of the axial fission source convergence. Calculations are carried out not only for cases proposed in the benchmark but also for additional cases assuming symmetric burnup profile. The actinide-only approach supposed for first domestic introduction of burnup credit into criticality evaluation is also considered in addition to the actinide plus fission product approach adopted in the benchmark. The calculated results show that k{sub eff} and the end effect increase almost linearly with increasing burnup axial offset that is defined as one of typical parameters showing the intensity of axial burnup asymmetry. The end effect is more sensitive to the asymmetry of burnup profile for the higher burnup. For an axially distributed burnup, the axial fission source distribution becomes strongly asymmetric as its peak shifts toward the top end of the fuel's active zone where the local burnup is less than that of the bottom end. The peak of fission source distribution becomes higher with the increase of either the asymmetry of burnup profile or the assembly-averaged burnup. The conservatism of the assumption of uniform axial burnup based on the actinide-only approach is estimated quantitatively in comparison with the k{sub eff} result calculated with experiment-based strongest asymmetric axial burnup profile with the actinide plus fission product approach. (author)

  8. Applicability of the MCNP-ACAB system to inventory prediction in high-burnup fuels: sensitivity/uncertainty estimates

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Herranz, N.; Cabellos, O. [Madrid Polytechnic Univ., Dept. of Nuclear Engineering (Spain); Cabellos, O.; Sanz, J. [Madrid Polytechnic Univ., 2 Instituto de Fusion Nuclear (Spain); Sanz, J. [Univ. Nacional Educacion a Distancia, Dept. of Power Engineering, Madrid (Spain)

    2005-07-01

    We present a new code system which combines the Monte Carlo neutron transport code MCNP-4C and the inventory code ACAB as a suitable tool for high burnup calculations. Our main goal is to show that the system, by means of ACAB capabilities, enables us to assess the impact of neutron cross section uncertainties on the inventory and other inventory-related responses in high burnup applications. The potential impact of nuclear data uncertainties on some response parameters may be large, but only very few codes exist which can treat this effect. In fact, some of the most reported effective code systems in dealing with high burnup problems, such as CASMO-4, MCODE and MONTEBURNS, lack this capability. As first step, the potential of our system, ruling out the uncertainty capability, has been compared with that of those code systems, using a well referenced high burnup pin-cell benchmark exercise. It is proved that the inclusion of ACAB in the system allows to obtain results at least as reliable as those obtained using other inventory codes, such as ORIGEN2. Later on, the uncertainty analysis methodology implemented in ACAB, including both the sensitivity-uncertainty method and the uncertainty analysis by the Monte Carlo technique, is applied to this benchmark problem. We estimate the errors due to activation cross section uncertainties in the prediction of the isotopic content up to the high-burnup spent fuel regime. The most relevant uncertainties are remarked, and some of the most contributing cross sections to those uncertainties are identified. For instance, the most critical reaction for Am{sup 242m} is Am{sup 241}(n,{gamma}-m). At 100 MWd/kg, the cross-section uncertainty of this reaction induces an error of 6.63% on the Am{sup 242m} concentration.The uncertainties in the inventory of fission products reach up to 30%.

  9. Comparison of nuclear data uncertainty propagation methodologies for PWR burn-up simulations

    CERN Document Server

    Diez, Carlos Javier; Hoefer, Axel; Porsch, Dieter; Cabellos, Oscar

    2014-01-01

    Several methodologies using different levels of approximations have been developed for propagating nuclear data uncertainties in nuclear burn-up simulations. Most methods fall into the two broad classes of Monte Carlo approaches, which are exact apart from statistical uncertainties but require additional computation time, and first order perturbation theory approaches, which are efficient for not too large numbers of considered response functions but only applicable for sufficiently small nuclear data uncertainties. Some methods neglect isotopic composition uncertainties induced by the depletion steps of the simulations, others neglect neutron flux uncertainties, and the accuracy of a given approximation is often very hard to quantify. In order to get a better sense of the impact of different approximations, this work aims to compare results obtained based on different approximate methodologies with an exact method, namely the NUDUNA Monte Carlo based approach developed by AREVA GmbH. In addition, the impact ...

  10. Development of burnup dependent fuel rod model in COBRA-TF

    Science.gov (United States)

    Yilmaz, Mine Ozdemir

    The purpose of this research was to develop a burnup dependent fuel thermal conductivity model within Pennsylvania State University, Reactor Dynamics and Fuel Management Group (RDFMG) version of the subchannel thermal-hydraulics code COBRA-TF (CTF). The model takes into account first, the degradation of fuel thermal conductivity with high burnup; and second, the fuel thermal conductivity dependence on the Gadolinium content for both UO2 and MOX fuel rods. The modified Nuclear Fuel Industries (NFI) model for UO2 fuel rods and Duriez/Modified NFI Model for MOX fuel rods were incorporated into CTF and fuel centerline predictions were compared against Halden experimental test data and FRAPCON-3.4 predictions to validate the burnup dependent fuel thermal conductivity model in CTF. Experimental test cases from Halden reactor fuel rods for UO2 fuel rods at Beginning of Life (BOL), through lifetime without Gd2O3 and through lifetime with Gd 2O3 and a MOX fuel rod were simulated with CTF. Since test fuel rod and FRAPCON-3.4 results were based on single rod measurements, CTF was run for a single fuel rod surrounded with a single channel configuration. Input decks for CTF were developed for one fuel rod located at the center of a subchannel (rod-centered subchannel approach). Fuel centerline temperatures predicted by CTF were compared against the measurements from Halden experimental test data and the predictions from FRAPCON-3.4. After implementing the new fuel thermal conductivity model in CTF and validating the model with experimental data, CTF model was applied to steady state and transient calculations. 4x4 PWR fuel bundle configuration from Purdue MOX benchmark was used to apply the new model for steady state and transient calculations. First, one of each high burnup UO2 and MOX fuel rods from 4x4 matrix were selected to carry out single fuel rod calculations and fuel centerline temperatures predicted by CTF/TORT-TD were compared against CTF /TORT-TD /FRAPTRAN

  11. Effects of microstructural constraints on the transport of fission products in uranium dioxide at low burnups

    Science.gov (United States)

    Lim, Harn Chyi; Rudman, Karin; Krishnan, Kapil; McDonald, Robert; Dickerson, Patricia; Gong, Bowen; Peralta, Pedro

    2016-08-01

    Diffusion of fission gases in UO2 is studied at low burnups, before bubble growth and coalescence along grain boundaries (GBs) become dominant, using a 3-D finite element model that incorporates actual UO2 microstructures. Grain boundary diffusivities are assigned based on crystallography with lattice and GB diffusion coupled with temperature to account for temperature gradients. Heterogeneity of GB properties and connectivity can induce regions where concentration is locally higher than without GB diffusion. These regions are produced by "bottlenecks" in the GB network because of lack of connectivity among high diffusivity GBs due to crystallographic constraints, and they can lead to localized swelling. Effective diffusivities were calculated assuming a uniform distribution of high diffusivity among GBs. Results indicate an increase over the bulk diffusivity with a clear grain size effect and that connectivity and properties of different GBs become important factors on the variability of fission product concentration at the microscale.

  12. Accident source terms for boiling water reactors with high burnup cores.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Powers, Dana Auburn; Leonard, Mark Thomas

    2007-11-01

    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 product release to containment described by NUREG-1465. This secondary objective is motivated by an interest to understand the extent to which research into the release and behaviors of radionuclides under accident conditions has altered best-estimate calculations of the integral response of BWRs to severe core damage sequences and the resulting radiological source terms to containment. This report, therefore, documents specific results of fission product source term analyses that will form the basis for the HBU supplement to NUREG-1465. However, commentary is also provided on observed differences between the composite results of the source term calculations performed here and those reflected NUREG-1465 itself.

  13. Propagation of Uncertainty in System Parameters of a LWR Model by Sampling MCNPX Calculations - Burnup Analysis

    Science.gov (United States)

    Campolina, Daniel de A. M.; Lima, Claubia P. B.; Veloso, Maria Auxiliadora F.

    2014-06-01

    For all the physical components that comprise a nuclear system there is an uncertainty. Assessing the impact of uncertainties in the simulation of fissionable material systems is essential for a best estimate calculation that has been replacing the conservative model calculations as the computational power increases. The propagation of uncertainty in a simulation using a Monte Carlo code by sampling the input parameters is recent because of the huge computational effort required. In this work a sample space of MCNPX calculations was used to propagate the uncertainty. The sample size was optimized using the Wilks formula for a 95th percentile and a two-sided statistical tolerance interval of 95%. Uncertainties in input parameters of the reactor considered included geometry dimensions and densities. It was showed the capacity of the sampling-based method for burnup when the calculations sample size is optimized and many parameter uncertainties are investigated together, in the same input.

  14. Burn-Up Determination by High Resolution Gamma Spectrometry: Axial and Diametral Scanning Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Forsyth, R.S.; Blackadder, W.H.; Ronqvist, N.

    1967-02-15

    In the gamma spectrometric determination of burn-up the use of a single fission product as a monitor of the specimen fission rate is subject to errors caused by activity saturation or, in certain cases, fission product migration. Results are presented of experiments in which all the resolvable gamma peaks in the fission product spectrum have been used to calculate the fission rate; these results form a pattern which reflect errors in the literature values of the gamma branching ratios, fission yields etc., and also represent a series of empirical correction factors. Axial and diametral scanning experiments on a long-irradiated low-enrichment fuel element are also described and demonstrate that it is possible to differentiate between fissions in U-235 and in Pu-239 respectively by means of the ratios of the Ru-106 activity to the activities of the other fission products.

  15. Data Mining Techniques to Estimate Plutonium, Initial Enrichment, Burnup, and Cooling Time in Spent Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Trellue, Holly Renee [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fugate, Michael Lynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tobin, Stephen Joesph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-03-19

    The Next Generation Safeguards Initiative (NGSI), Office of Nonproliferation and Arms Control (NPAC), National Nuclear Security Administration (NNSA) of the U.S. Department of Energy (DOE) has sponsored a multi-laboratory, university, international partner collaboration to (1) detect replaced or missing pins from spent fuel assemblies (SFA) to confirm item integrity and deter diversion, (2) determine plutonium mass and related plutonium and uranium fissile mass parameters in SFAs, and (3) verify initial enrichment (IE), burnup (BU), and cooling time (CT) of facility declaration for SFAs. A wide variety of nondestructive assay (NDA) techniques were researched to achieve these goals [Veal, 2010 and Humphrey, 2012]. In addition, the project includes two related activities with facility-specific benefits: (1) determination of heat content and (2) determination of reactivity (multiplication). In this research, a subset of 11 integrated NDA techniques was researched using data mining solutions at Los Alamos National Laboratory (LANL) for their ability to achieve the above goals.

  16. Investigation of the fundamental constants stability based on the reactor Oklo burn-up analysis

    CERN Document Server

    Onegin, M S

    2010-01-01

    The burn-up for SC56-1472 sample of the natural Oklo reactor zone 3 was calculated using the modern Monte Carlo codes. We reconstructed the neutron spectrum in the core by means of the isotope ratios: $^{147}$Sm/$^{148}$Sm and $^{176}$Lu/$^{175}$Lu. These ratios unambiguously determine the spectrum index and core temperature. The effective neutron absorption cross section of $^{149}$Sm calculated using this spectrum was compared with experimental one. The disagreement between these two values allows to limit a possible shift of the low laying resonance of $^{149}$Sm even more . Then, these limits were converted to the limits for the change of the fine structure constant $\\alpha$. We found that for the rate of $\\alpha$ change the inequality $|\\delta \\dot{\\alpha}/\\alpha| \\le 5\\cdot 10^{-18}$ is fulfilled, which is of the next higher order than our previous limit.

  17. Investigation of the Fundamental Constants Stability Based on the Reactor Oklo Burn-Up Analysis

    Science.gov (United States)

    Onegin, M. S.; Yudkevich, M. S.; Gomin, E. A.

    2012-12-01

    The burn-up of few samples of the natural Oklo reactor zones 3, 5 was calculated using the modern Monte Carlo code. We reconstructed the neutron spectrum in the core by means of the isotope ratios: 147Sm/148Sm and 176Lu/175Lu. These ratios unambiguously determine the water content and core temperature. The isotope ratio of the 149Sm in the sample calculated using this spectrum was compared with experimental one. The disagreement between these two values allows one to limit a possible shift of the low lying resonance of 149Sm. Then, these limits were converted to the limits for the change of the fine structure constant α. We have found out, that for the rate of α change, the inequality ěrt˙ {α }/α ěrt<= 5× 10-18 is fulfilled, which is one order higher than our previous limit.

  18. Development, implementation, and verification of multicycle depletion perturbation theory for reactor burnup analysis

    Energy Technology Data Exchange (ETDEWEB)

    White, J.R.

    1980-08-01

    A generalized depletion perturbation formulation based on the quasi-static method for solving realistic multicycle reactor depletion problems is developed and implemented within the VENTURE/BURNER modular code system. The present development extends the original formulation derived by M.L. Williams to include nuclide discontinuities such as fuel shuffling and discharge. This theory is first described in detail with particular emphasis given to the similarity of the forward and adjoint quasi-static burnup equations. The specific algorithm and computational methods utilized to solve the adjoint problem within the newly developed DEPTH (Depletion Perturbation Theory) module are then briefly discussed. Finally, the main features and computational accuracy of this new method are illustrated through its application to several representative reactor depletion problems.

  19. Utilizing the burnup capability in MCNPX to perform depletion analysis of an MNSR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Boafo, Emmanuel [Ghana atomic Energy Commission, Accra (Ghana)

    2013-07-01

    The burnup capability in the MCNPX code was utilized to perform fuel depletion analysis of the MNSR LEU core by estimating the amount of fissile material (U-235) consumed as well as the amount of plutonium formed after the reactor core expected life. The decay heat removal rate for the MNSR after reactor shutdown was also investigated due to its significance to reactor safety. The results show that 0.568 % of U-235 was burnt up after 200 days of reactor operation while the amount of plutonium formed was not significant. The study also found that the decay heat decreased exponentially after reactor shutdown confirming that the decay heat will be removed from the system by natural circulation after shut down and hence safety of the reactor is assured.

  20. Impact of Reactor Operating Parameters on Cask Reactivity in BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Betzler, Benjamin R [ORNL; Ade, Brian J [ORNL

    2017-01-01

    This paper discusses the effect of reactor operating parameters used in fuel depletion calculations on spent fuel cask reactivity, with relevance for boiling-water reactor (BWR) burnup credit (BUC) applications. Assessments that used generic BWR fuel assembly and spent fuel cask configurations are presented. The considered operating parameters, which were independently varied in the depletion simulations for the assembly, included fuel temperature, bypass water density, specific power, and operating history. Different operating history scenarios were considered for the assembly depletion to determine the effect of relative power distribution during the irradiation cycles, as well as the downtime between cycles. Depletion, decay, and criticality simulations were performed using computer codes and associated nuclear data within the SCALE code system. Results quantifying the dependence of cask reactivity on the assembly depletion parameters are presented herein.

  1. Evaluation of fission product worth margins in PWR spent nuclear fuel burnup credit calculations.

    Energy Technology Data Exchange (ETDEWEB)

    Blomquist, R.N.; Finck, P.J.; Jammes, C.; Stenberg, C.G.

    1999-02-17

    Current criticality safety calculations for the transportation of irradiated LWR fuel make the very conservative assumption that the fuel is fresh. This results in a very substantial overprediction of the actual k{sub eff} of the transportation casks; in certain cases, this decreases the amount of spent fuel which can be loaded in a cask, and increases the cost of transporting the spent fuel to the repository. Accounting for the change of reactivity due to fuel depletion is usually referred to as ''burnup credit.'' The US DOE is currently funding a program aimed at establishing an actinide only burnup credit methodology (in this case, the calculated reactivity takes into account the buildup or depletion of a limited number of actinides). This work is undergoing NRC review. While this methodology is being validated on a significant experimental basis, it implicitly relies on additional margins: in particular, the absorption of neutrons by certain actinides and by all fission products is not taken into account. This provides an important additional margin and helps guarantee that the methodology is conservative provided these neglected absorption are known with reasonable accuracy. This report establishes the accuracy of fission product absorption rate calculations: (1) the analysis of European fission product worth experiments demonstrates that fission product cross-sections available in the US provide very good predictions of fission product worth; (2) this is confirmed by a direct comparison of European and US cross section evaluations; (3) accuracy of Spent Nuclear Fuel (SNF) fission product content predictions is established in a recent ORNL report where several SNF isotopic assays are analyzed; and (4) these data are then combined to establish in a conservative manner the fraction of the predicted total fission product absorption which can be guaranteed based on available experimental data.

  2. Neutron transport-burnup code MCORGS and its application in fusion fission hybrid blanket conceptual research

    Science.gov (United States)

    Shi, Xue-Ming; Peng, Xian-Jue

    2016-09-01

    Fusion science and technology has made progress in the last decades. However, commercialization of fusion reactors still faces challenges relating to higher fusion energy gain, irradiation-resistant material, and tritium self-sufficiency. Fusion Fission Hybrid Reactors (FFHR) can be introduced to accelerate the early application of fusion energy. Traditionally, FFHRs have been classified as either breeders or transmuters. Both need partition of plutonium from spent fuel, which will pose nuclear proliferation risks. A conceptual design of a Fusion Fission Hybrid Reactor for Energy (FFHR-E), which can make full use of natural uranium with lower nuclear proliferation risk, is presented. The fusion core parameters are similar to those of the International Thermonuclear Experimental Reactor. An alloy of natural uranium and zirconium is adopted in the fission blanket, which is cooled by light water. In order to model blanket burnup problems, a linkage code MCORGS, which couples MCNP4B and ORIGEN-S, is developed and validated through several typical benchmarks. The average blanket energy Multiplication and Tritium Breeding Ratio can be maintained at 10 and 1.15 respectively over tens of years of continuous irradiation. If simple reprocessing without separation of plutonium from uranium is adopted every few years, FFHR-E can achieve better neutronic performance. MCORGS has also been used to analyze the ultra-deep burnup model of Laser Inertial Confinement Fusion Fission Energy (LIFE) from LLNL, and a new blanket design that uses Pb instead of Be as the neutron multiplier is proposed. In addition, MCORGS has been used to simulate the fluid transmuter model of the In-Zinerater from Sandia. A brief comparison of LIFE, In-Zinerater, and FFHR-E will be given.

  3. Triton burnup study using scintillating fiber detector on JT-60U

    Energy Technology Data Exchange (ETDEWEB)

    Harano, Hideki [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1997-09-01

    The DT fusion reactor cannot be realized without knowing how the fusion-produced 3.5 MeV {alpha} particles behave. The {alpha} particles` behavior can be simulated using the 1 MeV triton. To investigate the 1 MeV triton`s behavior, a new type of directional 14 MeV neutron detector, scintillating fiber (Sci-Fi) detector has been developed and installed on JT-60U in the cooperation with LANL as part of a US-Japan collaboration. The most remarkable feature of the Sci-Fi detector is that the plastic scintillating fibers are employed for the neutron sensor head. The Sci-Fi detector measures and extracts the DT neutrons from the fusion radiation field in high time resolution (10 ms) and wide dynamic range (3 decades). Triton burnup analysis code TBURN has been made in order to analyze the time evolution of DT neutron emission rate obtained by the Sci-Fi detector. The TBURN calculations reproduced the measurements fairly well, and the validity of the calculation model that the slowing down of the 1 MeV triton was classical was confirmed. The Sci-Fi detector`s directionality indicated the tendency that the DT neutron emission profile became more and more peaked with the time progress. In this study, in order to examine the effect of the toroidal field ripple on the triton burnup, R{sub p}-scan and n{sub e}-scan experiments have been performed. The R{sub p}-scan experiment indicates that the triton`s transport was increased as the ripple amplitude over the triton became larger. In the n{sub e}-scan experiment, the DT neutron emission showed the characteristic changes after the gas puffing injection. It was theoretically confirmed that the gas puffing was effective for the collisionality scan. (J.P.N.) 127 refs.

  4. High burnup effects on fuel behaviour under accident conditions: the tests CABRI REP-Na

    Science.gov (United States)

    Schmitz, Franz; Papin, Joelle

    A large, performance based, knowledge and experience in the field of nuclear fuel behaviour is available for nominal operation conditions. The database is continuously completed and precursor assembly irradiations are performed for testing of new materials and innovative designs. This procedure produces data and arguments to extend licencing limits in the permanent research for economic competitiveness. A similar effort must be devoted to the establishment of a database for fuel behaviour under off-normal and accident conditions. In particular, special attention must be given to the so-called design-basis-accident (DBA) conditions. Safety criteria are formulated for these situations and must be respected without consideration of the occurrence probability and the risk associated to the accident situation. The introduction of MOX fuel into the cores of light water reactors and the steadily increasing goal burnup of the fuel call for research work, both experimental and analytical, in the field of fuel response to DBA conditions. In 1992, a significant programme step, CABRI REP-Na, has been launched by the French Nuclear Safety and Protection Institute (IPSN) in the field of the reactivity initiated accident (RIA). After performing the nine experiments of the initial test matrix it can be concluded that important new findings have been evidenced. High burnup clad corrosion and the associated degradation of the mechanical properties of the ZIRCALOY4 clad is one of the key phenomena of the fuel behaviour under accident conditions. Equally important is the evidence that transient, dynamic fission gas effects resulting from the close to adiabatic heating introduces a new explosive loading mechanism which may lead to clad rupture under RIA conditions, especially in the case of heterogeneous MOX fuel.

  5. Transport and Burnup Numerical Simulation on the Liquid Blanket Burnup of In-Zinerater%In-Zinerater液态包层输运燃耗数值模拟

    Institute of Scientific and Technical Information of China (English)

    师学明; 杨俊云; 刘成安

    2014-01-01

    Z-Pinch惯性约束聚变是未来一种有竞争力的能源候选方案。Z-Pinch驱动的聚变裂变混合堆可高效地嬗变反应堆乏燃料中分离出的超铀元素。对美国Sandia国家实验室提出的In-Zinerater混合堆概念进行了中子学分析和数值模拟。在三维输运燃耗耦合程序MCORGS中增加了处理在线添加燃料与去除裂变产物的功能,实现了对液态燃料燃耗过程的模拟。增加6Li丰度和燃料初装量保持寿期初反应性不变,可以减缓寿期内反应性下降趋势。逐步增加包层内超铀元素装量,可以控制整个寿期内反应性基本恒定。聚变功率取20 MW,通过反应性控制,5年内包层能量放大倍数在160∼180之间,氚增殖比在1.5∼1.7之间,优于In-Zinerater基准设计方案。%Z-Pinch Inertial confinement fusion is a competitive candidate for future energy solution. A fusion-fission hybrid driven by Z-Pinch can be used to transmute transuranic elements from spent fuels of reactors efficiently. Analysis and numerical simulation of blanket neutronics of In-Zinerater, which is a fusion-fission hybrid concept design in Sandia National Laboratories, is given in this paper. Modification to the three dimension transport and burnup code MCORGS are done, so as to simulate continuous feeding and continuous chemical processing of the liquid fuel. Different combination of initial enrichment of 6Li and fuels loading in the blanket are selected to keep the same reactivity at begin of core. By this way, the decreasing trend of reactivity at life of the core can be lowered. The reactivity can be maintained constant by increasing the fuel loading in the core gradually as the burnup deepens. Given a 20 MW fusion power, by reactivity control, the blanket energy multiplication is around 160∼180 and tritium breed ratio 1.5∼1.7 in 5 years, which is a better result than Sandia’s original design.

  6. Qualification of the B and W Mark B fuel assembly for high burnup. Third semi-annual progress report, July-December 1979

    Energy Technology Data Exchange (ETDEWEB)

    Coleman, T.A.

    1980-03-01

    Five Babcock and Wilcox-designed Mark B (15 x 15) pressurized water reactor fuel assemblies were irradiated to extended burnups in Duke Power Company's Oconee Unit 1 reactor. An assembly average burnup of 40,000 MWd/mtU, which is about 29% greater than previous discharge burnups at Oconee 1, was attained. The nondestructive examination of these five assemblies, which have been irradiated for four fuel cycles, was begun. Data obtained included fuel assembly and fuel dimensions, water channel spacings, fuel rod surface deposit samples, and holddown spring preload forces. Visual examination of the assemblies indicated that good fuel performance was maintained through four cycles of irradiation.

  7. Burnup determination of a fuel element concerning different cooling times; Seguimiento del quemado de un elemento combustible, para diferentes tiempos de enfriamento

    Energy Technology Data Exchange (ETDEWEB)

    Henriquez, C.; Navarro, G.; Pereda, C.; Mutis, O. [Comision Chilena de Energia Nuclear, Santiago (Chile). Dept. de Aplicaciones Nucleares. Unidad de Reactores; Terremoto, Luis A.A.; Zeituni, Carlos A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear

    2002-07-01

    In this work we report a complete set of measurements and some relevant results regarding the burnup process of a fuel element containing low enriched nuclear fuel. This fuel element was fabricated at the Plant of Fuel Elements of the Chilean Nuclear Energy Commission (CCHEN). Measurements were carried out using gamma-ray spectroscopy and the absolute burnup of the fuel element was determined. (author)

  8. Evaluation technology for burnup and generated amount of plutonium by measurement of xenon isotopic ratio in dissolver off-gas at reprocessing facility (Joint research)

    OpenAIRE

    岡野 正紀; 久野 剛彦; 高橋 一朗; 白水 秀知; Charlton, W. S.; Wells, C. A.; Hemberger, P. H.; 山田 敬二; 酒井 敏雄

    2006-01-01

    The amount of Pu in the spent fuel was evaluated from Xe isotopic ratio in off-gas in reprocessing facility, is related to burnup. Six batches of dissolver off-gas at spent fuel dissolution process were sampled from the main stack in Tokai Reprocessing Plant during BWR fuel reprocessing campaign. Xenon isotopic ratio was determined with GC/MS. Burnup and generated amount of Pu were evaluated with Noble Gas Environmental Monitoring Application code (NOVA), developed by Los Alamos National Labo...

  9. THE INVESTIGATION OF BURNUP CHARACTERISTICS USING THE SERPENT MONTE CARLO CODE FOR A SODIUM COOLED FAST REACTOR

    Directory of Open Access Journals (Sweden)

    MEHMET E. KORKMAZ

    2014-06-01

    Full Text Available In this research, we investigated the burnup characteristics and the conversion of fertile 232Th into fissile 233U in the core of a Sodium-Cooled Fast Reactor (SFR. The SFR fuel assemblies were designed for burning 232Th fuel (fuel pin 1 and 233U fuel (fuel pin 2 and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method and TTA (Transmutation Trajectory Analysis method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff was between 0.964 and 0.954 and peaking factor is 1.88867.

  10. The investigation of burnup characteristics using the serpent Monte Carlo code for a sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Korkmaz, Mehmet E.; Agar, Osman [Karamanoglu Mehmetbey University, Faculty of Kamil Oezdag Science, Karaman (Turkmenistan)

    2014-06-15

    In this research, we investigated the burnup characteristics and the conversion of fertile {sup 232}Th into fissile {sup 233}U in the core of a Sodium-Cooled Fast Reactor (SFR). The SFR fuel assemblies were designed for burning {sup 232}Th fuel (fuel pin 1) and {sup 233}U fuel (fuel pin 2) and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method) and TTA (Transmutation Trajectory Analysis) method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff) was between 0.964 and 0.954 and peaking factor is 1.88867.

  11. Study of irradiation induced restructuring of high burnup fuel - Use of computer and accelerator for fuel science and engineering -

    Energy Technology Data Exchange (ETDEWEB)

    Sataka, M.; Ishikawa, N.; Chimn, Y.; Nakamura, J.; Amaya, M. [Japan Atomic Energy Agency, Naka Gun (Japan); Iwasawa, M.; Ohnuma, T.; Sonoda, T. [Central Research Institute of Electric Power Industry, Tokyo (Japan); Kinoshita, M.; Geng, H. Y.; Chen, Y.; Kaneta, Y. [The Univ. of Tokyo, Tokyo (Japan); Yasunaga, K.; Matsumura, S.; Yasuda, K. [Kyushu Univ., Motooka (Japan); Iwase [Osaka Prefecture Univ., Osaka (Japan); Ichinomiya, T.; Nishiuran, Y. [Hokkaido Univ., Kitaku (Japan); Matzke, HJ. [Academy of Ceramics, Karlsruhe (Germany)

    2008-10-15

    In order to develop advanced fuel for future LWR reactors, trials were made to simulate the high burnup restructuring of the ceramics fuel, using accelerator irradiation out of pile and with computer simulation. The target is to reproduce the principal complex process as a whole. The reproduction of the grain subdivision (sub grain formation) was successful at experiments with sequential combined irradiation. It was made by recovery process of the accumulated dislocations, making cells and sub-boundaries at grain boundaries and pore surfaces. Details of the grain sub division mechanism is now in front of us outside of the reactor. Extensive computer science studies, first principle and molecular dynamics gave behavior of fission gas atoms and interstitial oxygen, assisting the high burnup restructuring.

  12. Coolant Density and Control Blade History Effects in Extended BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian J [ORNL; Marshall, William BJ J [ORNL; Bowman, Stephen M [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL; Martinez-Gonzalez, Jesus S [ORNL

    2015-01-01

    Oak Ridge National Laboratory and the US Nuclear Regulatory Commission have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase investigates the applicability of peak reactivity methods currently used for spent fuel pools to spent fuel storage and transportation casks and the validation of reactivity (keff) calculations and predicted spent fuel compositions. The second phase focuses on extending BUC beyond peak reactivity. This paper documents work performed to date investigating some aspects of extended BUC. (The technical basis for application of peak reactivity methods to BWR fuel in storage and transportation systems is presented in a companion paper.) Two reactor operating parameters are being evaluated to establish an adequate basis for extended BWR BUC: (1) the effect of axial void profile and (2) the effect of control blade utilization during operation. A detailed analysis of core simulator data for one cycle of a modern operating BWR plant was performed to determine the range of void profiles and the variability of the profile experienced during irradiation. Although a single cycle does not provide complete data, the data obtained are sufficient to determine the primary effects and to identify conservative modeling approaches. These data were used in a study of the effect of axial void profile. The first stage of the study was determination of the necessary moderator density temporal fidelity in depletion modeling. After the required temporal fidelity was established, multiple void profiles were used to examine the effect on cask reactivity. The results of these studies are being used to develop recommendations for conservatively modeling the void profile effects for BWR depletion calculations. The second operational parameter studied was control blade history. Control blades are inserted in

  13. Temperature and burnup correlated fuel-cladding chemical interaction in U-10ZR metallic fuel

    Science.gov (United States)

    Carmack, William J.

    Metallic fuels are proposed for use in advanced sodium cooled fast reactors and provide a number of advantages over other fuel types considering their fabricability, performance, recyclability, and safety. Resistance to cladding "breach" and subsequent release of fission products and fuel constituents to the nuclear power plant primary coolant system is a key performance parameter for a nuclear fuel system. In metallic fuel, FCCI weakens the cladding, especially at high power-high temperature operation, contributing to fuel pin breach. Empirical relationships for FCCI have been developed from a large body of data collected from in-pile (EBR-II) and out-of-pile experiments [1]. However, these relationships are unreliable in predicting FCCI outside the range of EBR-II experimental data. This dissertation examines new FCCI data extracted from the MFF-series of prototypic length metallic fuel irradiations performed in the Fast Flux Test Facility (FFTF). The fuel in these assemblies operated a temperature and burnup conditions similar to that in EBR-II but with axial fuel height three times longer than EBR-II experiments. Comparing FCCI formation data from FFTF and EBR-II provides new insight into FCCI formation kinetics. A model is developed combining both production and diffusion of lanthanides to the fuel-cladding interface and subsequent reaction with the cladding. The model allows these phenomena to be influenced by fuel burnup (lanthanide concentrations) and operating temperature. Parameters in the model are adjusted to reproduce measured FCCI layer thicknesses from EBR-II and FFTF. The model predicts that, under appropriate conditions, rate of FCCI formation can be controlled by either fission product transport or by the reaction rate of the interaction species at the fuel-cladding interface. This dissertation will help forward the design of metallic fuel systems for advanced sodium cooled fast reactors by allowing the prediction of FCCI layer formation in full

  14. A feasibility study to determine cooling time and burnup of ATR fuel using a nondestructive technique and three types of gamma-ray detectors

    Energy Technology Data Exchange (ETDEWEB)

    Navarro, J.; Aryaeinejad, R.; Nigg, D.W. [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID 83415 (United States)

    2011-07-01

    The goal of this work was to perform a feasibility study and establish measurement techniques to determine the burnup of the Advanced Test Reactor (ATR) fuels at the Idaho National Laboratory (INL). Three different detectors of high purity germanium (HPGe), lanthanum bromide (LaBr{sub 3}), and high pressure xenon (HPXe) in two detection system configurations of below and above the water pool were used in this study. The last two detectors were used for the first time in fuel burnup measurements. The results showed that a better quality spectra can be achieved with the above the water pool configuration. Both short and long cooling time fuels were investigated in order to determine which measurement technique, absolute or fission product ratio, is better suited in each scenario and also to establish what type of detector should be used in each case for the best burnup measurement. The burnup and cooling time calibrations were established using experimental absolute activities or isotopic ratios and ORIGEN burnup calculations. A method was developed to do burnup and cooling time calibrations using fission isotopes activities without the need to know the exact geometry. (authors)

  15. Quantification of the computational accuracy of code systems on the burn-up credit using experimental re-calculations; Quantifizierung der Rechengenauigkeit von Codesystemen zum Abbrandkredit durch Experimentnachrechnungen

    Energy Technology Data Exchange (ETDEWEB)

    Behler, Matthias; Hannstein, Volker; Kilger, Robert; Moser, Franz-Eberhard; Pfeiffer, Arndt; Stuke, Maik

    2014-06-15

    In order to account for the reactivity-reducing effect of burn-up in the criticality safety analysis for systems with irradiated nuclear fuel (''burnup credit''), numerical methods to determine the enrichment and burnup dependent nuclide inventory (''burnup code'') and its resulting multiplication factor k{sub eff} (''criticality code'') are applied. To allow for reliable conclusions, for both calculation systems the systematic deviations of the calculation results from the respective true values, the bias and its uncertainty, are being quantified by calculation and analysis of a sufficient number of suitable experiments. This quantification is specific for the application case under scope and is also called validation. GRS has developed a methodology to validate a calculation system for the application of burnup credit in the criticality safety analysis for irradiated fuel assemblies from pressurized water reactors. This methodology was demonstrated by applying the GRS home-built KENOREST burnup code and the criticality calculation sequence CSAS5 from SCALE code package. It comprises a bounding approach and alternatively a stochastic, which both have been exemplarily demonstrated by use of a generic spent fuel pool rack and a generic dry storage cask, respectively. Based on publicly available post irradiation examination and criticality experiments, currently the isotopes of uranium and plutonium elements can be regarded for.

  16. In Comparative Analysis for Fuel Burnup of Fuel Assembly Designs for the 300 kW Small Medical Reactor

    Science.gov (United States)

    Sambuu, Odmaa; Nanzad, Norov

    2009-03-01

    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 burnup calculations were carried out.

  17. Recent view to the results of pulse tests in the IGR reactor with high burn-up fuel

    Energy Technology Data Exchange (ETDEWEB)

    Asmolov, V.; Yegorova, L. [Russian Research Centre, Moscow (Russian Federation)

    1996-03-01

    Testing of 43 fuel elements (13 fuel elements with high burn-up fuel, 10 fuel elements with preirradiated cladding and fresh fuel, and 20 non-irradiated fuel elements) was carried out in the IGR pulse reactor with a half width of the reactor power pulse of about 0.7 sec. Tests were conducted in capsules with no coolant flow and with standard initial conditions in the capsule of 20{degrees}C and 0.2 MPa. Two types of coolant were used: water and air. One purpose of the test program was to determine the thresholds and mechanisms of fuel rod failure under RIA conditions for VVER fuel rods over their entire exposure range, from zero to high burn-up. These failure thresholds are often used in safety analyses. The tests and analyses were designed to reveal the influence on fuel rod failure of (1) the mechanical properties of the cladding, (2) the pellet-to-cladding gap, (3) fuel burn-up, (4) fuel-to-coolant heat transfer, and other parameters. The resulting data base can also be used for validation of computer codes used for analyzing fuel rod behavior. Three types of test specimens were used in the tests, and diagrams of these specimens are shown in Fig. 1. {open_quotes}Type-C{close_quotes} specimens were re-fabricated from commercial fuel rods of the VVER-1000 type that had been subjected to many power cycles of operation in the Novovoronezh Nuclear Power Plant (NV NPP). {open_quotes}Type-D{close_quotes} specimens were fabricated from the same commercial fuel rods used above, but the high burn-up oxide fuel was removed from the cladding and was replaced with fresh oxide fuel pellets. {open_quotes}Type-D{close_quotes} specimens thus provided a means of separating the effects of the cladding and the oxide fuel pellets and were used to examine cladding effects only.

  18. Characterization of the non-uniqueness of used nuclear fuel burnup signatures through a Mesh-Adaptive Direct Search

    Energy Technology Data Exchange (ETDEWEB)

    Skutnik, Steven E., E-mail: sskutnik@utk.edu; Davis, David R.

    2016-05-01

    The use of passive gamma and neutron signatures from fission indicators is a common means of estimating used fuel burnup, enrichment, and cooling time. However, while characteristic fission product signatures such as {sup 134}Cs, {sup 137}Cs, {sup 154}Eu, and others are generally reliable estimators for used fuel burnup within the context where the assembly initial enrichment and the discharge time are known, in the absence of initial enrichment and/or cooling time information (such as when applying NDA measurements in a safeguards/verification context), these fission product indicators no longer yield a unique solution for assembly enrichment, burnup, and cooling time after discharge. Through the use of a new Mesh-Adaptive Direct Search (MADS) algorithm, it is possible to directly probe the shape of this “degeneracy space” characteristic of individual nuclides (and combinations thereof), both as a function of constrained parameters (such as the assembly irradiation history) and unconstrained parameters (e.g., the cooling time before measurement and the measurement precision for particular indicator nuclides). In doing so, this affords the identification of potential means of narrowing the uncertainty space of potential assembly enrichment, burnup, and cooling time combinations, thereby bounding estimates of assembly plutonium content. In particular, combinations of gamma-emitting nuclides with distinct half-lives (e.g., {sup 134}Cs with {sup 137}Cs and {sup 154}Eu) in conjunction with gross neutron counting (via {sup 244}Cm) are able to reasonably constrain the degeneracy space of possible solutions to a space small enough to perform useful discrimination and verification of fuel assemblies based on their irradiation history.

  19. The calculational VVER burnup Credit Benchmark No.3 results with the ENDF/B-VI rev.5 (1999)

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez Gual, Maritza [Centro de Tecnologia Nuclear, La Habana (Cuba). E-mail: mrgual@ctn.isctn.edu.cu

    2000-07-01

    The purpose of this papers to present the results of CB3 phase of the VVER calculational benchmark with the recent evaluated nuclear data library ENDF/B-VI Rev.5 (1999). This results are compared with the obtained from the other participants in the calculations (Czech Republic, Finland, Hungary, Slovaquia, Spain and the United Kingdom). The phase (CB3) of the VVER calculation benchmark is similar to the Phase II-A of the OECD/NEA/INSC BUC Working Group benchmark for PWR. The cases without burnup profile (BP) were performed with the WIMS/D-4 code. The rest of the cases have been carried with DOTIII discrete ordinates code. The neutron library used was the ENDF/B-VI rev. 5 (1999). The WIMS/D-4 (69 groups) is used to collapse cross sections from the ENDF/B-VI Rev. 5 (1999) to 36 groups working library for 2-D calculations. This work also comprises the results of CB1 (obtained with ENDF/B-VI rev. 5 (1999), too) and CB3 for cases with Burnup of 30 MWd/TU and cooling time of 1 and 5 years and for case with Burnup of 40 MWd/TU and cooling time of 1 year. (author)

  20. OECD/NEA burnup credit criticality benchmarks phase IIIA: Criticality calculations of BWR spent fuel assemblies in storage and transport

    Energy Technology Data Exchange (ETDEWEB)

    Okuno, Hiroshi; Naito, Yoshitaka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ando, Yoshihira [Toshiba Corp., Kawasaki, Kanagawa (Japan)

    2000-09-01

    The report describes the final results of Phase IIIA Benchmarks conducted by the Burnup Credit Criticality Calculation Working Group under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD/NEA). The benchmarks are intended to confirm the predictive capability of the current computer code and data library combinations for the neutron multiplication factor (k{sub eff}) of a layer of irradiated BWR fuel assembly array model. In total 22 benchmark problems are proposed for calculations of k{sub eff}. The effects of following parameters are investigated: cooling time, inclusion/exclusion of FP nuclides and axial burnup profile, and inclusion of axial profile of void fraction or constant void fractions during burnup. Axial profiles of fractional fission rates are further requested for five cases out of the 22 problems. Twenty-one sets of results are presented, contributed by 17 institutes from 9 countries. The relative dispersion of k{sub eff} values calculated by the participants from the mean value is almost within the band of {+-}1%{delta}k/k. The deviations from the averaged calculated fission rate profiles are found to be within {+-}5% for most cases. (author)

  1. SEM Characterization of the High Burn-up Microstructure of U-7Mo Alloy

    Energy Technology Data Exchange (ETDEWEB)

    Dennis D. Keiser, Jr.; Jan-Fong Jue; Jian Gan; Brandon Miller; Adam Robinson; Pavel Medvedev; James Madden; Dan Wachs; M. Teague

    2014-04-01

    During irradiation, the microstructure of U-7Mo evolves until at a fission density near 5x1021 f/cm3 a high-burnup microstructure exists that is very different than what was observed at lower fission densities. This microstructure is dominated by randomly distributed, relatively large, homogeneous fission gas bubbles. The bubble superlattice has collapsed in many microstructural regions, and the fuel grain sizes, in many areas, become sub-micron in diameter with both amorphous fuel and crystalline fuel present. Solid fission product precipitates can be found inside the fission gas bubbles. To generate more information about the characteristics of the high-fission density microstructure, three samples irradiated in the RERTR-7 experiment have been characterized using a scanning electron microscope equipped with a focused ion beam. The FIB was used to generate samples for SEM imaging and to perform 3D reconstruction of the microstructure, which can be used to look for evidence of possible fission gas bubble interlinkage.

  2. Propagation of Nuclear Data Uncertainties for ELECTRA Burn-up Calculations

    Science.gov (United States)

    Sjöstrand, H.; Alhassan, E.; Duan, J.; Gustavsson, C.; Koning, A. J.; Pomp, S.; Rochman, D.; Österlund, M.

    2014-04-01

    The European Lead-Cooled Training Reactor (ELECTRA) has been proposed as a training reactor for fast systems within the Swedish nuclear program. It is a low-power fast reactor cooled by pure liquid lead. In this work, we propagate the uncertainties in 239Pu transport data to uncertainties in the fuel inventory of ELECTRA during the reactor lifetime using the Total Monte Carlo approach (TMC). Within the TENDL project, nuclear models input parameters were randomized within their uncertainties and 740 239Pu nuclear data libraries were generated. These libraries are used as inputs to reactor codes, in our case SERPENT, to perform uncertainty analysis of nuclear reactor inventory during burn-up. The uncertainty in the inventory determines uncertainties in: the long-term radio-toxicity, the decay heat, the evolution of reactivity parameters, gas pressure and volatile fission product content. In this work, a methodology called fast TMC is utilized, which reduces the overall calculation time. The uncertainty of some minor actinides were observed to be rather large and therefore their impact on multiple recycling should be investigated further. It was also found that, criticality benchmarks can be used to reduce inventory uncertainties due to nuclear data. Further studies are needed to include fission yield uncertainties, more isotopes, and a larger set of benchmarks.

  3. Fuel burnup analysis of the TRIGA Mark II Reactor at the University of Pavia

    CERN Document Server

    Chiesa, Davide; Pozzi, Stefano; Previtali, Ezio; Sisti, Monica; Alloni, Daniele; Magrotti, Giovanni; Manera, Sergio; Prata, Michele; Salvini, Andrea; Cammi, Antonio; Zanetti, Matteo; Sartori, Alberto

    2015-01-01

    A time evolution model was developed to study fuel burnup for the TRIGA Mark II reactor at the University of Pavia. The results were used to predict the effects of a complete core reconfiguration and the accuracy of this prediction was tested experimentally. We used the Monte Carlo code MCNP5 to reproduce system neutronics in different operating conditions and to analyse neutron fluxes in the reactor core. The software that took care of time evolution, completely designed in-house, used the neutron fluxes obtained by MCNP5 to evaluate fuel consumption. This software was developed specifically to keep into account some features that differentiate experimental reactors from power ones, such as the daily ON/OFF cycle and the long fuel lifetime. These effects can not be neglected to properly account for neutron poison accumulation. We evaluated the effect of 48 years of reactor operation and predicted a possible new configuration for the reactor core: the objective was to remove some of the fuel elements from the...

  4. Cladding stress during extended storage of high burnup spent nuclear fuel

    Science.gov (United States)

    Raynaud, Patrick A. C.; Einziger, Robert E.

    2015-09-01

    In an effort to assess the potential for low temperature creep and delayed hydride cracking failures in high burnup spent fuel cladding during extended dry storage, the U.S. NRC analytical fuel performance tools were used to predict cladding stress during a 300 year dry storage period for UO2 fuel burned up to 65 GWd/MTU. Fuel swelling correlations were developed and used along with decay gas production and release fractions to produce circumferential average cladding stress predictions with the FRAPCON-3.5 fuel performance code. The resulting stresses did not result in cladding creep failures. The maximum creep strains accumulated were on the order of 0.54-1.04%, but creep failures are not expected below at least 2% strain. The potential for delayed hydride cracking was assessed by calculating the critical flaw size required to trigger this failure mechanism. The critical flaw size far exceeded any realistic flaw expected in spent fuel at end of reactor life.

  5. Fuel failure and fission gas release in high burnup PWR fuels under RIA conditions

    Science.gov (United States)

    Fuketa, Toyoshi; Sasajima, Hideo; Mori, Yukihide; Ishijima, Kiyomi

    1997-09-01

    To study the fuel behavior and to evaluate the fuel enthalpy threshold of fuel rod failure under reactivity initiated accident (RIA) conditions, a series of experiments using pulse irradiation capability of the Nuclear Safety Research Reactor (NSRR) has been performed. During the experiments with 50 MWd/kg U PWR fuel rods (HBO test series; an acronym for high burnup fuels irradiated in Ohi unit 1 reactor), significant cladding failure occurred. The energy deposition level at the instant of the fuel failure in the test is 60 cal/g fuel, and is considerably lower than those expected and pre-evaluated. The result suggests that mechanical interaction between the fuel pellets and the cladding tube with decreased integrity due to hydrogen embrittlement causes fuel failure at the low energy deposition level. After the pulse irradiation, the fuel pellets were found as fragmented debris in the coolant water, and most of these were finely fragmented. This paper describes several key observations in the NSRR experiments, which include cladding failure at the lower enthalpy level, possible post-failure events and large fission gas release.

  6. Development and verification of fuel burn-up calculation model in a reduced reactor geometry

    Energy Technology Data Exchange (ETDEWEB)

    Sembiring, Tagor Malem [Center for Reactor Technology and Nuclear Safety (PTKRN), National Nuclear Energy Agency (BATAN), Kawasan PUSPIPTEK Gd. No. 80, Serpong, Tangerang 15310 (Indonesia)], E-mail: tagorms@batan.go.id; Liem, Peng Hong [Research Laboratory for Nuclear Reactor (RLNR), Tokyo Institute of Technology (Tokyo Tech), O-okayama, Meguro-ku, Tokyo 152-8550 (Japan)

    2008-02-15

    A fuel burn-up model in a reduced reactor geometry (2-D) is successfully developed and implemented in the Batan in-core fuel management code, Batan-FUEL. Considering the bank mode operation of the control rods, several interpolation functions are investigated which best approximate the 3-D fuel assembly radial power distributions across the core as function of insertion depth of the control rods. Concerning the applicability of the interpolation functions, it can be concluded that the optimal coefficients of the interpolation functions are not very sensitive to the core configuration and core or fuel composition in RSG GAS (MPR-30) reactor. Consequently, once the optimal interpolation function and its coefficients are derived then they can be used for 2-D routine operational in-core fuel management without repeating the expensive 3-D neutron diffusion calculations. At the selected fuel elements (at H-9 and G-6 core grid positions), the discrepancy of the FECFs (fuel element channel power peaking factors) between the 2-D and 3-D models are within the range of 3.637 x 10{sup -4}, 3.241 x 10{sup -4} and 7.556 x 10{sup -4} for the oxide, silicide cores with 250 g {sup 235}U/FE and the silicide core with 300 g {sup 235}U/FE, respectively.

  7. TEM Characterization of High Burn-up Microstructure of U-7Mo Alloy

    Energy Technology Data Exchange (ETDEWEB)

    Jian Gan; Brandon Miller; Dennis Keiser; Adam Robinson; James Madden; Pavel Medvedev; Daniel Wachs

    2014-04-01

    As an essential part of global nuclear non-proliferation effort, the RERTR program is developing low enriched U-Mo fuels (< 20% U-235) for use in research and test reactors that currently employ highly enriched uranium fuels. One type of fuel being developed is a dispersion fuel plate comprised of U-7Mo particles dispersed in Al alloy matrix. Recent TEM characterizations of the ATR irradiated U-7Mo dispersion fuel plates include the samples with a local fission densities of 4.5, 5.2, 5.6 and 6.3 E+21 fissions/cm3 and irradiation temperatures of 101-136?C. The development of the irradiated microstructure of the U-7Mo fuel particles consists of fission gas bubble superlattice, large gas bubbles, solid fission product precipitates and their association to the large gas bubbles, grain subdivision to tens or hundreds of nanometer size, collapse of bubble superlattice, and amorphisation. This presentation will describe the observed microstructures specifically focusing on the U-7Mo fuel particles. The impact of the observed microstructure on the fuel performance and the comparison of the relevant features with that of the high burn-up UO2 fuels will be discussed.

  8. Fuel Burnup and Fuel Pool Shielding Analysis for Bushehr Nuclear Reactor VVER-1000

    Science.gov (United States)

    Hadad, Kamal; Ayobian, Navid

    Bushehr Nuclear power plant (BNPP) is currently under construction. The VVER-1000 reactor will be loaded with 126 tons of about 4% enriched fuel having 3-years life cycle. The spent fuel (SF) will be transferred into the spent fuel pool (SPF), where it stays for 8 years before being transferred to Russia. The SPF plays a crucial role during 8 years when the SP resides in there. This paper investigates the shielding of this structure as it is designed to shield the SF radiation. In this study, the SF isotope inventory, for different cycles and with different burnups, was calculated using WIMS/4D transport code. Using MCNP4C nuclear code, the intensity of γ rays was obtained in different layers of SFP shields. These layers include the water above fuel assemblies (FA) in pool, concrete wall of the pool and water laid above transferring fuels. Results show that γ rays leakage from the shield in the mentioned layers are in agreement with the plant's PSAR data. Finally we analyzed an accident were the water height above the FA in the pool drops to 47 cm. In this case it was observed that exposure dose above pool, 10 and 30 days from the accident, are still high and in the levels of 1000 and 758 R/hr.

  9. MCNPX Monte Carlo burnup simulations of the isotope correlation experiments in the NPP Obrigheim

    Energy Technology Data Exchange (ETDEWEB)

    Cao Yan, E-mail: ycao@anl.go [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Gohar, Yousry [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Broeders, Cornelis H.M. [Forschungszentrum Karlsruhe, Institute for Neutron Physics and Reactor Technology, P.O. Box 3640, 76021 Karlsruhe (Germany)

    2010-10-15

    This paper describes the simulation work of the Isotope Correlation Experiment (ICE) using the MCNPX Monte Carlo computer code package. The Monte Carlo simulation results are compared with the ICE-Experimental measurements for burnup up to 30 GWD/t. The comparison shows the good capabilities of the MCNPX computer code package for predicting the depletion of the uranium fuel and the buildup of the plutonium isotopes in a PWR thermal reactor. The Monte Carlo simulation results show also good agreements with the experimental data for calculating several long-lived and stable fission products. However, for the americium and curium actinides, it is difficult to judge the predication capabilities for these actinides due to the large uncertainties in the ICE-Experimental data. In the MCNPX numerical simulations, a pin cell model is utilized to simulate the fuel lattice of the nuclear power reactor. Temperature dependent libraries based on JEFF3.1 nuclear data files are utilized for the calculations. In addition, temperature dependent libraries based ENDF/B-VII nuclear data files are utilized and the obtained results are very close to the JEFF3.1 results, except for {approx}10% differences in the prediction of the minor actinide isotopes buildup.

  10. Propagation of nuclear data uncertainties for ELECTRA burn-up calculations

    CERN Document Server

    ostrand, H; Duan, J; Gustavsson, C; Koning, A; Pomp, S; Rochman, D; Osterlund, M

    2013-01-01

    The European Lead-Cooled Training Reactor (ELECTRA) has been proposed as a training reactor for fast systems within the Swedish nuclear program. It is a low-power fast reactor cooled by pure liquid lead. In this work, we propagate the uncertainties in Pu-239 transport data to uncertainties in the fuel inventory of ELECTRA during the reactor life using the Total Monte Carlo approach (TMC). Within the TENDL project the nuclear models input parameters were randomized within their uncertainties and 740 Pu-239 nuclear data libraries were generated. These libraries are used as inputs to reactor codes, in our case SERPENT, to perform uncertainty analysis of nuclear reactor inventory during burn-up. The uncertainty in the inventory determines uncertainties in: the long-term radio-toxicity, the decay heat, the evolution of reactivity parameters, gas pressure and volatile fission product content. In this work, a methodology called fast TMC is utilized, which reduces the overall calculation time. The uncertainty in the ...

  11. A feasibility and optimization study to determine cooling time and burnup of advanced test reactor fuels using a nondestructive technique

    Energy Technology Data Exchange (ETDEWEB)

    Navarro, Jorge [Univ. of Utah, Salt Lake City, UT (United States)

    2013-12-01

    The goal of this study presented is to determine the best available non-destructive technique necessary to collect validation data as well as to determine burn-up and cooling time of the fuel elements onsite at the Advanced Test Reactor (ATR) canal. This study makes a recommendation of the viability of implementing a permanent fuel scanning system at the ATR canal and leads3 to the full design of a permanent fuel scan system. The study consisted at first in determining if it was possible and which equipment was necessary to collect useful spectra from ATR fuel elements at the canal adjacent to the reactor. Once it was establish that useful spectra can be obtained at the ATR canal the next step was to determine which detector and which configuration was better suited to predict burnup and cooling time of fuel elements non-destructively. Three different detectors of High Purity Germanium (HPGe), Lanthanum Bromide (LaBr3), and High Pressure Xenon (HPXe) in two system configurations of above and below the water pool were used during the study. The data collected and analyzed was used to create burnup and cooling time calibration prediction curves for ATR fuel. The next stage of the study was to determine which of the three detectors tested was better suited for the permanent system. From spectra taken and the calibration curves obtained, it was determined that although the HPGe detector yielded better results, a detector that could better withstand the harsh environment of the ATR canal was needed. The in-situ nature of the measurements required a rugged fuel scanning system, low in maintenance and easy to control system. Based on the ATR canal feasibility measurements and calibration results it was determined that the LaBr3 detector was the best alternative for canal in-situ measurements; however in order to enhance the quality of the spectra collected using this scintillator a deconvolution method was developed. Following the development of the deconvolution method

  12. Thermal property change of MOX and UO2 irradiated up to high burnup of 74 GWd/t

    Science.gov (United States)

    Nakae, Nobuo; Akiyama, Hidetoshi; Miura, Hiromichi; Baba, Toshikazu; Kamimura, Katsuichiro; Kurematsu, Shigeru; Kosaka, Yuji; Yoshino, Aya; Kitagawa, Takaaki

    2013-09-01

    Thermal property is important because it controls fuel behavior under irradiation. The thermal property change at high burnup of more than 70 GWd/t is examined. Two kinds of MOX fuel rods, which were fabricated by MIMAS and SBR methods, and one referenced UO2 fuel rod were used in the experiment. These rods were taken from the pre-irradiated rods (IFA 609/626, of which irradiation test were carried out by Japanese PWR group) and re-fabricated and re-irradiated in HBWR as IFA 702 by JNES. The specification of fuel corresponds to that of 17 × 17 PWR type fuel and the axially averaged linear heat rates (LHR) of MOX rods are 25 kW/m (BOL of IFA 702) and 20 kW/m (EOL of IFA 702). The axial peak burnups achieved are about 74 GWd/t for both of MOX and UO2. Centerline temperature and plenum gas pressure were measured in situ during irradiation. The measured centerline temperature is plotted against LHR at the position where thermocouples are fixed. The slopes of MOX are corresponded to each other, but that of UO2 is higher than those of MOX. This implies that the thermal conductivity of MOX is higher than that of UO2 at high burnup under the condition that the pellet-cladding gap is closed during irradiation. Gap closure is confirmed by the metallography of the postirradiation examinations. It is understood that thermal conductivity of MOX is lower than that of UO2 before irradiation since phonon scattering with plutonium in MOX becomes remarkable. A phonon scattering with plutonium decreases in MOX when burnup proceeds. Thus, thermal conductivity of MOX becomes close to that of UO2. A reverse phenomenon is observed at high burnup region. The phonon scattering with fission products such as Nd and Zr causes a degradation of thermal conductivity of burnt fuel. It might be speculated that this scattering effect causes the phenomenon and the mechanism is discussed here.

  13. Temperature and Burnup Correlated FCCI in U-10Zr Metallic Fuel

    Energy Technology Data Exchange (ETDEWEB)

    William J. Carmack

    2012-05-01

    Metallic fuels are proposed for use in advanced sodium cooled fast reactors. The experience basis for metallic fuels is extensive and includes development and qualification of fuels for the Experimental Breeder Reactor I, the Experimental Breeder Reactor II, FERMI-I, and the Fast Flux Test Facility (FFTF) reactors. Metallic fuels provide a number of advantages over other fuel types in terms of fabricability, performance, recyclability, and safety. Key to the performance of all nuclear fuel systems is the resistance to “breach” and subsequent release of fission products and fuel constituents to the primary coolant system of the nuclear power plant. In metallic fuel, the experience is that significant fuel-cladding chemical (FCCI) interaction occurs and becomes prevalent at high power-high temperature operation and ultimately leads to fuel pin breach and failure. Empirical relationships for metallic fuel pin failure have been developed from a large body of in-pile and out of pile research, development, and experimentation. It has been found that significant in-pile acceleration of the FCCI rate is experienced over similar condition out-of-pile experiments. The study of FCCI in metallic fuels has led to the quantification of in-pile failure rates to establish an empirical time and temperature dependent failure limit for fuel elements. Up until now the understanding of FCCI layer formation has been limited to data generated in EBR-II experiments. This dissertation provides new FCCI data extracted from the MFF-series of metallic fuel irradiations performed in the FFTF. These fuel assemblies contain valuable information on the formation of FCCI in metallic fuels at a variety of temperature and burnup conditions and in fuel with axial fuel height three times longer than EBR-II experiments. The longer fuel column in the FFTF and the fuel pins examined have significantly different flux, power, temperature, and FCCI profiles than that found in similar tests conducted in

  14. Source Term Analysis for Reactor Coolant System with Consideration of Fuel Burnup

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yu Jong; Ahn, Joon Gi; Hwang, Hae Ryong [KEPCO EnC, Daejeon (Korea, Republic of)

    2015-10-15

    The radiation source terms in reactor coolant system (RCS) of pressurized water reactor (PWR) are basic design information for ALARA design such as radiation protection and shielding. Usually engineering companies own self-developed computer codes to estimate the source terms in RCS. DAMSAM and FIPCO are the codes developed by engineering companies. KEPCO E and C has developed computer code, RadSTAR, for use in the Radiation Source Term Analysis for Reactor coolant system during normal operation. The characteristics of RadSTAR are as follows. (1) RadSTAR uses fuel inventory data calculated by ORIGEN, such as ORIGEN2 or ORIGEN-S to consider effects of the fuel burnup. (2) RadSTAR estimates fission products by using finite differential method and analytic method to minimize numerical error. (3) RadSTAR enhances flexibility by adding the function to build the nuclide data library (production pathway library) for user-defined nuclides from ORIGEN data library. (4) RadSTAR consists of two modules. RadSTAR-BL is to build the nuclide data library. RadSTAR-ST is to perform numerical analysis on source terms. This paper includes descriptions on the numerical model, the buildup of nuclide data library, and the sensitivity analysis and verification of RadSTAR. KEPCO E and C developed RadSTAR to calculate source terms in RCS during normal operation. Sensitivity analysis and accuracy verification showed that RadSTAR keeps stability at Δt of 0.1 day and gives more accurate results in comparison with DAMSAM. After development, RadSTAR will replace DAMSAM. The areas, necessary to further development of RadSTAR, are addition of source term calculations for activation products and for shutdown operation.

  15. FY14 Status Report: CIRFT Testing Results on High Burnup UNF

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL; Wang, Hong [ORNL; Jiang, Hao [ORNL

    2014-09-01

    The objective of this project is to perform a systematic study of SNF/UNF (spent nuclear fuel/or used nuclear fuel) integrity under simulated transportation environments by using hot cell testing technology developed recently at Oak Ridge National Laboratory (ORNL), CIRFT (Cyclic Integrated Reversible-Bending Fatigue Tester). Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmarking tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. With support from the US Department of Energy and the NRC, CIRFT testing has been continued. The CIRFT testing was conducted on three HBR rods (R3, R4, and R5), with two specimens failed and one specimen un-failed. The total number of cycles in the test of un-failed specimens went over 2.23 107; the test was stopped as because the specimen did not show any sign of failure. The data analysis on all the HBR SNF rods demonstrated that it is necessary to characterize the fatigue life of used fuel rods in terms of both the curvature amplitude and the maximum of absolute of curvature extremes. The latter is significant because the maxima of extremes signify the maximum of tensile stress of the outer fiber of the bending rod. So far, a large variety of hydrogen contents has been covered in the CIRFT testing on HBR rods. It has been shown that the load amplitude is the dominant factor that controls the lifetime of bending rods, but the hydrogen content also has an important effect on the lifetime attained, according to the load range tested.

  16. Design and construction of a prototype advanced on-line fuel burn-up monitoring system for the modular pebble bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Su, Bingjing; Hawari, Ayman, I.

    2004-03-30

    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 End-of-Life burn-up limit. Unlike the situation with a conventional light water reactor, depending solely on computational methods to perform in-core fuel management for MPBR will be highly inaccurate. An on-line measurement system is needed to accurately assess whether a given pebble has reached its End-of-Life burn-up limit and thereby provide an on-line, automated go/no-go decision on fuel disposition on a pebble-by-pebble basis. This project investigated approaches to analyzing fuel pebbles in real time using gamma spectroscopy and possibly using passive neutron counting of spontaneous fission neutrons to provide the speed, accuracy, and burn-up range required for burnup determination of MPBR. It involved all phases necessary to develop and construct a burn-up monitor, including a review of the design requirements of the system, identification of detection methodologies, modeling and development of potential designs, and finally, the construction and testing of an operational detector system. Based upon the research work performed in this project, the following conclusions are made. In terms of using gamma spectrometry, two possible approaches were identified for burnup assay. The first approach is based on the measurement of the absolute activity of Cs-137. However, due to spectral interference and the need for absolute calibration of the spectrometer, the uncertainty in burnup determination using this approach was found to range from {approx} {+-}40% at beginning of life to {approx} {+-}10% at the discharge burnup. An alternative approach is to use a relative burnup indicator. In this

  17. Development of Monteburns: A Code That Links MCNP and ORIGEN2 in an Automated Fashion for Burnup Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Holly R. Trellue

    1998-12-01

    Monteburns is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code 0RIGEN2. Monteburns produces many criticality and burnup computational parameters based on material feed/removal specifications, power(s), and time intervals. This code processes input from the user indicating the system geometry, initial material compositions, feed/removal, and other code-specific parameters. Results from MCNP, 0RIGEN2, and other calculations are then output successively as the code runs. The principle function of monteburns is to first transfer one-group cross sections and fluxes from MCNP to 0RIGEN2, and then transfer the resulting material compositions (after irradiation and/or decay) from 0RIGEN2 back to MCNP in a repeated, cyclic fashion. The main requirement of the code is that the user have a working MCNP input file and other input parameters; all interaction with 0RIGEN2 and other calculations are performed by monteburns. This report presents the results obtained from the benchmarking of monteburns to measured and previously obtained data from traditional Light Water Reactor systems. The majority of the differences seen between the two were less than five percent. These were primarily a result of variances in cross sections between MCNP, cross section libraries used by other codes, and observed values. With this understanding, this code can now be used with confidence for burnup calculations in three-dimensional systems. It was designed for use in the Accelerator Transmutation of Waste project at Los Alamos National Laboratory but is also being applied to the analysis of isotopic production/destruction of transuranic actinides in a reactor system. The code has now been shown to sufficiently support these calculations.

  18. Accident source terms for pressurized water reactors with high-burnup cores calculated using MELCOR 1.8.5.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Powers, Dana Auburn; Ashbaugh, Scott G.; Leonard, Mark Thomas; Longmire, Pamela

    2010-04-01

    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 emphasis is placed on estimating the effects of high fuel burnup in contrast with low burnup on fission product releases to the containment. Supporting this emphasis, recent data available on fission product release from high-burnup (HBU) fuel from the French VERCOR project are used in this study. The results of these analyses are treated as samples from a population of accident sequences in order to employ approximate order statistics characterization of the results. These trends and tendencies are then compared to the NUREG-1465 alternative source term prescription used today for regulatory applications. In general, greater differences are observed between the state-of-the-art calculations for either HBU or low-burnup (LBU) fuel and the NUREG-1465 containment release fractions than exist between HBU and LBU release fractions. Current analyses suggest that retention of fission products within the vessel and the reactor coolant system (RCS) are greater than contemplated in the NUREG-1465 prescription, and that, overall, release fractions to the containment are therefore lower across the board in the present analyses than suggested in NUREG-1465. The decreased volatility of Cs2MoO4 compared to CsI or CsOH increases the predicted RCS retention of cesium, and as a result, cesium and iodine do not follow identical behaviors with respect to distribution among vessel, RCS, and containment. With respect to the regulatory alternative source term, greater differences are observed between the NUREG-1465 prescription and both HBU and LBU predictions than exist between HBU and LBU

  19. Study of the triton-burnup process in different JET scenarios using neutron monitor based on CVD diamond

    Science.gov (United States)

    Nemtsev, G.; Amosov, V.; Meshchaninov, S.; Popovichev, S.; Rodionov, R.

    2016-11-01

    We present the results of analysis of triton burn-up process using the data from diamond detector. Neutron monitor based on CVD diamond was installed in JET torus hall close to the plasma center. We measure the part of 14 MeV neutrons in scenarios where plasma current varies in a range of 1-3 MA. In this experiment diamond neutron monitor was also able to detect strong gamma bursts produced by runaway electrons arising during the disruptions. We can conclude that CVD diamond detector will contribute to the study of fast particles confinement and help predict the disruption events in future tokamaks.

  20. Efficient and Accurate Calculation of Burnup Problems with Short-Lived Nuclides by a Krylov Subspace Method with the Newton Divided Difference

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yoon Hee; Cho, Nam Zin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-05-15

    Nowadays lattice physics codes tend to utilize a detailed burnup chain including short-lived nuclides in order to perform more accurate burnup calculations. But, since production codes, for example, ORIGEN2, take account of nuclides which have relatively long half-life, it is inappropriate for such detailed burnup chain calculation. To enhance that drawback, many matrix exponential calculation methods have been developed. Recently, a Krylov subspace method with the PADE approximation was used. In this paper, a Krylov subspace method based on spectral decomposition property of the matrix function theory with the Newton divided difference (NDD) is introduced. It is tested with a sample problem and compared with simple Taylor expansion method

  1. Chemical states of fission products in irradiated (U 0.3Pu 0.7)C 1+ x fuel at high burn-ups

    Science.gov (United States)

    Agarwal, Renu; Venugopal, V.

    2006-12-01

    The chemical states of fission products have been theoretically determined for the irradiated carbide fuel of Fast Breeder Test Reactor (FBTR) at Kalpakkam, India, at different burn-ups. The SOLGASMIX-PV computer code was used to determine the equilibrium chemical composition of the fuel. The system was assumed to be composed of a gaseous phase at one atmosphere pressure, and various solid phases. The distribution of elements in these phases and their chemical states at different temperatures were calculated as a function of burn-up. The FBTR fuel, (U 0.3Pu 0.7)C 1+ x, was loaded with C/M values in the range, 1.03-1.06. The present calculations indicated that even for the lowest starting C/M of 1.03 in the FBTR fuel, the liquid metal phase of (U, Pu), should not appear at a burn-up as high as 150 GWd/t.

  2. Instant release of fission products in leaching experiments with high burn-up nuclear fuels in the framework of the Euratom project FIRST- Nuclides

    Science.gov (United States)

    Lemmens, K.; González-Robles, E.; Kienzler, B.; Curti, E.; Serrano-Purroy, D.; Sureda, R.; Martínez-Torrents, A.; Roth, O.; Slonszki, E.; Mennecart, T.; Günther-Leopold, I.; Hózer, Z.

    2017-02-01

    The instant release of fission products from high burn-up UO2 fuels and one MOX fuel was investigated by means of leach tests. The samples covered PWR and BWR fuels at average rod burn-up in the range of 45-63 GWd/tHM and included clad fuel segments, fuel segments with opened cladding, fuel fragments and fuel powder. The tests were performed with sodium chloride - bicarbonate solutions under oxidizing conditions and, for one test, in reducing Ar/H2 atmosphere. The iodine and cesium release could be partially explained by the differences in sample preparation, leading to different sizes and properties of the exposed surface areas. Iodine and cesium releases tend to correlate with FGR and linear power rating, but the scatter of the data is significant. Although the gap between the fuel and the cladding was closed in some high burn-up samples, fissures still provide possible preferential transport pathways.

  3. Accident source terms for light-water nuclear power plants using high-burnup or MOX fuel.

    Energy Technology Data Exchange (ETDEWEB)

    Salay, Michael (U.S. Nuclear Regulatory Commission, Washington, D.C.); Gauntt, Randall O.; Lee, Richard Y. (U.S. Nuclear Regulatory Commission, Washington, D.C.); Powers, Dana Auburn; Leonard, Mark Thomas

    2011-01-01

    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 release fractions of nine chemical classes of radionuclides as calculated with the MELCOR 1.8.5 accident analysis computer code. The accident phases are those defined in the NUREG-1465 Source Term - gap release, in-vessel release, ex-vessel release, and late in-vessel release. Important differences among the accident source terms derived here and the NUREG-1465 Source Term are not attributable to either fuel burnup or use of MOX fuel. Rather, differences among the source terms are due predominantly to improved understanding of the physics of core meltdown accidents. Heat losses from the degrading reactor core prolong the process of in-vessel release of radionuclides. Improved understanding of the chemistries of tellurium and cesium under reactor accidents changes the predicted behavior characteristics of these radioactive elements relative to what was assumed in the derivation of the NUREG-1465 Source Term. An additional radionuclide chemical class has been defined to account for release of cesium as cesium molybdate which enhances molybdenum release relative to other metallic fission products.

  4. Neutron Transport and Nuclear Burnup Analysis for the Laser Inertial Confinement Fusion-Fission Energy (LIFE) Engine

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, K J; Latkowski, J F; Abbott, R P; Boyd, J K; Powers, J J; Seifried, J E

    2008-10-24

    Lawrence Livermore National Laboratory is currently developing a hybrid fusion-fission nuclear energy system, called LIFE, to generate power and burn nuclear waste. We utilize inertial confinement fusion to drive a subcritical fission blanket surrounding the fusion chamber. It is composed of TRISO-based fuel cooled by the molten salt flibe. Low-yield (37.5 MJ) targets and a repetition rate of 13.3 Hz produce a 500 MW fusion source that is coupled to the subcritical blanket, which provides an additional gain of 4-8, depending on the fuel. In the present work, we describe the neutron transport and nuclear burnup analysis. We utilize standard analysis tools including, the Monte Carlo N-Particle (MCNP) transport code, ORIGEN2 and Monteburns to perform the nuclear design. These analyses focus primarily on a fuel composed of depleted uranium not requiring chemical reprocessing or enrichment. However, other fuels such as weapons grade plutonium and highly-enriched uranium are also under consideration. In addition, we have developed a methodology using {sup 6}Li as a burnable poison to replace the tritium burned in the fusion targets and to maintain constant power over the lifetime of the engine. The results from depleted uranium analyses suggest up to 99% burnup of actinides is attainable while maintaining full power at 2GW for more than five decades.

  5. Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

    Science.gov (United States)

    Afifah, Maryam; Miura, Ryosuke; Su'ud, Zaki; Takaki, Naoyuki; Sekimoto, H.

    2015-09-01

    Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don't need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design. The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.

  6. Benchmark calculation of SCALE-PC 4.3 CSAS6 module and burnup credit criticality analysis

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Hee Sung; Ro, Seong Gy; Shin, Young Joon; Kim, Ik Soo [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-12-01

    Calculation biases of SCALE-PC CSAS6 module for PWR spent fuel, metallized spent fuel and solution of nuclear materials have been determined on the basis of the benchmark to be 0.01100, 0.02650 and 0.00997, respectively. With the aid of the code system, nuclear criticality safety analysis for the spent fuel storage pool has been carried out to determine the minimum burnup of spent fuel required for safe storage. The criticality safety analysis is performed using three types of isotopic composition of spent fuel: ORIGEN2-calculated isotopic compositions; the conservative inventory obtained from the multiplication of ORIGEN2-calculated isotopic compositions by isotopic correction factors; the conservative inventory of only U, Pu and {sup 241}Am. The results show that the minimum burnup for three cases are 990,6190 and 7270 MWd/tU, respectively in the case of 5.0 wt% initial enriched spent fuel. (author). 74 refs., 68 figs., 35 tabs.

  7. Development of a method for xenon determination in the microstructure of high burn-up nuclear fuel[Dissertation 17527

    Energy Technology Data Exchange (ETDEWEB)

    Horvath, M. I

    2008-07-01

    In nuclear fuel, in approximately one quarter of the fissions, one of the two formed fission products is gaseous. These are mainly the noble gases xenon and krypton with isotopes of xenon contributing up to 90% of the product gases. These noble fission gases do not combine with other species, and have a low solubility in the normally used uranium oxide matrix. They can be dissolved in the fuel matrix or precipitate in nanometer-sized bubbles within the fuel grain, in micrometer-sized bubbles at the grain boundaries, and a fraction also precipitates in fuel pores, coming from fuel fabrication. A fraction of the gas can also be released into the plenum of the fuel rod. With increasing fission, and therefore burn-up, the ceramic fuel material experiences a transformation of its structure in the 'cooler' rim region of the fuel. A subdivision occurs of the original fuel grains of few microns size into thousands of small grains of sub-micron sizes. Additionally, larger pores are formed, which also leads into an increasing porosity in the fuel rim, called high burn-up structure. In this structure, only a small fraction of the fission gas remains in the matrix, the major quantity is said to accumulate in these pores. Because of this accumulation, the knowledge of the quantities of gas within these pores is of major interest in consideration to burn-up, fuel performance and especially for safety issues. In case of design based accidents, i.e. rapidly increasing temperature transients, the behavior of the fuel has to be estimated. Various analytical techniques have been used to determine the Xe concentration in nuclear fuel samples. The capabilities of EPMA (Electron Probe Micro-Analyser) and SIMS (Secondary Ion Mass Spectrometry) have been studied and provided some qualitative information, which has been used for determining Xe-matrix concentrations. First approaches combining these two techniques to estimate pore pressures have been recently reported. However

  8. Burn-up calculation of different thorium-based fuel matrixes in a thermal research reactor using MCNPX 2.6 code

    Directory of Open Access Journals (Sweden)

    Gholamzadeh Zohreh

    2014-12-01

    Full Text Available Decrease of the economically accessible uranium resources and the inherent proliferation resistance of thorium fuel motivate its application in nuclear power systems. Estimation of the nuclear reactor’s neutronic parameters during different operational situations is of key importance for the safe operation of nuclear reactors. In the present research, thorium oxide fuel burn-up calculations for a demonstrative model of a heavy water- -cooled reactor have been performed using MCNPX 2.6 code. Neutronic parameters for three different thorium fuel matrices loaded separately in the modelled thermal core have been investigated. 233U, 235U and 239Pu isotopes have been used as fissile element in the thorium oxide fuel, separately. Burn-up of three different fuels has been calculated at 1 MW constant power. 135X and 149Sm concentration variations have been studied in the modelled core during 165 days burn-up. Burn-up of thorium oxide enriched with 233U resulted in the least 149Sm and 135Xe productions and net fissile production of 233U after 165 days. The negative fuel, coolant and void reactivity of the used fuel assures safe operation of the modelled thermal core containing (233U-Th O2 matrix. Furthermore, utilisation of thorium breeder fuel demonstrates several advantages, such as good neutronic economy, 233U production and less production of long-lived α emitter high radiotoxic wastes in biological internal exposure point of view

  9. Impacts of burnup-dependent swelling of metallic fuel on the performance of a compact breed-and-burn fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hartanto, Donny; Heo, Woong; Kim, Chi Hyung; Kim, Yong Hee [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology (KAIST), Daejeon (Korea, Republic of)

    2016-04-15

    The U-Zr or U-TRU-Zr cylindrical metallic fuel slug used in fast reactors is known to swell significantly and to grow during irradiation. In neutronics simulations of metallic-fueled fast reactors, it is assumed that the slug has swollen and contacted cladding, and the bonding sodium has been removed from the fuel region. In this research, a realistic burnup-dependent fuel-swelling simulation was performed using Monte Carlo code McCARD for a single-batch compact sodium-cooled breed-and-burn reactor by considering the fuel-swelling behavior reported from the irradiation test results in EBR-II. The impacts of the realistic burnup-dependent fuel swelling are identified in terms of the reactor neutronics performance, such as core lifetime, conversion ratio, axial power distribution, and local burnup distributions. It was found that axial fuel growth significantly deteriorated the neutron economy of a breed-and-burn reactor and consequently impaired its neutronics performance. The bonding sodium also impaired neutron economy, because it stayed longer in the blanket region until the fuel slug reached 2% burnup.

  10. Estimate of fuel burnup spatial a multipurpose reactor in computer simulation; Estimativa da queima espacial do combustivel de um reator multiproposito por simulacao computacional

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Nadia Rodrigues dos, E-mail: nadia.santos@ifrj.edu.br [Instituto Federal de Educacao, Ciencia e Tecnologia do Rio de Janeiro (IFRJ), Paracambi, RJ (Brazil); Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes, E-mail: malu@ien.gov.br, E-mail: zrlima@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    In previous research, which aimed, through computer simulation, estimate the spatial fuel burnup for the research reactor benchmark, material test research - International Atomic Energy Agency (MTR/IAEA), it was found that the use of the code in FORTRAN language, based on the diffusion theory of neutrons and WIMSD-5B, which makes cell calculation, bespoke be valid to estimate the spatial burnup other nuclear research reactors. That said, this paper aims to present the results of computer simulation to estimate the space fuel burnup of a typical multipurpose reactor, plate type and dispersion. the results were considered satisfactory, being in line with those presented in the literature. for future work is suggested simulations with other core configurations. are also suggested comparisons of WIMSD-5B results with programs often employed in burnup calculations and also test different methods of interpolation values obtained by FORTRAN. Another proposal is to estimate the burning fuel, taking into account the thermohydraulics parameters and the appearance of xenon. (author)

  11. Corrosion studies with high burnup light water reactor fuel. Release of nuclides into simulated groundwater during accumulated contact time of up to two years

    Energy Technology Data Exchange (ETDEWEB)

    Zwicky, Hans-Urs (Zwicky Consulting GmbH, Remigen (Switzerland)); Low, Jeanett; Ekeroth, Ella (Studsvik Nuclear AB, Nykoeping (Sweden))

    2011-03-15

    In the framework of comprehensive research work supporting the development of a Swedish concept for the disposal of highly radioactive waste and spent fuel, Studsvik has performed a significant number of spent fuel corrosion studies under a variety of different conditions. These experiments, performed between 1990 and 2002, covered a burnup range from 27 to 49 MWd/kgU, which was typical for fuel to be disposed at that time. As part of this work, the so called Series 11 tests were performed under oxidising conditions in synthetic groundwater with fuel samples from a rod irradiated in the Ringhals 1 Boiling Water Reactor (BWR). In the meantime, Swedish utilities tend to increase the discharge burnup of fuel operated in their reactors. This means that knowledge of spent fuel corrosion performance has to be extended to higher burnup as well. Therefore, a series of experiments has been started at Studsvik, aiming at extending the data base acquired in the Series 11 corrosion tests to higher burnup fuel. Fuel burnup leads to complex and significant changes in the composition and properties of the fuel. The transformed microstructure, which is referred to as the high burnup structure or rim structure in the outer region of the fuel, consists of small grains of submicron size and a high concentration of pores of typical diameter 1 to 2 mum. This structure forms in UO{sub 2} fuel at a local burnup above 50 MWd/kgU, as long as the temperature is below 1,000-1,100 deg C. The high burnup at the pellet periphery is the consequence of plutonium build-up by neutron capture in 238U followed by fission of the formed plutonium. The amount of fission products in the fuel increases more or less linearly with burnup, in contrast to alpha emitting actinides that increase above average. As burnup across a spent fuel pellet is not uniform, but increases towards the periphery, the radiation field is also larger at the pellet surface. At the same time, it is easier for water to access the

  12. A study of fuel failure behavior in high burnup HTGR fuel. Analysis by STRESS3 and STAPLE codes

    Energy Technology Data Exchange (ETDEWEB)

    Martin, David G.; Sawa, Kazuhiro; Ueta, Shouhei; Sumita, Junya [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2001-05-01

    In current high temperature gas-cooled reactors (HTGRs), Tri-isotropic coated fuel particles are employed as fuel. In safety design of the HTGR fuels, it is important to retain fission products within particles so that their release to primary coolant does not exceed an acceptable level. From this point of view, the basic design criteria for the fuel are to minimize the failure fraction of as-fabricated fuel coating layers and to prevent significant additional fuel failures during operation. This report attempts to model fuel behavior in irradiation tests using the U.K. codes STRESS3 and STAPLE. Test results in 91F-1A and HRB-22 capsules irradiation tests, which were carried out at the Japan Materials Testing Reactor of JAERI and at the High Flux Isotope Reactor of Oak Ridge National Laboratory, respectively, were employed in the calculation. The maximum burnup and fast neutron fluence were about 10%FIMA and 3 x 10{sup 25} m{sup -2}, respectively. The fuel for the irradiation tests was called high burnup fuel, whose target burnup and fast neutron fluence were higher than those of the first-loading fuel of the High Temperature Engineering Test Reactor. The calculation results demonstrated that if only mean fracture stress values of PyC and SiC are used in the calculation it is not possible to predict any particle failures, by which is meant when all three load bearing layers have failed. By contrast, when statistical variations in the fracture stresses and particle specifications are taken into account, as is done in the STAPLE code, failures can be predicted. In the HRB-22 irradiation test, it was concluded that the first two particles which had failed were defective in some way, but that the third and fourth failures can be accounted for by the pressure vessel model. In the 91F-1A irradiation test, the result showed that 1 or 2 particles had failed towards the end of irradiation in the upper capsule and no particles failed in the lower capsule. (author)

  13. A Feasibility Study to Determine Cooling Time and Burnup of ATR Fuel Using a Nondestructive Technique and Three Types of Gamma-ray Detectors

    Energy Technology Data Exchange (ETDEWEB)

    Jorge Navarro; Rahmat Aryaeinejad,; David W. Nigg

    2011-05-01

    A Feasibility Study to Determine Cooling Time and Burnup of ATR Fuel Using a Nondestructive Technique1 Rahmat Aryaeinejad, Jorge Navarro, and David W Nigg Idaho National Laboratory Abstract Effective and efficient Advanced Test Reactor (ATR) fuel management require state of the art core modeling tools. These new tools will need isotopic and burnup validation data before they are put into production. To create isotopic, burn up validation libraries and to determine the setup for permanent fuel scanner system a feasibility study was perform. The study consisted in measuring short and long cooling time fuel elements at the ATR canal. Three gamma spectroscopy detectors (HPGe, LaBr3, and HPXe) and two system configurations (above and under water) were used in the feasibility study. The first stage of the study was to investigate which detector and system configuration would be better suited for different scenarios. The second stage of the feasibility study was to create burnup and cooling time calibrations using experimental isotopic data collected and ORIGEN 2.2 burnup data. The results of the study establish that a better spectra resolution is achieve with an above the water configuration and that three detectors can be used in the permanent fuel scanner system for different situations. In addition it was conclude that a number of isotopic ratios and absolute measurements could be used to predict ATR fuel burnup and cooling times. 1This work was supported by the U.S. Depart¬ment of Energy (DOE) under Battelle Energy Alliance, LLC Contract No. DE-AC07-05ID14517.

  14. Fully Coupled Modeling of Burnup-Dependent (U1- y , Pu y )O2- x Mixed Oxide Fast Reactor Fuel Performance

    Science.gov (United States)

    Liu, Rong; Zhou, Wenzhong; Zhou, Wei

    2016-03-01

    During the fast reactor nuclear fuel fission reaction, fission gases accumulate and form pores with the increase of fuel burnup, which decreases the fuel thermal conductivity, leading to overheating of the fuel element. The diffusion of plutonium and oxygen with high temperature gradient is also one of the important fuel performance concerns as it will affect the fuel material properties, power distribution, and overall performance of the fuel pin. In order to investigate these important issues, the (U1- y Pu y )O2- x fuel pellet is studied by fully coupling thermal transport, deformation, oxygen diffusion, fission gas release and swelling, and plutonium redistribution to evaluate the effects on each other with burnup-dependent models, accounting for the evolution of fuel porosity. The approach was developed using self-defined multiphysics models based on the framework of COMSOL Multiphysics to manage the nonlinearities associated with fast reactor mixed oxide fuel performance analysis. The modeling results showed a consistent fuel performance comparable with the previous results. Burnup degrades the fuel thermal conductivity, resulting in a significant fuel temperature increase. The fission gas release increased rapidly first and then steadily with the burnup increase. The fuel porosity increased dramatically at the beginning of the burnup and then kept constant as the fission gas released to the fuel free volume, causing the fuel temperature to increase. Another important finding is that the deviation from stoichiometry of oxygen affects greatly not only the fuel properties, for example, thermal conductivity, but also the fuel performance, for example, temperature distribution, porosity evolution, grain size growth, fission gas release, deformation, and plutonium redistribution. Special attention needs to be paid to the deviation from stoichiometry of oxygen in fuel fabrication. Plutonium content will also affect the fuel material properties and performance

  15. French investigations of high burnup effect on LOCA thermomechanical behavior: Part 1. Experimental programmes in support of LOCA design methodologies

    Energy Technology Data Exchange (ETDEWEB)

    Waeckel, N. [EDF/SEPTEN Villeurbanne (France); GrandJean, C. [IPSN, Cadarache (France); Cauvin, R.; Lebuffe, C. [EDF/SCMI, Chinon (France)

    1997-01-01

    Within the framework of Burn-Up extension request, EDF, FRAMATOME, CEA and IPSN have carried out experimental programmes in order to provide the design of fuel rods under LOCA conditions with relevant data. The design methods used in France for LOCA are based on standard Appendix K methodology updated to take into account some penalties related to the actual conditions of the Nuclear Power Plant. Best-Estimate assessments are used as well. Experimental programmes concern plastic deformation and burst behavior of advanced claddings (EDGAR) and thermal shock quenching behavior of highly irradiated claddings (TAGCIR). The former reveals the important role played by the {alpha}/{beta} transformation kinetics related to advanced alloys (Niobium alloys) and the latter the significative impact of hydrogen charged during in-reactor corrosion on oxidation kinetics and failure behavior in terms of cooling rates.

  16. Verification of spectral burn-up codes on 2D fuel assemblies of the GFR demonstrator ALLEGRO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Čerba, Štefan, E-mail: stefan.cerba@stuba.sk [Slovak University of Technology in Bratislava, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava (Slovakia); Vrban, Branislav; Lüley, Jakub [Slovak University of Technology in Bratislava, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava (Slovakia); Dařílek, Petr [VUJE a.s., Okružná 5, 918 64 Trnava (Slovakia); Zajac, Radoslav, E-mail: radoslav.zajac@vuje.sk [VUJE a.s., Okružná 5, 918 64 Trnava (Slovakia); Nečas, Vladimír; Haščik, Ján [Slovak University of Technology in Bratislava, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava (Slovakia)

    2014-02-15

    Highlights: • Verification of the MCNPX, HELIOS and SCALE codes. • MOX and ceramic fuel assembly. • Gas-cooled fast reactor. • Burnup calculation. - Abstract: The gas-cooled fast reactor, which is one of the six GEN IV reactor concepts, is characterized by high operational temperatures and a hard neutron spectrum. The utilization of commonly used spectral codes, developed mainly for LWR reactors operated in the thermal/epithermal neutron spectrum, may be connected with systematic deviations since the main development effort of these codes has been focused on the thermal part of the neutron spectrum. To be able to carry out proper calculations for fast systems the used codes have to account for neutron resonances including the self-shielding effect. The presented study aims at verifying the spectral HELIOS, MCNPX and SCALE codes on the basis of depletion calculations of 2D MOX and ceramic fuel assemblies of the ALLEGRO gas-cooled fast reactor demonstrator in infinite lattice.

  17. Analysis of Corrosion Residues Collected from the Aluminum Basket Rails of the High-Burnup Demonstration Cask.

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-03-01

    On September, 2015, an inspection was performed on the TN-32B cask that will be used for the high-burnup demonstration project. During the survey, wooden cribbing that had been placed within the cask eleven years earlier to prevent shifting of the basket during transport was removed, revealing two areas of residue on the aluminum basket rails, where they had contacted the cribbing. The residue appeared to be a corrosion product, and concerns were raised that similar attack could exist at more difficult-to-inspect locations in the canister. Accordingly, when the canister was reopened, samples of the residue were collected for analysis. This report presents the results of that assessment, which determined that the corrosion was due to the presence of the cribbing. The corrosion was associated with fungal material, and fungal activity likely contributed to an aggressive chemical environment. Once the cask has been cleaned, there will be no risk of further corrosion.

  18. Development of a method for xenon determination in the microstructure of high burn-up nuclear fuel[Dissertation 17527

    Energy Technology Data Exchange (ETDEWEB)

    Horvath, M. I

    2008-07-01

    In nuclear fuel, in approximately one quarter of the fissions, one of the two formed fission products is gaseous. These are mainly the noble gases xenon and krypton with isotopes of xenon contributing up to 90% of the product gases. These noble fission gases do not combine with other species, and have a low solubility in the normally used uranium oxide matrix. They can be dissolved in the fuel matrix or precipitate in nanometer-sized bubbles within the fuel grain, in micrometer-sized bubbles at the grain boundaries, and a fraction also precipitates in fuel pores, coming from fuel fabrication. A fraction of the gas can also be released into the plenum of the fuel rod. With increasing fission, and therefore burn-up, the ceramic fuel material experiences a transformation of its structure in the 'cooler' rim region of the fuel. A subdivision occurs of the original fuel grains of few microns size into thousands of small grains of sub-micron sizes. Additionally, larger pores are formed, which also leads into an increasing porosity in the fuel rim, called high burn-up structure. In this structure, only a small fraction of the fission gas remains in the matrix, the major quantity is said to accumulate in these pores. Because of this accumulation, the knowledge of the quantities of gas within these pores is of major interest in consideration to burn-up, fuel performance and especially for safety issues. In case of design based accidents, i.e. rapidly increasing temperature transients, the behavior of the fuel has to be estimated. Various analytical techniques have been used to determine the Xe concentration in nuclear fuel samples. The capabilities of EPMA (Electron Probe Micro-Analyser) and SIMS (Secondary Ion Mass Spectrometry) have been studied and provided some qualitative information, which has been used for determining Xe-matrix concentrations. First approaches combining these two techniques to estimate pore pressures have been recently reported. However

  19. PLUTON: Three-group neutronic code for burnup analysis of isotope generation and depletion in highly irradiated LWR fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Lemehov, Sergei E; Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-08-01

    PLUTON is a three-group neutronic code analyzing, as functions of time and burnup, the change of radial profiles, together with average values, of power density, burnup, concentration of trans-uranium elements, plutonium buildup, depletion of fissile elements, and fission product generation in water reactor fuel rod with standard UO{sub 2}, UO{sub 2}-Gd{sub 2}O{sub 3}, inhomogeneous MOX, and UO{sub 2}-ThO{sub 2}. The PLUTON code, which has been designed to be run on Windows PC, has adopted a theoretical shape function of neutron attenuation in pellet, which enables users to perform a very fast and accurate calculation easily. The present code includes the irradiation conditions of the Halden Reactor which gives verification data for the code. The total list of trans-uranium elements included in the calculations consists of {sub 92}U{sup 233-239}, {sub 93}Np{sup 237-239}, {sub 94}Pu{sup 238-243}, {sub 95}Am{sup 241-244} (including isomers), and {sub 96}Cm{sup 242-245}. Poisoning fission products are represented by {sub 54}Xe{sup 131,133,135}, {sub 48}Cd{sup 113}, {sub 62}Sm{sup 149,151,152}, {sub 64}Gd{sup 154-160}, {sub 63}Eu{sup 153,155}, {sub 36}Kr{sup 83,85}, {sub 42}Mo{sup 95}, {sub 43}Tc{sup 99}, {sub 45}Rh{sup 103}, {sub 47}Ag{sup 109}, {sub 53}I{sup 127,129,131}, {sub 55}Cs{sup 133}, {sub 57}La{sup 139}, {sub 59}Pr{sup 141}, {sub 60}Nd{sup 143-150}, {sub 61}Pm{sup 147}. Fission gases and volatiles included in the code are {sub 36}Kr{sup 83-86}, {sub 54}Xe{sup 129-136}, {sub 52}Te{sup 125-130}, {sub 53}I{sup 127-131}, {sub 55}Cs{sup 133-137}, and {sub 56}Ba{sup 135-140}. Verification has been performed up to 83 GWd/tU, and a satisfactory agreement has been obtained. (author)

  20. Determination of burnup grade of fuel plates by gamma spectrometry; Determinacao do grau de queima em elementos combustiveis tipo placa por meio de espectrometria gama

    Energy Technology Data Exchange (ETDEWEB)

    Terremoto, Luis A.A.; Zeituni, Carlos A.; Perrotta, Jose A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Div. de Engenharia do Nucleo

    1999-11-01

    This work describes absolute burnup measurements on spent MTR fuel elements by means of non-destructive gamma-ray spectroscopy which correlates activities of radioactive fission products with the fissioned mass of {sup 235} U. Experiments based on such method were performed at the storage pool area of the IEA-R1 research reactor. The obtained results were compared with calculational ones based on neutronics. (author) 7 refs., 6 figs., 1 tab.; e-mail: laaterre at net.ipen.br

  1. Isotopic analyses and calculation by use of JENDL-3.2 for high burn-up UO{sub 2} and MOX spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Sasahara, Akihiro; Matsumura, Tetsuo [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab.; Nicolaou, G.; Betti, M.; Walker, C.T.

    1997-03-01

    The post irradiation examinations (PIE) were carried out for high burn-up UO{sub 2} spent fuel (3.8%U235, average burn-up:60GWd/t) and mixed oxide (MOX) spent fuel (5.07%Pu, average burn-up:45GWd/t). The PIE includes, (a) isotopic analysis, (b) electron probe microanalysis (EPMA) in pellet cross section and so on. The results of isotopic analyses and EPMA were compared with ORIGEN2/82 and VIM-BURN calculation results. In VIM-BURN calculation, the nuclear data of actinides were proceeded from new data file, JENDL-3.2. The sensitivities of power history and moderator density to nuclides composition were investigated by VIM-BURN calculation and consequently power history mainly effected on Am241 and Am242m and moderator density effected on fissile nuclides. From EPMA results of U and Pu distribution in pellet, VIM-BURN calculation showed reasonable distribution in pellet cross section. (author)

  2. Raman micro-spectroscopy of UOX and MOX spent nuclear fuel characterization and oxidation resistance of the high burn-up structure

    Science.gov (United States)

    Jegou, C.; Gennisson, M.; Peuget, S.; Desgranges, L.; Guimbretière, G.; Magnin, M.; Talip, Z.; Simon, P.

    2015-03-01

    Raman micro-spectroscopy was applied to study the structure and oxidation resistance of UO2 (burnup 60 GWd/tHM) and MOX (burnup 47 GWd/tHM) irradiated fuels. The Raman technique, adapted to working under extreme conditions, enabled structural information to be obtained at the cubic micrometer scale in various zones of interest within irradiated fuel (central and zones like the Rim for UOX60, and the plutonium-enriched agglomerates for MOX47 characterized by a high burn-up structure), and the study of their oxidation resistance. As regards the structural information after irradiation, the spectra obtained make up a set of data consistent with the systematic presence of the T2g band characteristic of the fluorite structure, and of a triplet band located between 500 and 700 cm-1. The existence of this triplet can be attributed to the presence of defects originating in changes to the fuel chemistry occurring in the reactor (presence of fission products) and to the accumulation of irradiation damage. As concerns the oxidation resistance of the different zones of interest, Raman spectroscopy results confirmed the good stability of the restructured zones (plutonium-enriched agglomerates and Rim) rich in fission products compared to the non-restructured UO2 grains. A greater structural stability was noticed in the case of high plutonium content agglomerates, as this element favors the maintenance of the fluorite structure.

  3. Measurement of the composition of noble-metal particles in high-burnup CANDU fuel by wavelength dispersive X-ray microanalysis

    Energy Technology Data Exchange (ETDEWEB)

    Hocking, W.H.; Szostak, F.J

    1999-09-01

    An investigation of the composition of the metallic inclusions in CANDU fuel, which contain Mo, Tc, Ru, Rh and Pd, has been conducted as a function of burnup by wavelength dispersive X-ray (WDX) microanalysis. Quantitative measurements were performed on micrometer sized particles embedded in thin sections of fuel using elemental standards and the ZAF method. Because the fission yields of the noble metals change with burnup, as a consequence of a shift from almost entirely {sup 235}U fission to mainly {sup 239}Pu fission, their inventories were calculated from the fuel power histories using the WIMS-Origin code for comparison with experiment. Contrary to expectations that the oxygen potential would be buffered by progressive Mo oxidation, little evidence was obtained for reduced incorporation of Mo in the noble-metal particles at high burnup. These surprising results are discussed with respect to the oxygen balance in irradiated CANDU fuels and the likely intrinsic and extrinsic sinks for excess oxygen. (author)

  4. Analysis of high burnup fuel behavior under control rod ejection accident in Korea standard nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bok; Lee, Chung Chan; Kim, Oh Hwan; Kim, Jong Jin [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-07-01

    Test results of high burnup fuel behavior under RIA(reactivity insertion accident) indicated that fuel might fail at the fuel enthalpy lower than that in the current fuel failure criteria was derived by the conservative assumptions and analysis of fuel failure mechanisms, and applied to the analysis of control rod ejection accident in the 1,000 MWe Korea standard PWR. Except that three dimensional core analysis was performed instead of conventional zero dimensional analysis, all the other conservative assumptions were kept. Analysis results showed that less than on percent of the fuel rods in the core has failed which was much less than the conventional fuel failure fraction, 9.8 %, even though a newly derived fuel failure criteria -Fuel failure occurs at the power level lower than that in the current fuel failure criteria. - was applied, since transient fuel rod power level was significantly decreased by analyzing the transient fuel rod power level was significantly decreased by analyzing the transient core three dimensionally. Therefore, it can be said that results of the radiological consequence analysis for the control rod ejection accident in the FSAR where fuel failure fraction was assumed 9.8 % is still bounding. 18 tabs., 48 figs., 39 refs. (Author).

  5. Thermochemical prediction of chemical form distributions of fission products in LWR oxide fuels irradiated to high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Moriyama, Kouki; Furuya, Hirotaka [Kyushu Univ., Fukuoka (Japan). Faculty of Engineering

    1997-09-01

    Based on the result of micro-gamma scanning of a fuel pin irradiated to high burnup in a commercial PWR, the radial distribution of chemical forms of fission products (FPs) in LWR fuel pins was theoretically predicted by a thermochemical computer code SOLGASMIX-PV. The absolute amounts of fission products generated in the fuel was calculated by ORIGEN-2 code, and the radial distributions of temperature and oxygen potential were calculated by taking the neutron depression and oxygen redistribution in the fuel into account. A fuel pellet was radially divided into 51 sections and chemical forms of FPs were calculated in each section. In addition, the effects of linear heat rating (LHR) and average O/U ratio on radial distribution of chemical form were evaluated. It was found that approximately 13 mole% of the total amount of Cs compounds exists as CsI and virtually remaining fraction as Cs{sub 2}MoO{sub 4} under the operation condition of LHR below 400 W/cm. On the other hand, when LHR is beyond 400 W/cm under the transient operation condition, its distribution did not change so much from the one under normal operation condition. (author)

  6. Core burnup calculation and accidents analyses of a pressurized water reactor partially loaded with rock-like oxide fuel

    Science.gov (United States)

    Akie, H.; Sugo, Y.; Okawa, R.

    2003-06-01

    A rock-like oxide (ROX) fuel - light water reactor (LWR) burning system has been studied for efficient plutonium transmutation. For the improvement of small negative reactivity coefficients and severe transient behaviors of ROX fueled LWRs, a partial loading core of ROX fuel assemblies with conventional UO 2 assemblies was considered. As a result, although the reactivity coefficients could be improved, the power peaking tends to be large in this heterogeneous core configuration. The reactivity initiated accident (RIA) and loss of coolant accident (LOCA) behaviors were not sufficiently improved. In order to reduce the power peaking, the fuel composition and the assembly design of the ROX fuel were modified. Firstly, erbium burnable poison was added as Er 2O 3 in the ROX fuel to reduce the burnup reactivity swing. Then pin-by-pin Pu enrichment and Er content distributions within the ROX fuel assembly were considered. In addition, the Er content distribution was also considered in the axial direction of the ROX fuel pin. With these modifications, a power peaking factor even lower than the one in a conventional UO 2 fueled core can be obtained. The RIA and LOCA analyses of the modified core have also shown the comparable transient behaviors of ROX partial loading core to those of the UO 2 core.

  7. Separation of metallic residues from the dissolution of a high-burnup BWR fuel using nitrogen trifluoride

    Energy Technology Data Exchange (ETDEWEB)

    McNamara, Bruce K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Buck, Edgar C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Soderquist, Chuck Z. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Smith, Frances N. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Mausolf, Edward J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Scheele, Randall D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2014-03-23

    Nitrogen trifluoride (NF3) was used to fluorinate the metallic residue from the dissolution of a high burnup, boiling water reactor fuel (~70 MWd/kgU). The metallic residue included the noble metal phase (containing ruthenium, rhodium, palladium, technetium, and molybdenum), and smaller amounts of zirconium, selenium, tellurium, and silver. Exposing the noble metal phase to 10% NF3 in argon between 400 and 550°C, removed molybdenum and technetium near 400°C as their volatile fluorides, and ruthenium near 500C as its volatile fluoride. The events were thermally and temporally distinct and the conditions specified are a recipe to separate these transition metals from each other and from the noble metal phase nonvolatile residue. Depletion of the volatile fluorides resulted in substantial exothermicity. Thermal excursion behavior was recorded under non-adiabatic, isothermal conditions that typically minimize heat release. Physical characterization of the metallic noble phase and its thermal behavior are consistent with high kinetic velocity reactions encouraged by the nanoparticulate phase or perhaps catalytic influences of the mixed platinum metals with nearly pure phase structure. Post-fluorination, only two phases were present in the residual nonvolatile fraction. These were identified as a nano-crystalline, metallic palladium cubic phase and a hexagonal rhodium trifluoride (RhF3) phase. The two phases were distinct as the sub-µm crystallites of metallic palladium were in contrast to the RhF3 phase, which grew from the parent nano-crystalline noble-metal phase during fluorination, to acicular crystals exceeding 20-µm in length.

  8. Development and validation of burnup dependent computational schemes for the analysis of assemblies with advanced lattice codes

    Science.gov (United States)

    Ramamoorthy, Karthikeyan

    The main aim of this research is the development and validation of computational schemes for advanced lattice codes. The advanced lattice code which forms the primary part of this research is "DRAGON Version4". The code has unique features like self shielding calculation with capabilities to represent distributed and mutual resonance shielding effects, leakage models with space-dependent isotropic or anisotropic streaming effect, availability of the method of characteristics (MOC), burnup calculation with reaction-detailed energy production etc. Qualified reactor physics codes are essential for the study of all existing and envisaged designs of nuclear reactors. Any new design would require a thorough analysis of all the safety parameters and burnup dependent behaviour. Any reactor physics calculation requires the estimation of neutron fluxes in various regions of the problem domain. The calculation goes through several levels before the desired solution is obtained. Each level of the lattice calculation has its own significance and any compromise at any step will lead to poor final result. The various levels include choice of nuclear data library and energy group boundaries into which the multigroup library is cast; self shielding of nuclear data depending on the heterogeneous geometry and composition; tracking of geometry, keeping error in volume and surface to an acceptable minimum; generation of regionwise and groupwise collision probabilities or MOC-related information and their subsequent normalization thereof, solution of transport equation using the previously generated groupwise information and obtaining the fluxes and reaction rates in various regions of the lattice; depletion of fuel and of other materials based on normalization with constant power or constant flux. Of the above mentioned levels, the present research will mainly focus on two aspects, namely self shielding and depletion. The behaviour of the system is determined by composition of resonant

  9. An extended version of the SERPENT-2 code to investigate fuel burn-up and core material evolution of the Molten Salt Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aufiero, M.; Cammi, A.; Fiorina, C. [Politecnico di Milano, Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via Ponzio, 34/3, I-20133 Milano (Italy); Leppänen, J. [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT (Finland); Luzzi, L., E-mail: lelio.luzzi@polimi.it [Politecnico di Milano, Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via Ponzio, 34/3, I-20133 Milano (Italy); Ricotti, M.E. [Politecnico di Milano, Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via Ponzio, 34/3, I-20133 Milano (Italy)

    2013-10-15

    In this work, the Monte Carlo burn-up code SERPENT-2 has been extended and employed to study the material isotopic evolution of the Molten Salt Fast Reactor (MSFR). This promising GEN-IV nuclear reactor concept features peculiar characteristics such as the on-line fuel reprocessing, which prevents the use of commonly available burn-up codes. Besides, the presence of circulating nuclear fuel and radioactive streams from the core to the reprocessing plant requires a precise knowledge of the fuel isotopic composition during the plant operation. The developed extension of SERPENT-2 directly takes into account the effects of on-line fuel reprocessing on burn-up calculations and features a reactivity control algorithm. It is here assessed against a dedicated version of the deterministic ERANOS-based EQL3D procedure (PSI-Switzerland) and adopted to analyze the MSFR fuel salt isotopic evolution. Particular attention is devoted to study the effects of reprocessing time constants and efficiencies on the conversion ratio and the molar concentration of elements relevant for solubility issues (e.g., trivalent actinides and lanthanides). Quantities of interest for fuel handling and safety issues are investigated, including decay heat and activities of hazardous isotopes (neutron and high energy gamma emitters) in the core and in the reprocessing stream. The radiotoxicity generation is also analyzed for the MSFR nominal conditions. The production of helium and the depletion in tungsten content due to nuclear reactions are calculated for the nickel-based alloy selected as reactor structural material of the MSFR. These preliminary evaluations can be helpful in studying the radiation damage of both the primary salt container and the axial reflectors.

  10. An extended version of the SERPENT-2 code to investigate fuel burn-up and core material evolution of the Molten Salt Fast Reactor

    Science.gov (United States)

    Aufiero, M.; Cammi, A.; Fiorina, C.; Leppänen, J.; Luzzi, L.; Ricotti, M. E.

    2013-10-01

    In this work, the Monte Carlo burn-up code SERPENT-2 has been extended and employed to study the material isotopic evolution of the Molten Salt Fast Reactor (MSFR). This promising GEN-IV nuclear reactor concept features peculiar characteristics such as the on-line fuel reprocessing, which prevents the use of commonly available burn-up codes. Besides, the presence of circulating nuclear fuel and radioactive streams from the core to the reprocessing plant requires a precise knowledge of the fuel isotopic composition during the plant operation. The developed extension of SERPENT-2 directly takes into account the effects of on-line fuel reprocessing on burn-up calculations and features a reactivity control algorithm. It is here assessed against a dedicated version of the deterministic ERANOS-based EQL3D procedure (PSI-Switzerland) and adopted to analyze the MSFR fuel salt isotopic evolution. Particular attention is devoted to study the effects of reprocessing time constants and efficiencies on the conversion ratio and the molar concentration of elements relevant for solubility issues (e.g., trivalent actinides and lanthanides). Quantities of interest for fuel handling and safety issues are investigated, including decay heat and activities of hazardous isotopes (neutron and high energy gamma emitters) in the core and in the reprocessing stream. The radiotoxicity generation is also analyzed for the MSFR nominal conditions. The production of helium and the depletion in tungsten content due to nuclear reactions are calculated for the nickel-based alloy selected as reactor structural material of the MSFR. These preliminary evaluations can be helpful in studying the radiation damage of both the primary salt container and the axial reflectors.

  11. Modeling of PWR fuel at extended burnup; Estudo de modelos para o comportamento a altas queimas de varetas combustiveis de reatores a agua leve pressurizada

    Energy Technology Data Exchange (ETDEWEB)

    Dias, Raphael Mejias

    2016-11-01

    This work studies the modifications implemented over successive versions in the empirical models of the computer program FRAPCON used to simulate the steady state irradiation performance of Pressurized Water Reactor (PWR) fuel rods under high burnup condition. In the study, the empirical models present in FRAPCON official documentation were analyzed. A literature study was conducted on the effects of high burnup in nuclear fuels and to improve the understanding of the models used by FRAPCON program in these conditions. A steady state fuel performance analysis was conducted for a typical PWR fuel rod using FRAPCON program versions 3.3, 3.4, and 3.5. The results presented by the different versions of the program were compared in order to verify the impact of model changes in the output parameters of the program. It was observed that the changes brought significant differences in the results of the fuel rod thermal and mechanical parameters, especially when they evolved from FRAPCON-3.3 version to FRAPCON-3.5 version. Lower temperatures, lower cladding stress and strain, lower cladding oxide layer thickness were obtained in the fuel rod analyzed with the FRAPCON-3.5 version. (author)

  12. Neutronic Study of Burnup, Radiotoxicity, Decay Heat and Basic Safety Parameters of Mono-Recycling of Americium in French Pressurised Water Reactors

    Directory of Open Access Journals (Sweden)

    Robert Bright Mawuko Sogbadji

    2017-03-01

    Full Text Available The reprocessing of actinides with long half-life has been non-existent except for plutonium (Pu. This work looks at reducing the actinides inventory nuclear fuel waste meant for permanent disposal. The uranium oxide fuel (UOX assembly, as in the open cycle system, was designed to reach a burnup of 46GWd/T and 68GWd/T using the MURE code. The MURE code is based on the coupling of a static Monte Carlo code and the calculation of the evolution of the fuel during irradiation and cooling periods. The MURE code has been used to address two different questions concerning the mono-recycling of americium (Am in present French pressurised water reactors (PWR. These are reduction of americium in the clear fuel cycle and the safe quantity of americium that can be introduced into mixed oxide (MOX as fuel. The spent UOX was reprocessed to fabricate MOX assemblies, by the extraction of plutonium and addition of depleted uranium to reach burnups of 46GWd/T and 68GWd/T, taking into account various cooling times of the spent UOX assembly in the repository. The effect of cooling time on burnup and radiotoxicity was then ascertained. After 30 years of cooling in the repository, the spent UOX fuel required a higher concentration of Pu to be reprocessed into MOX fuel due to the decay of Pu-241. Americium, with a mean half-life of 432 years, has a high radiotoxicity level, high mid-term residual heat and is a precursor for other long-lived isotopes. An innovative strategy would be to reprocess not only the plutonium from the UOX spent fuel but also the americium isotopes, which presently dominate the radiotoxicity of waste. The mono-recycling of Am is not a definitive solution because the once-through MOX cycle transmutation of Am in a PWR is not enough to destroy all americium. The main objective is to propose a ‘waiting strategy’ for both Am and Pu in the spent fuel so that they can be made available for further transmutation strategies. The MOX and

  13. Burn-up Function of Fuel Management Code for Aqueous Homogeneous Reactors and Its Validation%溶液堆物理计算程序FMCAHR燃耗功能及其验证

    Institute of Scientific and Technical Information of China (English)

    汪量子; 姚栋; 王侃

    2011-01-01

    介绍了FMCAHR程序的燃耗计算模型及流程,并使用燃耗基准题和DRAGON程序对燃耗计算结果进行验证.验证结果表明,FMCAHR燃耗计算功能的准确性较高,适用于溶液堆的燃耗计算分析.%Fuel Management Code for Aqueous Homogeneous Reactors(FMCAHR)is developed based on the Monte Carlo transport method,to analyze the physics characteristics of aqueous homogeneous reactors. FMCAHR has the ability of doing resonance treatment,searching for critical rod heights,thermal hydraulic parameters calculation,radiolytic-gas bubbles' calculation and burn-up calculation. This paper introduces the theory model and scheme of its bum-up function,and then compares its calculation results with benchmarks and with DRAGON'S burn-up results,which confirms its burn-up computing precision and its applicability in the burn-up calculation and analysis for aqueous solution reactors.

  14. Development of Burnup Calculation Function in Reactor Monte Carlo Code RMC%堆用蒙卡程序燃耗计算功能开发

    Institute of Scientific and Technical Information of China (English)

    佘顶; 王侃; 余纲林

    2012-01-01

    This paper presents the burnup calculation capability of RMC, which is a new Monte Carlo (MC) neutron transport code developed by Reactor Engineering Analysis Laboratory (REAL) in Tsinghua university of China. Unlike most of existing MC depletion codes which explicitly couple the depletion module, RMC incorporates ORIGEN 2.1 in an implicit way. Different burn step strategies, including the middle-of-step approximation and the predictor-corrector method, are adopted by RMC to assure the accuracy under large burnup step size. RMC employs a spectrum-based method of tallying one-group cross section, which can considerably saves computational time with negligible accuracy loss. According to the validation results of benchmarks and examples, it is proved that the burnup function of RMC performs quite well in accuracy and efficiency.%堆用蒙卡程序(RMC)是由清华大学工程物理系REAL实验室自主开发的用于反应堆物理分析的中子输运蒙卡程序,本文主要介绍其燃耗计算功能的开发与验证.RMC的燃耗计算功能具有的特点:内部耦合ORIGEN,相比于外耦合方式,更加灵活和高效;使用基于能谱的单群截面统计方法,可在保证精度的前提下,显著提高计算效率;采取预估修正和中点近似等多种燃耗步策略,减小大燃耗步长时的计算误差.通过计算压水堆栅元、沸水堆组件、快堆等一系列基准题和算例,验证了RMC燃耗计算的正确性和速度优势.

  15. Preliminary Content Evaluation of the North Anna High Burn-Up Sister Fuel Rod Segments for Transportation in the 10-160B and NAC-LWT

    Energy Technology Data Exchange (ETDEWEB)

    Ketusky, E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-08-09

    The U.S. Department of Energy’s (DOE’s) Used Fuel Disposition Campaign (UFDC) Program has transported high-burnup nuclear sister fuel rods from a commercial nuclear power plant for purposes of evaluation and testing. The evaluation and testing of high-burnup used nuclear fuel is integral to DOE initiatives to collect information useful in determining the integrity of fuel cladding for future safe transportation of the fuel, and for determining the effects of aging, on the integrity of UNF subjected to extended storage and subsequent transportation. The UFDC Program, in collaboration with the U.S. Nuclear Regulatory Commission and the commercial nuclear industry, has obtained individual used nuclear fuel rods for testing. The rods have been received at Oak Ridge National Laboratory (ORNL) for both separate effects testing (SET) and small-scale testing (SST). To meet the research objectives, testing on multiple 6 inch fuel rod pins cut from the rods at ORNL will be performed at Pacific Northwest National Laboratory (PNNL). Up to 10 rod equivalents will be shipped. Options were evaluated for multiple shipments using the 10-160B (based on 4.5 rod equivalents) and a single shipment using the NAC-LWT. Based on the original INL/Virginia Power transfer agreement, the rods are assumed to 152 inches in length with a 0.374-inch diameter. This report provides a preliminary content evaluation for use of the 10-160B and NAC-LWT for transporting those fuel rod pins from ORNL to PNNL. This report documents the acceptability of using these packagings to transport the fuel segments from ORNL to PNNL based on the following evaluations: enrichment, A2 evaluation, Pu-239 FGE evaluation, heat load, shielding (both gamma and neutron), and content weight/structural evaluation.

  16. The effect of dissolved hydrogen on the dissolution of {sup 233}U doped UO{sub 2}(s) high burn-up spent fuel and MOX fuel

    Energy Technology Data Exchange (ETDEWEB)

    Carbol, P. [Inst. for Transuranium Elements, Karlsruhe (Germany); Spahiu, K. (ed.) [and others

    2005-03-01

    In this report the results of the experimental work carried out in a large EU-research project (SFS, 2001-2004) on spent fuel stability in the presence of various amounts of near field hydrogen are presented. Studies of the dissolution of {sup 233}U doped UO{sub 2}(s) simulating 'old' spent fuel were carried out as static leaching tests, autoclave tests with various hydrogen concentrations and electrochemical tests. The results of the leaching behaviour of a high burn-up spent fuel pellet in 5 M NaCl solutions in the presence of 3.2 bar H{sub 2} pressure and of MOX fuel in dilute synthetic groundwater under 53 bar H{sub 2} pressure are also presented. In all the experimental studies carried out in this project, a considerable effect of hydrogen in the dissolution rates of radioactive materials was observed. The experimental results obtained in this project with a-doped UO{sub 2}, high burn-up spent fuel and MOX fuel together with literature data give a reliable background to use fractional alteration/dissolution rates for spent fuel of the order of 10{sup -6}/yr - 10{sup -8}/yr with a recommended value of 4x10{sup -7}/yr for dissolved hydrogen concentrations above 10{sup -3} M and Fe(II) concentrations typical for European repository concepts. Finally, based on a review of the experimental data and available literature data, potential mechanisms of the hydrogen effect are also discussed. The work reported in this document was performed as part of the Project SFS of the European Commission 5th Framework Programme under contract no FIKW-CT-2001-20192 SFS. It represents the deliverable D10 of the experimental work package 'Key experiments using a-doped UO{sub 2} and real spent fuel', coordinated by SKB with the participation of ITU, FZK-INE, ENRESA, CIEMAT, ARMINES-SUBATECH and SKB.

  17. Calculation of Effect of Burnup History on Spent Fuel Reactivity Based on CASMO5%基于CASMO5的燃耗历史对乏燃料反应性的影响计算

    Institute of Scientific and Technical Information of China (English)

    李晓波; 夏兆东; 朱庆福

    2015-01-01

    Based on the burnup credit of actinides+fission products (APU‐2) which are usually considered in spent fuel package ,the effect of power density and operating histo‐ry on k∞ was studied .All the burnup calculations are based on the two‐dimensional fuel assembly burnup program CASMO5 . The results show that taking the core average power density of specified power plus a bounding margin of 0.002 3 to k∞ ,and taking the operating history of specified power without shutdown during cycle and between cycles plus a bounding margin of 0.004 5 to k∞ can meet the bounding principle of burnup credit .%基于乏燃料贮存领域常用的锕系加裂变产物(APU‐2)级燃耗信任制,应用二维组件燃耗计算程序CASMO5,计算了燃耗过程中功率密度和运行历史对乏燃料 k∞的影响。结果表明:燃耗计算中,选择堆芯额定功率对应的平均功率密度,同时k∞附加0.0023的包络裕度,运行历史选择循环内及循环间无停堆额定功率运行,同时 k∞附加0.0045的包络裕度,可满足燃耗信任制中包络性原则。

  18. Development of Burnup Calculation Code for Pebble-bed High Temperature Reactor at Equilibrium State%球床高温堆平衡态燃耗计算程序的开发

    Institute of Scientific and Technical Information of China (English)

    朱贵凤; 邹杨; 李明海; 严睿; 彭红花; 徐洪杰

    2015-01-01

    The burnup calculation code PBRE coupling MCNP5 and ORIGEN2 was developed for pebble‐bed high temperature reactor at equilibrium state ,and it can be used to analyze the neutronic performance of equilibrium core .The iteration method was optimized in order to save Monte Carlo calculation time ,and the convergence can be reached in 10 iterative steps .The average discharged burnup for HTR‐10 is consistent with literature ,and it indicates that the PBRE is suitable to analyze the burnup for pebble‐bed reactor at equilibrium state .%基于MCNP5和ORIGEN2耦合方法,开发了平衡态下球床高温堆的燃耗计算程序PBRE ,用于堆的性能价值分析。为节省蒙特卡罗计算时间,对迭代收敛的方法进行优化,使之可在10个迭代步内收敛。使用PBRE对清华大学H T R‐10进行建模计算,得到的平均卸料燃耗深度与文献报道值一致,表明PBRE程序适用于球床堆平衡态的燃耗分析。

  19. Chemical states of fission products in irradiated (U{sub 0.3}Pu{sub 0.7})C{sub 1+x} fuel at high burn-ups

    Energy Technology Data Exchange (ETDEWEB)

    Agarwal, Renu [Fuel Chemistry Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)]. E-mail: arenu@barc.gov.in; Venugopal, V. [Fuel Chemistry Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2006-12-01

    The chemical states of fission products have been theoretically determined for the irradiated carbide fuel of Fast Breeder Test Reactor (FBTR) at Kalpakkam, India, at different burn-ups. The SOLGASMIX-PV computer code was used to determine the equilibrium chemical composition of the fuel. The system was assumed to be composed of a gaseous phase at one atmosphere pressure, and various solid phases. The distribution of elements in these phases and their chemical states at different temperatures were calculated as a function of burn-up. The FBTR fuel (U{sub 0.3}Pu{sub 0.7})C{sub 1+x}, was loaded with C/M values in the range, 1.03-1.06. The present calculations indicated that even for the lowest starting C/M of 1.03 in the FBTR fuel, the liquid metal phase of (U, Pu), should not appear at a burn-up as high as 150 GWd/t.

  20. Electron probe microanalysis of a METAPHIX UPuZr metallic alloy fuel irradiated to 7.0 at.% burn-up

    Science.gov (United States)

    Brémier, S.; Inagaki, K.; Capriotti, L.; Poeml, P.; Ogata, T.; Ohta, H.; Rondinella, V. V.

    2016-11-01

    The METAPHIX project is a collaboration between CRIEPI and JRC-ITU investigating safety and performance of a closed fuel cycle option based on fast reactor metal alloy fuels containing Minor Actinides (MA). The aim of the project is to investigate the behaviour of this type of fuel and demonstrate the transmutation of MA under irradiation. A UPuZr metallic fuel sample irradiated to a burn-up of 7 at.% was examined by electron probe microanalysis. The fuel sample was extensively characterised qualitatively and quantitatively using elemental X-ray imaging and point analysis techniques. The analyses reveal a significant redistribution of the fuel components along the fuel radius highlighting a nearly complete depletion of Zr in the central part of the fuel. Numerous rare earth and fission products secondary phases are present in various compositions. Fuel cladding chemical interaction was observed with creation of a number of intermediary layers affecting a cladding depth of 15-20 μm and migration of cladding elements to the fuel.

  1. Heterogeneous UO2 fuel irradiated up to a high burn-up: Investigation of the HBS and of fission product releases

    Science.gov (United States)

    Noirot, J.; Lamontagne, J.; Nakae, N.; Kitagawa, T.; Kosaka, Y.; Tverberg, T.

    2013-11-01

    A UO2 fuel with a heterogeneous distribution of 235U was irradiated up to a high burn-up in the Halden Boiling Water Reactor (HBWR). The last 100 days of irradiation were performed with an increased level of linear power. The effect of the heterogeneous fissile isotope distribution on the formation of the HBS was studied free of the possible influence of Pu which exists in heterogeneous MOX fuels. The HBS formed in 235U-rich agglomerates and its main characteristics were very similar to those of the HBS formed in Pu-rich agglomerates of heterogeneous MOX fuels. The maximum local contents of Nd and Xe before HBS formation were studied in this fuel. In addition to a Pu effect that promotes the HBS phenomenon, comparison with previous results for heterogeneous MOX fuels showed that the local fission product concentration was not the only parameter that has to be taken into consideration. It appears that the local actinide depletion by fission and/or the energy locally deposited through electronic interactions in the fission fragment recoils also have an effect on the HBS formation threshold. Moreover, a major release of fission gases from the peripheral 235U-rich agglomerates of HBS bubbles and a Cs radial movement are also evidenced in this heterogeneous UO2. Cs deposits on the peripheral grain boundaries, including the HBS grain boundaries, are considered to reveal the release paths. SUP>235U-rich agglomerates, SUP>235U-poor areas, an intermediate phase with intermediate 235U concentrations. Short fuel rods were fabricated with these pellets. The main characteristics of these fuel rods are shown in Table 1.These rods were irradiated to high burn-ups in the IFA-609/626 of the HBWR and then one was irradiated in the IFA-702 for 100 days. Fig. 2 shows the irradiation history of this fuel. The final average burn-up of the rod was 69 GWd/tU. Due to the flux differences along the rod, however, the average burn-up of the cross section examined was 63 GWd/tU. This fuel

  2. Postirradiation examinations of fuel pins from the GCFR F-1 series of mixed-oxide fuel pins at 5. 5 at. % burnup

    Energy Technology Data Exchange (ETDEWEB)

    Strain, R V; Johnson, C E

    1978-05-01

    Postirradiation examinations were performed on five fuel pins from the Gas-Cooled Fast-Breeder Reactor F-1 experiment irradiated in EBR-II to a peak burnup of approximately 5.5 at. %. These encapsulated fuel pins were irradiated at peak-power linear ratings from approximately 13 to 15 kW/ft and peak cladding inside diameter temperatures from approximately 625 to 760/sup 0/C. The maximum diametral change that occurred during irradiation was 0.2% ..delta..D/D/sub 0/. The maximum fuel-cladding chemical interaction depth was 2.6 mils in fuel pin G-1 and 1 mil or less in the other three pins examined destructively. Significant migration of the volatile fission products occurred axially to the fuel-blanket interfaces. Teh postirradiation examination data indicate that fuel melted at the inner surface of the annular fuel pellets in the two highest power rating fuel pins, but little axial movement of fuel occurred.

  3. On the oxidation state of UO 2 nuclear fuel at a burn-up of around 100 MWd/kgHM

    Science.gov (United States)

    Walker, C. T.; Rondinella, V. V.; Papaioannou, D.; Winckel, S. Van; Goll, W.; Manzel, R.

    2005-10-01

    Results for the radial distribution of the oxygen potential and stoichiometry of a PWR fuel with an average pellet burn-up of 102 MWd/kgHM are presented. The local Δ G bar (O2) of the fuel was measured using a miniature solid state galvanic cell, the local O/U ratio was calculated from the lattice parameter measured by micro-X-ray diffraction and the local O/M ratio was derived from the fuel composition determined by ICP-MS. During irradiation the O/U ratio of the fuel decreased from 2.005 to 1.991 ± 0.008. The average fuel O/M ratio was 1.973 compared with the stoichiometric value of 1.949. The amount of free oxygen in the fuel, represented by the difference between these two quantities, increased from the centre to periphery of the pellet. Similarly, the Δ G bar (O2) of the fuel increased from -370 kJ mol-1 at r/r0 = 0.1 to -293 kJ mol-1 at r/r0 = 0.975. Thus, the Δ G bar (O2) of the fuel had not been buffered by the oxidation of fission product Mo. About one-quarter of the free oxygen accumulated during the irradiation had been gettered by the Zircaloy cladding.

  4. Determination of plutonium content in high burnup pressurized water reactor fuel samples and its use for isotope correlations for isotopic composition of plutonium.

    Science.gov (United States)

    Joe, Kihsoo; Jeon, Young-Shin; Han, Sun-Ho; Lee, Chang-Heon; Ha, Yeong-Keong; Song, Kyuseok

    2012-06-01

    The content of plutonium isotopes in high burnup pressurized water reactor fuel samples was examined using both alpha spectrometry and mass spectrometry after anion exchange separation. The measured values were compared with results calculated by the ORIGEN-2 code. On average, the ratios (m/c) of the measured values (m) over the calculated values (c) were 1.22±0.16 for (238)Pu, 1.02±0.14 for (239)Pu, 1.08±0.06 for (240)Pu, 1.06±0.16 for (241)Pu, and 1.13±0.08 for (242)Pu. Using the Pu data obtained in this work, correlations were derived between the alpha activity ratios of (238)Pu/((239)Pu+(240)Pu), the alpha specific activities of Pu, and the atom % abundances of the Pu isotopes. Using these correlations, the atom % abundances of the plutonium isotopes in the target samples were calculated. These calculated results agreed within a range from 2 to 8% of the experimentally derived values according to the isotopes of plutonium.

  5. Bias estimates used in lieu of validation of fission products and minor actinides in MCNP Keff calculations for PWR burnup credit casks

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, Don [ORNL; Marshall, William BJ J [ORNL; Wagner, John C [ORNL; Bowen, Douglas G [ORNL

    2015-09-01

    The U.S. Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation recently issued Interim Staff Guidance (ISG) 8, Revision 3. This ISG provides guidance for burnup credit (BUC) analyses supporting transport and storage of PWR pressurized water reactor (PWR) fuel in casks. Revision 3 includes guidance for addressing validation of criticality (keff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MA). Based on previous work documented in NUREG/CR-7109, recommendation 4 of ISG-8, Rev. 3, includes a recommendation to use 1.5 or 3% of the FP&MA worth to conservatively cover the bias due to the specified FP&MAs. This bias is supplementary to the bias and bias uncertainty resulting from validation of keff calculations for the major actinides in SNF and does not address extension to actinides and fission products beyond those identified herein. The work described in this report involves comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII based nuclear data and supports use of the 1.5% FP&MA worth bias when either SCALE or MCNP codes are used for criticality calculations, provided the other conditions of the recommendation 4 are met. The method used in this report may also be applied to demonstrate the applicability of the 1.5% FP&MA worth bias to other codes using ENDF/B V, VI or VII based nuclear data. The method involves use of the applicant s computational method to generate FP&MA worths for a reference SNF cask model using specified spent fuel compositions. The applicant s FP&MA worths are then compared to reference values provided in this report. The applicants FP&MA worths should not exceed the reference results by more than 1.5% of the reference FP&MA worths.

  6. 多群蒙卡输运与点燃耗耦合程序系统TRITON基准验证%Benchmark Verification of Multi-group Monte Carlo Transport and Point-Burnup Codes Coupling System TRITON

    Institute of Scientific and Technical Information of China (English)

    武祥; 若夕子; 于涛; 谢金森; 陈昊威

    2014-01-01

    TRITON couples multi group Monte Carlo Transport code KENO V. a and point-burnup code ORIGEN-S. It features adaptability on complex geometries,flexible processing ability on cross section and rapid calculating speed. Based on the thorium-based fuel cell benchmark of Idaho National Laboratory ( INL) ,the verification on TRITON burnup calcu-lation was performed,which showed good coincidence with the result of MOCUP code by INL. Furthermore, the results of burnup isotopes selection schemes in TRITON showed that,for thorium based fuel,only important nuclides on Th-U cycle was included,correct results can be obtained by TRITON. Conclusions in the present paper will support further applications of TRITON.%TRITON程序系统耦合了多群蒙特卡罗输运程序KENO V. a与点燃耗程序ORIGEN-S,具有几何适应性强、截面处理能力灵活、计算速度快等显著特点.本文基于爱达荷国家实验室( INL)钍基燃料元件燃耗基准题,开展了TRITON程序燃耗功能的验证,结果与INL采用MOCUP程序给出的结果吻合很好.同时,燃耗核素选取对TRITON计算结果的影响分析表明对于钍基燃料,只有在考虑Th-U循环重要核素的前提下,TRITON才能给出正确结果.上述结论为TRITON程序的应用奠定了基础.

  7. An investigation into the origen of the interference generated during the measurement of the reactivity in a high burn-up reactor core%高燃耗堆芯反应性测量的干扰源研究

    Institute of Scientific and Technical Information of China (English)

    陈雄月; 吕大军; 裘希春; 韩承慈; 夏应军; 邓朝平; 张仲元

    2012-01-01

    回顾了1980年9月实验前,在反应堆噪声分析领域的技术发展概况.展示了在实验动力堆燃耗末、卸料前,用双探测器互相关频谱分析法(CCFS)测得的一组数据;和经过离线去本底拟合计算后,获得的动力学参数测量结果:αc=(144.57±2.09)s-1.介绍了数据获取过程中出现的异常情况;离线处理的方法;本底谱选定;拟合计算程序;计算结果和结论.还简要介绍了干扰源的来源及其强度计算概况.数据处理结果证明:在长期燃耗后的堆芯上应用噪声分析法,除了要克服大γ场的干扰外,还要严格消除本底中子场产生的不相关噪声干扰.%After a general review for the technical development before 1980's in the area of nuclear reactor noise analysis,a reactor dynamic parameter,ac = (144. 57 + 2. 09)s~1 , obtained through off-line background processing, is shown. The processed data is measured through double- detector cross correlation frequency spectral analysis (CCFS) for the experimental nuclear power reactor at the burn-up end in sept. 1980. This paper also presents the abnormal situations for data acquisition, the off-line data processing method,the background spectra selection for data processing and the program for the least-squares fit calculation. Here also explains how neutron background is generated and how its strength is calculated. This verifies the fact that after a long-term burn-up run, large y field must be suppressed and also more attention must be paid to the uncorrelated neutron noise from the fuel burn-up.

  8. Burnup extension and evolution in the fuel management of EDF's nuclear power plants; Accroissement des taux de combustion et impact des evolutions de gestion sur l'exploitation des reacteurs du parc EDF

    Energy Technology Data Exchange (ETDEWEB)

    Provost, J.L.; Thibault, X. [Electricite de France (EDF/DPN), 93 - Saint-Denis (France); Debes, M. [Electricite de France (EDF/DCN), 92 - Clamart (France); Kaplan, P. [Cogema, 78 - Velizy Villacoublay (France)

    2004-07-01

    Today the use of enhanced nuclear fuels that can sustain higher burnups has allowed a better optimization of the fuel management in nuclear power plants. The optimization for the near future is based on 3 aims: -) a better competitiveness of nuclear energy, longer campaigns mean a higher availability and less refueling so it has a direct impact on costs, -) a better flexibility to meet energy demand: a modulation of cycle lengths by more or less 2 months is possible by introducing or withdrawing 8 assemblies in the refueling load, this modulation will allow an optimization of the scheduling of the refueling shutdowns with respect to the seasonal energy demand peaks, -) a reduced volume of spent fuels (but with a higher level of radioactivity). (A.C.)

  9. Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    Energy Technology Data Exchange (ETDEWEB)

    Favalli, A., E-mail: afavalli@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM (United States); Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Schwalbach, P. [European Commission, DG Energy, Euratom Safeguards Luxemburg, Luxemburg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Tobin, S.J.; Trellue, H. [Los Alamos National Laboratory, Los Alamos, NM (United States); Vaccaro, S. [European Commission, DG Energy, Euratom Safeguards Luxemburg, Luxemburg (Luxembourg)

    2016-06-01

    The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute {sup 137}Cs count rate and the {sup 154}Eu/{sup 137}Cs, {sup 134}Cs/{sup 137}Cs, {sup 106}Ru/{sup 137}Cs, and {sup 144}Ce/{sup 137}Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

  10. Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    Science.gov (United States)

    Favalli, A.; Vo, D.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S. J.; Trellue, H.; Vaccaro, S.

    2016-06-01

    The purpose of the Next Generation Safeguards Initiative (NGSI)-Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/137Cs, 134Cs/137Cs, 106Ru/137Cs, and 144Ce/137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity's behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

  11. Research and verification of Monte Carlo burnup calculations based on Chebyshev rational approximation method%基于切比雪夫有理逼近方法的蒙特卡罗燃耗计算研究与验证

    Institute of Scientific and Technical Information of China (English)

    范文玎; 孙光耀; 张彬航; 陈锐; 郝丽娟

    2016-01-01

    燃耗计算在反应堆设计、分析研究中起着重要作用.相比于传统点燃耗算法,切比雪夫有理逼近方法(Chebyshev rational approximation method,CRAM)具有计算速度快、精度高的优点.基于超级蒙特卡罗核计算仿真软件系统SuperMC(Super Monte Carlo Simulation Program for Nuclear and Radiation Process),采用切比雪夫有理逼近方法和桶排序能量查找方法,进行了蒙特卡罗燃耗计算的初步研究与验证.通过燃料棒燃耗例题以及IAEA-ADS(International Atomic Energy Agency-Accelerator Driven Systems)国际基准题,初步验证了该燃耗计算方法的正确性,且IAEA-ADS基准题测试表明,与统一能量网格方法相比,桶排序能量查找方法在保证了计算效率的同时减少了内存开销.%Background:Burnup calculation is the key point of reactor design and analysis. It's significant to calculate the burnup situation and isotopic atom density accurately while a reactor is being designed.Purpose:Based on the Monte Carlo particle simulation code SuperMC (Super Monte Carlo Simulation Program for Nuclear and Radiation Process), this paper aimed to conduct preliminary study and verification on Monte Carlo burnup calculations. Methods:For the characteristics of accuracy, this paper adopted Chebyshev rational approximation method (CRAM) as the point-burnup algorithm. Moreover, instead of the union energy grids method, this paper adopted an energy searching method based on bucket sort algorithm, which reduced the memory overhead on the condition that the calculation efficiency is ensured.Results:By calculating the fuel rod burnup problem and the IAEA-ADS (International Atomic Energy Agency - Accelerator Driven Systems) international benchmark, the simulation results were basically consistent with Serpent and other counties' results, respectively. In addition, the bucket sort energy searching method reduced about 95% storage space compared with union energy grids method for IAEA

  12. 1D Burnup Calculation of Fusion-Fission Hybrid Energy Reactor%聚变-裂变混合能源堆一维计算模型燃耗分析

    Institute of Scientific and Technical Information of China (English)

    李茂生; 师学明; 伊炜伟

    2012-01-01

    Fusion-fission hybrid energy reactor is driven by Tokamak fusion source for energy production. Its subcritical zone uses the natural uranium as fuel and water as coolant. The neutron multiplication constant keff, energy multiplication factor M and tritium breeding ratio TBR of the ID hybrid energy reactor model were calculated by transport burnup code MCORGS. The neutron spectrum and nuclear density changing as a function of time show the characteristics of the hybrid energy reactors, which differs from the hybrid reactor for breed nuclear fuel and for spent fuel transmutation. The definition and results may be a reference to the other conceptual analysis.%聚变-裂变混合能源堆包括聚变中子源和以天然铀为燃料、水为冷却剂的次临界包层,主要目标是生产电力.利用输运燃耗耦合程序系统MCORGS计算了混合能源堆一维模型的燃耗,给出了中子有效增殖因数keff、能量放大倍数M、氚增殖比TBR等物理量随时间的变化.通过分析能谱和重要核素随燃耗时间的变化,说明混合能源堆与核燃料增殖、核废料嬗变混合堆的不同特点.本文给出的结果可作为混合堆中子输运、燃耗分析程序校验的参考数据,为混合堆概念研究提供了基础数据.

  13. Results of irradiation of (U0.55Pu0.45)N and (U0.4Pu0.6)N fuels in BOR-60 up to ˜12 at.% burn-up

    Science.gov (United States)

    Rogozkin, B. D.; Stepennova, N. M.; Fedorov, Yu. Ye.; Shishkov, M. G.; Kryukov, F. N.; Kuzmin, S. V.; Nikitin, O. N.; Belyaeva, A. V.; Zabudko, L. M.

    2013-09-01

    In the article presented are the results of post-irradiation tests of helium bonded fuel pins with mixed mononitride fuel (U0.55Pu0.45)N and (U0.4Pu0.6)N having 85% density irradiated in BOR-60 reactor. Achieved maximum burn-up was, respectively, equal to 9.4 and 12.1 at.% with max linear heat rates 41.9 and 54.5 kW/m. Maximum irradiation dose was 43 dpa. No damage of claddings made of ChS-68 steel (20% cold worked) was observed, and ductility margin existed. Maximum depth of cladding corrosion was within 15 μm. Swelling rates of (U0.4Pu0.6)N and (U0.55Pu0.45)N were, respectively, ˜1.1% and ˜0.68% per 1 at.%. Gas release rate did not exceed 19.3% and 19%. Pattern of porosity distribution in the fuel influenced fuel swelling and gas release rates. Plutonium and uranium are uniformly distributed in the fuel, local minimum values of their content being caused by pores and cracks in the pellets. The observable peaks in content distribution are probably connected with the local formation of isolated phases (e.g. Mo, Pd) while the minimum values refer to fuel pores and cracks. Xenon and cesium tend to migrate from the hot sections of fuel, and therefore their min content is observed in the central section of the fuel pellets. Phase composition of the fuel was determined with X-ray diffractometer. The X-ray patterns of metallographic specimens were obtained by the scanning method (the step was 0.02°, the step exposition was equal to 2 s). From the X-ray diffraction analysis data, it follows that the nitrides of both fuel types have the single-phase structure with an FCC lattice (see Table 6).

  14. Detailed Burnup Calculations for Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Leszczynski, F. [Centro Atomico Bariloche (CNEA), 8400 S. C. de Bariloche (Argentina)

    2011-07-01

    A general method (RRMCQ) has been developed by introducing a microscopic burn up scheme which uses the Monte Carlo calculated spatial power distribution of a research reactor core and a depletion code for burn up calculations, as a basis for solving nuclide material balance equations for each spatial region in which the system is divided. Continuous energy dependent cross-section libraries and full 3D geometry of the system is input for the calculations. The resulting predictions for the system at successive burn up time steps are thus based on a calculation route where both geometry and cross-sections are accurately represented, without geometry simplifications and with continuous energy data. The main advantage of this method over the classical deterministic methods currently used is that RRMCQ System is a direct 3D method without the limitations and errors introduced on the homogenization of geometry and condensation of energy of deterministic methods. The Monte Carlo and burn up codes adopted until now are the widely used MCNP5 and ORIGEN2 codes, but other codes can be used also. For using this method, there is a need of a well-known set of nuclear data for isotopes involved in burn up chains, including burnable poisons, fission products and actinides. For fixing the data to be included on this set, a study of the present status of nuclear data is performed, as part of the development of RRMCQ method. This study begins with a review of the available cross-section data of isotopes involved in burn up chains for research nuclear reactors. The main data needs for burn up calculations are neutron cross-sections, decay constants, branching ratios, fission energy and yields. The present work includes results of selected experimental benchmarks and conclusions about the sensitivity of different sets of cross-section data for burn up calculations, using some of the main available evaluated nuclear data files. Basically, the RRMCQ detailed burn up method includes four tasks for each burn up step: 1) Monte Carlo criticality calculation of the full system tallying spatial power distribution for each spatial region of interest. 2) Preparation of depletion code input and cross- section libraries from Monte Carlo calculation output and other auxiliary code, including normalized power density of each spatial zone with an auxiliary program. The 1 group cross section library needed for depletion calculations can be obtained with a cell code such as DRAGON4 vs. burn up. 3) Depletion calculations of isotope concentrations on the input burn up time-step. 4) Preparation of Monte Carlo calculation input with the new isotope concentrations output of depletion calculation with other auxiliary program. This sequence is implemented in an automatic way. On the first stages of RRMCQ development, a simplified version has been tested with a set of dependent numerical and experimental benchmarks using standard nuclear data libraries at lattice cell level. Then a full core model has been developed and it is to day used on RA6 reactor of Bariloche Atomic Centre. (author)

  15. Fission gas release from rock-like fuels, PuO{sub 2}-ZrO{sub 2}(Y){l_brace}or ThO{sub 2}{r_brace}-Al{sub 2}O{sub 3}-MgO at burn-up of 20 MWd/kg

    Energy Technology Data Exchange (ETDEWEB)

    Yanagisawa, Kazuaki; Ohmichi, Toshihiko; Kanazawa, Hiroyuki; Amano, Hidetoshi; Yamahara, Takeshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-11-01

    Two types of fuels, that is, a 20w/o PuO{sub 2} incorporated into ThO{sub 2}-Al{sub 2}O{sub 3}-MgAl{sub 2}O{sub 4} and a 23w/o PuO{sub 2} incorporated into ZrO{sub 2}(Y)-Al{sub 2}O{sub 3}-MgAl{sub 2}O{sub 4} were fabricated into a disk form (outer diameter by 3mm x thickness by 1mm) by JAERI`s self-established development technique. The two were provided for steady-state irradiation test at the Japan Research Reactor 3M (JRR-3M) up to the average burn-up of 20 MWd/kg (27 MWd/kg in peak) to understanding fuel behaviour. Post-irradiation examination (PIE) was carried out and the followings were revealed. (1) Despite of low irradiation temperature <1000degC, there occurred significant fission gas release (FGR) which could not be explained by diffusion mechanism. A possible explanation obtained from fuel microstructural study is direct escape of FP gas from fuel matrix to gas gap via open pores. (2) A nuclide Cs migrated from fuel matrix to plenum region. Its amount was roughly 20% of totally produced. This was partially due to slight temperature gradient across disk fuel and partly due to low Cs retentiveness in fabricated fuel matrix. A magnitude of Cs-137 migration was smaller in ThO{sub 2}-Al{sub 2}O{sub 3}-MgAl{sub 2}O{sub 4} disk fuel than that in ZrO{sub 2}(Y)-Al{sub 2}O{sub 3}-MgAl{sub 2}O{sub 4} disk fuel. (3) Gas bubble swelling rate estimated by total porosity increase was 14% for ZrO{sub 2}(Y)-Al{sub 2}O{sub 3}-MgAl{sub 2}O{sub 4} disk fuel and 11% for ThO{sub 2}-Al{sub 2}O{sub 3}-MgAl{sub 2}O{sub 4} disk fuel each per 10 MWd/kg. (4) Slight bonding between disk fuels and spacers (Nb-1w/oZr) and that between disk fuels and sheath (Nb-1w/oZr) occurred. The mechanism was attributed to mutual diffusion mainly between Al compound from fuel and Nb one from spacer or sheath. (author)

  16. Renault tackling new designs for fuel burnup and pollution cleanup

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    2000-02-01

    Over the past years, auto-makers have made great strides in gasoline and diesel motorization. Indeed, new cars burn up less fuel and release smaller amounts of polluting emissions. The Renault group has long been addressing an environmentally friendly policy, and accordingly manufacturing vehicles that burn up less fuel. Renault developments have spurred the most recent advances in this area. The group is now tackling new designs, such as the ADIVI or the Camless engine. The auto-maker is now working on substitute fuels such as natural gas, and on advanced post-treatment solutions. Renault has already engineered a Scenic 1.6 16V, low emissions demonstrator. (authors)

  17. The high burn-up structure in nuclear fuel

    Directory of Open Access Journals (Sweden)

    Vincenzo V. Rondinella

    2010-12-01

    Full Text Available During its operating life in the core of a nuclear reactor nuclear fuel is subjected to significant restructuring processes determined by neutron irradiation directly through nuclear reactions and indirectly through the thermo-mechanical conditions established as a consequence of such reactions. In today's light water reactors, starting after ∼4 years of operation the cylindrical UO2 fuel pellet undergoes a transformation that affects its outermost radial region. The discovery of a newly forming structure necessitated the answering of important questions concerning the safety of extended fuel operation and still today poses the fascinating scientific challenge of fully understanding the microstructural mechanisms responsible for its formation.

  18. Error Analysis and Experimental Research of HTR-10 Burn-up Measurement System Based on MCNP Modeling%基于MCNP模型10MW高温气冷实验堆燃耗测量误差分析与实验研究

    Institute of Scientific and Technical Information of China (English)

    马涛; 夏冰; 王军令; 陈晓明; 江二东

    2012-01-01

    10MW高温气冷实验堆(HTR-10)的燃耗测量系统通过测量燃料球内裂变产物137Cs发出的γ射线进而间接确定燃料球的燃耗,测量结果的准确性直接影响着反应堆的安全性和经济性.利用HTR-10现有的设备条件,设计并实施了提升器偏转实验,使燃料球逐步偏离正常测量位,改变球心与准直器轴线的相对位置,得到了偏离角度与计数率之间的对应关系,进而确定燃料球球心与准直器轴线的周向偏移量.通过MCNP程序建立HTR-10燃耗测量系统模型,模拟γ光子从燃料球发出,经过提升器、密封法兰、准直器直到被HPGe晶体探测器捕捉的全过程.利用MCNP模型可以模拟在不同径向偏离情况下的实验过程,通过与实验结果的对比,确定燃料球球心偏离准直器轴线的径向偏移量.%With the burn-up measurement system fur the 10MW high temperature gas-rooletl reuctor. ihe hum-up of the spherical fuel element is obtained indirectly hy measuring the gamma-ray from the fission product 137Cs. The accuracy of results will affect the security and the economy of the reactor. In this paper, on the basis uf the existing equipment conditions of HTR-10, the diversion experiment of the elevator is designed and implemented. By making the spherical fuel element gradually deviate tiie running measurement place, the relative position between the center of the spherical fuel element and the axis of the rnllimalor is changed, and the relationship between the angle of she diversion and the count rale is obtained anri is then used lo determine the offset in the circumferential direction between the center of the spherical fuel element and the collimator axis. The model of the HTR-10 bum-up measurement system, established by the MCNP program, can he used to simulate the gamma photon transporting process from the spherical fuel element, through the elevator, the sealed flange and the collimator, to the HPCe detector. The MCNP model eon be

  19. Analysis of fresh fuel critical experiments appropriate for burnup credit validation

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.; Bowman, S.M.

    1995-10-01

    The ANS/ANS-8.1 standard requires that calculational methods used in determining criticality safety limits for applications outside reactors be validated by comparison with appropriate critical experiments. This report provides a detailed description of 34 fresh fuel critical experiments and their analyses using the SCALE-4.2 code system and the 27-group ENDF/B-IV cross-section library. The 34 critical experiments were selected based on geometry, material, and neutron interaction characteristics that are applicable to a transportation cask loaded with pressurized-water-reactor spent fuel. These 34 experiments are a representative subset of a much larger data base of low-enriched uranium and mixed-oxide critical experiments. A statistical approach is described and used to obtain an estimate of the bias and uncertainty in the calculational methods and to predict a confidence limit for a calculated neutron multiplication factor. The SCALE-4.2 results for a superset of approximately 100 criticals are included in uncertainty analyses, but descriptions of the individual criticals are not included.

  20. Propagation of nuclear data uncertainties for ELECTRA burn-up calculations

    OpenAIRE

    Sjöstrand, Henrik; Alhassan, Erwin; Duan, Junfeng; Gustavsson, Cecilia; KONING Arjan J.; Pomp, Stephan; Rochman, Dimitri; Österlund, Michael

    2013-01-01

    The European Lead-Cooled Training Reactor (ELECTRA) has been proposed as a training reactor for fast systems within the Swedish nuclear program. It is a low-power fast reactor cooled by pure liquid lead. In this work, we propagate the uncertainties in Pu-239 transport data to uncertainties in the fuel inventory of ELECTRA during the reactor life using the Total Monte Carlo approach (TMC). Within the TENDL project the nuclear models input parameters were randomized within their uncertainties a...

  1. An attempt to reproduce high burn-up structure by ion irradiation of SIMFUEL

    Energy Technology Data Exchange (ETDEWEB)

    Baranov, V.G. [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Kashirskoye Shosse 31, Moscow 115409 (Russian Federation); Lunev, A.V., E-mail: AVLunev@mephi.ru [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Kashirskoye Shosse 31, Moscow 115409 (Russian Federation); Reutov, V.F. [Joint Institute for Nuclear Research (JINR), Flerov Laboratory of Nuclear Reactions (FLNR), 141980 Dubna, Moscow Region (Russian Federation); Tenishev, A.V.; Isaenkova, M.G.; Khlunov, A.V. [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Kashirskoye Shosse 31, Moscow 115409 (Russian Federation)

    2014-09-15

    Experiments in IC-100 and U-400 cyclotrons were conducted with SIMFUEL pellets (11.47 wt.% of fission products simulators) to reproduce some aspects of the long-term irradiation conditions in epithermal reactors. Pellets were irradiated with Xe{sup 16+}, Xe{sup 24+} and He{sup +} at energies ranging from 20 keV (He{sup +}) to 320 keV (Xe{sup 16+}) and 1–90 MeV (Xe{sup 24+}). Some samples were subsequently annealed to obtain larger grain sizes and to study defects recovery. The major microstructural changes consisted in grain sub-division observed on SEM and AFM images and change in composition registered by EPMA (pellets irradiated with 1–90 MeV Xe{sup 24+} ions at fluence of 5 × 10{sup 15} cm{sup −2}). Lattice distortion and increase in dislocation density is also noted according to X-ray data. At low energies and high fluences formation of bubbles (20 keV He{sup +} at 5.5 × 10{sup 17} cm{sup −2}) was observed. Grain sub-division exhibits full coverage of the grain body and preservation of former grain boundaries. The size of sub-grains depends on local dislocation density and changes from 200 nm to 400 nm along the irradiated surface. Beneath it the size ranges from 150 to 600 nm. Sub-grains are not observed in samples irradiated by low-energy ions even at high dislocation densities.

  2. Design and burn-up analyses of new type holder for silicon neutron transmutation doping.

    Science.gov (United States)

    Komeda, Masao; Arai, Masaji; Tamai, Kazuo; Kawasaki, Kozo

    2016-07-01

    We have developed a new silicon irradiation holder with a neutron filter to increase the irradiation efficiency. The neutron filter is made of an alloy of aluminum and B4C particles. We fabricated a new holder based on the results of design analyses. This filter has limited use in applications requiring prolonged use due to a decrease in the amount of (10)B in B4C particles. We investigated the influence of (10)B reduction on doping distribution in a silicon ingot by using the Monte Carlo Code MVP.

  3. Closed fuel cycle with increased fuel burn-up and economy applying of thorium resources

    Science.gov (United States)

    Kulikov, G. G.; Apse, V. A.

    2017-01-01

    The possible role of existing thorium reserves in the Russian Federation on engaging thorium in being currently closed (U-Pu)-fuel cycle of nuclear power of the country is considered. The application efficiency of thermonuclear neutron sources with thorium blanket for the economical use of existing thorium reserves is demonstrated. The aim of the work is to find solutions of such major tasks as the reduction of both front-end and back-end of nuclear fuel cycle and an enhancing its protection against the uncontrolled proliferation of fissile materials by means of the smallest changes in the fuel cycle. During implementation of the work we analyzed the results obtained earlier by the authors, brought new information on the number of thorium available in the Russian Federation and made further assessments. On the basis of proposal on the inclusion of hybrid reactors with Th-blanket into the future nuclear power for the production of light uranium fraction 232+233+234U, and 231Pa, we obtained the following results: 1. The fuel cycle will shift from fissile 235U to 233U which is more attractive for thermal power reactors. 2. The light uranium fraction is the most "protected" in the uranium component of fuel and mixed with regenerated uranium will in addition become a low enriched uranium fuel, that will weaken the problem of uncontrolled proliferation of fissile materials. 3. 231Pa doping into the fuel stabilizes its multiplying properties that will allow us to implement long-term fuel residence time and eventually to increase the export potential of all nuclear power technologies. 4. The thorium reserves being near city Krasnoufimsk (Russia) are large enough for operation of large-scale nuclear power of the Russian Federation of 70 GWe capacity during more than a quarter century under assumption that thorium is loaded into blankets of hybrid TNS only. The general conclusion: the inclusion of a small number of hybrid reactors with Th-blanket into the future nuclear power will allow us substantially to solve its problems, as well as to increase its export potential.

  4. Highest average burnups achieved by MTR fuel elements of the IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Damy, Margaret A.; Terremoto, Luis A.A.; Silva, Jose E.R.; Silva, Antonio Teixeira e; Castanheira, Myrthes; Teodoro, Celso A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear (CEN)]. E-mail: madamy@ipen.br

    2007-07-01

    Different nuclear fuels were employed in the manufacture of plate type at IPEN , usually designated as Material Testing Reactor (MTR) fuel elements. These fuel elements were used at the IEA-R1 research reactor. This work describes the main characteristics of these nuclear fuels, emphasizing the highest average burn up achieved by these fuel elements. (author)

  5. Using Coupled Mesoscale Experiments and Simulations to Investigate High Burn-Up Oxide Fuel Thermal Conductivity

    Science.gov (United States)

    Teague, Melissa C.; Fromm, Bradley S.; Tonks, Michael R.; Field, David P.

    2014-12-01

    Nuclear energy is a mature technology with a small carbon footprint. However, work is needed to make current reactor technology more accident tolerant and to allow reactor fuel to be burned in a reactor for longer periods of time. Optimizing the reactor fuel performance is essentially a materials science problem. The current understanding of fuel microstructure have been limited by the difficulty in studying the structure and chemistry of irradiated fuel samples at the mesoscale. Here, we take advantage of recent advances in experimental capabilities to characterize the microstructure in 3D of irradiated mixed oxide (MOX) fuel taken from two radial positions in the fuel pellet. We also reconstruct these microstructures using Idaho National Laboratory's MARMOT code and calculate the impact of microstructure heterogeneities on the effective thermal conductivity using mesoscale heat conduction simulations. The thermal conductivities of both samples are higher than the bulk MOX thermal conductivity because of the formation of metallic precipitates and because we do not currently consider phonon scattering due to defects smaller than the experimental resolution. We also used the results to investigate the accuracy of simple thermal conductivity approximations and equations to convert 2D thermal conductivities to 3D. It was found that these approximations struggle to predict the complex thermal transport interactions between metal precipitates and voids.

  6. Burn-up and Operation Time of Fuel Elements Produced in IPEN

    Science.gov (United States)

    Tondin, Julio Benedito Marin; Filho, Tufic Madi

    2011-08-01

    The aim of this paper is to present the developed work along the operational and reliability tests of fuel elements produced in the Institute of Energetic and Nuclear Research, IPEN-CNEN/SP, from the 1980's. The study analyzed the U-235 burn evolution and the element remain in the research reactor IEA-R1. The fuel elements are of the type MTR (Material Testing Reactor), the standard with 18 plates and a 12-plate control, with a nominal mean enrichment of 20%.

  7. Analysis of Fresh Fuel Critical Experiments Appropriate for Burnup Credit Validation

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    1995-01-01

    The ANS/ANS-8.1 standard requires that calculational methods used in determining criticality safety limits for applications outside reactors be validated by comparison with appropriate critical experiments. This report provides a detailed description of 34 fresh fuel critical experiments and their analyses using the SCALE-4.2 code system and the 27-group ENDF/B-IV cross-section library. The 34 critical experiments were selected based on geometry, material, and neutron interaction characteristics that are applicable to a transportation cask loaded with pressurized-water-reactor spent fuel. These 34 experiments are a representative subset of a much larger data base of low-enriched uranium and mixed-oxide critical experiments. A statistical approach is described and used to obtain an estimate of the bias and uncertainty in the calculational methods and to predict a confidence limit for a calculated neutron multiplication factor. The SCALE-4.2 results for a superset of approximately 100 criticals are included in uncertainty analyses, but descriptions of the individual criticals are not included.

  8. Investigation of the fundamental constants stability based on the reactor Oklo burn-up analysis

    CERN Document Server

    Onegin, M S

    2014-01-01

    New severe constraints on the variation of the fine structure constant have been obtained from reactor Oklo analysis in our previous work. We investigate here how these constraints confine the parameter of BSBM model of varying $\\alpha$. Integrating the coupled system of equations from the Big Bang up to the present time and taking into account the Oklo limits we have obtained the following margin on the combination of the parameters of BSBM model: $$ |\\zeta_m (\\frac{l}{l_{pl}})^2|<6\\cdot 10^{-7}, $$ where $l_{pl}=(\\frac{G\\hbar}{c^3})^{\\frac{1}{2}} \\approx 1.6 \\cdot 10^{-33}$ cm is a Plank length and $l$ is the characteristic length of the BSBM model. The natural value of the parameter $\\zeta_m$ - the fraction of electromagnetic energy in matter - is about $10^{-4}$. As a result it is followed from our analysis that the characteristic length $l$ of BSBM theory should be considerably smaller than the Plank length to fulfill the Oklo constraints on $\\alpha$ variation.

  9. A state of the Art report on Manufacturing technology of high burn-up fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyeong Ho; Nam, Cheol; Baek, Jong Hyuk; Choi, Byung Kwon; Park, Sang Yoon; Lee, Myung Ho; Jeong, Yong Hwan

    1999-09-01

    In order to manufacturing the prototype fuel cladding, overall manufacturing processes and technologies should be thoroughly understood on the manufacturing processes and technologies of foreign cladding tubes. Generally, the important technology related to fuel cladding tube manufacturing processes for PWRs/PHWRs is divided into three stages. The first stage is to produce the zirconium sponge from zirconium sand, the second stage is to produce the zircaloy shell or TREX from zirconium sponge ingot and finally, cladding is produced from TREX or zircaloy shell. Therefore, the manufacturing processes including the first and second stages are described in brief in this technology report in order to understand the whole fuel cladding manufacturing processes. (author)

  10. 77 FR 26050 - Burnup Credit in the Criticality Safety Analyses of Pressurized Water Reactor Spent Fuel in...

    Science.gov (United States)

    2012-05-02

    ... acceptance criteria contained in NUREG-1536, Revision 1, ``Standard Review Plan for Spent Fuel Dry Storage Systems at a General License Facility,'' NUREG-1567, ``Standard Review Plan for Spent Fuel Dry Storage Facilities,'' and NUREG-1617, ``Standard Review Plan for Transportation Packages for Spent Nuclear...

  11. Optimization of FRAM precision for isotopic measurements on large samples of low-burnup PuO2

    Energy Technology Data Exchange (ETDEWEB)

    Vo, Duc T [Los Alamos National Laboratory; Wenz, Tracy R [Los Alamos National Laboratory; Sampson, Thomas E [Los Alamos National Laboratory

    2009-01-01

    The gamma ray spectrum of plutonium contains measurable gamma rays ranging in energy from 60 keV to above 1 MeV. The FRAM gamma ray isotopic analysis code can analyze data from all types of HPGe detectors in this energy range typically using planar detectors in the energy range 60-210 keV or 120-451 keV and using coaxial detectors in the energy ranges 120-451 keV or 200-1001 keV. The statistical measurement precision depends upon the detector/energy range combination as well as the characteristics of the sample and any addition filters. In this paper we carry out the optimization of measurement precision for the important case of a multi-kg sample of low bumup PuO{sub 2} contained in a DOE 3013 Standard-compatible long-teml storage container.

  12. High-Pressure Liquid Chromatography of Irradiated Nuclear Fue - Separation of Neodymium for Burn-up Determination

    DEFF Research Database (Denmark)

    Larsen, N. R.

    1979-01-01

    Neodymium is separated from solutions of spent nuclear fuel by high-pressure liquid chromatography in methanol-nitric acid-water media using an anion-exchange column. Chromatograms obtained by monitoring at 280 nm, illustrate the difficulties especially with the fission product ruthenium in nuclear...... chemistry. Preseparation of the rare earths and trivalent actinides using a di(2-ethylhexyl)phosphoric acid/kieselguhr column is described....

  13. 输运-燃耗程序系统及ADS基准题的计算%Transport-Burnup Code System and Its Application for IAEA ADS Benchmark

    Institute of Scientific and Technical Information of China (English)

    蒋校丰; 谢仲生

    2004-01-01

    为了对加速器驱动次临界系统(Accwlerator Driven System)的中子经济性和安全性做进一步研究,IAEA启动了ADS合作研究计划,提出了ADS基准题.基准题分为几个阶段,第一阶段主要致力于ADS的中子性能和核数据库及计算程序的验证.本文开发了ANS N-DOT4.2-ORIGEN2输运-燃耗程序系统,并对该基准题进行了给定次临界度下的富集度、零燃耗下的径向与轴向功率分布、空泡效应、外源价值和燃耗的计算,取得了满意的结果.

  14. Using Finite Model Analysis and Out of Hot Cell Surrogate Rod Testing to Analyze High Burnup Used Nuclear Fuel Mechanical Properties

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL; Jiang, Hao [ORNL; Wang, Hong [ORNL

    2014-07-01

    Based on a series of FEA simulations, the discussions and the conclusions concerning the impact of the interface bonding efficiency to SNF vibration integrity are provided in this report; this includes the moment carrying capacity distribution between pellets and clad, and the impact of cohesion bonding on the flexural rigidity of the surrogate rod system. As progressive de-bonding occurs at the pellet-pellet interfaces and at the pellet-clad interface, the load ratio of the bending moment carrying capacity gradually shifts from the pellets to the clad; the clad starts to carry a significant portion of the bending moment resistance until reaching the full de-bonding state at the pellet-pellet interface regions. This results in localized plastic deformation of the clad at the pellet-pellet-clad interface region; the associated plastic deformations of SS clad leads to a significant degradation in the stiffness of the surrogate rod. For instance, the flexural rigidity was reduced by 39% from the perfect bond state to the de-bonded state at the pellet-pellet interfaces.

  15. Twenty-second water reactor safety information meeting. Volume 2: Severe accident research, thermal hydraulic research for advanced passive LWRs, high-burnup fuel behavior

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.

    1995-04-01

    This three-volume report contains papers presented at the Twenty-Second Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 24-26, 1994. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Finland, France, Italy, Japan, Russia, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting.

  16. Burn-Up Determination by High Resolution Gamma Spectrometry: Spectra from Slightly-Irradiated Uranium and Plutonium between 400-830 keV

    Energy Technology Data Exchange (ETDEWEB)

    Forsyth, R.S.; Ronqvist, N.

    1966-08-15

    Previously published studies of the short-cooled fission product spectra of irradiated uranium have been severely restricted by the poor energy resolution of the sodium iodide detectors used. In this report are presented fission product spectra of irradiated uranium and plutonium obtained by means of a lithium-drifted germanium detector. The resolved gamma peaks have been assigned to various fission products by correlation of measured energy and half-life values with published data. By simultaneous study of the spectra of two irradiated mixtures of plutonium and uranium, the possibility of using the activities of Ru-103 and Ru-106 as a measure of the relative fission rate in U-235 and Pu-239 has been briefly examined.

  17. Effect of Burnable Absorbers on Inert Matrix Fuel Performance and Transuranic Burnup in a Low Power Density Light-Water Reactor

    Directory of Open Access Journals (Sweden)

    Geoff Recktenwald

    2013-04-01

    Full Text Available Zirconium dioxide has received particular attention as a fuel matrix because of its ability to form a solid solution with transuranic elements, natural radiation stability and desirable mechanical properties. However, zirconium dioxide has a lower coefficient of thermal conductivity than uranium dioxide and this presents an obstacle to the deployment of these fuels in commercial reactors. Here we show that axial doping of a zirconium dioxide based fuel with erbium reduces power peaking and fuel temperature. Full core simulations of a modified AP1000 core were done using MCNPX 2.7.0. The inert matrix fuel contained 15 w/o transuranics at its beginning of life and constituted 28% of the assemblies in the core. Axial doping reduced power peaking at startup by more than ~23% in the axial direction and reduced the peak to average power within the core from 1.80 to 1.44. The core was able to remain critical between refueling while running at a simulated 2000 MWth on an 18 month refueling cycle. The results show that the reactor would maintain negative core average reactivity and void coefficients during operation. This type of fuel cycle would reduce the overall production of transuranics in a pressurized water reactor by 86%.

  18. Development of the MCNPX depletion capability: A Monte Carlo linked depletion method that automates the coupling between MCNPX and CINDER90 for high fidelity burnup calculations

    Science.gov (United States)

    Fensin, Michael Lorne

    Monte Carlo-linked depletion methods have gained recent interest due to the ability to more accurately model complex 3-dimesional geometries and better track the evolution of temporal nuclide inventory by simulating the actual physical process utilizing continuous energy coefficients. The integration of CINDER90 into the MCNPX Monte Carlo radiation transport code provides a high-fidelity completely self-contained Monte-Carlo-linked depletion capability in a well established, widely accepted Monte Carlo radiation transport code that is compatible with most nuclear criticality (KCODE) particle tracking features in MCNPX. MCNPX depletion tracks all necessary reaction rates and follows as many isotopes as cross section data permits in order to achieve a highly accurate temporal nuclide inventory solution. This work chronicles relevant nuclear history, surveys current methodologies of depletion theory, details the methodology in applied MCNPX and provides benchmark results for three independent OECD/NEA benchmarks. Relevant nuclear history, from the Oklo reactor two billion years ago to the current major United States nuclear fuel cycle development programs, is addressed in order to supply the motivation for the development of this technology. A survey of current reaction rate and temporal nuclide inventory techniques is then provided to offer justification for the depletion strategy applied within MCNPX. The MCNPX depletion strategy is then dissected and each code feature is detailed chronicling the methodology development from the original linking of MONTEBURNS and MCNP to the most recent public release of the integrated capability (MCNPX 2.6.F). Calculation results of the OECD/NEA Phase IB benchmark, H. B. Robinson benchmark and OECD/NEA Phase IVB are then provided. The acceptable results of these calculations offer sufficient confidence in the predictive capability of the MCNPX depletion method. This capability sets up a significant foundation, in a well established and supported radiation transport code, for further development of a Monte Carlo-linked depletion methodology which is essential to the future development of advanced reactor technologies that exceed the limitations of current deterministic based methods.

  19. Fiber optic Cerenkov radiation sensor system to estimate burn-up of spent fuel: characteristic evaluation of the system using Co-60 source

    Science.gov (United States)

    Shin, S. H.; Jang, K. W.; Jeon, D.; Hong, S.; Kim, S. G.; Sim, H. I.; Yoo, W. J.; Park, B. G.; Lee, B.

    2013-09-01

    Cerenkov radiation occurs when charged particles are moving faster than the speed of light in a transparent dielectric medium. In optical fibers, the Cerenkov light also can be generated due to their dielectric components. Accordingly, the radiation-induced light signals can be obtained using optical fibers without any scintillating material. In this study, to measure the intensities of Cerenkov radiation induced by gamma-rays, we have fabricated the fiber-optic Cerenkov radiation sensor system using silica optical fibers, plastic optical fibers, multi-anode photomultiplier tubes, and a scanning system. To characterize the Cerenkov radiation generated in optical fibers, the spectra of Cerenkov radiation generated in the silica and plastic optical fibers were measured. Also, the intensities of Cerenkov radiation induced by gamma-rays generated from a cylindrical Co-60 source with or without lead shielding were measured using the fiberoptic Cerenkov radiation sensor system.

  20. Analysis of the burnup of the control rods with the COREMASTER-Presto code; Analisis del quemado de barras de control con el codigo COREMASTER-PRESTO

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez, J.L.; Alonso, G.; Perusquia, R.; Montes, J.L.; Hernandez, H. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: jlhm@nuclear.inin-mx

    2003-07-01

    An evaluation of the capacity of the COREMASTER-Presto code, to evaluate generically the burnt of the control bars in the Laguna Verde reactors plant (CLV) is made. It was found that the code only reports burnt values of the control rods in MWD/TM, in spite of having with a second order polynomial model, for the conversion to remainder of the Boron-10 (B-10). It was observed that said model is adequate only for burnt smaller to 45,000 MWD/TM. To evaluate the burnt of the control rods it was reproduced the balance cycle of 18 months for the CLV, executing Cm-Presto during 13 consecutive cycles. First without rod burnt, taking this as the base case. Later on, cases with 1, 2 and up to 13 cycles with rod burnt were generated. When comparing results it was observed that the control rods pattern it loses reactivity lineally with the burnt one. By each 10 G Wd/T of burnt of the nucleus it is decreased the reactivity of the pattern rods {approx} 1 pcm in hot condition and of {approx} 20 pcm in cold condition. When burning three cycles those rods more burnt reached the 13,900 MWD/TM, equivalent to 36% of B-10 reduction, near value to 34% proposed by aging in the one lost study of B-10. It was observed that Cm-Presto it doesn't burn the superior node of the control rods when these are completely extracted. A one big lost of B-10, of the order of 50%, it represents only a decrease of 11% of the reactivity value of the rod. One can affirm that even when it is strongly decreased the content of B-10, the rod is continue considering as a black absorber, that is to say, thermal neutron that enters in the neutron rod that is absorbed. (Author)

  1. Qinshan Nuclear Power Plant PWR Fuel Assembly(3×3-2)In-pile Test for Extending Burnup

    Institute of Scientific and Technical Information of China (English)

    1996-01-01

    QinshanNuclearPowerPlantPWRFuelAssembly(3×3-2)In-pileTestforExtendingBurnupWangHuarong;ZhangPeishengThefirstin-pileirradiatio...

  2. Radionuclide inventories : ORIGEN2.2 isotopic depletion calculation for high burnup low-enriched uranium and weapons-grade mixed-oxide pressurized-water reactor fuel assemblies.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Ross, Kyle W. (Los Alamos National Laboratory, Los Alamos, NM); Smith, James Dean; Longmire, Pamela

    2010-04-01

    The Oak Ridge National Laboratory computer code, ORIGEN2.2 (CCC-371, 2002), was used to obtain the elemental composition of irradiated low-enriched uranium (LEU)/mixed-oxide (MOX) pressurized-water reactor fuel assemblies. Described in this report are the input parameters for the ORIGEN2.2 calculations. The rationale for performing the ORIGEN2.2 calculation was to generate inventories to be used to populate MELCOR radionuclide classes. Therefore the ORIGEN2.2 output was subsequently manipulated. The procedures performed in this data reduction process are also described herein. A listing of the ORIGEN2.2 input deck for two-cycle MOX is provided in the appendix. The final output from this data reduction process was three tables containing the radionuclide inventories for LEU/MOX in elemental form. Masses, thermal powers, and activities were reported for each category.

  3. 基于蒙特卡罗方法的三维燃耗计算研究%Study on three-dimensional burnup calculation based on Monte Carlo method

    Institute of Scientific and Technical Information of China (English)

    薛小刚; 周培德; 柯国土; 赵守智; 杨勇

    2005-01-01

    采用通过编写连接MCNP程序和ORIGEN2程序的接口处理程序的方法进行快中子系统的燃耗计算.由MCNP、ORIGEN2、接口处理程序和截面文件组成的软件系统可用于燃料或堆芯非均匀布置快中子系统的燃料同位素成分和燃耗反应性损失计算,在燃耗反应性损失计算中采用了伪裂变产物的方法.介绍程序系统的研制情况,并给出用该软件系统计算中国实验快堆首炉堆芯和OECD/NEAMOX燃料快堆基准题的燃耗计算结果.

  4. 蒙卡-燃耗程序系统及ADS基准题的计算%Monte Carlo-burnup code system and it's application to IAEA ADS benchmark

    Institute of Scientific and Technical Information of China (English)

    蒋校丰; 谢仲生

    2003-01-01

    为对核废物进行嬗变,提出了加速器驱动的次临界系统(ADS).由于外中子源的存在及中子注量率的各向异性,ADS的核计算和常规反应堆的有较大的差别,原则上需要应用中子输运理论的方法,目前尚无成熟的计算方法与程序.为此,IAEA提出了ADS基准题.基准题分为几个阶段,第一阶段主要致力于ADS的中子性能分析,检验现有核数据库、计算方法和程序的可靠性及不确定性.开发了MC-NT-ORIGEN2程序系统,并对该基准题进行了给定次临界点下的富集度、零燃耗下的径向与轴向的功率分布、反应性空泡效应、外源价值和燃耗的计算,取得了满意的结果.

  5. 板型燃料元件燃耗自动测量系统设计%Design of Automatic Measurement System for Plate-Type Fuel Element Burn-up

    Institute of Scientific and Technical Information of China (English)

    张冒; 何朝明; 程珮珮

    2014-01-01

    现有的板型燃料元件的燃耗无损测量自动化水平较低,只适用于单点测量.结合板型燃料元件的结构特点及燃耗测量要求,设计板型燃料元件自动测量系统.自动测量系统集成了交互式测点规划、运动系统控制、γ能量谱仪控制、数据处理及结果可视化功能,实现了板型燃料元件多点、分段、自动化、无损燃耗测量.系统运行结果表明:该系统可靠性高,操作简单,能够有效地提高测量效率、精度,降低操作人员的工作量.

  6. Incorporation of the variation in conductivity with burnup in the stability of code predictive LAPUR; Incroporacion de la variacion de la conductividad con el quemado en el codigo de estabilidad predictivo LAPUR

    Energy Technology Data Exchange (ETDEWEB)

    Escriba, A.; Munoz-cobo, J. L.; Merino, R.; Melara, J.; Albendea, M.

    2013-07-01

    In the field of nuclear safety, the analysis of the stability of boiling water reactors is one of the biggest challenges for researchers. LAPUR code that allows to obtain the parameters of stability of the plant (Decay rate and frequency), being one of the programs used by IBERDROLA can be used for these calculations. With the collaboration of the research group TIN of the Polytechnic University of Valencia, a model of loss of conductivity of uranium has joined with the burned LAPUR. This update allows you to play the phenomenon in a more realistic way. This improvement has been validated and verified contrasting results with reference values.

  7. Gamma-ray spectroscopy on irradiated fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Terremoto, Luis Antonio Albiac [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear], e-mail: laaterre@ipen.br

    2009-07-01

    The recording of gamma-ray spectra along an irradiated fuel rod allows the fission products to be qualitatively and quantitatively examined. Among all nondestructive examinations performed on irradiated fuel rods by gamma-ray spectroscopy, the most comprehensive one is the average burnup measurement, which is quantitative. Moreover, burnup measurements by means of gamma-ray spectroscopy are less time-consuming and waste-generating than burnup measurements by radiochemical, destructive methods. This work presents the theoretical foundations and experimental techniques necessary to measure, using nondestructive gamma-ray spectroscopy, the average burnup of irradiated fuel rods in a laboratory equipped with hot cells. (author)

  8. 76 FR 52356 - Indiana Michigan Power Company, Donald C. Cook Nuclear Plant, Unit 1; Environmental Assessment...

    Science.gov (United States)

    2011-08-22

    ... accidents that involve damage or melting of the fuel in the reactor core, fuel rod integrity has been shown...\\ fuel to a peak rod average burnup limit of 62 GWD/MTU would allow for more effective fuel management... use Optimized ZIRLO\\TM\\ fuel design with a peak rod average burnup as high as 62 GWD/MTU; the...

  9. COMPARATIVE ANALYSIS OF STRUCTURAL CHANGES IN U-MO DISPERSED FUEL OF FULL-SIZE FUEL ELEMENTS AND MINI-RODS IRRADIATED IN THE MIR REACTOR

    Directory of Open Access Journals (Sweden)

    ALEKSEY. L. IZHUTOV

    2013-12-01

    The full-size fuel rods were irradiated up to an average burnup of ∼ 60%235U; the mini-rods were irradiated to an average burnup of ∼ 85%235U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%.

  10. 基于MCNP和ORIGEN2耦合程序的IHNI-1型堆裂变产物中毒及燃耗分析%The fission product poisoning and burnup calculation for IHNI-1 reactor based on coupled code of MCNP-ORIGEN2

    Institute of Scientific and Technical Information of China (English)

    张信一; 赵柱民; 江新标; 郭和伟; 陈立新; 周永茂

    2012-01-01

    To calculate the fission product poisoning and bumup of the reactor accurately, the paper sets up the coupled calculation methods based on MCNP code and ORIGEN2 code and program data translation, cross section revision and date interface codes. Making use of elaborate reactor model to calculate the fission product poisoning and bumup for in-hospital neutron irradiator mark 1 reactor.%为了准确地计算反应堆的裂变产物中毒和燃耗问题,开发了一套蒙特卡罗方法程序系统.利用通用的燃耗计算方法,基于MCNP和ORIGEN2,编写了相关的数据转换、截面修正、数据接口程序,实现了MCNP和ORIGEN2程序的耦合.采用堆芯精细结构划分,对医院中子照射器Ⅰ型堆裂变产物中毒和燃耗进行了计算分析.

  11. A comparative study of MONTEBURNS and MCNPX 2.6.0 codes in ADS simulations

    Energy Technology Data Exchange (ETDEWEB)

    Barros, Graiciany P.; Pereira, Claubia; Veloso, Maria A.F.; Velasquez, Carlos E.; Costa, Antonella L., E-mail: gbarros@ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2013-07-01

    The possible use of the MONTEBURNS and MCNPX 2.6.0 codes in Accelerator-driven systems (ADSs) simulations for fuel evolution description is discussed. ADSs are investigated for fuel breeding and long-lived fission product transmutation so simulations of fuel evolution have a great relevance. The burnup/depletion capability is present in both studied codes. MONTEBURNS code links Monte Carlo N-Particle Transport Code (MCNP) to the radioactive decay burnup code ORIGEN2, whereas MCNPX depletion/ burnup capability is a linked process involving steady-state flux calculations by MCNPX and nuclide depletion calculations by CINDER90. A lead-cooled accelerator-driven system fueled with thorium was simulated and the results obtained using MONTEBURNS code and the results from MCNPX 2.6.0 code were compared. The system criticality and the variation of the actinide inventory during the burnup were evaluated and the results indicate a similar behavior between the results of each code. (author)

  12. FY09 PROGRESS: MULTI-ISOTOPE PROCESS (MIP) MONITOR

    Energy Technology Data Exchange (ETDEWEB)

    Schwantes, Jon M.; Orton, Christopher R.; Fraga, Carlos G.; Christensen, Richard; Laspe, Amy R.; Ward, Rebecca M.

    2009-10-18

    Model and experimental estimates of the Multi-Isotope Process Monitor performance for determining burnup after dissolution and acid concentration during solvent extraction steps during reprocessing of spent nuclear fuel are presented.

  13. Separation and Purification of Fissiogenic Ruthenium From Irradiated Uranium

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    Ruthenium is an important fission product. Its isotopic composition may reflect the burnup or the initial uranium enrichment of nuclear fuel. So the separation and purification method of fission products of Ruthenium from irradiated uranium was studied and established.

  14. Power level effects on thorium-based fuels in pressure-tube heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bromley, B.P.; Edwards, G.W.R., E-mail: blair.bromley@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Sambavalingam, P. [Univ. of Ontario Inst. of Technology, Oshawa, Ontario (Canada)

    2016-06-15

    Lattice and core physics modeling and calculations have been performed to quantify the impact of power/flux levels on the reactivity and achievable burnup for 35-element fuel bundles made with Pu/Th or U-233/Th. The fissile content in these bundles has been adjusted to produce on the order of 20 MWd/kg burnup in homogeneous cores in a 700 MWe-class pressure-tube heavy water reactor, operating on a once-through thorium cycle. Results demonstrate that the impact of the power/flux level is modest for Pu/Th fuels but significant for U-233/Th fuels. In particular, high power/flux reduces the breeding and burnup potential of U-233/Th fuels. Thus, there may be an incentive to operate reactors with U-233/Th fuels at a lower power density or to develop alternative refueling schemes that will lower the time-average specific power, thereby increasing burnup.(author)

  15. Reactivity follow of the two first loadings of the Jose Cabrera Reactor; Seguimiento de la ractividad durante las dos primeras cargas del Reactor de la Central Nuclear Jose Cabrera

    Energy Technology Data Exchange (ETDEWEB)

    Bru, A.

    1975-07-01

    In this paper the first two cores together with the in-core measurements taken during the operation of the Nuclear Power Station Jose Cabrera are described. The results of this measurements have been processed with the INCORE and FOLLOW codes. The peaking factors and the boron concentration versus burn-up are displayed. The final burn-up of the fuel elements in these two loading are given, too. (Author)

  16. 44 BWR Waste Package Loading Curve Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    J.M. Scaglione

    2001-11-05

    The objective of this calculation is to evaluate the required minimum burnup as a function of average initial boiling water reactor (BWR) assembly enrichment that would permit loading of fuel into a potential 44 BWR waste package (WP). The potential WP design is illustrated in Attachment I. The scope of this calculation covers a range of initial enrichments from 1.5 through 5.0 weight percent U-235, and a burnup range of 0 through 50 GWd/mtU.

  17. Reference Spent Fuel and Its Source Terms for a Design of Deep Geological Disposal System

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Dong Keun

    2005-12-15

    In this study, current status and future trend of domestic spent fuels were analyzed to propose reference spent nuclear fuel. And then, source terms needed for design of a deep geological disposal system were calculated using ORIGEN-ARP. The reference spent fuels selected based on assembly physical dimension, inventory projection, trend of initial enrichment of 235U, discharge burnup are as follows; The 17x17 Korean Optimized Fuel Assembly with initial enrichment of 4.0 wt.% 235U and discharge burnup of 45 GWD/MTU was adopted as a low-burnup representative fuel. For the high-burnup representative fuel, 16x16 Korean Standard Fuel Assembly with initial enrichment of 4.5 wt.% 235U and discharge burnup of 55 GWD/MTU was chosen. CANDU fuel with initial enrichment of 0.711 wt.% 235U and discharge burnup of 7.5 GWD/MTU was also considered. For these reference fuels, decay heat, radiation intensity and spectrum, nuclide concentration, and individual nuclide radioactivity were calculated using ORIGEN-ARP for a disposal system design. It is expected that the source terms estimated in this study will be applied to the disposal system development in the future.

  18. Verify Super Double-Heterogeneous Spherical Lattice Model for Equilibrium Fuel Cycle Analysis AND HTR Spherical Super Lattice Model for Equilibrium Fuel Cycle Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Gray S. Chang

    2005-11-01

    The currently being developed advanced High Temperature gas-cooled Reactors (HTR) is able to achieve a simplification of safety through reliance on innovative features and passive systems. One of the innovative features in these HTRs is reliance on ceramic-coated fuel particles to retain the fission products even under extreme accident conditions. Traditionally, the effect of the random fuel kernel distribution in the fuel pebble / block is addressed through the use of the Dancoff correction factor in the resonance treatment. However, the Dancoff correction factor is a function of burnup and fuel kernel packing factor, which requires that the Dancoff correction factor be updated during Equilibrium Fuel Cycle (EqFC) analysis. An advanced KbK-sph model and whole pebble super lattice model (PSLM), which can address and update the burnup dependent Dancoff effect during the EqFC analysis. The pebble homogeneous lattice model (HLM) is verified by the burnup characteristics with the double-heterogeneous KbK-sph lattice model results. This study summarizes and compares the KbK-sph lattice model and HLM burnup analyzed results. Finally, we discuss the Monte-Carlo coupling with a fuel depletion and buildup code - ORIGEN-2 as a fuel burnup analysis tool and its PSLM calculated results for the HTR EqFC burnup analysis.

  19. Preliminary design report: Babcock and Wilcox BR-100 100-ton rail/barge spent fuel shipping cask

    Energy Technology Data Exchange (ETDEWEB)

    None

    1990-02-01

    The purpose of this document is to provide information on burnup credit as applied to the preliminary design of the BR-100 shipping cask. There is a brief description of the preliminary basket design and the features used to maintain a critically safe system. Following the basket description is a discussion of various criticality analyses used to evaluate burnup credit. The results from these analyses are then reviewed in the perspective of fuel burnups expected to be shipped to either the final repository or a Monitored Retrievable Storage (MRS) facility. The hurdles to employing burnup credit in the certification of any cask are then outlines and reviewed. the last section gives conclusions reached as to burnup credit for the BR-100 cask, based on our analyses and experience. All information in this study refers to the cask configured to transport PWR fuel. Boiling Water Reactor (BWR) fuel satisfies the criticality requirements so that burnup credit is not needed. All calculations generated in the preparation of this report were based upon the preliminary design which will be optimized during the final design. 8 refs., 19 figs., 16 tabs.

  20. Fuel Behaviour at High During RIA and LOCA Accidents; Comportamiento del Combustible de Alto Quemado en Accidents RIA y LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Barrio del Juanes, M.T.; Garcia Cuesta, J.C.; Vallejo Diaz, I.; Herranz Puebla

    2001-07-01

    Safety analysis of high burnup fuel requires ensuring the acceptable performance under design basis accidents, in particular during conditions representative of Reactivity Accidents (RIA) and Loss-of-Coolant Accidents (LOCA). The report's objective is to compile the state of the art on these issues. This is mainly focused in the effort made to define the applicability of safety criteria to the high burnup fuel. Irradiation damage modifies fuel rod properties, thus the probability of fuel to withstand thermal and mechanical loads during an accident could be quite different compared with unirradiated fuel. From the thermal point of view, fuel conductivity is the most affected property, decreasing notably with irradiation. From the mechanical point of view, a change in the pellet microstructure at its periphery is observed at high burnup (remiffect). Cladding is also effected during operation, showing a significant external and internal corrosion. All these phenomena result in the decrease of efficiency in heat transfer an in the reduction of capability to accommodate mechanical loads; this situation is especially significant at high burnup, when pellet-cladding mechanical interaction is present. Knowledge about these phenomena is not possible without appropriate experimental programmes. The most relevant have been performed in France, Japan, United States and Russia. Results obtained with fuel at high burnup show significant differences with respect to the phenomena observed in fuel at the present discharge burnup. Indeed, this is the encouragement to research about this occurrence. This study is framed within the CSN-CIEMAT agreement, about Fuel Thermo-Mechanical Behaviour at High Burnup. (Author) 172 refs.

  1. Characterization of used nuclear fuel with multivariate analysis for process monitoring

    Science.gov (United States)

    Dayman, Kenneth J.; Coble, Jamie B.; Orton, Christopher R.; Schwantes, Jon M.

    2014-01-01

    This paper presents initial development of a reactor-type classifier that is used to select a reactor-specific partial least squares model to predict used nuclear fuel burnup. Nuclide activities for prototypic used fuel samples were generated in ORIGEN-ARP and used to investigate techniques to characterize used nuclear fuel in terms of reactor type (pressurized or boiling water reactor) and burnup. A variety of reactor type classification algorithms, including k-nearest neighbors, linear and quadratic discriminant analyses, and support vector machines, were evaluated to differentiate used fuel from pressurized and boiling water reactors. Then, reactor type-specific partial least squares models were developed to predict the burnup of the fuel. Using these reactor type-specific models instead of a model trained for all light water reactors improved the accuracy of burnup predictions. The developed classification and prediction models were combined and applied to a large dataset that included eight fuel assembly designs, two of which were not used in training the models, and spanned the range of the initial 235U enrichment, cooling time, and burnup values expected of future commercial used fuel for reprocessing. Error rates were consistent across the range of considered enrichment, cooling time, and burnup values. Average absolute relative errors in burnup predictions for validation data both within and outside the training space were 0.0574% and 0.0597%, respectively. The errors seen in this work are artificially low, because the models were trained, optimized, and tested on simulated, noise-free data. However, these results indicate that the developed models may generalize well to new data and that the proposed approach constitutes a viable first step in developing a fuel characterization algorithm based on gamma spectra.

  2. Post-irradiation examinations and high-temperature tests on undoped large-grain UO2 discs

    Science.gov (United States)

    Noirot, J.; Pontillon, Y.; Yagnik, S.; Turnbull, J. A.

    2015-07-01

    Within the Nuclear Fuel Industry Research (NFIR) programme, several fuel variants -in the form of thin circular discs - were irradiated in the Halden Boiling Water Reactor (HBWR) at burn-ups up to ∼100 GWd/tHM. The design of the fuel assembly was similar to that used in other HBWR programmes: the assembly contained several rods with fuel discs sandwiched between Mo discs, which limited temperature differences within each fuel disc. One such variant was made of large-grain UO2 discs (3D grain size = ∼45 μm) which were subjected to three burn-ups: 42, 72 and 96 GWd/tHM. Detailed characterizations of some of these irradiated large-grain UO2 discs were performed in the CEA Cadarache LECA-STAR hot laboratory. The techniques used included electron probe microanalysis (EPMA), scanning electron microscopy (SEM) and secondary ion mass spectrometry (SIMS). Comparisons were then carried out with more standard grain size UO2 discs irradiated under the same conditions. Examination of the high burn-up large-grain UO2 discs revealed the limited formation of a high burn-up structure (HBS) when compared with the standard-grain UO2 discs at similar burn-up. High burn-up discs were submitted to temperature transients up to 1200 °C in the heating test device called Merarg at a relatively low temperature ramp rate (0.2 °C/s). In addition to the total gas release during these tests, the release peaks throughout the temperature ramp were monitored. Tests at 1600 °C were also conducted on the 42 GWd/tHM discs. The fuels were then characterized with the same microanalysis techniques as those used before the tests, to investigate the effects of these tests on the fuel's microstructure and on the fission gas behaviour. This paper outlines the high resistance of this fuel to gas precipitation at high temperature and to HBS formation at high burn-up. It also shows the similarity of the positions, within the grains, where HBS forms at high burn-up and where bubbles appear during the low

  3. Comparative Analysis of Structural Changes In U-Mo Dispersed Fuel of Full-Size Fuel Elements And Mini-Rods Irradiated In The MIR Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Izhutov, Aleksey L.; Iakovlev, Valeriy V.; Novoselov, Andrey E. and others

    2013-12-15

    The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60%{sup 235}U; the mini-rods were irradiated to an average burnup of ∼ 85%{sup 235}U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%.

  4. Recent results on the RIA test in IGR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Asmolov, V.; Yegorova, L. [Nuclear Safety Institute, Moscow (Russian Federation)

    1997-01-01

    At the 23d WRSM meeting the data base characterizing results of VVER high burnup fuel rods tests under reactivity-initiated accident (RIA) conditions was presented. Comparison of PWR and VVER failure thresholds was given also. Additional analysis of the obtained results was being carried out during 1996. The results of analysis show that the two different failure mechanisms were observed for PWR and VVER fuel rods. Some factors which can be as the possible reasons of these differences are presented. First of them is the state of preirradiated cladding. Published test data for PWR high burnup fuel rods demonstrated that the PWR high burnup fuel rods failed at the RIA test are characterized by very high level of oxidation and hydriding for the claddings. Corresponding researches were performed at Institute of Atomic Reactors (RLAR, Dimitrovgrad, Russia) for large set of VVER high burnup fuel rods. Results of these investigations show that preirradiated commercial Zr-1%Nb claddings practically keep their initial levels of oxidation and H{sub 2} concentration. Consequently the VVER preirradiated cladding must keep the high level of mechanical properties. The second reason leading to differences between failure mechanisms for two types of high burnup fuel rods can be the test conditions. Now such kind of analysis have been performed by two methods.

  5. Gamma-ray spectroscopy on irradiated MTR fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Terremoto, L.A.A. E-mail: laaterre@net.ipen.br; Zeituni, C.A.; Perrotta, J.A.; Silva, J.E.R. da

    2000-08-11

    The availability of burnup data is an important requirement in any systematic approach to the enhancement of safety, economics and performance of a nuclear research reactor. This work presents the theory and experimental techniques applied to determine, by means of nondestructive gamma-ray spectroscopy, the burnup of Material Testing Reactor (MTR) fuel elements irradiated in the IEA-R1 research reactor. Burnup measurements, based on analysis of spectra that result from collimation and detection of gamma-rays emitted in the decay of radioactive fission products, were performed at the reactor pool area. The measuring system consists of a high-purity germanium (HPGe) detector together with suitable fast electronics and an on-line microcomputer data acquisition module. In order to achieve absolute burnup values, the detection set (collimator tube+HPGe detector) was previously calibrated in efficiency. The obtained burnup values are compared with ones provided by reactor physics calculations, for three kinds of MTR fuel elements with different cooling times, initial enrichment grades and total number of fuel plates. Both values show good agreement within the experimental error limits.

  6. Mechanical behaviors of the dispersion nuclear fuel plates induced by fuel particle swelling and thermal effect I: Effects of variations of the fuel particle volume fractions

    Science.gov (United States)

    Wang, Qiming; Yan, Xiaoqing; Ding, Shurong; Huo, Yongzhong

    2010-05-01

    A new method of modeling the in-pile mechanical behaviors of dispersion nuclear fuel elements is proposed. Considering the irradiation swelling together with the thermal effect, numerical simulations of the in-pile mechanical behaviors are performed with the developed finite element models for different fuel particle volume fractions of the fuel meat. The effects of the particle volume fractions on the mechanical performances of the fuel element are studied. The research results indicate that: (1) the maximum Mises stresses and equivalent plastic strains at the matrix increase with the particle volume fractions at each burnup; the locations of the maximum first principal stresses shift with increasing burnup; at low burnups, the maximum first principal stresses increase with the particle volume fractions; while at high burnups, the 20% volume fraction case holds the lowest value; (2) at the cladding, the maximum equivalent plastic strains and the tensile principal stresses increase with the particle volume fractions; while the maximum Mises stresses do not follow this order at high burnups; (3) the maximum Mises stresses at the fuel particles increase with the particle volume fractions, and the particles will engender plastic strains until the particle volume fraction reaches high enough.

  7. Reactor-specific spent fuel discharge projections: 1986 to 2020

    Energy Technology Data Exchange (ETDEWEB)

    Heeb, C.M.; Walling, R.C.; Purcell, W.L.

    1987-03-01

    The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from US commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent-fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water reactors (BWR). The projections are based on individual reactor information supplied by the US reactor owners. The basic information is adjusted to conform to Energy Information Agency (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: (1) No new orders with extended burnup, (2) No new orders with constant burnup, (3) Upper reference (which assumes extended burnup), (4) Upper reference with constant burnup, and (5) Lower reference (which assumes extended burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum-at-reactor storage, and for storage requirements assuming maximum-at-reactor plus intra-utility transshipment of spent fuel. 6 refs., 8 figs., 8 tabs.

  8. Reactor-specific spent fuel discharge projections, 1987-2020

    Energy Technology Data Exchange (ETDEWEB)

    Walling, R.C.; Heeb, C.M.; Purcell, W.L.

    1988-03-01

    The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from U.S. commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water eactors (BWR), and one existing high temperature gas reactor (HTGR). The projections are based on individual reactor information supplied by the U.S. reactor owners. The basic information is adjusted to conform to Energy Information Administration (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: No New Orders (assumes increasing burnup), No New Orders with No Increased Burnup, Upper Reference (assumes increasing burnup), Upper Reference with No Increased Burnup, and Lower Reference (assumes increasing burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum at-reactor storage, and for storage requirements assuming maximum at-reactor storage plus intra-utility transshipment of spent fuel. 8 refs., 8 figs., 10 tabs.

  9. Sequestration of radioactive iodine in silver-palladium phases in commercial spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Buck, Edgar C.; Mausolf, Edward J.; McNamara, Bruce K.; Soderquist, Chuck Z.; Schwantes, Jon M.

    2016-12-01

    Radioactive iodine is the Achilles’ heel in the design for the safe geological disposal of spent UO2 nuclear fuel. Iodine’s high solubility and anticipated instant release during waste package compromise jeopardize performance assessment calculations. However, dissolution studies have indicated that the instant release fraction (IRF) of radioiodine (I) does not correlate with increasing fuel burn-up. In fact, there is a peak in the release iodine at around 50-60 Mwd/kgU and with increasing burn-up the instant release of iodine decreases. Detailed electron microscopy analysis of high burn-up fuel (~80 MWd/kgU) has revealed the presence of (Pd,Ag)(I,Br) nano-particles. As UO2 fuels are irradiated, the Ag and Pd content increases, from 239Pu fission, enabling radioiodine to be retained. The occurrence of these phases in nuclear fuels may have significant implications for the long-term behavior of iodine.

  10. Fuel performance annual report for 1991. Volume 9

    Energy Technology Data Exchange (ETDEWEB)

    Painter, C.L.; Alvis, J.M.; Beyer, C.E. [Pacific Northwest Lab., Richland, WA (United States); Marion, A.L. [Oregon State Univ., Corvallis, OR (United States). Dept. of Nuclear Engineering; Payne, G.A. [Northwest Coll. and Univ. Association for Science, Richland, WA (United States); Kendrick, E.D. [Nuclear Regulatory Commission, Washington, DC (United States)

    1994-08-01

    This report is the fourteenth in a series that provides a compilation of information regarding commercial nuclear fuel performance. The series of annual reports were developed as a result of interest expressed by the public, advising bodies, and the US Nuclear Regulatory Commission (NRC) for public availability of information pertaining to commercial nuclear fuel performance. During 1991, the nuclear industry`s focus regarding fuel continued to be on extending burnup while maintaining fuel rod reliability. Utilities realize that high-burnup fuel reduces the amount of generated spent fuel, reduces fuel costs, reduces operational and maintenance costs, and improves plant capacity factors by extending operating cycles. Brief summaries of fuel operating experience, fuel design changes, fuel surveillance programs, high-burnup experience, problem areas, and items of general significance are provided.

  11. Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

    Energy Technology Data Exchange (ETDEWEB)

    Monado, Fiber, E-mail: fiber.monado@gmail.com [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung, Indonesia and Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Ariani, Menik [Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Su' ud, Zaki; Waris, Abdul; Basar, Khairul; Permana, Sidik [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung (Indonesia); Aziz, Ferhat [National Nuclear Energy Agency of Indonesia (BATAN) (Indonesia); Sekimoto, Hiroshi [CRINES, Tokyo Institute of Technology, O-okoyama, Meguro-ku, Tokyo 152-8550 (Japan)

    2014-02-12

    A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance.

  12. Fabrication, properties, and tritium recovery from solid breeder materials

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, C.E. (Argonne National Lab., IL (USA)); Kondo, T. (Japan Atomic Energy Research Inst., Tokyo (Japan)); Roux, N. (CEA Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)); Tanaka, S. (Tokyo Univ. (Japan)); Vollath, D. (Kernforschungszentrum Karlsruhe GmbH (Germany, F.R.))

    1991-01-01

    The breeding blanket is a key component of the fusion reactor because it directly involves tritium breeding and energy extraction, both of which are critical to development of fusion power. The lithium ceramics continue to show promise as candidate breeder materials. This promise was recognized by the International Thermonuclear Experimental Reactor (ITER) design team in its selection of ceramics as the first option for the ITER breeder material. Blanket design studies have indicated properties in the candidate materials data base that need further investigation. Current studies are focusing on tritium release behavior at high burnup, changes in thermophysical properties with burnup, compatibility between the ceramic breeder and beryllium multiplier, and phase changes with burnup. Laboratory and in-reactor tests, some as part of an international collaboration for development of ceramic breeder materials, are underway. 133 refs., 1 fig.

  13. Methodology of Fuel Burn Up Fitting in VVER-1000 Reactor Core by Using New Ex-Vessel Neutron Dosimetry and In-Core Measurements and its Application for Routine Reactor Pressure Vessel Fluence Calculations

    Directory of Open Access Journals (Sweden)

    Borodkin Pavel

    2016-01-01

    Full Text Available Paper describes the new approach of fitting axial fuel burn-up patterns in peripheral fuel assemblies of VVER-1000 type reactors, on the base of ex-core neutron leakage measurements, neutron-physical calculations and in-core SPND measured data. The developed approach uses results of new ex-vessel measurements on different power units through different reactor cycles and their uncertainties to clear the influence of a fitted fuel burn-up profile to the RPV neutron fluence calculations. The new methodology may be recommended to be included in the routine fluence calculations used in RPV lifetime management and may be taken into account during VVER-1000 core burn-up pattern correction.

  14. Methodology of Fuel Burn Up Fitting in VVER-1000 Reactor Core by Using New Ex-Vessel Neutron Dosimetry and In-Core Measurements and its Application for Routine Reactor Pressure Vessel Fluence Calculations

    Science.gov (United States)

    Borodkin, Pavel; Borodkin, Gennady; Khrennikov, Nikolay

    2016-02-01

    Paper describes the new approach of fitting axial fuel burn-up patterns in peripheral fuel assemblies of VVER-1000 type reactors, on the base of ex-core neutron leakage measurements, neutron-physical calculations and in-core SPND measured data. The developed approach uses results of new ex-vessel measurements on different power units through different reactor cycles and their uncertainties to clear the influence of a fitted fuel burn-up profile to the RPV neutron fluence calculations. The new methodology may be recommended to be included in the routine fluence calculations used in RPV lifetime management and may be taken into account during VVER-1000 core burn-up pattern correction.

  15. Research on high-burn fuel: the Cabri water loop reactor project; La investigacion sobre combustible de alto quemado: El proyecto del reactor Cabri con lazo de agua

    Energy Technology Data Exchange (ETDEWEB)

    Conde Lopez, J. M.; Recio Santamaria, M. [Consejo de Seguridad Nuclear. Madrid (Spain)

    2000-07-01

    This paper describes the present status of the nuclear fuel utilization both in the national and international arenas. Details on the fuel-related research programs actually ongoing are given. The interests and strategic lines stablished by the CSN regarding high burnup fuel research are described. Specifically, an analysis is made of the reasons underlying the CSN's decision to participate in the IPSN's CABRI Water Loop high burnup fuel research program. The core of the paper is devoted to the description of the CABRI project contents and of the technical aspects in the behaviour of high burnup fuel that will be studied through the foreseen integral and separate effects tests. Finally, a summary of the project organization, schedule and technical program is included. (Author)

  16. In search of the ideal burn; A la busqueda del quemado ideal

    Energy Technology Data Exchange (ETDEWEB)

    Molina, D. [IBERDROLA. Madrid (Spain)

    2000-07-01

    This paper shows a two steps method able to establish optimum operating cycle length and associated fuel batches discharge burnup. Application results (performed for Cofrentes NPP) indicate a potential improvement linked to burnup increases. An international research effort called Robust Fuel Program has been launched to facilitate, among other issues, high burnup licensing. This paper summarizes bases, objectives and expectations for such a program. IBERDROLA has developed a method to analyse economical cycle performance by using relevant concepts within a fix and variable cost model. Relevant concepts means power operation, refueling and fuel. In the other hand. fix costs are independent of production while variable costs are proportional to it. Taking into account model uncertainties, it can be established that reference case (Cofrentes NPP) optimum is obtained for equilibrium cycle lengths between 21 and 22 months. Once cycle energy and reload batch fraction are established, average enrichment can also be known by using a 3D nodal simulator. For such applications, IBERDROLA has already developed a system called EVACOM able to perform very detailed fuel cycle scoping calculations covering from current conditions to equilibrium. By means of neutronic, thermohydraulic and economic coupling models some advantages of burnup increases have been analysed in a three steps schedule: available pin burnup margin, available enrichment limit and mechanical performance limit. Robust Fuel Program (RFP) was launched to facilitate technical resolution of any issue linked to design and licensing of advanced fuel, particularly high burnup fuel. Initiative for such development was taken by US electrical utilities (with the participation of UNESA) looking to a target of increasing fuel reliability and operating margins to obtain power operation enhanced economical results. (Author)

  17. Managing Wastes With and Without Plutonium Separation

    Science.gov (United States)

    1999-01-01

    Typical fuel bumup is now around 42,500 MWD/T, and this is the burnup used in the other chapters of this report. However, ORIGEN2 does not provide cross...42,500 MWD/T burnup twice—once using ORIGEN2 ’s 33,000 MWD/T cross section library and once using the 50,000 MWD/T cross section library. The difference... calculations are easier to trace. Based on the current nuclear costs and trends, MOX-OT is more likely than SGR. But, this study will still compare SGR with

  18. Chemical state of fission products in irradiated uranium carbide fuel

    Science.gov (United States)

    Arai, Yasuo; Iwai, Takashi; Ohmichi, Toshihiko

    1987-12-01

    The chemical state of fission products in irradiated uranium carbide fuel has been estimated by equilibrium calculation using the SOLGASMIX-PV program. Solid state fission products are distributed to the fuel matrix, ternary compounds, carbides of fission products and intermetallic compounds among the condensed phases appearing in the irradiated uranium carbide fuel. The chemical forms are influenced by burnup as well as stoichiometry of the fuel. The results of the present study almost agree with the experimental ones reported for burnup simulated carbides.

  19. RERTR-12 Post-irradiation Examination Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Rice, Francine [Idaho National Lab. (INL), Idaho Falls, ID (United States); Williams, Walter [Idaho National Lab. (INL), Idaho Falls, ID (United States); Robinson, Adam [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason [Idaho National Lab. (INL), Idaho Falls, ID (United States); Meyer, Mitch [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabin, Barry [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-02-01

    The following report contains the results and conclusions for the post irradiation examinations performed on RERTR-12 Insertion 2 experiment plates. These exams include eddy-current testing to measure oxide growth; neutron radiography for evaluating the condition of the fuel prior to sectioning and determination of fuel relocation and geometry changes; gamma scanning to provide relative measurements for burnup and indication of fuel- and fission-product relocation; profilometry to measure dimensional changes of the fuel plate; analytical chemistry to benchmark the physics burnup calculations; metallography to examine the microstructural changes in the fuel, interlayer and cladding; and microhardness testing to determine the material-property changes of the fuel and cladding.

  20. An evaluation of the metal fuel core performance for commercial FBR requirements

    Energy Technology Data Exchange (ETDEWEB)

    Ohta, H.; Yokoo, T. [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan)

    2001-07-01

    Neutronic and thermal hydraulic design studies are conducted on the 3,900 MWt and 800 MWt metal fuel fast breeder reactor (FBR) cores that achieve a high burnup of 150 GWd/t, which is considered to be one of the goals of future commercial FBRs. The results show that a large-scale metal fuel core with a homogeneous configuration is a promising design for future commercial FBRs from the view point of economics, core safety, effective use of uranium resources and reduction of environmental load. In small-scale high-burnup cores, a radial heterogeneous design that can improve neutronic performance and safety parameters is suitable. (author)

  1. Neutron intensity of fast reactor spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Takamatsu, Misao; Aoyama, Takafumi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-03-01

    Neutron intensity of spent fuel of the JOYO Mk-II core with a burnup of 62,500 MWd/t and cooling time of 5.2 years was measured at the spent fuel storage pond. The measured data were compared with the calculated values based on the JOYO core management code system `MAGI`, and the average C/E approximately 1.2 was obtained. It was found that the axial neutron intensity didn`t simply follow the burnup distribution, and the neutron intensity was locally increased at the bottom end of the fuel region due to an accumulation of {sup 244}Cm. (author)

  2. Modeling of Reactor Kinetics and Dynamics

    Energy Technology Data Exchange (ETDEWEB)

    Matthew Johnson; Scott Lucas; Pavel Tsvetkov

    2010-09-01

    In order to model a full fuel cycle in a nuclear reactor, it is necessary to simulate the short time-scale kinetic behavior of the reactor as well as the long time-scale dynamics that occur with fuel burnup. The former is modeled using the point kinetics equations, while the latter is modeled by coupling fuel burnup equations with the kinetics equations. When the equations are solved simultaneously with a nonlinear equation solver, the end result is a code with the unique capability of modeling transients at any time during a fuel cycle.

  3. Propagation of Isotopic Bias and Uncertainty to Criticality Safety Analyses of PWR Waste Packages

    Energy Technology Data Exchange (ETDEWEB)

    Radulescu, Georgeta [ORNL

    2010-06-01

    Burnup credit methodology is economically advantageous because significantly higher loading capacity may be achieved for spent nuclear fuel (SNF) casks based on this methodology as compared to the loading capacity based on a fresh fuel assumption. However, the criticality safety analysis for establishing the loading curve based on burnup credit becomes increasingly complex as more parameters accounting for spent fuel isotopic compositions are introduced to the safety analysis. The safety analysis requires validation of both depletion and criticality calculation methods. Validation of a neutronic-depletion code consists of quantifying the bias and the uncertainty associated with the bias in predicted SNF compositions caused by cross-section data uncertainty and by approximations in the calculational method. The validation is based on comparison between radiochemical assay (RCA) data and calculated isotopic concentrations for fuel samples representative of SNF inventory. The criticality analysis methodology for commercial SNF disposal allows burnup credit for 14 actinides and 15 fission product isotopes in SNF compositions. The neutronic-depletion method for disposal criticality analysis employing burnup credit is the two-dimensional (2-D) depletion sequence TRITON (Transport Rigor Implemented with Time-dependent Operation for Neutronic depletion)/NEWT (New ESC-based Weighting Transport code) and the 44GROUPNDF5 crosssection library in the Standardized Computer Analysis for Licensing Evaluation (SCALE 5.1) code system. The SCALE 44GROUPNDF5 cross section library is based on the Evaluated Nuclear Data File/B Version V (ENDF/B-V) library. The criticality calculation code for disposal criticality analysis employing burnup credit is General Monte Carlo N-Particle (MCNP) Transport Code. The purpose of this calculation report is to determine the bias on the calculated effective neutron multiplication factor, k{sub eff}, due to the bias and bias uncertainty associated with

  4. FY 2014 Status Report: of Vibration Testing of Clad Fuel (M4FT-14OR0805033)

    Energy Technology Data Exchange (ETDEWEB)

    Bevard, Bruce Balkcom [ORNL

    2014-03-28

    The DOE Used Fuel Disposition Campaign (UFDC) tasked Oak Ridge National Laboratory (ORNL) to investigate the behavior of light-water-reactor (LWR) fuel cladding material performance related to extended storage and transportation of UNF. ORNL has been tasked to perform a systematic study on UNF integrity under simulated normal conditions of transportation (NCT) by using the recently developed hot-cell testing equipment, Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT). To support the testing on actual high-burnup UNF, fast-neutron irradiation of pre-hydrided zirconium-alloy cladding in the High Flux Isotope Reactor (HFIR) at elevated temperatures will be used to simulate the effects of high-burnup on fuel cladding for help in understanding the cladding materials properties relevant to extended storage and subsequent transportation. The irradiated pre-hydrided metallic materials testing will generate baseline data to benchmark hot-cell testing of the actual high-burnup UNF cladding. More importantly, the HFIR-irradiated samples will be free of alpha contamination and can be provided to researchers who do not have hot cell facilities to handle highly contaminated high-burnup UNF cladding to support their research projects for the UFDC.

  5. Performance tests for integral reactor nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Dong-Seong; Yim, Jeong-Sik; Lee, Chong-Tak; Kim, Han-Soo; Koo, Yang-Hyun; Lee, Byung-Ho; Cheon, Jin-Sik; Oh, Je-Yong

    2006-02-15

    An integral type reactor SMART plans to utilize metallic Zr-U fuel which is Zr-based alloy with 34{approx}38 wt% U. In order to verify the technologies for the design and manufacturing of the fuel and get a license, performance tests were carried out. Experimental Fuel Assembly (EFA) manufactured in KAERI is being successfully irradiated in the MIR reactor of RIAR from September 4 2004, and it has achieved burnup of 0.21 g/cc as of January 25 2006. Thermal properties of irradiated Zr-U fuel were measured. Up to the phase transformation temperature, thermal diffusivity increased linearly in proportion to temperature. However its dependence on the burnup was not significant. RIA tests with 4 unirradiated Zr-U fuel rods were performed in Kurchatov Institute to establish a safety criterion. In the case of the un-irradiated Zr-U fuel, the energy deposition during the control rod ejection accident should be less than 172 cal/g to prevent the failure accompanying fuel fragmentation and dispersal. Finally the irradiation tests of fuel rods have been performed at HANARO. The HITE-2 test was successfully completed up to a burnup of 0.31 g/cc. The HITE-3 test began in February 2004 and will be continued up to a target burnup of 0.6 g/cc.

  6. Measurement of Flow Resistance Coefficient of Sodium Valve in Traveling Wave Reactor

    Institute of Scientific and Technical Information of China (English)

    LV; Ming-yu; WANG; Chong; YU; Hua-jin

    2015-01-01

    Traveling wave reactor(TWR)whose fundamental philosophy is achieving breeding and incineration of 238 U which is in the majority in natural uranium at original location through deepened burnup is a new concept fast reactor.It can effectively enhance utilization rate of uranium resources.TWR has much characteristics,such as

  7. Gamma spectrometry of irradiated fuel plates; Espectrometria gama em elementos combustiveis tipo placa irradiados

    Energy Technology Data Exchange (ETDEWEB)

    Terremoto, Luis A.A.; Zeituni, Carlos A.; Perrotta, Jose A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Div. de Engenharia do Nucleo

    1997-10-01

    This work describes the fundamental aspects of a method which uses gamma-ray spectroscopy in order to perform non-destructive burnup measurements in irradiated MTR fuel elements. Experiments based on such method will be conducted at the storage pool area of the IEA-R1 research reactor. Some preliminary results are presented. (author). 10 refs., 5 figs., 1 tab.

  8. Updated Neutron Nuclear Data Evaluation for 238Np

    Institute of Scientific and Technical Information of China (English)

    CHEN; Guo-chang; YU; Bao-sheng

    2012-01-01

    <正>The nuclear data with high accuracy for minor actinides are playing an important role in nuclear technology applications, including reactor design and operation, fuel cycle concepts, estimation of the amount of minor actinides in high burn-up reactors and the minor actinides transmutation. Base on a new

  9. Spent nuclear fuel. A review of properties of possible relevance to corrosion processes

    Energy Technology Data Exchange (ETDEWEB)

    Forsyth, R. [Caledon Consult AB, Nykoeping (Sweden)

    1995-04-01

    The report reviews the properties of spent fuel which are considered to be of most importance in determining the corrosion behaviour in groundwaters. Pellet cracking and fragment size distribution are discussed, together with the available results of specific surface area measurements on spent fuel. With respect to the importance of fuel microstructure, emphasis is placed on recent work on the so called structural rim effect, which consists of the formation of a zone of high porosity, and the polygonization of fuel grains to form many sub-grains, at the pellet rim, and appears to be initiated when the average pellet burnup exceeds a threshold of about 40 MWd/kgU. Due to neutron spectrum effects, the pellet rim is also associated with the buildup of plutonium and other actinides, which results in an enhanced local burnup and specific activity of both beta-gamma and alpha radiation, thus representing a greater potential for radiolysis effects in ingressed groundwater. The report presents and discusses the results of quantitative determination of the radial profiles of burnup and alpha activity on spent fuel with average burnups from 21.2 to 49 MWd/kgU. In addition to the radial variation of fission product and actinide inventories due to the effects mentioned above, migration, redistribution and release of some fission products can occur during reactor irradiation and the report concludes with a short review of these processes.

  10. 77 FR 51071 - Indiana Michigan Power Company, Donald C. Cook Nuclear Plant, Unit 2, Environmental Assessment...

    Science.gov (United States)

    2012-08-23

    ... reactor core remains intact, the increased burnup may slightly change the mix of fission products that... Manager, Plant Licensing Branch III-1, Division of Operating Reactor Licensing, Office of Nuclear Reactor... From the Federal Register Online via the Government Publishing Office NUCLEAR...

  11. Research on the interfacial behaviors of plate-type dispersion nuclear fuel elements

    Science.gov (United States)

    Wang, Qiming; Yan, Xiaoqing; Ding, Shurong; Huo, Yongzhong

    2010-04-01

    The three-dimensional constitutive relations are constructed, respectively, for the fuel particles, the metal matrix and the cladding of dispersion nuclear fuel elements, allowing for the effects of large deformation and thermal-elastoplasticity. According to the constitutive relations, the method of modeling their irradiation behaviors in ABAQUS is developed and validated. Numerical simulations of the interfacial performances between the fuel meat and the cladding are implemented with the developed finite element models for different micro-structures of the fuel meat. The research results indicate that: (1) the interfacial tensile stresses and shear stresses for some cases will increase with burnup, but the relative stresses will decrease with burnup for some micro-structures; (2) at the lower burnups, the interfacial stresses increase with the particle sizes and the particle volume fractions; however, it is not the case at the higher burnups; (3) the particle distribution characteristics distinctly affect the interfacial stresses, and the face-centered cubic case has the best interfacial performance of the three considered cases.

  12. Noble gas atmospheric monitoring at reprocessing facilities

    Energy Technology Data Exchange (ETDEWEB)

    Nakhleh, C.W.; Perry, R.T. Jr.; Poths, J.; Stanbro, W.D.; Wilson, W.B.; Fearey, B.L.

    1997-05-01

    The discovery in Iraq after the Gulf War of the existence of a large clandestine nuclear-weapon program has led to an across-the-board international effort, dubbed Programme 93+2, to improve the effectiveness and efficiency of International Atomic Energy Agency (IAEA) safeguards. One particularly significant potential change is the introduction of environmental monitoring (EM) techniques as an adjunct to traditional safeguards methods. Monitoring of stable noble gas (Kr, Xe) isotopic abundances at reprocessing plant stacks appears to be able to yield information on the burnup and type of the fuel being processed. To estimate the size of these signals, model calculations of the production of stable Kr, Xe nuclides in reactor fuel and the subsequent dilution of these nuclides in the plant stack are carried out for two case studies: reprocessing of PWR fuel with a burnup of 35 GWd/tU, and reprocessing of CAND fuel with a burnup of 1 GWd/tU. For each case, a maximum-likelihood analysis is used to determine the fuel burnup and type from the isotopic data.

  13. Application of ex-vessel neutron dosimetry combined with in-core measurements for correction of neutron source used for RPV fluence calculations

    Energy Technology Data Exchange (ETDEWEB)

    Borodkin, P.G.; Borodkin, G.I.; Khrennikov, N.N. [Scientific and Engineering Centre for Nuclear and Radiation Safety SEC NRS, Malaya Krasnoselskaya ul., 2/8, Bld. 5, 107140 Moscow (Russian Federation); Konheiser, J. [Helmholz Zentrum Dresden-Rossendorf HZDR, Postfach 510119, D-01314 Dresden (Germany)

    2011-07-01

    This paper deals with calculated and semi-analytical evaluations of VVER-1000 reactor core neutron source distributions and their influence on measurements and calculations of the integral through-vessel neutron leakage. Neutron activation measurements analyzed in the paper were carried out in an ex-vessel air cavity at different nuclear power plant units with VVER-1000 during different fuel cycles. The time-integrated neutron source distributions used for DORT calculations were prepared via two different approaches based on (a) calculated fuel burnup (standard routine procedure) and (b) in-core measurements by means of self-powered detectors (SPDs) and thermocouples (TCs) (new approach). Considering that fuel burnup distributions in operating VVER may be evaluated now by the use of analytical methods (calculations) only, it is necessary to develop new approaches for the testing and correction of calculated evaluations of a neutron source. The results presented in this paper allow one to consider the reverse task of the alternative estimation of fuel burnup distributions. The proposed approach is based on the adjustment (fitting) of time-integrated neutron source distributions, and thus fuel burnup patterns, in some part of the reactor core, taking into account neutron leakage measurements, neutron-physical calculations, and in-core SPD and TC measurement data. (authors)

  14. Fuel element concept for long life high power nuclear reactors

    Science.gov (United States)

    Mcdonald, G. E.; Rom, F. E.

    1969-01-01

    Nuclear reactor fuel elements have burnups that are an order of magnitude higher than can currently be achieved by conventional design practice. Elements have greater time integrated power producing capacity per unit volume. Element design concept capitalizes on known design principles and observed behavior of nuclear fuel.

  15. A Multi-Layered Ceramic Composite for Impermeable Fuel Cladding for COmmercial Wate Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Feinroth, Herbert

    2008-03-03

    A triplex nuclear fuel cladding is developed to further improve the passive safety of commercial nuclear plants, to increase the burnup and durablity of nuclear fuel, to improve the power density and economics of nuclear power, and to reduce the amount of spent fuel requiring disposal or recycle.

  16. Conditioning the γ spectrometer for activity measurement at very high background

    Institute of Scientific and Technical Information of China (English)

    YAN Wei-Hua; ZHANG Li-Guo; ZHANG Zhao; XIAO Zhi-Gang

    2012-01-01

    The application of a high purity Germanium (HPGe) γ spectrometer in determining the fuel element burnup in a future reactor is studied.The HPGe detector is exposed by a 60Co source with varying irradiation rate from 10× 103 s-1 to 150× 103 s-1 to simulate the input counting rate in real reactor environment.A 137Cs and a 152Eu source are positioned at given distances to generate a certain event rate in the detector with the former being proposed as a labeling nuclide to measure the burnup of a fuel element.It is shown that both the energy resolution slightly increasing with the irradiation rate and the passthrough rate at high irradiation level match the requirement of the real application.The influence of the background is studied in the different parameter sets used in the specially developed procedure of background subtraction.It is demonstrated that with the typical input irradiation rate and 137Cs intensity relevant to a deep burnup situation,the precision of the 137Cs counting rate in the current experiment is consistently below 2.8%,indicating a promising feasibility of utilizing an HPGe detector in the burnup measurement in future bed-like reactors.

  17. Impact of statistical uncertainty of the neutron spectrum in the isotopic evolution of fuel; Impacto de la incertidumbre estadistica del espectro neutronico en la evoluacion isotopica del combustible

    Energy Technology Data Exchange (ETDEWEB)

    Ortega, P.

    2012-07-01

    The results obtained and presented in this study for different calculation conditions (number of stories, number of steps burning, etc.) and their simultaneous impact on neutron spectrum and isotopic composition and a methodology is proposed to determine the minimum parameters for calculation given uncertainty in the results of isotopic composition with high burnup, both UO{sub 2} and MOX fuel.

  18. High-temperature materials program. Progress report No. 34, Part B

    Energy Technology Data Exchange (ETDEWEB)

    1964-04-15

    Information is presented concerning coated fuel particle development and evaluation; burnup capabilities of Y/sub 2/O/sub 3/-stabilized UO/sub 2/ and stabilized fuel - BeO matrix materials; high-temperature studies of substoichiometric urania and urania solid solutions; high-temperature carbides and borides research; and high-flux reactor materials gaseous fuels research.

  19. Fuel performance annual report for 1990. Volume 8

    Energy Technology Data Exchange (ETDEWEB)

    Preble, E.A.; Painter, C.L.; Alvis, J.A.; Berting, F.M.; Beyer, C.E.; Payne, G.A. [Pacific Northwest Lab., Richland, WA (United States); Wu, S.L. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Technology

    1993-11-01

    This annual report, the thirteenth in a series, provides a brief description of fuel performance during 1990 in commercial nuclear power plants. Brief summaries of fuel design changes, fuel surveillance programs, fuel operating experience and trends, fuel problems high-burnup fuel experience, and items of general significance are provided . References to additional, more detailed information, and related NRC evaluations are included where appropriate.

  20. 10 CFR 766.3 - Definitions.

    Science.gov (United States)

    2010-01-01

    ... defined as follows: CPI-U means the Consumer Price Index for all-urban consumers published by the... consumers. DOE means the United States Department of Energy and its predecessor agencies. Domestic utility... invoicing and historical tracking of SWU deliveries. Use and burnup charges mean lease charges for...

  1. Sequestration of radioactive iodine in silver-palladium phases in commercial spent nuclear fuel

    Science.gov (United States)

    Buck, Edgar C.; Mausolf, Edward J.; McNamara, Bruce K.; Soderquist, Chuck Z.; Schwantes, Jon M.

    2016-12-01

    Radioactive iodine is the Achilles' heel in the design for the safe geological disposal of spent uranium oxide (UO2) nuclear fuel. Furthermore, iodine's high volatility and aqueous solubility were mainly responsible for the high early doses released during the accident at Fukushima Daiichi in 2011. Studies Kienzler et al., however, have indicated that the instant release fraction (IRF) of radioiodine (131/129I) does not correlate directly with increasing fuel burn-up. In fact, there is a peak in the release of iodine at around 50-60 MW d/kgU, and with increasing burn-up, the IRF of 131/129I decreases. The reasons for this decrease have not fully been understood. We have performed microscopic analysis of chemically processed high burn-up UO2 fuel (80 MW d/kgU) and have found recalcitrant nano-particles containing, Pd, Ag, I, and Br, possibly consistent with a high pressure phase of silver iodide in the undissolved residue. It is likely that increased levels of Ag and Pd from 239Pu fission in high burnup fuels leads to the formation of these metal halides. The occurrence of these phases in UO2 nuclear fuels may reduce the impact of long-lived 129I on the repository performance assessment calculations.

  2. Introduction to Reactor Statics Modules, RS-1. Nuclear Engineering Computer Modules.

    Science.gov (United States)

    Edlund, Milton C.

    The nine Reactor Statics Modules are designed to introduce students to the use of numerical methods and digital computers for calculation of neutron flux distributions in space and energy which are needed to calculate criticality, power distribution, and fuel burn-up for both slow neutron and fast neutron fission reactors. The diffusion…

  3. Development of X-ray Fluorescence Spectrometer for Determination of Tc

    Institute of Scientific and Technical Information of China (English)

    SONG; You; ZHENG; Wei-ming; LIU; Gui-jiao; CHANG; Zhi-yuan

    2012-01-01

    <正>With the deepening of the spent nuclear fuel burnup in power reactor, one of the important fission products, technetium content also increases significantly in the spent fuel. Due to its complex behavior during uranium and plutonium separation, technetium distribution must be monitored in process of spent fuel reprocessing.

  4. Comparison of Leaching Rates of Glass-Ceramic and Glass

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    <正>With the increase of the burn-up of the nuclear fuel, the amounts of the long-lived radionuclides increase. The solubility of actinides such as plutonium in glass is very limited. Glass-ceramic as the new

  5. Coupling between the differential and perturbation theory methods for calculating sensitivity coefficients in nuclear transmutation problems; Acoplamento entre os metodos diferencial e da teoria da perturbacao para o calculo dos coeficientes de sensibilidade em problemas de transmutacao nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Rossi, Lubianka Ferrari Russo

    2014-07-01

    The main target of this study is to introduce a new method for calculating the coefficients of sensibility through the union of differential method and generalized perturbation theory, which are the two methods generally used in reactor physics to obtain such variables. These two methods, separated, have some issues turning the sensibility coefficients calculation slower or computationally exhaustive. However, putting them together, it is possible to repair these issues and build a new equation for the coefficient of sensibility. The method introduced in this study was applied in a PWR reactor, where it was performed the sensibility analysis for the production and {sup 239}Pu conversion rate during 120 days (1 cycle) of burnup. The computational code used for both burnup and sensibility analysis, the CINEW, was developed in this study and all the results were compared with codes widely used in reactor physics, such as CINDER and SERPENT. The new mathematical method for calculating the sensibility coefficients and the code CINEW provide good numerical agility and also good efficiency and security, once the new method, when compared with traditional ones, provide satisfactory results, even when the other methods use different mathematical approaches. The burnup analysis, performed using the code CINEW, was compared with the code CINDER, showing an acceptable variation, though CINDER presents some computational issues due to the period it was built. The originality of this study is the application of such method in problems involving temporal dependence and, not least, the elaboration of the first national code for burnup and sensitivity analysis. (author)

  6. 75 FR 52039 - Biweekly Notice Applications and Amendments to Facility Operating Licenses Involving No...

    Science.gov (United States)

    2010-08-24

    ... application for amendment which is available for public inspection at the Commission's PDR, located at One..., ``Both Isolation Actuation Instrument Channels Inoperable'' (Agencywide Documents Access and Management... restricts the maximum fuel rod average burnup. Deletion of this condition would allow the maximum fuel...

  7. Simulation of irradiation hardening of Zircaloy within plate-type dispersion nuclear fuel elements

    Science.gov (United States)

    Jiang, Yijie; Wang, Qiming; Cui, Yi; Huo, Yongzhong; Ding, Shurong

    2011-06-01

    Within plate-type dispersion nuclear fuel elements, the metal matrix and cladding attacked continuously by fast neutrons undergo irradiation hardening, which might have remarkable effects upon the mechanical behaviors within fuel elements. In this paper, with the irradiation hardening effect of metal materials mainly considered together with irradiation growth effect of the cladding, the three-dimensional large-deformation constitutive relations for the metal matrix and cladding are developed. The method of virtual temperature increase in the previous studies is further developed to model the irradiation swelling of fuel particles; the method of anisotropic thermal expansion is introduced to model irradiation growth of the cladding; and a method of multi-step-temperature loading is proposed to simulate the coupling features of irradiation-induced swelling of the fuel particles together with irradiation growth of the cladding. Above all, based on the developed relationship between irradiation growth at certain burnup and the loaded virtual temperatures, with considering that certain burnup corresponds to certain fast neutron fluence, the time-dependent constitutive relation due to irradiation hardening effect is replaced by the virtual-temperature-dependent one which is introduced into the commercial software to simulate the irradiation hardening effects of the matrix and cladding. Numerical simulations of the irradiation-induced mechanical behaviors are implemented with the finite element method in consideration of the micro-structure of the fuel meat. The obtained results indicate that when the irradiation hardening effects are introduced into the constitutive relations of the metal matrix and cladding: (1) higher maximum Mises stresses for certain burnup at the matrix exist with the equivalent plastic strains remaining almost the same at lower burnups; (2) the maximum Mises stresses for certain burnup at the cladding are enhanced while the maximum equivalent

  8. The study of capability natural uranium as fuel cycle input for long life gas cooled fast reactors with helium as coolant

    Science.gov (United States)

    Ariani, Menik; Satya, Octavianus Cakra; Monado, Fiber; Su'ud, Zaki; Sekimoto, Hiroshi

    2016-03-01

    The objective of the present research is to assess the feasibility design of small long-life Gas Cooled Fast Reactor with helium as coolant. GCFR included in the Generation-IV reactor systems are being developed to provide sustainable energy resources that meet future energy demand in a reliable, safe, and proliferation-resistant manner. This reactor can be operated without enrichment and reprocessing forever, once it starts. To obtain the capability of consuming natural uranium as fuel cycle input modified CANDLE burn-up scheme was adopted in this system with different core design. This study has compared the core with three designs of core reactors with the same thermal power 600 MWth. The fuel composition each design was arranged by divided core into several parts of equal volume axially i.e. 6, 8 and 10 parts related to material burn-up history. The fresh natural uranium is initially put in region 1, after one cycle of 10 years of burn-up it is shifted to region 2 and the region 1 is filled by fresh natural uranium fuel. This concept is basically applied to all regions, i.e. shifted the core of the region (i) into region (i+1) region after the end of 10 years burn-up cycle. The calculation results shows that for the burn-up strategy on "Region-8" and "Region-10" core designs, after the reactors start-up the operation furthermore they only needs natural uranium supply to the next life operation until one period of refueling (10 years).

  9. New European cross-section data libraries for ORIGEN-S based on JEF2.2 and EAF3

    Energy Technology Data Exchange (ETDEWEB)

    Kloosterman, J.L.

    1995-03-01

    Cross-section data libraries for the ORIGEN-S fuel depletion code have been updated with cross sections from the JEF2.2 evaluated nuclear data file and the EAF3 activation file. For both an LWR and a LMFBR, cross-section data for 517 light elements, 65 actinides and 319 fission products have been renewed or added to the libraries (in total cross sections of 604 different nuclides). The nuclides for which the cross sections have been updated were subdivided in two groups. One group of nuclides for which resonance shielding has been applied, and one group for which cross sections of nuclides at infinite dilution have been used. For the LWR, the first group consisted of 22 actinides and 32 fission products, for the LMFBR this group contained 22 actinides (the same as for the PWR) and 26 fission products. For the LWR, the French PWR-N4 reactor was used as a reference design. Spectrum-integrated cross sections of nuclides in the first group (with corrections for resonance shielding) have been calculated as a function of burnup, and the cross sections at average burnup have been used to update the libraries. Spectrum-integrated cross sections of all other nuclides (without resonance shielding corrections), have been obtained with the neutron spectrum at average burnup, and the few-group cross sections have been used to update the libraries. For the LMFBR, the Superphenix fast reactor was used as a reference design. Also for this reactor, the nuclide cross sections have been calculated as a function of burnup, but the burnup dependence of these cross sections turned out to be very small. The procedures for both nuclide groups were similar to those of the LWR. (orig./GL).

  10. Development of ORIGEN libraries for mixed oxide (MOX) fuel assembly designs

    Energy Technology Data Exchange (ETDEWEB)

    Mertyurek, Ugur, E-mail: mertyureku@ornl.gov; Gauld, Ian C., E-mail: gauldi@ornl.gov

    2016-02-15

    Highlights: • ORIGEN MOX library generation process is described. • SCALE burnup calculations are validated against measured MOX fuel samples from the MALIBU program. • ORIGEN MOX libraries are verified using the OECD Phase IV-B benchmark. • There is good agreement for calculated-to-measured isotopic distributions. - Abstract: ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup. The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. The nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.

  11. Regulatory Technology Development Plan - Sodium Fast Reactor. Mechanistic Source Term - Metal Fuel Radionuclide Release

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David [Argonne National Lab. (ANL), Argonne, IL (United States); Bucknor, Matthew [Argonne National Lab. (ANL), Argonne, IL (United States); Jerden, James [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-02-01

    The development of an accurate and defensible mechanistic source term will be vital for the future licensing efforts of metal fuel, pool-type sodium fast reactors. To assist in the creation of a comprehensive mechanistic source term, the current effort sought to estimate the release fraction of radionuclides from metal fuel pins to the primary sodium coolant during fuel pin failures at a variety of temperature conditions. These release estimates were based on the findings of an extensive literature search, which reviewed past experimentation and reactor fuel damage accidents. Data sources for each radionuclide of interest were reviewed to establish release fractions, along with possible release dependencies, and the corresponding uncertainty levels. Although the current knowledge base is substantial, and radionuclide release fractions were established for the elements deemed important for the determination of offsite consequences following a reactor accident, gaps were found pertaining to several radionuclides. First, there is uncertainty regarding the transport behavior of several radionuclides (iodine, barium, strontium, tellurium, and europium) during metal fuel irradiation to high burnup levels. The migration of these radionuclides within the fuel matrix and bond sodium region can greatly affect their release during pin failure incidents. Post-irradiation examination of existing high burnup metal fuel can likely resolve this knowledge gap. Second, data regarding the radionuclide release from molten high burnup metal fuel in sodium is sparse, which makes the assessment of radionuclide release from fuel melting accidents at high fuel burnup levels difficult. This gap could be addressed through fuel melting experimentation with samples from the existing high burnup metal fuel inventory.

  12. Sustainability of thorium-uranium in pebble-bed fluoride salt-cooled high temperature reactor

    Directory of Open Access Journals (Sweden)

    Zhu Guifeng

    2016-01-01

    Full Text Available Sustainability of thorium fuel in a Pebble-Bed Fluoride salt-cooled High temperature Reactor (PB-FHR is investigated to find the feasible region of high discharge burnup and negative Flibe (2LiF-BeF2 salt Temperature Reactivity Coefficient (TRC. Dispersion fuel or pellet fuel with SiC cladding and SiC matrix is used to replace the tristructural-isotropic (TRISO coated particle system for increasing fuel loading and decreasing excessive moderation. To analyze the neutronic characteristics, an equilibrium calculation method of thorium fuel self-sustainability is developed. We have compared two refueling schemes (mixing flow pattern and directional flow pattern and two kinds of reflector materials (SiC and graphite. This method found that the feasible region of breeding and negative Flibe TRC is between 20 vol% and 62 vol% fuel loading in the fuel. A discharge burnup could be achieved up to about 200 MWd/kgHM. The case with directional flow pattern and SiC reflector showed superior burnup characteristics but the worst radial power peak factor, while the case with mixing flow pattern and SiC reflector, which was the best tradeoff between discharge burnup and radial power peak factor, could provide burnup of 140 MWd/kgHM and about 1.4 radial power peak factor with 50 vol% dispersion fuel. In addition, Flibe salt displays good neutron properties as a coolant of quasi-fast reactors due to the strong 9Be(n,2n reaction and low neutron absorption of 6Li (even at 1000 ppm in fast spectrum. Preliminary thermal hydraulic calculation shows good safety margin. The greatest challenge of this reactor may be the decades irradiation time of the pebble fuel.

  13. Graphical User Interface Software for Gross Defect Detection at the Atucha-I Plant

    Energy Technology Data Exchange (ETDEWEB)

    Wong, A C; Sitaraman, S; Ham, Y S; Peixoto, O

    2012-05-10

    At the Atucha-I pressurized heavy water reactor in Argentina, fuel assemblies in the spent fuel pools are stored by suspending them in two vertically stacked layers. This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Movement of fuel, especially from the lower layer, would involve a major effort on the part of the operator. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 w% {sup 235}U, and has been in operation since 1974, a wide range of burnups and cooling times can exist in any given pool. Additionally, while fuel assemblies are grouped together in a uniform fashion, the packing density from group to group can vary within a single pool. A tool called the Spent Fuel Neutron Counter (SFNC) was developed and successfully tested at the site to verify, in an in-situ condition, the presence of fuel up to burnups of 8,000 MWd/t. Since the neutron source term becomes a nonlinear function of burnup beyond this burnup, a new algorithm was developed to predict expected response from the SFNC at measurement locations covering the entire range of burnups, cooling times, and initial enrichments. With the aid of a static database of parameters including intrinsic sources and energy group-wise detector response functions, as well as explicit spent fuel information including burnups, cooling times, enrichment types, and spacing between fuel assemblies, an expected response for any given location can be calculated by summing the contributions from the relevant neighboring fuel assemblies. Thus, the new algorithm maps the expected responses across the various pools providing inspectors with a visual aid in verifying the presence of the spent fuel assemblies. This algorithm has been fully integrated into a standalone application built in LabVIEW. The GUI uses a step-by-step approach to allow the end-user to first calibrate the predicted database against a set of

  14. 考虑静水压力的 HBS辐照肿胀本构关系开发及验证%Development and Validation of Constitutive Equation of HBS Irradiation Swelling Considering Hydrostatic Pressure

    Institute of Scientific and Technical Information of China (English)

    高利军; 姜胜耀; 陈炳德; 肖忠; 俞冀阳

    2014-01-01

    分析了静水压力对U O2高燃耗结构辐照肿胀产生影响的机理,据此给出了开发辐照肿胀本构关系的3个基本假设,得出静水压力主要通过压缩高燃耗结构气孔体积而影响辐照肿胀这一结论。在已完成开发的高燃耗结构辐照肿胀关系式的基础上,定量计算静水压力引起的气孔收缩和高燃耗结构辐照肿胀减小量,成功地开发了考虑静水压力的高燃耗结构辐照肿胀本构关系。利用现有高燃耗结构辐照肿胀数据对本构关系做了初步验证,验证结果表明了其合理性。%The mechanism of hydrostatic pressure affecting the irradiation swelling of UO2 high burnup structure was analyzed .Three basic assumptions used to develop the constitutive equation of irradiation swelling were made accordingly .It is concluded that hydrostatic pressure imposes an important impact on irradiation swelling mainly through compressing the UO2 high burnup structure pores . Based on the already developed correlation of the irradiation swelling of UO 2 high burnup structure ,pore shrinkage due to the application of hydrostatic pressure and thus the reduction of irradiation swelling of UO2 high burnup structure were determined quantitatively , and the constitutive equation of irradiation swelling of UO2 high burnup structure considering the hydrostatic pressure was constructed successfully . The constitutive equation is validated using available irradiation swelling data of UO2 high burnup structure ,which demonstrates its reasonability .

  15. Status and verification strategy for ITER neutronics

    Energy Technology Data Exchange (ETDEWEB)

    Loughlin, Michael, E-mail: michael.loughlin@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Angelone, Maurizio [Associazione EURATOM-ENEA Sulla Fusione, Via E. Fermi 45, I-00044 Frascati, Roma (Italy); Batistoni, Paola [Associazione EURATOM-ENEA Sulla Fusione, Via E. Fermi 45, I-00044 Frascati, Roma (Italy); JET-EFDA, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Bertalot, Luciano [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Eskhult, Jonas [Studsvik Nuclear AB, SE-611 Nyköping (Sweden); Konno, Chikara [Japan Atomic Energy Agency Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan); Pampin, Raul [F4E Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, Barcelona 08019 (Spain); Polevoi, Alexei; Polunovskiy, Eduard [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2014-10-15

    The paper summarizes the current status of neutronics at ITER and a first set of proposals for experimental programmes to be conducted in the early operational life-time of ITER are described for the more crucial areas. These include a TF coils heating benchmark, a streaming benchmark and streaming measurements by activation on ITER itself. Also on ITER the measurement of activated water from triton burn-up should be planned and performed. This will require the measurement of triton burn-up in DD phase. Measurements of neutron flux in the tokamak building during DD operations should also be carried out. The use of JET for verification of shut down dose rate estimates is desirable. Other facilities to examine the production and behaviour of activated corrosion products and the shielding properties of concretes to high energy (6 MeV) gamma-rays are recommended.

  16. Nuclear data uncertainty analysis on a minor actinide burner for transmuting spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hangbok

    1998-08-01

    A comprehensive sensitivity and uncertainty analysis was performed on a 1200 MWt minor actinides burner designed for a low burnup reactivity swing, negative doppler coefficient, and low sodium void worth. Sensitivities of the performance parameters were generated using depletion perturbation methods for the constrained close fuel cycle of the reactor. The uncertainty analysis was performed using the sensitivity and covariance data taken from ENDF-B/V and other published sources. The uncertainty analysis of a liquid metal reactor for burning minor actinide has shown that uncertainties in the nuclear data of several key minor actinide isotopes can introduce large uncertainties in the predicted performance of the core. The relative uncertainties in the burnup swing, doppler coefficient, and void worth were conservatively estimated to be 180 %, 97 %, and 46 %, respectively. An analysis was performed to prioritize the minor actinide reactions for reducing the uncertainties. (author). 41 refs., 17 tabs., 1 fig.

  17. A Pebble-Bed Breed-and-Burn Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Greenspan, Ehud [Univ. of California, Berkeley, CA (United States)

    2016-03-31

    The primary objective of this project is to use three-dimensional fuel shuffling in order to reduce the minimum peak radiation damage of ~550 dpa present Breed-and-Burn (B&B) fast nuclear reactor cores designs (they feature 2-D fuel shuffling) call for to as close as possible to the presently accepted value of 200 dpa thereby enabling earlier commercialization of B&B reactors which could make substantial contribution to energy sustainability and economic stability without need for fuel recycling. Another objective is increasing the average discharge burnup for the same peak discharge burnup thereby (1) increasing the fuel utilization of 2-D shuffled B&B reactors and (2) reducing the reprocessing capacity required to support a given capacity of FRs that are to recycle fuel.

  18. Special Nuclear Material Gamma-Ray Signatures for Reachback Analysts

    Energy Technology Data Exchange (ETDEWEB)

    Karpius, Peter Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Myers, Steven Charles [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-08-29

    These are slides on special nuclear material gamma-ray signatures for reachback analysts for an LSS Spectroscopy course. The closing thoughts for this presentation are the following: SNM materials have definite spectral signatures that should be readily recognizable to analysts in both bare and shielded configurations. One can estimate burnup of plutonium using certain pairs of peaks that are a few keV apart. In most cases, one cannot reliably estimate uranium enrichment in an analogous way to the estimation of plutonium burnup. The origin of the most intense peaks from some SNM items may be indirect and from ‘associated nuclides.' Indirect SNM signatures sometimes have commonalities with the natural gamma-ray background.

  19. Hybrid fusion-fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors

    Science.gov (United States)

    Shmelev, A. N.; Kulikov, G. G.; Kurnaev, V. A.; Salahutdinov, G. H.; Kulikov, E. G.; Apse, V. A.

    2015-12-01

    Discussions are currently going on as to whether it is suitable to employ thorium in the nuclear fuel cycle. This work demonstrates that the 231Pa-232U-233U-Th composition to be produced in the thorium blanket of a hybrid thermonuclear reactor (HTR) as a fuel for light-water reactors opens up the possibility of achieving high, up to 30% of heavy metals (HM), or even ultrahigh fuel burnup. This is because the above fuel composition is able to stabilize its neutron-multiplying properties in the process of high fuel burnup. In addition, it allows the nuclear fuel cycle (NFC) to be better protected against unauthorized proliferation of fissile materials owing to an unprecedentedly large fraction of 232U (several percent!) in the uranium bred from the Th blanket, which will substantially hamper the use of fissile materials in a closed NFC for purposes other than power production.

  20. Strategy for Fuel Rod Receipt, Characterization, Sample Allocation for the Demonstration Sister Rods

    Energy Technology Data Exchange (ETDEWEB)

    Marschman, Steven C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Warmann, Stephan A. [Portage, Inc., Idaho Falls, ID (United States); Rusch, Chris [NAC International, Inc., Norcross, GA (United States)

    2014-03-01

    The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technology, has established the Used Fuel Disposition Campaign (UFDC) to conduct the research and development activities related to storage, transportation, and disposal of used nuclear fuel and high-level radioactive waste. The mission of the UFDC is to identify alternatives and conduct scientific research and technology development to enable storage, transportation and disposal of used nuclear fuel (UNF) and wastes generated by existing and future nuclear fuel cycles. The UFDC Storage and Transportation staffs are responsible for addressing issues regarding the extended or long-term storage of UNF and its subsequent transportation. The near-term objectives of the Storage and Transportation task are to use a science-based approach to develop the technical bases to support the continued safe and secure storage of UNF for extended periods, subsequent retrieval, and transportation. While low burnup fuel [that characterized as having a burnup of less than 45 gigawatt days per metric tonne uranium (GWD/MTU)] has been stored for nearly three decades, the storage of high burnup used fuels is more recent. The DOE has funded a demonstration project to confirm the behavior of used high burnup fuel under prototypic conditions. The Electric Power Research Institute (EPRI) is leading a project team to develop and implement the Test Plan to collect this data from a UNF dry storage system containing high burnup fuel. The Draft Test Plan for the demonstration outlines the data to be collected; the high burnup fuel to be included; the technical data gaps the data will address; and the storage system design, procedures, and licensing necessary to implement the Test Plan. To provide data that is most relevant to high burnup fuel in dry storage, the design of the test storage system must closely mimic real conditions high burnup SNF experiences during all stages of dry storage: loading, cask drying

  1. Effect of indium addition in U-Zr metallic fuel on lanthanide migration

    Science.gov (United States)

    Kim, Yeon Soo; Wiencek, T.; O'Hare, E.; Fortner, J.; Wright, A.; Cheon, J. S.; Lee, B. O.

    2017-02-01

    Advanced fast reactor concepts to achieve ultra-high burnup (∼50%) require prevention of fuel-cladding chemical interaction (FCCI). Fission product lanthanide accumulation at high burnup is substantial and significantly contributes to FCCI upon migration to the cladding interface. Diffusion barriers are typically used to prevent interaction of the lanthanides with the cladding. A more active method has been proposed which immobilizes the lanthanides through formation of stable compounds with an additive. Theoretical analysis showed that indium, thallium, and antimony are good candidates. Indium was the strongest candidate because of its low reactivity with iron-based cladding alloys. Characterization of the as-fabricated alloys was performed to determine the effectiveness of the indium addition in forming compounds with lanthanides, represented by cerium. Tests to examine how effectively the dopant prevents lanthanide migration under a thermal gradient were also performed. The results showed that indium effectively prevented cerium migration.

  2. Characterization of Nuclear Fuel using Multivariate Statistical Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Robel, M; Robel, M; Robel, M; Kristo, M J; Kristo, M J

    2007-11-27

    Various combinations of reactor type and fuel composition have been characterized using principle components analysis (PCA) of the concentrations of 9 U and Pu isotopes in the 10 fuel as a function of burnup. The use of PCA allows the reduction of the 9-dimensional data (isotopic concentrations) into a 3-dimensional approximation, giving a visual representation of the changes in nuclear fuel composition with burnup. Real-world variation in the concentrations of {sup 234}U and {sup 236}U in the fresh (unirradiated) fuel was accounted for. The effects of reprocessing were also simulated. The results suggest that, 15 even after reprocessing, Pu isotopes can be used to determine both the type of reactor and the initial fuel composition with good discrimination. Finally, partial least squares discriminant analysis (PSLDA) was investigated as a substitute for PCA. Our results suggest that PLSDA is a better tool for this application where separation between known classes is most important.

  3. Hybrid fusion–fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shmelev, A. N., E-mail: shmelan@mail.ru; Kulikov, G. G., E-mail: ggkulikov@mephi.ru; Kurnaev, V. A., E-mail: kurnaev@yandex.ru; Salahutdinov, G. H., E-mail: saip07@mail.ru; Kulikov, E. G., E-mail: egkulikov@mephi.ru; Apse, V. A., E-mail: apseva@mail.ru [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute) (Russian Federation)

    2015-12-15

    Discussions are currently going on as to whether it is suitable to employ thorium in the nuclear fuel cycle. This work demonstrates that the {sup 231}Pa–{sup 232}U–{sup 233}U–Th composition to be produced in the thorium blanket of a hybrid thermonuclear reactor (HTR) as a fuel for light-water reactors opens up the possibility of achieving high, up to 30% of heavy metals (HM), or even ultrahigh fuel burnup. This is because the above fuel composition is able to stabilize its neutron-multiplying properties in the process of high fuel burnup. In addition, it allows the nuclear fuel cycle (NFC) to be better protected against unauthorized proliferation of fissile materials owing to an unprecedentedly large fraction of {sup 232}U (several percent!) in the uranium bred from the Th blanket, which will substantially hamper the use of fissile materials in a closed NFC for purposes other than power production.

  4. Fission gas release behaviour of a 103 GWd/tHM fuel disc during a 1200 °C annealing test

    Science.gov (United States)

    Noirot, J.; Pontillon, Y.; Yagnik, S.; Turnbull, J. A.; Tverberg, T.

    2014-03-01

    Within the Nuclear Fuel Industry Research (NFIR) program, several fuel variants, in the form of thin circular discs, were irradiated in the Halden Boiling Water Reactor (HBWR) to a range of burn-ups ˜100 GWd/tHM. The design of the assembly was similar to that used in other HBWR programs: the assembly contained several rods with fuel discs sandwiched between Mo discs, which limited temperature gradients within the fuel discs. One such rod contained standard grain UO2 discs (3D grain size = 18 μm) reaching a burn-up of 103 GWd/tHM. After the irradiation, the gas release upon rod puncturing was measured to be 2.9%.

  5. Sensitivity of Candu-Scwr Reactor Physics Calculations to Nuclear Data Files

    Science.gov (United States)

    Kozier, K. S.; Dyck, G. R.

    2006-04-01

    A simplified MCNP model of a CANDU-SCWR lattice was used to test the sensitivity of the calculated reactivity to various nuclear data files involving issues of current interest. These tests were performed for cooled and voided conditions, with and without axial neutron leakage, for a uniform lattice of mid-life fuel and a mixed lattice with high-burnup and low-burnup fuel in alternate channels. Results were compared using different room-temperature data files for deuterium, various thermal-scattering-law data files for hydrogen bound in light water and deuterium bound in heavy water, and for pre-ENDF/B-VII and ENDF/B-VI.8 data for uranium. The reactivity differences observed were small (typically <1 mk) and increased with axial neutron leakage.

  6. Testing of uranium nitride fuel in T-111 cladding at 1200 K cladding temperature

    Science.gov (United States)

    Rohal, R. G.; Tambling, T. N.; Smith, R. L.

    1973-01-01

    Two groups of six fuel pins each were assembled, encapsulated, and irradiated in the Plum Brook Reactor. The fuel pins employed uranium mononitride (UN) in a tantalum alloy clad. The first group of fuel pins was irradiated for 1500 hours to a maximum burnup of 0.7-atom-percent uranium. The second group of fuel pins was irradiated for about 3000 hours to a maximum burnup of 1.0-atom-percent uranium. The average clad surface temperature during irradiation of both groups of fuel pins was approximately 1200 K. The postirradiation examination revealed the following: no clad failures or fuel swelling occurred; less than 1 percent of the fission gases escaped from the fuel; and the clad of the first group of fuel pins experienced clad embrittlement whereas the second group, which had modified assembly and fabrication procedures to minimize contamination, had a ductile clad after irradiation.

  7. Standard test method for atom percent fission in uranium and plutonium fuel (Neodymium-148 Method)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    1996-01-01

    1.1 This test method covers the determination of stable fission product 148Nd in irradiated uranium (U) fuel (with initial plutonium (Pu) content from 0 to 50 %) as a measure of fuel burnup (1-3). 1.2 It is possible to obtain additional information about the uranium and plutonium concentrations and isotopic abundances on the same sample taken for burnup analysis. If this additional information is desired, it can be obtained by precisely measuring the spike and sample volumes and following the instructions in Test Method E267. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  8. Mol 7C/6; Mol 7C/6

    Energy Technology Data Exchange (ETDEWEB)

    Aberle, J.; Schleisiek, K.; Schmuck, I.; Schmidt, L.; Romer, O.; Weih, G.

    1995-08-01

    The Mol 7C/6 coolant blockage experiment in the Belgian BR2 reactor yielded results different from Mol 7C experiments with low burnup pins: At 10% burnup local failure is not self-limiting, but requires active systems for detection and scram. The Mol 7C series was finished in 1991. In each of the test bundles Mol 7C/4, /5 and /6, 30 Mk I pins pre-irradiated in KNK II were used. The central blockage consisted of enriched UO{sub 2} covering 30 percent of the bundle cross-section, with a height of 40 mm. The most important system for timely detection of coolant blockages of the type studied in Mol 7C/6 is based on DND. (orig.)

  9. Severe-damage experiments in irradiated mixed-oxide subassemblies

    Energy Technology Data Exchange (ETDEWEB)

    Schleisiek, K.; Weimar, P.; Doerr, L.; Romer, O. (Kernforschungsanlage Karlsruhe (Germany))

    1992-01-01

    Severe-damage experiments (Mol-7C series) have been performed in the BR-2 reactor at Mol, Belgium. The main objective was to investigate the consequences of local cooling disturbances in liquid-metal-cooled reactor fuel subassemblies. The cooling deficit was generated by a central blockage. During irradiation, this blockage caused significant local damage, including melting of cladding and fuel. As the most important parameter, the burnup of the tested bundles was changed. According to Table I, the burnups varied from 1.7 to 5 and 10 at.%. In this paper, preliminary results of the postirradiation examination (PIE) of Mol-7C/6 are presented and discussed in comparison to the previous experiments.

  10. The uncertainty analysis of a liquid metal reactor for burning minor actinides from light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    The neutronics analysis of a liquid metal reactor for burning minor actinides has shown that uncertainties in the nuclear data of several key minor actinide isotopes can introduce large uncertainties in the predicted performance of the core. A comprehensive sensitivity and uncertainty analysis was performed on a 1200 MWth actinide burner designed for a low burnup reactivity swing, negative doppler coefficient, and low sodium void worth. Sensitivities were generated using depletion perturbation methods for the equilibrium cycle of the reactor and covariance data was taken ENDF-B/V and other published sources. The relative uncertainties in the burnup swing, doppler coefficient, and void worth were conservatively estimated to be 180%, 97%, and 46%, respectively. 5 refs., 1 fig., 3 tabs. (Author)

  11. New Monte Carlo-based method to evaluate fission fraction uncertainties for the reactor antineutrino experiment

    Science.gov (United States)

    Ma, X. B.; Qiu, R. M.; Chen, Y. X.

    2017-02-01

    Uncertainties regarding fission fractions are essential in understanding antineutrino flux predictions in reactor antineutrino experiments. A new Monte Carlo-based method to evaluate the covariance coefficients between isotopes is proposed. The covariance coefficients are found to vary with reactor burnup and may change from positive to negative because of balance effects in fissioning. For example, between 235U and 239Pu, the covariance coefficient changes from 0.15 to -0.13. Using the equation relating fission fraction and atomic density, consistent uncertainties in the fission fraction and covariance matrix were obtained. The antineutrino flux uncertainty is 0.55%, which does not vary with reactor burnup. The new value is about 8.3% smaller.

  12. The Halden reactor, a facility open to the international nuclear community; Halden, un reacteur ouvert a la communaute internationale

    Energy Technology Data Exchange (ETDEWEB)

    Vitanza, C. [Organisation for Economic Co-Operation and Development, Nuclear Energy Agency (OECD/NEA), 75 - Paris (France)

    2005-07-01

    The Halden test reactor is a boiling-type reactor moderated and cooled by heavy water, it yields a thermal power of 20 MW. The reactor operates for 2 periods of about 100 days each year. The Halden reactor has been in operation for more than 45 years and is the largest OECD-NEA project, it carries out the OECD joint program and bilateral contract work. Its experimental programs are supported by about 100 organisations in 20 countries. The fuel and materials programs for the years to come focus on the following main areas: -) fuel high burn-up capabilities in normal operating conditions, -) fuel response to transients aiming at generating experimental data on the behaviour of high burn-up fuels in short duration transients and on phenomena occurring during loss of coolant accident and coolant flow oscillations, -) cladding corrosion and water chemistry issues, and -) pressure vessel embrittlement and irradiation assisted stress corrosion cracking of reactor internals. (A.C.)

  13. Peach Bottom test element program. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Saurwein, J.J.; Holzgraf, J.F.; MIller, C.M.; Myers, B.F.; Wallroth, C.F.

    1982-11-01

    Thirty-three test elements were irradiated in the Peach Bottom high-temperature gas-cooled reactor (HTGR) as part of the testing program for advanced HTGRs. Extensive postirradiation examinations and evaluations of 21 of these irradiation experiments were performed. The test element irradiations were simulated using HTGR design codes and data. Calculated fuel burnups, power profiles, fast neutron fluences, and temperatures were verified via destructive burnup measurements, gamma scanning, and in-pile thermocouple readings corrected for decalibration effects. Analytical techniques were developed to improve the quality of temperature predictions through feedback of nuclear measurements into thermal calculations. Dimensional measurements, pressure burst tests, diametral compression tests, ring-cutting tests, strip-cutting tests, and four-point bend tests were performed to measure residual stress, strain, and strength distributions in H-327 graphite structures irradiated in the test elements.

  14. Development and verification of NRC`s single-rod fuel performance codes FRAPCON-3 AND FRAPTRAN

    Energy Technology Data Exchange (ETDEWEB)

    Beyer, C.E.; Cunningham, M.E.; Lanning, D.D. [Pacific Northwest National Lab., Richland, WA (United States)

    1998-03-01

    The FRAPCON and FRAP-T code series, developed in the 1970s and early 1980s, are used by the US Nuclear Regulatory Commission (NRC) to predict fuel performance during steady-state and transient power conditions, respectively. Both code series are now being updated by Pacific Northwest National Laboratory to improve their predictive capabilities at high burnup levels. The newest versions of the codes are called FRAPCON-3 and FRAPTRAN. The updates to fuel property and behavior models are focusing on providing best estimate predictions under steady-state and fast transient power conditions up to extended fuel burnups (> 55 GWd/MTU). Both codes will be assessed against a data base independent of the data base used for code benchmarking and an estimate of code predictive uncertainties will be made based on comparisons to the benchmark and independent data bases.

  15. ASSESSMENT OF THE RADIONUCLIDE COMPOSITION OF "HOT PARTICLES" SAMPLED IN THE CHERNOBYL NUCLEAR POWER PLANT FOURTH REACTOR UNIT

    Energy Technology Data Exchange (ETDEWEB)

    Farfan, E.; Jannik, T.; Marra, J.

    2011-10-01

    Fuel-containing materials sampled from within the Chernobyl Nuclear Power Plant (ChNPP) 4th Reactor Unit Confinement Shelter were spectroscopically studied for gamma and alpha content. Isotopic ratios for cesium, europium, plutonium, americium, and curium were identified and the fuel burnup in these samples was determined. A systematic deviation in the burnup values based on the cesium isotopes, in comparison with other radionuclides, was observed. The conducted studies were the first ever performed to demonstrate the presence of significant quantities of {sup 242}Cm and {sup 243}Cm. It was determined that there was a systematic underestimation of activities of transuranic radionuclides in fuel samples from inside of the ChNPP Confinement Shelter, starting from {sup 241}Am (and going higher), in comparison with the theoretical calculations.

  16. Preliminary study of the tight lattice pressured heavy water reactor loaded with Pu/U and Th/U mixed fuels

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    To improve nuclear fuel utilization efficiency and prolong fuel cycle burn-up, a tight pitch lattice pressured heavy water reactor was investigated as an alternative of next generation of power reactors. It is shown that the high conversion ratio and negative coolant void reactivity coefficient are challenges in the reactor core physics designs. Various techniques were proposed to solve these problems. In this work, a tight pitch lattice and mixed fuel assemblies pressured heavy water reactor concept was investigated. By utilizing numerical simulation technique, it is demonstrated that reactor core mixed with Pu/U and Th/U assemblies can achieve high conversion ratio (0.98), long burn-up (60 GWD/t) and negative void reactivity coefficients.

  17. Swelling behavior detection of irradiated U-10Zr alloy fuel using indirect neutron radiography

    Science.gov (United States)

    Sun, Yong; Huo, He-yong; Wu, Yang; Li, Jiangbo; Zhou, Wei; Guo, Hai-bing; Li, Hang; Cao, Chao; Yin, Wei; Wang, Sheng; Liu, Bin; Feng, Qi-jie; Tang, Bin

    2016-11-01

    It is hopeful that fusion-fission hybrid energy system will become an effective approach to achieve long-term sustainable development of fission energy. U-10Zr alloy (which means the mass ratio of Zr is 10%) fuel is the key material of subcritical blanket for fusion-fission hybrid energy system which the irradiation performance need to be considered. Indirect neutron radiography is used to detect the irradiated U-10Zr alloy because of the high residual dose in this paper. Different burnup samples (0.1%, 0.3%, 0.5% and 0.7%) have been tested with a special indirect neutron radiography device at CMRR (China Mianyang Research Reactor). The resolution of the device is better than 50 μm and the quantitative analysis of swelling behaviors was carried out. The results show that the swelling behaviors relate well to burnup character which can be detected accurately by indirect neutron radiography.

  18. Fuel performance improvement program. Quarterly/annual progress report, October 1978-September 1979

    Energy Technology Data Exchange (ETDEWEB)

    Crouthamel, C.E. (comp.)

    1979-10-01

    The objective of the Fuel Performance Improvement Program is to develop, test, and demonstrate basically two advanced fuel designs with the capability for improved power ramping performance and thus increase the capability of achieving extended burnup levels to better utilize uranium resources. The irradiations are being supported by out-of-reactor experiments to evaluate the effect of graphite coatings to inhibit stress-corrosion-cracking type cladding failures that are related to pellet-cladding interaction. Instrumented test irradiations in the Halden Boiling Water Reactor (HBWR) have achieved peak burnups of 697 GJ/kgM (8.1 MWd/kgM) with reference, annular-coated-pressurized, and sphere-pac rods.

  19. M3FT-15OR0202212: SUBMIT SUMMARY REPORT ON THERMODYNAMIC EXPERIMENT AND MODELING

    Energy Technology Data Exchange (ETDEWEB)

    McMurray, Jake W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brese, Robert G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Silva, Chinthaka M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Besmann, Theodore M. [Univ. of South Carolina, Columbia, SC (United States)

    2015-09-01

    Modeling the behavior of nuclear fuel with a physics-based approach uses thermodynamics for key inputs such as chemical potentials and thermal properties for phase transformation, microstructure evolution, and continuum transport simulations. Many of the lanthanide (Ln) elements and Y are high-yield fission products. The U-Y-O and U-Ln-O ternaries are therefore key subsystems of multi-component high-burnup fuel. These elements dissolve in the dominant urania fluorite phase affecting many of its properties. This work reports on an effort to assess the thermodynamics of the U-Pr-O and U-Y-O systems using the CALPHAD (CALculation of PHase Diagrams) method. The models developed within this framework are capable of being combined and extended to include additional actinides and fission products allowing calculation of the phase equilibria, thermochemical and material properties of multicomponent fuel with burnup.

  20. A new MC-based method to evaluate the fission fraction uncertainty at reactor neutrino experiment

    CERN Document Server

    Ma, X B; Chen, Y X

    2016-01-01

    Uncertainties of fission fraction is an important uncertainty source for the antineutrino flux prediction in a reactor antineutrino experiment. A new MC-based method of evaluating the covariance coefficients between isotopes was proposed. It was found that the covariance coefficients will varying with reactor burnup and which may change from positive to negative because of fissioning balance effect, for example, the covariance coefficient between $^{235}$U and $^{239}$Pu changes from 0.15 to -0.13. Using the equation between fission fraction and atomic density, the consistent of uncertainty of fission fraction and the covariance matrix were obtained. The antineutrino flux uncertainty is 0.55\\% which does not vary with reactor burnup, and the new value is about 8.3\\% smaller.

  1. Impact of nuclear library difference on neutronic characteristics of thorium-loaded light water reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Unesaki, H. [Research Reactor Inst., Kyoto Univ., Asashiro-Nishi 2, Kumatori-cho, Sennan-gun, Osaka 590-0494 (Japan); Dept. of Socio-Environmental Energy Science, Graduate School of Energy Science, Kyoto Univ., Yoshida-Honmachi, Sakyo-ku, Kyoto 606-8501 (Japan); Isaka, S. [Dept. of Socio-Environmental Energy Science, Graduate School of Energy Science, Kyoto Univ., Yoshida-Honmachi, Sakyo-ku, Kyoto 606-8501 (Japan); Nakagome, Y. [Research Reactor Inst., Kyoto Univ., Asashiro-Nishi 2, Kumatori-cho, Sennan-gun, Osaka 590-0494 (Japan); Dept. of Socio-Environmental Energy Science, Graduate School of Energy Science, Kyoto Univ., Yoshida-Honmachi, Sakyo-ku, Kyoto 606-8501 (Japan)

    2006-07-01

    Impact of nuclear library difference on neutronic characteristics of thorium-loaded light water reactor fuel is investigated through cell burnup calculations using SRAC code system. Comparison of k{sub {infinity}} and nuclide composition was made between the results obtained by JENDL-3.3, ENDF/B-VI.8 and JEFF3.0 for (U, Th)O{sub 2} fuels as well as UO{sub 2} fuels, with special interest on the burnup dependence of the neutronic characteristics. The impact of nuclear data library difference on k{sub {infinity}} of (U, Th)O{sub 2} fuels was found to be significantly large compared to that of UO{sub 2} fuels. Notable difference was also found in nuclide concentration of TRU nuclides. (authors)

  2. Evaluation and demonstration of methods for improved fuel utilization. Fort Calhoun Poolside Inspection Programs: end-of-cycles 4 and 5

    Energy Technology Data Exchange (ETDEWEB)

    LaVake, J.C.; Smith, G.P.

    1980-08-01

    The purpose of this program is to demonstrate two techniques for improving fuel utilization in current Pressurized Water Reactors (PWR): (1) more efficient fuel management, and (2) high burnup. These improvements are being demonstrated in the Fort Calhoun reactor, a 1420 Mwt PWR. A more efficient fuel management scheme called SAVFUEL (Shimmed And Very Flexible Uranium Element Loading), will be demonstrated which reduces neutron leakage and is expected to reduce uranium requirements. Another part of the demonstration program will increase the fuel burnup of the current 14 x 14 fuel design to further reduce uranium requirements. The program is a joint Department of Energy (DOE), Omaha Public Power District (OPPD) and Combustion Engineering (C-E) endeavor and is part of a national effort to improve uranium utilization in light water reactors.

  3. Gamma spectroscopic examination of Peach Bottom HTGR core components

    Energy Technology Data Exchange (ETDEWEB)

    Holzgraf, J.F.; McCord, F.; Wallroth, C.F.

    1978-04-01

    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, power profile determinations, and fission product release and redistribution studies. The results showed that the predicted and measured burnups had a +-7 percent root mean square deviation on an element-to-element basis and were within +-0.7 percent (1 sigma) on a core average basis. The element-to-element variation of +-7 percent is within the generally stated +-3 percent to 8 percent accuracy for nuclear predictions.

  4. Calculations of 3D full-scale VVER fuel assembly and core models using MCU and BIPR-7A codes

    Energy Technology Data Exchange (ETDEWEB)

    Aleshin, Sergey S.; Bikeev, Artem S.; Bolshagin, Sergey N.; Kalugin, Mikhail A.; Kosourov, Evgeniy K.; Pavlovichev, Aleksandr M.; Pryanichnikov, Aleksandr V.; Sukhino-Khomenko, Evgenia A.; Shcherenko, Anna I.; Shcherenko, Anastasia I.; Shkarovskiy, Denis A. [Nuclear Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

    2015-09-15

    Two types of calculations were made to compare BIPR-7A and MCU results for 3D full-scale models. First EPS (emergency protection system) efficiency and in-core power distributions were analyzed for an equilibrium fuel load of VVER-1000 assuming its operation within an 18-month cycle. Computations were performed without feedbacks and with fuel burnup distributed over the core. After 3D infinite lattices of full-scale VVER-1000 fuel assemblies (A's) with uranium fuel 4.4% enrichment and uranium-erbium fuel 4.4% enrichment and Er{sub 2}O{sub 3} 1 % wt were considered. Computations were performed with feedbacks and fuel burnup at the constant power level. For different time moments effective multiplication factor and power distribution were obtained. EPS efficiency and reactivity effects at chosen time moments were analyzed.

  5. Study on core design for reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Okubo, Tsutomu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    The Reduced-Moderation Water Reactor (RMWR) is a water-cooled reactor with the harder neutron spectrum comparing with the LWR, resulting from low neutron moderation due to reduced water volume fraction. Based on the difference from the spectrum from the LWR, the conversion from U-238 to Pu-239 is promoted and the new cores preferable to effective utilization of uranium resource can be possible Design study of the RMWR core started in 1997 and new four core concepts (three BWR cores and one PWR core) are recently evaluated in terms of control rod worths, plutonium multiple recycle, high burnup and void coefficient. Comparative evaluations show needed incorporation of control rod programming and simplified PUREX process as well as development of new fuel cans for high burnup of 100 GW-d/t. Final choice of design specifications will be made at the next step aiming at realization of the RMWR. (T. Tanaka)

  6. Parametric Analysis of PWR Spent Fuel Depletion Parameters for Long-Term-Disposal Criticality Safety

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    1999-08-01

    Utilization of burnup credit in criticality safety analysis for long-term disposal of spent nuclear fuel allows improved design efficiency and reduced cost due to the large mass of fissile material that will be present in the repository. Burnup-credit calculations are based on depletion calculations that provide a conservative estimate of spent fuel contents (in terms of criticality potential), followed by criticality calculations to assess the value of the effective neutron multiplication factor (k(sub)eff) for the a spent fuel cask or a fuel configuration under a variety of probabilistically derived events. In order to ensure that the depletion calculation is conservative, it is necessary to both qualify and quantify assumptions that can be made in depletion models.

  7. Measurements of Fission Cross Sections for the Isotopes relevant to the Thorium Fuel Cycle

    CERN Multimedia

    2002-01-01

    The present concern about a sustainable energy supply is characterised by a considerable uncertainty: the green house effect and foreseeable limits in fossil fuel resources on the one hand, the concern about the environmental impact of nuclear fission energy and the long term fusion research on the other hand, have led to the consideration of a variety of advanced strategies for the nuclear fuel cycle and related nuclear energy systems. The present research directories concern such strategies as the extension of the life span of presently operating reactors, the increase of the fuel burn-up, the plutonium recycling, and in particular the incineration of actinides and long-Lived fission products, the accelerator driven systems (ADS), like the "Energy Amplifier" (EA) concept of C. Rubbia, and the possible use of the Thorium fuel cycle. The detailed feasibility study and safety assessment of these strategies requires the accurate knowledge of neutron nuclear reaction data. Both, higher fuel burn-up and especiall...

  8. Damage production and accumulation in SiC structures in inertial and magnetic fusion systems

    Science.gov (United States)

    Sawan, M. E.; Ghoniem, N. M.; Snead, L.; Katoh, Y.

    2011-10-01

    Radiation damage parameters in SiC/SiC composite structures are determined in both magnetic (MFE) and inertial (IFE) confinement fusion systems. Variations in the geometry, neutron energy spectrum, and pulsed nature of neutron production result in significant differences in damage parameters between the two systems. With the same neutron wall loading, the displacement damage rate in the first wall in an IFE system is ˜10% lower than in an MFE system, while gas production and burnup rates are a factor of 2 lower. Self-cooled LiPb and Flibe blankets were analyzed. While using LiPb results in higher displacement damage, Flibe yields higher gas production and burnup rates. The effects of displacement damage and helium production on defect accumulation in SiC/SiC composites are also discussed.

  9. Nuclear Data Uncertainty Propagation in Depletion Calculations Using Cross Section Uncertainties in One-group or Multi-group

    Science.gov (United States)

    Díez, C. J.; Cabellos, O.; Martínez, J. S.

    2015-01-01

    Several approaches have been developed in last decades to tackle nuclear data uncertainty propagation problems of burn-up calculations. One approach proposed was the Hybrid Method, where uncertainties in nuclear data are propagated only on the depletion part of a burn-up problem. Because only depletion is addressed, only one-group cross sections are necessary, and hence, their collapsed one-group uncertainties. This approach has been applied successfully in several advanced reactor systems like EFIT (ADS-like reactor) or ESFR (Sodium fast reactor) to assess uncertainties on the isotopic composition. However, a comparison with using multi-group energy structures was not carried out, and has to be performed in order to analyse the limitations of using one-group uncertainties.

  10. Proposal for the construction of a high resolution tomography installation in the BWR-type reactor Dodewaard, Netherlands. Voorstel voor de bouw van een hoge resolutie tomografie-opstelling in de centrale Dodewaard

    Energy Technology Data Exchange (ETDEWEB)

    Jaspers, J.E.; Tanke, R.H.J. (KEMA Onderzoek en Ontwikkeling, Arnhem (Netherlands))

    1993-01-01

    KEMA carried out several measurements on fission fuel rods and fission fuel elements of the nuclear reactor Dodewaard within the framework of the project Tomography Fission Fuel. The use of tomography provides information on the integrity of fission fuel, the burnup, the diffusion of fission fuel gases and the fission fuel/cladding interaction. It appeared that the present installation is not appropriate for high resolution tomography. Therefore it has been decided to design and construct a new tomography installation. Also attention is paid to the estimation of the costs for the chosen installation. The results of the title study can be used for the inspection of fission fuel elements in nuclear power plants and for local burnup analysis in fission fuel elements.

  11. Spent LWR fuel leach tests: Waste Isolation Safety Assessment program

    Energy Technology Data Exchange (ETDEWEB)

    Katayama, Y.B.

    1979-04-01

    Spent light-water-reactor (LWR) fuels with burnups of 54.5, 28 and 9 MWd/kgU were leach-tested in deionized water at 25/sup 0/C. Fuel burnup has no apparent effect on the calculated leach rates based upon the behavior of /sup 137/Cs and /sup 239 +240/Pu. A leach test of 54.5 MWd/kgU spent fuel in synthetic sea brine showed that the cesium-based leach rate is lower in sea brine than in deionized water. A rise in the leach rate was observed after approximately 600 d of cumulative leaching. During the rise, the leach rate for all the measured radionuclides become nearly equal. Evidence suggests that exposure of new surfaces to the leachant may cause the increase. As a result, experimental work to study leaching mechanisms of spent fuel has been initiated. 22 figures.

  12. Application of Thermochemical Modeling to Assessment/Evaluation of Nuclear Fuel Behavior

    Energy Technology Data Exchange (ETDEWEB)

    Besmann, Theodore M [University of South Carolina, Columbia; McMurray, Jake W [ORNL; Simunovic, Srdjan [ORNL

    2016-01-01

    The combination of new fuel compositions and higher burn-ups envisioned for the future means that representing fuel properties will be much more important, and yet more complex. Behavior within the oxide fuel rods will be difficult to model owing to the high temperatures, and the large number of elements generated and their significant concentrations that are a result of fuels taken to high burn-up. This unprecedented complexity offers an enormous challenge to the thermochemical understanding of these systems and opportunities to advance solid solution models to describe these materials. This paper attempts to model and simulate that behavior using an oxide fuels thermochemical description to compute the equilibrium phase state and oxygen potential of LWR fuel under irradiation.

  13. The Application of High Performance Liquid Chromatography in the Analysis of Trace Rare Earth

    Institute of Scientific and Technical Information of China (English)

    WANG; Xiu-feng; DING; You-qian

    2012-01-01

    <正>Rare earth elements are very important in the field of radioanalytical chemistry, for it must be separated and determined in the measurements of burn-up and fission yield. High performance liquid chromatography has become a main method in the separation of rare earth elements due to its obvious advantages, this is, high speed of analysis, high efficiency and easy automation. The ion exchange chromatography is the main means to separate rare earth elements, especially the cation exchange

  14. Improve Design of Fuel Shear for Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    GAO; Wei; OUYANG; Ying-gen; LI; Wei-min

    2012-01-01

    <正>Due to the deeper burnup and higher fuel swelling, fast reactor metal fuel rod using 316 stainless steel cladding, replacing the traditional zirconia cladding. The diameter of fuel rod of fast reactor is much longer than that of PWR, and the cladding of stainless steel has better ductility than zirconia cladding. Using the existing shear still will cause several aspects of problem: 1) Longer diameter of rod leads to

  15. Study for 228Th reduction in thermal reactor with Th-U fuel cycls

    Institute of Scientific and Technical Information of China (English)

    1999-01-01

    By using computercode WIMS/CENDL, the effects of some parameters, core configuration such as fuel element structure, neutron flux and burn-up, are discussed in thispaper.It is shown that high neutron flux, small fuel rod diameter,large volume ratio of coolant to fuel, seed-blank heterogeneous corearrangement and 231Pa chemical separation are necessary for reducing 228Th production in reactor.

  16. The Serpent Monte Carlo Code: Status, Development and Applications in 2013

    Science.gov (United States)

    Leppänen, Jaakko; Pusa, Maria; Viitanen, Tuomas; Valtavirta, Ville; Kaltiaisenaho, Toni

    2014-06-01

    The Serpent Monte Carlo reactor physics burnup calculation code has been developed at VTT Technical Research Centre of Finland since 2004, and is currently used in 100 universities and research organizations around the world. This paper presents the brief history of the project, together with the currently available methods and capabilities and plans for future work. Typical user applications are introduced in the form of a summary review on Serpent-related publications over the past few years.

  17. Using ORIGEN2 to Predict Nuclear Reactor Fuel Compositions.

    Science.gov (United States)

    1988-03-01

    Although the principal use of ORIGEN2 is to calculate the isotopic composition of nuclear materials, the following parameters may also be computed with...V’v inal vector by a second vector before storing in the destination vector. BUP: Burnup calculation . Identifies the beginning and end of a series of...has no effect on the accuracy of the calculations . Pressurized Water Reactor (PWR), 33 GWd/MTIHM The ORIGEN2 PWR models are based on a Westinghouse

  18. Variants of closing the nuclear fuel cycle

    Science.gov (United States)

    Andrianova, E. A.; Davidenko, V. D.; Tsibulskiy, V. F.; Tsibulskiy, S. V.

    2015-12-01

    Influence of the nuclear energy structure, the conditions of fuel burnup, and accumulation of new fissile isotopes from the raw isotopes on the main parameters of a closed fuel cycle is considered. The effects of the breeding ratio, the cooling time of the spent fuel in the external fuel cycle, and the separation of the breeding area and the fissile isotope burning area on the parameters of the fuel cycle are analyzed.

  19. Post irradiation examination of thermal reactor fuels

    Science.gov (United States)

    Sah, D. N.; Viswanathan, U. K.; Ramadasan, E.; Unnikrishnan, K.; Anantharaman, S.

    2008-12-01

    The post irradiation examination (PIE) facility at the Bhabha Atomic Research Centre (BARC) has been in operation for more than three decades. Over these years this facility has been utilized for examination of experimental fuel pins and fuels from commercial power reactors operating in India. In a program to assess the performance of (U,Pu)O 2 MOX fuel prior to its introduction in commercial reactors, three experimental MOX fuel clusters irradiated in the pressurized water loop (PWL) of CIRUS up to burnup of 16 000 MWd/tU were examined. Fission gas release from these pins was measured by puncture test. Some of these fuel pins in the cluster contained controlled porosity pellets, low temperature sintered (LTS) pellets, large grain size pellets and annular pellets. PIE has also been carried out on natural UO 2 fuel bundles from Indian PHWRs, which included two high burnup (˜15 000 MWd/tU) bundles. Salient investigations carried out consisted of visual examination, leak testing, axial gamma scanning, fission gas analysis, microstructural examination of fuel and cladding, β, γ autoradiography of the fuel cross-section and fuel central temperature estimation from restructuring. A ThO 2 fuel bundle irradiated in Kakrapar Atomic Power Station (KAPS) up to a nominal fuel burnup of ˜11 000 MWd/tTh was also examined to evaluate its in-pile performance. The performance of the BWR fuel pins of Tarapur Atomic Power Stations (TAPS) was earlier assessed by carrying out PIE on 18 fuel elements selected from eight fuel assemblies irradiated in the two reactors. The burnup of these fuel elements varied from 5000 to 29 000 MWd/tU. This paper provides a brief review of some of the fuels examined and the results obtained on the performance of natural UO 2, enriched UO 2, MOX, and ThO 2 fuels.

  20. Light-water reactors: preliminary safety and environmental information document. Volume I

    Energy Technology Data Exchange (ETDEWEB)

    1980-01-01

    Information is presented concerning the reference PWR reactor system; once-through, low-enrichment uranium-235 fuel, 30 MWD per kilogram (PWR LEU(5)-OT); once-through, low-enrichment, high-burnup uranium fuel (PWR LEU(5)-Mod OT); self-generated plutonium spiked recycle (PWR LEU(5)-Pu-Spiked Recycle); denatured uranium-233/thorium cycle (PWR DU(3)-Th Recycle DU(3)); and plutonium/thorium cycle (Pu/ThO/sub 2/ Burner).

  1. Comment: collection of assay data on isotopic composition in LWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Naito, Yoshitaka; Kurosawa, Masayoshi; Suyama, Kenya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-03-01

    Many assay data of LWR spent fuels have been collected from reactors in the world and some of them are already stored in the database SFCOMPO which was constructed on a personal computer IBM PC/AT. On the other hand, Group constant libraries for burnup calculation code ORIGEN-II were generated from the nuclear data file JENDL3.2. These libraries were evaluated by using the assay data in SFCOMPO. (author)

  2. Cladding embrittlement during postulated loss-of-coolant accidents.

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M.; Yan, Y.; Burtseva, T.; Daum, R.; Nuclear Engineering Division

    2008-07-31

    The effect of fuel burnup on the embrittlement of various cladding alloys was examined with laboratory tests conducted under conditions relevant to loss-of-coolant accidents (LOCAs). The cladding materials tested were Zircaloy-4, Zircaloy-2, ZIRLO, M5, and E110. Tests were performed with specimens sectioned from as-fabricated cladding, from prehydrided (surrogate for high-burnup) cladding, and from high-burnup fuel rods which had been irradiated in commercial reactors. The tests were designed to determine for each cladding material the ductile-to-brittle transition as a function of steam oxidation temperature, weight gain due to oxidation, hydrogen content, pre-transient cladding thickness, and pre-transient corrosion-layer thickness. For short, defueled cladding specimens oxidized at 1000-1200 C, ring compression tests were performed to determine post-quench ductility at {le} 135 C. The effect of breakaway oxidation on embrittlement was also examined for short specimens oxidized at 800-1000 C. Among other findings, embrittlement was found to be sensitive to fabrication processes--especially surface finish--but insensitive to alloy constituents for these dilute zirconium alloys used as cladding materials. It was also demonstrated that burnup effects on embrittlement are largely due to hydrogen that is absorbed in the cladding during normal operation. Some tests were also performed with longer, fueled-and-pressurized cladding segments subjected to LOCA-relevant heating and cooling rates. Recommendations are given for types of tests that would identify LOCA conditions under which embrittlement would occur.

  3. Fission-product retention in HTGR fuels

    Energy Technology Data Exchange (ETDEWEB)

    Homan, F.J.; Kania, M.J.; Tiegs, T.N.

    1982-01-01

    Retention data for gaseous and metallic fission products are presented for both Triso-coated and Biso-coated HTGR fuel particles. Performance trends are established that relate fission product retention to operating parameters, such as temperature, burnup, and neutron exposure. It is concluded that Biso-coated particles are not adequately retentive of fission gas or metallic cesium, and Triso-coated particles which retain cesium still lose silver. Design implications related to these performance trends are identified and discussed.

  4. 3. world TRIGA users conference. Papers and abstracts

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2006-07-01

    The Conference is focused on TRIGA reactors operation and applications. The main topics are: use of the reactor as a research tool; inspection of spent fuel elements; integrity of fuel rods cladding checks; evaluation of corrosion of aluminum-base fuel cladding materials; Pitting behavior of Aluminum alloys; Monte Carlo simulation of TRIGA: reactivity worth, burnup, flux and power; irradiation facilities; thermal hydraulics analyses etc.

  5. Fissuration des matériaux à gradient de propriétés. Application au Zircaloy hydruré.

    OpenAIRE

    Perales, Frédéric

    2005-01-01

    This thesis is devoted to the dynamic fracture of functionally graded materials. More particularly, it deals with the toughness of nuclear cladding at high burnup submitted to transient loading. The fracture is studied at local scale using cohesive zone model in a multibody approach. Cohesive zone models include frictional contact to take into account mixed mode fracture. Non smooth dynamics problems are treated within the Non-Smooth Contact Dynamics framework. A multiscale study is necessary...

  6. Progress and applications of the variational nodal method

    Energy Technology Data Exchange (ETDEWEB)

    Carrico, C.B. [Argonne National Lab., IL (United States); Palmiotti, G. [CEA Centre d`Etudes Nucleaires de Cadarache, 13 - Saint-Paul-lez-Durance (France); Lewis, E.E. [Northwestern Univ., Evanston, IL (United States). Dept. of Mechanical Engineering

    1995-07-01

    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 calculations, burnup calculations, highly heterogeneous cores, higher-order spherical harmonics approximations, perturbation theory and heterogeneous nodes.

  7. Materials science research for sodium cooled fast reactors

    Indian Academy of Sciences (India)

    Baldev Raj

    2009-06-01

    The paper gives an insight into basic as well as applied research being carried out at the Indira Gandhi Centre for Atomic Research for the development of advanced materials for sodium cooled fast reactors towards extending the life of reactors to nearly 100 years and the burnup of fuel to 2,00,000 MWd/t with an objective of providing fast reactor electricity at an affordable and competitive price.

  8. Identification of unknown nuclear material

    Energy Technology Data Exchange (ETDEWEB)

    Nicolaou, G. [University of Thrace, Department of Electrical and Computer Engineering, Laboratory of Nuclear Technology, Kimmerria Campus, 67100 Xanthi (Greece)

    2010-07-01

    Aim: provenance determination of unknown nuclear material: - demonstrated for spent nuclear fuel; - information sought for unknown: fuel type, reactor type where fuel was irradiated, final burnup; Using an isotopic finger-printing method: - U, Pu or Pu isotopics or fission products; - simulations of fuel evolution during irradiation, using ORIGEN; - multivariate statistical tools. Fuel considered: simulated commercial spent fuel for a range of burnups: - PWR UO{sub 2} 3.1% and 3.5% {sup 235}U, - PWR thermal MOX, - BWR UO{sub 2} 3.2% {sup 235}U, - CANDU-N natural U, - CANDU-S UO{sub 2} 3.2% {sup 235}U, - fast Reactor MOX; simulated commercial spent fuel for a range of burnups: - PWR UO{sub 2} 3.1% and 3.5% {sup 235}U, - PWR thermal MOX, - BWR UO{sub 2} 3.2% {sup 235}U, - CANDU-N natural U, - CANDU-S UO{sub 2} 3.2% {sup 235}U, - fast Reactor MOX; 'unknown' spent fuel: - PWR 1: UO{sub 2} 3.1% {sup 235}U (26 GWd/t), - PWR 2: UO{sub 2} 3.1% {sup 235}U (32 GWd/t). Procedures: U, Pu or Pu isotopic compositions or fission products: - isotopic composition of unknown spent fuel, - simulated for commercial spent fuel from a range of nuclear power reactors {yields} comparison of compositions through factor analysis {yields} unknown has the provenance of the commercial spent fuel with which it exhibits the most similar composition. In conclusion: different reactor-fuel types well resolved; fuel and reactor type accurately predicted; burnup predicted to within 5% of declared; different reactor-fuel types. (authors)

  9. 1987 calibration of the TFTR neutron spectrometers

    Energy Technology Data Exchange (ETDEWEB)

    Barnes, C.W.; Strachan, J.D. (Los Alamos National Lab., NM (USA); Princeton Univ., NJ (USA). Plasma Physics Lab.)

    1989-12-01

    The {sup 3}He neutron spectrometer used for measuring ion temperatures and the NE213 proton recoil spectrometer used for triton burnup measurements were absolutely calibrated with DT and DD neutron generators placed inside the TFTR vacuum vessel. The details of the detector response and calibration are presented. Comparisons are made to the neutron source strengths measured from other calibrated systems. 23 refs., 19 figs., 6 tabs.

  10. SAS2H Generated Isotopic Concentrations For B&W 15X15 PWR Assembly (SCPB:N/A)

    Energy Technology Data Exchange (ETDEWEB)

    J.W. Davis

    1996-08-29

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide pressurized water reactor (PWR) isotopic composition data as a function of time for use in criticality analyses. The objectives of this evaluation are to generate burnup and decay dependant isotopic inventories and to provide these inventories in a form which can easily be utilized in subsequent criticality calculations.

  11. Fessenheim: 30 years of nuclear fuel operation; Fessenheim: 30 ans d'exploitation du combustible nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    Abgrall, G. [Centrale de Fessenheim, EDF, 68 - Dessenheim (France)

    2008-03-15

    Fessenheim units 1 and 2 are the first two 900 MW PWR put into operation in France (1977). This article reviews 30 years of change, optimization and feedback experience from Fessenheim, concerning: -) fuel assemblies (particularly the design of some components like grids, ends and guide tubes), -) the reload fuel management (to get a higher unloading burn-up), -) the refueling machine and tools (heavy modifications to reduce the human factor), and -) work organization (work shifts and staff training). (A.C.)

  12. Development of Improved Models and Designs for Coated-Particle Gas Reactor Fuels -- Final Report under the International Nuclear Energy Research Initiative (I-NERI)

    Energy Technology Data Exchange (ETDEWEB)

    Petti, David [Idaho National Lab. (INL), Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab. (INEEL); Martin, Philippe [Commissariat a l' Energie Atomique et aux Energies Alternatives (CEA-Saclay), Gif-sur-Yvette (France); Phelip, Mayeul [Commissariat a l' Energie Atomique et aux Energies Alternatives (CEA-Saclay), Gif-sur-Yvette (France); Ballinger, Ronald [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    2004-12-01

    The objective of this INERI project was to develop improved fuel behavior models for gas reactor coated-particle fuels and to explore improved coated-particle fuel designs that could be used reliably at very high burnups and potentially in gas-cooled fast reactors. Project participants included the Idaho National Engineering Laboratory (INEEL), Centre Étude Atomique (CEA), and the Massachusetts Institute of Technology (MIT). To accomplish the project objectives, work was organized into five tasks.

  13. Determination of Neptunium, Americium and Curium in Spent Nuclear Fuel Samples by Alpha Spectrometry Using {sup 239}Np and {sup 243}Am as a Spike and a Tracer

    Energy Technology Data Exchange (ETDEWEB)

    Jeo, Kih-Soo; Song, Byung-Chul; Kim, Young-Bok; Han, Sun-Ho; Jeon, Young-Shin; Jung, Euo-Chang; Jee, Kwang-Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2007-07-01

    Determination of actinide elements and fission products in spent nuclear fuels is of importance for a burnup determination and source term evaluation. Especially, the amounts of uranium and plutonium isotopes are used for the evaluation of a burnup credit in spent nuclear fuels. Additionally, other actinides such as Np, Am and Cm in spent nuclear fuel samples is also required for the purposes mentioned above. In this study, {sup 237}Np, {sup 241}Am and {sup 244}Cm were determined by an alpha spectrometry for the source term data for high burnup spent nuclear fuels ranging from 37 to 62.9 GWD/MtU as a burnup. Generally, mass spectrometry has been known as the most powerful method for isotope determinations such as high concentrations of uranium and plutonium. However, in the case of minor actinides such as Np, Am and Cm, alpha spectrometry would be recommended instead. Determination of the transuranic elements in spent nuclear fuel samples is different from that for environmental samples because the amount of each nuclide in the spent fuel samples is higher and the relative ratios between each nuclide are also different from those for environmental samples. So, it is important to select an appropriate tracer and an optimum sample size depending on the nuclides and analytical method. In this study {sup 237}Np was determined by an isotope dilution alpha(gamma) spectrometry using {sup 239}Np as a spike, and {sup 241}Am and curium isotopes were determined by alpha spectrometry using {sup 243}Am as a tracer. The content of each nuclide was compared with that by the Origen-2 code.

  14. Probabilistic methods applied to the evaluation of the influence of the steady state uncertainties in the modeling of RIA scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Sagrado, I. C.; Herranz, L. E.

    2014-07-01

    Because of the need of evaluating safety criteria for higher burn-up levels during RIA accidents, considerable efforts have been invested to assess the actual codes capabilities. One of the main related issues is the evaluation of uncertainties affecting their predictions. While Monte Carlo (MC) methods are the reference, the use of techniques involving lower computation time, as statistical order approaches or Surrogate Models (SM), is increasing. (Author)

  15. Applications of Neutrino Physics

    OpenAIRE

    Christensen, Eric Kurt

    2014-01-01

    Neutrino physics has entered a precision era in which understanding backgrounds and systematic uncertainties is particularly important. With a precise understanding of neutrino physics, we can better understand neutrino sources. In this work, we demonstrate dependency of single detector oscillation experiments on reactor neutrino flux model. We fit the largest reactor neutrino flux model error, weak magnetism, using data from experiments. We use reactor burn-up simulations in combination with...

  16. Fuel performance annual report for 1989

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, W.J.; Berting, F.M. (Pacific Northwest Lab., Richland, WA (United States)); Wu, S. (Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Technology)

    1992-06-01

    This annual report, the twelfth in a series, provides a brief description of fuel performance during 1989 in commercial nuclear power plants and an indication of trends. Brief summaries of fuel design changes, fuel surveillance programs, fuel operating experience, fuel problems, high-burnup fuel experience, and items of general significance are provided. References to more detailed information and related US Nuclear Regulatory Commission evaluations are included.

  17. Determination of fission gas yields from isotope ratios

    DEFF Research Database (Denmark)

    Mogensen, Mogens Bjerg

    1983-01-01

    This paper describes a method of calculating the actual fission yield of Kr and Xe in nuclear fuel including the effect of neutron capture reactions and decay. The bases for this calculation are the cumulative yields (ref. 1) of Kr and Xe isotopes (or pairs of isotopes) which are unaffected...... by neutron capture reactions, and measured Kr and Xe isotope ratios. Also the burnup contribution from the different fissile heavy isotopes must be known in order to get accurate fission gas yields....

  18. Criticality safety evaluation of disposing of K Basin sludge in double-shell tank AW-105

    Energy Technology Data Exchange (ETDEWEB)

    ROGERS, C.A.

    1999-06-04

    A criticality safety evaluation is made of the disposal of K Basin sludge in double-shell tank (DST) AW-105 located in the 200 east area of Hanford Site. The technical basis is provided for limits and controls to be used in the development of a criticality prevention specification (CPS). A model of K Basin sludge is developed to account for fuel burnup. The iron/uranium mass ration required to ensure an acceptable magrin of subcriticality is determined.

  19. Gross gamma-ray measurements of light water reactor spent-fuel assemblies in underwater storage arrays

    Energy Technology Data Exchange (ETDEWEB)

    Moss, C.E.; Lee, D.M.

    1980-12-01

    Two gross gamma-ray detection systems have been developed for rapid measurement of spent-fuel assemblies in underwater storage racks. One system uses a scintillator as the detector and has a 2% crosstalk between a fuel assembly and an adjacent void. The other system uses an ion chamber as the detector. The measurements with both detectors correlate well with operator-declared burnup and cooling-time values.

  20. RERTR-8 Irradiation Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    D. M. Perez; M. A. Lillo; G. S. Chang; G. A. Roth; N. E. Woolstenhulme; D. M. Wachs

    2011-12-01

    The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-8, was designed to test monolithic mini-fuel plates fabricated via hot isostatic pressing (HIP), the effect of molybdenum (Mo) content on the monolithic fuel behavior, and the efficiency of ternary additions to dispersion fuel particles on the interaction layer behavior at higher burnup. The following report summarizes the life of the RERTR-8 experiment through end of irradiation, including as-run neutronic analysis, thermal analysis and hydraulic testing results.

  1. Design of unique pins for irradiation of higher actinides in a fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Basmajian, J.A.; Birney, K.R.; Weber, E.T.; Adair, H.L.; Quinby, T.C.; Raman, S.; Butler, J.K.; Bateman, B.C.; Swanson, K.M.

    1982-03-01

    The actinides produced by transmutation reactions in nuclear reactor fuels are a significant factor in nuclear fuel burnup, transportation and reprocessing. Irradiation testing is a primary source of data of this type. A segmented pin design was developed which provides for incorporation of multiple specimens of actinide oxides for irradiation in the UK's Prototype Fast Reactor (PFR) at Dounreay Scotland. Results from irradiation of these pins will extend the basic neutronic and material irradiation behavior data for key actinide isotopes.

  2. Fission products, activity calculation of spent-fuel

    Energy Technology Data Exchange (ETDEWEB)

    Souka, N.; El-Hakiem, M.N.

    1981-01-01

    This work is a scrutiny of the activity of burned up fuel elements of the ET-RR-1. A knowledge of this activity as well as its decay with time is quite helpful in shielding calculations related to construction purposes of hot facilities. The present treatment is based on a knowledge of: fuel composition, percentage burnup, and fission yields of produced isotopes. Cooling periods ranging from 1 hr to 10 years were considered.

  3. Key safety parameters in the optimization of fuel management

    Energy Technology Data Exchange (ETDEWEB)

    Kollmar, W.; Boehm, R.; Dernedde, I.; Haase, H.; Kiehlmann, H.D.; Neufert, A.

    1988-08-01

    Nuclear design related key safety parameters and admissible parameter ranges are defined for reload cycles which are so similar in safety terms as to allow these to be covered by generic reload safety analyses in advance. The conceptual frame of such safety analyses together with the resulting economic benefits are illustrated by four concrete applications demonstrating reduction of excessive safety margins, increase in discharge burnup, streamlining of steam break analysis, and increase in operational flexibility of first cores.

  4. Pre-irradiation testing and analysis to support the LWRS Hybrid SiC-CMC-Zircaloy-04 unfueled rodlet irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Isabella J van Rooyen

    2012-09-01

    Nuclear fuel performance is a significant driver of nuclear power plant operational performance, safety, economics and waste disposal requirements. The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Pathway focuses on improving the scientific knowledge basis to enable the development of high-performance, high burn-up fuels with improved safety and cladding integrity and improved nuclear fuel cycle economics. To achieve significant improvements, fundamental changes are required in the areas of nuclear fuel composition, cladding integrity, and fuel/cladding interaction.

  5. Pre-irradiation testing and analysis to support the LWRS Hybrid SiC-CMC-Zircaloy-04 unfueled rodlet irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Isabella J van Rooyen

    2013-01-01

    Nuclear fuel performance is a significant driver of nuclear power plant operational performance, safety, economics and waste disposal requirements. The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Pathway focuses on improving the scientific knowledge basis to enable the development of high-performance, high burn-up fuels with improved safety and cladding integrity and improved nuclear fuel cycle economics. To achieve significant improvements, fundamental changes are required in the areas of nuclear fuel composition, cladding integrity, and fuel/cladding interaction.

  6. RERTR-12 Post-irradiation Examination Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Rice, Francine; Williams, Walter; Robinson, Adam; Harp, Jason; Meyer, Mitch; Rabin, Barry

    2015-02-01

    The following report contains the results and conclusions for the post irradiation examinations performed on RERTR-12 Insertion 2 experiment plates. These exams include eddy-current testing to measure oxide growth; neutron radiography for evaluating the condition of the fuel prior to sectioning and determination of fuel relocation and geometry changes; gamma scanning to provide relative measurements for burnup and indication of fuel- and fission-product relocation; profilometry to measure dimensional changes of the fuel plate; analytical chemistry to benchmark the physics burnup calculations; metallography to examine the microstructural changes in the fuel, interlayer and cladding; and microhardness testing to determine the material-property changes of the fuel and cladding. These characterization activities are tailored specifically to define: • The mechanical response of fuel meat, cladding, and interlayers, including diffusion barrier integrity • Whether geometry is stable and predictable; that changes in channel gap do not compromise ability to cool fuel • That fuel performance is known and predictable • A limited set of physical properties that are important for the analysis of fuel burnup limits • Whether swelling is stable and predictable.

  7. Full Core 3-D Simulation of a Partial MOX LWR Core

    Energy Technology Data Exchange (ETDEWEB)

    S. Bays; W. Skerjanc; M. Pope

    2009-05-01

    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 and fuel cycle calculations. For the purpose of isotopic generation for fuel cycle analyses, the approach of using a 2-D lattice code (i.e., fuel assembly in infinite lattice) gave reasonable predictions of uranium and plutonium isotope concentrations at the predicted 3-D core simulation batch average discharge burnup. However, it was found that the 2-D lattice calculation can under-predict the power of pins located along a shared edge between MOX and UO2 by as much as 20%. In this analysis, this error did not occur in the peak pin. However, this was a coincidence and does not rule out the possibility that the peak pin could occur in a lattice position with high calculation uncertainty in future un-optimized studies. Another important consideration in realistic fuel design is the prediction of the peak axial burnup and neutron fluence. The use of 3-D core simulation gave peak burnup conditions, at the pellet level, to be approximately 1.4 times greater than what can be predicted using back-of-the-envelope assumptions of average specific power and irradiation time.

  8. Extrapolating power-ramp performance criteria for current and advanced CANDU fuels

    Energy Technology Data Exchange (ETDEWEB)

    Tayal, M.; Chassie, G.G

    2000-06-01

    To improve the precision and accuracy of power-ramp performance criteria for high-burnup fuel, we have examined in-reactor fuel performance data as well as out-reactor test data. The data are consistent with some of the concepts used in the current formulations for defining fuel failure thresholds, such as size of power-ramp and extent of burnup. Our review indicates that there is a need to modify some other aspects of the current formulations; therefore, a modified formulation is presented in this paper. The improvements mainly concern corrodent concentration and its relationships with threshold stress for failure. The new formulation is consistent with known and expected trends such as strength of Zircaloy in corrosive environment, timing of the release of fission products to the pellet-to-sheath gap, CANLUB coating, and fuel burnup. Because of the increased precision and accuracy, the new formulation is better able to identify operational regimes that are at risk of power-ramp failures; this predictive ability provides enhanced protection to fuel against power-ramp defects. At die same time, by removing unnecessary conservatisms in other areas, the new formulation permits a greater range of defect-free operational envelope as well as larger operating margins in regions that are, in fact, not prone to power-ramp failures. (author)

  9. PWR core stablity aganst xenon-induced spatial power oscillation

    Energy Technology Data Exchange (ETDEWEB)

    Moon, H.J.; Han, K.I. (Korea Advanced Energy Research Inst., Seoul (Republic of Korea))

    1982-06-01

    Stability of a PWR core against xenon-induced axial power oscillation is studied using one-dimensional xenon transient analysis code, DD1D, that has been developed and verified at KAERI. Analyzed by DD1D utilizing the Kori Unit 1 design and operating data is the sensitivity of axial stability in a PWR core to the changes in core physical parameters including core power level, moderator temperature coefficient, core inlet temperature, doppler power coefficient and core average burnup. Through the sensitivity study the Kori Unit 1 core is found to be stable against axial xenon oscillation at the beginning of cycle 1. But, it becomes less stable as burnup progresses, and unstable at the end of cycle. Such a decrease in stability is mainly due to combined effect of changes in axial power distribution, moderator temperature coefficient and doppler power coefficient as core burnup progresses. It is concluded from the stability analysis of the Kori Unit 1 core that design of a large PWR with high power density and increased dimension can not avoid xenon-induced axial power instabilites to some extents, especially at the end of cycle.

  10. Performance of low smeared density sodium-cooled fast reactor metal fuel

    Science.gov (United States)

    Porter, D. L.; Chichester, H. J. M.; Medvedev, P. G.; Hayes, S. L.; Teague, M. C.

    2015-10-01

    An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at.% burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactor designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low melting points and gaseous precursors (Cs and Rb). A model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.

  11. AFC-1 Transmutation Fuels Post-Irradiation Hot Cell Examination 4-8 at.% - Final Report (Irradiation Experiments AFC-1B, -1F and -1Æ)

    Energy Technology Data Exchange (ETDEWEB)

    Bruce Hilton; Douglas Porter; Steven Hayes

    2006-09-01

    The AFC-1B, AFC-1F and AFC-1Æ irradiation tests are part of a series of test irradiations designed to evaluate the feasibility of the use of actinide bearing fuel forms in advanced fuel cycles for the transmutation of transuranic elements from nuclear waste. The tests were irradiated in the Idaho National Laboratory’s (INL) Advanced Test Reactor (ATR) to an intermediate burnup of 4 to 8 at% (2.7 - 6.8 x 1020 fiss/cm3). The tests contain metallic and nitride fuel forms with non-fertile (i.e., no uranium) and low-fertile (i.e., uranium bearing) compositions. Results of postirradiation hot cell examinations of AFC-1 irradiation tests are reported for eleven metallic alloy transmutation fuel rodlets and five nitride transmutation fuel rodlets. Non-destructive examinations included visual examination, dimensional inspection, gamma scan analysis, and neutron radiography. Detailed examinations, including fission gas puncture and analysis, metallography / ceramography and isotopics and burnup analyses, were performed on five metallic alloy and three nitride transmutation fuels. Fuel performance of both metallic alloy and nitride fuel forms was best correlated with fission density as a burnup metric rather than at.% depletion. The actinide bearing transmutation metallic alloy compositions exhibit irradiation performance very similar to U-xPu-10Zr fuel at equivalent fission densities. The irradiation performance of nitride transmutation fuels was comparable to limited data published on mixed nitride systems.

  12. A lattice-based Monte Carlo evaluation of Canada Deuterium Uranium-6 safety parameters

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yong Hee; Hartanto, Donny; Kim, Woo Song [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology (KAIST), Daejeon (Korea, Republic of)

    2016-06-15

    Important safety parameters such as the fuel temperature coefficient (FTC) and the power coefficient of reactivity (PCR) of the CANada Deuterium Uranium (CANDU-6) reactor have been evaluated using the Monte Carlo method. For accurate analysis of the parameters, the Doppler broadening rejection correction scheme was implemented in the MCNPX code to account for the thermal motion of the heavy uranium-238 nucleus in the neutron-U scattering reactions. In this work, a standard fuel lattice has been modeled and the fuel is depleted using MCNPX. The FTC value is evaluated for several burnup points including the mid-burnup representing a near-equilibrium core. The Doppler effect has been evaluated using several cross-section libraries such as ENDF/B-VI.8, ENDF/B-VII.0, JEFF-3.1.1, and JENDL-4.0. The PCR value is also evaluated at mid-burnup conditions to characterize the safety features of an equilibrium CANDU-6 reactor. To improve the reliability of the Monte Carlo calculations, we considered a huge number of neutron histories in this work and the standard deviation of the k-infinity values is only 0.5-1 pcm.

  13. Isotopic Generation and Confirmation of the PWR Application Model 

    Energy Technology Data Exchange (ETDEWEB)

    L.B. Wimmer

    2003-11-10

    The objective of this calculation is to establish an isotopic database to represent commercial spent nuclear fuel (CSNF) from pressurized water reactors (PWRs) in criticality analyses performed for the proposed Monitored Geologic Repository at Yucca Mountain, Nevada. Confirmation of the conservatism with respect to criticality in the isotopic concentration values represented by this isotopic database is performed as described in Section 3.5.3.1.2 of the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2000). The isotopic database consists of the set of 14 actinides and 15 fission products presented in Section 3.5.2.1.1 of YMP 2000 for use in CSNF burnup credit. This set of 29 isotopes is referred to as the principal isotopes. The oxygen isotope from the UO{sub 2} fuel is also included in the database. The isotopic database covers enrichments of {sup 235}U ranging from 1.5 to 5.5 weight percent (wt%) and burnups ranging from approximately zero to 75 GWd per metric ton of uranium (mtU). The choice of fuel assembly and operating history values used in generating the isotopic database are provided is Section 5. Tables of isotopic concentrations for the 29 principal isotopes (plus oxygen) as a function of enrichment and burnup are provided in Section 6.1. Results of the confirmation of the conservatism with respect to criticality in the isotopic concentration values are provided in Section 6.2.

  14. Verification of the Mimir-N2 Joyo plant dynamic code

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, Akihiro [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Kuroha, Takaya [Nuclear Energy System Inc., Tokyo (Japan)

    2002-06-01

    Passive safety systems at Experimental Fast Reactor JOYO were studied to demonstrate the inherent safety of MOX fueled sodium cooled fast breeder reactors. The Mimir-N2 analysis code, developed to analyze JOYO plant kinetics, was selected as the standard code. To increase the reliability, Mimir-N2 was modified based on data from plant characteristics and natural circulation tests in JOYO. JOYO operational data suggest that the burn-up dependency of the power reactivity coefficient could be due to the reactivity shift caused by decrease of fuel pellet thermal expansion in the axial direction. Based on the relationship between the measured power reactivity coefficient and the core averaged burn-up, burn-up dependency was estimated and introduced to the Mimir-N2 model. This brought good correspondence between calculated and measured values for a step reactivity response test. Calculated plant parameters including power range neutron monitor response and fuel subassembly outlet coolant temperature corresponded to measured values. Mimir-N2 could simulate plant dynamics such as the perturbations due to core support plate thermal expansion. (author)

  15. An upgraded Monju core concept for demonstration of future FBR technologies

    Energy Technology Data Exchange (ETDEWEB)

    Kinjo, Hidehito; Nishi, Hiroshi [Japan Nuclear Cycle Development Inst., Tsuruga Head Office, Tsuruga, Fukui (Japan); Kageyama, Takeshi [Nuclear Energy System Inc., Tokyo (Japan)

    2002-03-01

    Upgrading of the Monju core is planned for increased fuel burn-up and longer operating cycles. It is also proposed that this upgraded core be utilized as an irradiation test bed to demonstrate the performance of the advanced fuel subassembly concepts proposed by the 'Feasibility Study on Commercialized Fast Reactor Cycle Systems (FS)'. A conceptual design study has been performed at the International Cooperation and Technology Development Center to investigate the feasibility of demonstrating the commercial FBR fuel subassembly concept proposed in the first phase of the above study. The specification of the subassembly with an increased fuel pin diameter irradiated in Monju will have fewer pins than the full-scale commercial subassembly due to the smaller wrapper diameter. The effects of loading different numbers of commercial-type subassemblies in different patterns on the major core characteristics have been evaluated along with the reactivity parameters related to core safety. The results show that in a core loading scheme with up to 30 subassemblies around the core center, the target discharge burn-up of 150 GWd/t could be achieved in 5 years, without causing significant drawbacks on the core neutronics and safety aspects, such as the maximum linear power, burnup reactivity, control-rod worth and Doppler reactivity. Thus the feasibility of demonstrating the performance of the commercial FBR fuel designs of this type in the upgraded core has been confirmed. (author)

  16. Post-irradiation examinations of Li{sub 4}SiO{sub 4} pebbles irradiated in the EXOTIC-8 experiment

    Energy Technology Data Exchange (ETDEWEB)

    Piazza, G. E-mail: giovanni.piazza@jet.efda.org; Erbe, A.; Rolli, R.; Romer, O

    2004-08-01

    In the EXOTIC-8 irradiation experiment in the High Flux Reactor (HFR) in Petten, lithium orthosilicate (OSi) pebbles with 50% {sup 6}Li-content were irradiated with the aim at investigating the behaviour of breeder ceramics at DEMO relevant burn-up. Specifically, changes in the mechanical stability as well as in the tritium release properties (in-pile and out-of-pile) had to be studied. After the completion of the irradiation, lithium orthosilicate pebble samples were shipped to the Research Centre of Karlsruhe (FZK) for post-irradiation examinations. The pebbles irradiated up to about 11% {sup 6}Li burn-up showed an increase in smaller cracks in the bulk and the presence of larger through-cracks, but the amount of fragments is quite small and the pebbles maintained a very good crush load. On the basis of these results, it can be concluded that the OSi pebbles can withstand high burn-up irradiation without an unduly high fragmentation.

  17. Validation of the Physics Analysis used to Characterize the AGR-1 TRISO Fuel Irradiation Test

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James W.; Harp, Jason M.; Demkowicz, Paul A.; Hawkes, Grant L.; Chang, Gray S.

    2015-05-01

    The results of a detailed physics depletion calculation used to characterize the AGR-1 TRISO-coated particle fuel test irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory are compared to measured data for the purpose of validation. The particle fuel was irradiated for 13 ATR power cycles over three calendar years. The physics analysis predicts compact burnups ranging from 11.30-19.56% FIMA and cumulative neutron fast fluence from 2.21?4.39E+25 n/m2 under simulated high-temperature gas-cooled reactor conditions in the ATR. The physics depletion calculation can provide a full characterization of all 72 irradiated TRISO-coated particle compacts during and post-irradiation, so validation of this physics calculation was a top priority. The validation of the physics analysis was done through comparisons with available measured experimental data which included: 1) high-resolution gamma scans for compact activity and burnup, 2) mass spectrometry for compact burnup, 3) flux wires for cumulative fast fluence, and 4) mass spectrometry for individual actinide and fission product concentrations. The measured data are generally in very good agreement with the calculated results, and therefore provide an adequate validation of the physics analysis and the results used to characterize the irradiated AGR-1 TRISO fuel.

  18. Domain Decomposition Strategy for Pin-wise Full-Core Monte Carlo Depletion Calculation with the Reactor Monte Carlo Code

    Directory of Open Access Journals (Sweden)

    Jingang Liang

    2016-06-01

    Full Text Available Because of prohibitive data storage requirements in large-scale simulations, the memory problem is an obstacle for Monte Carlo (MC codes in accomplishing pin-wise three-dimensional (3D full-core calculations, particularly for whole-core depletion analyses. Various kinds of data are evaluated and quantificational total memory requirements are analyzed based on the Reactor Monte Carlo (RMC code, showing that tally data, material data, and isotope densities in depletion are three major parts of memory storage. The domain decomposition method is investigated as a means of saving memory, by dividing spatial geometry into domains that are simulated separately by parallel processors. For the validity of particle tracking during transport simulations, particles need to be communicated between domains. In consideration of efficiency, an asynchronous particle communication algorithm is designed and implemented. Furthermore, we couple the domain decomposition method with MC burnup process, under a strategy of utilizing consistent domain partition in both transport and depletion modules. A numerical test of 3D full-core burnup calculations is carried out, indicating that the RMC code, with the domain decomposition method, is capable of pin-wise full-core burnup calculations with millions of depletion regions.

  19. A neutronic study of the cycle PWR-CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alberto da; Pereira, Claubia; Veloso, Maria Auxiliadora Fortini; Fortini, Angela; Pinheiro, Ricardo Brant [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear]. E-mail: albertomoc@terra.com.br; claubia@nuclear.ufmg.br; dora@nuclear.ufmg.br; fortini@nuclear.ufmg.br; rbp@nuclear.ufmg.br

    2007-07-01

    The cycle PWR-CANDU was simulated using the WIMSD-5B and ORIGEN2.1 codes. It was simulated a fuel burnup of 33,000 MWd/t for UO{sub 2} with enrichment of 3.2% and a fuel extended burnup of 45,000 MWd/t for UO{sub 2} with enrichments of 3.5%, 4.0% and 5.0% in a PWR reactor. The PWR discharged fuel was submitted to the simulation of deposition for five years. After that, it was submitted to AYROX reprocessing and used to produce a fuel to CANDU reactor. Then, it was simulated the burnup in the CANDU. Parameters such as infinite medium multiplication factor, k{sub inf}, fuel temperature coefficient of reactivity, {alpha}{sub TF}, moderator temperature coefficient of reactivity, {alpha}{sub TM}, the ratio rapid flux/total flux and the isotopic composition in the begin and the end of life were evaluated. The results showed that the fuels analyzed could be used on PWR and CANDU reactors without the need of change on the design of these reactors. (author)

  20. SARAPAN—A Simulated-Annealing-Based Tool to Generate Random Patterned-Channel-Age in CANDU Fuel Management Analyses

    Directory of Open Access Journals (Sweden)

    Doddy Kastanya

    2017-02-01

    Full Text Available In any reactor physics analysis, the instantaneous power distribution in the core can be calculated when the actual bundle-wise burnup distribution is known. Considering the fact that CANDU (Canada Deuterium Uranium utilizes on-power refueling to compensate for the reduction of reactivity due to fuel burnup, in the CANDU fuel management analysis, snapshots of power and burnup distributions can be obtained by simulating and tracking the reactor operation over an extended period using various tools such as the *SIMULATE module of the Reactor Fueling Simulation Program (RFSP code. However, for some studies, such as an evaluation of a conceptual design of a next-generation CANDU reactor, the preferred approach to obtain a snapshot of the power distribution in the core is based on the patterned-channel-age model implemented in the *INSTANTAN module of the RFSP code. The objective of this approach is to obtain a representative snapshot of core conditions quickly. At present, such patterns could be generated by using a program called RANDIS, which is implemented within the *INSTANTAN module. In this work, we present an alternative approach to derive the patterned-channel-age model where a simulated-annealing-based algorithm is used to find such patterns, which produce reasonable power distributions.

  1. Integrated Simulation Studies of Plasma Performances and Fusion Reactions in the Deuterium Experiment of LHD

    Science.gov (United States)

    Murakami, S.; Yamaguchi, H.; Homma, M.; Maeta, S.; Saito, Y.; Fukuyama, A.; Nagaoka, K.; Takahashi, H.; Nakano, H.; Osakabe, M.; Yokoyama, M.; Tanaka, K.; Ida, K.; Yoshinuma, M.; Isobe, M.; Tomita, H.; Ogawa, K.; LHD Exp Group Team

    2016-10-01

    The deuterium experiment project from 2017 is planned in LHD, where the deuterium NBI heating beams with the power more than 30MW are injected into the deuterium plasma. Principal objects of this project are to clarify the isotope effect on the heat and particle transport in the helical plasma and to study energetic particle confinement in a helical magnetic configuration measuring triton burn-up neutrons. We study the deuterium experiment plasma of LHD applying the integrated simulation code, TASK3D [Murakami, PPCF2015], and the 5-D drift kinetic equation solver, GNET [Murakami, NF2006]. (i) More than 20% of ion temperature increment is obtained in the deuterium plasma (nD /nH +nD = 0.8) due to the isotope effect assuming the turbulent transport model based on the H/He plasma experiment of LHD. (ii) The triton burn-up simulation shows the triton slowing down distribution and the strong magnetic configuration dependency of the triton burn-up ratio in LHD. This work was supported by JSPS KAKENHI Grant Number 26420851.

  2. Domain Decomposition strategy for pin-wise full-core Monte Carlo depletion calculation with the reactor Monte Carlo Code

    Energy Technology Data Exchange (ETDEWEB)

    Liang, Jingang; Wang, Kan; Qiu, Yishu [Dept. of Engineering Physics, LiuQing Building, Tsinghua University, Beijing (China); Chai, Xiao Ming; Qiang, Sheng Long [Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu (China)

    2016-06-15

    Because of prohibitive data storage requirements in large-scale simulations, the memory problem is an obstacle for Monte Carlo (MC) codes in accomplishing pin-wise three-dimensional (3D) full-core calculations, particularly for whole-core depletion analyses. Various kinds of data are evaluated and quantificational total memory requirements are analyzed based on the Reactor Monte Carlo (RMC) code, showing that tally data, material data, and isotope densities in depletion are three major parts of memory storage. The domain decomposition method is investigated as a means of saving memory, by dividing spatial geometry into domains that are simulated separately by parallel processors. For the validity of particle tracking during transport simulations, particles need to be communicated between domains. In consideration of efficiency, an asynchronous particle communication algorithm is designed and implemented. Furthermore, we couple the domain decomposition method with MC burnup process, under a strategy of utilizing consistent domain partition in both transport and depletion modules. A numerical test of 3D full-core burnup calculations is carried out, indicating that the RMC code, with the domain decomposition method, is capable of pin-wise full-core burnup calculations with millions of depletion regions.

  3. Mechanical behaviors of the dispersion nuclear fuel plates induced by fuel particle swelling and thermal effect II: Effects of variations of the fuel particle diameters

    Science.gov (United States)

    Ding, Shurong; Wang, Qiming; Huo, Yongzhong

    2010-02-01

    In order to predict the irradiation mechanical behaviors of plate-type dispersion nuclear fuel elements, the total burnup is divided into two stages: the initial stage and the increasing stage. At the initial stage, the thermal effects induced by the high temperature differences between the operation temperatures and the room temperature are mainly considered; and at the increasing stage, the intense mechanical interactions between the fuel particles and the matrix due to the irradiation swelling of fuel particles are focused on. The large-deformation thermo-elasto-plasticity finite element analysis is performed to evaluate the effects of particle diameters on the in-pile mechanical behaviors of fuel elements. The research results indicate that: (1) the maximum Mises stresses and equivalent plastic strains at the matrix increase with the fuel particle diameters; the effects of particle diameters on the maximum first principal stresses vary with burnup, and the considered case with the largest particle diameter holds the maximum values all along; (2) at the cladding near the interface between the fuel meat and the cladding, the Mises stresses and the first principal stresses undergo major changes with increasing burnup, and different variations exist for different particle diameter cases; (3) the maximum Mises stresses at the fuel particles rise with the particle diameters.

  4. OTUS - Reactor inventory management system based on ORIGEN2

    Energy Technology Data Exchange (ETDEWEB)

    Poellaenen, R.; Toivonen, H.; Lahtinen, J.; Ilander, T.

    1995-10-01

    ORIGEN2 is a computer code that calculates nuclide composition and other characteristics of nuclear fuel. The use of ORIGEN2 requires good knowledge in reactor physics. However, once the input has been defined for a particular reactor type, the calculations can be easily repeated for any burnup and decay time. This procedure produces large output files that are difficult to handle manually. A new computer code, known as OTUS, was designed to facilitate the postprocessing of the data. OTUS makes use of the inventory files precalculated with ORIGEN2 in a way that enables their versatile treatment for different safety analysis purposes. A data base is created containing a comprehensive set of ORIGEN2 calculations as a function of fuel burnup and decay time. OTUS is a reactor inventory management system for a microcomputer with Windows interface. Four major data operations are available: (1) Build data modifies ORIGEN2 output data into a suitable format, (2) View data enables flexible presentation of the data as such, (3) Different calculations, such as nuclide ratios and hot particle characteristics, can be performed for severe accident analyses, consequence analyses and research purposes, (4) Summary files contain both burnup dependent and decay time dependent inventory information related to the nuclide and the reactor specified. These files can be used for safeguards, radiation monitoring and safety assessment. (orig.) (22 refs., 29 figs.).

  5. Artificial neural network modeling for fission gas release in LWR UO 2 fuel under RIA conditions

    Science.gov (United States)

    Koo, Yang-Hyun; Oh, Jae-Yong; Lee, Byung-Ho; Tahk, Young-Wook; Song, Kun-Woo

    2010-10-01

    A fission gas release (FGR) model was developed by using an artificial neural network method to predict fission gas release in UO 2 fuel under reactivity initiated accident (RIA) conditions. Based on the test data obtained in the CABRI test reactor and nuclear safety research reactor, the model takes into account the effect of the five parameters: pellet average burnup, peak fuel enthalpy, the ratio of peak fuel enthalpy to pulse width, fission gas release during base-irradiation, and grain size of a fuel pellet. The parametric study of the model, producing a physically reasonable trend of FGR for each parameter, shows that the pellet average burnup and the ratio of peak fuel enthalpy to pulse width are two of the most important parameters. Depending on the combination of input values for the five parameters, the application of the model to a fuel rod under typical RIA conditions of light water reactor produces 1.7-14.0% of FGR for the pellet average burnup ranging from 20 to 70 MW d/kg U.

  6. Reprocessed and combined thorium fuel cycles in a PER system with a micro heterogeneous approaches

    Energy Technology Data Exchange (ETDEWEB)

    Monteiro, Fabiana B.A.; Castro, Victor F.; Faria, Rochkhudson B. de; Pereira, Claubia; Fortini, Angela, E-mail: fabianabeghini@yahoo.com.br, E-mail: victorfariacastro@gmail.com, E-mail: rochkdefaria@yahoo.com.br, E-mail: claubia@nuclear.ufmg.br, E-mail: fortini@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2015-07-01

    A micro heterogeneous approaches were used to study the behavior of reprocessed fuel spiked with thorium in a PWR fuel element considering (TRU-Th) cycle. The goal is to achieve a higher burnup using three different configurations to model the fuel element using SCALE 6.0. The reprocessed fuels were obtained using the ORIGEN 2.1 code from a spent PWR standard fuel (33,000 MWd/tHM burned), with 3.1% of initial enrichment. The spent fuel remained in the cooling pool for five years and then reprocessed using the UREX+ technique. Three configurations of micro heterogeneous approaches were analyzed, and the k{sub inf} and plutonium evolution during the burnup were evaluated. The preliminary results show that the behavior of advanced fuel based on transuranic elements spiked with thorium, and micro heterogeneous approach are satisfactory in PWRs, and the configuration that use a combination of Th and TRU (configuration 1) seems to be the most promising once has higher values for k{sub inf} during the burnup, compared with other configurations. (author)

  7. Development of output user interface software to support analysis

    Science.gov (United States)

    Wahanani, Nursinta Adi; Natsir, Khairina; Hartini, Entin

    2014-09-01

    Data processing software packages such as VSOP and MCNPX are softwares that has been scientifically proven and complete. The result of VSOP and MCNPX are huge and complex text files. In the analyze process, user need additional processing like Microsoft Excel to show informative result. This research develop an user interface software for output of VSOP and MCNPX. VSOP program output is used to support neutronic analysis and MCNPX program output is used to support burn-up analysis. Software development using iterative development methods which allow for revision and addition of features according to user needs. Processing time with this software 500 times faster than with conventional methods using Microsoft Excel. PYTHON is used as a programming language, because Python is available for all major operating systems: Windows, Linux/Unix, OS/2, Mac, Amiga, among others. Values that support neutronic analysis are k-eff, burn-up and mass Pu239 and Pu241. Burn-up analysis used the mass inventory values of actinide (Thorium, Plutonium, Neptunium and Uranium). Values are visualized in graphical shape to support analysis.

  8. Geometrical α- and β-dose distributions and production rates of radiolysis products in water in contact with spent nuclear fuel

    Science.gov (United States)

    Nielsen, Fredrik; Jonsson, Mats

    2006-12-01

    A mathematical model for the dose distribution and production rates of radiolysis products in water surrounding spent nuclear fuel has been developed, based on the geometrical and energetic properties of radiation. The nuclear fuel particle is divided into layers, from which the radiation emits. The water is likewise divided into layers, where the doses are distributed. The doses are stored in vectors which are added to determine the total dose rate. A complete inventory with over 200 radionuclides has been used as input data for the model. The purpose of the model is to describe the geometrical dose distribution as a function of fuel age and burn-up, to be used as input data for kinetic modeling of the fuel dissolution. The results show that the β-dose contribution close to the spent fuel surface is negligible. Also, the variation in the relative α/β dose contribution between different ages and burn-ups is insignificant. The α- and β-dose rates vary between different burn-ups of the same age; the younger the fuel is, the larger is the difference. Exponential functions have been fitted to the relations between fuel age and average dose rate, giving useful expressions for determining average dose rates for fuel ages other than those covered in this work.

  9. DESIGN OF LSDS FOR ISOTOPIC FISSILE ASSAY IN SPENT FUEL

    Directory of Open Access Journals (Sweden)

    YONGDEOK LEE

    2013-12-01

    Full Text Available A future nuclear energy system is being developed at Korea Atomic Energy Research Institute (KAERI, the system involves a Sodium Fast Reactor (SFR linked with the pyro-process. The pyro-process produces a source material to fabricate a SFR fuel rod. Therefore, an isotopic fissile content assay is very important for fuel rod safety and SFR economics. A new technology for an analysis of isotopic fissile content has been proposed using a lead slowing down spectrometer (LSDS. The new technology has several features for a fissile analysis from spent fuel: direct isotopic fissile assay, no background interference, and no requirement from burnup history information. Several calculations were done on the designed spectrometer geometry: detection sensitivity, neutron energy spectrum analysis, neutron fission characteristics, self shielding analysis, and neutron production mechanism. The spectrum was well organized even at low neutron energy and the threshold fission chamber was a proper choice to get prompt fast fission neutrons. The characteristic fission signature was obtained in slowing down neutron energy from each fissile isotope. Another application of LSDS is for an optimum design of the spent fuel storage, maximization of the burnup credit and provision of the burnup code correction factor. Additionally, an isotopic fissile content assay will contribute to an increase in transparency and credibility for the utilization of spent fuel nuclear material, as internationally demanded.

  10. Development of output user interface software to support analysis

    Energy Technology Data Exchange (ETDEWEB)

    Wahanani, Nursinta Adi, E-mail: sintaadi@batan.go.id; Natsir, Khairina, E-mail: sintaadi@batan.go.id; Hartini, Entin, E-mail: sintaadi@batan.go.id [Center for Development of Nuclear Informatics - National Nuclear Energy Agency, PUSPIPTEK, Serpong, Tangerang, Banten (Indonesia)

    2014-09-30

    Data processing software packages such as VSOP and MCNPX are softwares that has been scientifically proven and complete. The result of VSOP and MCNPX are huge and complex text files. In the analyze process, user need additional processing like Microsoft Excel to show informative result. This research develop an user interface software for output of VSOP and MCNPX. VSOP program output is used to support neutronic analysis and MCNPX program output is used to support burn-up analysis. Software development using iterative development methods which allow for revision and addition of features according to user needs. Processing time with this software 500 times faster than with conventional methods using Microsoft Excel. PYTHON is used as a programming language, because Python is available for all major operating systems: Windows, Linux/Unix, OS/2, Mac, Amiga, among others. Values that support neutronic analysis are k-eff, burn-up and mass Pu{sup 239} and Pu{sup 241}. Burn-up analysis used the mass inventory values of actinide (Thorium, Plutonium, Neptunium and Uranium). Values are visualized in graphical shape to support analysis.

  11. Experimental validation of the DARWIN2.3 package for fuel cycle applications

    Energy Technology Data Exchange (ETDEWEB)

    San-Felice, L.; Eschbach, R.; Bourdot, P. [DEN, DER, CEA-Cadarache, F-13108 ST Paul-Lez-Durance (France); Tsilanizara, A.; Huynh, T. D. [DEN, DM2S, CEA-Saclay, F-91191 Gif sur Yvette (France); Ourly, H. [EDF, R and D, 1 av. General de Gaulle, F-92131 Clamart Cedex (France); Thro, J. F. [AREVA, Tour AREVA, F-92084 Paris la Defense (France)

    2012-07-01

    The DARWIN package, developed by the CEA and its French partners (AREVA and EDF) provides the required parameters for fuel cycle applications: fuel inventory, decay heat, activity, neutron, {gamma}, {alpha}, {beta} sources and spectrum, radiotoxicity. This paper presents the DARWIN2.3 experimental validation for fuel inventory and decay heat calculations on Pressurized Water Reactor (PWR). In order to validate this code system for spent fuel inventory a large program has been undertaken, based on spent fuel chemical assays. This paper deals with the experimental validation of DARWIN2.3 for the Pressurized Water Reactor (PWR) Uranium Oxide (UOX) and Mixed Oxide (MOX) fuel inventory calculation, focused on the isotopes involved in Burn-Up Credit (BUC) applications and decay heat computations. The calculation - experiment (C/E-1) discrepancies are calculated with the latest European evaluation file JEFF-3.1.1 associated with the SHEM energy mesh. An overview of the tendencies is obtained on a complete range of burn-up from 10 to 85 GWd/t (10 to 60 GWcVt for MOX fuel). The experimental validation of the DARWIN2.3 package for decay heat calculation is performed using calorimetric measurements carried out at the Swedish Interim Spent Fuel Storage Facility for Pressurized Water Reactor (PWR) assemblies, covering a large burn-up (20 to 50 GWd/t) and cooling time range (10 to 30 years). (authors)

  12. Comparison of VNEM to measured data from Ringhals unit 3. (Phase 3)

    Energy Technology Data Exchange (ETDEWEB)

    Tsuiki, M.; Mullet, S. (Institute for Energy Technology, OECD Halden Project, Kjeller (Norway))

    2011-01-15

    1. PWR. Comparisons have been made of a PWR core simulator CYGNUS with VNEM neutronics module to the measured data obtained from Ringhals unit 3 NPP through the cycle 1A (core average burnup = 0 through 10,507MWD/MT). The results can be summarized as: core eigenvalue = 0.99937 +/- 0.00086 before intermediate 5 months shutdown core eigenvalue = 0.99647 +/- 0.00029 after intermediate 5 months shutdown. The reason of core eigenvalue drop after the intermediate shutdown is estimated to be the build-up of fissile elements during the long shutdown. A calculation model to track some important isotopes in addition to Xe135 and Sm149 (these isotopes are tracked in the present version of CYGNUS) has to be implemented. As for the comparison of the neutron detector readings, the agreement was excellent throughout the cycle 1A as observed in Phase 1 and 2 (2008, 2009). The burnup tilt effect was not observed during the cycle 1A. The verification of the burnup tilt model of CYGNUS will be performed in the next phase of the project. 2. BWR. A preliminary 2D numerical benchmarking was performed for BWR cores. The problems were generated imitating the NEACRP MOX PWR 2D benchmark problems. The results of comparisons of VNEM to a reference transport code (FCM2D), based on the method of characteristics, were as good as those obtained in the case of PWR cores for similar benchmarking. (Author)

  13. A Stochastic Method for Estimating the Effect of Isotopic Uncertainties in Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    2001-08-24

    This report describes a novel approach developed at the Oak Ridge National Laboratory (ORNL) for the estimation of the uncertainty in the prediction of the neutron multiplication factor for spent nuclear fuel. This technique focuses on burnup credit, where credit is taken in criticality safety analysis for the reduced reactivity of fuel irradiated in and discharged from a reactor. Validation methods for burnup credit have attempted to separate the uncertainty associated with isotopic prediction methods from that of criticality eigenvalue calculations. Biases and uncertainties obtained in each step are combined additively. This approach, while conservative, can be excessive because of a physical assumptions employed. This report describes a statistical approach based on Monte Carlo sampling to directly estimate the total uncertainty in eigenvalue calculations resulting from uncertainties in isotopic predictions. The results can also be used to demonstrate the relative conservatism and statistical confidence associated with the method of additively combining uncertainties. This report does not make definitive conclusions on the magnitude of biases and uncertainties associated with isotopic predictions in a burnup credit analysis. These terms will vary depending on system design and the set of isotopic measurements used as a basis for estimating isotopic variances. Instead, the report describes a method that can be applied with a given design and set of isotopic data for estimating design-specific biases and uncertainties.

  14. Characterization of Delayed-Particle Emission Signatures for Pyroprocessing. Part 1: ABTR Fuel Assembly.

    Energy Technology Data Exchange (ETDEWEB)

    Durkee, Jr., Joe W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-06-19

    A three-part study is conducted using the MCNP6 Monte Carlo radiation-transport code to calculate delayed-neutron (DN) and delayed-gamma (DG) emission signatures for nondestructive assay (NDA) metal-fuel pyroprocessing. In Part 1, MCNP6 is used to produce irradiation-induced used nuclear fuel (UNF) isotopic inventories for an Argonne National Laboratory (ANL) Advanced Burner Test Reactor (ABTR) preconceptual design fuel assembly (FA) model. The initial fuel inventory consists of uranium mixed with light-water-reactor transuranic (TRU) waste and 10 wt% zirconium (U-LWR-SFTRU-10%Zr). To facilitate understanding, parametric evaluation is done using models for 3% and 5% initial 235U a% enrichments, burnups of 5, 10, 15, 20, 30, …, 120 GWd/MTIHM, and 3-, 5-, 10-, 20-, and 30- year cooling times. Detailed delayed-particle radioisotope source terms for the irradiate FA are created using BAMF-DRT and SOURCES3A. Using simulation tallies, DG activity ratios (DGARs) are developed for 134Cs/137Cs 134Cs/154Eu, and 154Eu/137Cs markers as a function of (1) burnup and (2) actinide mass, including elemental uranium, neptunium, plutonium, americium, and curium. Spectral-integrated DN emission is also tallied. The study reveals a rich assortment of DGAR behavior as a function of DGAR type, enrichment, burnup, and cooling time. Similarly, DN emission plots show variation as a function of burnup and of actinide mass. Sensitivity of DGAR and DN signatures to initial 235U enrichment, burnup, and cooling time is evident. Comparisons of the ABTR radiation signatures and radiation signatures previously reported for a generic Westinghouse oxide-fuel assembly indicate that there are pronounced differences in the ABTR and Westinghouse oxide-fuel DN and DG signatures. These differences are largely attributable to the initial TRU inventory in the ABTR fuel. The actinide and nonactinide inventories for the

  15. The effects of spectral shift absorbers on the design and safety of fast spectrum space reactors

    Science.gov (United States)

    King, Jeffrey Charles

    Spectral Shift Absorbers (SSAs) are incorporated into space reactors to maintain them sufficiently subcritical when submerged in seawater or wet sand and subsequently flooded, following a launch abort accident. The effect of four SSAs (samarium-149, europium-151, gadolinium-155, and gadolinium-157) on the submersion criticality, operation, and temperature reactivity feedback of the thermal spectrum reactors developed in the Systems for Nuclear Auxilary Power (SNAP) program is extensively documented. Recent work on SSAs in fast spectrum space reactors, preferred for compactness and higher powers, has focused on rhenium as the primary SSA. In addition to identifying additional SSAs, the present work investigates the effects of SSAs on the overall size and mass, temperature reactivity feedback, and operational lifetime of fast spectrum space reactors. The fast spectrum S4 reactor has a sectored Mo-14%Re solid-core, loadedwith UN fuel, cooled by He-30%Xe, and designed to avoid single point failures at a steady thermal power of 550 kWth. The addition of SSAs to the reactor core increases the fuel enrichment and decreases the size and mass of the reactor and the radiation shadow shield. SSA additions of boron-10, europium-151, gadolinium-155 and iridium result in the smallest and lightest S4 reactors. The effects of SSA additions on the operational lifetime and the temperature and burnup reactivity coefficients of the S^4 reactor are studied. An increasein fuel enrichment with SSAs markedly increases the operational lifetime by decreasing the burnup reactivity coefficient with only a slight decrease in the temperature reactivity feedback coefficient. With no SSAs, the UN fuel enrichment is lowest (58.5 wt%), the temperature and burnup reactivity coefficients are the highest (-0.2709 ¢/K and -1.3470 /atom%), and the estimated operating lifetime is the shortest (7.6 years). The temperature and burnup reactivity coefficients decrease to -0.2649 ¢/K and -1.0230 /atom%, and

  16. Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    Carmack, W. J.; Chichester, H. M.; Porter, D. L.; Wootan, D. W.

    2016-05-01

    Abstract The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important potential comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The irradiations were the beginning tests to qualify U-10wt%Zr as a driver fuel for FFTF. The FFTF core, with a 91.4 cm tall fuel column and a chopped cosine neutron flux profile, operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This places the peak fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in previous EBR-II experiments that had a 32-cm height core. The MFF-3 and MFF-5 qualification assemblies operated in FFTF to >10 at% burnup, and performed very well with no cladding breaches. The MFF-3 assembly operated to 13.8 at% burnup with a peak inner cladding temperature of 643°C, and the MFF-5 assembly operated to 10.1 at% burnup with a peak inner cladding temperature of 651°C. Because of the very high operating temperatures for both the fuel and the cladding, data from the MFF assemblies are most comparable to the data obtained from the EBR-II X447 experiment, which experienced two pin breaches. The X447 breaches were strongly influenced by a large amount of fuel/cladding chemical interaction (FCCI). The MFF pins benefitted from different axial locations of high burnup and peak cladding temperature, which helped to reduce interdiffusion between rare earth fission products and stainless steel cladding. Post-irradiation examination evidence illustrates this advantage. Comparing other performance data of the long MFF pins to prior EBR-II test data, the MFF fuel inside the cladding grew less axially, and the gas release data did not reveal a definitive difference.

  17. HTR Spherical Super Lattice Model for Equilibrium Fuel Cycle Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Gray S. Cahng

    2005-09-01

    Advanced High Temperature gas-cooled Reactors (HTR) currently being developed (GFR, VHTR - Very High Temperature gas-cooled Reactor, PBMR, and GT-MHR) are able to achieve a simplification of safety through reliance on innovative features and passive systems. One of the innovative features in these HTRs is reliance on ceramic-coated fuel particles to retain the fission products even under extreme accident conditions. The effect of the random fuel kernel distribution in the fuel pebble / block is addressed through the use of the Dancoff correction factor in the resonance treatment. In addition, the Dancoff correction factor is a function of burnup and fuel kernel packing factor, which requires that the Dancoff correction factor be updated during Equilibrium Fuel Cycle (EqFC) analysis. Although HTR fuel is rather homogeneously dispersed in the fuel graphite matrix, the heterogeneity effects in between fuel kernels and pebbles cannot be ignored. The double-heterogeneous lattice model recently developed at the Idaho National Engineering and Environmental Laboratory (INEEL) contains tens of thousands of cubic fuel kernel cells, which makes it very difficult to deplete the fuel, kernel by kernel (KbK), for the EqFC analysis. In addition, it is not possible to preserve the cubic size and packing factor in a spherical fuel pebble. To avoid these difficulties, a newly developed and validated HTR pebble-bed Kernel-by-Kernel spherical (KbK-sph) model, has been developed and verified in this study. The objective of this research is to introduce the KbK-sph model and super whole Pebble lattice model (PLM). The verified double-heterogeneous KbK-sph and pebble homogeneous lattice model (HLM) are used for the fuel burnup chracteristics analysis and important safety parameters validation. This study summarizes and compares the KbK-sph and HLM burnup analyzed results. Finally, we discus the Monte-Carlo coupling with a fuel depletion and buildup code - Origen-2 as a fuel burnup

  18. ORNL Interim Progress Report on Hydride Reorientation CIRFT Tests

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yan, Yong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-10-28

    A systematic study of H. B. Robinson (HBR) high burnup spent nuclear fuel (SNF) vibration integrity was performed in Phase I project under simulated transportation environments, using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot cell testing technology developed at Oak Ridge National Laboratory in 2013–14. The data analysis on the as-irradiated HBR SNF rods demonstrated that the load amplitude is the dominant factor that controls the fatigue life of bending rods. However, previous studies have shown that the hydrogen content and hydride morphology has an important effect on zirconium alloy mechanical properties. To address the effect of radial hydrides in SNF rods, in Phase II a test procedure was developed to simulate the effects of elevated temperatures, pressures, and stresses during transfer-drying operations. Pressurized and sealed fuel segments were heated to the target temperature for a preset hold time and slow-cooled at a controlled rate. The procedure was applied to both non-irradiated/prehydrided and high-burnup Zircaloy-4 fueled cladding segments using the Nuclear Regulatory Commission-recommended 400°C maximum temperature limit at various cooling rates. Before testing high-burnup cladding, four out-of-cell tests were conducted to optimize the hydride reorientation (R) test condition with pre-hydride Zircaloy-4 cladding, which has the same geometry as the high burnup fuel samples. Test HR-HBR#1 was conducted at the maximum hoop stress of 145 MPa, at a 400°C maximum temperature and a 5°C/h cooling rate. On the other hand, thermal cycling was performed for tests HR-HBR#2, HR-HBR#3, and HR-HBR#4 to generate more radial hydrides. It is clear that thermal cycling increases the ratio of the radial hydride to circumferential hydrides. The internal pressure also has a significant effect on the radial hydride morphology. This report describes a procedure and experimental results of the four out-of-cell hydride reorientation tests of

  19. Underwater characterization of control rods for waste disposal using SMOPY

    Energy Technology Data Exchange (ETDEWEB)

    Gallozzi-Ulmann, A.; Couturier, P.; Amgarou, K.; Rothan, D.; Menaa, N. [CANBERRA France,1 rue des Herons, 78182 ST Quentin Yvelines Cedex (France); Chard, P. [CANBERRA UK, Lower Dunbeath House, Forss Business Park, Thurso, Caithness KW14 7UZ (United Kingdom)

    2015-07-01

    Storage of spent fuel assemblies in cooling ponds requires careful control of the geometry and proximity of adjacent assemblies. Measurement of the fuel burnup makes it possible to optimise the storage arrangement of assemblies taking into account the effect of the burnup on the criticality safety margins ('burnup credit'). Canberra has developed a measurement system for underwater measurement of spent fuel assemblies. This system, known as 'SMOPY', performs burnup measurements based on gamma spectroscopy (collimated CZT detector) and neutron counting (fission chamber). The SMOPY system offers a robust and waterproof detection system as well as the needed capability of performing radiometric measurements in the harsh high dose - rate environments of the cooling ponds. The gamma spectroscopy functionality allows powerful characterization measurements to be performed, in addition to burnup measurement. Canberra has recently performed waste characterisation measurements at a Nuclear Power Plant. Waste activity assessment is important to control costs and risks of shipment and storage, to ensure that the activity level remains in the range allowed by the facility, and to declare activity data to authorities. This paper describes the methodology used for the SMOPY measurements and some preliminary results of a radiological characterisation of AIC control rods. After describing the features and normal operation of the SMOPY system, we describe the approach used for establishing an optimum control rod geometric scanning approach (optimum count time and speed) and the method of the gamma spectrometry measurements as well as neutron check measurements used to verify the absence of neutron sources in the waste. We discuss the results obtained including {sup 60}Co, {sup 110m}Ag and {sup 108m}Ag activity profiles (along the length of the control rods) and neutron results including Total Measurement Uncertainty evaluations. Full self-consistency checks were

  20. Strategy for Used Fuel Acquisition

    Energy Technology Data Exchange (ETDEWEB)

    Steven C. Marschman; Chris Rusch

    2013-09-01

    The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technology, has established the Used Fuel Disposition Campaign (UFDC) to conduct the research and development activities related to storage, transportation, and disposal of used nuclear fuel and high-level radioactive waste. The mission of the UFDC is to identify alternatives and conduct scientific research and technology development to enable storage, transportation and disposal of used nuclear fuel (UNF) and wastes generated by existing and future nuclear fuel cycles. The Storage and Transportation staffs within the UFDC are responsible for addressing issues regarding the extended or long-term storage of UNF and its subsequent transportation. The near-term objectives of the Storage and Transportation task are to use a science-based approach to develop the technical bases to support the continued safe and secure storage of UNF for extended periods, subsequent retrieval, and transportation. While both wet and dry storage have been shown to be safe options for storing UNF, the focus of the program is on dry storage at reactor or centralized locations. Because limited information is available on the properties of high burnup fuel (exceeding 45 gigawatt-days per metric tonne of uranium [GWd/MTU]), and because much of the fuel currently discharged from today’s reactors exceeds this burnup threshold, a particular emphasis of this program is on high burnup fuels. Since high burnup used fuels have only been loaded into dry storage systems in the past decade or so, these materials are available to the UFDC for testing in only very limited quantities. Much of what is available has come via NRC testing programs. Some of these fuels may have achieved "high burnup," but that does not mean they were designed for high burnup use (e.g. lower enrichments, smaller plenum spaces, extra reactor cycles). The handling and transfer of these materials from utility to laboratory has not always been

  1. On-line interrogation of pebble bed reactor fuel using passive gamma-ray spectrometry

    Science.gov (United States)

    Chen, Jianwei

    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 burnup limit (˜80,000--100,000 MWD/MTU). Unlike the situation with conventional light water reactors (LWRs), depending solely on computational methods to perform in-core fuel management will be highly inaccurate. As a result, an on-line measurement approach becomes the only accurate method to assess whether a particular pebble has reached its end-of-life burnup limit. In this work, an investigation was performed to assess the feasibility of passive gamma-ray spectrometry assay as an approach for on-line interrogation of PBR fuel for the simultaneous determination of burnup and enrichment on a pebble-by-pebble basis. Due to the unavailability of irradiated or fresh pebbles, Monte Carlo simulations were used to study the gamma-ray spectra of the PBR fuel at various levels of burnup. A pebble depletion calculation was performed using the ORIGEN code, which yielded the gamma-ray source term that was introduced into the input of an MCNP simulation. The MCNP simulation assumed the use of a high-purity coaxial germanium detector. Due to the lack of one-group high temperature reactor cross sections for ORIGEN, a heterogeneous MCNP model was developed to describe a typical PBR core. Subsequently, the code MONTEBURNS was used to couple the MCNP model and ORIGEN. This approach allowed the development of the burnup-dependent, one-group spectral-averaged PBR cross sections to be used in the ORIGEN pebble depletion calculation. Based on the above studies, a relative approach for performing the measurements was established. The approach is based on using the relative activities of Np-239/I-132 in combination

  2. Advanced characterization of MIMAS MOX fuel microstructure to quantify the HBS formation

    Energy Technology Data Exchange (ETDEWEB)

    Bouloré, Antoine, E-mail: antoine.boulore@cea.fr [CEA, DEN, DEC Fuel Research Department, Cadarache, F13108 Saint-Paul-lez-Durance (France); Aufore, Laurence; Federici, Eric [CEA, DEN, DEC Fuel Research Department, Cadarache, F13108 Saint-Paul-lez-Durance (France); Blanpain, Patrick [AREVA NP SAS, 10 rue Juliette Récamier, F-69456 Lyon (France); Blachier, Rémi [EDF, SEPTEN, 12-14 Av. Dutrievoz, F-69628 Villeurbanne (France)

    2015-01-15

    Highlights: • An advanced characterization of MIMAS MOX fuel based only on fresh fuel pellet characterization. • A probabilistic approach to model the High Burnup Structure formation in oxide fuels. • Validation of the method by comparing to experimental data obtained on fuel irradiated in the Halden reactor. - Abstract: Fission gas behaviour in accidental situations is closely related to the location of fission gas before the accident. More precisely, most of the fission gas in intergranular position is released during the accident and HBS zones contribute a lot to this intergranular quantity. So a methodology to characterize the HBS zones a priori from examination of unirradiated pellet has been developed at CEA. Characterization of plutonium distribution in MIMAS MOX fresh fuel pellets can be performed by image analysis on 1 mm{sup 2} X-ray mappings of plutonium acquired using Electron Probe Micro Analysis (EPMA). The specific software developed to describe the fuel using Pu X-ray mapping (ANACONDA) has been improved in order to simulate the fission products (FP) production and recoil during a given irradiation of the fuel, taking into account the evolution of the plutonium due to neutron irradiation. This simulation results from calculations with our fuel performance code ALCYONE combined with image processing. The final result is a mapping of local burn-up, but also the distribution of the relative FP concentration as a function of the local burn-up. A validation of this simulation process has been done by comparing the simulated mapping of neodymium to one measured on the same fuel batch after irradiation. Using previous studies of mechanisms for HBS formation, a probabilistic criterion for HBS formation has been proposed, based on the EPMA measurements of the decrease of the xenon signal as a function of the local burn-up. Combining the simulated FP cartography with this probabilistic HBS formation criterion, it is possible to calculate the surface

  3. Nuclide Inventory Calculation Using MCNPX for Wolsung Unit 1 Reactor Decommissioning

    Energy Technology Data Exchange (ETDEWEB)

    Rabir, Mohamad Hairie; Noh, Kyoung Ho; Hah, Chang Joo [KEPCO International Nuclear Graduate School, Daejeon (Korea, Republic of)

    2014-05-15

    The CINDER90 computation process involves utilizing linear Markovian chains to determine the time dependent nuclide densities. The CINDER90 depletion algorithm is implemented the MCNPX code package. The coupled depletion process involves a Monte-Carlo steady-state reaction rate calculation linked to a deterministic depletion calculation. The process is shown in Fig.1. MCNPX runs a steady state calculation to determine the system eigenvalue collision densities, recoverable energies from fission and neutrons per fission events. In order to generate number densities for the next time step, the CINDER90 code takes the MCNPX generated values and performs a depletion calculation. MCNPX then takes the new number densities and caries out a new steady-stated calculation. The process repeats itself until the final time step. This paper describe the preliminary source term and nuclide inventory calculation of Candu single fuel channel using MCNPX, as a part of the activities to support the equilibrium core model development and decommissioning evaluation process of a Candu reactor. The aim of this study was to apply the MCNPX code for source term and nuclide inventory calculation of Candu single fuel channel. Nuclide inventories as a function of burnup will be used to model an equilibrium core for Candu reactor. The core lifetime neutron fluence obtained from the model is used to estimate radioactivity at the stage of decommisioning. In general, as expected, the actinides and fission products build up increase with increasing burnup. Despite the fact that the MCNPX code is still in development we can conclude that the code is capable of obtaining relevant results in burnup and source term calculation. It is recommended that in the future work, the calculation has to be verified on the basis of experimental data or comparison with other codes.

  4. Fission Product Monitoring of TRISO Coated Fuel For The Advanced Gas Reactor -1 Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Dawn M. Scates; John (Jack) K Hartwell; John B. Walter

    2008-09-01

    The US Department of Energy has embarked on a series of tests of TRISO-coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burn up of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B’s) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  5. Numerical simulation of radioisotope's dependency on containment performance for large dry PWR containment under severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Mehboob, Khurram, E-mail: khurramhrbeu@gmail.com [College of Nuclear Science and Technology, Harbin Engineering University, 145-31 Nantong Street, Nangang District, Harbin, Heilongjiang 150001 (China); Xinrong, Cao [College of Nuclear Science and Technology, Harbin Engineering University, 145-31 Nantong Street, Nangang District, Harbin, Heilongjiang 150001 (China); Ahmed, Raheel [College of Automation, Harbin Engineering University, 145-31 Nantong Street, Nangang District, Harbin, Heilongjiang 150001 (China); Ali, Majid [College of Nuclear Science and Technology, Harbin Engineering University, 145-31 Nantong Street, Nangang District, Harbin, Heilongjiang 150001 (China)

    2013-09-15

    Highlights: • Calculation and comparison of activity of BURN-UP code with ORIGEN2 code. • Development of SASTC computer code. • Radioisotopes dependency on containment ESFs. • Mitigation in atmospheric release with ESFs operation. • Variation in radioisotopes source term with spray flow and pH value. -- Abstract: During the core melt accidents large amount of fission products can be released into the containment building. These fission products escape into the environment to contribute in accident source term. The mitigation in environmental release is demanded for such radiological consequences. Thus, countermeasures to source term, mitigations of release of radioactivity have been studied for 1000 MWe PWR reactor. The procedure of study is divided into five steps: (1) calculation and verification of core inventory, evaluated by BURN-UP code, (2) containment modeling based on radioactivity removal factors, (3) selection of potential accidents initiates the severe accident, (4) calculation of release of radioactivity, (5) study the dependency of release of radioactivity on containment engineering safety features (ESFs) inducing mitigation. Loss of coolant accident (LOCA), small break LOCA and flow blockage accidents (FBA) are selected as initiating accidents. The mitigation effect of ESFs on source term has been studied against ESFs performance. Parametric study of release of radioactivity has been carried out by modeling and simulating the containment parameters in MATLAB, which takes BURN-UP outcomes as input along with the probabilistic data. The dependency of iodine and aerosol source term on boric and caustic acid spray has been determined. The variation in source term mitigation with the variation of containment spray flow rate and pH values have been studied. The variation in containment retention factor (CRF) has also been studied with the ESF performance. A rapid decrease in source term is observed with the increase in pH value.

  6. Minor Actinides Loading Optimization for Proliferation Resistant Fuel Design - BWR

    Energy Technology Data Exchange (ETDEWEB)

    G. S. Chang; Hongbin Zhang

    2009-09-01

    One approach to address the United States Nuclear Power (NP) 2010 program for the advanced light water reactor (LWR) (Gen-III+) intermediate-term spent fuel disposal need is to reduce spent fuel storage volume while enhancing proliferation resistance. One proposed solution includes increasing burnup of the discharged spent fuel and mixing minor actinide (MA) transuranic nuclides (237Np and 241Am) in the high burnup fuel. Thus, we can reduce the spent fuel volume while increasing the proliferation resistance by increasing the isotopic ratio of 238Pu/Pu. For future advanced nuclear systems, MAs are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. A typical boiling water reactor (BWR) fuel unit lattice cell model with UO2 fuel pins will be used to investigate the effectiveness of adding MAs (237Np and/or 241Am) to enhance proliferation resistance and improve fuel cycle performance for the intermediate-term goal of future nuclear energy systems. However, adding MAs will increase plutonium production in the discharged spent fuel. In this work, the Monte-Carlo coupling with ORIGEN-2.2 (MCWO) method was used to optimize the MA loading in the UO2 fuel such that the discharged spent fuel demonstrates enhanced proliferation resistance, while minimizing plutonium production. The axial averaged MA transmutation characteristics at different burnup were compared and their impact on neutronics criticality and the ratio of 238Pu/Pu discussed.

  7. Regulatory perspective on incomplete control rod insertions

    Energy Technology Data Exchange (ETDEWEB)

    Chatterton, M.

    1997-01-01

    The incomplete control rod insertions experienced at South Texas Unit 1 and Wolf Creek are of safety concern to the NRC staff because they represent potential precursors to loss of shutdown margin. Even before it was determined if these events were caused by the control rods or by the fuel there was an apparent correlation of the problem with high burnup fuel. It was determined that there was also a correlation between high burnup and high drag forces as well as with rod drop time histories and lack of rod recoil. The NRC staff initial actions were aimed at getting a perspective on the magnitude of the problem as far as the number of plants and the amount of fuel that could be involved, as well as the safety significance in terms of shutdown margin. As tests have been performed and data has been analyzed the focus has shifted more toward understanding the problem and the ways to eliminate it. At this time the staff`s understanding of the phenomena is that it was a combination of factors including burnup, power history and temperature. The problem appears to be very sensitive to these factors, the interaction of which is not clearly understood. The model developed by Westinghouse provides a possible explanation but there is not sufficient data to establish confidence levels and sensitivity studies involving the key parameters have not been done. While several fixes to the problem have been discussed, no definitive fixes have been proposed. Without complete understanding of the phenomena, or fixes that clearly eliminate the problem the safety concern remains. The safety significance depends on the amount of shutdown margin lost due to incomplete insertion of the control rods. Were the control rods to stick high in the core, the reactor could not be shutdown by the control rods and other means such as emergency boration would be required.

  8. Proof of concept experiments of the multi-isotope process monitor: An online, nondestructive, near real-time monitor for spent nuclear fuel reprocessing facilities

    Science.gov (United States)

    Orton, Christopher R.; Fraga, Carlos G.; Christensen, Richard N.; Schwantes, Jon M.

    2012-04-01

    Operators, national regulatory agencies and the IAEA will require the development of advanced technologies to efficiently control and safeguard nuclear material at increasingly large-scale nuclear recycling facilities. Ideally, the envisioned technologies would be capable of non-destructive, near real-time (NRT), autonomous process monitoring. This paper describes results from proof-of-principle experiments designed to test the multi-isotope process (MIP) monitor, a novel approach to monitoring and safeguarding reprocessing facilities. The MIP Monitor combines the detection of intrinsic gamma ray signatures emitted from process solutions with multivariate analysis to detect off-normal conditions in process streams nondestructively and in NRT. Commercial spent nuclear fuel of various irradiation histories was dissolved and separated using a PUREX-based batch solvent extraction. Extractions were performed at various nitric acid concentrations to mimic both normal and off-normal industrial plant operating conditions. Principal component analysis (PCA) was applied to the simulated gamma spectra to investigate pattern variations as a function of acid concentration, burnup and cooling time. Partial least squares (PLS) regression was applied to attempt to quantify both the acid concentration and burnup of the dissolved spent fuel during the initial separation stage of recycle. The MIP Monitor demonstrated sensitivity to induced variations of acid concentration, including the distinction of ±1.3 M variation from normal process conditions by way of PCA. Acid concentration was predicted using measurements from the organic extract and PLS resulting in predictions with <0.7 M relative error. Quantification of burnup levels from dissolved fuel spectra using PLS was demonstrated to be within 2.5% of previously measured values.

  9. The multi-isotope process monitor: Non-destructive, near-real-time nuclear safeguards monitoring at a reprocessing facility

    Science.gov (United States)

    Orton, Christopher Robert

    The IAEA will require advanced technologies to effectively safeguard nuclear material at envisioned large scale nuclear reprocessing plants. This dissertation describes results from simulations and experiments designed to test the Multi-Isotope Process (MIP) Monitor, a novel safeguards approach for process monitoring in reprocessing plants. The MIP Monitor combines the detection of intrinsic gamma ray signatures emitted from process solutions with multivariate analysis to detect off-normal conditions in process streams, nondestructively and in near-real time (NRT). Three different models were used to predict spent nuclear fuel composition, estimate chemical distribution during separation, and simulate spectra from a variety of gamma detectors in product and raffinate streams for processed fuel. This was done for fuel with various irradiation histories and under a variety of plant operating conditions. Experiments were performed to validate the results from the model. Three segments of commercial spent nuclear fuel with variations in burnup and cooling time were dissolved and subjected to a batch PUREX method to separate the uranium and plutonium from fission and activation products. Gamma spectra were recorded by high purity germanium (HPGe) and cadmium zinc telluride (CZT) detectors. Hierarchal Cluster Analysis (HCA) and Principal Component Analysis (PCA) were applied to spectra from both model and experiment to investigate spectral variations as a function of acid concentration, burnup level and cooling time. Partial Least Squares was utilized to extract quantitative information about process variables, such as acid concentration or burnup. The MIP Monitor was found to be sensitive to the induced variations of the process and was capable of extracting quantitative process information from the analyzed spectra.

  10. Evaluation of maximum allowable temperature inside basket of dry storage module for CANDU spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyung Ho; Yoon, Jeong Hyoun; Chae, Kyoung Myoung; Choi, Byung Il; Lee, Heung Young; Song, Myung Jae [Nuclear Environment Technology Institute, Taejon (Korea, Republic of); Cho, Gyu Seong [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2002-10-01

    This study provides a maximum allowable fuel temperature through a preliminary evaluation of the UO{sub 2} weight gain that may occur on a failed (breached sheathing) element of a fuel bundle. Intact bundles would not be affected as the UO{sub 2} would not be in contact with the air for the fuel storage basket. The analysis is made for the MACSTOR/KN-400 to be operated in Wolsong ambient air temperature conditions. The design basis fuel is a 6-year cooled fuel bundle that, on average has reached a burnup of 7,800 MWd/MTU. The fuel bundle considered for analysis is assumed to have a high burnup of 12,000 MWd/MTU and be located in a hot basket. The MACSTOR/KN-400 has the same air circuit as the MACSTOR and the air circuit will require a slightly higher temperature difference to exit the increased heat load. The maximum temperature of a high burnup bundle stored in the new MACSTOR/KN-400 is expected to be about 9 .deg. C higher than the fuel temperature of the MACSTOR at an equivalent constant ambient temperature. This temperature increase will in turn increase the UO{sub 2} weight gain from 0.06% (MACSTOR for Wolsong conditions) to an estimated 0.13% weight gain for the MACSTOR/KN-400. Compared to an acceptable UO{sub 2} weight gain of 0.6%, we are thus expecting to maintain a very acceptable safety factor of 4 to 5 for the new module against unacceptable stresses in the fuel sheathing. For the UO{sub 2} weight gain, the maximum allowable fuel temperature was shown by 164 .deg. C.

  11. Proof of concept experiments of the multi-isotope process monitor: An online, nondestructive, near real-time monitor for spent nuclear fuel reprocessing facilities

    Energy Technology Data Exchange (ETDEWEB)

    Orton, Christopher R., E-mail: christopher.orton@pnnl.gov [Pacific Northwest National Laboratory, 902 Battelle Boulevard, P.O. Box 999, Richland, WA 99354 (United States); Fraga, Carlos G., E-mail: carlos.fraga@pnnl.gov [Pacific Northwest National Laboratory, 902 Battelle Boulevard, P.O. Box 999, Richland, WA 99354 (United States); Christensen, Richard N., E-mail: christensen.3@osu.edu [The Ohio State University, 201W. 19th Avenue, Columbus, Ohio 43210 (United States); Schwantes, Jon M., E-mail: jon.schwantes@pnnl.gov [Pacific Northwest National Laboratory, 902 Battelle Boulevard, P.O. Box 999, Richland, WA 99354 (United States)

    2012-04-21

    Operators, national regulatory agencies and the IAEA will require the development of advanced technologies to efficiently control and safeguard nuclear material at increasingly large-scale nuclear recycling facilities. Ideally, the envisioned technologies would be capable of non-destructive, near real-time (NRT), autonomous process monitoring. This paper describes results from proof-of-principle experiments designed to test the multi-isotope process (MIP) monitor, a novel approach to monitoring and safeguarding reprocessing facilities. The MIP Monitor combines the detection of intrinsic gamma ray signatures emitted from process solutions with multivariate analysis to detect off-normal conditions in process streams nondestructively and in NRT. Commercial spent nuclear fuel of various irradiation histories was dissolved and separated using a PUREX-based batch solvent extraction. Extractions were performed at various nitric acid concentrations to mimic both normal and off-normal industrial plant operating conditions. Principal component analysis (PCA) was applied to the simulated gamma spectra to investigate pattern variations as a function of acid concentration, burnup and cooling time. Partial least squares (PLS) regression was applied to attempt to quantify both the acid concentration and burnup of the dissolved spent fuel during the initial separation stage of recycle. The MIP Monitor demonstrated sensitivity to induced variations of acid concentration, including the distinction of {+-}1.3 M variation from normal process conditions by way of PCA. Acid concentration was predicted using measurements from the organic extract and PLS resulting in predictions with <0.7 M relative error. Quantification of burnup levels from dissolved fuel spectra using PLS was demonstrated to be within 2.5% of previously measured values.

  12. Proof of Concept Experiments of the Multi-Isotope Process Monitor: An Online, Nondestructive, Near Real-Time Monitor for Spent Nuclear Fuel Reprocessing Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Orton, Christopher R.; Fraga, Carlos G.; Christensen, Richard; Schwantes, Jon M.

    2012-04-21

    Operators, national regulatory agencies and the IAEA will require the development of advanced technologies to efficiently control and safeguard nuclear material at increasingly large-scale nuclear recycling facilities. Ideally, the envisioned technologies would be capable of non-destructive, near-real-time (NRT), autonomous process monitoring. This paper describes results from proof-of-principle experiments designed to test the Multi-Isotope Process (MIP) Monitor, a novel approach to safeguarding reprocessing facilities. The MIP Monitor combines the detection of intrinsic gamma ray signatures emitted from process solutions with multivariate analysis to detect off-normal conditions in process streams nondestructively and in NRT. Commercial spent nuclear fuel of various irradiation histories was dissolved and separated using a PUREX-based batch solvent extraction. Extractions were performed at various nitric acid concentrations to mimic both normal and off-normal industrial plant operating conditions. Principal Component Analysis (PCA) was applied to the simulated gamma spectra to investigate pattern variations as a function of acid concentration, burnup and cooling time. Partial Least Squares (PLS) regression was applied to attempt to quantify both the acid concentration and burnup of the dissolved spent fuel during the initial separation stage of recycle. The MIP Monitor demonstrated sensitivity to induced variations of acid concentration, including the distinction of {+-} 1.3 M variation from normal process conditions by way of PCA. Acid concentration was predicted using measurements from the organic extract and PLS resulting in predictions with <0.7 M relative error. Quantification of burnup levels from dissolved fuel spectra using PLS was demonstrated to be within 2.5% of previously measured values.

  13. A Phased Development of Breed-and-Burn Reactors for Enhanced Nuclear Energy Sustainability

    Directory of Open Access Journals (Sweden)

    Ehud Greenspan

    2012-10-01

    Full Text Available Several options for designing fast reactors to operate in the Breed-and-Burn (B&B mode are compared and a strategy is outlined for early introduction of B&B reactors followed by a gradual increase in the fuel utilization of such reactors. In the first phase the fast reactor core will consist of a subcritical B&B blanket driven by a relatively small critical seed. As the required discharge burnup/radiation-damage to both driver and blanket fuel had already been proven, and as the depleted uranium fueled B&B blanket could generate close to 2/3 of the core power and will have very low fuel cycle cost, the deployment of such fast reactors could start in the near future. The second phase consists of deploying self-sustaining stationary wave B&B reactors. It will require development of fuel technology that could withstand peak burnups of ~30% and peak radiation damage to the cladding of ~550 dpa. The third phase requires development of a fuel reconditioning technology that will enable using the fuel up to an average burnup of ~50%—the upper bound permitted by neutron balance considerations when most of the fission products are not separated from the fuel. The increase in the uranium ore utilization relative to that provided by contemporary power reactors is estimated to be 20, 40 and 100 folds for, respectively, phase 1, 2 and 3. The energy value of the depleted uranium stockpiles (“waste” accumulated in the US is equivalent to, when used in the B&B reactors, up to 20 centuries of the total 2010 USA supply of electricity. Therefore, a successful development of B&B reactors could provide a great measure of energy sustainability and cost stability.

  14. Passive gamma analysis of the boiling-water-reactor assemblies

    Science.gov (United States)

    Vo, D.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S.; Trellue, H.; Vaccaro, S.

    2016-09-01

    This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden's Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative-Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: 137Cs, 154Eu, 134Cs, and to a lesser extent, 106Ru and 144Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.

  15. Passive gamma analysis of the boiling-water-reactor assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Vo, D., E-mail: ducvo@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM (United States); Favalli, A. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Schwalbach, P. [European Atomic Energy Community (EURATOM), Luxemburg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Tobin, S.; Trellue, H. [Los Alamos National Laboratory, Los Alamos, NM (United States); Vaccaro, S. [European Atomic Energy Community (EURATOM), Luxemburg (Luxembourg)

    2016-09-11

    This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden's Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative–Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: {sup 137}Cs, {sup 154}Eu, {sup 134}Cs, and to a lesser extent, {sup 106}Ru and {sup 144}Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.

  16. Thorium-based mixed oxide fuel in a pressurized water reactor: A feasibility analysis with MCNP

    Science.gov (United States)

    Tucker, Lucas Powelson

    This dissertation investigates techniques for spent fuel monitoring, and assesses the feasibility of using a thorium-based mixed oxide fuel in a conventional pressurized water reactor for plutonium disposition. Both non-paralyzing and paralyzing dead-time calculations were performed for the Portable Spectroscopic Fast Neutron Probe (N-Probe), which can be used for spent fuel interrogation. Also, a Canberra 3He neutron detector's dead-time was estimated using a combination of subcritical assembly measurements and MCNP simulations. Next, a multitude of fission products were identified as candidates for burnup and spent fuel analysis of irradiated mixed oxide fuel. The best isotopes for these applications were identified by investigating half-life, photon energy, fission yield, branching ratios, production modes, thermal neutron absorption cross section and fuel matrix diffusivity. 132I and 97Nb were identified as good candidates for MOX fuel on-line burnup analysis. In the second, and most important, part of this work, the feasibility of utilizing ThMOX fuel in a pressurized water reactor (PWR) was first examined under steady-state, beginning of life conditions. Using a three-dimensional MCNP model of a Westinghouse-type 17x17 PWR, several fuel compositions and configurations of a one-third ThMOX core were compared to a 100% UO2 core. A blanket-type arrangement of 5.5 wt% PuO2 was determined to be the best candidate for further analysis. Next, the safety of the ThMOX configuration was evaluated through three cycles of burnup at several using the following metrics: axial and radial nuclear hot channel factors, moderator and fuel temperature coefficients, delayed neutron fraction, and shutdown margin. Additionally, the performance of the ThMOX configuration was assessed by tracking cycle length, plutonium destroyed, and fission product poison concentration.

  17. Nuclear data requirements for the ADS conceptual design EFIT: Uncertainty and sensitivity study

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Herranz, N., E-mail: nuria@din.upm.e [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid (Spain); Instituto de Fusion Nuclear, Universidad Politecnica de Madrid (Spain); Cabellos, O. [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid (Spain); Instituto de Fusion Nuclear, Universidad Politecnica de Madrid (Spain); Alvarez-Velarde, F. [CIEMAT (Spain); Sanz, J. [Instituto de Fusion Nuclear, Universidad Politecnica de Madrid (Spain); Departamento de Ingenieria Energetica, UNED (Spain); Gonzalez-Romero, E.M. [CIEMAT (Spain); Juan, J. [Laboratorio de Estadistica, Universidad Politecnica de Madrid (Spain)

    2010-11-15

    In this paper, we assess the impact of activation cross-section uncertainties on relevant fuel cycle parameters for a conceptual design of a modular European Facility for Industrial Transmutation (EFIT) with a 'double strata' fuel cycle. Next, the nuclear data requirements are evaluated so that the parameters can meet the assigned design target accuracies. Different discharge burn-up levels are considered: a low burn-up, corresponding to the equilibrium cycle, and a high burn-up level, simulating the effects on the fuel of the multi-recycling scenario. In order to perform this study, we propose a methodology in two steps. Firstly, we compute the uncertainties on the system parameters by using a Monte Carlo simulation, as it is considered the most reliable approach to address this problem. Secondly, the analysis of the results is performed by a sensitivity technique, in order to identify the relevant reaction channels and prioritize the data improvement needs. Cross-section uncertainties are taken from the EAF-2007/UN library since it includes data for all the actinides potentially present in the irradiated fuel. Relevant uncertainties in some of the fuel cycle parameters have been obtained, and we conclude with recommendations for future nuclear data measurement programs, beyond the specific results obtained with the present nuclear data files and the limited available covariance information. A comparison with the uncertainty and accuracy analysis recently published by the WPEC-Subgroup26 of the OECD using BOLNA covariance matrices is performed. Despite the differences in the transmuter reactor used for the analysis, some conclusions obtained by Subgroup26 are qualitatively corroborated, and improvements for additional cross sections are suggested.

  18. Improving Natural Uranium Utilization By Using Thorium in Low Moderation PWRs - A Preliminary Neutronic Scoping Study

    Energy Technology Data Exchange (ETDEWEB)

    Gilles Youinou; Ignacio Somoza

    2010-10-01

    The Th-U fuel cycle is not quite self-sustainable when used in water-cooled reactors and with fuel burnups higher than a few thousand of MWd/t characteristic of CANDU reactors operating with a continuous refueling. For the other industrially mature water-cooled reactors (i.e. PWRs and BWRs) it is economically necessary that the fuel has enough reactivity to reach fuel burnups of the order of a few tens of thousand of MWd/t. In this particular case, an additional input of fissile material is necessary to complement the bred fissile U-233. This additional fissile material could be included in the form of Highly Enriched Uranium (HEU) at the fabrication of the Th-U fuel. The objective of this preliminary neutronic scoping study is to determine (1) how much HEU and, consequently, how much natural uranium is necessary in such Th-U fuel cycle with U recycling and (2) how much TRansUranics (TRU=Pu, Np, Am and Cm) are produced. These numbers are then compared with those of a standard UO2 PWR. The thorium reactors considered have a homogeneous hexagonal lattice made up of the same (Th-U)O2 pins. Furthermore, at this point, we are not considering the use of blankets inside or outside the core. The lattice pitch has been varied to estimate the effect of the water-to-fuel volume ratio, and light water as well as heavy water have been considered. For most cases, an average burnup at discharge of 45,000 MWd/t has been considered.

  19. Experimental study of the burned of nuclear fuel by the gamma spectroscopy method; Estudio experimental del quemado de combustible nuclear por el metodo de espectrometria gamma

    Energy Technology Data Exchange (ETDEWEB)

    Amador V, P.

    2009-07-01

    Accurate information on nuclear fuel burnup is of vital importance in reactor operation, fuel management and fuel-characteristics studies. Conventionally fuel management of the TRIGA III Reactor from the National Institute of Nuclear Research (ININ) is done through the thermal balance method (management) of the power generated during reactor operation, since it is known that with 1.24 grams of {sup 235}U is possible to generate a power or 1 MW per day during the reactor operation. On the other hand, it is possible to calculate the operation time in days during a power of 1 MW with the help of the data registered in logs. With the information just mentioned one can calculate the quantity of {sup 235}U consumed in the fuel during a complete period of irradiation. In order to compare and prove that the burnup values, calculated through the thermal balance method, are correct, the ININ implemented, for the first time, the gamma-ray spectroscopy method as an experimental technique to calculate the burnup of several fuel elements. Gamma-ray spectroscopy is a nondestructive method, so that the integrity of the fuel element is not affected which is of great importance. Since there is a direct relation between the activity of {sup 137}Cs contained in the fuel elements and a series of constants which are unique for the radioisotope and for the high resolution system, the problem just simplifies in measuring the {sup 137}Cs activities. Furthermore the {sup 137}Cs concentration equation was developed theoretically and I wrote a computer program (AMAVAL) in Fortran. The task of this program is to calculate the concentrations and the activity through the use of the equation just mentioned and the history of each fuel element. The purpose of this is to compare and validate the experimental activities with the theoretical ones for each fuel element. (Author)

  20. Light water reactor fuel analysis code FEMAXI-V (Ver.1)

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-09-01

    A light water fuel analysis code FEMAXI-V is an advanced version which has been produced by integrating FEMAXI-IV(Ver.2), high burn-up fuel code EXBURN-I, and a number of functional improvements and extensions, to predict fuel rod behavior in normal and transient (not accident) conditions. The present report describes in detail the basic theories and structure, models and numerical solutions applied, improvements and extensions, and the material properties adopted in FEMAXI-V(Ver.1). FEMAXI-V deals with a single fuel rod. It predicts thermal and mechanical response of fuel rod to irradiation, including FP gas release. The thermal analysis predicts rod temperature distribution on the basis of pellet heat generation, changes in pellet thermal conductivity and gap thermal conductance, (transient) change in surface heat transfer to coolant, using radial one-dimensional geometry. The heat generation density profile of pellet can be determined by adopting the calculated results of burning analysis code. The mechanical analysis performs elastic/plastic, creep and PCMI calculations by FEM. The FP gas release model calculates diffusion of FP gas atoms and accumulation in bubbles, release and increase in internal pressure of rod. In every analysis, it is possible to allow some materials properties and empirical equations to depend on the local burnup or heat flux, which enables particularly analysis of high burnup fuel behavior and boiling transient of BWR rod. In order to facilitate effective and wide-ranging application of the code, formats and methods of input/output of the code are also described, and a sample output in an actual form is included. (author)

  1. Fuel utilization improvements in a once-through PWR fuel cycle. Final report on Task 6

    Energy Technology Data Exchange (ETDEWEB)

    Dabby, D.

    1979-06-01

    In studying the position of the United States Department of Energy, Non-proliferation Alternative Systems Assessment Program, this report determines the uranium saving associated with various improvement concepts applicable to a once-through fuel cycle of a standard four-loop Westinghouse Pressurized Water Reactor. Increased discharged fuel burnup from 33,000 to 45,000 MWD/MTM could achieve a 12% U/sub 3/O/sub 8/ saving by 1990. Improved fuel management schemes combined with coastdown to 60% power, could result in U/sub 3/O/sub 8/ savings of 6%.

  2. Bounce-free Spherical Hydrodynamic Implosion

    CERN Document Server

    Kagan, Grigory; Hsu, Scott C; Awe, Thomas J

    2011-01-01

    In a bounce-free spherical hydrodynamic implosion, the post-stagnation hot core plasma does not expand against the imploding flow. Such an implosion scheme has the advantage of improving the dwell time of the burning fuel, resulting in a higher fusion burn-up fraction. The existence of bounce-free spherical implosions is demonstrated by explicitly constructing a family of self-similar solutions to the spherically symmetric ideal hydrodynamic equations. When applied to a specific example of plasma liner driven magneto-inertial fusion, the bounce-free solution is found to produce at least a factor of four improvement in dwell time and fusion energy gain.

  3. Application of Origen2.1 in the decay photon spectrum calculation of spallation products

    OpenAIRE

    Hong, Shuang; Yang, Yong-Wei; Xu, Hu-Shan; Meng, Hai-Yan; Zhang, Lu; Liu, Zhao-Qing; Gao, Yu-Cui; Chen, Kang

    2016-01-01

    Origen2.1 is a widely used computer code for calculating the burnup, decay, and processing of radioactive materials. However, the nuclide library of Origen2.1 is used for existing reactors like pressurized water reactor, to calculate the photon spectrum released by the decay of spallation products, we have made specific libraries for the ADS tungsten spallation target, based on the results given by a Monte Carlo code: FLUKA. All the data used to make the Origen2.1 libraries is obtained from N...

  4. Sensitivity analysis on various parameters for lattice analysis of DUPIC fuel with WIMS-AECL code

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Gyu Hong; Choi, Hang Bok; Park, Jee Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    The code WIMS-AECL has been used for the lattice analysis of DUPIC fuel. The lattice parameters calculated by the code is sensitive to the choice of number of parameters, such as the number of tracking lines, number of condensed groups, mesh spacing in the moderator region, other parameters vital to the calculation of probabilities and burnup analysis. We have studied this sensitivity with respect to these parameters and recommend their proper values which are necessary for carrying out the lattice analysis of DUPIC fuel.

  5. Fission product inventory calculation by a CASMO/ORIGEN coupling program

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Heon; Kim, Jong Kyung [Hanyang University, Seoul (Korea, Republic of); Choi, Hang Bok; Roh, Gyu Hong; Jung, In Ha [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    A CASMO/ORIGEN coupling utility program was developed to predict the composition of all the fission products in spent PWR fuels. The coupling program reads the CASMO output file, modifies the ORIGEN cross section library and reconstructs the ORIGEN input file at each depletion step. In ORIGEN, the burnup equation is solved for actinides and fission products based on the fission reaction rates and depletion flux of CASMO. A sample calculation has been performed using a 14 x 14 PWR fuel assembly and the results are given in this paper. 3 refs., 1 fig., 1 tab. (Author)

  6. The effect of the composition of plutonium loaded on the reactivity change and the isotopic composition of fuel produced in a fast reactor

    Science.gov (United States)

    Blandinskiy, V. Yu.

    2014-12-01

    This paper presents the results of a numerical investigation into burnup and breeding of nuclides in metallic fuel consisting of a mixture of plutonium and depleted uranium in a fast reactor with sodium coolant. The feasibility of using plutonium contained in spent nuclear fuel from domestic thermal reactors and weapons-grade plutonium is discussed. It is shown that the largest production of secondary fuel and the least change in the reactivity over the reactor lifetime can be achieved when employing plutonium contained in spent nuclear fuel from a reactor of the RBMK-1000 type.

  7. Annual report of nuclear code evaluation committee for fiscal 2000 year

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-03-01

    In this report, research results discussed in fiscal 2000 year at Nuclear Code Evaluation Committee of Nuclear Code Research Committee were summarized. In 2000, papers mainly on the three topics of (1) present status of burnup credit evaluation methods, (2) issues concerning convergence of criticality calculation and (3) estimation methods for errors associated with criticality calculation based on nuclear data covariance file, are presented and discussed. These results are sorted to grasp the present status of related technology and described in this report. (author)

  8. Development of core fuel management code system for WWER-type reactors

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    In this article, a core fuel management program for hexagonal pressurized water type WWER reactors (CFMHEX) has been developed, which is based on advanced three-dimensional nodal method and integrated with thermal hydraulic code to realize the coupling of neutronics and thermal-hydraulics. In CFMHEX, all these feedback effects such as burnup, power distribution, moderator density, and control rod insertion are considered. The verification and validation of the code system have been examined through the IAEA WWER-1000-type Kalinin NPP benchmark problem. The numerical results are in good agreement with measurements and are close to those of other international institutes.

  9. Characterization of intergranular fission gas bubbles in U-Mo fuel.

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y. S.; Hofman, G.; Rest, J.; Shevlyakov, G. V.; Nuclear Engineering Division; SSCR RIAR

    2008-04-14

    This report can be divided into two parts: the first part, which is composed of sections 1, 2, and 3, is devoted to report the analyses of fission gas bubbles; the second part, which is in section 4, is allocated to describe the mechanistic model development. Swelling data of irradiated U-Mo alloy typically show that the kinetics of fission gas bubbles is composed of two different rates: lower initially and higher later. The transition corresponds to a burnup of {approx}0 at% U-235 (LEU) or a fission density of {approx}3 x 10{sup 21} fissions/cm{sup 3}. Scanning electron microscopy (SEM) shows that gas bubbles appear only on the grain boundaries in the pretransition regime. At intermediate burnup where the transition begins, gas bubbles are observed to spread into the intragranular regions. At high burnup, they are uniformly distributed throughout fuel. In highly irradiated U-Mo alloy fuel large-scale gas bubbles form on some fuel particle peripheries. In some cases, these bubbles appear to be interconnected and occupy the interface region between fuel and the aluminum matrix for dispersion fuel, and fuel and cladding for monolithic fuel, respectively. This is a potential performance limit for U-Mo alloy fuel. Microscopic characterization of the evolution of fission gas bubbles is necessary to understand the underlying phenomena of the macroscopic behavior of fission gas swelling that can lead to a counter measure to potential performance limit. The microscopic characterization data, particularly in the pre-transition regime, can also be used in developing a mechanistic model that predicts fission gas bubble behavior as a function of burnup and helps identify critical physical properties for the future tests. Analyses of grain and grain boundary morphology were performed. Optical micrographs and scanning electron micrographs of irradiated fuel from RERTR-1, 2, 3 and 5 tests were used. Micrographic comparisons between as-fabricated and as-irradiated fuel revealed

  10. Sustainability of thorium-uranium in pebble-bed fluoride salt-cooled high temperature reactor

    OpenAIRE

    Zhu Guifeng; Zou Yang; Xu Hongjie

    2016-01-01

    Sustainability of thorium fuel in a Pebble-Bed Fluoride salt-cooled High temperature Reactor (PB-FHR) is investigated to find the feasible region of high discharge burnup and negative Flibe (2LiF-BeF2) salt Temperature Reactivity Coefficient (TRC). Dispersion fuel or pellet fuel with SiC cladding and SiC matrix is used to replace the tristructural-isotropic (TRISO) coated particle system for increasing fuel loading and decreasing excessive moderation. To analyze the neutronic characteristics,...

  11. Development of computational methods to describe the mechanical behavior of PWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Wanninger, Andreas; Seidl, Marcus; Macian-Juan, Rafael [Technische Univ. Muenchen, Garching (Germany). Dept. of Nuclear Engineering

    2016-10-15

    To investigate the static mechanical response of PWR fuel assemblies (FAs) in the reactor core, a structural FA model is being developed using the FEM code ANSYS Mechanical. To assess the capabilities of the model, lateral deflection tests are performed for a reference FA. For this purpose we distinguish between two environments, in-laboratory and in-reactor for different burn-ups. The results are in qualitative agreement with experimental tests and show the stiffness decrease of the FAs during irradiation in the reactor core.

  12. Used Nuclear Fuel Loading and Structural Performance Under Normal Conditions of Transport- Demonstration of Approach and Results on Used Fuel Performance Characterization

    Energy Technology Data Exchange (ETDEWEB)

    Adkins, Harold [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Geelhood, Ken [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Koeppel, Brian [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Coleman, Justin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bignell, John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Flores, Gregg [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wang, Jy-An [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sanborn, Scott [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Spears, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States); Klymyshyn, Nick [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-09-30

    This document addresses Oak Ridge National Laboratory milestone M2FT-13OR0822015 Demonstration of Approach and Results on Used Nuclear Fuel Performance Characterization. This report provides results of the initial demonstration of the modeling capability developed to perform preliminary deterministic evaluations of moderate-to-high burnup used nuclear fuel (UNF) mechanical performance under normal conditions of storage (NCS) and normal conditions of transport (NCT) conditions. This report also provides results from the sensitivity studies that have been performed. Finally, discussion on the long-term goals and objectives of this initiative are provided.

  13. 加速器驱动的次临界10MW气冷快堆物理方案研究%Physical Scheme Study of 10 MW Accelerator-driven Sub-critical Gas-cooled Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    秦长平; 顾龙; 李金阳

    2013-01-01

    本文采用MCNPX与ORIGEN耦合的COUPLE2.0程序对加速器驱动的次临界气冷快堆进行了物理方案的设计和研究。计算得到了该方案350d燃耗期间的kef、质子效率、加速器束流强度、1500K温度下的多普勒系数、功率峰因子等参数,并研究了该方案的安全特性;通过计算得到了该方案燃耗并分析了该方案的嬗变能力。结果表明,该方案在350d燃耗期间的kef、加速器束流强度、功率峰因子变化较小,在假想事故下仍保持较深的次临界状态,系统安全性能较好;燃耗较浅,嬗变支持比为20.28,具有较好的嬗变效果。%A physical scheme design of a 10MW accelerator-driven sub-critical gas-cooled fast reactor was studied .The coupling program COUPLE2.0 was applied ,which couples MCNPX and ORIGEN .The physical parameters such as kef ,proton efficiency , accelerator current ,power peak factor and Doppler coefficient of 1500K varying with burnup time were obtained ,and the safety feature of the system was analyzed .By further calculation the 350 d burnup of the system was obtained and the transmutation capability was analyzed . It is found that during the 350 d burnup , kef , accelerator current ,power peak factor variations are relatively small .The system has good safety feature with deep sub-critical status in hypothetical accident . The total burnup is relatively shallow , but the system has good transmutation capability with the transmutation support ratio of 20.28 .

  14. Development of two-step softening heat-treatment in ODS ferritic steel claddings with recrystallized grain structure

    Energy Technology Data Exchange (ETDEWEB)

    Narita, Takeshi; Ukai, Shigeharu; Kaito, Takeji; Ohtsuka, Satoshi [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Kobayashi, Toshimi [Sumitomo Metal Technology, Inc., Tokyo (Japan)

    2003-03-01

    Oxide dispersion-strengthened (ODS) ferritic steels have been developed as prospective cladding materials for attaining higher burnup and higher coolant outlet temperature in advanced fast reactors. In the cladding manufacturing process, two-step softening heat-treatment was developed as an effective method for application to the recrystalizing heat-treatment, which made possible the improvement of formability, ductility and the disappearance of strength anisotropy in the ODS ferritic steel claddings. A mass production process is predicted and cost reduction is concomitantly expected for manufacturing ODS ferritic steel claddings with an adequate strength level at 700degC for commercialized fast reactors. (author)

  15. Spent Nuclear Fuel Transport Reliability Study

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL; Wang, Hong [ORNL; Jiang, Hao [ORNL

    2016-01-01

    This conference paper was orignated and shorten from the following publisehd PTS documents: 1. Jy-An Wang, Hao Jiang, and Hong Wang, Dynamic Deformation Simulation of Spent Nuclear Fuel Assembly and CIRFT Deformation Sensor Stability Investigation, ORNL/SPR-2015/662, November 2015. 2. Jy-An Wang, Hong Wang, Mechanical Fatigue Testing of High-Burnup Fuel for Transportation Applications, NUREG/CR-7198, ORNL/TM-2014/214, May 2015. 3. Jy-An Wang, Hong Wang, Hao Jiang, Yong Yan, Bruce Bevard, Spent Nuclear Fuel Vibration Integrity Study 16332, WM2016 Conference, March 6 10, 2016, Phoenix, Arizona.

  16. Proceedings of the first symposium on Monte Carlo simulation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-01-01

    The first symposium on Monte Carlo simulation was held at Mitsubishi Research Institute, Otemachi, Tokyo, on 10th and 11st of September, 1998. This symposium was organized by Nuclear Code Research Committee at Japan Atomic Energy Research Institute. In the sessions, were presented orally 21 papers on code development, parallel calculation, reactor physics, burn-up, criticality, shielding safety, dose evaluation, nuclear fusion reactor, thermonuclear fusion plasma, nuclear transmutation, electromagnetic cascade, fuel cycle facility. Those presented papers are compiled in this proceedings. The 21 of the presented papers are indexed individually. (J.P.N.)

  17. Calculation Analysis of San Onofre Depletion MOX Fuel Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Pavlovichev, AM

    2001-08-31

    The report provides calculation results of isotopic composition of spent MOX fuel irradiated in Sun Onofre PWR reactor. The calculation was performed by means of the MCU/BURNUP Monte Carlo code. The code is developed in Kurchatov Institute, Russia. The predicted isotope contents are compared with the measured ones. A purpose of this work is a verification both the code and the model of experiment description. Predicted plutonium content exceeds the measured one approximately by 3%. It is arise mainly from error of {sup 239}Pu isotope. Isotopic contents of the main plutonium and uranium isotopes are predicted with satisfactory precision.

  18. Experimental needs for water cooled reactors. Reactor and nuclear fuel; Les besoins experimentaux pour les reacteurs a eau legere. Reacteur et combustible

    Energy Technology Data Exchange (ETDEWEB)

    Waeckel, N. [Electricite de France (EDF/SEPTEN), 69 - Villeurbanne (France); Beguin, S. [Electricite de France (EDF/SEPTEN), 50 - Cherbourg (France); Assedo [AREVA Framatome ANP, 92 - Paris La Defense (France)

    2005-07-01

    In order to improve the competitiveness of nuclear reactors, the trend will be to increase the fuel burn-up, the fuel enrichment, the length of the irradiation cycle and the global thermal power of the reactor. In all cases the fuel rod will be more acted upon. Experimental programs involving research reactors able to irradiate in adequate conditions instrumented fuel rods will stay necessary for the validation of new practices or new nuclear fuel materials in normal or accidental conditions. (A.C.)

  19. DOE plutonium disposition study: Pu consumption in ALWRs. Volume 2, Final report

    Energy Technology Data Exchange (ETDEWEB)

    1993-05-15

    The Department of Energy (DOE) has contracted with Asea Brown Boveri-Combustion Engineering (ABB-CE) to provide information on the capability of ABB-CE`s System 80 + Advanced Light Water Reactor (ALWR) to transform, through reactor burnup, 100 metric tonnes (MT) of weapons grade plutonium (Pu) into a form which is not readily useable in weapons. This information is being developed as part of DOE`s Plutonium Disposition Study, initiated by DOE in response to Congressional action. This document Volume 2, provides a discussion of: Plutonium Fuel Cycle; Technology Needs; Regulatory Considerations; Cost and Schedule Estimates; and Deployment Strategy.

  20. Development of data base with mechanical properties of un- and pre-irradiated VVER cladding

    Energy Technology Data Exchange (ETDEWEB)

    Asmolov, V.; Yegorova, L.; Kaplar, E.; Lioutov, K. [Nuclear Safety Inst. of Russian Research Centre, Moscow (Russian Federation). Kurchatov Inst.; Smirnov, V.; Prokhorov, V.; Goryachev, A. [State Research Centre, Dimitrovgrad (Russian Federation). Research Inst. of Atomic Reactors

    1998-03-01

    Analysis of recent RIA test with PWR and VVER high burnup fuel, performed at CABRI, NSRR, IGR reactors has shown that the data base with mechanical properties of the preirradiated cladding is necessary to interpret the obtained results. During 1997 the corresponding cycle of investigations for VVER clad material was performed by specialists of NSI RRC KI and RIAR in cooperation with NRC (USA), IPSN (France) in two directions: measurements of mechanical properties of Zr-1%Nb preirradiated cladding versus temperature and strain rate; measurements of failure parameters for gas pressurized cladding tubes. Preliminary results of these investigations are presented in this paper.

  1. On collapsing the Pu94242 average number of neutrons released per fission from the IAEA.LIB library with the WIMSD-5b code

    Energy Technology Data Exchange (ETDEWEB)

    Caldeira, Alexandre D. [Centro Tecnico Aeroespacial (CTA), Instituto de Estudos Avancados (IEAv), 12231-970 Sao Jose dos Campos, SP (Brazil)]. E-mail: alexdc@ieav.cta.br; Claro, Luiz H. [Centro Tecnico Aeroespacial (CTA), Instituto de Estudos Avancados (IEAv), 12231-970 Sao Jose dos Campos, SP (Brazil)

    2007-01-15

    It was verified after a fuel burnup calculation with the WIMSD-5b code using the IAEA.LIB library that the computed average number of neutrons released per fission of Pu94242 shows up as a Not-a-Number (NaN) for some energy groups. As this problem does not permit the use of the generated multigroup microscopic cross sections by a reactor calculation code, the value of 1.0E-38 barns was attributed to all energy groups of the IAEA.LIB library that have null values of multigroup microscopic fission cross sections for this material.

  2. Investigation of fuel behaviour at different power levels (SKI-bump II). Final report

    Energy Technology Data Exchange (ETDEWEB)

    Lysell, G.; Bengtsson, Sven

    1994-12-31

    Fuel microstructure and fission product redistribution in the fuel and to the cladding inside were studied after a `bump` irradiation of a high burnup rod up to maximum Linear Heat Rating (LHR) of 43 kW/m. Fission gas induced swelling of the fuel imposed a large strain on the cladding. Total gas release in the rod as well as the local release at two different LHRs were measured. Reference fuel not subject to the bump was examined with the same techniques as the bumped fuel.

  3. ANL pre analysis of the SHEBA/CERES experiments.

    Energy Technology Data Exchange (ETDEWEB)

    Palmiotti, G.; Smith, M.; Klann, R.; Fujita, E.; Imel, G.

    2000-05-05

    The French and British nuclear programs have prepared a series of natural uranium oxide fuel samples spiked with small amounts of the individual fission products which makeup a large fraction of the total neutron absorption by fission products in spent nuclear fuel. Both programs have utilized these samples in experimental reactors and have inferred the worth of the individual fission products. These results have been used to validate the cross sections used in criticality safety calculations. These measurements constitute a major element in support of spent fuel burnup credit in these countries.

  4. Theory of Antineutrino Monitoring of Burning MOX Plutonium Fuels

    CERN Document Server

    Hayes, A C; Nieto, Michael Martin; WIlson, W B

    2011-01-01

    This letter presents the physics and feasibility of reactor antineutrino monitoring to verify the burnup of plutonium loaded in the reactor as a Mixed Oxide (MOX) fuel. It examines the magnitude and temporal variation in the antineutrino signals expected for different MOX fuels, for the purposes of nuclear accountability and safeguards. The antineutrino signals from reactor-grade and weapons-grade MOX are shown to be distinct from those from burning low enriched uranium. Thus, antineutrino monitoring could be used to verify the destruction of plutonium in reactors, though verifying the grade of the plutonium being burned is found to be more challenging.

  5. Very High Temperature Reactor (VHTR) Deep Burn Core and Fuel Analysis -- Complete Design Selection for the Pebble Bed Reactor

    Energy Technology Data Exchange (ETDEWEB)

    B. Boer; A. M. Ougouag

    2010-09-01

    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 multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup, while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239-Pu, 240-Pu and 241-Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a ”standard,” UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge

  6. Boron coating on boron nitride coated nuclear fuels by chemical vapor deposition

    Science.gov (United States)

    Durmazuçar, Hasan H.; Gündüz, Güngör

    2000-12-01

    Uranium dioxide-only and uranium dioxide-gadolinium oxide (5% and 10%) ceramic nuclear fuel pellets which were already coated with boron nitride were coated with thin boron layer by chemical vapor deposition to increase the burn-up efficiency of the fuel during reactor operation. Coating was accomplished from the reaction of boron trichloride with hydrogen at 1250 K in a tube furnace, and then sintering at 1400 and 1525 K. The deposited boron was identified by infrared spectrum. The morphology of the coating was studied by using scanning electron microscope. The plate, grainy and string (fiber)-like boron structures were observed.

  7. ARP: Automatic rapid processing for the generation of problem dependent SAS2H/ORIGEN-s cross section libraries

    Energy Technology Data Exchange (ETDEWEB)

    Leal, L.C.; Hermann, O.W.; Bowman, S.M.; Parks, C.V.

    1998-04-01

    In this report, a methodology is described which serves as an alternative to the SAS2H path of the SCALE system to generate cross sections for point-depletion calculations with the ORIGEN-S code. ARP, Automatic Rapid Processing, is an algorithm that allows the generation of cross-section libraries suitable to the ORIGEN-S code by interpolation over pregenerated SAS2H libraries. The interpolations are carried out on the following variables: burnup, enrichment, and water density. The adequacy of the methodology is evaluated by comparing measured and computed spent fuel isotopic compositions for PWR and BWR systems.

  8. Development and applications of methodologies for the neutronic design of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR)

    Science.gov (United States)

    Fratoni, Massimiliano

    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 conventional helium coolant, is not transparent to neutrons. The flibe occupies 40% of the PB-AHTR core volume and absorbs ˜8% of the neutrons, but also acts as an effective neutron moderator. Two novel methodologies were developed for calculating the time dependent and equilibrium core composition: (1) a simplified single pebble model that is relatively fast; (2) a full 3D core model that is accurate and flexible but computationally intensive. A parametric analysis was performed spanning a wide range of fuel kernel diameters and graphite-to-heavy metal atom ratios to determine the attainable burnup and reactivity coefficients. Using 10% enriched uranium ˜130 GWd/tHM burnup was found to be attainable, when the graphite-to-heavy metal atom ratio (C/HM) is in the range of 300 to 400. At this or smaller C/HM ratio all reactivity coefficients examined---coolant temperature, coolant small and full void, fuel temperature, and moderator temperature, were found to be negative. The PB-AHTR performance was compared to that of alternative options for HTRs, including the helium-cooled pebble-bed reactor and prismatic fuel reactors, both gas-cooled and flibe-cooled. The attainable burnup of all designs was found to be similar. The PB-AHTR generates at least 30% more energy per pebble than the He-cooled pebble-bed reactor. Compared to LWRs the PB-AHTR requires 30% less natural uranium and 20% less separative work per unit of electricity generated. For deep burn TRU fuel made from recycled LWR spent fuel, it was found that in a single pass through the core ˜66% of the TRU can be

  9. Parametric study on a natural circulation cooled U-battery

    Energy Technology Data Exchange (ETDEWEB)

    De Zwaan, Sieuwert; Kloosterman, Jan Leen [Delft University of Technology, Delft (Netherlands); Van Uitert, Gert [The Hague (Netherlands)

    2008-07-01

    A feasibility study has been performed on a natural circulation cooled small nuclear reactor with a molten salt or tin as a coolant. This reactor is called the U-Battery. The study included neutronics calculations to obtain the minimum dimensions required for a critical system during burnup, the calculation of coolant temperature and core temperature reactivity coefficients, and an investigation of the thermal hydraulics to asses the possibilities for natural circulation cooling. For every coolant, core designs are feasible within the dimensions imposed and with natural circulation of the coolant. (authors)

  10. Used Fuel Testing Transportation Model

    Energy Technology Data Exchange (ETDEWEB)

    Ross, Steven B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Best, Ralph E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Maheras, Steven J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jensen, Philip J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); England, Jeffery L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); LeDuc, Dan [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2014-09-25

    This report identifies shipping packages/casks that might be used by the Used Nuclear Fuel Disposition Campaign Program (UFDC) to ship fuel rods and pieces of fuel rods taken from high-burnup used nuclear fuel (UNF) assemblies to and between research facilities for purposes of evaluation and testing. Also identified are the actions that would need to be taken, if any, to obtain U.S. Nuclear Regulatory (NRC) or other regulatory authority approval to use each of the packages and/or shipping casks for this purpose.

  11. Focused technology: Nuclear propulsion

    Science.gov (United States)

    Miller, Thomas J.

    1991-01-01

    The topics presented are covered in viewgraph form and include: nuclear thermal propulsion (NTP), which challenges (1) high temperature fuel and materials, (2) hot hydrogen environment, (3) test facilities, (4) safety, (5) environmental impact compliance, and (6) concept development, and nuclear electric propulsion (NEP), which challenges (1) long operational lifetime, (2) high temperature reactors, turbines, and radiators, (3) high fuel burn-up reactor fuels, and designs, (4) efficient, high temperature power conditioning, (5) high efficiency, and long life thrusters, (6) safety, (7) environmental impact compliance, and (8) concept development.

  12. Behavior of irradiated BWR fuel under reactivity-initiated-accident conditions; Results of tests FK-1, -2 and -3

    OpenAIRE

    2004-01-01

    Boiling water reactor (BWR) fuels with burnups of 41 to 45 GWd/tU were pulse-irradiated in the Nuclear Safety Research Reactor (NSRR) to investigate fuel behavior under cold startup reactivity-initiated-accident (RIA) conditions. BWR fuel segment rods of 8times8BJ (STEP I) type from Fukushima-Daiichi Unit 3 nuclear power plant were refabricated into short test rods, and they were subjected to prompt enthalpy insertion from 293 to 607 J/g (70 to 145 cal/g) within about 20 ms. The fuel cladding...

  13. Depletion analysis and sensitivity study of PHENIX fuel-irradiation experiments

    Energy Technology Data Exchange (ETDEWEB)

    Biswas, D.; Kallfelz, J.M.; White, J.R.

    1981-01-01

    The experimental results are in the form of various U and Pu atom density ratios (R/sub E/) and burnup (BU) values. The results were for samples irradiated during the first three cycles in the central zone of PHENIX. The time-dependent sensitivity study was performed with the depletion generalized perturbation code DEPTH-CHARGE, to investigate the sensitivity of R/sub E/ to cross sections and to absolute flux level changes. The depletion analysis was performed using ENDS/B-IV data, an (R-Z) model, and the VENTURE Code system. 2 tables.

  14. Nuclear Reactor Simulations for Unveiling Diversion Scenarios: capabilities of the antineutrino probe

    Energy Technology Data Exchange (ETDEWEB)

    Bui, V.M.; Fallot, M.; Giot, L.; Guillon, B.; Martino, J.; Yermia, F. [SUBATECH - CNRS-IN2P3 - Univ. of Nantes - EMN, Nantes (France); Nuttin, A. [LPSC - CNRS-IN2P3/UJF/INPG, Grenoble (France)

    2009-06-15

    After many years of fundamental research, physicists have a good understanding of the neutrinos detection techniques. It is now possible to apply neutrino physics as a new tool to monitor nuclear power plants. We already know that modest size detectors are achievable to fulfill that task such as the SONGS 1 and the future Nucifer detectors. In parallel, sophisticated simulations of reactors and their associated antineutrino flux and energy spectrum have been developed to predict the neutrino signature of the fuel burnup and of a diversion. Taking advantage of the tremendous quantity of information available nowadays in nuclear databases, the total {beta} spectrum of a reactor is built by adding the contributions of all the {beta} branches involved in the decay of all fission products (FP). A package called MCNP Utility for Reactor Evolution (MURE) computes the fuel and FP inventories by simulating the neutronics and time evolution of a reactor core. MURE, initially developed by CNRS/IN2P3/LPSC Grenoble and IPN Orsay to study Generation IV reactors, is a precision code written in C++ which automates the preparation and computation of successive MCNP calculations either for precision burnup or thermal-hydraulics purpose. MURE will be soon available at NEA. The only user-defined inputs driving the time evolution of the isotopic composition of the core are the initial fuel composition, the refueling scheme, and the thermal power. The evolution of the antineutrino flux and energy spectrum with the fuel burnup, as well as the effect of neutron capture on various nuclei are taken into account. Nonproliferation scenarios and burnup monitoring with antineutrinos have been studied using these tools for PWR and Candu reactors. A full core simulation of an N4-PWR will be presented in a first part. Gross unveiling diversion scenarios using a PWR have been simulated in order to test the ability of the antineutrino probe. A channel of a Heavy Water Reactor (Candu 600) loaded with

  15. Computer simulation of tritium systems for fusion technology

    Energy Technology Data Exchange (ETDEWEB)

    Gabowitsch, E.; Spannagel, G. (Kernforschungszentrum Karlsruhe GmbH Institut fur Datenverarbeitung in der Technik Postfach 3640, D-7500 Karlsruhe 1 (DE))

    1989-09-01

    The KATRIM computer code is presented. It calculates key values of tritium systems, especially those related to complete fuel cycles. First, a deterministic model is discussed. Then, a stochastic model is presented based on dynamic systems with different dynamic states, each with its own system of equations. Such an approach allows the modeling of reactors with different degrees of availability and/or different operational strategies. Results of simulations for different availabilities, variable frequencies of interruptions in reactor operation, and changing tritium burnup in the plasma are presented.

  16. Analysis of the impact of coolant density variations in the high efficiency channel of a pressure tube super critical water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Scriven, M.G.; Hummel, D.W.; Novog, D.R.; Luxat, J.C. [McMaster Univ., Hamilton, Ontario (Canada)

    2012-07-01

    The Pressure Tube (PT) Supercritical Water Reactor (SCWR) is based on a light water coolant operating at pressures above the thermodynamic critical pressure; a separate low temperature and low pressure moderator. The coolant density changes by an order of magnitude depending on its local enthalpy in the porous ceramic insulator tube. This causes significant changes in the neutron transport characteristics, axially and radially, in the fuel channel. This work performs lattice physics calculations for a 78-element Pu-Th fuel at zero burnup and examines the effect of assumptions related to coolant density in the radial direction of a HEC, using the neutron transport code WIMS-AECL. (author)

  17. Nuclear design report for system-integrated modular advanced reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Yoon; Lee, Chung Chan; Zee, Sung Quun; Chang, Moon Hee [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-02-01

    This report presents nuclear characteristics analysis results for SMART. Information is given on fuel loading, power density distributions, reactivity coefficients and control rod worths. The core consists of 57 modified Korean Standard Fuel Assemblies (m-KOFAs). and all fuel assemblies contain burnable absorbers to control the power distribution and the excess reactivity that is required for soluble boron-free and ultra longer cycle operation. The cycle length of SMART amounts to 990 EFPD corresponding to a cycle burnup of 26,160 MWD/MTU. 4 refs., 92 figs., 5 tabs. (Author)

  18. Development of the fabrication technology of the simulated fuel-I, 15,000MWd/tU

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kweon Ho; Kim, D. J.; Kim, H. S.; Lee, J. W.; Yang, M. S

    2001-04-01

    It is important to get basic data to analysis physical properties, behavior in reactor and performance of the DUPIC fuel because physical properties, fission gas release, grain growth and et al. of the DUPIC fuel is different from the commercial UO2 fuel. But what directly measures physical properties et al. of DUPIC fuel being resinterred simulated spent fuel through OREOX process is very difficult in laboratory owing to its high level radiation. Then fabrication of simulated DUPIC fuel is needed to measure its properties. In this study, the sintering characterization of wet milled powder for 24 hours to fabricate 15,000MWd/tU equivalent burnup simulated fuel.

  19. Determination of the rod-wise fission gas release fraction in a complete fuel assembly using non-destructive gamma emission tomography

    Science.gov (United States)

    Holcombe, Scott; Andersson, Peter; Svärd, Staffan Jacobsson; Hallstadius, Lars

    2016-11-01

    A gamma tomography instrument has been developed at the Halden Boiling Water Reactor (HBWR) in cooperation between the Institute for Energy Technology, Westinghouse (Sweden) and Uppsala University. The instrument is used to record the gamma radiation field surrounding complete fuel assemblies and consists of a shielded enclosure with fixtures to accurately position the fuel and detector relative to each other. A High Purity Germanium detector is used for acquiring high-resolution spectroscopic data, allowing for analysis of multiple gamma-ray peaks. Using the data extracted from the selected peaks, tomographic reconstruction algorithms are used to reproduce the corresponding spatial gamma-ray source distributions within the fuel assembly. With this method, rod-wise data can be can be deduced without the need to dismantle the fuel. In this work, the tomographic device has been experimentally benchmarked for non-destructive rod-wise determination of the Fission Gas Release (FGR) fraction. Measurements were performed on the fuel-stack and gas-plenum regions of a complete fuel assembly, and quantitative tomographic reconstructions of the measurement data were performed in order to determine the rod-wise ratio of 85Kr in the gas plenum to 137Cs in the fuel stack. The rod-wise ratio of 85Kr/137Cs was, in turn, used to calculate the rod-wise FGR fraction. In connection to the tomographic measurements, the fuel rods were also measured individually using gamma scanning in order to provide an experimental benchmark for the tomographic method. Fuel rods from two donor driver fuel assemblies were placed into a nine-rod HBWR driver fuel assembly configuration. In order to provide a challenging measurement object and thus an appropriate benchmark for the tomographic method, five rods were taken from an assembly with a burnup of 51 MWd/kgUO2, and four rods were from an assembly with a burnup of 26 MWd/kgUO2. At the time of the measurements, the nine rods had cooled for

  20. Dynamic leaching studies of 48 MWd/kgU UO2 commercial spent nuclear fuel under oxic conditions

    OpenAIRE

    Serrano Purroy, D.; Casas Pons, Ignasi; Gonzalez Robles, E.; Glatz, Jean Paul; Wegen, D.H.; Clarens Blanco, Frederic; Giménez Izquierdo, Francisco Javier; Pablo Ribas, Joan de; Martínez Esparza, A.

    2013-01-01

    The leaching of a high-burn-up spent nuclear fuel (48 MWd/KgU) has been studied in a carbonate-containing solution and under oxic conditions using a Continuously Stirred Tank Flow-Through Reactor (CSTR). Two samples of the fuel, one prepared from the centre of the pellet (labelled CORE) and another one from the fuel pellet periphery, enriched with the so-called High Burn-Up Structure (HBS, labelled OUT) have been used. For uranium and actinides, the results showed that U, Np, Am and Cm ga...