Two dimensional burn-up calculation of TRIGA core

International Nuclear Information System (INIS)

Persic, A.; Ravnik, M.; Slavic, S.

1996-01-01

TRIGLAV is a new computer program for burn-up calculation of mixed core of research reactors. The code is based on diffusion model in two dimensions and iterative procedure is applied for its solution. The material data used in the model are calculated with the transport program WIMS. In regard to fission density distribution and energy produced by the reactor the burn-up increment of fuel elements is determined. In this paper the calculation model of diffusion constants and burn-up calculation are described and some results of calculations for TRIGA MARK II reactor are presented. (author)

Technical development on burn-up credit for spent LWR fuels

International Nuclear Information System (INIS)

Nakahara, Yoshinori; Suyama, Kenya; Suzaki, Takenori

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)

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)

Reactivity effect of spent fuel depending on burn-up history

International Nuclear Information System (INIS)

Hayashi, Takafumi; Suyama, Kenya; Nomura, Yasushi

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)

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.

Burn-up Credit Criticality Safety Benchmark Phase III-C. Nuclide Composition and Neutron Multiplication Factor of a Boiling Water Reactor Spent Fuel Assembly for Burn-up Credit and Criticality Control of Damaged Nuclear Fuel

International Nuclear Information System (INIS)

Suyama, K.; Uchida, Y.; Kashima, T.; Ito, T.; Miyaji, T.

2016-01-01

longer process time (CPU) is required. Treatment of the gadolinium rod is still a key issue. The difference of the neutron multiplication factor generated by the burn-up calculation results was confirmed by the analysis using the same criticality calculation code, MVP. It was less than 3% when the latest code system was used, including continuous-energy Monte Carlo codes and deterministic codes. This is the first time this kind of value has been shown by an extensive international benchmark problem. These results show that even if calculation codes are benchmarked using the well-qualified experimental data before being adopted in the safety review process, it should be understood that some uncertainty in the evaluation of the neutron multiplication factor arising from the uncertainty of the burn-up calculation methodology used still remains

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.

A Study for Burn-up Calculation applied on 400MWth PBMR Core

International Nuclear Information System (INIS)

Luu, Nam Hai; Kim, Hong Chul; Kim, Soon Young; Kim, Jong Kyung; Noh, Jae Man

2007-01-01

The 400MWth Pebble-bed Modular Reactor (PBMR) is an advanced high temperature gas cooled-reactor (HTGR). It possesses a very high efficiency and attractive economics without compromising the high levels of passive safety expected of advanced nuclear designs. With this reason, PBMR is a target which researchers especially in nuclear engineering field study carefully and therefore it is regarded as the leader in the power generation field. There are many research results about benchmark problems but results of the burn-up process are still poor. Hence, in this study a burn-up calculation was performed with PBMR using MONTEBURNS code in which MCNP modeling linked a depletion systems is used

Effect of burn-up on the radioactivation behavior of cladding hull materials studied using the ORIGEN-S code

International Nuclear Information System (INIS)

Min Ku Jeon; Chang Hwa Lee; Jung Hoon Choi; In Hak Cho; Kweon Ho Kang; Hwan-Seo Park; Geun Il Park; Chang Je Park

2013-01-01

The effect of fuel burn-up on the radioactivation behavior of cladding hull materials was investigated using the ORIGEN-S code for various materials of Zircaloy-4, Zirlo, HANA-4, and HANA-6 and for various fuel burn-ups of 30, 45, 60, and 75 GWD/MTU. The Zircaloy-4 material is the only one that does not contain Nb as an alloy constituent, and it was revealed that 125 Sb, 125m Te, and 55 Fe are the major sources of radioactivity. On the other hand, 93m Nb was identified as the most radioactive nuclide for the other materials although minor radioactive nuclides varied owing to their different initial constituents. The radioactivity of 94 Nb was of particular focus owing to its acceptance limit against a Korean intermediate-/low-level waste repository. The radioactivation calculation results revealed that only Zircaloy-4 is acceptable for the Korean repository, while the other materials required at least 4,900 of Nb decontamination factor owing to the high radioactivity of 94 Nb regardless of the fuel burn-up. A discussion was also made on the feasibility of Zr recovery methods (chlorination and electrorefining) for selective recovery of Zr so that it can be disposed of in the Korean repository. (author)

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.

Nondestructive, fast methods for burn-up study

International Nuclear Information System (INIS)

Schaechter, L.; Hacman, D.; Mot, O.

1977-01-01

Nondestructive methods, based on high resolution-spectrometry successfully applied at Institute for Atomic Physics are presented. These methods are preferred to destructive chemical methods; the latter being costly and lengthy and not suitable for statistical prediction of nuclear fuel behaviour. The following methods are developed: methods for determining the burn up of fuel elements and fuel assemblies; a method for determining the U 235 and Pu 239 contributions to the burn up and a code written in FORTRAN IV for numerical calculation of Pu 239 fission vs. burn up; a high precision method for burnup determination by adding burnable poison; a method for prediction of specific power distribution in the fuel elements of a research or power reactors; a method for determining the power output of the fuel element in an operating power reactor; a method for determining the content of Pu 239 of the fuel element irradiated in a reactor. The results which were obtained by these methods improved the fuel management at the VVR-S reactor at Institute for Atomic Physics, Bucharest and may be applied to other reactor types [fr

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

International Nuclear Information System (INIS)

Garcia-Herranz, Nuria; Cabellos, Oscar; Sanz, Javier; Juan, Jesus; Kuijper, Jim C.

2008-01-01

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

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.

Burnup code for fuel assembly by Monte Carlo code. MKENO-BURN

International Nuclear Information System (INIS)

Naito, Yoshitaka; Suyama, Kenya; Masukawa, Fumihiro; Matsumoto, Kiyoshi; Kurosawa, Masayoshi; Kaneko, Toshiyuki.

1996-12-01

The evaluation of neutron spectrum is so important for burnup calculation of the heterogeneous geometry like recent BWR fuel assembly. MKENO-BURN is a multi dimensional burnup code that based on the three dimensional monte carlo neutron transport code 'MULTI-KENO' and the routine for the burnup calculation of the one dimensional burnup code 'UNITBURN'. MKENO-BURN analyzes the burnup problem of arbitrary regions after evaluating the neutron spectrum and making one group cross section in three dimensional geometry with MULTI-KENO. It enables us to do three dimensional burnup calculation. This report consists of general description of MKENO-BURN and the input data. (author)

Burnup code for fuel assembly by Monte Carlo code. MKENO-BURN

Energy Technology Data Exchange (ETDEWEB)

Naito, Yoshitaka; Suyama, Kenya; Masukawa, Fumihiro; Matsumoto, Kiyoshi; Kurosawa, Masayoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Toshiyuki

1996-12-01

The evaluation of neutron spectrum is so important for burnup calculation of the heterogeneous geometry like recent BWR fuel assembly. MKENO-BURN is a multi dimensional burnup code that based on the three dimensional monte carlo neutron transport code `MULTI-KENO` and the routine for the burnup calculation of the one dimensional burnup code `UNITBURN`. MKENO-BURN analyzes the burnup problem of arbitrary regions after evaluating the neutron spectrum and making one group cross section in three dimensional geometry with MULTI-KENO. It enables us to do three dimensional burnup calculation. This report consists of general description of MKENO-BURN and the input data. (author)

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)

Calculations of fuel burn-up and radionuclide inventory in the syrian miniature neutron source reactor using the WIMSD4 code

International Nuclear Information System (INIS)

Khattab, K.

2005-01-01

Calculations of the fuel burn up and radionuclide inventory in the Miniature Neutron Source Reactor after 10 years (the reactor core expected life) of the reactor operating time are presented in this paper. The WIMSD4 code is used to generate the fuel group constants and the infinite multiplication factor versus the reactor operating time for 10, 20, and 30 kW operating power levels. The amounts of uranium burnt up and plutonium produced in the reactor core, the concentrations and radioactivities of the most important fission product and actinide radionuclides accumulated in the reactor core, and the total radioactivity of the reactor core are calculated using the WIMSD4 code as well

Analysis of some antecipated transients without scram for PWR type reactors by coupling of the CORAN code to the ALMOD code system

International Nuclear Information System (INIS)

Carvalho, F. de A.T. de.

1985-01-01

This study investigates some antecipated transients without scram for a pressurized water cooled reactor, using coupling of the containment CORAN code to the ALMOD code system, under severe random conditions. This coupling has the objective of including containment model as part of an unified code system. These severe conditions include failure of reactor scram, following a station black-out and emergency power initiation for the burn-up status at the beginning and end of the cycle. Furthermore, for the burn-up status at the end of the cycle, a failure in the closure of the pressurizer relief valve was also investigated. (Author) [pt

Challenges in the application of burn-up credit to the criticality safety of the THORP reprocessing plant

International Nuclear Information System (INIS)

Mayson, R.T.H.; Gunston, K.J.

1999-01-01

Since 1991 BNFL has made a significant investment in the development of the burn-up credit method and the application to its operations. It has recently demonstrated that using this method for the THORP dissolvers, it is possible to justify operating safety with reduced neutron poison concentrations and this has now been submitted to the regulators. The continued challenges the criticality safety community is facing are to show that we are not reducing safety levels because we are using burn-up credit. The burn-up credit method that has been developed can be summarized as follows. It consists of performing reactivity calculations for irradiated fuel using compositions generated by and inventory prediction code, generally in order to determine the limiting burn-up required for that fuel in a particular environment. In addition, it has always been envisaged that a confirmatory measurement of burn-up would be required to be made prior to certain operations such as the sharing of fuel into a dissolver. The burn-up credit method therefore relies upon three key components of inventory prediction, reactivity calculation code and the quantification and verification of burn-up. (J.P.N.)

Actinide-only and full burn-up credit in criticality assessment of RBMK-1500 spent nuclear fuel storage cask using axial burn-up profile

Energy Technology Data Exchange (ETDEWEB)

Barkauskas, V., E-mail: vytenis.barkauskas@ftmc.lt; Plukiene, R., E-mail: rita.plukiene@ftmc.lt; Plukis, A., E-mail: arturas.plukis@ftmc.lt

2016-10-15

Highlights: • RBMK-1500 fuel burn-up impact on k{sub eff} in the SNF cask was calculated using SCALE 6.1. • Positive end effect was noticed at certain burn-up for the RBMK-1500 spent nuclear fuel. • The non-uniform uranium depletion is responsible for the end effect in RBMK-1500 SNF. • k{sub eff} in the SNF cask does not exceed a value of 0.95 which is set in the safety requirements. - Abstract: Safe long-term storage of spent nuclear fuel (SNF) is one of the main issues in the field of nuclear safety. Burn-up credit application in criticality analysis of SNF reduces conservatism of usually used fresh fuel assumption and implies a positive economic impact for the SNF storage. Criticality calculations of spent nuclear fuel in the CONSTOR® RBMK-1500/M2 cask were performed using pre-generated ORIGEN-ARP spent nuclear fuel composition libraries, and the results of the RBMK-1500 burn-up credit impact on the effective neutron multiplication factor (k{sub eff}) have been obtained and are presented in the paper. SCALE 6.1 code package with the STARBUCKS burn-up credit evaluation tool was used for modeling. Pre-generated ARP (Automatic Rapid Processing) crosssection libraries based on ENDF/B-VII cross section library were used for fast burn-up inventory modeling. Different conditions in the SNF cask were modeled: 2.0% and 2.8% initial enrichment fuel of various burn-up and water density inside cavities of the SNF cask. The fuel composition for the criticality analysis was chosen taking into account main actinides and most important fission products used in burn-up calculations. A significant positive end effect is noticed from 15 GWd/tU burn-up for 2.8% enrichment fuel and from 9 GWd/tU for 2.0% enrichment fuel applying the actinide-only approach. The obtained results may be applied in further evaluations of the RBMK type reactor SNF storage as well as help to optimize the SNF storage volume inside the CONSTOR® RBMK-1500/M2 cask without compromising criticality

Actinide-only and full burn-up credit in criticality assessment of RBMK-1500 spent nuclear fuel storage cask using axial burn-up profile

International Nuclear Information System (INIS)

Barkauskas, V.; Plukiene, R.; Plukis, A.

2016-01-01

Highlights: • RBMK-1500 fuel burn-up impact on k_e_f_f in the SNF cask was calculated using SCALE 6.1. • Positive end effect was noticed at certain burn-up for the RBMK-1500 spent nuclear fuel. • The non-uniform uranium depletion is responsible for the end effect in RBMK-1500 SNF. • k_e_f_f in the SNF cask does not exceed a value of 0.95 which is set in the safety requirements. - Abstract: Safe long-term storage of spent nuclear fuel (SNF) is one of the main issues in the field of nuclear safety. Burn-up credit application in criticality analysis of SNF reduces conservatism of usually used fresh fuel assumption and implies a positive economic impact for the SNF storage. Criticality calculations of spent nuclear fuel in the CONSTOR® RBMK-1500/M2 cask were performed using pre-generated ORIGEN-ARP spent nuclear fuel composition libraries, and the results of the RBMK-1500 burn-up credit impact on the effective neutron multiplication factor (k_e_f_f) have been obtained and are presented in the paper. SCALE 6.1 code package with the STARBUCKS burn-up credit evaluation tool was used for modeling. Pre-generated ARP (Automatic Rapid Processing) crosssection libraries based on ENDF/B-VII cross section library were used for fast burn-up inventory modeling. Different conditions in the SNF cask were modeled: 2.0% and 2.8% initial enrichment fuel of various burn-up and water density inside cavities of the SNF cask. The fuel composition for the criticality analysis was chosen taking into account main actinides and most important fission products used in burn-up calculations. A significant positive end effect is noticed from 15 GWd/tU burn-up for 2.8% enrichment fuel and from 9 GWd/tU for 2.0% enrichment fuel applying the actinide-only approach. The obtained results may be applied in further evaluations of the RBMK type reactor SNF storage as well as help to optimize the SNF storage volume inside the CONSTOR® RBMK-1500/M2 cask without compromising criticality safety.

Calculational prediction of fuel burn-up for the Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Nguyen Phuoc Lan; Do Quang Binh

2016-01-01

In this paper, the method of expanding operators and functions in the neutron diffusion equations as chains of time variable is used for calculation of fuel burn-up of the Dalat nuclear reactors. A computer code, named BURREF, programmed in language Fortran-77 running on IBM PC-AT, has been developed based on this method to predict the fuel burn-up of the Dalat reactor. Some results will be presented here. (author)

Burn-up TRIGA Mark II benchmark experiment

International Nuclear Information System (INIS)

Persic, A.; Ravnik, M.; Zagar, T.

1998-01-01

Different reactor codes are used for calculations of reactor parameters. The accuracy of the programs is tested through comparison of the calculated values with the experimental results. Well-defined and accurately measured benchmarks are required. The experimental results of reactivity measurements, fuel element reactivity worth distribution and fuel-up measurements are presented in this paper. The experiments were performed with partly burnt reactor core. The experimental conditions were well defined, so that the results can be used as a burn-up benchmark test case for a TRIGA Mark II reactor calculations.(author)

SRAC95; general purpose neutronics code system

International Nuclear Information System (INIS)

Okumura, Keisuke; Tsuchihashi, Keichiro; Kaneko, Kunio.

1996-03-01

SRAC is a general purpose neutronics code system applicable to core analyses of various types of reactors. Since the publication of JAERI-1302 for the revised SRAC in 1986, a number of additions and modifications have been made for nuclear data libraries and programs. Thus, the new version SRAC95 has been completed. The system consists of six kinds of nuclear data libraries(ENDF/B-IV, -V, -VI, JENDL-2, -3.1, -3.2), five modular codes integrated into SRAC95; collision probability calculation module (PIJ) for 16 types of lattice geometries, Sn transport calculation modules(ANISN, TWOTRAN), diffusion calculation modules(TUD, CITATION) and two optional codes for fuel assembly and core burn-up calculations(newly developed ASMBURN, revised COREBN). In this version, many new functions and data are implemented to support nuclear design studies of advanced reactors, especially for burn-up calculations. SRAC95 is available not only on conventional IBM-compatible computers but also on scalar or vector computers with the UNIX operating system. This report is the SRAC95 users manual which contains general description, contents of revisions, input data requirements, detail information on usage, sample input data and list of available libraries. (author)

SRAC95; general purpose neutronics code system

Energy Technology Data Exchange (ETDEWEB)

Okumura, Keisuke; Tsuchihashi, Keichiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Kunio

1996-03-01

SRAC is a general purpose neutronics code system applicable to core analyses of various types of reactors. Since the publication of JAERI-1302 for the revised SRAC in 1986, a number of additions and modifications have been made for nuclear data libraries and programs. Thus, the new version SRAC95 has been completed. The system consists of six kinds of nuclear data libraries(ENDF/B-IV, -V, -VI, JENDL-2, -3.1, -3.2), five modular codes integrated into SRAC95; collision probability calculation module (PIJ) for 16 types of lattice geometries, Sn transport calculation modules(ANISN, TWOTRAN), diffusion calculation modules(TUD, CITATION) and two optional codes for fuel assembly and core burn-up calculations(newly developed ASMBURN, revised COREBN). In this version, many new functions and data are implemented to support nuclear design studies of advanced reactors, especially for burn-up calculations. SRAC95 is available not only on conventional IBM-compatible computers but also on scalar or vector computers with the UNIX operating system. This report is the SRAC95 users manual which contains general description, contents of revisions, input data requirements, detail information on usage, sample input data and list of available libraries. (author).

Burn-up measurement in the HTR-module-reactor

International Nuclear Information System (INIS)

Gerhards, E.

1993-05-01

The burn-up status of spherical HTR-fuel elements is determined by a γ-spectrometric analysis of Cs-137 activity. The γ-spectrum recorded by a semiconductor detector up to now is analyzed by complex mathematical and time-consuming methods. For the operation of the HTR-Module-Reactor, however, a fast evaluation of the burn-up status is necessary. It is shown that this can be ensured by a comparison between the measured spectra and simulation results. Using the computer-program HTROGEN and the program system SPECCALC especially developed for this problem the γ-spectra are evaluated as a function of the burn-up status. The method is applied to results available from the operation of the AVR-reactor. The burn-up status determined with different methods corresponds very well within the limits of accuracy. (orig.)

Modeling of WWER-440 Fuel Pin Behavior at Extended Burn-up

International Nuclear Information System (INIS)

El-Koliel, M.S.; Abou-Zaid, A.A.; El-Kafas, A.A.

2004-01-01

Currently, there is an ongoing effort to increase fuel discharge burn-up of all LWRs fuel including WWER's as much as possible in order to decrease power production cost. Therefore, burn-up is expected to be increased to 60 to 70 Mwd/kg U. The change in the fuel radial power distribution as a function of fuel burn up can affect the radial fuel temperature distribution as well as the fuel microstructure in the fuel pellet rim. In this paper, the radial burn-up and fissile products distributions of WWER-440 UO 2 fuel pin were evaluated using MCNP 4B and ORIGEN2 codes. The impact of the thermal conductivity on predicted fission gas release calculations is needed. For the analysis, a typical WWER-440 fuel pin and surrounding water moderator are considered in a hexagonal pin cell well. The thermal release and the athermal release from the pellet rim were modeled separately. The fraction of the rim structure and the excessive porosity in the rim structure in isothermal irradiation as a function of the fuel burn-up was predicted. a computer program; RIMSC-01, is developed to perform the required FGR calculations. Finally, the relevant phenomena and the corresponding models together with their validation are presented

Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

Directory of Open Access Journals (Sweden)

Muhammad Atta

2011-01-01

Full Text Available The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease.

Estimating NIRR-1 burn-up and core life time expectancy using the codes WIMS and CITATION

Science.gov (United States)

Yahaya, B.; Ahmed, Y. A.; Balogun, G. I.; Agbo, S. A.

The Nigeria Research Reactor-1 (NIRR-1) is a low power miniature neutron source reactor (MNSR) located at the Centre for Energy Research and Training, Ahmadu Bello University, Zaria Nigeria. The reactor went critical with initial core excess reactivity of 3.77 mk. The NIRR-1 cold excess reactivity measured at the time of commissioning was determined to be 4.97 mk, which is more than the licensed range of 3.5-4 mk. Hence some cadmium poison worth -1.2 mk was inserted into one of the inner irradiation sites which act as reactivity regulating device in order to reduce the core excess reactivity to 3.77 mk, which is within recommended licensed range of 3.5 mk and 4.0 mk. In this present study, the burn-up calculations of the NIRR-1 fuel and the estimation of the core life time expectancy after 10 years (the reactor core expected cycle) have been conducted using the codes WIMS and CITATION. The burn-up analyses carried out indicated that the excess reactivity of NIRR-1 follows a linear decreasing trend having 216 Effective Full Power Days (EFPD) operations. The reactivity worth of top beryllium shim data plates was calculated to be 19.072 mk. The result of depletion analysis for NIRR-1 core shows that (7.9947 ± 0.0008) g of U-235 was consumed for the period of 12 years of operating time. The production of the build-up of Pu-239 was found to be (0.0347 ± 0.0043) g. The core life time estimated in this research was found to be 30.33 years. This is in good agreement with the literature

JOYO MK-III performance test. Criticality test, excess reactivity measurement and burn-up coefficient measurement

International Nuclear Information System (INIS)

Maeda, Shigetaka; Sekine, Takashi; Kitano, Akihiro; Nagasaki, Hideaki

2005-03-01

The MK-III performance test began in June 2003 to fully characterize the upgraded core and heat transfer system of the experimental fast reactor JOYO. This paper describes the results of the approach to criticality, the excess reactivity evaluation and the burn-up coefficient measurement. In the approach to criticality test, the MK-III core achieved initial criticality at the control rod bank position of 412.8 mm on 14:03 July 2nd, 2003. Because the replacement of the outer two rows of reflector subassemblies with shielding subassemblies reduced the source range monitor signals by a factor of 3 at the same reactor power compared with those in the MK-II core, we measured the change of the monitor's response and determined the count rate 2x10 4 cps.' as an appropriate value judging the zero power criticality. In the excess reactivity evaluation, the zero power excess reactivity at 250degC was 2.99±0.10%Δk/kk' based on the measured critical rod bank position and the measured control rod worths. The predicted value by the JOYO core management code system HESTIA was 3.13±0.16%Δk/kk', showing good agreement with the measured value. The measured excess reactivity was within the safety requirement limit. In the burn-up coefficient measurement, the excess reactivity change versus the reactor burn-up was evaluated. The measurement method adopted was to measure the control rod positions during the rated power operation. A value of -2.12x10 -4 Δk/kk'/MWd was obtained as a measured burn-up coefficient. The value calculated by HESTIA was -2.12x10 -4 Δk/kk'/MWd, and it agreed well with the measured value. All technical safety requirements for MK-III core were satisfied and the calculation accuracy of the core management code system HESTIA was confirmed. (author)

Parameterized representation of macroscopic cross section in the PWR fuel element considering burn-up cycles

International Nuclear Information System (INIS)

Belo, Thiago F.; Fiel, Joao Claudio B.

2015-01-01

Nuclear reactor core analysis involves neutronic modeling and the calculations require problem dependent nuclear data generated with few neutron energy groups, as for instance the neutron cross sections. The methods used to obtain these problem-dependent cross sections, in the reactor calculations, generally uses nuclear computer codes that require a large processing time and computational memory, making the process computationally very expensive. Presently, analysis of the macroscopic cross section, as a function of nuclear parameters, has shown a very distinct behavior that cannot be represented by simply using linear interpolation. Indeed, a polynomial representation is more adequate for the data parameterization. To provide the cross sections of rapidly and without the dependence of complex systems calculations, this work developed a set of parameterized cross sections, based on the Tchebychev polynomials, by fitting the cross sections as a function of nuclear parameters, which include fuel temperature, moderator temperature and density, soluble boron concentration, uranium enrichment, and the burn-up. In this study is evaluated the problem-dependent about fission, scattering, total, nu-fission, capture, transport and absorption cross sections for a typical PWR fuel element reactor, considering burn-up cycle. The analysis was carried out with the SCALE 6.1 code package. The results of comparison with direct calculations with the SCALE code system and also the test using project parameters, such as the temperature coefficient of reactivity and fast fission factor, show excellent agreements. The differences between the cross-section parameterization methodology and the direct calculations based on the SCALE code system are less than 0.03 percent. (author)

Parameterized representation of macroscopic cross section in the PWR fuel element considering burn-up cycles

Energy Technology Data Exchange (ETDEWEB)

Belo, Thiago F.; Fiel, Joao Claudio B., E-mail: thiagofbelo@hotmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil)

2015-07-01

Nuclear reactor core analysis involves neutronic modeling and the calculations require problem dependent nuclear data generated with few neutron energy groups, as for instance the neutron cross sections. The methods used to obtain these problem-dependent cross sections, in the reactor calculations, generally uses nuclear computer codes that require a large processing time and computational memory, making the process computationally very expensive. Presently, analysis of the macroscopic cross section, as a function of nuclear parameters, has shown a very distinct behavior that cannot be represented by simply using linear interpolation. Indeed, a polynomial representation is more adequate for the data parameterization. To provide the cross sections of rapidly and without the dependence of complex systems calculations, this work developed a set of parameterized cross sections, based on the Tchebychev polynomials, by fitting the cross sections as a function of nuclear parameters, which include fuel temperature, moderator temperature and density, soluble boron concentration, uranium enrichment, and the burn-up. In this study is evaluated the problem-dependent about fission, scattering, total, nu-fission, capture, transport and absorption cross sections for a typical PWR fuel element reactor, considering burn-up cycle. The analysis was carried out with the SCALE 6.1 code package. The results of comparison with direct calculations with the SCALE code system and also the test using project parameters, such as the temperature coefficient of reactivity and fast fission factor, show excellent agreements. The differences between the cross-section parameterization methodology and the direct calculations based on the SCALE code system are less than 0.03 percent. (author)

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.

SRAC2006: A comprehensive neutronics calculation code system

International Nuclear Information System (INIS)

Okumura, Keisuke; Kugo, Teruhiko; Kaneko, Kunio; Tsuchihashi, Keichiro

2007-02-01

The SRAC is a code system applicable to neutronics analysis of a variety of reactor types. Since the publication of the second version of the users manual (JAERI-1302) in 1986 for the SRAC system, a number of additions and modifications to the functions and the library data have been made to establish a comprehensive neutronics code system. The current system includes major neutron data libraries (JENDL-3.3, JENDL-3.2, ENDF/B-VII, ENDF/B-VI.8, JEFF-3.1, JEF-2.2, etc.), and integrates five elementary codes for neutron transport and diffusion calculation; PIJ based on the collision probability method applicable to 16 kind of lattice models, S N transport codes ANISN(1D) and TWOTRN(2D), diffusion codes TUD(1D) and CITATION(multi-D). The system also includes an auxiliary code COREBN for multi-dimensional core burn-up calculation. (author)

Ultrasonic measurement of high burn-up fuel elastic properties

International Nuclear Information System (INIS)

Laux, D.; Despaux, G.; Augereau, F.; Attal, J.; Gatt, J.; Basini, V.

2006-01-01

The ultrasonic method developed for the evaluation of high burn-up fuel elastic properties is presented hereafter. The objective of the method is to provide data for fuel thermo-mechanical calculation codes in order to improve industrial nuclear fuel and materials or to design new reactor components. The need for data is especially crucial for high burn-up fuel modelling for which the fuel mechanical properties are essential and for which a wide range of experiments in MTR reactors and high burn-up commercial reactor fuel examinations have been included in programmes worldwide. To contribute to the acquisition of this knowledge the LAIN activity is developing in two directions. First one is development of an ultrasonic focused technique adapted to active materials study. This technique was used few years ago in the EdF laboratory in Chinon to assess the ageing of materials under irradiation. It is now used in a hot cell at ITU Karlsruhe to determine the elastic moduli of high burnup fuels from 0 to 110 GWd/tU. Some of this work is presented here. The second on going programme is related to the qualification of acoustic sensors in nuclear environments, which is of a great interest for all the methods, which work, in a hostile nuclear environment

Technical description of the burn-up software system MOP

International Nuclear Information System (INIS)

Schutte, C.K.

1991-05-01

The burn-up software system MOP is a research tool primary intended to study the behaviour of fission products in any reactor composition. Input data are multi-group cross-sections and data concerning the nuclide chains. An option is available to calculate a fundamental mode neutron spectrum for the specified reactor composition. A separate program can test the consistency of the specified nuclide chains. Options are available to calculate time-dependent cross-sections of lumped fission products and to take account of the leakage of gaseous fission products from the reactor core. The system is written in FORTRAN77 for a CYBER computer, using the operating system NOS/BE. The report gives a detailed technical description of the applied algorithms and the flow and storage of data. Information is provided for adapting the system to other computer configurations. (author). 5 refs.; 11 figs

Technique for sensitivity analysis of space- and energy-dependent burn-up calculations

International Nuclear Information System (INIS)

Williams, M.L.; White, J.R.

1979-01-01

A practical method is presented for sensitivity analysis of the very complex, space-energy dependent burn-up equations, in which the neutron and nuclide fields are coupled nonlinearly. The adjoint burn-up equations that are given are in a form which can be directly implemented into multi-dimensional depletion codes, such as VENTURE/BURNER. The data sensitivity coefficients can be used to determine the effect of data uncertainties on time-dependent depletion responses. Initial condition sensitivity coefficients provide a very effective method for computing the change in end of cycle parameters (such as k/sub eff/, fissile inventory, etc.) due to changes in nuclide concentrations at beginning of cycle

Analysis of some antecipated transients without scram for a pressurized water cooled reactor (PWR) using coupling of the containment code CORAN to the system model code ALMOD

International Nuclear Information System (INIS)

Carvalho, F. de A.T. de.

1985-01-01

Some antecipated transients without scram (ATWS) for a pressurized water cooled reactor, model KWU 1300 MWe, are studied using coupling of the containment code CORAN to the system model code ALMOD, under severe random conditions. This coupling has the objective of including containment model as part of a unified code system. These severe conditions include failure of reactor scram, following a station black-out and emergency power initiation for the burn-up status at the beginning and end of the cycle. Furthermore, for the burn-up status at the end of the cycle a failure in the closure of the pressurizer relief valve was also investigated. For the beginning of the cycle, the containment participates actively during the transient. It is noted that the effect of the burn-up in the fuel is to reduce the seriousness of these transients. On the other hand, the failure in the closure of the pressurized relief valve makes this transients more severe. Moreover, the containment safety or radiological public safety is not affected in any of the cases. (Author) [pt

Development of a FBR fuel bundle-duct interaction analysis code-BAMBOO. Analysis model and verification by Phenix high burn-up fuel subassemblies

International Nuclear Information System (INIS)

Uwaba, Tomoyuki; Ito, Masahiro; Ukai, Shigeharu

2005-01-01

The bundle-duct interaction analysis code ''BAMBOO'' has been developed for the purpose of predicting deformation of a wire-wrapped fuel pin bundle of a fast breeder reactor (FBR). The BAMBOO code calculates helical bowing and oval-distortion of all the fuel pins in a fuel subassembly. We developed deformation models in order to precisely analyze the irradiation induced deformation by the code: a model to analyze fuel pin self-bowing induced by circumferential gradient of void swelling as well as thermal expansion, and a model to analyze dispersion of the orderly arrangement of a fuel pin bundle. We made deformation analyses of high burn-up fuel subassemblies in Phenix reactor and compared the calculated results with the post irradiation examination data of these subassemblies for the verification of these models. From the comparison we confirmed that the calculated values of the oval-distortion and bowing reasonably agreed with the PIE results if these models were used in the analysis of the code. (author)

Burn-Up Calculation of the Fuel Element in RSG-GAS Reactor using Program Package BATAN-FUEL

International Nuclear Information System (INIS)

Mochamad Imron; Ariyawan Sunardi

2012-01-01

Calculation of burn lip distribution of 2.96 gr U/cc Silicide fuel element at the 78 th reactor cycle using computer code program of BATAN-FUEL has been done. This calculation uses inputs such as generated power, operation time and a core assumption model of 5/1. Using this calculation model burn up for the entire fuel elements at the reactor core are able to be calculated. From the calculation it is obtained that the minimum burn up of 6.82% is RI-50 at the position of A-9, while the maximum burn up of 57.57% is RI 467 at the position of 8-7. Based on the safety criteria as specified in the Safety Analysis Report (SAR) RSG-GAS reactor, the maximum fuel burn up allowed is 59.59%. It then can be concluded that pattern that elements placement at the reactor core are properly and optimally done. (author)

Sensitivity change of rhodium self -powered detectors with burn-up

International Nuclear Information System (INIS)

Girgis, R.; Akimov, I.S.; Hamouda, I.

1976-01-01

The scope of the present paper is to obtain the calculation formulae to evaluate the rate of sensitivity change of the neutron self-powered detectors with burn-up. A code written in FORTRAN 4 was developed to be operational on the IBM-1130 computer. It has been established in the case of rhodium detectors that neglecting the β-particle absorption in the calculations leads to the underestimation of the detector sensitivity decrease up to 40%. The derived formulae can be used for other self-powered detectors. (author)

Optimalisation Of Oxide Burn-Up Enhanced For RSG-Gas Core

International Nuclear Information System (INIS)

Tukiran; Sembiring, Tagor Malem

2000-01-01

Strategy of fuel management of the RSG-Gas core has been changed from 6/1 to 5/1 pattern so the evaluation of fuel management is necessary to be done. The aim of evaluation is to look for the optimal fuel management so that the fuel can be stayed longer in the core and finally can save cost of operation. Using Batan-EQUIL-2D code did the evaluation of fuel management with 5/1 pattern. The result of evaluation is used to choose which one is more advantage without break the safety margin which is available in the Safety Analysis Report (SAR) firstly, the fuel management was calculated with core excess reactivity of 9,2% criteria. Secondly, fuel burn-up maximum of 56% criteria and the last, fuel burn-up maximum of 64% criteria. From the result of fuel management calculation of the RSG-Gas equilibrium core can be concluded that the optimal RSG-Gas equilibrium core with 5/1 pattern is if the fuel burn-up maximum 64% and the energy in a cycle of operation is 715 MWD. The fuel can be added one more step in the core without break any safety margin. It means that the RSG-Gas equilibrium core can save fuel and cost reduction

Three dimensional Burn-up program parallelization using socket programming

International Nuclear Information System (INIS)

Haliyati R, Evi; Su'ud, Zaki

2002-01-01

A computer parallelization process was built with a purpose to decrease execution time of a physics program. In this case, a multi computer system was built to be used to analyze burn-up process of a nuclear reactor. This multi computer system was design need using a protocol communication among sockets, i.e. TCP/IP. This system consists of computer as a server and the rest as clients. The server has a main control to all its clients. The server also divides the reactor core geometrically to in parts in accordance with the number of clients, each computer including the server has a task to conduct burn-up analysis of 1/n part of the total reactor core measure. This burn-up analysis was conducted simultaneously and in a parallel way by all computers, so a faster program execution time was achieved close to 1/n times that of one computer. Then an analysis was carried out and states that in order to calculate the density of atoms in a reactor of 91 cm x 91 cm x 116 cm, the usage of a parallel system of 2 computers has the highest efficiency

Nuclear fuel burn-up economy

International Nuclear Information System (INIS)

Matausek, M.

1984-01-01

In the period 1981-1985, for the needs of Utility Organization, Beograd, and with the support of the Scientific Council of SR Srbija, work has been performed on the study entitled 'Nuclear Fuel Burn-up Economy'. The forst [phase, completed during the year 1983 comprised: comparative analysis of commercial NPP from the standpoint of nuclear fuel requirements; development of methods for fuel burn-up analysis; specification of elements concerning the nuclear fuel for the tender documentation. The present paper gives the short description of the purpose, content and results achieved in the up-to-now work on the study. (author)

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...

Simulation of the neutron-physical properties of the classical UO2 fuel and of MOX fuel during the burn-up by Transuranus

International Nuclear Information System (INIS)

Breza, J. jr.; Necas, V.; Daoeilek, P.

2005-01-01

The classical nuclear fuel UO 2 is well known for VVER reactors. Nevertheless, in the near future it will be possible to replace this fuel by novel, advanced kinds of fuel, for instance MOX, inert matrices fuel, etc., that will allow to increase the level of burn-up and minimize the amount of hazardous waste. The code Transuranus [2], designed at ITU Karlsruhe, is intended for thermal and mechanical analyses of fuel elements in nuclear reactors. We have utilized the code Transuranus to simulate the neutron-physical properties of the classical UO 2 fuel and of MOX fuel during the burn-up to a level of 40 MWd/kgHM. We compare obtained results of uranium and plutonium nuclides concentrations, their changes during burn-up, with results obtained by code HELIOS [3], which is well-validated code for this kind of applications. We performed calculations of fission gasses concentrations, namely xenon and krypton. (author)

Development of external coupling for calculation of the control rod worth in terms of burn-up for a WWER-1000 nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Noori-Kalkhoran, Omid, E-mail: o_noori@yahoo.com [Reactor Research School, Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of); Yarizadeh-Beneh, Mehdi [Faculty of Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of); Ahangari, Rohollah [Reactor Research School, Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of)

2016-08-15

Highlights: • Calculation of control rod worth in term of burn-up. • Calculation of differential and integral control rod worth. • Developing an external couple. • Modification of thermal-hydraulic profiles in calculations. - Abstract: One of the main problems relating to operation of a nuclear reactor is its safety and controlling system. The most widely used control systems for thermal reactors are neutron absorbent rods. In this study a code based method has been developed for calculation of integral and differential control rod worth in terms of burn-up for a WWER-1000 nuclear reactor. External coupling of WIMSD-5B, PARCS V2.7 and COBRA-EN has been used for this purpose. WIMSD-5B has been used for cell calculation and handling burn-up of the core in various days. PARCS V2.7 has been used for neutronic calculation of core and critical boron concentration search. Thermal-hydraulic calculation has been performed by COBRA-EN. An external coupling algorithm has been developed by MATLAB to couple and transfer suitable data between these codes in each step. Steady-State Power Picking Factors (PPFs) of the core and control rod worth for different control rod groups have been calculated from Beginning Of Cycle (BOC) to 289.7 Effective Full Power Days (EFPDs) in some steps. Results have been compared with the results of Bushehr Nuclear Power Plant (BNPP) Final Safety Analysis Report (FSAR). The results show a good agreement and confirm the ability of developed coupling in calculation of control rod worth in terms of burn-up.

Burn-up determination of irradiated thoria samples by isotope dilution-thermal ionisation mass spectrometry

International Nuclear Information System (INIS)

Aggarwal, S.K.; Jaison, P.G.; Telmore, V.M.; Shah, R.V.; Sant, V.L.; Sasibhushan, K.; Parab, A.R.; Alamelu, D.

2010-03-01

Burn-up was determined experimentally using thermal ionization mass spectrometry for two samples from ThO 2 bundles irradiated in KAPS-2. This involved quantitative dissolution of the irradiated fuel samples followed by separation and determination of Th, U and a stable fission product burn-up monitor in the dissolved fuel solution. Stable fission product 148 Nd was used as a burn-up monitor for determining the number of fissions. Isotope Dilution-Thermal Ionisation Mass Spectrometry (ID-TIMS) using natural U, 229 Th and enriched 142 Nd as spikes was employed for the determination of U, Th and Nd, respectively. Atom % fission values of 1.25 ± 0.03 were obtained for both the samples. 232 U content in 233 U determined by alpha spectrometry was about 500 ppm and this was higher by a factor of 5 compared to the theoretically predicted value by ORIGEN-2 code. (author)

FUEL BURN-UP CALCULATION FOR WORKING CORE OF THE RSG-GAS RESEARCH REACTOR AT BATAN SERPONG

Directory of Open Access Journals (Sweden)

Tukiran Surbakti

2017-12-01

Full Text Available The neutronic parameters are required in the safety analysis of the RSG-GAS research reactor. The RSG-GAS research reactor, MTR (Material Testing Reactor type is used for research and also in radioisotope production. RSG-GAS has been operating for 30 years without experiencing significant obstacles. It is managed under strict requirements, especially fuel management and fuel burn-up calculations. The reactor is operated under the supervision of the Regulatory Body (BAPETEN and the IAEA (International Atomic Energy Agency. In this paper, the experience of managing RSG-GAS core fuels will be discussed, there are hundred possibilities of fuel placements on the reactor core and the strategy used to operate the reactor will be crucial. However, based on strict calculation and supervision, there is no incorrect placement of the fuels in the core. The calculations were performed on working core by using the WIMSD-5B computer code with ENDFVII.0 data file to generate the macroscopic cross-section of fuel and BATAN-FUEL code were used to obtain the neutronic parameter value such as fuel burn-up fractions. The calculation of the neutronic core parameters of the RSG-GAS research reactor was carried out for U3Si2-Al fuel, 250 grams of mass, with an equilibrium core strategy. The calculations show that on the last three operating cores (T90, T91, T92, all fuels meet the safety criteria and the fuel burn-up does not exceed the maximum discharge burn-up of 59%. Maximum fuel burn-up always exists in the fuel which is close to the position of control rod.

Estimation of the impact of manufacturing tolerances on burn-up calculations using Monte Carlo techniques

Energy Technology Data Exchange (ETDEWEB)

Bock, M.; Wagner, M. [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH, Garching (Germany). Forschungszentrum

2012-11-01

tool SUnCISTT (Sensitivities and Uncertainties in Criticality Inventory and Source Term Tool). The SUnCISTT defines an interface between the well established GRS tool for uncertainty and sensitivity analyses SUSA and codes used in the nuclear fuel cycle. In the context of the analysis presented here, the GRS burn-up system OREST is coupled. The coupling between these two tools will be outlined, the available features of this new application will be presented, and exemplary results will be shown. Finally, an outlook on future developments and future applications will be given. (orig.)

Improvement of JRR-4 core management code system

International Nuclear Information System (INIS)

Izumo, H.; Watanabe, S.; Nagatomi, H.; Hori, N.

2000-01-01

In the modification of JRR-4, the fuel was changed from 93% high enrichment uranium aluminized fuel to 20% low enriched uranium silicide fuel in conformity with the framework of reduced enrichment program on JAERI research reactors. As changing of this, JRR-4 core management code system which estimates excess reactivity of core, fuel burn-up and so on, was improved too. It had been difficult for users to operate the former code system because its input-output form was text-form. But, in the new code system (COMMAS-JRR), users are able to operate the code system without using difficult text-form input. The estimation results of excess reactivity of JRR-4 LEU fuel core were showed very good agreements with the measured value. It is the strong points of this new code system to be operated simply by using the windows form pictures act on a personal workstation equip with the graphical-user-interface (GUI), and to estimate accurately the specific characteristics of the LEU core. (author)

Fundamental burn-up mode in a pebble-bed type reactor

International Nuclear Information System (INIS)

Chen, Xue-Nong; Kiefhaber, Edgar; Maschek, Werner

2008-01-01

This paper deals with a pebble-bed type reactor, in which the fuel is loaded from one side (top) and discharged from the other side (bottom). A boundary value problem of a single group diffusion equation coupled with simplified burn-up equations is studied, where the natural radioactive decay processes are neglected in the burn-up modelling. An asymptotic burning wave solution is found analytically in the one-dimensional case, which is called as fundamental burn-up mode. Among this solution family there are two particular cases, namely, a classic fundamental solution with a zero burn-up and a partial solitary burn-up wave solution with a highest burn-up. An example of Th-U conversion is considered and the solutions are presented in order to show the mechanism of the burning wave. (author)

Detonation of high explosives in Lagrangian hydrodynamic codes using the programmed burn technique

International Nuclear Information System (INIS)

Berger, M.E.

1975-09-01

Two initiation methods were developed for improving the programmed burn technique for detonation of high explosives in smeared-shock Lagrangian hydrodynamic codes. The methods are verified by comparing the improved programmed burn with existing solutions in one-dimensional plane, converging, and diverging geometries. Deficiencies in the standard programmed burn are described. One of the initiation methods has been determined to be better for inclusion in production hydrodynamic codes

User's guide for FRMOD, a zero dimensional FRM burn code

International Nuclear Information System (INIS)

Driemeryer, D.; Miley, G.H.

1979-01-01

The zero-dimensional FRM plasma burn code, FRMOD is written in the FORTRAN language and is currently available on the Control Data Corporation (CDC) 7600 computer at the Magnetic Fusion Energy Computer Center (MFECC), sponsored by the US Department of Energy, in Livermore, CA. This guide assumes that the user is familiar with the system architecture and some of the utility programs available on the MFE-7600 machine, since online documentation is available for system routines through the use of the DOCUMENT utility. Users may therefore refer to it for answers to system related questions

The DACC system. Code burnup of cell for projection of the fuel elements in the power net work PWR and BWR

International Nuclear Information System (INIS)

Cepraga, D.; Boeriu, St.; Gheorghiu, E.; Cristian, I.; Patrulescu, I.; Cimporescu, D.; Ciuvica, P.; Velciu, E.

1975-01-01

The calculation system DACC-5 is a zero-dimensional reactor physics code used to calculate the criticality and burn-up of light-water reactors. The code requires as input essential extensive reactor parameters (fuel rod radius, water density, etc.). The nuclear constants (intensive parameters) are calculated with a five-group model (2 thermal and 3 fast groups). A fitting procedure is systematically employed to reduce the computation time of the code. Zero-dimensional burn-up calculations are made in an automatic way. Part one of the paper contains the code physical model and computer structure. Part two of the paper will contain tests of DACC-5 credibility for different light-water power lattices

Monte Carlo sampling on technical parameters in criticality and burn-up-calculations

International Nuclear Information System (INIS)

Kirsch, M.; Hannstein, V.; Kilger, R.

2011-01-01

The increase in computing power over the recent years allows for the introduction of Monte Carlo sampling techniques for sensitivity and uncertainty analyses in criticality safety and burn-up calculations. With these techniques it is possible to assess the influence of a variation of the input parameters within their measured or estimated uncertainties on the final value of a calculation. The probabilistic result of a statistical analysis can thus complement the traditional method of figuring out both the nominal (best estimate) and the bounding case of the neutron multiplication factor (k eff ) in criticality safety analyses, e.g. by calculating the uncertainty of k eff or tolerance limits. Furthermore, the sampling method provides a possibility to derive sensitivity information, i.e. it allows figuring out which of the uncertain input parameters contribute the most to the uncertainty of the system. The application of Monte Carlo sampling methods has become a common practice in both industry and research institutes. Within this approach, two main paths are currently under investigation: the variation of nuclear data used in a calculation and the variation of technical parameters such as manufacturing tolerances. This contribution concentrates on the latter case. The newly developed SUnCISTT (Sensitivities and Uncertainties in Criticality Inventory and Source Term Tool) is introduced. It defines an interface to the well established GRS tool for sensitivity and uncertainty analyses SUSA, that provides the necessary statistical methods for sampling based analyses. The interfaced codes are programs that are used to simulate aspects of the nuclear fuel cycle, such as the criticality safety analysis sequence CSAS5 of the SCALE code system, developed by Oak Ridge National Laboratories, or the GRS burn-up system OREST. In the following, first the implementation of the SUnCISTT will be presented, then, results of its application in an exemplary evaluation of the neutron

Application of reactivity method to MTR fuel burn-up measurement

International Nuclear Information System (INIS)

Zuniga, A.; Ravnik, M.; Cuya, R.

2001-01-01

Fuel element burn-up has been measured for the first time by reactivity method in a MTR reactor. The measurement was performed in RP-10 reactor of Peruvian Institute for Nuclear Energy (IPEN) in Lima. It is a pool type 10MW material testing reactor using standard 20% enriched uranium plate type fuel elements. A fresh element and an element with well defined burn-up were selected as reference elements. Several elements in the core were selected for burn-up measurement. Each of them was replaced in its original position by both reference elements. Change in excess reactivity was measured using control rod calibration curve. The burn-up reactivity worth of fuel elements was plotted as a function of their calculated burnup. Corrected burn-up values of the measured fuel elements were calculated using the fitting function at experimental reactivity for all elements. Good agreement between measured and calculated burn-up values was observed indicating that the reactivity method can be successfully applied also to MTR fuel element burn-up determination.(author)

High-fidelity plasma codes for burn physics

Energy Technology Data Exchange (ETDEWEB)

Cooley, James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Graziani, Frank [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Marinak, Marty [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Murillo, Michael [Michigan State Univ., East Lansing, MI (United States)

2016-10-19

Accurate predictions of equation of state (EOS), ionic and electronic transport properties are of critical importance for high-energy-density plasma science. Transport coefficients inform radiation-hydrodynamic codes and impact diagnostic interpretation, which in turn impacts our understanding of the development of instabilities, the overall energy balance of burning plasmas, and the efficacy of self-heating from charged-particle stopping. Important processes include thermal and electrical conduction, electron-ion coupling, inter-diffusion, ion viscosity, and charged particle stopping. However, uncertainties in these coefficients are not well established. Fundamental plasma science codes, also called high-fidelity plasma codes, are a relatively recent computational tool that augments both experimental data and theoretical foundations of transport coefficients. This paper addresses the current status of HFPC codes and their future development, and the potential impact they play in improving the predictive capability of the multi-physics hydrodynamic codes used in HED design.

Windows user-friendly code package development for operation of research reactors

International Nuclear Information System (INIS)

Hoang Anh Tuan

1998-01-01

The content of the project was to developed: 1. MS Windows interface to spectral codes like THERMOS, PEACO-COLLIS, GRACE and burn-up code. 2. MS Windows C-language burn-up diffusion hexagonal lattice code. The overall scope of the project was to develop a PC-based MS Windows code package for operation of Dalat research reactor. Various problems relating to neutronic physics like thermalization, resonance treatment, fast spectral treatment, change of isotopic concentration during burn-up time as well as burn-up distribution in the reactor core are considered in parallel to application of informatics technique. The developing process is a subject of the concept of user-friendly interface between end-users and the code package. High level input features through system of icon, menu, dialog box with regard to Common User Access (CUA) convention and sophisticated graphical output in MS Windows environment was used. The user-computer interface is also enhanced by using both keyboard and mouse, which creates a very natural manner for end-user. (author)

Use of system code to estimate equilibrium tritium inventory in fusion DT machines, such as ARIES-AT and components testing facilities

International Nuclear Information System (INIS)

Wong, C.P.C.; Merrill, B.

2014-01-01

Highlights: • With the use of a system code, tritium burn-up fraction (f burn ) can be determined. • Initial tritium inventory for steady state DT machines can be estimated. • f burn of ARIES-AT, CFETR and FNSF-AT are in the range of 1–2.8%. • Respective total tritium inventories of are 7.6 kg, 6.1 kg, and 5.2 kg. - Abstract: ITER is under construction and will begin operation in 2020. This is the first 500 MW fusion class DT device, and since it is not going to breed tritium, it will consume most of the limited supply of tritium resources in the world. Yet, in parallel, DT fusion nuclear component testing machines will be needed to provide technical data for the design of DEMO. It becomes necessary to estimate the tritium burn-up fraction and corresponding initial tritium inventory and the doubling time of these machines for the planning of future supply and utilization of tritium. With the use of a system code, tritium burn-up fraction and initial tritium inventory for steady state DT machines can be estimated. Estimated tritium burn-up fractions of FNSF-AT, CFETR-R and ARIES-AT are in the range of 1–2.8%. Corresponding total equilibrium tritium inventories of the plasma flow and tritium processing system, and with the DCLL blanket option are 7.6 kg, 6.1 kg, and 5.2 kg for ARIES-AT, CFETR-R and FNSF-AT, respectively

MTR fuel element burn-up measurements by the reactivity method

International Nuclear Information System (INIS)

Zuniga, A.; Cuya, T.R.; Ravnik, M.

2003-01-01

Fuel element burn-up was measured by the reactivity method in the 10 MW Peruvian MTR reactor RP-10. The main purpose of the experiment was testing the reactivity method for an MTR reactor as the reactivity method was originally developed for TRIGA reactors. The reactivity worth of each measured fuel element was measured in its original core position in order to measure the burn-up of the fuel elements that were part of the experimental core. The burn-up of each measured fuel element was derived by interpolating its reactivity worth from the reactivity worth of two reference fuel elements of known burn-up, whose reactivity worth was measured in the position of the measured fuel element. The accuracy of the method was improved by separating the reactivity effect of burn-up from the effect of the position in the core. The results of the experiment showed that the modified reactivity method for fuel element burn-up determination could be applied also to MTR reactors. (orig.)

ETR/ITER systems code

Energy Technology Data Exchange (ETDEWEB)

Barr, W.L.; Bathke, C.G.; Brooks, J.N.; Bulmer, R.H.; Busigin, A.; DuBois, P.F.; Fenstermacher, M.E.; Fink, J.; Finn, P.A.; Galambos, J.D.; Gohar, Y.; Gorker, G.E.; Haines, J.R.; Hassanein, A.M.; Hicks, D.R.; Ho, S.K.; Kalsi, S.S.; Kalyanam, K.M.; Kerns, J.A.; Lee, J.D.; Miller, J.R.; Miller, R.L.; Myall, J.O.; Peng, Y-K.M.; Perkins, L.J.; Spampinato, P.T.; Strickler, D.J.; Thomson, S.L.; Wagner, C.E.; Willms, R.S.; Reid, R.L. (ed.)

1988-04-01

A tokamak systems code capable of modeling experimental test reactors has been developed and is described in this document. The code, named TETRA (for Tokamak Engineering Test Reactor Analysis), consists of a series of modules, each describing a tokamak system or component, controlled by an optimizer/driver. This code development was a national effort in that the modules were contributed by members of the fusion community and integrated into a code by the Fusion Engineering Design Center. The code has been checked out on the Cray computers at the National Magnetic Fusion Energy Computing Center and has satisfactorily simulated the Tokamak Ignition/Burn Experimental Reactor II (TIBER) design. A feature of this code is the ability to perform optimization studies through the use of a numerical software package, which iterates prescribed variables to satisfy a set of prescribed equations or constraints. This code will be used to perform sensitivity studies for the proposed International Thermonuclear Experimental Reactor (ITER). 22 figs., 29 tabs.

ETR/ITER systems code

International Nuclear Information System (INIS)

Barr, W.L.; Bathke, C.G.; Brooks, J.N.

1988-04-01

A tokamak systems code capable of modeling experimental test reactors has been developed and is described in this document. The code, named TETRA (for Tokamak Engineering Test Reactor Analysis), consists of a series of modules, each describing a tokamak system or component, controlled by an optimizer/driver. This code development was a national effort in that the modules were contributed by members of the fusion community and integrated into a code by the Fusion Engineering Design Center. The code has been checked out on the Cray computers at the National Magnetic Fusion Energy Computing Center and has satisfactorily simulated the Tokamak Ignition/Burn Experimental Reactor II (TIBER) design. A feature of this code is the ability to perform optimization studies through the use of a numerical software package, which iterates prescribed variables to satisfy a set of prescribed equations or constraints. This code will be used to perform sensitivity studies for the proposed International Thermonuclear Experimental Reactor (ITER). 22 figs., 29 tabs

Development of a BWR core burn-up calculation code COREBN-BWR

International Nuclear Information System (INIS)

Morimoto, Yuichi; Okumura, Keisuke

1992-05-01

In order to evaluate core performances of BWR type reactors, the three dimensional core burnup calculation code COREBN-BWR and the fuel management code HIST-BWR have been developed. In analyses of BWR type reactors, thermal hydraulics calculations must be coupled with neutronics calculations to evaluate core performances, because steam void distribution changes according to the change of the power distribution. By installing new functions as follows to the three dimensional core burnup code COREBN2 developed in JAERI for PWR type reactor analyses, the code system becomes to be applicable to burnup analyses of BWR type reactors. (1) Macroscopic cross section calculation function taking into account of coolant void distribution. (2) Thermal hydraulics calculation function to evaluate core flow split, coolant void distribution and thermal margin. (3) Burnup calculation function under the Haling strategy. (4) Fuel management function to incorporate the thermal hydraulics information. This report consists of the general description, calculational models, input data requirements and their explanations, detailed information on usage and sample input. (author)

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.

Determination of burn-up of irradiated nuclear fuels using mass spectrometry

International Nuclear Information System (INIS)

Jagadish Kumar, S.; Telmore, V.M.; Shah, R.V.; Sasi Bhushan, K.; Paul, Sumana; Kumar, Pranaw; Rao, Radhika M.; Jaison, P.G.

2017-01-01

Burn-up defined as the atom percent fission, is a vital parameter used for assessing the performance of nuclear fuel during its irradiation in the reactor. Accurate data on the actinide isotopes are also essential for the reliable accountability of nuclear materials and for nuclear safeguards. Both destructive and non-destructive methods are employed in the post-irradiation analysis for the burn-up measurements. Though non-destructive methods are preferred from the point view of remote handling of irradiated fuels with high radioactivity, they do not provide the high accuracy as achieved by the chemical analysis methods. Thus destructive radiochemical and chemical analyses are still the established reference methods for accurate and reliable burn-up determination of irradiated nuclear fuels. In the destructive method, burn-up of irradiated nuclear fuel is determined by correlating the amount of a fission product formed during irradiation with that of heavy elements. Thus the destructive experimental determination of burn-up involves the dissolution of irradiated fuel samples followed by the separation and determination of heavy elements and fission product(s) to be used as burn-up monitor(s). Another approach for the experimental determination of burn-up is based on the changes in the abundances of the heavy element isotopes. A widely accepted method for burn-up determination is based on stable "1"4"8Nd and "1"3"9La as burn-up monitors. Several properties such as non-volatility, nearly same yields for thermal fissions of "2"3"5U and "2"3"9Pu etc justifies the selection of "1"4"8Nd as a burn-up monitor

Increased fuel burn-up and fuel cycle equilibrium

International Nuclear Information System (INIS)

Debes, M.

2001-01-01

Improvement of nuclear competitiveness will rely mainly on increased fuel performance, with higher burn-up, and reactors sustained life. Regarding spent fuel management, the EDF current policy relies on UO 2 fuel reprocessing (around 850 MTHM/year at La Hague) and MOX recycling to ensure plutonium flux adequacy (around 100 MTHM/year, with an electricity production equivalent to 30 TWh). This policy enables to reuse fuel material, while maintaining global kWh economy with existing facilities. It goes along with current perspective to increase fuel burn-up up to 57 GWday/t mean in 2010. The following presentation describes the consequences of higher fuel burn-up on fuel cycle and waste management and implementation of a long term and global equilibrium for decades in spent fuel management resulting from this strategy. (author)

Use of system code to estimate equilibrium tritium inventory in fusion DT machines, such as ARIES-AT and components testing facilities

Energy Technology Data Exchange (ETDEWEB)

Wong, C.P.C., E-mail: wongc@fusion.gat.com [General Atomics, San Diego, CA (United States); Merrill, B. [Idaho National Laboratory, Idaho Falls, ID (United States)

2014-10-15

Highlights: • With the use of a system code, tritium burn-up fraction (f{sub burn}) can be determined. • Initial tritium inventory for steady state DT machines can be estimated. • f{sub burn} of ARIES-AT, CFETR and FNSF-AT are in the range of 1–2.8%. • Respective total tritium inventories of are 7.6 kg, 6.1 kg, and 5.2 kg. - Abstract: ITER is under construction and will begin operation in 2020. This is the first 500 MW{sub fusion} class DT device, and since it is not going to breed tritium, it will consume most of the limited supply of tritium resources in the world. Yet, in parallel, DT fusion nuclear component testing machines will be needed to provide technical data for the design of DEMO. It becomes necessary to estimate the tritium burn-up fraction and corresponding initial tritium inventory and the doubling time of these machines for the planning of future supply and utilization of tritium. With the use of a system code, tritium burn-up fraction and initial tritium inventory for steady state DT machines can be estimated. Estimated tritium burn-up fractions of FNSF-AT, CFETR-R and ARIES-AT are in the range of 1–2.8%. Corresponding total equilibrium tritium inventories of the plasma flow and tritium processing system, and with the DCLL blanket option are 7.6 kg, 6.1 kg, and 5.2 kg for ARIES-AT, CFETR-R and FNSF-AT, respectively.

Hydrogen burn assessment with the CONTAIN code

International Nuclear Information System (INIS)

van Rij, H.M.

1986-01-01

The CONTAIN computer code was developed at Sandia National Laboratories, under contract to the US Nuclear Regulatory Commission (NRC). The code is intended for calculations of containment loads during severe accidents and for prediction of the radioactive source term in the event that the containment leaks or fails. A strong point of the CONTAIN code is the continuous interaction of the thermal-hydraulics phenomena, aerosol behavior and fission product behavior. The CONTAIN code can be used for Light Water Reactors as well as Liquid Metal Reactors. In order to evaluate the CONTAIN code on its merits, comparisons between the code and experiments must be made. In this paper, CONTAIN calculations for the hydrogen burn experiments, carried out at the Nevada Test Site (NTS), are presented and compared with the experimental data. In the Large-Scale Hydrogen Combustion Facility at the NTS, 21 tests have been carried out. These tests were sponsored by the NRC and the Electric Power Research Institute (EPRI). The tests, carried out by EG and G, were performed in a spherical vessel 16 m in diameter with a design pressure of 700 kPa, substantially higher than that of most commercial nuclear containment buildings

Treatment and follow-up results of children with electrical burn who observed in burn intensive care unit

Directory of Open Access Journals (Sweden)

Çiğdem Aliosmanoğlu

2011-06-01

Full Text Available Electrical burns are infrequent relative to other injuries, but they are associated with high morbidity and mortality. The aim of this study was to assess management and follow-up results of pediatric patients’ who observed in intensive care unit and also review the precautions for preventing electrical burns.Materials and methods: Totally 22 patients aged under 17 years who were observed in the burn intensive care unit of Şanlıurfa Education and Research Hospital during the period between July 2009-October 2010. Cases were investigated retrospectively. The patients’ age, gender, total burn surface area, length of stay in hospital, musculo-skeletal system complication, cardiovascular system complication, kidney damage and attempts were recorded.Results: Of the 22 cases, 19 (86.3% were male and 3 (13.7% were female. The mean age of the patients was 11.5 years. In 10 (45.4% children burns were occurred in workplace and working area and 12 (54.6% were occurred in the home environment. Depth of burns were third degree in 10 (45.4% children and second degree in 12 (54.6%. The mean percentage of burn surface area was 25.9%. The mean length of stay in hospital was 17 days. Debridement and grafting were performed to 12 (54.6% cases and 10 (45.4% children were treated with dressings. No patient had increased creatinine kinase levels, oliguria, myoglobuinuria and arrhythmia. The mean hospitalization time was 17 days.Conclusion: Nearly half of patients underwent debridement plus grafting. None of our patients developed renal failure other severe system dysfunction.

Observations on the CANDLE burn-up in various geometries

International Nuclear Information System (INIS)

Seifritz, W.

2007-01-01

We have looked at all geometrical conditions under which an auto catalytically propagating burnup wave (CANDLE burn-up) is possible. Thereby, the Sine Gordon equation finds a new place in the burn-up theory of nuclear fission reactors. For a practical reactor design the axially burning 'spaghetti' reactor and the azimuthally burning 'pancake' reactor, respectively, seem to be the most promising geometries for a practical reactor design. Radial and spherical burn-waves in cylindrical and spherical geometry, respectively, are principally impossible. Also, the possible applicability of such fission burn-waves on the OKLO-phenomenon and the GEOREACTOR in the center of Earth, postulated by Herndon, is discussed. A fast CANDLE-reactor can work with only depleted uranium. Therefore, uranium mining and uranium-enrichment are not necessary anymore. Furthermore, it is also possible to dispense with reprocessing because the uranium utilization factor is as high as about 40%. Thus, this completely new reactor type can open a new era of reactor technology

High-burn-up fuels for fast reactors. Past experience and novel applications

International Nuclear Information System (INIS)

Weaver, Kevan D.; Gilleland, John; Whitmer, Charles; Zimmerman, George

2009-01-01

Fast reactors in the U.S. routinely achieved fuel burn-ups of 10%, with some fuel able to reach peak burn-ups of 20%, notably in the Experimental Breeder Reactor II and the Fast Flux Test Facility. Maximum burn-up has historically been constrained by chemical and mechanical interactions between the fuel and its cladding, and to some extent by radiation damage and thermal effects (e.g., radiation-induced creep, thermal creep, and radiation embrittlement) that cause the cladding to weaken. Although fast reactors have used several kinds of fuel - including oxide, metal alloy, carbide, and nitride - the vast majority of experience with fast reactors has been using oxide (including mixed oxide) and metal-alloy fuels based on uranium. Our understanding of high-burn-up operation is also limited by the fact that breeder reactor programs have historically assumed that their fuel would eventually undergo reprocessing; the programs thus have not made high burn-up a top priority. Recently a set of novel designs have emerged for fast reactors that require little initial enrichment and no reprocessing. These reactors exploit a concept known as a traveling wave (sometimes referred to as a breed-and-burn wave, fission wave, or nuclear-burning wave). By breeding and using its own fuel in place as it operates, a traveling-wave reactor can obtain burn-ups that approach 50%, well beyond the current base of knowledge and experience. Our computational work on the physics of traveling-wave reactors shows that they require metal-alloy fuel to provide the margins of reactivity necessary to sustain a breed-and-burn wave. This paper reviews operating experience with high-burn-up fuels and the technical feasibility of moving to a qualitatively new burn-up regime. We discuss our calculations on traveling-wave reactors, including those concerning the possible use of thorium. The challenges associated with high burn-up and fluence in fuels and materials are also discussed. (author)

Comparison of measured and calculated burn-up of AVR-Fuel-Elements

Energy Technology Data Exchange (ETDEWEB)

Wagemann, R.

1974-03-15

Burn-up comparisons are made for small batches of three types of AVR fuel elements using a coupled EREBUS-MUPO neutronic analysis compared against test results from both nondestructive gamma-ray measurements of cesium-137 activity and destructive mass spectrometry measurements of the ratio of U-233 to U-235. The comparisons are relatively good for average burn-up and reasonably good for burn-up distributions.

Fuel burn-up distribution and transuranic nuclide contents produced at the first cycle operation of AP1000

International Nuclear Information System (INIS)

Jati Susilo; Jupiter Sitorus Pane

2016-01-01

AP1000 reactor core was designed with nominal power of 1154 MWe (3415 MWth), operated within life time of 60 years and cycle length of 18 months. For the first cycle, the AP1000 core uses three kinds of UO 2 enrichment, they are 2.35 w/o, 3.40 w/o and 4.45 w/o. Absorber materials such as ZrB 2 , Pyrex and Boron solution are used to compensate the excess reactivity at the beginning of cycle. In the core, U-235 fuels are burned by fission reaction and produce energy, fission products and new neutron. Because of the U-238 neutron absorption reaction, the high level radioactive waste of heavy nuclide transuranic such as Pu, Am, Cm and Np are also generated. They have a very long half life. The purpose of this study is to evaluate the result of fuel burn-up distribution and heavy nuclide transuranic contents produced by AP1000 at the end of first cycle operation (EOFC). Calculation of ¼ part of the AP1000 core in the 2 dimensional model has been done using SRAC2006 code with the module of COREBN/HIST. The input data called the table of macroscopic cross section, is calculated using module of PIJ. The result shows that the maximum fuel assembly (FA) burn-up is 27.04 GWD/MTU, that is still lower than allowed maximum burn-up of 62 GWD/MTU. Fuel loading position at the center/middle of the core will produce bigger burn-up and transuranic nuclide than one at the edges the of the core. The use of IFBA fuel just give a small effect to lessen the fuel burn-up and transuranic nuclide production. (author)

Effect of local burn-up variation on computed mean nuclide concentrations

International Nuclear Information System (INIS)

Moeller, W.

1982-01-01

Mean concentrations of U-235, U-236, U-238, Pu-239, Pu-240, Pu-241 and Pu-242 in some volume areas of WWER-440 fuel assemblies have been calculated from corresponding burn-up microdistribution data and compared with those calculated from burn-up mean values. Differences occurring were below 3% for the uranium nuclides but, at low burn-ups, considerable for Pu-241 and Pu-242. (author)

Review of high burn-up RIA and LOCA database and criteria

International Nuclear Information System (INIS)

Vitanza, C.; Hrehor, M.

2006-01-01

This document is intended to provide regulators, their technical support organizations and industry with a concise review of existing fuel experimental data at RIA and LOCA conditions and considerations on how these data affect fuel safety criteria at increasing burn-up. It mostly addresses experimental results relevant to BWR and PWR fuel and it encompasses several contributions from the various experts that participated in the CSNI SEGFSM activities. It also covers the information presented at the joint CSNI/CNRA Topical Discussion on high burn-up fuel issues that took place on this subject in December 2004. The report is organized in the following way: the CABRI RIA database (14 tests), the NSRR database (26 tests) and other databases, RIA failure thresholds, comparison of failure thresholds for the HZP case, LOCA database ductility tests and quench tests, LOCA safety limit, provisional burn-up dependent criterion for Zr-4. The conclusions are as follows. On RIA, there is a well-established testing method and a significant and relatively consistent database from NSRR and Cabri tests, especially on high burn-up Zr-2 and Zr-4 cladding. It is encouraging that several correlations have been proposed for the RIA fuel failure threshold. Their predictions are compared and discussed in this paper for a representative PWR case. On LOCA, there are two different test methods, one based on ductility determinations and the other based on 'integral' quench tests. The LOCA database at high burn-up is limited to both testing methods. Ductility tests carried out with pre-hydrided non-irradiated cladding show a pronounced hydrogen effect. Data for actual high burn-up specimens are being gathered in various laboratories and will form the basis for a burn-up dependent LOCA limit. A provisional burn-up dependent criterion is discussed in the paper

Effect of error propagation of nuclide number densities on Monte Carlo burn-up calculations

International Nuclear Information System (INIS)

Tohjoh, Masayuki; Endo, Tomohiro; Watanabe, Masato; Yamamoto, Akio

2006-01-01

As a result of improvements in computer technology, the continuous energy Monte Carlo burn-up calculation has received attention as a good candidate for an assembly calculation method. However, the results of Monte Carlo calculations contain the statistical errors. The results of Monte Carlo burn-up calculations, in particular, include propagated statistical errors through the variance of the nuclide number densities. Therefore, if statistical error alone is evaluated, the errors in Monte Carlo burn-up calculations may be underestimated. To make clear this effect of error propagation on Monte Carlo burn-up calculations, we here proposed an equation that can predict the variance of nuclide number densities after burn-up calculations, and we verified this equation using enormous numbers of the Monte Carlo burn-up calculations by changing only the initial random numbers. We also verified the effect of the number of burn-up calculation points on Monte Carlo burn-up calculations. From these verifications, we estimated the errors in Monte Carlo burn-up calculations including both statistical and propagated errors. Finally, we made clear the effects of error propagation on Monte Carlo burn-up calculations by comparing statistical errors alone versus both statistical and propagated errors. The results revealed that the effects of error propagation on the Monte Carlo burn-up calculations of 8 x 8 BWR fuel assembly are low up to 60 GWd/t

Experimental studies of spent fuel burn-up in WWR-SM reactor

Energy Technology Data Exchange (ETDEWEB)

Alikulov, Sh. A.; Baytelesov, S.A.; Boltaboev, A.F.; Kungurov, F.R. [Institute of Nuclear Physics, Ulughbek township, 100214, Tashkent (Uzbekistan); Menlove, H.O.; O’Connor, W. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Osmanov, B.S., E-mail: bari_osmanov@yahoo.com [Research Institute of Applied Physics, Vuzgorodok, 100174 Tashkent (Uzbekistan); Salikhbaev, U.S. [Institute of Nuclear Physics, Ulughbek township, 100214, Tashkent (Uzbekistan)

2014-10-01

Highlights: • Uranium burn-up measurement from {sup 137}Cs activity in spent reactor fuel. • Comparison to reference sample with known burn-up value (ratio method). • Cross-check of the approach with neutron-based measurement technique. - Abstract: The article reports the results of {sup 235}U burn-up measurements using {sup 137}Cs activity technique for 12 nuclear fuel assemblies of WWR-SM research reactor after 3-year cooling time. The discrepancy between the measured and the calculated burn-up values was about 3%. To increase the reliability of the data and for cross-check purposes, neutron measurement approach was also used. Average discrepancy between two methods was around 12%.

UO2 fuel behaviour at rod burn-ups up to 105 MWd/kgHM. A review of 10 years of high burn-up examinations commissioned by AREVA NP

International Nuclear Information System (INIS)

Goll, W.; Hoffmann, P.B.; Hellwig, C.; Sauser, W.; Spino, J.; Walker, C.T.

2007-01-01

Irradiation experience gained on fuel rods with burn-ups greater than 60 MWd/kgHM irradiated in the Nuclear Power Plant Goesgen, Switzerland, is described. Emphasis is placed on the fuel behaviour, which has been analysed by hot cell examinations at the Institute for Transuranium Elements and the Paul-Scherrer-Institute. Above 60 MWd/kgHM, the so-called high burn-up structure (HBS) forms and the fission gas release increases with burn-up and rod power. Examinations performed in the outer region of the fuel revealed that most if not all of the fission gas created was retained in the HBS, even at 25% porosity. Furthermore, the HBS has a relatively low swelling rate, greatly increased plasticity, and its thermal conductivity is higher than expected from the porosity. The post-irradiation examinations showed that the HBS has no detrimental effects on the performance of stationary irradiated PWR fuel irradiated to the high burn-ups that can be achieved with 5 wt% U-235 enrichment. On the contrary, the HBS results in fuel performance that is generally better than it would have been if the HBS had not formed. (orig.)

A one-dimensional transport code for the simulation of D-T burning tokamak plasma

International Nuclear Information System (INIS)

Tone, Tatsuzo; Maki, Koichi; Kasai, Masao; Nishida, Hidetsugu

1980-11-01

A one-dimensional transport code for D-T burning tokamak plasma has been developed, which simulates the spatial behavior of fuel ions(D, T), alpha particles, impurities, temperatures of ions and electrons, plasma current, neutrals, heating of alpha and injected beam particles. The basic transport equations are represented by one generalized equation so that the improvement of models and the addition of new equations may be easily made. A model of burn control using a variable toroidal field ripple is employed. This report describes in detail the simulation model, numerical method and the usage of the code. Some typical examples to which the code has been applied are presented. (author)

Plasma burn-through simulations using the DYON code and predictions for ITER

International Nuclear Information System (INIS)

Kim, Hyun-Tae; Sips, A C C; De Vries, P C

2013-01-01

This paper will discuss simulations of the full ionization process (i.e. plasma burn-through), fundamental to creating high temperature plasma. By means of an applied electric field, the gas is partially ionized by the electron avalanche process. In order for the electron temperature to increase, the remaining neutrals need to be fully ionized in the plasma burn-through phase, as radiation is the main contribution to the electron power loss. The radiated power loss can be significantly affected by impurities resulting from interaction with the plasma facing components. The DYON code is a plasma burn-through simulator developed at Joint European Torus (JET) (Kim et al and EFDA-JET Contributors 2012 Nucl. Fusion 52 103016, Kim, Sips and EFDA-JET Contributors 2013 Nucl. Fusion 53 083024). The dynamic evolution of the plasma temperature and plasma densities including the impurity content is calculated in a self-consistent way using plasma wall interaction models. The recent installation of a beryllium wall at JET enabled validation of the plasma burn-through model in the presence of new, metallic plasma facing components. The simulation results of the plasma burn-through phase show a consistent good agreement against experiments at JET, and explain differences observed during plasma initiation with the old carbon plasma facing components. In the International Thermonuclear Experimental Reactor (ITER), the allowable toroidal electric field is restricted to 0.35 (V m −1 ), which is significantly lower compared to the typical value (∼1 (V m −1 )) used in the present devices. The limitation on toroidal electric field also reduces the range of other operation parameters during plasma formation in ITER. Thus, predictive simulations of plasma burn-through in ITER using validated model is of crucial importance. This paper provides an overview of the DYON code and the validation, together with new predictive simulations for ITER using the DYON code. (paper)

Development of a parallel processing couple for calculations of control rod worth in terms of burn-up in a WWER-1000 reactor

Energy Technology Data Exchange (ETDEWEB)

Noori-Kalkhoran, Omid; Ahangari, R. [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Reactor Research school; Shirani, A.S. [Shahid Beheshti Univ., Tehran (Iran, Islamic Republic of). Faculty of Engineering

2017-03-15

In this study a code based method has been developed for calculation of integral and differential control rod worth in terms of burn-up for a WWER-1000 reactor. Parallel processing of WIMSD-5B, PARCS V2.7 and COBRA-EN has been used for this purpose. WIMSD-5B has been used for cell calculation and handling burn-up of core at different days. PARCS V2.7?has been used for neutronic calculation of core and critical boron concentration search. Thermal-hydraulic calculation has been performed by COBRA-EN. A Parallel processing algorithm has been developed by MATLAB to couple and transfer suitable data between these codes in each step. Steady-State Power Picking Factors (PPFs) of the core and Control rod worth have been calculated from Beginning Of Cycle (BOC) to 289.7 Effective full Power Days (EFPDs) in some steps. Results have been compared with Bushehr Nuclear Power Plant (BNPP) Final Safety Analysis Report (FSAR) results. The results show great similarity and confirm the ability of developed coupling in calculation of control rod worth in terms of burn-up.

Determination of nuclear fuel burn-up using mass spectrometric techniques

International Nuclear Information System (INIS)

Saha, B.; Bagyalakshmi, R.; Periaswami, G.; Kavimandan, V.D.; Chitambar, S.A.; Jain, H.C.; Mathews, C.K.

1977-01-01

Determination of burn-up using a stable fission product monitor such as 148 Nd and heavy elements, determined by isotope dilution mass spectrometry gives the most accurate data. This report describes the work carried out to standardise the conditions for burn-up determination. Some typical results are given. (author)

Current applications of actinide-only burn-up credit within the Cogema group and R and D programme to take fission products into account

International Nuclear Information System (INIS)

Toubon, H.; Guillou, E.; Cousinou, P.; Barbry, F.; Grouiller, J.P.; Bignan, G.

2001-01-01

Burn-up credit can be defined as making allowance for absorbent radioactive isotopes in criticality studies, in order to optimise safety margins and avoid over-engineering of nuclear facilities. As far as the COGEMA Group is concerned, the three fields in which burn-up credit proves to be an advantage are the transport of spent fuel assemblies, their interim storage in spent fuel pools and reprocessing. In the case of transport, burn-up credit means that cask size do not need to be altered, despite an increase in the initial enrichment of the fuel assemblies. Burn-up credit also makes it possible to offer new cask designs with higher capacity. Burn-up credit means that fuel assemblies with a higher initial enrichment can be put into interim storage in existing facilities and opens the way to the possibility of more compact ones. As far as reprocessing is concerned, burn-up credit makes it possible to keep up current production rates, despite an increase in the initial enrichment of the fuel assemblies being reprocessed. In collaboration with the French Atomic Energy Commission and the Institute for Nuclear Safety and Protection, the COGEMA Group is participating in an extensive experimental programme and working to qualify criticality and fuel depletion computer codes. The research programme currently underway should mean that by 2003, allowance will be made for fission products in criticality safety analysis

Current applications of actinide-only burn-up credit within the Cogema group and R and D programme to take fission products into account

Energy Technology Data Exchange (ETDEWEB)

Toubon, H. [Cogema, 78 - Saint Quentin en Yvelines (France); Guillou, E. [Cogema Etablissement de la Hague, D/SQ/SMT, 50 - Beaumont Hague (France); Cousinou, P. [CEA Fontenay aux Roses, Inst. de Protection et de Surete Nucleaire, 92 (France); Barbry, F. [CEA Valduc, Inst. de Protection et de Surete Nucleaire, 21 - Is sur Tille (France); Grouiller, J.P.; Bignan, G. [CEA Cadarache, 13 - Saint Paul lez Durance (France)

2001-07-01

Burn-up credit can be defined as making allowance for absorbent radioactive isotopes in criticality studies, in order to optimise safety margins and avoid over-engineering of nuclear facilities. As far as the COGEMA Group is concerned, the three fields in which burn-up credit proves to be an advantage are the transport of spent fuel assemblies, their interim storage in spent fuel pools and reprocessing. In the case of transport, burn-up credit means that cask size do not need to be altered, despite an increase in the initial enrichment of the fuel assemblies. Burn-up credit also makes it possible to offer new cask designs with higher capacity. Burn-up credit means that fuel assemblies with a higher initial enrichment can be put into interim storage in existing facilities and opens the way to the possibility of more compact ones. As far as reprocessing is concerned, burn-up credit makes it possible to keep up current production rates, despite an increase in the initial enrichment of the fuel assemblies being reprocessed. In collaboration with the French Atomic Energy Commission and the Institute for Nuclear Safety and Protection, the COGEMA Group is participating in an extensive experimental programme and working to qualify criticality and fuel depletion computer codes. The research programme currently underway should mean that by 2003, allowance will be made for fission products in criticality safety analysis.

Zone-plate coded imaging of thermonuclear burn

International Nuclear Information System (INIS)

Ceglio, N.M.

1978-01-01

The first high-resolution, direct images of the region of thermonuclear burn in laser fusion experiments have been produced using a novel, two-step imaging technique called zone-plate coded imaging. This technique is extremely versatile and well suited for the microscopy of laser fusion targets. It has a tomographic capability, which provides three-dimensional images of the source distribution. It is equally useful for imaging x-ray and particle emissions. Since this technique is much more sensitive than competing imaging techniques, it permits us to investigate low-intensity sources

Criterion for burn-up conditions in gas-cooled cryogenic current leads

International Nuclear Information System (INIS)

Bejan, A.; Cluss, E.M. Jr.

1976-01-01

Superconducting magnets are energized through helium vapour-cooled cryogenic current leads operating at high ratios of current to mass flow. The high current operation where lead temperature, runaway, and eventual burn-up are likely to occur is investigated. A simple criterion for estimating the burn-up operation conditions (current, mass flow) for a given lead geometry (cross-sectional area, length, heat exchanger area) is presented. This article stresses the role played by the available heat exchanger area in avoiding burn-up at high ratios of current to mass flow. (author)

MISER-I: a computer code for JOYO fuel management

International Nuclear Information System (INIS)

Yamashita, Yoshioki

1976-06-01

A computer code ''MISER-I'' is for a nuclear fuel management of Japan Experimental Fast Breeder Reactor JOYO. The nuclear fuel management in JOYO can be regarded as a fuel assembly management because a handling unit of fuel in JOYO plant is a fuel subassembly (core and blanket subassembly), and so the recording of material balance in computer code is made with each subassembly. The input information into computer code is given with each subassembly for a transfer operation, or with one reactor cycle and every one month for a burn-up in reactor core. The output information of MISER-I code is the fuel assembly storage record, fuel storage weight record in each material balance subarea at any specified day, and fuel subassembly transfer history record. Change of nuclear fuel composition and weight due to a burn-up is calculated with JOYO-Monitoring Code by off-line computation system. MISER-I code is written in FORTRAN-IV language for FACOM 230-48 computer. (auth.)

Calculation using MVP and MVP-BURN in JRR-3

International Nuclear Information System (INIS)

Komeda, Masao; Kato, Tomoaki; Murayama, Yoji; Yamashita, Kiyonobu

2007-01-01

MVP is the particle-transport Monte Carlo code that has been developed in JAEA. MVP-BURN is an added function to do burn-up calculation. It is easy to built complex structure like core for MVP. And it is easy to do calculations of keff, any reaction rate, flux, burn-up and so on. In this report, it is introduced MVP and MVP-BURN. And some sample calculations of JRR-3 are shown. (author)

Nuclear fuel burn-up economy; Ekonomija izgaranja nuklearnog goriva

Energy Technology Data Exchange (ETDEWEB)

Matausek, M [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1984-07-01

In the period 1981-1985, for the needs of Utility Organization, Beograd, and with the support of the Scientific Council of SR Srbija, work has been performed on the study entitled 'Nuclear Fuel Burn-up Economy'. The forst [phase, completed during the year 1983 comprised: comparative analysis of commercial NPP from the standpoint of nuclear fuel requirements; development of methods for fuel burn-up analysis; specification of elements concerning the nuclear fuel for the tender documentation. The present paper gives the short description of the purpose, content and results achieved in the up-to-now work on the study. (author)

Managing the fusion burn to improve symbiotic system performance

International Nuclear Information System (INIS)

Renier, J.P.; Martin, J.G.

1979-01-01

Symbiotic power systems, in which fissile fuel is produced in fusion-powered factories and burned in thermal reactors characterized by high conversion ratios, constitute an interesting near-term fusion application. It is shown that the economic feasibility of such systems depend on adroit management of the fusion burn. The economics of symbiotes is complex: reprocessing and fabrication of the fusion reactor blankets are important components of the production cost of fissile fuel, but burning fissile material in the breeder blanket raises overall costs and lowers the support ratio. Analyses of factories which assume that the fusion power is constant during an irradiation cycle underestimate their potential. To illustrate the effect of adroit engineering of the fusion burn, this paper analyzes systems based on D-T and semi-catalyzed D-D fusion-powered U-233 breeders. To make the D-T symbiote self-sufficient, tritium is bred in separate lithium blankets designed so as to minimize overall costs. All blankets are assumed to have spherical geometry, with 85% closure. Neutronics depletion calculations were performed with a revised version of the discrete ordinates code XSDRN-PM, using multigroup (100 neutron, 21 gamma-ray groups) coupled cross-section libraries

Fast-ion diffusion measurements from radial triton burn up studies

International Nuclear Information System (INIS)

McCauley, J.S.; Budny, R.; McCune, D.; Strachan, J.D.

1993-08-01

A fast-ion diffusion coefficient of 0.1 ± 0.1 m 2 s -1 has been deduced from the triton burnup neutron emission profile measured by a collimated array of helium-4 spectrometers. The experiment was performed with high-power deuterium discharges produced by Princeton University's Tokamak Fusion Test Reactor (TFTR). The fast ions monitored were the 1.0 MeV tritons produced from the d(d,t)p. These tritons ''burn up'' with deuterons and emit a 14 MeV neutron by the d(t,α)n reaction. The ratio of the measured to calculated DT yield is typically 70%. The measured DT profile width is comparable to that predicted by the TRANSP transport code during neutral beam heating and narrower after the beam heating ended

The build-up and characterization of nuclear burn-up wave in a fast ...

Indian Academy of Sciences (India)

K V Anoop

2018-02-07

Feb 7, 2018 ... evaluating the quality of the wave by the researchers working in the field of nuclear burn-up wave build-up and propagation. Keywords. ... However, there are concerns relating to the nuclear safety, ... Simulation studies have.

Fuel cycles with high fuel burn-up: analysis of reactivity coefficients

International Nuclear Information System (INIS)

Kryuchkov, E.F.; Shmelev, A.N.; Ternovykh, M.J.; Tikhomirov, G.V.; Jinhong, L.; Saito, M.

2003-01-01

Fuel cycles of light-water reactors (LWR) with high fuel burn-up (above 100 MWd/kg), as a rule, involve large amounts of fissionable materials. It leads to forming the neutron spectrum harder than that in traditional LWR. Change of neutron spectrum and significant amount of non-traditional isotopes (for example, 237 Np, 238 Pu, 231 Pa, 232 U) in such fuel compositions can alter substantially reactivity coefficients as compared with traditional uranium-based fuel. The present work addresses the fuel cycles with high fuel burn-up which are based on Th-Pa-U and U-Np-Pu fuel compositions. Numerical analyses are carried out to determine effective neutron multiplication factor and void reactivity coefficient (VRC) for different values of fuel burn-up and different lattice parameters. The algorithm is proposed for analysis of isotopes contribution to these coefficients. Various ways are considered to upgrade safety of nuclear fuel cycles with high fuel burn-up. So, the results obtained in this study have demonstrated that: -1) Non-traditional fuel compositions developed for achievement of high fuel burn-up in LWR can possess positive values of reactivity coefficients that is unacceptable from the reactor operation safety point of view; -2) The lattice pitch of traditional LWR is not optimal for non-traditional fuel compositions, the increased value of the lattice pitch leads to larger value of initial reactivity margin and provides negative VRC within sufficiently broad range of coolant density; -3) Fuel burn-up has an insignificant effect on VRC dependence on coolant density, so, the measures undertaken to suppress positive VRC of fresh fuel will be effective for partially burnt-up fuel compositions also and; -4) Increase of LWR core height and introduction of additional moderators into the fuel lattice can be used as the ways to reach negative VRC values for full range of possible coolant density variations

High explosive programmed burn in the FLAG code

Energy Technology Data Exchange (ETDEWEB)

Mandell, D.; Burton, D.; Lund, C.

1998-02-01

The models used to calculate the programmed burn high-explosive lighting times for two- and three-dimensions in the FLAG code are described. FLAG uses an unstructured polyhedra grid. The calculations were compared to exact solutions for a square in two dimensions and for a cube in three dimensions. The maximum error was 3.95 percent in two dimensions and 4.84 percent in three dimensions. The high explosive lighting time model described has the advantage that only one cell at a time needs to be considered.

Burn-up calculation of fusion-fission hybrid reactor using thorium cycle

International Nuclear Information System (INIS)

Shido, S.; Matsunaka, M.; Kondo, K.; Murata, I.; Yamamoto, Y.

2006-01-01

A burn-up calculation system has been developed to estimate performance of blanket in a fusion-fission hybrid reactor which is a fusion reactor with a blanket region containing nuclear fuel. In this system, neutron flux is calculated by MCNP4B and then burn-up calculation is performed by ORIGEN2. The cross-section library for ORIGEN2 is made from the calculated neutron flux and evaluated nuclear data. The 3-dimensional ITER model was used as a base fusion reactor. The nuclear fuel (reprocessed plutonium as the fission materials mixed with thorium as the fertile materials), transmutation materials (minor actinides and long-lived fission products) and tritium breeder were loaded into the blanket. Performances of gas-cooled and water-cooled blankets were compared with each other. As a result, the proposed reactor can meet the requirement for TBP and power density. As far as nuclear waste incineration is concerned, the gas-cooled blanket has advantages. On the other hand, the water cooled-blanket is suited to energy production. (author)

Surviving "Payment by Results": a simple method of improving clinical coding in burn specialised services in the United Kingdom.

Science.gov (United States)

Wallis, Katy L; Malic, Claudia C; Littlewood, Sonia L; Judkins, Keith; Phipps, Alan R

2009-03-01

Coding inpatient episodes plays an important role in determining the financial remuneration of a clinical service. Insufficient or incomplete data may have very significant consequences on its viability. We created a document that improves the coding process in our Burns Centre. At Yorkshire Regional Burns Centre an inpatient summary sheet was designed to prospectively record and present essential information on a daily basis, for use in the coding process. The level of care was also recorded. A 3-month audit was conducted to assess the efficacy of the new forms. Forty-nine patients were admitted to the Burns Centre with a mean age of 27.6 years and TBSA ranging from 0.5% to 65%. The total stay in the Burns Centre was 758 days, of which 22% were at level B3-B5 and 39% at level B2. The use of the new discharge document identified potential income of about 500,000 GB pound sterling at our local daily tariffs for high dependency and intensive care. The new form is able to ensure a high quality of coding with a possible direct impact on the financial resources accrued for burn care.

Full MOX high burn-up PWR

Energy Technology Data Exchange (ETDEWEB)

Okubo, Tsutomu; Kugo, Teruhiko; Shimada, Shoichiro; Araya, Fumimasa; Ochiai, Masaaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1999-12-01

As a part of conceptual investigation on advanced light water reactors for the future, a light water reactor with the high burn-up of 100 GWd/t, the long cycle operation of 3 years and the full MOX core is being studied, aiming at the improvement on economical aspects, the reduction of the spent fuel production, the utilization of Plutonium and so forth. The present report summarizes investigation on PWR-type reactors. The core with the increased moderation of the moderator-to-fuel volume ratio of 2.6 {approx} 3.0 has been proposed be such a core that accomplishes requirements mentioned above. Through the neutronic and the thermo-hydrodynamic evaluation, the performances of the core have been evaluated. Also, the safety designing is underway considering the reactor system with the passive safety features. (author)

Improved Algorithms Speed It Up for Codes

International Nuclear Information System (INIS)

Hazi, A

2005-01-01

Huge computers, huge codes, complex problems to solve. The longer it takes to run a code, the more it costs. One way to speed things up and save time and money is through hardware improvements--faster processors, different system designs, bigger computers. But another side of supercomputing can reap savings in time and speed: software improvements to make codes--particularly the mathematical algorithms that form them--run faster and more efficiently. Speed up math? Is that really possible? According to Livermore physicist Eugene Brooks, the answer is a resounding yes. ''Sure, you get great speed-ups by improving hardware,'' says Brooks, the deputy leader for Computational Physics in N Division, which is part of Livermore's Physics and Advanced Technologies (PAT) Directorate. ''But the real bonus comes on the software side, where improvements in software can lead to orders of magnitude improvement in run times.'' Brooks knows whereof he speaks. Working with Laboratory physicist Abraham Szoeke and others, he has been instrumental in devising ways to shrink the running time of what has, historically, been a tough computational nut to crack: radiation transport codes based on the statistical or Monte Carlo method of calculation. And Brooks is not the only one. Others around the Laboratory, including physicists Andrew Williamson, Randolph Hood, and Jeff Grossman, have come up with innovative ways to speed up Monte Carlo calculations using pure mathematics

Fuel cycles with high fuel burn-up: analysis of reactivity coefficients

Energy Technology Data Exchange (ETDEWEB)

Kryuchkov, E.F.; Shmelev, A.N.; Ternovykh, M.J.; Tikhomirov, G.V.; Jinhong, L. [Moscow Engineering Physics Institute (State University) (Russian Federation); Saito, M. [Tokyo Institute of Technology (Japan)

2003-07-01

Fuel cycles of light-water reactors (LWR) with high fuel burn-up (above 100 MWd/kg), as a rule, involve large amounts of fissionable materials. It leads to forming the neutron spectrum harder than that in traditional LWR. Change of neutron spectrum and significant amount of non-traditional isotopes (for example, {sup 237}Np, {sup 238}Pu, {sup 231}Pa, {sup 232}U) in such fuel compositions can alter substantially reactivity coefficients as compared with traditional uranium-based fuel. The present work addresses the fuel cycles with high fuel burn-up which are based on Th-Pa-U and U-Np-Pu fuel compositions. Numerical analyses are carried out to determine effective neutron multiplication factor and void reactivity coefficient (VRC) for different values of fuel burn-up and different lattice parameters. The algorithm is proposed for analysis of isotopes contribution to these coefficients. Various ways are considered to upgrade safety of nuclear fuel cycles with high fuel burn-up. So, the results obtained in this study have demonstrated that: -1) Non-traditional fuel compositions developed for achievement of high fuel burn-up in LWR can possess positive values of reactivity coefficients that is unacceptable from the reactor operation safety point of view; -2) The lattice pitch of traditional LWR is not optimal for non-traditional fuel compositions, the increased value of the lattice pitch leads to larger value of initial reactivity margin and provides negative VRC within sufficiently broad range of coolant density; -3) Fuel burn-up has an insignificant effect on VRC dependence on coolant density, so, the measures undertaken to suppress positive VRC of fresh fuel will be effective for partially burnt-up fuel compositions also and; -4) Increase of LWR core height and introduction of additional moderators into the fuel lattice can be used as the ways to reach negative VRC values for full range of possible coolant density variations.

SRAC: JAERI thermal reactor standard code system for reactor design and analysis

International Nuclear Information System (INIS)

Tsuchihashi, Keichiro; Takano, Hideki; Horikami, Kunihiko; Ishiguro, Yukio; Kaneko, Kunio; Hara, Toshiharu.

1983-01-01

The SRAC (Standard Reactor Analysis Code) is a code system for nuclear reactor analysis and design. It is composed of neutron cross section libraries and auxiliary processing codes, neutron spectrum routines, a variety of transport, 1-, 2- and 3-D diffusion routines, dynamic parameters and cell burn-up routines. By making the best use of the individual code function in the SRAC system, the user can select either the exact method for an accurate estimate of reactor characteristics or the economical method aiming at a shorter computer time, depending on the purpose of study. The user can select cell or core calculation; fixed source or eigenvalue problem; transport (collision probability or Sn) theory or diffusion theory. Moreover, smearing and collapsing of macroscopic cross sections are separately done by the user's selection. And a special attention is paid for double heterogeneity. Various techniques are employed to access the data storage and to optimize the internal data transfer. Benchmark calculations using the SRAC system have been made extensively for the Keff values of various types of critical assemblies (light water, heavy water and graphite moderated systems, and fast reactor systems). The calculated results show good prediction for the experimental Keff values. (author)

Effect of burn-up on the thermal conductivity of uranium dioxide up to 100.000 MWd t-1

International Nuclear Information System (INIS)

Ronchi, C.; Sheindlin, M.; Staicu, D.; Kinoshita, M.

2004-01-01

The thermal diffusivity and specific heat of reactor-irradiated UO 2 fuel have been measured. Starting from end-of-life conditions at various burn-ups, measurements under thermal annealing cycles were performed in order to investigate the recovery of the thermal conductivity as a function of temperature. The separate effects of soluble fission products, of fission gas frozen in dynamical solution and of radiation damage were determined. In this context, particular emphasis was given to the behaviour of samples displaying the high burn-up rim structure. Recovery stages could be thoroughly investigated in samples that were irradiated at low burn-ups and/or at high irradiation temperatures. Other samples, in particular those exhibiting the characteristic rim structure, disintegrated at temperatures slightly higher than the irradiation temperature. Finally, from a database of several thousand measurements, an accurate formula for the in-pile thermal conductivity of UO 2 up to 100 GWd t -1 was developed, taking into account all the relevant effects and structural changes induced by reactor burn-up

Model comparisons of the reactive burn model SURF in three ASC codes

Energy Technology Data Exchange (ETDEWEB)

Whitley, Von Howard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stalsberg, Krista Lynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Reichelt, Benjamin Lee [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Shipley, Sarah Jayne [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2018-01-12

A study of the SURF reactive burn model was performed in FLAG, PAGOSA and XRAGE. In this study, three different shock-to-detonation transition experiments were modeled in each code. All three codes produced similar model results for all the experiments modeled and at all resolutions. Buildup-to-detonation time, particle velocities and resolution dependence of the models was notably similar between the codes. Given the current PBX 9502 equations of state and SURF calibrations, each code is equally capable of predicting the correct detonation time and distance when impacted by a 1D impactor at pressures ranging from 10-16 GPa, as long as the resolution of the mesh is not too coarse.

Evaluation and Parameter Analysis of Burn up Calculations for the Assessment of Radioactive Waste - 13187

Energy Technology Data Exchange (ETDEWEB)

Fast, Ivan; Aksyutina, Yuliya; Tietze-Jaensch, Holger [Product Quality Control Office for Radioactive Waste (PKS) at the Institute of Energy and Climate Research, Nuclear Waste Management and Reactor Safety Research, IEK-6, Forschungszentrum Juelich (Germany)

2013-07-01

Burn up calculations facilitate a determination of the composition and nuclear inventory of spent nuclear fuel, if operational history is known. In case this information is not available, the total nuclear inventory can be determined by means of destructive or, even on industrial scale, nondestructive measurement methods. For non-destructive measurements however only a few easy-to-measure, so-called key nuclides, are determined due to their characteristic gamma lines or neutron emission. From these measured activities the fuel burn up and cooling time are derived to facilitate the numerical inventory determination of spent fuel elements. Most regulatory bodies require an independent assessment of nuclear waste properties and their documentation. Prominent part of this assessment is a consistency check of inventory declaration. The waste packages often contain wastes from different types of spent fuels of different history and information about the secondary reactor parameters may not be available. In this case the so-called characteristic fuel burn up and cooling time are determined. These values are obtained from a correlations involving key-nuclides with a certain bandwidth, thus with upper and lower limits. The bandwidth is strongly dependent on secondary reactor parameter such as initial enrichment, temperature and density of the fuel and moderator, hence the reactor type, fuel element geometry and plant operation history. The purpose of our investigation is to look into the scaling and correlation limitations, to define and verify the range of validity and to scrutinize the dependencies and propagation of uncertainties that affect the waste inventory declarations and their independent verification. This is accomplished by numerical assessment and simulation of waste production using well accepted codes SCALE 6.0 and 6.1 to simulate the cooling time and burn up of a spent fuel element. The simulations are benchmarked against spent fuel from the real reactor

Pediatric superficial scald burns--reassessment of our follow-up protocol.

Science.gov (United States)

Egro, Francesco M; O'Neill, Jennifer K; Briard, Robert; Cubison, Tania C S; Kay, Alan R; Estela, Catalina M; Burge, Timothy S

2010-01-01

The most common pediatric burn injury is a superficial scald. The current follow-up protocol for such burns includes review of the patient at 2 weeks postinjury and then 2 months later. The authors decided to review the protocol to assess the need for this second follow-up. A retrospective study reviewed the case notes of patients younger than 16 years at the time of their injury presenting with a scald over 5% TBSA. The progress of healing and scar development up to 5 years follow-up was assessed. This study showed that scalds healing within 2 weeks following injury rarely became hypertrophic. A prospective study was performed over a 10-month period. All children who suffered a superficial partial-thickness scald injury were included. At the 2-week appointment, the need for further follow-up was predicted. The accuracy of this prediction was assessed 2 months later. This study showed that an experienced member of the burns team could reliably predict at 2-week appointment those children who could be safely discharged with no subsequent need for scar management. This study suggests that it will be safe to modify the follow-up protocol, reducing the number of clinic attendances.

Numerical solution of matrix exponential in burn-up equation using mini-max polynomial approximation

International Nuclear Information System (INIS)

Kawamoto, Yosuke; Chiba, Go; Tsuji, Masashi; Narabayashi, Tadashi

2015-01-01

Highlights: • We propose a new numerical solution of matrix exponential in burn-up depletion calculations. • The depletion calculation with extremely short half-lived nuclides can be done numerically stable with this method. • The computational time is shorter than the other conventional methods. - Abstract: Nuclear fuel burn-up depletion calculations are essential to compute the nuclear fuel composition transition. In the burn-up calculations, the matrix exponential method has been widely used. In the present paper, we propose a new numerical solution of the matrix exponential, a Mini-Max Polynomial Approximation (MMPA) method. This method is numerically stable for burn-up matrices with extremely short half-lived nuclides as the Chebyshev Rational Approximation Method (CRAM), and it has several advantages over CRAM. We also propose a multi-step calculation, a computational time reduction scheme of the MMPA method, which can perform simultaneously burn-up calculations with several time periods. The applicability of these methods has been theoretically and numerically proved for general burn-up matrices. The numerical verification has been performed, and it has been shown that these methods have high precision equivalent to CRAM

Systemization of burnup sensitivity analysis code

International Nuclear Information System (INIS)

Tatsumi, Masahiro; Hyoudou, Hideaki

2004-02-01

To practical use of fact reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoints of improvements on plant efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by development of adjusted nuclear library using the cross-section adjustment method, in which the results of critical experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor core 'JOYO'. The analysis of burnup characteristics is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, development of a analysis code for burnup sensitivity, SAGEP-BURN, has been done and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to user due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functionalities in the existing large system. It is not sufficient to unify each computational component for some reasons; computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion. For this

Satisfaction with life after burn: A Burn Model System National Database Study.

Science.gov (United States)

Goverman, J; Mathews, K; Nadler, D; Henderson, E; McMullen, K; Herndon, D; Meyer, W; Fauerbach, J A; Wiechman, S; Carrougher, G; Ryan, C M; Schneider, J C

2016-08-01

While mortality rates after burn are low, physical and psychosocial impairments are common. Clinical research is focusing on reducing morbidity and optimizing quality of life. This study examines self-reported Satisfaction With Life Scale scores in a longitudinal, multicenter cohort of survivors of major burns. Risk factors associated with Satisfaction With Life Scale scores are identified. Data from the National Institute on Disability, Independent Living, and Rehabilitation Research (NIDILRR) Burn Model System (BMS) database for burn survivors greater than 9 years of age, from 1994 to 2014, were analyzed. Demographic and medical data were collected on each subject. The primary outcome measures were the individual items and total Satisfaction With Life Scale (SWLS) scores at time of hospital discharge (pre-burn recall period) and 6, 12, and 24 months after burn. The SWLS is a validated 5-item instrument with items rated on a 1-7 Likert scale. The differences in scores over time were determined and scores for burn survivors were also compared to a non-burn, healthy population. Step-wise regression analysis was performed to determine predictors of SWLS scores at different time intervals. The SWLS was completed at time of discharge (1129 patients), 6 months after burn (1231 patients), 12 months after burn (1123 patients), and 24 months after burn (959 patients). There were no statistically significant differences between these groups in terms of medical or injury demographics. The majority of the population was Caucasian (62.9%) and male (72.6%), with a mean TBSA burned of 22.3%. Mean total SWLS scores for burn survivors were unchanged and significantly below that of a non-burn population at all examined time points after burn. Although the mean SWLS score was unchanged over time, a large number of subjects demonstrated improvement or decrement of at least one SWLS category. Gender, TBSA burned, LOS, and school status were associated with SWLS scores at 6 months

Reactivity management and burn-up management on JRR-3 silicide-fuel-core

International Nuclear Information System (INIS)

Kato, Tomoaki; Araki, Masaaki; Izumo, Hironobu; Kinase, Masami; Torii, Yoshiya; Murayama, Yoji

2007-08-01

On the conversion from uranium-aluminum-dispersion-type fuel (aluminide fuel) to uranium-silicon-aluminum-dispersion-type fuel (silicide fuel), uranium density was increased from 2.2 to 4.8 g/cm 3 with keeping uranium-235 enrichment of 20%. So, burnable absorbers (cadmium wire) were introduced for decreasing excess reactivity caused by the increasing of uranium density. The burnable absorbers influence reactivity during reactor operation. So, the burning of the burnable absorbers was studied and the influence on reactor operation was made cleared. Furthermore, necessary excess reactivity on beginning of operation cycle and the time limit for restart after unplanned reactor shutdown was calculated. On the conversion, limit of fuel burn-up was increased from 50% to 60%. And the fuel exchange procedure was changed from the six-batch dispersion procedure to the fuel burn-up management procedure. The previous estimation of fuel burn-up was required for the planning of fuel exchange, so that the estimation was carried out by means of past operation data. Finally, a new fuel exchange procedure was proposed for effective use of fuel elements. On the procedure, burn-up of spent fuel was defined for each loading position. The average length of fuel's staying in the core can be increased by two percent on the procedure. (author)

Neutronics analysis of Dalat Nuclear Research Reactor by MVP/GMVP code

International Nuclear Information System (INIS)

Nguyen Kien Cuong; Toru Obara

2008-01-01

The paper presents neutronics calculation for Dalat Nuclear Research Reactor (DNRR) to validate MVP/GMVP Code. Beside fresh core calculation, burnt core and burn up distribution were also carried out and compared with experimental data or result obtained from other codes. With complex geometry and operating history like DNRR, burn up calculation by Monte Carlo Method is the better choice owing to the use of exact geometry description and continuous neutron energy in calculation. The discrepancy between calculated data and experimental data is good to compare. By using Monte Carlo method, continuous neutron energy from JENDL3.3 library and combined with burn up calculation, MVP/GMVP Code is a very useful tool for reactor calculation. (author)

Establishing a PWR burn-up library

International Nuclear Information System (INIS)

Lutz, D.C.

1981-01-01

Starting out from data file ENDF/B IV /1/, a cross-section library has been established for the calculation of operating conditions in pressurized water reactors of the type used in BIBLIS B. The library includes macroscopic, homogenized 2-group cross-sections for all types of fuel elements used in this reactor, including those equipped with boron glass rods. For their calculation the previous irradiation of the fuel has been taken into consideration by approximation. Information on fuel consumption from cell burn-up calculations has been stored in a separate data file. It was designed as a base for the determination of cross sections to be used in the calculation of the incident ''main-steam pipe fracture''. For this library the description of cross sections as a function of the moderator status chose the water densities at 300 0 C/155 bar, 190 0 C/140 bar and 100 0 C/100 bar as fixed values. The burn-up library has been tested by a three-dimensional calculation for the 1sup(st) cycle of the BIBLIS B-reactor using program QUABOX /2/. This showed variances with the anticipated course concerning critically, which can be explained almost quantitatively by known deficiencies of the ENDF/b-IV library. (orig.) [de

Total surface area change of Uranium dioxide fuel in function of burn-up and its impact on fission gas release during neutron irradiation for small, intermediate and high burn-up

International Nuclear Information System (INIS)

Szuta, M.

2011-01-01

In the early published papers it was observed that the fractional fission gas release from the specimen have a tendency to increase with the total surface area of the specimen - a fairy linear relationship was indicated. Moreover it was observed that the increase of total surface area during irradiation occurs in the result of connection the closed porosity with the open porosity what in turn causes the increase of fission gas release. These observations let us surmise that the process of knock-out release is the most significant process of fission gas release since its quantity is proportional to the total surface area. Review of the experiments related to the increase of total surface area in function of burn-up is presented in the paper. For very high burn-up the process of grain sub-division (polygonization) occurs under condition that the temperature of irradiated fuel lies below the temperature of grain re-crystallization. Simultaneously with the process of polygonization, the increase in local porosity and the decrease in local density in function of burn-up occurs, which leads to the increase of total surface area. It is suggested that the same processes take place in the transformed fuel as in the original fuel, with the difference that the total surface area is so big that the whole fuel can be treated as that affected by the knock-out process. This leads to explanation of the experimental data that for very high burn-up (>120 MWd/kgU) the concentration of xenon is constant. An explanation of the grain subdivision process in function of burn-up in the 'athermal' rim region in terms of total surface area, initial grain size and knock-out release is undertaken. Correlation of the threshold burn-up, the local fission gas concentration, local total surface area, initial and local grain size and burn-up in the rim region is expected. (author)

Burn-up measurements of spent fuel using gamma spectrometry technique

International Nuclear Information System (INIS)

Pereda, C.; Henriquez, C.; Klein, J.; Medel, J.

2005-01-01

Burn-up results obtained for HEU (45% of 235 U) fuel assemblies of the RECH-1 Research Reactor using gamma spectrometry technique are presented. The spectra were got from an in-pool facility built in the reactor to be mainly used to measure the burnup of irradiated fuel assemblies with short cooling time, where 95 Zr is being evaluated as possible fission monitor. A program to measure all spent fuel assemblies of the RECH-1 reactor was initiated in the frame of the Regional Project RLA/4/018: 'Management of Spent Fuel from Research Reactors'. The results presented here were obtained from HEU spent fuel assemblies with cooling time greater than 100 days and 137 Cs was used as fission monitor. The efficiency of the in-pool system was determined using a slightly burnt experimental fuel assembly, which has one fuel plate (one of the outer plates) and the rest are dummy plates. An average burn-up of 2.8% of 235 U was previously measured for the experimental fuel assembly utilizing a facility installed in a hot cell and 137 Cs was used as monitor. (author)

Verification to the RSG-GAS fuel discharge burn-up using SRAC2006 module of COREBN/HIST

International Nuclear Information System (INIS)

J-Susilo; T-M-Sembiring; G-R-Sunaryo; M-Imron

2018-01-01

For 30 years operation, some of the modifications to the RSG GAS core has been done, that are changes included the type of fuel from U 3 O 8 -Al to U 3 Si 2 -Al with the same density 2.96 gU/cc, the loading pattern of standard fuel elements/fuel control elements from 6/1 & 6/2 to 5/1 pattern, and in core fuel management calculation tool has been change from IAFUEL to BATAN-FUEL. To obtain an extension of the operating license for the next 10 years, the RSG-GAS Periodic Safety Assessment Document is need to prepared. According to the Regulatory Body Chairman Regulation No. 2 2015, RSG-GAS safety assessment should be done independently. As part of this assessment the fuel discharge burn-up must be estimated. In this research, to ensure that the misposition of fuel element in the core has not occurred, the investigation to the document operating report related the fuel placement has been done. Therefore, by using 78 th to 93 rd operation data, verify of the fuel discharge burn-up of the RSG-GAS has been performed by using SRAC2006 module of COREBN/HIST. In addition, the results of these calculations are also made comparative with the operating report data that is calculated by using BATAN-FUEL. Maximum fuel discharge burn-up (57.73 % of U-235) was verified still under permissible value determined by the regulatory body (<60 % of U-235). Maximum differences value between two computer codes was about 2.12 % of U-235 (3.80 %) that is fuel at the B-7 position. Fuel discharge burn-up of RSG-GAS showed almost the same value for each the operation cycle, range of 1.52 % of U-235. So it can be concluded that the RSG-GAS core operation over the last ten years was in good fuel management performance, in accordance with the design. BATAN-FUEL has been conformed well enough with COREBN/HIST. (author)

Burn-Up Determination by High Resolution Gamma Spectrometry: Fission Product Migration Studies

Energy Technology Data Exchange (ETDEWEB)

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

1967-04-15

The migration of solid fission products, in particular caesium and ruthenium, in high temperature oxide fuel can create a severe problem during the application of non-destructive burn-up methods employing gamma spectrometry, since caesium-137 is otherwise the most convenient long-lived burn-up monitor and ruthenium-106 can be used to distinguish between fissions in U-235 and Pu-239. As part of an experimental programme to develop burn-up methods, gamma scanning experiments have been performed on slices of irradiated UO{sub 2} pellets using a lithium-drifted germanium detector. The usefulness of the technique for migration studies has been demonstrated by comparing the fission product distribution curves across the specimen diameters with the microstructure of the specimens after polishing and etching.

Study of nuclear fuel burn-up

International Nuclear Information System (INIS)

Pavelescu, M.; Borza, M.

1975-01-01

The authors approach theoretical treatment of isotopic composition changement for nuclear fuel in nuclear reactors. They show the difficulty of exhaustive treatment of burn-up problems and introduce the principal simplifying principles. Due to these principles they write and solve analytically the evolution equations of the concentration for the principal nuclides both in the case of fast and thermal reactors. Finally, they expose and comment the results obtained in the case of a power fast reactor. (author)

Deuterides of light elements: low-temperature thermonuclear burn-up and applications to thermonuclear fusion problems

International Nuclear Information System (INIS)

Frolov, A.M.; Smith, V.H.; Smith, G.T.

2002-01-01

Thermonuclear burn-up and thermonuclear applications are discussed for a number of deuterides and DT hydrides of light elements. These deuterides and corresponding DT hydrides are often used as thermonuclear fuels or components of such fuels. In fact, only for these substances thermonuclear energy gain exceeds (at some densities and temperatures) the bremsstrahlung loss and other high-temperature losses, i.e., thermonuclear burn-up is possible. Herein, thermonuclear burn-up in these deuterides and DT hydrides is considered in detail. In particular, a simple method is proposed to determine the critical values of the burn-up parameter x c for these substances and their mixtures at different temperatures and densities. The results for equimolar DT mixtures coincide quite well with the results of previous calculations. Also, the natural or Z limit is determined for low-temperature thermonuclear burn-up in the deuterides of light elements. (author)

Fuel management codes for fast reactors

International Nuclear Information System (INIS)

Sicard, B.; Coulon, P.; Mougniot, J.C.; Gouriou, A.; Pontier, M.; Skok, J.; Carnoy, M.; Martin, J.

The CAPHE code is used for managing and following up fuel subassemblies in the Phenix fast neutron reactor; the principal experimental results obtained since this reactor was commissioned are analyzed with this code. They are mainly concerned with following up fuel subassembly powers and core reactivity variations observed up to the beginning of the fifth Phenix working cycle (3/75). Characteristics of Phenix irradiated fuel subassemblies calculated by the CAPHE code are detailed as at April 1, 1975 (burn-up steel damage)

On-line extraction of the variance caused by burn-up in in-core three-dimensional power distribution

International Nuclear Information System (INIS)

Wang Yaqi; Luo Zhengpei; Li Fu; Liu Wenfeng

2001-01-01

In most of PWRs, the ex-core ion-chambers are the sole real-time sensors to respond to in-core power and its axial offset. However, the calibration coefficient of the ion-chambers depends on the (3D) power distribution and varies with the burn-up. People expect to know the variance in distribution caused by burn-up directly from the signals of ion-chambers. This expectation is not realized as yet, because an ion-chamber almost only responds to its nearest fuel assemblies. The authors then developed a two-step method for burn-up characteristic extraction: the harmonics synthesis method and harmonics' burn-up grouping. Using the extracted burn-up characteristics, the relationship between the readings of the ex-core ion-chambers and the in-core 3D power distribution is set up. Through the simulation on the heating reactor, the method of burn-up characteristic extraction is verified under engineering conditions. It is possible to on-line extract the variance caused by burn-up in 3D power distribution

HELIAS module development for systems codes

Energy Technology Data Exchange (ETDEWEB)

Warmer, F., E-mail: Felix.Warmer@ipp.mpg.de; Beidler, C.D.; Dinklage, A.; Egorov, K.; Feng, Y.; Geiger, J.; Schauer, F.; Turkin, Y.; Wolf, R.; Xanthopoulos, P.

2015-02-15

In order to study and design next-step fusion devices such as DEMO, comprehensive systems codes are commonly employed. In this work HELIAS-specific models are proposed which are designed to be compatible with systems codes. The subsequently developed models include: a geometry model based on Fourier coefficients which can represent the complex 3-D plasma shape, a basic island divertor model which assumes diffusive cross-field transport and high radiation at the X-point, and a coil model which combines scaling aspects based on the HELIAS 5-B reactor design in combination with analytic inductance and field calculations. In addition, stellarator-specific plasma transport is discussed. A strategy is proposed which employs a predictive confinement time scaling derived from 1-D neoclassical and 3-D turbulence simulations. This paper reports on the progress of the development of the stellarator-specific models while an implementation and verification study within an existing systems code will be presented in a separate work. This approach is investigated to ultimately allow one to conduct stellarator system studies, develop design points of HELIAS burning plasma devices, and to facilitate a direct comparison between tokamak and stellarator DEMO and power plant designs.

Calculation of heat rating and burn-up for test fuel pins irradiated in DR 3

International Nuclear Information System (INIS)

Bagger, C.; Carlsen, H.; Hansen, K.

1980-01-01

A summary of the DR 3 reactor and HP1 rig design is given followed by a detailed description of the calculation procedure for obtaining linear heat rating and burn-up values of fuel pins irradiated in HP1 rigs. The calculations are carried out rather detailed, especially regarding features like end pellet contribution to power as a function of burn-up, gamma heat contributions, and evaluation of local values of heat rating and burn-up. Included in the report is also a description of the fast flux- and cladding temperature calculation techniques currently used. A good agreement between measured and calculated local burn-up values is found. This gives confidence to the detailed treatment of the data. (author)

The burn-up credit physics and the 40. Minerve anniversary; La physique du credit Burn-Up et le 40. anniversaire de Minerve

Energy Technology Data Exchange (ETDEWEB)

Santamarina, A [CEA/Cadarache, Departement d' Etudes des Reacteurs, DER/SPRC, 13 - Saint-Paul-lez-Durance (France); Toubon, H [Cogema, 78 - Velizy Villacoublay (France); Trakas, C [FRAMATOME, 92 - Paris La Defense (France); and others

2000-03-21

The technical meeting organized by the SFEN on the burn-up credit (CBU) physics, took place the 23 november 1999 at Cadarache. the first presentation dealt with the economic interest and the neutronic problems of the CBU. Then two papers presented how taking into account the CBU in the industry in matter of transport, storage in pool, reprocessing and criticality calculation (MCNP4/Apollo2-F benchmark). An experimental method for the reactivity measurement through oscillations in the Minerve reactor, has been presented with an analysis of the possible errors. The future research program OSMOSE, taking into account the minor actinides in the CBU, was also developed. The last paper presented the national and international research programs in the CBU domain, in particular experiments realized in CEA/Valduc and the OECD Burn-up Criticality Benchmark Group activities. (A.L.B.)

ERANOS 2.0, Modular code and data system for fast reactor neutronics analyses

International Nuclear Information System (INIS)

2008-01-01

1 - Description of program or function: The European Reactor Analysis Optimized calculation System, ERANOS, has been developed and validated with the aim of providing a suitable basis for reliable neutronic calculations of current as well as advanced fast reactor cores. It consists of data libraries, deterministic codes and calculation procedures which have been developed within the European Collaboration on Fast Reactors over the past 20 years or so, in order to answer the needs of both industrial and R and D organisations. The whole system counts roughly 250 functions and 3000 subroutines totalling 450000 lines of FORTRAN-77 and ESOPE instructions. ERANOS is written using the ALOS software which requires only standard FORTRAN compilers and includes advanced programming features. A modular structure was adopted for easier evolution and incorporation of new functionalities. Blocks of data (SETs) can be created or used by the modules themselves or by the user via the LU control language. Programming, and dynamic memory allocation, are performed by means of the ESOPE language. External temporary storage and permanent storage capabilities are provided by the GEMAT and ARCHIVE functions, respectively. ESOPE, LU, GEMAT and ARCHIVE are all part of the ALOS software. This modular structure allows different modules to be linked together in procedures corresponding to recommended calculation routes ranging from fast-running and moderately-accurate 'routine' procedures to slow-running but highly-accurate 'reference' procedures. The main contents of the ERANOS-2.0 package are: nuclear data libraries (multigroup cross-sections from the JEF-2.2 evaluated nuclear data file, and other specific data files), a cell and lattice code (ECCO), reactor flux solvers (diffusion, Sn transport, nodal variational transport), a burn-up module, various processing modules (material and neutron balance, breeding gains,...), tools related to perturbation theory and sensitivity analysis, core

Oxygen stoichiometry shift of irradiated LWR-fuels at high burn-ups: Review of data and alternative interpretation of recently published results

International Nuclear Information System (INIS)

Spino, J.; Peerani, P.

2008-01-01

The available oxygen potential data of LWR-fuels by the EFM-method have been reviewed and compared with thermodynamic data of equivalent simulated fuels and mixed oxide systems, combined with the analysis of lattice parameter data. Up to burn-ups of 70-80 GWd/tM the comparison confirmed traditional predictions anticipating the fuels to remain quasi stoichiometric along irradiation. However, recent predictions of a fuel with average burn-up around 100 GWd/tM becoming definitely hypostoichiometric were not confirmed. At average burn-ups around 80 GWd/tM and above, it is shown that the fuels tend to acquire progressively slightly hyperstoichiometric O/M ratios. The maximum derived O/M ratio for an average burn-up of 100 GWd/tM lies around 2.001 and 2.002. Though slight, the stoichiometry shift may have a measurable accelerating impact on fission gas diffusion and release

Determination of nuclear fuel burn-up

International Nuclear Information System (INIS)

Kristak, J.; Vobecky, M.

1973-01-01

Samples containing a known content of 235 U were irradiated with several different neutron doses and activities were determined of radionuclides including 125 Sb, 144 Ce, 134 Cs, 154 Eu, 103 Ru, 95 Zr. The values thus obtained were divided by the 137 Cs activity value. The resulting neutron dose-dependent value is plotted into a calibration graph. The degree of nuclear fuel burn-up is obtained from the graph using an experimentally determined ratio of the activities of the above radionuclides. (B.S.)

High burn-up structure in nuclear fuel: impact on fuel behavior - 4005

International Nuclear Information System (INIS)

Noirot, J.; Pontillon, Y.; Zacharie-Aubrun, I.; Hanifi, K.; Bienvenu, P.; Lamontagne, J.; Desgranges, L.

2016-01-01

When UO 2 and (U,Pu)O 2 fuels locally reach high burn-up, a major change in the microstructure takes place. The initial grains are replaced by thousands of much smaller grains, fission gases form micrometric bubbles and metallic fission products form precipitates. This occurs typically at the rim of the pellets and in heterogeneous MOX fuel Pu rich agglomerates. The high burn-up at the rim of the pellets is due to a high capture of epithermal neutrons by 238 U leading locally to a higher concentration of fissile Pu than in the rest of the pellet. In the heterogeneous MOX fuels, this rim effect is also active, but most of the high burn-up structure (HBS) formation is linked to the high local concentration of fissile Pu in the Pu agglomerates. This Pu distribution leads to sharp borders between HBS and non-HBS areas. It has been shown that the size of the new grains, of the bubbles and of the precipitates increase with the irradiation local temperatures. Other parameters have been shown to have an influence on the HBS initiation threshold, such as the irradiation density rate, the fuel composition with an effect of the Pu presence, but also of the Gd concentration in poisoned fuels, some of the studied additives, like Cr, and, maybe some of the impurities. It has been shown by indirect and direct approaches that HBS formation is not the main contributor to the increase of fission gas release at high burn-up and that the HBS areas are not the main source of the released gases. The impact of HBS on the fuel behavior during ramp on high burn-up fuels is still unclear. This short paper is followed by the slides of the presentation

Investigating the burning characteristics of electric cables used in the nuclear power plant by way of 3-D transient FDS code

Energy Technology Data Exchange (ETDEWEB)

Ferng, Y.M., E-mail: ymferng@ess.nthu.edu.t [Department of Engineering and System Science, Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Sec. 2. Kuang-Fu Rd., Hsinchu 30013, Taiwan (China); Liu, C.H. [Department of Engineering and System Science, Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Sec. 2. Kuang-Fu Rd., Hsinchu 30013, Taiwan (China)

2011-01-15

Burning characteristics of electrical cables are one of the key parameters for the fire hazard assessment of nuclear power plants (NPPs) since the cables are the essential sources of fire in the plants. A three-dimensional (3-D) transient computational fluid dynamics (CFD) code{sub F}DS is adopted in this paper to simulate these characteristics related to the cable burning. Being one of the NRC licensing fire codes, the FDS includes the thermal-hydraulic equations, the turbulence model and the chemical combustion model, etc. In order to assess the CFD fire models used in this code, a burning test using the control cable with the outer jacket of polyvinylchloride (PVC) and the inner insulation of cross-linked polyethylene (XLPE) is conducted. The measured parameters associated with the burning characteristics include the heat release rate (HRR), O{sub 2} depletion, and CO and CO{sub 2} production, etc. Except the amount of O{sub 2} consumption, the predicted transient behaviors of other parameters can reproduce the measured data. Based on the chemical combustion model in the FDS code, this discrepancy may be essentially resulted from the default value of hydrogen fraction (H{sub frac}) contained in the soot since the soot yield for the burning of PVC material is high enough that the uncertainty in the H{sub frac} value has a prominent effect on the amount of O{sub 2} consumption. This explanation can be confirmed by a benchmark calculation for simulating a burning test with the polymethylmethacrylate (PMMA) fuel of low-soot yield. The present simulation works can provide the useful information for the plant staff or the researcher as they would perform the fire hazard analysis in the NPPs using the FDS code.

Development of Coolant Radioactivity Interpretation Code

International Nuclear Information System (INIS)

Kim, Kiyoung; Jung, Youngsuk; Kim, Kyounghyun; Kim, Jangwook

2013-01-01

In Korea, the coolant radioactivity analysis has been performed by using the computer codes of foreign companies such as CADE (Westinghouse), IODYNE and CESIUM (ABB-CE). However, these computer codes are too conservative and have involved considerable errors. Furthermore, since these codes are DOS-based program, their easy operability is not satisfactory. Therefore it is required development of an enhanced analysis algorithm applying an analytical method reflecting the change of operational environments of domestic nuclear power plants and a fuel failure evaluation software considering user' conveniences. We have developed a nuclear fuel failure evaluation code able to estimate the number of failed fuel rods and the burn-up of failed fuels during nuclear power plant operation cycle. A Coolant Radio-activity Interpretation Code (CRIC) for LWR has been developed as the output of the project 'Development of Fuel Reliability Enhanced Technique' organized by Korea Institute of Energy Technology Evaluation and Planning (KETEP). The CRIC is Windows based-software able to evaluate the number of failed fuel rods and the burn-up of failed fuel region by analyzing coolant radioactivity of LWR in operation. The CRIC is based on the model of fission products release commonly known as 'three region model' (pellet region, gap region, and coolant region), and we are verifying the CRIC results based on the cases of domestic fuel failures. CRIC users are able to estimate the number of failed fuel rods, burn-up and regions of failed fuel considered enrichment and power distribution of fuel region by using operational cycle data, coolant activity data, fuel loading pattern, Cs-134/Cs-137 ratio according to burn-up and U-235 enrichment provided in the code. Due to development of the CRIC, it is secured own unique fuel failure evaluation code. And, it is expected to have the following significant meaning. This is that the code reflecting a proprietary technique for quantitatively

Monte-Carlo code calculation of 3D reactor core model with usage of burnt fuel isotopic compositions, obtained by engineering codes

Energy Technology Data Exchange (ETDEWEB)

Aleshin, Sergey S.; Gorodkov, Sergey S.; Shcherenko, Anna I. [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation)

2016-09-15

A burn-up calculation of large systems by Monte-Carlo code (MCU) is complex process and it requires large computational costs. Previously prepared isotopic compositions are proposed to be used for the Monte-Carlo code calculations of different system states with burnt fuel. Isotopic compositions are calculated by an approximation method. The approximation method is based on usage of a spectral functionality and reference isotopic compositions, that are calculated by the engineering codes (TVS-M, BIPR-7A and PERMAK-A). The multiplication factors and power distributions of FAs from a 3-D reactor core are calculated in this work by the Monte-Carlo code MCU using earlier prepared isotopic compositions. The separate conditions of the burnt core are observed. The results of MCU calculations were compared with those that were obtained by engineering codes.

Development of high-strength aluminum alloys for basket in transport and storage cask for high burn-up spent fuel

International Nuclear Information System (INIS)

Maeguchi, T.; Sakaguchi, Y.; Kamiwaki, Y.; Ishii, M.; Yamamoto, T.

2004-01-01

Mitsubishi Heavy Industries, Ltd. (MHI) has developed high-strength borated aluminum alloys (high-strength B-Al alloys), suitable for application to baskets in transport and storage casks for high burn-up spent fuels. Aluminum is a suitable base material for the baskets due to its low density and high thermal conductivity. The aluminum basket would reduce weight of the cask, and effectively release heat generated by spent fuels. MHI had already developed borated aluminum alloys (high-toughness B-Al alloy), and registered them as ASME Code Case ''N-673''. However, there has been a strong demand for basket materials with higher strength in the case of MSF (Mitsubishi Spent Fuel) casks for high-burn up spent fuels, since the basket is required to stand up to higher stress at higher temperature. The high-strength basket material enables the design of a compact cask under a limitation of total size and weight. MHI has developed novel high-strength B-Al alloys which meet these requirements, based on a new manufacturing process. The outline of mechanical and metallurgical characteristics of the high-strength B-Al alloys is described in this paper

The Contribution of the Dyadic Parent-Child Interaction Coding System (DPICS) Warm-Up Segments in Assessing Parent-Child Interactions

Science.gov (United States)

Shanley, Jenelle R.; Niec, Larissa N.

2011-01-01

This study evaluated the inclusion of uncoded segments in the Dyadic Parent-Child Interaction Coding System, an analogue observation of parent-child interactions. The relationships between warm-up and coded segments were assessed, as well as the segments' associations with parent ratings of parent and child behaviors. Sixty-nine non-referred…

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

International Nuclear Information System (INIS)

Čerba, Štefan; Vrban, Branislav; Lüley, Jakub; Dařílek, Petr; Zajac, Radoslav; Nečas, Vladimír; Haščik, Ján

2014-01-01

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

LWR high burn-up operation and MOX introduction. Fuel cycle performance from the viewpoint of waste management

International Nuclear Information System (INIS)

Inagaki, Yaohiro; Iwasaki, Tomohiko; Niibori, Yuichi; Sato, Seichi; Ohe, Toshiaki; Kato, Kazuyuki; Torikai, Seishi; Nagasaki, Shinya; Kitayama, Kazumi

2009-01-01

From the viewpoint of waste management, a quantitative evaluation of LWR nuclear fuel cycle system performance was carried out, considering both higher burn-up operation of UO 2 fuel coupled with the introduction of MOX fuel. A major parameter to quantify this performance is the number of high-level waste (HLW) glass units generated per GWd (gigawatt-day based on reactor thermal power generation before electrical conversion). This parameter was evaluated for each system up to a maximum burn-up of 70GWd/THM (gigawatt-day per ton of heavy metal) assuming current conventional reprocessing and vitrification conditions where the waste loading of glass is restricted by the heat generation rate, the MoO 3 content, or the noble metal content. The results showed that higher burn-up operation has no significant influence on the number of glass units generated per GWd for UO 2 fuel, though the number of glass units per THM increases linearly with burn-up and is restricted by the heat generation rate. On the other hand, the introduction of MOX fuel causes the number of glass units per GWd to double owing to the increase in the heat generation rate. An extended cooling period of the spent fuel prior to reprocessing effectively reduces the heat generation rate for UO 2 fuel, while a separation of minor actinides (Np, Am, and Cm) from the high-level waste provides additional reduction for MOX fuel. However, neither of these leads to a substantial reduction in the number of glass units, since the MoO 3 content or the noble metal content restricts the number of glass units rather than the heat generation rate. These results suggest that both the MoO 3 content and the noble metal content provide the key to reducing the amount of waste glass that is generated, leading to an overall improvement in fuel cycle system performance. (author)

Fission gas release from UO2 pellet fuel at high burn-up

International Nuclear Information System (INIS)

Vitanza, C.; Kolstad, E.; Graziani, U.

1979-01-01

Analysis of in-reactor measurements of fuel center temperature and rod internal pressure at the OECD Halden Reactor Project has led to the development of an empirical fission gas release model, which is described. The model originally derived from data obtained in the low and intermediate burn-up range, appears to give good predictions for rods irradiated to high exposures as well. PIE puncturing data from seven fuel rods, operated at relatively constant powers and peak center temperatures between 1900 and 2000 0 C up to approx. 40,000 MWd/t UO 2 , did not exhibit any burn-up enhancement on the fission gas release rate

NACOM - a code for sodium spray fire analysis

International Nuclear Information System (INIS)

Rao, P.M.; Kannan, S.E.

2002-01-01

Full text: In liquid metal fast breeder reactors (LMFBR), leakage of sodium can result in a spray fire. Because of higher burning rates in droplet form combustion of sodium in spray fire, thermal consequences are more severe than that in a sodium pool fire. The code NACOM was developed for the analysis of sodium spray fires in LMFBRs facilities. The code uses the validated model for estimating the falling droplet burning rates in pre-ignition and vapour phase combustion stages. It uses a distribution system to generate the droplet groups of different diameters that represent the spray. The code requires about 20 input parameters like sodium leak rates, sodium temperature, initial cell conditions like oxygen concentration, temperature and dimensions. NACOM is a validated code based on experiments with sodium inventory up to 650 kg in 0 to 21 % O 2 atmospheres. The paper brings out the salient features of the code along with the sensitivity analysis of the main input parameters like spray volume mean diameter, oxygen concentration etc. based on the results obtained. The limitations of the code and the confidence margins applicable to results obtained are also brought out

Estimation of Emissions from Sugarcane Field Burning in Thailand Using Bottom-Up Country-Specific Activity Data

Directory of Open Access Journals (Sweden)

Wilaiwan Sornpoon

2014-09-01

Full Text Available Open burning in sugarcane fields is recognized as a major source of air pollution. However, the assessment of its emission intensity in many regions of the world still lacks information, especially regarding country-specific activity data including biomass fuel load and combustion factor. A site survey was conducted covering 13 sugarcane plantations subject to different farm management practices and climatic conditions. The results showed that pre-harvest and post-harvest burnings are the two main practices followed in Thailand. In 2012, the total production of sugarcane biomass fuel, i.e., dead, dry and fresh leaves, amounted to 10.15 million tonnes, which is equivalent to a fuel density of 0.79 kg∙m−2. The average combustion factor for the pre-harvest and post-harvest burning systems was determined to be 0.64 and 0.83, respectively. Emissions from sugarcane field burning were estimated using the bottom-up country-specific values from the site survey of this study and the results compared with those obtained using default values from the 2006 IPCC Guidelines. The comparison showed that the use of default values lead to underestimating the overall emissions by up to 30% as emissions from post-harvest burning are not accounted for, but it is the second most common practice followed in Thailand.

Status of the development of a fully integrated code system for the simulation of high temperature reactor cores

Energy Technology Data Exchange (ETDEWEB)

Kasselmann, Stefan, E-mail: s.kasselmann@fz-juelich.de [Institute of Energy and Climate Research, Nuclear Waste Management and Reactor Safety (IEK-6), Forschungszentrum Jülich GmbH, 52425 Jülich (Germany); Druska, Claudia [Institute of Energy and Climate Research, Nuclear Waste Management and Reactor Safety (IEK-6), Forschungszentrum Jülich GmbH, 52425 Jülich (Germany); Herber, Stefan [Institute of Energy and Climate Research, Nuclear Waste Management and Reactor Safety (IEK-6), Forschungszentrum Jülich GmbH, 52425 Jülich (Germany); Lehrstuhl für Reaktorsicherheit und -technik, RWTH Aachen, 52062 Aachen (Germany); Jühe, Stephan [Lehrstuhl für Reaktorsicherheit und -technik, RWTH Aachen, 52062 Aachen (Germany); Keller, Florian; Lambertz, Daniela; Li, Jingjing; Scholthaus, Sarah; Shi, Dunfu [Institute of Energy and Climate Research, Nuclear Waste Management and Reactor Safety (IEK-6), Forschungszentrum Jülich GmbH, 52425 Jülich (Germany); Xhonneux, Andre; Allelein, Hans-Josef [Institute of Energy and Climate Research, Nuclear Waste Management and Reactor Safety (IEK-6), Forschungszentrum Jülich GmbH, 52425 Jülich (Germany); Lehrstuhl für Reaktorsicherheit und -technik, RWTH Aachen, 52062 Aachen (Germany)

2014-05-01

The HTR code package (HCP) is a new code system, which couples a variety of stand-alone codes for the simulation of different aspects of HTR. HCP will allow the steady-state and transient operating conditions of a 3D reactor core to be simulated including new features such as spatially resolved fission product release calculations or production and transport of graphite dust. For this code the latest programming techniques and standards are applied. As a first step an object-oriented data model was developed which features a high level of readability because it is based on problem-specific data types like Nuclide, Reaction, ReactionHandler, CrossSectionSet, etc. Those classes help to encapsulate and therefore hide specific implementations, which are not relevant with respect to physics. HCP will make use of one consistent data library for which an automatic generation tool was developed. The new data library consists of decay information, cross sections, fission yields, scattering matrices etc. for all available nuclides (e.g. ENDF/B-VII.1). The data can be stored in different formats such as binary, ASCII or XML. The new burn up code TNT (Topological Nuclide Transmutation) applies graph theory to represent nuclide chains and to minimize the calculation effort when solving the burn up equations. New features are the use of energy-dependent fission yields or the calculation of thermal power for decay, fission and capture reactions. With STACY (source term analysis code system) the fission product release for steady state as well as accident scenarios can be simulated for each fuel batch. For a full-core release calculation several thousand fuel elements are tracked while passing through the core. This models the stochastic behavior of a pebble bed in a realistic manner. In this paper we report on the current status of the HCP and present first results, which prove the applicability of the selected approach.

Burn-up analysis of uranium silicide fuels 20% 235U, in the LFR facility

International Nuclear Information System (INIS)

Amor, Ricardo A.; Bouza, Edgardo; Cabrejas, Julian L.; Devida, Claudio A.; Gil, Daniel A.; Stankevicius, Alejandro; Gautier, Eduardo; Garavaglia, Ricardo N.; Lobo, Alfredo

2003-01-01

The LFR Facility is a laboratory designed and constructed with a Hot-Cells line, a Globe-Box and a Fume-Hood, all of them suited to work with radioactive materials such as samples of irradiated silicide MTR fuel elements. A series of dissolutions of this material was performed. From the resulting solutions, two fractions were separated by HPLC. One contained U + Pu, and other the fission product Nd. The concentrations of these elements were obtained by isotopic dilution and mass spectrometry (IDMS). It is concluded that this technique is very powerful and accurate when properly applied, and makes the validation of burn-up calculation codes possible. It is worth remarking the Lfr capacity to carry on different Research and Development (R + D) tasks in the Nuclear Fuel Cycle field. (author)

HTR core physics and transient analyses by the Panthermix code system

International Nuclear Information System (INIS)

Haas, J.B.M. de; Kuijper, J.C.; Oppe, J.

2005-01-01

At NRG Petten, core physics analyses on High Temperature gas-cooled Reactors (HTRs) are mainly performed by means of the PANTHERMIX code system. Since some years NRG is developing the HTR reactor physics code system WIMS/PANTHERMIX, based on the lattice code WIMS (Serco Assurance, UK), the 3-dimensional steady-state and transient core physics code PANTHER (British Energy, UK) and the 2-dimensional R-Z HTR thermal hydraulics code THERMIX-DIREKT (Research Centre FZJ Juelich, Germany). By means of the WIMS code nuclear data are being generated to suit the PANTHER code's neutronics. At NRG the PANTHER code has been interfaced with THERMIX-DIREKT to form PANTHERMIX, to enable core-follow/fuel management and transient analyses in a consistent manner on pebble bed type HTR systems. Also provisions have been made to simulate the flow of pebbles through the core of a pebble bed HTR, according to a given (R-Z) flow pattern. As examples of the versatility of the PANTHERMIX code system, calculations are presented on the PBMR, the South African pebble bed reactor design, to show the transient capabilities, and on a plutonium burning MEDUL-reactor, to demonstrate the core-follow/fuel management capabilities. For the investigated cases a good agreement is observed with the results of other HTR core physics codes

The burn-up credit physics and the 40. Minerve anniversary

International Nuclear Information System (INIS)

Santamarina, A.; Toubon, H.; Trakas, C.

2000-01-01

The technical meeting organized by the SFEN on the burn-up credit (CBU) physics, took place the 23 november 1999 at Cadarache. the first presentation dealt with the economic interest and the neutronic problems of the CBU. Then two papers presented how taking into account the CBU in the industry in matter of transport, storage in pool, reprocessing and criticality calculation (MCNP4/Apollo2-F benchmark). An experimental method for the reactivity measurement through oscillations in the Minerve reactor, has been presented with an analysis of the possible errors. The future research program OSMOSE, taking into account the minor actinides in the CBU, was also developed. The last paper presented the national and international research programs in the CBU domain, in particular experiments realized in CEA/Valduc and the OECD Burn-up Criticality Benchmark Group activities. (A.L.B.)

Recent improvements to TRIGLAV code

International Nuclear Information System (INIS)

Zagar, T.; Ravnik, M.; Persic, A.

1998-01-01

TRIGLAV code was developed for TRIGA research reactor calculations and is based on two-dimensional diffusion equation. The main purpose of the program is calculation of the fuel elements burn-up. Calculated core burn-up and excess reactivity results are compared with experimental values. New control rod model is introduced and tested in this paper. Calculated integral control rod worth and calculated integral reactivity curves are presented and compared with measured values. Comparison with measured fuel element worth values is presented as a test for two-dimensional flux distribution calculations.(author)

The estimation of the control rods absorber burn-up during the VVER-1000 operation

Energy Technology Data Exchange (ETDEWEB)

Bolshagin, Sergey N.; Gorodkov, Sergey S.; Sukhino-Khomenko, Evgeniya A. [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation)

2013-09-15

The isotopic composition of the control rods absorber changes under the neutron flux influence, so the control rods efficiency can decrease. In the VVER-1000 control rods boron carbide and dysprosium titanate are used as absorbing materials. In boric part the efficiency decreases due to the {sup 10}B isotope burn-up. Dysprosium isotopes turn into other absorbing isotopes, so the absorbing properties of dysprosium part decrease to a lesser degree. Also the control rod's shells may be deformed as a consequence of boron carbide radiation swelling. This fact should be considered in substantiation of control rods durability. For the estimation of the control rods absorber burn-up two models are developed: VVER-1000 3-D fuel assembly with control rods partially immersed (imitation of the control rods operation in the working group) and VVER-1000 3-D fuel assembly with control rods, located at the upper limit switch (imitation of the control rods operation in groups of the emergency shutdown system). (orig.)

MOSRA-SRAC. Lattice calculation module of the modular code system for nuclear reactor analyses MOSRA

International Nuclear Information System (INIS)

Okumura, Keisuke

2015-10-01

MOSRA-SRAC is a lattice calculation module of the Modular code System for nuclear Reactor Analyses (MOSRA). This module performs the neutron transport calculation for various types of fuel elements including existing light water reactors, research reactors, etc. based on the collision probability method with a set of the 200-group cross-sections generated from the Japanese Evaluated Nuclear Data Library JENDL-4.0. It has also a function of the isotope generation and depletion calculation for up to 234 nuclides in each fuel material in the lattice. In these ways, MOSRA-SRAC prepares the burn-up dependent effective microscopic and macroscopic cross-section data to be used in core calculations. A CD-ROM is attached as an appendix. (J.P.N.)

Determination of the burn-up of TRIGA fuel elements by calculation with new TRIGLAV program

International Nuclear Information System (INIS)

Zagar, T.; Ravnik, M.

1996-01-01

The results of fuel element burn-up calculations with new TRIGLAV program are presented. TRIGLAV program uses two dimensional model. Results of calculation are compared to results calculated with program, which uses one dimensional model. The results of fuel element burn-up measurements with reactivity method are presented and compared with the calculated results. (author)

Development of fast and accurate Monte Carlo code MVP

International Nuclear Information System (INIS)

Mori, Takamasa

2001-01-01

The development work of fast and accurate Monte Carlo code MVP has started at JAERI in late 80s. From the beginning, the code was designed to utilize vector supercomputers and achieved higher computation speed by a factor of 10 or more compared with conventional codes. In 1994, the first version of MVP was released together with cross section libraries based on JENDL-3.1 and JENDL-3.2. In 1996, minor revision was made by adding several functions such as treatments of ENDF-B6 file 6 data, time dependent problem, and so on. Since 1996, several works have been carried out for the next version of MVP. The main works are (1) the development of continuous energy Monte Carlo burn-up calculation code MVP-BURN, (2) the development of a system to generate cross section libraries at arbitrary temperature, and (3) the study on error estimations and their biases in Monte Carlo eigenvalue calculations. This paper summarizes the main features of MVP, results of recent studies and future plans for MVP. (author)

Burn-up credit applications for UO2 and MOX fuel assemblies in AREVA/COGEMA

International Nuclear Information System (INIS)

Toubon, H.; Riffard, C.; Batifol, M.; Pelletier, S.

2003-01-01

For the last seven years, AREVA/COGEMA has been implementing the second phase of its burn-up credit program (the incorporation of fission products). Since the early nineties, major actinides have been taken into account in criticality analyses first for reprocessing applications, then for transport and storage of fuel assemblies Next year (2004) COGEMA will take into account the six main fission products (Rh103, Cs133, Nd143, Sm149, Sm152 and Gd155) that make up 50% of the anti-reactivity of all fission products. The experimental program will soon be finished. The new burn-up credit methodology is in progress. After a brief overview of BUC R and D program and COGEMA's application of the BUC, this paper will focus on the new burn-up measurement for UO2 and MOX fuel assemblies. It details the measurement instrumentation and the measurement experiments on MOX fuels performed at La Hague in January 2003. (author)

HTR core physics and transient analyses by the Panthermix code system

Energy Technology Data Exchange (ETDEWEB)

Haas, J.B.M. de; Kuijper, J.C.; Oppe, J. [NRG - Fuels, Actinides and Isotopes group, Petten (Netherlands)

2005-07-01

At NRG Petten, core physics analyses on High Temperature gas-cooled Reactors (HTRs) are mainly performed by means of the PANTHERMIX code system. Since some years NRG is developing the HTR reactor physics code system WIMS/PANTHERMIX, based on the lattice code WIMS (Serco Assurance, UK), the 3-dimensional steady-state and transient core physics code PANTHER (British Energy, UK) and the 2-dimensional R-Z HTR thermal hydraulics code THERMIX-DIREKT (Research Centre FZJ Juelich, Germany). By means of the WIMS code nuclear data are being generated to suit the PANTHER code's neutronics. At NRG the PANTHER code has been interfaced with THERMIX-DIREKT to form PANTHERMIX, to enable core-follow/fuel management and transient analyses in a consistent manner on pebble bed type HTR systems. Also provisions have been made to simulate the flow of pebbles through the core of a pebble bed HTR, according to a given (R-Z) flow pattern. As examples of the versatility of the PANTHERMIX code system, calculations are presented on the PBMR, the South African pebble bed reactor design, to show the transient capabilities, and on a plutonium burning MEDUL-reactor, to demonstrate the core-follow/fuel management capabilities. For the investigated cases a good agreement is observed with the results of other HTR core physics codes.

Extended fuel swelling models and ultra high burn-up fuel behavior of U–Pu–Zr metallic fuel using FEAST-METAL

Energy Technology Data Exchange (ETDEWEB)

Karahan, Aydın, E-mail: karahan@alum.mit.edu [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering, Massachusetts Institute of Technology, 77 Massachusetts Avenue, 24-215, Cambridge, MA 02139 (United States); Andrews, Nathan C., E-mail: nandrews@mit.edu [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering, Massachusetts Institute of Technology, 77 Massachusetts Avenue, 24-215, Cambridge, MA 02139 (United States)

2013-05-15

Highlights: ► Improved fuel swelling models in phase structure dependent form. ► A probabilistic verification exercise for the open porosity formation threshold. ► Satisfactory validation effort for available EBR-II database. ► Ultra high burn-up behavior of U–6Zr fuel with 60% smear density fuel. -- Abstract: Computational models in FEAST-METAL U–Pu–Zr metallic fuel behavior code have been upgraded to improve fission gas, solid fission product swelling, and pore sintering behavior in a microstructure dependent form. First, fission gas bubble growth is modeled by selecting small and large bubble groups according to a fixed number of gas atoms per bubble group. Small bubbles nucleated at phase boundaries grow via gas migration and turn into large bubbles. Furthermore, bubble morphology for each phase structure is captured by selecting the number of atoms per bubble and the shape of the bubbles in a phase dependent form. The gas diffusion coefficients for the single gamma phase and effective dual (α + δ) and (β + γ) phase structures are modeled separately, using the activation energy of the corresponding phase structure. In this study, it is found that pressure sintering of the interconnected porosity in dual phases should be less effective than the reference model in order to match clad strain and fission gas release behavior. In addition to these improvements, a probabilistic approach is taken to verify the fission gas-swelling threshold at which interconnected porosity begins. This fracture problem is treated as a function of critical crack length formed via bubble coalescence. It was found that a 10% gas-swelling threshold is appropriate for a wide range of gas bubble sizes. The new version of FEAST-METAL predicts the burn-up, smear density, and axial variation of the clad hoop strain and fission gas release behavior satisfactorily for selected test pins under EBR-II conditions. The code is used to predict ultra-high burn-up U–Pu–6Zr vented

Extended fuel swelling models and ultra high burn-up fuel behavior of U–Pu–Zr metallic fuel using FEAST-METAL

International Nuclear Information System (INIS)

Karahan, Aydın; Andrews, Nathan C.

2013-01-01

Highlights: ► Improved fuel swelling models in phase structure dependent form. ► A probabilistic verification exercise for the open porosity formation threshold. ► Satisfactory validation effort for available EBR-II database. ► Ultra high burn-up behavior of U–6Zr fuel with 60% smear density fuel. -- Abstract: Computational models in FEAST-METAL U–Pu–Zr metallic fuel behavior code have been upgraded to improve fission gas, solid fission product swelling, and pore sintering behavior in a microstructure dependent form. First, fission gas bubble growth is modeled by selecting small and large bubble groups according to a fixed number of gas atoms per bubble group. Small bubbles nucleated at phase boundaries grow via gas migration and turn into large bubbles. Furthermore, bubble morphology for each phase structure is captured by selecting the number of atoms per bubble and the shape of the bubbles in a phase dependent form. The gas diffusion coefficients for the single gamma phase and effective dual (α + δ) and (β + γ) phase structures are modeled separately, using the activation energy of the corresponding phase structure. In this study, it is found that pressure sintering of the interconnected porosity in dual phases should be less effective than the reference model in order to match clad strain and fission gas release behavior. In addition to these improvements, a probabilistic approach is taken to verify the fission gas-swelling threshold at which interconnected porosity begins. This fracture problem is treated as a function of critical crack length formed via bubble coalescence. It was found that a 10% gas-swelling threshold is appropriate for a wide range of gas bubble sizes. The new version of FEAST-METAL predicts the burn-up, smear density, and axial variation of the clad hoop strain and fission gas release behavior satisfactorily for selected test pins under EBR-II conditions. The code is used to predict ultra-high burn-up U–Pu–6Zr vented

Performance optimization of spectral amplitude coding OCDMA system using new enhanced multi diagonal code

Science.gov (United States)

Imtiaz, Waqas A.; Ilyas, M.; Khan, Yousaf

2016-11-01

This paper propose a new code to optimize the performance of spectral amplitude coding-optical code division multiple access (SAC-OCDMA) system. The unique two-matrix structure of the proposed enhanced multi diagonal (EMD) code and effective correlation properties, between intended and interfering subscribers, significantly elevates the performance of SAC-OCDMA system by negating multiple access interference (MAI) and associated phase induce intensity noise (PIIN). Performance of SAC-OCDMA system based on the proposed code is thoroughly analyzed for two detection techniques through analytic and simulation analysis by referring to bit error rate (BER), signal to noise ratio (SNR) and eye patterns at the receiving end. It is shown that EMD code while using SDD technique provides high transmission capacity, reduces the receiver complexity, and provides better performance as compared to complementary subtraction detection (CSD) technique. Furthermore, analysis shows that, for a minimum acceptable BER of 10-9 , the proposed system supports 64 subscribers at data rates of up to 2 Gbps for both up-down link transmission.

Establishment of THERPRO Database and Estimation of the Effect of Fuel Burn-up on the Thermal Conductivity of Uranium Dioxide

International Nuclear Information System (INIS)

Lee, Hyun Seon

2005-02-01

Materials property data are an essential part of major disciplines in many engineering fields. To nuclear engineering, fundamental understanding of thermo-physical chemical mechanical properties of nuclear materials is very important. THERPRO data base that is re-designed and re-constructed through this study is a web-based on-line nuclear materials properties data base. For the future upgrade of the data base contemporary information technologies have been incorporated during the construction. Basically THERPRO data base has a hierarchical structure consisting of several levels: home page, element, compound, property, author, report, and bibliography level. All of data sets in each level are interconnected using network structure and thus every data can be easily retrieved including the bibliographical information by an appropriate query action. As a part of THERPRO DB utilization, the effect of fuel burn-up on the thermal conductivity of irradiated uranium dioxide is analyzed with the data contained in the data base as well as recent data published in the relevant journals. Their data are comparatively studied and the effect is estimated using FRAPCON-3 code with two in-pile data sets, BR-3 111i5 and Oconee rod 15309. The results show that the fuel center line temperature can differ 200 .deg. C∼400 .deg. C from thermal conductivity models depending on burn-up, which can significantly influence high burn-up fuel performance. In conclusion, it is demonstrated through this study that THERPRO data base can be a great utility for nuclear engineers and researchers, if appropriately utilized

Fusion core start-up, ignition, and burn simulations of reversed-field pinch (RFP) reactors

International Nuclear Information System (INIS)

Chu, Y.Y.

1988-01-01

A transient reactor simulation model is developed to investigate and simulate the start-up, ignition, and burn of a reversed-field pinch reactor. The simulation is based upon a spatially averaged plasma balance model with field profiles obtained from MHD quasi-equilibrium analysis. Alpha particle heating is estimated from Fokker-Planck calculations. The instantaneous plasma current is derived from a self-consistent circuit analysis for plasma/coil/eddy current interactions. The simulation code is applied to the TITAN RFP reactor design which features a compact, high-power-density reversed-field pinch fusion system. A contour analysis is performed using the steady-state global plasma balance. The results are presented with contours of constant plasma current. A saddle point is identified in the contour plot which determined the minimum value of plasma current required to achieve ignition. In the simulations of the TITAN RFP reactor, the OH-driven super-conducting EF coils are found to deviate from the required equilibrium values as the induced plasma current increases. A set of basic results from the simulation of TITAN RFP reactor yield a picture of RFP plasma operation in a reactor. Investigations of eddy currents are also presented and have very important in reactor design

CONTEMPT-DG containment analysis code

International Nuclear Information System (INIS)

Deem, R.E.; Rousseau, K.

1982-01-01

The assessment of hydrogen burning in a containment building during a degraded core event requires a knowledge of various system responses. These system responses (i.e. heat sinks, fan cooler units, sprays, etc.) can have a marked effect on the overall containment integrity results during a hydrogen burn. In an attempt to properly handle the various system responses and still retain the capability to perform sensitivity analysis on various parameters, the CONTEMPT-DG computer code was developed. This paper will address the historical development of the code, its various features, and the rationale for its development. Comparisons between results from the CONTEMPT-DG analyses and results from similar MARCH analyses will also be given

Validation of the VTT's reactor physics code system

International Nuclear Information System (INIS)

Tanskanen, A.

1998-01-01

At VTT Energy several international reactor physics codes and nuclear data libraries are used in a variety of applications. The codes and libraries are under constant development and every now and then new updated versions are released, which are taken in use as soon as they have been validated at VTT Energy. The primary aim of the validation is to ensure that the code works properly, and that it can be used correctly. Moreover, the applicability of the codes and libraries are studied in order to establish their advantages and weak points. The capability of generating program-specific nuclear data for different reactor physics codes starting from the same evaluated data is sometimes of great benefit. VTT Energy has acquired a nuclear data processing system based on the NJOY-94.105 and TRANSX-2.15 processing codes. The validity of the processing system has been demonstrated by generating pointwise (MCNP) and groupwise (ANISN) temperature-dependent cross section sets for the benchmark calculations of the Doppler coefficient of reactivity. At VTT Energy the KENO-VI three-dimensional Monte Carlo code is used in criticality safety analyses. The KENO-VI code and the 44GROUPNDF5 data library have been validated at VTT Energy against the ZR-6 and LR-0 critical experiments. Burnup Credit refers to the reduction in reactivity of burned nuclear fuel due to the change in composition during irradiation. VTT Energy has participated in the calculational VVER-440 burnup credit benchmark in order to validate criticality safety calculation tools. (orig.)

Study on small long-life LBE cooled fast reactor with CANDLE burn-up. Part 1. Steady state research

International Nuclear Information System (INIS)

Yan, Mingyu; Sekimoto, Hiroshi

2008-01-01

Small long-life reactor is required for some local areas. CANDLE small long-life fast reactor which does not require control rods, mining, enrichment and reprocessing plants can satisfy this demand. In a CANDLE reactor, the shapes of neutron flux, nuclide number densities and power density distributions remain constant and only shift in axial direction. The core with 1.0 m radius, 2.0 m length can realize CANDLE burn-up with nitride (enriched N-15) natural uranium as fresh fuel. Lead-Bismuth is used as coolant. From steady state analysis, we obtained the burn-up velocity, output power distribution, core temperature distribution, etc. The burn-up velocity is less than 1.0 cm/year that enables a long-life design easily. The core averaged discharged fuel burn-up is about 40%. (author)

Systemization of burnup sensitivity analysis code. 2

International Nuclear Information System (INIS)

Tatsumi, Masahiro; Hyoudou, Hideaki

2005-02-01

Towards the practical use of fast reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoint of improvements on plant efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by the development of adjusted nuclear library using the cross-section adjustment method, in which the results of criticality experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor 'JOYO'. The analysis of burnup characteristics is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, a burnup sensitivity analysis code, SAGEP-BURN, has been developed and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to users due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functions in the existing large system. It is not sufficient to unify each computational component for the following reasons; the computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore, it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion. For

Calculation of triton confinement and burn-up in tokamaks

International Nuclear Information System (INIS)

Anderson, D.; Battistoni, P.

1987-01-01

An analytical investigation is made of the confinement and subsequent burn-up of fusion produced tritons in a deuterium Tokamak plasma. Explicit approximations are obtained for the triton confinement factor, clearly displaying the scaling with physical parameters. The importance of pitch angle scattering losses during the triton slowing down is also estimated. A comparison with experiments and numerical calculations on the FT Tokamak slows good qualitative agreement. (authors)

Safety demonstration tests of air-ventilation system for the postulated explosive burning in a cell of fuel-reprocessing plant

International Nuclear Information System (INIS)

Takada, Junichi; Suzuki, Motoe; Tukamoto, Michio; Koike, Tadao; Nishio, Gunji

1995-03-01

Safety demonstration tests of an explosive burning in a cell in the reprocessing plant has been carried out in JAERI under the auspices of the Science and Technology Agency, to evaluate the safety of an air-ventilation system during the hypothetical explosion. The postulated explosive burning of organic solvent mixed with nitric acid was simulated by solid explosives. The demonstration test was performed using an industrial scale experimental facility simulating to the ventilation system of the large scale reprocessing plant in JAPAN. Propagations of pressure, temperature, and gas velocity through cells and ducts in the ventilation system were measured during the explosive burning under deflagration. Experimental data in this report can be used to evaluate the transport phenomena of radioactive materials in the ventilation system during the explosion, and also to verify computer code CELVA for the safety analysis of ventilation system in the event of explosion accidents. (author)

Base data for looking-up tables of calculation errors in JACS code system

International Nuclear Information System (INIS)

Murazaki, Minoru; Okuno, Hiroshi

1999-03-01

The report intends to clarify the base data for the looking-up tables of calculation errors cited in 'Nuclear Criticality Safety Handbook'. The tables were obtained by classifying the benchmarks made by JACS code system, and there are two kinds: One kind is for fuel systems in general geometry with a reflected and another kind is for fuel systems specific to simple geometry with a reflector. Benchmark systems were further categorized into eight groups according to the fuel configuration: homogeneous or heterogeneous; and fuel kind: uranium, plutonium and their mixtures, etc. The base data for fuel systems in general geometry with a reflected are summarized in this report for the first time. The base data for fuel systems in simple geometry with a reflector were summarized in a technical report published in 1987. However, the data in a group named homogeneous low-enriched uranium were further selected out later by the working group for making the Nuclear Criticality Safety Handbook. This report includes the selection. As a project has been organized by OECD/NEA for evaluation of criticality safety benchmark experiments, the results are also described. (author)

Burn up Theoretical Analysis of A Thorium Fuel Rod in Light Water Reactor

International Nuclear Information System (INIS)

Gaber, F.A.; Aziz, M.; Elsheikh, B.

2008-01-01

A computer model was designed to analyze the burn up and irradiation of both Th-Pu and Th-U fuel rod in a typical light water reactors conditions. MCNP computer model was designed to simulate the fuel rod burnup and evaluate neutron flux and group constants . A system of ordinary differential equations were solved numerically to evaluate the isotopic concentrations for both the two types of fuel using the previous calculated data from MCNP model. The results are analyzed and compared with published data where satisfactory agreement was found

Development of LWR fuel performance code FEMAXI-6

International Nuclear Information System (INIS)

Suzuki, Motoe

2006-01-01

LWR fuel performance code: FEMAXI-6 (Finite Element Method in AXIs-symmetric system) is a representative fuel analysis code in Japan. Development history, background, design idea, features of model, and future are stated. Characteristic performance of LWR fuel and analysis code, what is model, development history of FEMAXI, use of FEMAXI code, fuel model, and a special feature of FEMAXI model is described. As examples of analysis, PCMI (Pellet-Clad Mechanical Interaction), fission gas release, gap bonding, and fission gas bubble swelling are reported. Thermal analysis and dynamic analysis system of FEMAXI-6, function block at one time step of FEMAXI-6, analytical example of PCMI in the output increase test by FEMAXI-III, analysis of fission gas release in Halden reactor by FEMAXI-V, comparison of the center temperature of fuel in Halden reactor, and analysis of change of diameter of fuel rod in high burn up BWR fuel are shown. (S.Y.)

Behaviour of fission gas in the rim region of high burn-up UO2 fuel pellets with particular reference to results from an XRF investigation

International Nuclear Information System (INIS)

Mogensen, M.; 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. (orig.)

Introduction of thermal-hydraulic analysis code and system analysis code for HTGR

International Nuclear Information System (INIS)

Tanaka, Mitsuhiro; Izaki, Makoto; Koike, Hiroyuki; Tokumitsu, Masashi

1984-01-01

Kawasaki Heavy Industries Ltd. has advanced the development and systematization of analysis codes, aiming at lining up the analysis codes for heat transferring flow and control characteristics, taking up HTGR plants as the main object. In order to make the model of flow when shock waves propagate to heating tubes, SALE-3D which can analyze a complex system was developed, therefore, it is reported in this paper. Concerning the analysis code for control characteristics, the method of sensitivity analysis in a topological space including an example of application is reported. The flow analysis code SALE-3D is that for analyzing the flow of compressible viscous fluid in a three-dimensional system over the velocity range from incompressibility limit to supersonic velocity. The fundamental equations and fundamental algorithm of the SALE-3D, the calculation of cell volume, the plotting of perspective drawings and the analysis of the three-dimensional behavior of shock waves propagating in heating tubes after their rupture accident are described. The method of sensitivity analysis was added to the analysis code for control characteristics in a topological space, and blow-down phenomena was analyzed by its application. (Kako, I.)

Non-instrumented capsule design of HANARO irradiation test for the high burn-up large grain UO2 pellets

International Nuclear Information System (INIS)

Kim, D. H.; Lee, C. B.; Oh, D. S.

2001-01-01

Non-instrumented capsule was designed to irradiate the large grain UO 2 pellet developed for the high burn-up LWR fuel in the HANARO in-pile capsule. UO 2 pelletes will be irradiated up to the burn-up higher than 70 MWD/kgU in HANARO. To irradiate the UO 2 pellets up to the burn-up 70 MWD/kgU, need the time about 60 months and ensure the integrity of non-instrumented capsule for 30 months until replace the new capsule. In addition, to satisfy the safety criteria of HANARO such as prevention of ONB(Onset of Nucleate Boiling), fuel melting and wear damage of the capsule during the long term irradiation, design of the non-instrumented capsule was optimized

Isotopic analyses and calculation by use of JENDL-3.2 for high burn-up UO2 and MOX spent fuels

International Nuclear Information System (INIS)

Sasahara, Akihiro; Matsumura, Tetsuo; Nicolaou, G.; Betti, M.; Walker, C.T.

1997-01-01

The post irradiation examinations (PIE) were carried out for high burn-up UO 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)

On the thermal conductivity of UO2 nuclear fuel at a high burn-up of around 100 MWd/kgHM

International Nuclear Information System (INIS)

Walker, C.T.; Staicu, D.; Sheindlin, M.; Papaioannou, D.; Goll, W.; Sontheimer, F.

2006-01-01

A study of the thermal conductivity of a commercial PWR fuel with an average pellet burn-up of 102 MWd/kgHM is described. The thermal conductivity data reported were derived from the thermal diffusivity measured by the laser flash method. The factors determining the fuel thermal conductivity at high burn-up were elucidated by investigating the recovery that occurred during thermal annealing. It was found that the thermal conductivity in the outer region of the fuel was much higher than it would have been if the high burn-up structure were not present. The increase in thermal conductivity is a consequence of the removal of fission products and radiation defects from the fuel lattice during recrystallisation of the fuel grains (an integral part of the formation process of the high burn-up structure). The gas porosity in the high burn-up structure lowers the increase in thermal conductivity caused by recrystallisation

Development of long-lived radionuclide transmutation technology - Development of a code system for core analysis of the transmutation reactors

Energy Technology Data Exchange (ETDEWEB)

Cho, Nam Zin; Kim, Yong Hee; Kim, Tae Hyung; Jo, Chang Keun; Park, Chang Je [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1996-07-01

The objective of this study is to develop a code system for core analysis= of the critical transmutation reactors utilizing fast neutrons. Core characteristics of the transmutation reactors were identified and four codes, HANCELL for pincell calculation, PRISM and AFEN-H3D for core calculation, and MA{sub B}URN for depletion calculation, were developed. The pincell calculation code is based on one-dimensional collision probability method and may provide homogenized/condensed parameters of a pincell and also can homogenize the control assembly via a nonlinear iterative method. The core calculation codes, PRISM and AFEN-H3D, solve the multi-group, multi-dimensional neutron diffusion equations for a hexagonal geometry and they are based on the finite difference method and analytic function expansion nodal (AFEN) method, respectively. The MA{sub B}URN code san analyze the behavior of actinides and fission products in a reactor core. Through benchmarking, we confirmed that the newly developed codes provide accurate solutions. 30 refs., 10 tabs., 8 figs. (author)

Experimental methods for burn-up determination in nuclear fuels, 1

International Nuclear Information System (INIS)

Taddei, J.F. de A.C.; Rodrigues, C.

1977-01-01

A method is presented that allows the calculation of the total percentage of atoms having undergone fission ('burn up') in nuclear fuels, from the measurement of absolute amounts of fission product neodymium-148 and of uranium and plutoniun present in the spent fuel, the fission yield of neodymium-148 being known. These measurements are performed through the mass spectrometry- isotope dilution technique [pt

Risks for skin and other cancers up to 25 years after burn injuries

DEFF Research Database (Denmark)

Mellemkjaer, Lene; Hölmich, Lisbet R; Gridley, Gloria

2006-01-01

BACKGROUND: Malignant degeneration of chronic ulcers such as nonhealed burn wounds has been described in the literature, but this phenomenon has never been quantified in an epidemiologic study. We investigated the risks for skin and other cancers among patients with a prior burn. METHODS: We...... with that in the general population of Denmark. RESULTS: Patients with burn had 139 skin cancers, with 189 expected, yielding a standardized incidence ratio of 0.7 (95% confidence interval = 0.6-0.9). This reduced risk was due mainly to deficits of basal cell carcinoma and malignant melanoma, whereas the number...... of squamous cell carcinomas observed was close to expected. We saw no consistent increases in risk for skin cancer in the subgroups of patients with the most severe injuries or with the longest periods of follow up. CONCLUSIONS: The tendency to malignant degeneration of burn scars, described in previous...

Development of high performance liquid chromatography for rapid determination of burn-up of nuclear fuels

International Nuclear Information System (INIS)

Joseph, M.; Karunasagar, D.; Saha, B.

1996-01-01

Burn-up an important parameter during evaluation of the performance of any nuclear fuel. Among the various techniques available, the preferred one for its determination is based on accurate measurement of a suitable fission product monitor and the residual heavy elements. Since isotopes of rare earth elements are generally used as burn-up monitors, conditions were standardized for rapid separation (within 15 minutes) of light rare earths using high performance liquid chromatography based on either anion exchange (Partisil 10 SAX) in methanol-nitric acid medium or by cation exchange on a reverse phase column (Spherisorb 5-ODS-2 or Supelcosil LC-18) dynamically modified with 1-octane sulfonate or camphor-10-sulfonic acid (β). Both these methods were assessed for separation of individual fission product rare earths from their mixtures. A new approach has been examined in detail for rapid assay of neodymium, which appears promising for faster and accurate measurement of burn-up. (author)

Burn up determination of IEAR-1 fuel elements by non destructive gamma ray spectrometry method

International Nuclear Information System (INIS)

Soares, A.J.

1977-01-01

Measurement of nuclear fuel burn up by non destructive gamma ray spectrometry is discussed, and results of such measurements, made at the Instituto de Energia Atomica (IEA), are given. Specifically, the burn up of an MTR (Material Testing Reactor) fuel element removed from the IEAR-1 swimming pool reactor in 1958 is evaluated from the measured Cs-137 activity, which gives a single 661,6 keV gamma ray. Due to the long decay time of the test element, no other fission decay product activity could be detected. Analysis of measurements, made with a 3'' x 3'' NaI(Tl) detector at 330 distinct points of the element, showed the total burn up to 3.3 +- -+ 0.8 mg. This is in agreement with a calculated value. As the maximum temperature of IEAR-1 fuel elements is of the order of 40 0 C, migration effects of Cs-137 was not considered, this being significant only at fuel temperature in excess of 1000 0 C [pt

Criticality qualification of a new Monte Carlo code for reactor core analysis

International Nuclear Information System (INIS)

Catsaros, N.; Gaveau, B.; Jaekel, M.; Maillard, J.; Maurel, G.; Savva, P.; Silva, J.; Varvayanni, M.; Zisis, Th.

2009-01-01

In order to accurately simulate Accelerator Driven Systems (ADS), the utilization of at least two computational tools is necessary (the thermal-hydraulic problem is not considered in the frame of this work), namely: (a) A High Energy Physics (HEP) code system dealing with the 'Accelerator part' of the installation, i.e. the computation of the spectrum, intensity and spatial distribution of the neutrons source created by (p, n) reactions of a proton beam on a target and (b) a neutronics code system, handling the 'Reactor part' of the installation, i.e. criticality calculations, neutron transport, fuel burn-up and fission products evolution. In the present work, a single computational tool, aiming to analyze an ADS in its integrity and also able to perform core analysis for a conventional fission reactor, is proposed. The code is based on the well qualified HEP code GEANT (version 3), transformed to perform criticality calculations. The performance of the code is tested against two qualified neutronics code systems, the diffusion/transport SCALE-CITATION code system and the Monte Carlo TRIPOLI code, in the case of a research reactor core analysis. A satisfactory agreement was exhibited by the three codes.

Analysis of an XADS Target with the System Code TRACE

International Nuclear Information System (INIS)

Jaeger, Wadim; Sanchez Espinoza, Victor H.; Feng, Bo

2008-01-01

Accelerator-driven systems (ADS) present an option to reduce the radioactive waste of the nuclear industry. The experimental Accelerator-Driven System (XADS) has been designed to investigate the feasibility of using ADS on an industrial scale to burn minor actinides. The target section lies in the middle of the subcritical core and is bombarded by a proton beam to produce spallation neutrons. The thermal energy produced from this reaction requires a heat removal system for the target section. The target is cooled by liquid lead-bismuth-eutectics (LBE) in the primary system which in turn transfers the heat via a heat exchanger (HX) to the secondary coolant, Diphyl THT (DTHT), a synthetic diathermic fluid. Since this design is still in development, a detailed investigation of the system is necessary to evaluate the behavior during normal and transient operations. Due to the lack of experimental facilities and data for ADS, the analyses are mostly done using thermal hydraulic codes. In addition to evaluating the thermal hydraulics of the XADS, this paper also benchmarks a new code developed by the NRC, TRACE, against other established codes. The events used in this study are beam power switch-on/off transients and a loss of heat sink accident. The obtained results from TRACE were in good agreement with the results of various other codes. (authors)

Two-year follow-up of outcomes related to scarring and distress in children with severe burns.

Science.gov (United States)

Wurzer, Paul; Forbes, Abigail A; Hundeshagen, Gabriel; Andersen, Clark R; Epperson, Kathryn M; Meyer, Walter J; Kamolz, Lars P; Branski, Ludwik K; Suman, Oscar E; Herndon, David N; Finnerty, Celeste C

2017-08-01

We assessed the perception of scarring and distress by pediatric burn survivors with burns covering more than one-third of total body surface area (TBSA) for up to 2 years post-burn. Children with severe burns were admitted to our hospital between 2004 and 2012, and consented to this IRB-approved-study. Subjects completed at least one Scars Problems and/or Distress questionnaire between discharge and 24 months post burn. Outcomes were modeled with generalized estimating equations or using mixed linear models. Significance was accepted at p body areas over time (p self-conscious with respect to their body image even 2 years after burn injury. Implications for Rehabilitation According to self-assessment questionnaires, severely burned children perceive significant improvements in scarring and distress during the first 2 years post burn. Significant improvements were seen in reduction of pain, itching, sleeping disturbances, tightness, range of motion, and strength (p body areas. The rehabilitation team should provide access to wigs or other aids to pediatric burn survivors to address these needs.

Establishment of Technical Collaboration basis between Korea and France for the development of severe accident assessment computer code under high burnup condition

International Nuclear Information System (INIS)

Kim, H. D.; Kim, D. H.; Park, S. Y.; Park, J. H.

2005-10-01

This project was performed by KAERI in the frame of construction of the international cooperative basis on the nuclear energy. This was supported from MOST under the title of 'Establishment of Technical Collaboration basis between Korea and France for the development of severe accident assessment computer code under high burn up condition'. The current operating NPP are converting the burned fuel to the wasted fuel after burn up of 40 GWD/MTU. But in Korea, burn up of more than 60 GWD/MTU will be expected because of the high fuel efficiency but also cost saving for storing the wasted fuel safely. The domestic research for the purpose of developing the fuel and the cladding that can be used under the high burn up condition up to 100 GWD/MTU is in progress now. But the current computer code adopts the model and the data that are valid only up to the 40 GWD/MTU at most. Therefore the current model could not take into account the phenomena that may cause differences in the fission product release behavior or in the core damage process due to the high burn up operation (more than 40 GWD/MTU). To evaluate the safety of the NPP with the high burn up fuel, the improvement of current severe accident code against the high burn up condition is an important research item. Also it should start without any delay. Therefore, in this study, an expert group was constructed to establish the research basis for the severe accident under high burn up conditions. From this expert group, the research items regarding the high burn up condition were selected and identified through discussion and technical seminars. Based on these selected items, the meeting between IRSN and KAERI to find out the cooperative research items on the severe accident under the high burn up condition was held in the IRSN headquater in Paris. After the meeting, KAERI and IRSN agreed to cooperate with each other on the selected items, and to co-host the international seminar, and to develop the model and to

[Clinical effect of three dimensional human body scanning system BurnCalc in the evaluation of burn wound area].

Science.gov (United States)

Lu, J; Wang, L; Zhang, Y C; Tang, H T; Xia, Z F

2017-10-20

Objective: To validate the clinical effect of three dimensional human body scanning system BurnCalc developed by our research team in the evaluation of burn wound area. Methods: A total of 48 burn patients treated in the outpatient department of our unit from January to June 2015, conforming to the study criteria, were enrolled in. For the first 12 patients, one wound on the limbs or torso was selected from each patient. The stability of the system was tested by 3 attending physicians using three dimensional human body scanning system BurnCalc to measure the area of wounds individually. For the following 36 patients, one wound was selected from each patient, including 12 wounds on limbs, front torso, and side torso, respectively. The area of wounds was measured by the same attending physician using transparency tracing method, National Institutes of Health (NIH) Image J method, and three dimensional human body scanning system BurnCalc, respectively. The time for getting information of 36 wounds by three methods was recorded by stopwatch. The stability among the testers was evaluated by the intra-class correlation coefficient (ICC). Data were processed with randomized blocks analysis of variance and Bonferroni test. Results: (1) Wound area of patients measured by three physicians using three dimensional human body scanning system BurnCalc was (122±95), (121±95), and (123±96) cm(2,) respectively, and there was no statistically significant difference among them ( F =1.55, P >0.05). The ICC among 3 physicians was 0.999. (2) The wound area of limbs of patients measured by transparency tracing method, NIH Image J method, and three dimensional human body scanning system BurnCalc was (84±50), (76±46), and (84±49) cm(2,) respectively. There was no statistically significant difference in the wound area of limbs of patients measured by transparency tracing method and three dimensional human body scanning system BurnCalc ( P >0.05). The wound area of limbs of patients

Study on the sensitivity of Self-Powered Neutron Detectors (SPND) and its change due to burn-up

International Nuclear Information System (INIS)

Cho, Gyuseong; Lee, Wanno; Yoon, Jeong-Hyoun.

1996-01-01

Self-Powered Neutron Detectors (SPND) are currently used to estimate the power generation distribution and fuel burn-up in several nuclear power reactors in Korea. While they have several advantages such as small size, low cost, and relatively simple electronics required in conjunction with its usage, it has some intrinsic problems of the low level of output current, a slow response time, the rapid change of sensitivity which makes it difficult to use for a long term. In this paper, Monte Carlo simulation was accomplished to calculate the escape probability as a function of the birth position for the typical geometry of rhodium-based SPNDs. Using the simulation result, the burn-up profile of rhodium number density and the neutron sensitivity is calculated as a function of burn-up time in the reactor. The sensitivity of the SPND decreases non-linearly due to the high absorption cross-section and the non-uniform burn-up of rhodium in the emitter rod. The method used here can be applied to the analysis of other types of SPNDs and will be useful in the optimum design of new SPNDs for long-term usage. (author)

Development of destructive methods of burn-up determination and their application on WWER type nuclear fuels

International Nuclear Information System (INIS)

Hermann, A.; Stephan, H.; Nebel, D.

1984-03-01

Results are described of a cooperation between the Central Institute of Nuclear Research Rossendorf and the Radium Institute 'V.G. Chlopin' Leningrad in the field of destructive burn-up determination. Laboratory methods of burn-up determination using the classical monitors 137 Cs, 106 Ru, 148 Nd and isotopes of heavy metals (U, Pu) as well as the usefulness of 90 Sr, stable isotopes of Ru and Mo as monitors are dealt with. The analysis of the fuel components uranium (spectrophotometry, potentiometric titration, mass-spectrometric isotope dilution) and plutonium (spectrophotometry, coulometric titration, mass- and alpha-spectrometric isotope dilution) is fully described. Possibilities of increasing the reproducibility (automatic adjusting of measurement conditions) and the sensibility (ion impuls counting) of mass-spectrometric measurements are proposed and applied to a precise determination of Am and Cm isotopic composition. The methods have been used for burn-up analysis of spent WWER (especially WWER-440) fuel. (author)

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)

Comparison of the ENIGMA code with experimental data on thermal performance, stable fission gas and iodine release at high burnup

Energy Technology Data Exchange (ETDEWEB)

Killeen, J C [Nuclear Electric plc, Barnwood (United Kingdom)

1997-08-01

The predictions of the ENIGMA code have been compared with data from high burn-up fuel experiments from the Halden and RISO reactors. The experiments modelled were IFA-504 and IFA-558 from Halden and the test II-5 from the RISO power burnup test series. The code has well modelled the fuel thermal performance and has provided a good measure of iodine release from pre-interlinked fuel. After interlinkage the iodine predictions remain a good fit for one experiment, but there is significant overprediction for a second experiment (IFA-558). Stable fission gas release is also well modelled and the predictions are within the expected uncertainly band throughout the burn-up range. This report presents code predictions for stable fission gas release to 40GWd/tU, iodine release measurements to 50GWd/tU and thermal performance (fuel centre temperature) to 55GWd/tU. Fuel ratings of up to 38kW/m were modelled at the high burn-up levels. The code is shown to accurately or conservatively predict all these parameters. (author). 1 ref., 6 figs.

Computer code TOBUNRAD for PWR fuel bundle heat-up calculations

International Nuclear Information System (INIS)

Shimooke, Takanori; Yoshida, Kazuo

1979-05-01

The computer code TOBUNRAD developed is for analysis of ''fuel-bundle'' heat-up phenomena in a loss-of-coolant accident of PWR. The fuel bundle consists of fuel pins in square lattice; its behavior is different from that of individual pins during heat-up. The code is based on the existing TOODEE2 code which analyzes heat-up phenomena of single fuel pins, so that the basic models of heat conduction and transfer and coolant flow are the same as the TOODEE2's. In addition to the TOODEE2 features, unheated rods are modeled and radiation heat loss is considered between fuel pins, a fuel pin and other heat sinks. The TOBUNRAD code is developed by a new FORTRAN technique which makes it possible to interrupt a flow of program controls wherever desired, thereby attaching several subprograms to the main code. Users' manual for TOBUNRAD is presented: The basic program-structure by interruption method, physical and computational model in each sub-code, usage of the code and sample problems. (author)

CRISTAL V1: Criticality package for burn up credit calculations

International Nuclear Information System (INIS)

Gomit, Jean-Michel; Cousinou, Patrick; Gantenbein, Francoise; Diop, Cheikh; Fernandez de Grado, Guy; Mijuin, Dominique; Grouiller, Jean-Paul; Marc, Andre; Toubon, Herve

2003-01-01

The first version of the CRISTAL package, created and validated as part of a joint project between IRSN, COGEMA and CEA, was delivered to users in November 1999. This fruitful cooperation between IRSN, COGEMA and CEA has been pursued until 2003 with the development and the validation of the package CRISTAL V1, whose main objectives are to improve the criticality safety studies including the Burn up Credit effect. (author)

Development and verifications of fast reactor fuel design code ''Ceptar''

International Nuclear Information System (INIS)

Ozawa, T.; Nakazawa, H.; Abe, T.

2001-01-01

The annular fuel is very beneficial for fast reactors, because it is available for both high power and high burn-up. Concerning the irradiation behavior of the annular fuel, most of annular pellets irradiated up to high burn-up showed shrinkage of the central hole due to deformation and restructuring of the pellets. It is needed to predict precisely the shrinkage of the central hole during irradiation, because it has a great influence on power-to-melt. In this paper, outline of CEPTAR code (Calculation code to Evaluate fuel pin stability for annular fuel design) developed to meet this need is presented. In this code, the radial profile of fuel density can be computed by using the void migration model, and law of conservation of mass defines the inner diameter. For the mechanical analysis, the fuel and cladding deformation caused by the thermal expansion, swelling and creep is computed by the stress-strain analysis using the approximation of plane-strain. In addition, CEPTAR can also take into account the effect of Joint-Oxide-Gain (JOG) which is observed in fuel-cladding gap of high burn-up fuel. JOG has an effect to decrease the fuel swelling and to improve the gap conductance due to deposition of solid fission product. Based on post-irradiation data on PFR annular fuel, we developed an empirical model for JOG. For code verifications, the thermal and mechanical data obtained from various irradiation tests and post-irradiation examinations were compared with the predictions of this code. In this study, INTA (instrumented test assembly) test in JOYO, PTM (power-to-melt) test in JOYO, EBR-II, FFTF and MTR in Harwell laboratory, and post-irradiation examinations on a number of PFR fuels, were used as verification data. (author)

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

Energy Technology Data Exchange (ETDEWEB)

Venkiteswaran, C.N., E-mail: cnv@igcar.gov.in; 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.

DANDE: a linked code system for core neutronics/depletion analysis

International Nuclear Information System (INIS)

LaBauve, R.J.; England, T.R.; George, D.C.; MacFarlane, R.E.; Wilson, W.B.

1985-06-01

This report describes DANDE - a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of updating the nuclear data file during a calculation; a detailed treatment of depletion, burnable poisons as well as fuel; and the ability to make geometric changes such as control rod repositioning and fuel relocation in the course of a calculation. The detailed treatment of reactor fuel burnup, fission-product creation and decay, as well as inventories of higher-order actinides is a necessity when predicting the behavior of reactor fuel under increased burn conditions. The operation of the code system is made clear in this report by following a sample problem

DANDE: a linked code system for core neutronics/depletion analysis

International Nuclear Information System (INIS)

LaBauve, R.J.; England, T.R.; George, D.C.; MacFarlane, R.E.; Wilson, W.B.

1986-01-01

This report describes DANDE - a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of updating the nuclear data file during a calculation; a detailed treatment of depletion, burnable poisons as well as fuel; and the ability to make geometric changes such as control rod repositioning and fuel relocation in the cource of a calculation. The detailed treatment of reactor fuel burnup, fission-product creation and decay, as well as inventories of higher-order actinides is a necessity when predicting the behavior of reactor fuel under increased burn conditions. The operation of the code system is illustrated in this report by two sample problems. 25 refs

DANDE-a linked code system for core neutronics/depletion analysis

International Nuclear Information System (INIS)

LaBauve, R.J.; England, T.R.; George, D.C.; MacFarlane, R.E.; Wilson, W.B.

1986-01-01

This report describes DANDE-a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of updating the nuclear data file during a calculation; a detailed treatment of depletion, burnable poisons as well as fuel; and the ability to make geometric changes such as control rod repositioning and fuel relocation in the course of a calculation. The detailed treatment of reactor fuel burnup, fission-product creation and decay, as well as inventories of higher-order actinides is a necessity when predicting the behavior of the reactor fuel under increased burn conditions. The operation of the code system is illustrated in this report by two actual problems

1-D hybrid code for FRM start-up

International Nuclear Information System (INIS)

Stark, R.A.; Miley, G.H.

1982-01-01

A one-D hybrid has been developed to study the start-up of the FRM via neutral-beam injection. The code uses a multi-group numerical model originally developed by J. Willenberg to describe fusion product dynamics in a solenoidal plasma. Earlier we described such a model for use in determining self-consistent ion currents and magnetic fields in FRM start-up. However, consideration of electron dynamics during start-up indicate that the electron current will oppose the injected ion current and may even foil the attempt to achieve reversal. For this reason, we have combined the multi-group ion (model) with a fluid treatment for electron dynamics to form the hybrid code FROST (Field Reversed One-dimensional STart-up). The details of this merger, along with sample results of operation of FROST, are given

Calculation of burn-up data for spent LWR-fuels with respect to the design of spent fuel reprocessing plants

International Nuclear Information System (INIS)

Gasteiger, R.

1976-11-01

The design of spent fuel reprocessing plants makes necessary a detailed knowledge of the composition of the incoming fuels as a function of burn-up. This report gives a broad review on the composition of radionuclides in fuels (fission products, actinides) and structural materials for different burn-up data. (orig.) [de

Making of a burn unit: SOA burn center

Directory of Open Access Journals (Sweden)

Jayant Kumar Dash

2016-01-01

Full Text Available Each year in India, burn injuries account for more than 6 million hospital emergency department visits; of which many require hospitalization and are referred to specialized burn centers. There are few burn surgeons and very few burn centers in India. In our state, Odisha, there are only two burn centers to cater to more than 5000 burn victims per year. This article is an attempt to share the knowledge that I acquired while setting up a new burn unit in a private medical college of Odisha.

Modeling approach for annular-fuel elements using the ASSERT-PV subchannel code

International Nuclear Information System (INIS)

Dominguez, A.N.; Rao, Y.

2012-01-01

The internally and externally cooled annular fuel (hereafter called annular fuel) is under consideration for a new high burn-up fuel bundle design in Atomic Energy of Canada Limited (AECL) for its current, and its Generation IV reactor. An assessment of different options to model a bundle fuelled with annular fuel elements is presented. Two options are discussed: 1) Modify the subchannel code ASSERT-PV to handle multiple types of elements in the same bundle, and 2) coupling ASSERT-PV with an external application. Based on this assessment, the selected option is to couple ASSERT-PV with the thermalhydraulic system code CATHENA. (author)

Development of dynamic simulation code for fuel cycle fusion reactor

Energy Technology Data Exchange (ETDEWEB)

Aoki, Isao; Seki, Yasushi [Department of Fusion Engineering Research, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki (Japan); Sasaki, Makoto; Shintani, Kiyonori; Kim, Yeong-Chan

1999-02-01

A dynamic simulation code for fuel cycle of a fusion experimental reactor has been developed. The code follows the fuel inventory change with time in the plasma chamber and the fuel cycle system during 2 days pulse operation cycles. The time dependence of the fuel inventory distribution is evaluated considering the fuel burn and exhaust in the plasma chamber, purification and supply functions. For each subsystem of the plasma chamber and the fuel cycle system, the fuel inventory equation is written based on the equation of state considering the fuel burn and the function of exhaust, purification, and supply. The processing constants of subsystem for steady states were taken from the values in the ITER Conceptual Design Activity (CDA) report. Using this code, the time dependence of the fuel supply and inventory depending on the burn state and subsystem processing functions are shown. (author)

Changes of the inventory of radioactive materials in reactor fuel from uranium in changing to higher burn-up and determining the important effects of this

International Nuclear Information System (INIS)

Kirchner, G.; Schaefer, R.

1985-01-01

The knowledge of the nuclide composition during and after use in the reactor is an essential, in order to be able to determine the effects associated with the operation of nuclear plants. The missing reliable data on the inventory of radioactive materials resulting from the expected change to higher burn-ups of uranium fuels in West Germany are calculated. The reliability of the program system used for this, which permits a one-dimensional account taken of the fuel rod cell and measurement of the changes of specific sets of nuclear data depending on burn-up, is confirmed by the comparison with experimentally found concentrations of important nuclides in fuel samples at Obrigheim nuclear power station. Realistic conditions of use are defined for a range of burn-up of 33 GWd/t to 55 GWd/t and the effects of changes of the number of cycles and the use of types of fuel elements being developed on the composition of the inventory are determined. The plutonium compositions during use in the reactor are given and are tabulated with the inventory for decay times up to 30 years. Effects during change to higher burn-ups are examined and discussed for the maximum inventories during use of fuel and for heat generation during final storage. (orig./HP) [de

Direct measurement of burn up monitor by Pulsed Laser Deposition (PLD) followed by Isotopic Dilution Mass Spectrometry

International Nuclear Information System (INIS)

Sajimol, R.; Manoravi, P.; NaIini, S.; Balasubramanian, R.; Joseph, M.

2012-01-01

Burn-up measurement is an important aspect in the assessment of fuel performance especially for experimental nuclear fuels. Conventional mass spectrometric technique offer the best accuracy for determination of burn-up but they suffer from the labour intensive and time consuming chemical separation procedures followed by mass spectrometric analysis. Our laboratory has reported a potential laser mass spectrometric technique with advantages of (i) direct and fast measurement of ion intensities of selected rare earth element and residual heavy element atoms to deduce burn up and (ii) adaptability to remote handling of radioactive samples. Direct quantification of burn up monitor element in fuel in the form of pellet as well as liquid was probed by pulsed laser deposition followed by Isotopic Dilution Mass Spectrometric technique (IDMS). The procedure involving laser ablation of heavy element (namely U and Pu) and fission product (Nd, La etc) from a simulated spent fuel matrix followed by isotopic dilution mass spectrometry using thermal ionization mass spectrometry (TIMS) has been presently attempted to arrive at the rare earth element to heavy element ratio to deduce burn up using the methodology described in our earlier work. The details of IDMS technique has been reviewed by Heumann et al. Accurately weighed amounts of major rare earth fission products such as Nd, La, Ce and Sm in solution form were mixed with known quantity of uranium solution (all the weights are corresponding to their fission yields and the residual heavy element atoms after a given burn up) and mixed together to attain uniformity. The solution is then dried and resulting powder was pelletized and sintered. Subsequently, the pellet was ablated with pulsed laser (8 ns, 532 nm, Nd-YAG) and the plume was deposited on a glass plate. This deposit was dissolved in minimum amount of nitric acid. A known volume of the solution was mixed with spike (for e.g., 150 Nd/ 142 Nd, 233 U/ 238 U in this study

Follow-up care of children suffered from burns

Directory of Open Access Journals (Sweden)

Konstantin Aleksandrovich Afonichev

2015-03-01

Full Text Available Outcomes of III-VI AB degree burns in children,regardless of the nature of treatment in the acute andrecovery period, are the development of scar contractures and deformities of the joints. However, thecorrect organization of follow-up care and rehabilitation treatment can significantly reduce the severity and facilitates the full recovery of the affected segment. Based on the analysis of their own material, the author defines the early stage of rehabilitation in these patients before full maturation of scar tissue or before the formation of functionally significant joint contractures, and later period, when there are indications for surgical rehabilitation. In the early period, follow-up care is recommended in 1 month after discharge and then on a quarterly basis, and with the appearance of deformities - at least once in 2 months. At the2nd stage of rehabilitation, older children and children of secondary school age are subject to follow-up care at least 1 time per year of primary school age - atleast once in 6 months, preschool children - every3 months. The proposed assessment of scar tissuehelps to determine the terms of follow-up care. Usingthis scheme of follow-up care and appropriate treatment allowed the author to obtain excellent and goodresults in 87-90 % of cases at the stages of rehabilitaion.

Burn-up determinations and dimensional measurements of TRIGA-HEU fuel elements from the 14 MW steady-state core

International Nuclear Information System (INIS)

Toma, C.; Alexa, Al.; Craciunescu, T.; Pirvan, M.; Dobrin, R.

2008-01-01

In this paper there are presented the results of nondestructive examination in Post Irradiation Examination Laboratory for twenty five fuel rods selected from 14 MW steady state core. Gamma scanning and dimensional measurements were carried out in order to determine burn-up and diametric deflection of the fuel rods. Also, some comparisons with SSR Safety Report estimations for the maximum burn-up pin were made. (authors)

General features of the neutronics design code EQUICYCLE

International Nuclear Information System (INIS)

Jirlow, K.

1978-10-01

The neutronics code EQUICYCLE has been developed and improved over a long period of time. It is expecially adapted to survey type design calculations of large fast power reactors with particular emphasis on the nuclear parameters for a realistic equilibrium fuel cycle. Thus the code is used to evaluate the breeding performance, the power distributions and the uranium and plutonium mass balance for realistic refuelling schemes. In addition reactivity coefficients can be calculated and the influence of burnup could be assessed. The code is two-dimensional and treats the reactor core in R-Z geometry. The basic ideas of the calculating scheme are successive iterative improvement of cross-section sets and flux spectra and use of the mid-cycle flux for burning the fuel according to a specified refuelling scheme. Normally given peak burn-ups and maximum power densities are used as boundary conditions. The code is capable of handling the unconventional, so called heterogeneous cores. (author)

Cellular automata approach to investigation of high burn-up structures in nuclear reactor fuel

International Nuclear Information System (INIS)

Akishina, E.P.; Ivanov, V.V.; Kostenko, B.F.

2005-01-01

Micrographs of uranium dioxide (UO 2 ) corresponding to exposure times in reactor during 323, 953, 971, 1266 and 1642 full power days were investigated. The micrographs were converted into digital files isomorphous to cellular automata (CA) checkerboards. Such a representation of the fuel structure provides efficient tools for its dynamics simulation in terms of primary 'entities' imprinted in the micrographs. Besides, it also ensures a possibility of very effective micrograph processing by CA means. Interconnection between the description of fuel burn-up development and some exactly soluble models is ascertained. Evidences for existence of self-organization in the fuel at high burn-ups were established. The fractal dimension of microstructures is found to be an important characteristic describing the degree of radiation destructions

Bar-code automated waste tracking system

International Nuclear Information System (INIS)

Hull, T.E.

1994-10-01

The Bar-Code Automated Waste Tracking System was designed to be a site-Specific program with a general purpose application for transportability to other facilities. The system is user-friendly, totally automated, and incorporates the use of a drive-up window that is close to the areas dealing in container preparation, delivery, pickup, and disposal. The system features ''stop-and-go'' operation rather than a long, tedious, error-prone manual entry. The system is designed for automation but allows operators to concentrate on proper handling of waste while maintaining manual entry of data as a backup. A large wall plaque filled with bar-code labels is used to input specific details about any movement of waste

EBSD and TEM Characterization of High Burn-up Mixed Oxide Fuel

International Nuclear Information System (INIS)

Teague, Melissa C; Gorman, Brian P.; Miller, Brandon D; 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 approximately 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 approximately 2.5x 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 approximately 25 um 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

Systemization of burnup sensitivity analysis code (2) (Contract research)

International Nuclear Information System (INIS)

Tatsumi, Masahiro; Hyoudou, Hideaki

2008-08-01

Towards the practical use of fast reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoint of improvements on plant economic efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by the development of adjusted nuclear library using the cross-section adjustment method, in which the results of critical experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor 'JOYO'. The analysis of burnup characteristic is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, a burnup sensitivity analysis code, SAGEP-BURN, has been developed and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to users due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functions in the existing large system. It is not sufficient to unify each computational component for the following reasons: the computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore, it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion

Usage of burnt fuel isotopic compositions from engineering codes in Monte-Carlo code calculations

International Nuclear Information System (INIS)

Aleshin, Sergey S.; Gorodkov, Sergey S.; Shcherenko, Anna I.

2015-01-01

A burn-up calculation of VVER's cores by Monte-Carlo code is complex process and requires large computational costs. This fact makes Monte-Carlo codes usage complicated for project and operating calculations. Previously prepared isotopic compositions are proposed to use for the Monte-Carlo code (MCU) calculations of different states of VVER's core with burnt fuel. Isotopic compositions are proposed to calculate by an approximation method. The approximation method is based on usage of a spectral functionality and reference isotopic compositions, that are calculated by engineering codes (TVS-M, PERMAK-A). The multiplication factors and power distributions of FA and VVER with infinite height are calculated in this work by the Monte-Carlo code MCU using earlier prepared isotopic compositions. The MCU calculation data were compared with the data which were obtained by engineering codes.

LWR FA burn up: A challenge to optimize the entire fuel cycle to assure the envisaged benefit

Energy Technology Data Exchange (ETDEWEB)

Peehs, M [Siemens AG Unternehmensbereich KWU, Erlangen (Germany)

1997-12-01

Commercial LWR fuel will be limited to a maximum of U-235 content of 5% since the front end of the fuel cycle is licensed and prepared for that maximal enrichment. BWR- and PWR-reloads can be designed achieving batch average burn up over 60 GWd/tHM. In Germany the batch average burn up will presumably increase to this level, since the reload market is requesting further reductions in the fuel cycle inventories. However, it must be noted that the envisaged benefit can only be assured if the entire fuel cycle is optimized. Not all steps in the fuel cycle will bring a positive contribution bu the balance of all individual contributions must realize the envisaged integral benefit. In order to increase the burn up of the nuclear fuel beneficially further R and D both in the front end as well as in the back end of the fuel cycle is needed. An underestimation of the front end/back end interfaces may consume all benefits gained from isolated front optimizations. Back end R and D must be at once concentrated to avoid conservative enveloping licensing for the subsequent steps in the back end of the fuel cycle. Increasing burn up in the front end means making more and more use of the structural materials reserves.

LWR FA burn up: A challenge to optimize the entire fuel cycle to assure the envisaged benefit

International Nuclear Information System (INIS)

Peehs, M.

1997-01-01

Commercial LWR fuel will be limited to a maximum of U-235 content of 5% since the front end of the fuel cycle is licensed and prepared for that maximal enrichment. BWR- and PWR-reloads can be designed achieving batch average burn up over 60 GWd/tHM. In Germany the batch average burn up will presumably increase to this level, since the reload market is requesting further reductions in the fuel cycle inventories. However, it must be noted that the envisaged benefit can only be assured if the entire fuel cycle is optimized. Not all steps in the fuel cycle will bring a positive contribution bu the balance of all individual contributions must realize the envisaged integral benefit. In order to increase the burn up of the nuclear fuel beneficially further R and D both in the front end as well as in the back end of the fuel cycle is needed. An underestimation of the front end/back end interfaces may consume all benefits gained from isolated front optimizations. Back end R and D must be at once concentrated to avoid conservative enveloping licensing for the subsequent steps in the back end of the fuel cycle. Increasing burn up in the front end means making more and more use of the structural materials reserves

Contribution to the qualification of the Neptune system. Application to the follow-up of the Tihange reactor

International Nuclear Information System (INIS)

Tournier, Dominique.

1980-08-01

For the calculations of light water reactors the modular system Neptune has been developed. It includes transport, diffusion, thermohydraulic and kinetic codes and so allows the treatment of the various problems of core physics. The first part of this thesis is devoted to a comparison of the most usually used modulus of the neutron transport code (APOLLO). Two examples are considered: a PWR lattice and a BWR U-Pu mixed assembly. The consequences of the different hypotheses made to solve the Boltzmann's equation by a collision probability method can be appreciated on these practical cases. The second part is a check of the complete calculation scheme against experimental results obtained during the first cycle of Tihange (900 MWe PWR). The core calculation is a 3D-diffusion calculation taking into account the thermohydraulic feedbacks; the macroscopic cross-sections needed by the neutron calculation are obtained by the transport code and tabulated versus the burn-up, the fuel temperature and the water density. The results prove that Neptune can now be considered as a precise and reliable tool [fr

Opening up codings?

DEFF Research Database (Denmark)

Steensig, Jakob; Heinemann, Trine

2015-01-01

doing formal coding and when doing more “traditional” conversation analysis research based on collections. We are more wary, however, of the implication that coding-based research is the end result of a process that starts with qualitative investigations and ends with categories that can be coded...

Transients and burn dynamics in advanced tokamak fusion reactors

International Nuclear Information System (INIS)

Mantsinen, M.J.; Salomaa, R.R.E.

1994-01-01

Transient behavior of D 3 He-tokamak reactors is investigated numerically using a zero-dimensional code with prescribed profiles. Pure D 3 He start-up is compared to DT-assisted and DT-ignited start-ups. We have considered two categories of transients which could extinguish steady fusion burn: fuelling interruptions and sudden confinement changes similar to the L → H transients occurring in present-day tokamaks. Shutdown with various current and density ramp-down scenarios are studied, too. (author)

Burning minor actinides in a HTR energy spectrum

International Nuclear Information System (INIS)

Pohl, Christoph; Rütten, H. Jochem

2012-01-01

Highlights: ► Burn-up analysis for varying plutonium/minor actinide fuel compositions. ► The influence of varying heavy metal fuel element loads is investigated. ► Significant burn-up via radiative capture and subsequently fission is observed. ► Difference observed between fuel element burn-up and total actinide burning rate. - Abstract: The generation of nuclear energy by means of the existing nuclear reactor systems is based mainly on the fission of U-235. But this comes along with the capture of neutrons by the U-238 faction and results in a build-up of plutonium isotopes and minor actinides as neptunium, americium and curium. These actinides are dominant for the long time assessment of the radiological risk of a final disposal therefore a minimization of the long living isotopes is aspired. Burning the actinides in a high temperature helium cooled graphite moderated reactor (HTR) is one of these options. The use of plutonium isotopes to sustain the criticality of the system is intended to avoid on the one hand highly enriched uranium because of international regulations and on the other hand low enriched uranium because of the build up of new actinides from neutron capture in the U-238 fraction. Because initial minor actinide isotopes are typically not fissionable by thermal neutrons the idea is to fission instead the intermediate isotopes generated by the first neutron capture. This paper comprises calculations for plutonium/minor actinides/thorium fuel compositions and their correlated final burn-up for a generic pebble bed HTR based on the reference design of the 400 MW PBMR. In particular the cross sections and the neutron balance of the different minor actinide isotopes in the higher thermal energy spectrum of a HTR will be discussed. For a fuel mixture of plutonium and minor actinides a significant burn-up of these actinides up to 20% can be achieved but at the expense of a higher residual fraction of plutonium in the burned fuel. Combining

Multi-shape pulse pile-up correction: The MCPPU code

International Nuclear Information System (INIS)

Sabbatucci, Lorenzo; Scot, Viviana; Fernandez, Jorge E.

2014-01-01

In spectroscopic measurements with high counting rate, pulse pile-up (PPU) is a common distortion of the spectrum. It is fully ascribable to the pulse handling circuitry of the detector and it is not comprised in the detector response function which is well explained by a purely physical model. Since PPU occurs after the transport inside the detector, this is the first correction to perform in case of spectrum unfolding. Many producers include electronic rejection circuits to limit the appearance of PPU, but it is never suppressed completely. Therefore, it is always necessary to correct PPU distortions after the measurement. In the present work, it is described the post-processing tool MCPPU (Monte Carlo Pulse Pile-Up), based on the MC algorithm developed by Guo et al. (2004, 2005). MCPPU automatically determines the dead time of the counting system and corrects for PPU effects even in the presence of electronic suppression. The capability of allowing a user defined pulse shape makes the code suitable to be used with any kind of detector. The features of MCPPU are illustrated with some examples. - Highlights: • Pulse pile-up (PPU) is a common distortion in radiation detection. • MCPPU is a Monte Carlo code to perform post-processing PPU correction. • MCPPU evaluates automatically the dead time to use in the pile-up recovery. • The measured pulse shape can be introduced as a normalized discrete distribution. • MCPPU is compatible with detectors using electronic rejection circuitry

Burn-up measurements of LEU fuel for short cooling times

International Nuclear Information System (INIS)

Pereda B, C.; Henriquez A, C.; Klein D, J.; Medel R, J.

2005-01-01

The measurements presented in this work were made essentially at in-pool gamma-spectrometric facility, installed inside of the secondary pool of the RECH-1 research reactor, where the measured fuel elements are under 2 meters of water. The main reason for using the in-pool facility was because of its capability to measure the burning of fuel elements without having to wait so long, that is with only 5 cooling days, which are the usual times between reactor operations. Regarding these short cooling times, this work confirms again the possibility of using the 95 Zr as a promising burnup monitor, in spite of the rough approximations used to do it. These results are statistically reasonable within the range calculated using codes. The work corroborates previous results, presented in Santiago de Chile, and it suggests future improvements in that way. (author)

Development of assessment technology for hydrogen burn and fission product behavior in containment

International Nuclear Information System (INIS)

Kim, S. B.; Kim, J. T.; Ha, K. S.; Hong, S. W.; Song, Y. M.; Park, J. H.; Cho, Y. R.; Kang, H. S.

2012-04-01

Analysis tools for hydrogen burn was established to resolve the hydrogen issues in containment. To validate CFX commercial CFD(computational fluid dynamics) code, the hydrogen combustion experiments such as FLAME and ENACEFF for reactor containment were analyzed. And OpenFOAM hydrogen combustion code was developed and validated. Experiments for the flame propagation characteristics in IRWST and the run-up-distance for DDT(Deflagration to detonation transition) were performed and analytical model was evaluated to evaluation of the performance of hydrogen mitigation system, that is, PAR(Passive auto-catalistic re-combiner) To improvement of the fission product modelling in containment, separate analysis module for Iodine behavior and its application tool of K-IODIP (Korea IODIne Package) were developed. PHEBUS FPT-3 analysis was performed to validate MELCOR code. And also the characteristics of fission product behaviors in Future Reactors(GEN-IV) were compared

A validated methodology for evaluating burn-up credit in spent fuel casks

International Nuclear Information System (INIS)

Brady, M.C.; Sanders, T.L.

1992-01-01

The concept of allowing reactivity credit for the transmuted state of spent fuel offers both economic and risk incentives. This paper presents a general overview of the technical work being performed in support of the US Department of Energy (USDOE) programme to resolve issues related to the implementation of burn-up credit in spent fuel cask design. An analysis methodology is presented along with information representing the validation of the method against available experimental data. The experimental data that are applicable to burn-up credit include chemical assay data for the validation of the isotopic prediction models, fresh fuel critical experiments for the validation of criticality calculations for various cask geometries, and reactor re-start critical data to validate criticality calculations with spent fuel. The methodology has been specifically developed to be simple and generally applicable, therefore giving rise to uncertainties or sensitivities which are identified and quantified in terms of a percent bias effective multiplication (k eff ). Implementation issues affecting licensing requirements and operational procedures are discussed briefly. (Author)

Development of dynamic simulation code for fuel cycle of fusion reactor

International Nuclear Information System (INIS)

Aoki, Isao; Seki, Yasushi; Sasaki, Makoto; Shintani, Kiyonori; Kim, Yeong-Chan

1999-02-01

A dynamic simulation code for fuel cycle of a fusion experimental reactor has been developed. The code follows the fuel inventory change with time in the plasma chamber and the fuel cycle system during 2 days pulse operation cycles. The time dependence of the fuel inventory distribution is evaluated considering the fuel burn and exhaust in the plasma chamber, purification and supply functions. For each subsystem of the plasma chamber and the fuel cycle system, the fuel inventory equation is written based on the equation of state considering the fuel burn and the function of exhaust, purification, and supply. The processing constants of subsystem for steady states were taken from the values in the ITER Conceptual Design Activity (CDA) report. Using this code, the time dependence of the fuel supply and inventory depending on the burn state and subsystem processing functions are shown. (author)

A study of the effects of changing burn-up and gap gaseous compound on the gap convection coefficient (in a hot fuel pin) in VVER-1000 reactor

International Nuclear Information System (INIS)

Rahgoshay, M.; Rahmani, Y.

2007-01-01

In this article we worked on the result and process of calculation of the gap heat transfer coefficient for a hot fuel pin in accordance with burn-up changes in the VVER-1000 reactor at the Bushehr nuclear power plant (Iran). With regard to the fact that in calculating the fuel gap heat transfer coefficient, various parameters are effective and the need for designing a model is being felt, therefore, in this article we used Ross and Stoute gap model to study impacts of different effective parameters such as thermal expansion and gaseous fission products on the h gap change rate. Over time and with changes in fuel burn-up some gaseous fission products such as xenon, argon and krypton gases are released to the gas mixture in the gap, which originally contained helium. In this study, the composition of gaseous elements in the gap volume during different times of reactor operation was found using ORIGEN code. Considering that the thermal conduction of these gases is lower than that of helium, and by using the Ross and Stoute gap model, we find first that the changes in gaseous compounds in the gap reduce the values of gap thermal conductivity coefficient, but considering thermal expansion (due to burn-up alterations) of fuel and clad resulting in the reduction of gap thickness we find that the gap heat transfer coefficient will augment in a broad range of burn-up changes. These changes result in a higher rate of gap thickness reduction than the low rate of decrease of heat conduction coefficient of the gas in the gap during burn-up. Once these changes have been defined, we can proceed with the analysis of the results of calculations based on the Ross and Stoute model and compare the results obtained with the experimental results for a hot fuel pin as presented in the final safety analysis report of the VVER-1000 reactor at Bushehr. It is noteworthy that the results of accomplished calculations based on the Ross and Stoute model correspond well with the existing

Test of calorimetry for high burn-up plutonium

International Nuclear Information System (INIS)

Beets, C.; Carchon, R.; Fettweis, P.

1984-01-01

In recent times, the interest of applying calorimetry for safeguards purpose is steadily increasing. Calorimetric measurements have been performed on a set of high burn-up (25000 MWd/t) Pu samples, ranging in mass between 60 g and 2.5 kg Pu, distributed as PuO 2 powder embedded in stainless steel containers. The powers produced by these containers ranged between 0.8 W and 36 W. The calorimeter used was the Mound 150 type, and the isotopics and the Am content have been determined earlier by mass spectroscopy, completed with α and γ counting, and were later verified by the same methods. Watts/gram measurements were made on twelve 60 g samples of the same plutonium lot to demonstrate the Pu elemental and isotopic homogeneity, and hence, its suitability for subsequent NDA experiments. These samples were also measured in a stacked way to fill up the mass and wattage gaps between 60 g (0.8W) and 1 kg (14W). Calorimetric assay values, obtained with both isotopic measurements are discussed

Application of Integral Ex-Core and Differential In-Core Neutron Measurements for Adjustment of Fuel Burn-Up Distributions in VVER-1000

Science.gov (United States)

Borodkin, Pavel G.; Borodkin, Gennady I.; Khrennikov, Nikolay N.

2010-10-01

The paper deals with calculational 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. Time-integrated neutron source distributions used for DORT calculations were prepared by two different approaches based on a) calculated fuel burn-up (standard routine procedure) and b) in-core measurements by means of SPD & TC (new approach). Taking into account that fuel burn-up distributions in operating VVER may be evaluated now by analytical methods (calculations) only it is needed to develop new approaches for testing and correction of calculational evaluations. Results presented in this paper allow to consider a reverse task of alternative estimation of fuel burn-up distributions. The approach proposed is based on adjustment (fitting) of time-integrated neutron source distributions, and hence fuel burn-up patterns in some part of reactor core, on the base of ex-core neutron leakage measurement, neutron-physical calculation and in-core SPD & TC measurement data.

Tokamak Systems Code

International Nuclear Information System (INIS)

Reid, R.L.; Barrett, R.J.; Brown, T.G.

1985-03-01

The FEDC Tokamak Systems Code calculates tokamak performance, cost, and configuration as a function of plasma engineering parameters. This version of the code models experimental tokamaks. It does not currently consider tokamak configurations that generate electrical power or incorporate breeding blankets. The code has a modular (or subroutine) structure to allow independent modeling for each major tokamak component or system. A primary benefit of modularization is that a component module may be updated without disturbing the remainder of the systems code as long as the imput to or output from the module remains unchanged

Development of a numerical experimentation method for thermal hydraulics design and evaluation of high burn-up and innovative fuel pins

International Nuclear Information System (INIS)

Ninokata, Hisashi; Misawa, Takeharu; Baglietto, Emilio; Sorokin, A.P.; Maekawa, Isamu; Ohshima, Hiroyuki; Yamaguchi, Akira

2003-03-01

A method of large scale direct numerical simulation of turbulent flows in a high burn-up fuel pin bundle is proposed to evaluate wall shear stress and temperature distributions on the pin surfaces as well as detailed coolant velocity and temperature distributions inside subchannels under various thermal hydraulic conditions. This simulation is aimed at providing a tool to confirm margins to thermal hydraulics design limits of the nuclear fuels and at the same time to be used in design-by-analysis approaches. The method will facilitate thermal hydraulic design of high performance LMFR core fuels characterized by high burn-up, ultra long life, high reliable and safe performances, easiness of operation and maintenance, minimization of radio active wastes, without much relying on such empirical approach as hot spot factor and sub-factors, and above all the high cost mock up experiments. A pseudo direct numerical simulation of turbulence (DNS) code is developed, first on the Cartesian coordinates and then on the curvilinear boundary fit coordinates that enables us to reproduce thermal hydraulics phenomena in such a complicated flow channel as subchannels in a nuclear fuel pin assembly. The coordinate transformation is evaluated and demonstrated to yield correct physical quantities by carrying out computations and comparisons with experimental data with respect to the distributions of various physical quantities and turbulence statistics for fluid flow and heat transfers in various kinds of simple flow channel geometry. Then the boundary fitted pseudo DNS for flows inside an infinite pin array configuration is carried out and compared with available detailed experimental data. In parallel similar calculations are carried out using a commercial code STAR-CD to cross-check the DNS performances. As a results, the pseudo DNS showed reasonable comparisons with experiments as well as the STAR-CD results. Importance of the secondary flow influences is emphasized on the momentum

Development of a coupled code system based on system transient code, RETRAN, and 3-D neutronics code, MASTER

International Nuclear Information System (INIS)

Kim, K. D.; Jung, J. J.; Lee, S. W.; Cho, B. O.; Ji, S. K.; Kim, Y. H.; Seong, C. K.

2002-01-01

A coupled code system of RETRAN/MASTER has been developed for best-estimate simulations of interactions between reactor core neutron kinetics and plant thermal-hydraulics by incorporation of a 3-D reactor core kinetics analysis code, MASTER into system transient code, RETRAN. The soundness of the consolidated code system is confirmed by simulating the MSLB benchmark problem developed to verify the performance of a coupled kinetics and system transient codes by OECD/NEA

Multi-Accuracy-Level Burning Plasma Simulations

International Nuclear Information System (INIS)

Artaud, J. F.; Basiuk, V.; Garcia, J.; Giruzzi, G.; Huynh, P.; Huysmans, G.; Imbeaux, F.; Johner, J.; Scheider, M.

2007-01-01

The design of a reactor grade tokamak is based on a hierarchy of tools. We present here three codes that are presently used for the simulations of burning plasmas. At the first level there is a 0-dimensional code that allows to choose a reasonable range of global parameters; in our case the HELIOS code was used for this task. For the second level we have developed a mixed 0-D / 1-D code called METIS that allows to study the main properties of a burning plasma, including profiles and all heat and current sources, but always under the constraint of energy and other empirical scaling laws. METIS is a fast code that permits to perform a large number of runs (a run takes about one minute) and design the main features of a scenario, or validate the results of the 0-D code on a full time evolution. At the top level, we used the full 1D1/2 suite of codes CRONOS that gives access to a detailed study of the plasma profiles evolution. CRONOS can use a variety of modules for source terms and transport coefficients computation with different level of complexity and accuracy: from simple estimators to highly sophisticated physics calculations. Thus it is possible to vary the accuracy of burning plasma simulations, as a trade-off with computation time. A wide range of scenario studies can thus be made with CRONOS and then validated with post-processing tools like MHD stability analysis. We will present in this paper results of this multi-level analysis applied to the ITER hybrid scenario. This specific example will illustrate the importance of having several tools for the study of burning plasma scenarios, especially in a domain that present devices cannot access experimentally. (Author)

Reactivity loss validation of high burn-up PWR fuels with pile-oscillation experiments in MINERVE

Energy Technology Data Exchange (ETDEWEB)

Leconte, P.; Vaglio-Gaudard, C.; Eschbach, R.; Di-Salvo, J.; Antony, M.; Pepino, A. [CEA, DEN, DER, Cadarache, F-13108 Saint-Paul-Lez-Durance (France)

2012-07-01

The ALIX experimental program relies on the experimental validation of the spent fuel inventory, by chemical analysis of samples irradiated in a PWR between 5 and 7 cycles, and also on the experimental validation of the spent fuel reactivity loss with bum-up, obtained by pile-oscillation measurements in the MINERVE reactor. These latter experiments provide an overall validation of both the fuel inventory and of the nuclear data responsible for the reactivity loss. This program offers also unique experimental data for fuels with a burn-up reaching 85 GWd/t, as spent fuels in French PWRs never exceeds 70 GWd/t up to now. The analysis of these experiments is done in two steps with the APOLLO2/SHEM-MOC/CEA2005v4 package. In the first one, the fuel inventory of each sample is obtained by assembly calculations. The calculation route consists in the self-shielding of cross sections on the 281 energy group SHEM mesh, followed by the flux calculation by the Method Of Characteristics in a 2D-exact heterogeneous geometry of the assembly, and finally a depletion calculation by an iterative resolution of the Bateman equations. In the second step, the fuel inventory is used in the analysis of pile-oscillation experiments in which the reactivity of the ALIX spent fuel samples is compared to the reactivity of fresh fuel samples. The comparison between Experiment and Calculation shows satisfactory results with the JEFF3.1.1 library which predicts the reactivity loss within 2% for burn-up of {approx}75 GWd/t and within 4% for burn-up of {approx}85 GWd/t. (authors)

Burn-up credit criticality safety benchmark phase VII - UO2 fuel: study of spent fuel compositions for long-term disposal

International Nuclear Information System (INIS)

2012-01-01

After spent nuclear fuel (SNF) is discharged from a nuclear reactor, fuel composition and reactivity continue to vary as a function of time due to the decay of unstable nuclides. Accurate predictions of the concentrations of long-lived radionuclides in SNF, which represent a significant potential hazard to human beings and to the environment over a very long period, are particularly necessary for radiological dose assessments. This report assesses the ability of existing computer codes and associated nuclear data to predict isotopic compositions and their corresponding neutron multiplication factor (k eff ) values for pressurised-water-reactor (PWR) UO 2 fuel at 50 GWd/MTU burn-up in a generic spent fuel cask configuration. Fuel decay compositions and k eff values have been calculated for 30 post-irradiation time steps out to one million years

CONDOR: neutronic code for fuel elements calculation with rods

International Nuclear Information System (INIS)

Villarino, E.A.

1990-01-01

CONDOR neutronic code is used for the calculation of fuel elements formed by fuel rods. The method employed to obtain the neutronic flux is that of collision probabilities in a multigroup scheme on two-dimensional geometry. This code utilizes new calculation algorithms and normalization of such collision probabilities. Burn-up calculations can be made before the alternative of applying variational methods for response flux calculations or those corresponding to collision normalization. (Author) [es

The relevance of axial burn-up profiles for the criticality safety analysis of spent nuclear fuel in a final repository

International Nuclear Information System (INIS)

Kilger, R.; Gmal, B.; Moser, E.F.

2008-01-01

Due to inhomogeneous neutron flux and moderator density distributions in the reactor core, the burn-up of a nuclear fuel assembly is not homogeneous but shows an axial distribution, typically with lower partial burn-up and thus higher remaining reactivity at the fuel ends in particular at the assembly top end. Beyond a burn-up of about 15 to 20 GWd/tHM, the multiplication factor K of the whole assembly is dominated by this lower-burnt end regions, and is usually higher than for assuming a homogeneous uniform distribution of the averaged burn-up. This behaviour commonly referred to as positive ''end effect'' is well known in burn-up credit considerations for transportation and storage casks and is being investigated also in the context of criticality analyses for final disposition of spent nuclear fuel. Sign and value of the end effect depend on several parameters. Based on a generic model one may not conclude that criticality in a final repository is a likely or expected event, but nevertheless it draws the attention to the fact that criticality is not excluded per se but has to be considered in the analysis and probably has to be encountered by certain appropriate measures, maybe e.g. by limitation of the amount of fissile material inside one single cask, or a rigorous prove for prevention of water ingress. The authors also conclude that the higher partial reactivity of the fuel ends has to be accounted for carefully in more realistic analyses of post-closure scenarios with respect to criticality safety.

CONTAIN code calculations of the effects on the source term of CsI to I/sub 2/ conversion due to severe hydrogen burns

International Nuclear Information System (INIS)

Valdez, G.D.; Williams, D.C.

1986-01-01

In experiments conducted at Sandia National Laboratories large amounts of elemental iodine were produced when CsI-Al 2 O 3 aerosol was exposed to hydrogen/air combustion. To evaluate some of the implications of the iodide conversion (observed to occur with up to 75% efficiency) for the severe accident source term, computational simulations of representative accident sequences were conducted with the CONTAIN code. The following conclusions can be drawn from this preliminary source term assessment: (1) If the containment sprays are inoperative during the accident, or failed by the hydrogen burn, the late-time source term is almost tripled when the iodide is converted to I 2 . (2) With the sprays active, the amount released without conversion of the CsI aerosol is 63% higher than for the case when conversion occurs. (3) For the case where CsI is converted to I 2 continued operation of the sprays reduces the release by a factor of 40, relative to the case in which the sprays fail at the time of the hydrogen burn. When there is no conversion, the reduction factor for continued spray operation is about a factor of 9, relative to the failed spray case

A computer code for calculation of radioactive nuclide generation and depletion, decay heat and γ ray spectrum. FPGS90

International Nuclear Information System (INIS)

Ihara, Hitoshi; Katakura, Jun-ichi; Nakagawa, Tsuneo

1995-11-01

In a nuclear reactor radioactive nuclides are generated and depleted with burning up of nuclear fuel. The radioactive nuclides, emitting γ ray and β ray, play role of radioactive source of decay heat in a reactor and radiation exposure. In safety evaluation of nuclear reactor and nuclear fuel cycle, it is needed to estimate the number of nuclides generated in nuclear fuel under various burn-up condition of many kinds of nuclear fuel used in a nuclear reactor. FPGS90 is a code calculating the number of nuclides, decay heat and spectrum of emitted γ ray from fission products produced in a nuclear fuel under the various kinds of burn-up condition. The nuclear data library used in FPGS90 code is the library 'JNDC Nuclear Data Library of Fission Products - second version -', which is compiled by working group of Japanese Nuclear Data Committee for evaluating decay heat in a reactor. The code has a function of processing a so-called evaluated nuclear data file such as ENDF/B, JENDL, ENSDF and so on. It also has a function of making figures of calculated results. Using FPGS90 code it is possible to do all works from making library, calculating nuclide generation and decay heat through making figures of the calculated results. (author)

A computer code for calculation of radioactive nuclide generation and depletion, decay heat and {gamma} ray spectrum. FPGS90

Energy Technology Data Exchange (ETDEWEB)

Ihara, Hitoshi; Katakura, Jun-ichi; Nakagawa, Tsuneo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1995-11-01

In a nuclear reactor radioactive nuclides are generated and depleted with burning up of nuclear fuel. The radioactive nuclides, emitting {gamma} ray and {beta} ray, play role of radioactive source of decay heat in a reactor and radiation exposure. In safety evaluation of nuclear reactor and nuclear fuel cycle, it is needed to estimate the number of nuclides generated in nuclear fuel under various burn-up condition of many kinds of nuclear fuel used in a nuclear reactor. FPGS90 is a code calculating the number of nuclides, decay heat and spectrum of emitted {gamma} ray from fission products produced in a nuclear fuel under the various kinds of burn-up condition. The nuclear data library used in FPGS90 code is the library `JNDC Nuclear Data Library of Fission Products - second version -`, which is compiled by working group of Japanese Nuclear Data Committee for evaluating decay heat in a reactor. The code has a function of processing a so-called evaluated nuclear data file such as ENDF/B, JENDL, ENSDF and so on. It also has a function of making figures of calculated results. Using FPGS90 code it is possible to do all works from making library, calculating nuclide generation and decay heat through making figures of the calculated results. (author).

Development of code SFINEL (Spent fuel integrity evaluator)

Energy Technology Data Exchange (ETDEWEB)

Kim, Yong Soo; Min, Chin Young; Ohk, Young Kil; Yang, Yong Sik; Kim, Dong Ju; Kim, Nam Ku [Hanyang University, Seoul (Korea)

1999-01-01

SFINEL code, an integrated computer program for predicting the spent fuel rod integrity based on burn-up history and major degradation mechanisms, has been developed through this project. This code can sufficiently simulate the power history of a fuel rod during the reactor operation and estimate the degree of deterioration of spent fuel cladding using the recently-developed models on the degradation mechanisms. SFINEL code has been thoroughly benchmarked against the collected in-pile data and operating experiences: deformation and rupture, and cladding oxidation, rod internal pressure creep, then comprehensive whole degradation process. (author). 75 refs., 51 figs., 5 tabs.

High performance mixed optical CDMA system using ZCC code and multiband OFDM

Directory of Open Access Journals (Sweden)

Nawawi N. M.

2017-01-01

Full Text Available In this paper, we have proposed a high performance network design, which is based on mixed optical Code Division Multiple Access (CDMA system using Zero Cross Correlation (ZCC code and multiband Orthogonal Frequency Division Multiplexing (OFDM called catenated OFDM. In addition, we also investigate the related changing parameters such as; effective power, number of user, number of band, code length and code weight. Then we theoretically analyzed the system performance comprehensively while considering up to five OFDM bands. The feasibility of the proposed system architecture is verified via the numerical analysis. The research results demonstrated that our developed modulation solution can significantly enhanced the total number of user; improving up to 80% for five catenated bands compared to traditional optical CDMA system, with the code length equals to 80, transmitted at 622 Mbps. It is also demonstrated that the BER performance strongly depends on number of weight, especially with less number of users. As the number of weight increases, the BER performance is better.

High performance mixed optical CDMA system using ZCC code and multiband OFDM

Science.gov (United States)

Nawawi, N. M.; Anuar, M. S.; Junita, M. N.; Rashidi, C. B. M.

2017-11-01

In this paper, we have proposed a high performance network design, which is based on mixed optical Code Division Multiple Access (CDMA) system using Zero Cross Correlation (ZCC) code and multiband Orthogonal Frequency Division Multiplexing (OFDM) called catenated OFDM. In addition, we also investigate the related changing parameters such as; effective power, number of user, number of band, code length and code weight. Then we theoretically analyzed the system performance comprehensively while considering up to five OFDM bands. The feasibility of the proposed system architecture is verified via the numerical analysis. The research results demonstrated that our developed modulation solution can significantly enhanced the total number of user; improving up to 80% for five catenated bands compared to traditional optical CDMA system, with the code length equals to 80, transmitted at 622 Mbps. It is also demonstrated that the BER performance strongly depends on number of weight, especially with less number of users. As the number of weight increases, the BER performance is better.

The cluster burn up programme CCC and a comparison of its results with NPD experiments

International Nuclear Information System (INIS)

Hoejerup, C.F.

1976-10-01

A brief description is given of the computer programme CCC, which can be used for rod/rod cluster burn up calculations. A comparison of CCC results with some Canadian measurements on NPD fuel is also included. (author)

Up-date of the BCG code library

International Nuclear Information System (INIS)

Caldeira, A.D.; Garcia, R.D.M.

1990-01-01

Procedures for generating an up-date material library for the BCG code were established. A new library was generated by processing ENDF/B-IV data with the 89-1 version of the LINEAR, RECENT and SIGMA1 programs. The effect of library change in the neutron spectrum and effective multiplication factor of a fast reactor cell was analized. During the course of this study, an error was detected in the BCG code. Although localized in a narrow energy range, the discrepancies in neutron spectrum caused by the error were large enough to yield a difference of about 1% in the effective multiplication factor of the test cell. (author)

Performance of high burned PWR fuel during transient

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki; Fujishiro, Toshio

1992-01-01

In a majority of Japanese light water type commercial powder reactors (LWRs), UO 2 pellet sheathed by zircaloy cladding is used. Licensed discharged burn-up of the PWR fuel rod is going to be increased from 39 MWd/kgU to 48 MWd/kgU. This requests the increased reliability of cladding material as a strong barrier against fission product (FP). A long time usage in the neutron field and in the high temperature coolant will cause the zircaloy hardening and embrittlement. The cladding material is also degraded by waterside corrosion. These degradations are enhanced much by increased burn-up. A increased magnitude of the pellet-cladding mechanical interaction (PCMI) is of importance for increasing the stress of cladding material. In addition, aggressive FPs released from the fuel tends to attack the cladding material to cause stress corrosion cracking (SCC). At the Nuclear Safety Research Reactor (NSRR) in JAERI, 14 x 14 PWR type fuel rods preirradiation up to 42 MWd/kgU was prepared for the transient pulse irradiation under the simulated reactivity initiated accident (RIA) conditions. This will cause a prompt increase of the fuel temperature and stress on the highly burned cladding material. In the present paper, steady-state and transient behavior observed from the tested PWR fuel rod and calculational results obtained from the computer code FPRETAIN will be described. (author)

Comparison of MCB and MONTEBURNS Monte Carlo burnup codes on a one-pass deep burn

International Nuclear Information System (INIS)

Talamo, Alberto; Ji, Wei; Cetnar, Jerzy; Gudowski, Waclaw

2006-01-01

Numerical applications implemented on the Monte Carlo method have developed in line with the increase of computer power; nowadays, in the field of nuclear reactor physics, it is possible to perform burnup simulations in a detailed 3D geometry and a continuous energy description by the Monte Carlo method; moreover, the required computing time can be abundantly reduced by taking advantage of a computer cluster. In this paper we focused on comparing the results of the two major Monte Carlo burnup codes, MONTEBURNS and MCB, when they share the same MCNP geometry, nuclear data library, core thermal power, and they apply the same refueling and shuffling schedule. While simulating a total operation time of the Gas Turbine-Modular Helium Reactor of 2100 effective full power days and a one-pass deep burn in-core fuel management schedule, we have found that the two Monte Carlo codes produce very similar results both on the criticality value of the core and the transmutation of the key actinides

Comparison of MCB and MONTEBURNS Monte Carlo burnup codes on a one-pass deep burn

Energy Technology Data Exchange (ETDEWEB)

Talamo, Alberto [Royal Institute of Technology (KTH), Roslagstullsbacken 21, Stockholm S-10691 (Sweden)]. E-mail: alby@anl.gov; Ji, Wei [University of Michigan, Bonisteel Boulevard 2355, Ann Arbor, MI 48109-2104 (United States); Cetnar, Jerzy [AGH-University of Science and Technology, Al. Mickiewicza 30 Cracow (Poland); Gudowski, Waclaw [Royal Institute of Technology (KTH), Roslagstullsbacken 21, Stockholm S-10691 (Sweden)

2006-09-15

Numerical applications implemented on the Monte Carlo method have developed in line with the increase of computer power; nowadays, in the field of nuclear reactor physics, it is possible to perform burnup simulations in a detailed 3D geometry and a continuous energy description by the Monte Carlo method; moreover, the required computing time can be abundantly reduced by taking advantage of a computer cluster. In this paper we focused on comparing the results of the two major Monte Carlo burnup codes, MONTEBURNS and MCB, when they share the same MCNP geometry, nuclear data library, core thermal power, and they apply the same refueling and shuffling schedule. While simulating a total operation time of the Gas Turbine-Modular Helium Reactor of 2100 effective full power days and a one-pass deep burn in-core fuel management schedule, we have found that the two Monte Carlo codes produce very similar results both on the criticality value of the core and the transmutation of the key actinides.

Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes

International Nuclear Information System (INIS)

Baratta, A.J.

1997-01-01

To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts and engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together

Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes

Energy Technology Data Exchange (ETDEWEB)

Baratta, A.J.

1997-07-01

To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts and engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together.

A core management system for JRR-3

International Nuclear Information System (INIS)

Soyama, Kazuhiko; Tsuruta, Harumichi; Ichikawa, Hiroki; Nemoto, Hiroyuki.

1991-05-01

Japan Research Reactor No.3 (JRR-3) was upgraded to the thermal output with 20 MW by replacing the core, cooling system and utilization facilities. It is a water moderated and cooled, pool type reactor using 20% enriched U · Alx fuel. A core management system for JRR-3 has been made. This code system can manage of reactivity, power distribution and burn up in consideration of the position of control rod, fuel arrangement and operation pattern. This report is the user's manual of this code system. (author)

Application Of WIMS Code To Calculation Kartini Reactor Parameters By Pin-Cell And Cluster Method

International Nuclear Information System (INIS)

Sumarsono, Bambang; Tjiptono, T.W.

1996-01-01

Analysis UZrH fuel element parameters calculation in Kartini Reactor by WIMS Code has been done. The analysis is done by pin cell and cluster method. The pin cell method is done as a function percent burn-up and by 8 group 3 region analysis and cluster method by 8 group 12 region analysis. From analysis and calculation resulted K ∼ = 1.3687 by pin cell method and K ∼ = 1.3162 by cluster method and so deviation is 3.83%. By pin cell analysis as a function percent burn-up at the percent burn-up greater than 59.50%, the multiplication factor is less than one (k ∼ < 1) it is mean that the fuel element reactivity is negative

Modelling of pore coarsening in the high burn-up structure of UO{sub 2} fuel

Energy Technology Data Exchange (ETDEWEB)

Veshchunov, M.S.; Tarasov, V.I., E-mail: tarasov@ibrae.ac.ru

2017-05-15

The model for coalescence of randomly distributed immobile pores owing to their growth and impingement, applied by the authors earlier to consideration of the porosity evolution in the high burn-up structure (HBS) at the UO{sub 2} fuel pellet periphery (rim zone), was further developed and validated. Predictions of the original model, taking into consideration only binary impingements of growing immobile pores, qualitatively correctly describe the decrease of the pore number density with the increase of the fractional porosity, however notably underestimate the coalescence rate at high burn-ups attained in the outmost region of the rim zone. In order to overcome this discrepancy, the next approximation of the model taking into consideration triple impingements of growing pores was developed. The advanced model provides a reasonable consent with experimental data, thus demonstrating the validity of the proposed pore coarsening mechanism in the HBS.

Burn-up determination of irradiated uranium oxide by means of direct gama spectrometry and by radiochemical method

International Nuclear Information System (INIS)

Cunha, I.I.L.; Nastasi, M.J.C.; Lima, F.W.

1981-09-01

The burn-up of thermal neutrons irradiated U 3 O 8 (natural uranium) samples has been determined by using both direct gamma spectrometry and radiochemical methods and the results obtained were compared. The fission products 144 Ce, 103 Ru, 106 Ru, 137 Cs and 95 Zr were chosen as burn-up monitors. In order to isolate the radioisotopes chosen as monitors, a radiochemical separation procedure has been established, in which the solvent extraction technique was used to separate cerium, cesium and ruthenium one from the other and all of them from uranium. The separation between zirconium and niobium and of both elements from the other radioisotopes and uranium was accomplished by means of adsorption on a silica-gel column, followed by selective elution of zirconium and of niobium. When use was made of the direct gamma-ray spectrometry method, the radioactivity of each nuclide of interest was measured in presence of all others. For this purpose use was made of gamma-ray spectrometry and of a Ge-Li detector. Comparison of burn-up values obtained by both methods was made by means of Student's 't' test, and this showed that results obtained in each case are statistically equal. (Author) [pt

Development of dynamic simulation code for fuel cycle of fusion reactor. 1. Single pulse operation simulation

Energy Technology Data Exchange (ETDEWEB)

Aoki, Isao; Seki, Yasushi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Sasaki, Makoto; Shintani, Kiyonori; Kim, Yeong-Chan

1997-11-01

A dynamic simulation code for the fuel cycle of a fusion experimental reactor has been developed. The code follows the fuel inventory change with time in the plasma chamber and the fuel cycle system during a single pulse operation. The time dependence of the fuel inventory distribution is evaluated considering the fuel burn and exhaust in the plasma chamber, purification and supply functions. For each subsystem of the plasma chamber and the fuel cycle system, the fuel inventory equation is written based on the equation of state considering the function of fuel burn, exhaust, purification, and supply. The processing constants of subsystem for the steady states were taken from the values in the ITER Conceptual Design Activity (CDA) report. Using the code, the time dependence of the fuel supply and inventory depending on the burn state and subsystem processing functions are shown. (author)

[Epidemiological changes in burned children. A 10-year follow-up].

Science.gov (United States)

Rojas Goldsack, María de Los Ángeles; Saavedra Opazo, Rolando; Vicencio Pezo, Paulina; Solís Flores, Fresia

2016-01-01

The aim of the study was to compare the incidence and epidemiological characteristics of burns suffered by children in a district of Santiago of Chile over a period of ten years. An analytical study was conducted by checking through the medical files of children under 15 years of age from Pudahuel district who were admitted with burns to the Santiago Aid to Burned Children Corporation (COANIQUEM) during 2011. A comparison was made with the results obtained in a similar study performed in the same district in 2001. In 2011, 440 children were admitted, with an incidence rate of 700/100,000 <15 years old (95% CI: 635-765), a decrease of 25% compared to 2001(Incidence rate of 933/100,000; 95% CI: 856-1010). There were 52% males, 64.5% under 5 years old of age, 88% burned at home, or at other houses where they are been taking care of. There was a significant change in the causative agent, and included, increasing by their relative importance; hot objects (27.1%). The mechanism that mostly increased in occurrence were contact with stoves or heaters, and also emerge that caused by hair iron, and motorcycle exhaust. The most common location was the hand, increasing by 30.8%, and 66.4% showed an extension of the burn of <1% total body surface area (2001, 61%). A significant decline of 54% of deep burns was observed, and 23.2% were admitted to rehabilitation, a similar proportion to 2001. The rate of hospitalization and/or skin graft decreased from 104/100,000 to 62/100,000<15 years old (95% CI: 43-82). Burns incidence has decreased. Hot objects are now the main causal agent. The decrease in the rate of hospitalization and/or graft indicates a lower severity of burns. Copyright © 2015 Sociedad Chilena de Pediatría. Publicado por Elsevier España, S.L.U. All rights reserved.

Interpretation of the CABRI-RAFT LTX test up to pin failure based on detailed data evaluation and PARAS-2S code analysis

International Nuclear Information System (INIS)

Fukano, Yoshitaka; Sato, Ikken

2001-09-01

The CABRI-RAFT LTX test aims at a study on the fuel-pin-failure mechanism, in-pin fuel motion and post-failure fuel relocation with an annular fuel pin which was pre-irradiated up to peak burn-up of 6.4 at%. The transient test conditions similar to those of the LT4 test were selected in the LTX test using the same type of fuel pin, allowing an effective direct comparison between the two tests. In contrast to the LT4 test which showed a large PCMI-mitigation potential of the annular fuel-pin design, early pin failure occurred in the LTX test when fuel does not seem to have molten. In order to clarify the fuel pin failure mechanism, interpretation of the LTX test up to pin failure is performed in this study, through an experimental data evaluation and a PAPAS-2S-code analysis. The PAPAS-2S code simulates reasonably the fuel thermal conditions such as transient fuel-pin heat-up and fuel melting. The present detailed data evaluation shows that the earlier cladding failure compared with the LT4 test is mainly attributed to the local cladding heat-up. Under the high-temperature condition, plenum gas pressure has a certain potential to explain the observed failure. Fuel swelling-induced PCMI does not seem significant in the LTX test and it may have contributed to the early pin failure only to a limited extent, if any. (author)

Decoding linear error-correcting codes up to half the minimum distance with Gröbner bases

NARCIS (Netherlands)

Bulygin, S.; Pellikaan, G.R.; Sala, M.; Mora, T.; Perret, L.; Sakata, S.; Traverso, C.

2009-01-01

In this short note we show how one can decode linear error-correcting codes up to half the minimum distance via solving a system of polynomial equations over a finite field. We also explicitly present the reduced Gröbner basis for the system considered.

Burn Control Mechanisms in Tokamaks

Science.gov (United States)

Hill, M. A.; Stacey, W. M.

2015-11-01

Burn control and passive safety in accident scenarios will be an important design consideration in future tokamak reactors, in particular fusion-fission hybrid reactors, e.g. the Subcritical Advanced Burner Reactor. We are developing a burning plasma dynamics code to explore various aspects of burn control, with the intent to identify feedback mechanisms that would prevent power excursions. This code solves the coupled set of global density and temperature equations, using scaling relations from experimental fits. Predictions of densities and temperatures have been benchmarked against DIII-D data. We are examining several potential feedback mechanisms to limit power excursions: i) ion-orbit loss, ii) thermal instability density limits, iii) MHD instability limits, iv) the degradation of alpha-particle confinement, v) modifications to the radial current profile, vi) ``divertor choking'' and vii) Type 1 ELMs. Work supported by the US DOE under DE-FG02-00ER54538, DE-FC02-04ER54698.

Determination of the burn-up in fuels of the MTR type by means of gamma spectroscopy with crystal of INa(Tl)

International Nuclear Information System (INIS)

Kestelman, A.J.

1988-01-01

One of the responsibilities of the Laboratory of Analysis by Neutronic Activation of the RA-6 reactor is to determine the burn-up in fuels of the MTR type. In order to gain experience, up to the arrival of the hyperpure Germanium detector (HPGe) to be used in normal operation, preliminary measurements with a crystal of INa(Tl) were made. The fuel elements used are originated in the RA-3 reactor, with a decay superior to the thirteen years. For this reason, the unique visible photoelectric peak is the one of Cs-137, owing to the low resolution of the INa(Tl). After preliminary measurements, the profiles of burn-up, rectified by attenuation, were measured. Once the efficiency of the detector was determined, the calculation of the burn-up was made; for the element No. 144, a value of 21.6 ± 2.9 g was obtained to be compared with the value 21.9 g which was the evaluation made by the operators. (Author) [es

Insertion of control systems models in the Almod 3 computer code for the simulation of Angra I reactor start-up tests

International Nuclear Information System (INIS)

Camargo, C.T.M.

1981-09-01

The Almod 3 computer code was modified, aiming at the simulation of Angra I nuclear power plant behavior during some reactor start-up tests. The results obtained with the modified computer code (Almod 3W) are compared with those obtained with the Retran computer code. (E.G.) [pt

Demonstration test on the safety of a cell ventilation system during a hypothetical explosive burning in a fuel reprocessing plant

International Nuclear Information System (INIS)

Suzuki, Motoe; Nishio, Gunji; Takada, Junichi; Tsukamoto, Michio; Koike, Tadao

1993-01-01

To demonstrate the safety of an air ventilation system of cells in a fuel reprocessing plant under a postulated explosive burning caused by solvent fire or by thermal decomposition of nitrated solvent, four types of demonstration tests have been conducted using a large-scale facility simulating a cell ventilation system of an actual reprocessing plant, thus revealing effective mitigation by cell and duct structures on the pressure and temperature pulses generated by explosive burning. In boilover burning tests, solvent fire in a model cell was observed with various sizes of burning surface area as a main parameter, and analysis was performed on the factors dominating the magnitude of boilover burning, revealing that the magnitude strongly depends on accumulated amounts and their ratio of oxygen and solvent vapor present in the cell. In deflagration tests, solid rocket fuel was burned in the cell to simulate the explosive source. The generated pressure and temperature pulses were effectively declined by the cell and duct structures and the integrity of the ventilation system was kept. In blower tests, a centrifugal turbo blower was imposed by a lump of air with a larger flow rate than the rated one by about six times to observe the transient response of the blower fan and motor. It was found that integrity of the blower was kept. In pressure transient tests, compressed air was blown into the cell to induce a mild transient state of fluid dynamics inside the facility, and a variety of data were successfully obtained to be used for the verification and improvement of a computer code. In all the tests, transient overloading of gas caused no damage on HEPA filters, and overloading on the blower motor was avoided either by the slipping of transmission belt or by the acceleration of blower fan rotation during peak flow. (author)

Study on the thermal-hydraulic stability of high burn up STEP III fuel in Japan

International Nuclear Information System (INIS)

Ishikawa, M.; Kitamura, H.; Toba, A.; Omoto, A.

2004-01-01

Japanese BWR utilities have performed a joint study of the Thermal Hydraulic Stability of High Burn up STEP III Fuel. In this study, the parametric dependency of thermal hydraulic stability threshold was obtained. It was confirmed through experiments that the STEP III Fuel has sufficient stability characteristics. (author)

COMRAD96, Nuclear Fuel Burnup and Depletion Calculation System

International Nuclear Information System (INIS)

Suyama, K.; Masukawa, F.; Ido, M.; Enomoto, M.; Takyu, S.; Hara, T.

2002-01-01

1 - Description of program or function: Burn-up calculation of nuclear fuel. 2 - Methods: Matrix exponential method, Bateman Equation. 3 - Restrictions on the complexity of the problem: a) One-grouped cross section library should be prepared for the fuel system to be analyzed using UNITBURN. However, UNITBURN is not available now for UNIX systems. b) Gamma ray spectrometry calculation will fail using the attached piflib routine. This problem has already been rectified in the internal version. 4 - Typical running time: Two minutes for standard burn-up calculation on Sun ULTRA 30. 5 - Unusual features - a) Selection of Matrix exponential method, or Bateman Equation. b) JDDL, a detailed decay chain data based on ENSDF. 6 - Related or auxiliary programs: UNITBURN: Burnup calculation code unit cell system

Verification calculations for the WWER version of the TRANSURANUS code

International Nuclear Information System (INIS)

Elenkov, D.; Boneva, S.; Georgieva, M.; Georgiev, S.; Schubert, A.; Van Uffelen, P.

2006-01-01

The paper presents part of the work performed in the study project 'Research and Development for Licensing of Nuclear Fuel in Bulgaria'. The main objective of the project is to provide assistance for solving technical questions of the fuel licensing process in Bulgaria. One important issue is the extension of the predictive capabilities of fuel performance codes for Russian-type WWER reactors. In the last decade, a series of international projects has been based on the TRANSURANUS fuel performance code: Specific models for WWER fuel have been developed and implemented in the code in the late 90's. In 2000-2003, basic verification work was done by using experimental data of nuclear fuel irradiated in WWER-440 reactors. While the present paper focuses on the analysis of WWER-1000 standard fuel under normal operating conditions, the above study project covers additional tasks: 1) Post-irradiation calculations of ramp tests performed in the DR3 test reactor of the Risoe National Laboratory (five instrumented fuel rods of the Risoe 3 dataset contained in the IFPE database) using the TRANSURANUS code; 2) Compilation of cross-section libraries for isotope evolution calculations in WWER-440 and WWER-1000 fuel assemblies using the ORIGEN-S code; 3) Analysis of current situation and needs for an extension of the curriculum in Nuclear Engineering at the Technical University of Sofia. In this paper the post-irradiation calculations of steady-state irradiation experiments with nuclear fuel for Russian-type WWER-1000 reactors, using the latest release of the TRANSURANUS code (v1m1j03)are presented. Regarding a comprehensive verification of modern fuel performance codes, the burn-up region above 40 MWd/kgU is of increasing importance. A number of new phenomena emerge at high fuel burn-up, implying the need for enlarged databases of postirradiation examinations (PIE). For one fuel assembly irradiated in a WWER-1000 reactor with a rod discharge burn-up between 50 and 55 MWd

Calculation of pellet radial power distributions with a Monte Carlo burnup code

International Nuclear Information System (INIS)

Suzuki, Motomu; Yamamoto, Toru; Nakata, Tetsuo

2010-01-01

The Japan Nuclear Energy Safety Organization (JNES) has been working on an irradiation test program of high-burnup MOX fuel at Halden Boiling Water Reactor (HBWR). MOX and UO 2 fuel rods had been irradiated up to about 64 GWd/t (rod avg.) as a Japanese utilities research program (1st phase), and using those fuel rods, in-situ measurement of fuel pellet centerline temperature was done during the 2nd phase of irradiation as the JNES test program. As part of analysis of the temperature data, power distributions in a pellet radial direction were analyzed by using a Monte Carlo burnup code MVP-BURN. In addition, the calculated results of deterministic burnup codes SRAC and PLUTON for the same problem were compared with those of MVP-BURN to evaluate their accuracy. Burnup calculations with an assembly model were performed by using MVP-BURN and those with a pin cell model by using SRAC and PLUTON. The cell pitch and, therefore, fuel to moderator ratio in the pin cell calculation was determined from the comparison of neutron energy spectra with those of MVP-BURN. The fuel pellet radial distributions of burnup and fission reaction rates at the end of the 1st phase irradiation were compared between the three codes. The MVP-BURN calculation results show a large peaking in the burnup and fission rates in the pellet outer region for the UO 2 and MOX pellets. The SRAC calculations give very close results to those of the MVP-BURN. On the other hand, the PLUTON calculations show larger burnup for the UO 2 and lower burnup for the MOX pellets in the pellet outer region than those of MVP-BURN, which lead to larger fission rates for the UO 2 and lower fission rates for the MOX pellets, respectively. (author)

Approach to lithium burn-up effect in lithium ceramics

International Nuclear Information System (INIS)

Rasneur, B.

1994-01-01

The lithium burn-up in Li 2 ZrO 3 is simulated by removing lithium under Li 2 O form and trapping it in high specific surface area powder while heating during 15 days or 1 month at moderate temperature so that lithium mobility be large enough without causing any sintering neither of the specimens nor of the powder. In a first treatment at 775 deg C during 1 month. 30% of the lithium content could be removed inducing a lithium concentration gradient in the specimen and the formation of a lithium-free monoclinic ZrO 2 skin. Improvements led to similar results at 650 deg C and 600 deg C, the latter temperatures are closer to the operating temperature of the ceramic breeder blanket of a fusion reactor. (author) 4 refs.; 4 figs.; 1 tab

An integrated expert system for optimum in core fuel management

International Nuclear Information System (INIS)

Abd Elmoatty, Mona S.; Nagy, M.S.; Aly, Mohamed N.; Shaat, M.K.

2011-01-01

Highlights: → An integrated expert system constructed for optimum in core fuel management. → Brief discussion of the ESOIFM Package modules, inputs and outputs. → Package was applied on the DALAT Nuclear Research Reactor (0.5 MW). → The Package verification showed good agreement. - Abstract: An integrated expert system called Efficient and Safe Optimum In-core Fuel Management (ESOIFM Package) has been constructed to achieve an optimum in core fuel management and automate the process of data analysis. The Package combines the constructed mathematical models with the adopted artificial intelligence techniques. The paper gives a brief discussion of the ESOIFM Package modules, inputs and outputs. The Package was applied on the DALAT Nuclear Research Reactor (0.5 MW). Moreover, the data of DNRR have been used as a case study for testing and evaluation of ESOIFM Package. This paper shows the comparison between the ESOIFM Package burn-up results, the DNRR experimental burn-up data, and other DNRR Codes burn-up results. The results showed good agreement.

Burning nuclear wastes in fusion reactors

International Nuclear Information System (INIS)

Meldner, H.W.; Howard, W.M.

1979-01-01

A study was made up of actinide burn-up in ICF reactor pellets; i.e. 14 Mev neutron fission of the very long-lived actinides that pose storage problems. A major advantage of pellet fuel region burn-up is safety: only milligrams of highly toxic and active material need to be present in the fusion chamber, whereas blanket burn-up requires the continued presence of tons of actinides in a small volume. The actinide data tables required for Monte Carlo calculations of the burn-up of /sup 241/Am and /sup 243/Am are discussed in connection with a study of the sensitivity to cross section uncertainties. More accurate and complete cross sections are required for realistic quantitative calculations. 13 refs

Third Degree Skin Burns Caused by an MRI Compatible Electrocardiographic Monitoring System

DEFF Research Database (Denmark)

Brix, Lau; Isaksen, Christin Rosendahl Graff; Kristensen, Birgitte Hornbæk

of the assigned compatibility specifications of the leads due to the use of TFE sequences with high SAR values. MRI compatible monitoring systems are only safe when used with proper care. The presented burn cases may have been avoided if space had been provided between the ECG leads and the skin using a cloth....... This holds true even in cases in which the devices are MRI compatible and therefore safe in specified MRI environments. Of particular interest to this case report is skin burns caused by the ECG monitoring equipment. In this context, several cases of ECG electrode related burns have been reported, while...... burns caused by the ECG cables are less common [1]. This case report presents two unusual cases of skin burns which were caused by MRI safe ECG leads during scanning. Cases:Two patients suffered third degree burns using MRI approved ECG leads (Medrad® Veris MR Monitor system) in a Siemens Skyra 3...

SASSYS LMFBR systems code

International Nuclear Information System (INIS)

Dunn, F.E.; Prohammer, F.G.; Weber, D.P.

1983-01-01

The SASSYS LMFBR systems analysis code is being developed mainly to analyze the behavior of the shut-down heat-removal system and the consequences of failures in the system, although it is also capable of analyzing a wide range of transients, from mild operational transients through more severe transients leading to sodium boiling in the core and possible melting of clad and fuel. The code includes a detailed SAS4A multi-channel core treatment plus a general thermal-hydraulic treatment of the primary and intermediate heat-transport loops and the steam generators. The code can handle any LMFBR design, loop or pool, with an arbitrary arrangement of components. The code is fast running: usually faster than real time

Synthesis of intermetallic hydrides of Zr-Ni system in the burning regime

Energy Technology Data Exchange (ETDEWEB)

Akopyan, A.G.; Dolukhanyan, S.K.; Karapetyan, A.K.; Merzhanov, A.G.

1983-06-01

Conditions for production of intermetallides in the Zr-Ni system and their hydrides in the burning regime are studied. Burning regularities of Zr/sub 2/Ni and ZrNi intermetallides in hydrogen are studied, the burning mechanism is found. It is shown that burning proceeds at abnormally low temperatures. Optimum synthesis conditions for Zr/sub 2/NiH/sub 5/ and ZrNiH/sub 3/ hydrides are determined.

Coded diffraction system in X-ray crystallography using a boolean phase coded aperture approximation

Science.gov (United States)

Pinilla, Samuel; Poveda, Juan; Arguello, Henry

2018-03-01

Phase retrieval is a problem present in many applications such as optics, astronomical imaging, computational biology and X-ray crystallography. Recent work has shown that the phase can be better recovered when the acquisition architecture includes a coded aperture, which modulates the signal before diffraction, such that the underlying signal is recovered from coded diffraction patterns. Moreover, this type of modulation effect, before the diffraction operation, can be obtained using a phase coded aperture, just after the sample under study. However, a practical implementation of a phase coded aperture in an X-ray application is not feasible, because it is computationally modeled as a matrix with complex entries which requires changing the phase of the diffracted beams. In fact, changing the phase implies finding a material that allows to deviate the direction of an X-ray beam, which can considerably increase the implementation costs. Hence, this paper describes a low cost coded X-ray diffraction system based on block-unblock coded apertures that enables phase reconstruction. The proposed system approximates the phase coded aperture with a block-unblock coded aperture by using the detour-phase method. Moreover, the SAXS/WAXS X-ray crystallography software was used to simulate the diffraction patterns of a real crystal structure called Rhombic Dodecahedron. Additionally, several simulations were carried out to analyze the performance of block-unblock approximations in recovering the phase, using the simulated diffraction patterns. Furthermore, the quality of the reconstructions was measured in terms of the Peak Signal to Noise Ratio (PSNR). Results show that the performance of the block-unblock phase coded apertures approximation decreases at most 12.5% compared with the phase coded apertures. Moreover, the quality of the reconstructions using the boolean approximations is up to 2.5 dB of PSNR less with respect to the phase coded aperture reconstructions.

Fission gas release at high burn-up: beyond the standard diffusion model

International Nuclear Information System (INIS)

Landskron, H.; Sontheimer, F.; Billaux, M.R.

2002-01-01

At high burn-up standard diffusion models describing the release of fission gases from nuclear fuel must be extended to describe the experimental loss of xenon observed in the fuel matrix of the rim zone. Marked improvements of the prediction of integral fission gas release of fuel rods as well as of radial fission gas profiles in fuel pellets are achieved by using a saturation concept to describe fission gas behaviour not only in the pellet rim but also as an additional fission gas path in the whole pellet. (author)

High-Speed Turbo-TCM-Coded Orthogonal Frequency-Division Multiplexing Ultra-Wideband Systems

Directory of Open Access Journals (Sweden)

Wang Yanxia

2006-01-01

Full Text Available One of the UWB proposals in the IEEE P802.15 WPAN project is to use a multiband orthogonal frequency-division multiplexing (OFDM system and punctured convolutional codes for UWB channels supporting a data rate up to 480 Mbps. In this paper, we improve the proposed system using turbo TCM with QAM constellation for higher data rate transmission. We construct a punctured parity-concatenated trellis codes, in which a TCM code is used as the inner code and a simple parity-check code is employed as the outer code. The result shows that the system can offer a much higher spectral efficiency, for example, 1.2 Gbps, which is 2.5 times higher than the proposed system. We identify several essential requirements to achieve the high rate transmission, for example, frequency and time diversity and multilevel error protection. Results are confirmed by density evolution.

Forty-Year Follow-up of Full-Thickness Skin Graft After Thermal Burn Injury to the Volar Hand.

Science.gov (United States)

Weeks, Dexter; Kasdan, Morton L; Wilhelmi, Bradon J

2016-01-01

The hands are commonly affected in severe thermal burn injuries. Resulting contractures lead to significant loss of function. Burn contracture release and skin grafting are necessary to restore hand function. We report a case in which surgical reconstruction of a volar hand burn was performed with full-thickness skin grafting. The patient had a 40-year follow-up to assess the function and cosmesis of the repaired hand. We report a case in which a 15-month-old boy presented after receiving third-degree burns to the left volar hand, including the flexural aspects of the index, long, and ring fingers by placing it on a hot kitchen stove burner. The patient subsequently underwent scar contracture release and full-thickness skin grafting. Eleven years after reconstruction, further contractures developed associated with the patient's growth, which were reconstructed with repeat full-thickness skin graft from the inguinal region. No recurrence was witnessed afterward and 40 years after initial injury, the patient maintains full activities of daily living and use of his hand in his occupation. There is debate regarding the superiority of split-thickness versus full-thickness grafts during reconstruction. Our case strengthens the argument for durability of a full-thickness skin graft following thermal burn injury.

Development of the vacuum system pressure responce analysis code PRAC

International Nuclear Information System (INIS)

Horie, Tomoyoshi; Kawasaki, Kouzou; Noshiroya, Shyoji; Koizumi, Jun-ichi.

1985-03-01

In this report, we show the method and numerical results of the vacuum system pressure responce analysis code. Since fusion apparatus is made up of many vacuum components, it is required to analyze pressure responce at any points of the system when vacuum system is designed or evaluated. For that purpose evaluating by theoretical solution is insufficient. Numerical analysis procedure such as finite difference method is usefull. In the PRAC code (Pressure Responce Analysis Code), pressure responce is obtained solving derivative equations which is obtained from the equilibrium relation of throughputs and contain the time derivative of pressure. As it considers both molecular and viscous flows, the coefficients of the equation depend on the pressure and the equations become non-linear. This non-linearity is treated as piece-wise linear within each time step. Verification of the code is performed for the simple problems. The agreement between numerical and theoretical solutions is good. To compare with the measured results, complicated model of gas puffing system is analyzed. The agreement is well for practical use. This code will be a useful analytical tool for designing and evaluating vacuum systems such as fusion apparatus. (author)

Power decoding Reed-Solomon codes up to the Johnson radius

DEFF Research Database (Denmark)

Rosenkilde, Johan Sebastian Heesemann

2018-01-01

Power decoding, or "decoding using virtual interleaving" is a technique for decoding Reed-Solomon codes up to the Sudan radius. Since the method's inception, it has been an open question if it is possible to use this approach to decode up to the Johnson radius - the decoding radius of the Guruswami...

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

Energy Technology Data Exchange (ETDEWEB)

Lemmens, K., E-mail: klemmens@sckcen.be [Waste and Disposal Expert Group, Belgian Nuclear Research Centre (SCK-CEN), Boeretang 200, 2400 Mol (Belgium); González-Robles, E.; Kienzler, B. [Karlsruhe Institute of Technology Institute for Nuclear Waste Disposal (KIT-INE), PO Box 3640, D-76021 Karlsruhe (Germany); Curti, E. [Laboratory for Waste Management, Nuclear Energy and Safety Dept., Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Serrano-Purroy, D. [European Commission, DG Joint Research Centre - JRC, Directorate G - Nuclear Safety & Security, Department G.III, PO Box 2340, D-76125 Karlsruhe (Germany); Sureda, R.; Martínez-Torrents, A. [CTM Centre Tecnològic, Plaça de la Ciència 2, 08243 Manresa (Spain); Roth, O. [Studsvik, Nuclear AB, 611 82 Nyköping (Sweden); Slonszki, E. [Magyar Tudományos Akadémia Energiatudományi Kutatóközpont (MTA EK), PO Box 49, H-1525 Budapest (Hungary); Mennecart, T. [Waste and Disposal Expert Group, Belgian Nuclear Research Centre (SCK-CEN), Boeretang 200, 2400 Mol (Belgium); Günther-Leopold, I. [Laboratory for Waste Management, Nuclear Energy and Safety Dept., Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Hózer, Z. [Magyar Tudományos Akadémia Energiatudományi Kutatóközpont (MTA EK), PO Box 49, H-1525 Budapest (Hungary)

2017-02-15

The instant release of fission products from high burn-up UO{sub 2} 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/t{sub HM} 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/H{sub 2} 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. - Highlights: • Leach tests were performed to study the instant release of fission products from high burn-up UO{sub 2} fuels and one MOX fuel. • In these tests, the fission gas release given by the operator was a pessimistic estimator of the iodine and cesium release. • Iodine and cesium release is proportional to linear power rating beyond 200 W cm{sup −1}. • Closure of the fuel-cladding gap at high burn-up slows down the release. • The release rate decreases following an exponential equation.

Fusion PIC code performance analysis on the Cori KNL system

Energy Technology Data Exchange (ETDEWEB)

Koskela, Tuomas S. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). National Energy Research Scientific Computing Center (NERSC); Deslippe, Jack [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). National Energy Research Scientific Computing Center (NERSC); Friesen, Brian [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). National Energy Research Scientific Computing Center (NERSC); Raman, Karthic [INTEL Corp. (United States)

2017-05-25

We study the attainable performance of Particle-In-Cell codes on the Cori KNL system by analyzing a miniature particle push application based on the fusion PIC code XGC1. We start from the most basic building blocks of a PIC code and build up the complexity to identify the kernels that cost the most in performance and focus optimization efforts there. Particle push kernels operate at high AI and are not likely to be memory bandwidth or even cache bandwidth bound on KNL. Therefore, we see only minor benefits from the high bandwidth memory available on KNL, and achieving good vectorization is shown to be the most beneficial optimization path with theoretical yield of up to 8x speedup on KNL. In practice we are able to obtain up to a 4x gain from vectorization due to limitations set by the data layout and memory latency.

3D pin-by-pin power density profiles with high spatial resolution in the vicinity of a BWR control blade tip simulated with coupled neutronics/burn-up calculations

International Nuclear Information System (INIS)

Li, J.; Nünighoff, K.; Allelein, H.-J.

2011-01-01

Highlights: ► High spatial resolution neutronic and burn-up calculations of quarter BWR fuel element section. ► Coupled MCNP(X)–ORIGEN2.2 simulation using VESTA. ► Control blade history effect was taken into account. ► Determining local power excursion after instantaneous control rod movement. ► Correlation between control blade geometry and occurrence of local power excursions. - Abstract: Pellet cladding interaction (PCI) as well as pellet cladding mechanical interaction (PCMI) are well-known fuel failures in light water reactors, especially in boiling water reactors (BWR). Whereas the thermo-mechanical processes of PCI effects have been intensively investigated in the last decades, only rare information is available on the role of neutron physics. However, each power transient is primary due to neutron physics effects and thus knowledge of the neutron physical background is mandatory to better understand the occurrence of PCI effects in BWRs. This paper will focus on a study of local power excursions in a typical BWR fuel assembly during control rod movements. Burn-up and energy deposition were simulated with high spatial granularity, especially in the vicinity of the control blade tip. It could be shown, that the design of the control blade plays a dominant role for the occurrence of local power peaks while instantaneously moving down the control rod. The main result is, that the largest power peak occurs at the interface between steel handle and absorber rods. A full width half maximum (FWHM) of ±2.5 cm was observed. This means, the local power excursion due to neutron physics phenomena involve approximately five pellets. With the VESTA code coupled MCNP(X)/ORIGEN2.2 calculations were performed with more than 3400 burn-up zones in order to take history effects into account.

Influences on Prescribed Burning Activity and Costs in the National Forest System

Science.gov (United States)

David A. Cleaves; Jorge Martinez; Terry K. Haines

2000-01-01

The results of a survey concerning National Forest System prescribed burning activity and costs from 1985 to 1995 are examined. Ninety-five of one hundred and fourteen national forests responded. Acreage burned and costs for conducting burns are reported for four types of prescribed fires slash reduction; management-ignited fires; prescribed natural fires; and brush,...

Benchmarking of FA2D/PARCS Code Package

International Nuclear Information System (INIS)

Grgic, D.; Jecmenica, R.; Pevec, D.

2006-01-01

FA2D/PARCS code package is used at Faculty of Electrical Engineering and Computing (FER), University of Zagreb, for static and dynamic reactor core analyses. It consists of two codes: FA2D and PARCS. FA2D is a multigroup two dimensional transport theory code for burn-up calculations based on collision probability method, developed at FER. It generates homogenised cross sections both of single pins and entire fuel assemblies. PARCS is an advanced nodal code developed at Purdue University for US NRC and it is based on neutron diffusion theory for three dimensional whole core static and dynamic calculations. It is modified at FER to enable internal 3D depletion calculation and usage of neutron cross section data in a format produced by FA2D and interface codes. The FA2D/PARCS code system has been validated on NPP Krsko operational data (Cycles 1 and 21). As we intend to use this code package for development of IRIS reactor loading patterns the first logical step was to validate the FA2D/PARCS code package on a set of IRIS benchmarks, starting from simple unit fuel cell, via fuel assembly, to full core benchmark. The IRIS 17x17 fuel with erbium burnable absorber was used in last full core benchmark. The results of modelling the IRIS full core benchmark using FA2D/PARCS code package have been compared with reference data showing the adequacy of FA2D/PARCS code package model for IRIS reactor core design.(author)

Development of the versatile reactor analysis code system, MARBLE2

International Nuclear Information System (INIS)

Yokoyama, Kenji; Jin, Tomoyuki; Hazama, Taira; Hirai, Yasushi

2015-07-01

The second version of the versatile reactor analysis code system, MARBLE2, has been developed. A lot of new functions have been added in MARBLE2 by using the base technology developed in the first version (MARBLE1). Introducing the remaining functions of the conventional code system (JOINT-FR and SAGEP-FR), MARBLE2 enables one to execute almost all analysis functions of the conventional code system with the unified user interfaces of its subsystem, SCHEME. In particular, the sensitivity analysis functionality is available in MARBLE2. On the other hand, new built-in solvers have been developed, and existing ones have been upgraded. Furthermore, some other analysis codes and libraries developed in JAEA have been consolidated and prepared in SCHEME. In addition, several analysis codes developed in the other institutes have been additionally introduced as plug-in solvers. Consequently, gamma-ray transport calculation and heating evaluation become available. As for another subsystem, ORPHEUS, various functionality updates and speed-up techniques have been applied based on user experience of MARBLE1 to enhance its usability. (author)

On the condition of UO{sub 2} nuclear fuel irradiated in a PWR to a burn-up in excess of 110 MWd/kgHM

Energy Technology Data Exchange (ETDEWEB)

Restani, R.; Horvath, M. [Paul Scherrer Institut, CH-5232, Villigen PSI (Switzerland); Goll, W. [AREVA GmbH, P.O. Box 1109, DE-91001 Erlangen (Germany); Bertsch, J.; Gavillet, D.; Hermann, A. [Paul Scherrer Institut, CH-5232, Villigen PSI (Switzerland); Martin, M., E-mail: matthias.martin@psi.ch [Paul Scherrer Institut, CH-5232, Villigen PSI (Switzerland); Walker, C.T. [The Grange, 66 High Street, Swinderby, Lincoln LN6 9LU (United Kingdom)

2016-12-01

Post-irradiation examination results are presented for UO{sub 2} fuel from a PWR fuel rod that had been irradiated to an average burn-up of 105 MWd/kgHM and showed high fission gas release of 42%. The radial distribution of xenon and the partitioning of fission gas between bubbles and the fuel matrix was investigated using laser ablation inductively coupled plasma mass spectrometry (LA-ICP-MS) and electron probe microanalysis. It is concluded that release from the fuel at intermediate radial positions was mainly responsible for the high fission gas release. In this region thermal release had occurred from the high burn-up structure (HBS) at some point after the sixth irradiation cycle. The LA-ICP-MS results indicate that gas release had also occurred from the HBS in the vicinity of the pellet periphery. It is shown that the gas pressure in the HBS pores is well below the pressure that the fuel can sustain. - Highlights: • Gas retention measured by laser ablation induction coupled plasma mass spectrometry. • Thermal release from the high burn structure responsible for high gas release. • At a pellet burn-up of 115 MWd/kgHM the high burn-up structure is still evolving. • The gas pressure in HBS pores is well below the pressure that the fuel can sustain.

The NJOY nuclear data processing system: The MICROR module

International Nuclear Information System (INIS)

Mathews, D.R.; Stepanek, J.; Pelloni, S.; Higgs, C.E.

1984-12-01

The NJOY nuclear data processing system is a comprehensive computer code package for producing pointwise and multigroup neutron and photon cross sections and related nuclear parameters from ENDF/B-IV and V evaluated nuclear data. The MICROR overlay is a reformatting module that produces cross sections library files for the MICROX, MICROX-2 and MICROBURN postprocessor codes. Using the data on the pointwise and groupwise NJOY tapes, MICROR produces the tapes containing basic nuclear data, FDTAPE, GAR and GGTAPE used by two-region spectrum codes MICROX and MICROX-2 and by two-region spectrum burn-up code MICROBURN. (author)

The impact of major trauma network triage systems on patients with major burns.

Science.gov (United States)

Nizamoglu, Metin; O'Connor, Edmund Fitzgerald; Bache, Sarah; Theodorakopoulou, Evgenia; Sen, Sankhya; Sherren, Peter; Barnes, David; Dziewulski, Peter

2016-12-01

Trauma is a leading cause of death and disability worldwide. Patients presenting with severe trauma and burns benefit from specifically trained multidisciplinary teams. Regional trauma systems have shown improved outcomes for trauma patients. The aim of this study is to determine whether the development of major trauma systems have improved the management of patients with major burns. A retrospective study was performed over a four-year period reviewing all major burns in adults and children received at a regional burns centre in the UK before and after the implementation of the regional trauma systems and major trauma centres (MTC). Comparisons were drawn between three areas: (1) Patients presenting before the introduction of MTC and after the introduction of MTC. (2) Patients referred from MTC and non-MTC within the region, following the introduction of MTC. (3) Patients referred using the urban trauma protocol and the rural trauma protocol. Following the introduction of regional trauma systems and major trauma centres (MTC), isolated burn patients seen at our regional burns centre did not show any significant improvement in transfer times, admission resuscitation parameters, organ dysfunction or survival when referred from a MTC compared to a non-MTC emergency department. There was also no significant difference in survival when comparing referrals from all hospitals pre and post establishment of the major trauma network. No significant outcome benefit was demonstrated for burns patients referred via MTCs compared to non-MTCs. We suggest further research is needed to ascertain whether burns patients benefit from prolonged transfer times to a MTC compared to those seen at their local hospitals prior to transfer to a regional burns unit for further specialist care. Copyright © 2016 Elsevier Ltd and ISBI. All rights reserved.

Aircraft Engine Technology for Green Aviation to Reduce Fuel Burn

Science.gov (United States)

Hughes, Christopher E.; VanZante, Dale E.; Heidmann, James D.

2013-01-01

The NASA Fundamental Aeronautics Program Subsonic Fixed Wing Project and Integrated Systems Research Program Environmentally Responsible Aviation Project in the Aeronautics Research Mission Directorate are conducting research on advanced aircraft technology to address the environmental goals of reducing fuel burn, noise and NOx emissions for aircraft in 2020 and beyond. Both Projects, in collaborative partnerships with U.S. Industry, Academia, and other Government Agencies, have made significant progress toward reaching the N+2 (2020) and N+3 (beyond 2025) installed fuel burn goals by fundamental aircraft engine technology development, subscale component experimental investigations, full scale integrated systems validation testing, and development validation of state of the art computation design and analysis codes. Specific areas of propulsion technology research are discussed and progress to date.

Numerical analysis and simulation of behavior of high burn-up PWR fuel pulse-irradiated in reactivity-initiated accident conditions

International Nuclear Information System (INIS)

Suzuki, M.; Sugiyama, T.; Udagawa, Y.; Nagase, F.; Fuketa, T.

2010-01-01

The four cases of the NSRR experiments, consisting of two room temperature tests and two high temperature tests, using high burn-up PWR fuel rods are analyzed by using the RANNS code to discuss the fuel behavior in hypothetical pulse-irradiation conditions, and the results are compared with metallography observations of ruptured claddings. The cladding rupture occurred by a shear sliding which starts from the tip of incipient crack generated in the hydride dense layer. The analyses reveal that the onset of shear sliding leading to cladding rupture can be closely associated with the stress intensity factor KI at the crack tip and local plastic strain evolution around the tip as well, and that these two factors depend also on the temperature of cladding. Simulation calculations on the basis of experimental conditions reveals that the cladding stress is dependent on the height and half-width of pulse power, and for the same integral enthalpy of pulse a larger half-width mitigates the severity of transient and decreases KI to allow plastic strain by temperature rise, thus failure possibility would be markedly decreased

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.

Thermal-Hydraulic System Codes in Nulcear Reactor Safety and Qualification Procedures

Directory of Open Access Journals (Sweden)

Alessandro Petruzzi

2008-01-01

Full Text Available In the last four decades, large efforts have been undertaken to provide reliable thermal-hydraulic system codes for the analyses of transients and accidents in nuclear power plants. Whereas the first system codes, developed at the beginning of the 1970s, utilized the homogenous equilibrium model with three balance equations to describe the two-phase flow, nowadays the more advanced system codes are based on the so-called “two-fluid model” with separation of the water and vapor phases, resulting in systems with at least six balance equations. The wide experimental campaign, constituted by the integral and separate effect tests, conducted under the umbrella of the OECD/CSNI was at the basis of the development and validation of the thermal-hydraulic system codes by which they have reached the present high degree of maturity. However, notwithstanding the huge amounts of financial and human resources invested, the results predicted by the code are still affected by errors whose origins can be attributed to several reasons as model deficiencies, approximations in the numerical solution, nodalization effects, and imperfect knowledge of boundary and initial conditions. In this context, the existence of qualified procedures for a consistent application of qualified thermal-hydraulic system code is necessary and implies the drawing up of specific criteria through which the code-user, the nodalization, and finally the transient results are qualified.

[Application of a hydrosurgery system in debridement of various types of burn wounds].

Science.gov (United States)

Li, M Y; Mao, Y G; Guo, G H; Liu, D W

2016-09-20

Burn wound healing is closely associated with the depth of wound and early debridement. The traditional ways of debridement have certain limitations and often result in poor appearance and function of repaired area. At present, the hydrosurgery system has been applied clinically in burn field. This paper summarizes advantages and disadvantages of application of the hydrosurgery system in debridement of burn wound with different depths, different periods, extraordinary region, and uncommon agent.

Economic burden of burn injuries in the Netherlands: A 3 months follow-up study.

Science.gov (United States)

Hop, M Jenda; Wijnen, Ben F M; Nieuwenhuis, Marianne K; Dokter, Jan; Middelkoop, Esther; Polinder, Suzanne; van Baar, Margriet E

2016-01-01

Burn care has rapidly improved in the past decades. However, healthcare innovations can be expensive, demanding careful choices on their implementation. Obtaining knowledge on the extent of the costs of burn injuries is an essential first step for economic evaluations within burn care. The objective of this study was to determine the economic burden of patients with burns admitted to a burn centre and to identify important cost categories until 3 months post-burn. A prospective cohort study was conducted in the burn centre of Maasstad Hospital Rotterdam, the Netherlands, including all patients with acute burn related injuries from August 2011 until July 2012. Total costs were calculated from a societal perspective, until 3 months post injury. Subgroup analyses were performed to examine whether the mean total costs per patient differed by age, aetiology or percentage total body surface area (TBSA) burned. In our population, with a mean burn size of 8%, mean total costs were €26,540 per patient varying from €742 to €235,557. Most important cost categories were burn centre days (62%), surgical interventions (5%) and work absence (20%). Flame burns were significantly more costly than other types of burns, adult patients were significantly more costly than children and adolescents and a higher percentage TBSA burned also corresponded to significantly higher costs. Mean total costs of burn care in the first 3 months post injury were estimated at €26,540 and depended on age, aetiology and TBSA. Mean total costs in our population probably apply to other high-income countries as well, although we should realise that patients with burn injuries are diverse and represent a broad range of total costs. To reduce costs of burn care, future intervention studies should focus on a timely wound healing, reducing length of stay and enabling an early return to work. Copyright © 2015 Elsevier Ltd. All rights reserved.

Telemedicine and burns: an overview.

Science.gov (United States)

Atiyeh, B; Dibo, S A; Janom, H H

2014-06-30

Access to specialized burn care is becoming more difficult and is being restricted by the decreasing number of specialized burn centers. It is also limited by distance and resources for many patients, particularly those living in poverty or in rural medically underserved communities. Telemedicine is a rapidly evolving technology related to the practice of medicine at a distance through rapid access to remote medical expertise by telecommunication and information technologies. Feasibility of telemedicine in burn care has been demonstrated by various centers. Its use facilitates the delivery of care to patients with burn injuries of all sizes. It allows delivery of acute care and can be appropriately used for a substantial portion of the long-term management of patients after a burn by guiding less-experienced surgeons to treat and follow-up patients more appropriately. Most importantly, it allows better effective triage which reduces unnecessary time and resource demanding referrals that might overwhelm system capacities. However, there are still numerous barriers to the implementation of telemedicine, including technical difficulties, legal uncertainties, limited financial support, reimbursement issues, and an inadequate evidence base of its value and efficiency.

The role of grain boundary fission gases in high burn-up fuel under reactivity initiated accident conditions

International Nuclear Information System (INIS)

Lemoine, F.; Papin, J.; Frizonnet, J.M.; Cazalis, B.; Rigat, H.

2002-01-01

In the frame of reactivity-initiated accidents (RIA) studies, the CABRI REP-Na programme is currently performed, focused on high burn-up UO 2 and MOX fuel behaviour. From 1993 to 1998, seven tests were performed with UO 2 fuel and three with MOX fuel. In all these tests, particular attention has been devoted to the role of fission gases in transient fuel behaviour and in clad loading mechanisms. From the analysis of experimental results, some basic phenomena were identified and a better understanding of the transient fission gas behaviour was obtained in relation to the fuel and clad thermo-mechanical evolution in RIA, but also to the initial state of the fuel before the transient. A high burn-up effect linked to the increasing part of grain boundary gases is clearly evidenced in the final gas release, which would also significantly contribute to the clad loading mechanisms. (authors)

Ambient air pollution associated to domestic wood burning heating systems

International Nuclear Information System (INIS)

Friboulet, I.; Durif, M.; Malherbe, L.

2009-01-01

Main publications are considering effects of wood burning appliances on indoor air quality, which is a major issue in some countries. But impacts on ambient air, close environment and human exposure are rather poorly characterised so far. Besides, woods burning for domestic purpose may develop in the next years while promoting bio fuels. The aim of the ongoing study is to assess in which conditions associated air pollution and population exposure could be significant, this poster shows preliminary results of the impact of a village of 98 houses equipped with a wood burning heating system. (N.C.)

A study on hydrogen burn due to the operation of containment spray system

International Nuclear Information System (INIS)

Park, S.Y.; Kim, D.H.; Jin, Y.; Park, C.K.

1995-01-01

The bounding calculation for inflammable gas combustion due to the steam condensation by the operation of the containment spray system was performed. Sensitivity study was performed for two initiating events, station blackout and loss of coolant accident. The parameters for sensitivity study are the condition of cavity, wet or dry, and the timing of operation of the containment spray system. It is shown, based on MAAP4 analyses, that: for dry cavity, auto-ignition burn and hydrogen laden jet burn due to the high temperature in the reactor cavity consumes large amount of burnable gas in the containment and reduces the peak pressure at the global burn by flammability criteria; for wet cavity, large amount of hydrogen and carbon monoxide are generated after dryout of the reactor cavity, but burn is prohibited due to the low gas temperature in the high concentration of the steam. The late operation of the containment spray system condenses the steam rapidly, which results in the global burn at high concentration of burnable gas in the containment. The containment peak pressure from this burn is determined to be high enough to threaten the containment integrity significantly. (author). 3 refs., 3 tabs

Subchannel analysis of a boiloff experiment by a system thermalhydraulic code

International Nuclear Information System (INIS)

Bousbia-Salah, A.; D'Auria, F.

2001-01-01

This paper presents the results of system thermalhydraulic code using the sub-channel analysis approach in predicting the Neptun boil off experiments. This approach will be suitable for further works in view of coupling the system code with a 3D neutron kinetic one. The boil off tests were conducted in order to simulate the consequences of loss of coolant inventory leading to uncovery and heat up of fuel elements of a nuclear reactor core. In this framework, the Neptun low pressure test No5002, which is a good repeat experiment, is considered. The calculations were carried out using the system transient analysis code Relap5/Mod3.2. A detailed nodalization of the Neptun test section was developed. A reference case was run, and the overall data comparison shows good agreement between calculated and experimental thermalhydraulic parameters. A series of sensitivity analyses were also performed in order to assess the code prediction capabilities. The obtained results were almost satisfactory, this demonstrates, as well, the reasonable success of the subchannel analysis approach adopted in the present context for a system thermalhydraulic code.(author)

Modelling of thermal mechanical behaviour of high burn-Up VVER fuel at power transients with special emphasis on the impact of fission gas induced swelling of fuel pellets

International Nuclear Information System (INIS)

Novikov, V.; Medvedev, A.; Khvostov, G.; Bogatyr, S.; Kuzetsov, V.; Korystin, L.

2005-01-01

This paper is devoted to the modelling of unsteady state mechanical and thermo-physical behaviour of high burn-up VVER fuel at a power ramp. The contribution of the processes related to the kinetics of fission gas to the consequences of pellet-clad mechanical interaction is analysed by the example of integral VVER-440 rod 9 from the R7 experimental series, with a pellet burn-up in the active part at around 60 MWd/kgU. This fuel rod incurred ramp testing with a ramp value ΔW 1 ∼ 250 W/cm in the MIR research reactor. The experimentally revealed residual deformation of the clad by 30-40 microns in the 'hottest' portion of the rod, reaching a maximum linear power of up to 430 W/cm, is numerically justified on the basis of accounting for the unsteady state swelling and additional degradation of fuel thermal conductivity due to temperature-induced formation and development of gaseous porosity within the grains and on the grain boundaries. The good prediction capability of the START-3 code, coupled with the advanced model of fission gas related processes, with regard to the important mechanical (residual deformation of clad, pellet-clad gap size, central hole filling), thermal physical (fission gas release) and micro-structural (profiles of intra-granular concentration of the retained fission gas and fuel porosity across a pellet) consequences of the R7 test is shown. (authors)

Interim Status Closure Plan Open Burning Treatment Unit Technical Area 16-399 Burn Tray

Energy Technology Data Exchange (ETDEWEB)

Vigil-Holterman, Luciana R. [Los Alamos National Laboratory

2012-05-07

This closure plan describes the activities necessary to close one of the interim status hazardous waste open burning treatment units at Technical Area (TA) 16 at the Los Alamos National Laboratory (LANL or the Facility), hereinafter referred to as the 'TA-16-399 Burn Tray' or 'the unit'. The information provided in this closure plan addresses the closure requirements specified in the Code of Federal Regulations (CFR), Title 40, Part 265, Subparts G and P for the thermal treatment units operated at the Facility under the Resource Conservation and Recovery Act (RCRA) and the New Mexico Hazardous Waste Act. Closure of the open burning treatment unit will be completed in accordance with Section 4.1 of this closure plan.

Burn-up Credit Criticality Safety Benchmark-Phase II-E. Impact of Isotopic Inventory Changes due to Control Rod Insertions on Reactivity and the End Effect in PWR UO2 Fuel Assemblies

International Nuclear Information System (INIS)

Neuber, Jens Christian; Tippl, Wolfgang; Hemptinne, Gwendoline de; Maes, Philippe; Ranta-aho, Anssu; Peneliau, Yannick; Jutier, Ludyvine; Tardy, Marcel; Reiche, Ingo; Kroeger, Helge; Nakata, Tetsuo; Armishaw, Malcom; Miller, Thomas M.

2015-01-01

The report describes the final results of the Phase II-E Burn-up Credit Criticality Benchmark conducted by the Expert Group on Burn-up Credit Criticality Safety. The objective of Phase II of the Burn-up Credit Criticality Safety programme is to study the impact of axial burn-up profiles of PWR UO 2 spent fuel assemblies on the reactivity of PWR UO 2 spent fuel assembly configurations. The objective of the Phase II-E benchmark was to study the impact of changes on the spent nuclear fuel isotopic composition due to control rod insertion during depletion on the reactivity and the end effect of spent fuel assemblies with realistic axial burn-up profiles for different control rod insertion depths ranging from 0 cm (no insertion) to full insertion (i.e. to the case that the fuel assemblies were exposed to control rod insertion over their full active length). For this purpose two axial burn-up profiles have been extracted from an AREVA-NP-GmbH-owned 17x17-(24+1) PWR UO 2 spent fuel assembly burn-up profile database. One profile has an average burn-up of 30 MWd/kg U, the other profile is related to an average burn-up of 50 MWd/kg U. Two profiles with different average burn-up values were selected because the shape of the burn-up profile is affected by the average burn-up and the end effect depends on the average burn-up of the fuel. The Phase II-E benchmark exercise complements the Phase II-C and Phase II-D benchmark exercises. In Phase II-D different irradiation histories were analysed using different control rod insertion histories during depletion as well as irradiation histories without control rod insertion. But in all the histories analysed a uniform distribution of the burn-up and hence a uniform distribution of the isotopic composition were assumed; and in all the histories including any usage of control rods full insertion of the control rods was assumed. In Phase II-C the impact of the asymmetry of axial burn-up profiles on the reactivity and the end effect of

An investigation into fuel pulverization with specific reference to high burn-up LOCA

International Nuclear Information System (INIS)

Yagnik, Suresh; Turnbull, James; Noirot, Jean; Walker, Clive; Hallstadius, Lars; Waeckel, N.; Blanpain, P.

2014-01-01

To investigate the phenomenon of high burn-up fuel pellet material potentially disintegrating into powder under a rapid temperature transient, such as in a LOCA-type accident scenario, two independent scoping studies were commissioned. The first was to investigate the effect of hydrostatic restraint pressure on Fission Gas Release (FGR) from small samples of highly irradiated fuel (71 MWd/kgU) during a series of rapid temperature ramps. Experimentally, when the FGR increased rapidly during the temperature transients, the fuel was assumed to be 'pulverized', i.e., fragmented into powder. In the second series of experiments, laser heating of small samples was used to investigate the temperature at which fuel pulverization was initiated. Subsequent to fuel disintegration, there was always a spectrum of particle sizes present. The significance of this observation was recognized in the context of extended burn-up operation in commercial reactors. Based on the observation from these investigations, a fuel fragmentation threshold has been discussed and developed. We conclude that fuel disintegration could be of potential importance in limiting the performance and productive lifetime of nuclear fuel. However, since only fuel closely adjacent to ballooned or ruptured cladding would be released in a LOCA-type transient, expulsion of pulverized fuel from the ruptured fuel rod is not considered a safety issue; cooling of the defected assembly remains possible and there is no issue with respect to local criticality. (author)

Experimental modeling of high burn-up structure in SIMFUEL with ion irradiation

International Nuclear Information System (INIS)

Baranov, V.; Isaenkova, M.; Lunev, A.; Tenishev, A.; Khlunov, A.

2013-01-01

Experiments are conducted to simulate high burn-up structure in accelerator conditions. Three ion irradiation schemes are used: 1. Xe 27+ 160 MeV up to 5x10 15 cm -2 (thermal spikes). 2. Xe 16+ 320 keV up to 1x10 17 cm -2 (collision cascades). 3. He + 20 keV up to 5,5x10 17 cm -2 (implantation stage). Structural characterization performed by scanning electron microscopy, X-ray analysis and atomic force microscopy revealed prominent grain refinement in case of Xe 27+ irradiation. Artificial energy variation for incident ions showed varying size of subgrains. At maximum energy of incident ions, subgrain size amounts ∼ 320 nm. Moving to the edge of irradiated region changes the size to ∼ 170 nm. Typical size of coherent scattering regions matches subgrain size for high-energy irradiation. Low-energy irradiation results in less significant structural changes: flaky structure at random sites for samples irradiated with low-energy xenon ions and bubble nucleation for helium irradiation. Dislocation density increases significantly, and it is shown that a single fluence dependence exists for low- and high-energy irradiation. (authors)

ESCADRE and ICARE code systems

International Nuclear Information System (INIS)

Reocreux, M.; Gauvain, J.

1992-01-01

The French sever accident code development program is following two parallel approaches: the first one is dealing with ''integral codes'' which are designed for giving immediate engineer answers, the second one is following a more mechanistic way in order to have the capability of detailed analysis of experiments, in order to get a better understanding of the scaling problem and reach a better confidence in plant calculations. In the first approach a complete system has been developed and is being used for practical cases: this is the ESCADRE system. In the second approach, a set of codes dealing first with primary circuit is being developed: a mechanistic core degradation code, ICARE, has been issued and is being coupled with the advanced thermalhydraulic code CATHARE. Fission product codes have been also coupled to CATHARE. The ''integral'' ESCADRE system and the mechanistic ICARE and associated codes are described. Their main characteristics are reviewed and the status of their development and assessment given. Future studies are finally discussed. 36 refs, 4 figs, 1 tab

The CORSYS neutronics code system

International Nuclear Information System (INIS)

Caner, M.; Krumbein, A.D.; Saphier, D.; Shapira, M.

1994-01-01

The purpose of this work is to assemble a code package for LWR core physics including coupled neutronics, burnup and thermal hydraulics. The CORSYS system is built around the cell code WIMS (for group microscopic cross section calculations) and 3-dimension diffusion code CITATION (for burnup and fuel management). We are implementing such a system on an IBM RS-6000 workstation. The code was rested with a simplified model of the Zion Unit 2 PWR. (authors). 6 refs., 8 figs., 1 tabs

Prestudy of burn control in NET

International Nuclear Information System (INIS)

Anderson, D.; Hamnen, H.; Lisak, M.

1990-02-01

The present report describes our ongoing work on a number of selected topics, and the plans for the nearest future. In chapter 2 we have specialized the system of the previous report to form an easily tractable, second-order system. In this case one can give an explicit, analytical condition for stability. A code providing quick answers regarding stability, time scales and eigenvectors has been written and tested. The zerodimensional modelling of a burning plasma described by space dependent equations is often done in a heuristic way, with no clear relation between the two systems of equations. We have tried to put the approximation procedure involved in the transition to 0-D models on a more formal basis. This is the topic of chapter 3. The 1-D equilibrium solution is also investigated with respect to its stability properties, which are shown to be the same as those derived from the simplest 0-D space averaged model. Chapter 4 contains a few emerging thoughts on burn control. First, the limited swing of the auxiliary heating gives rise to limitations on the possibilities to intervene against temperature excursions by an auxiliary heating modulation. This problem becomes severe when one operates at high Q values. Another analysis concerns the problem of selecting a proper control action when a temperature profile differs from the equilibrium shape. A couple of alternative schemes for burn control, minor radius alterations and dynamic stabilization are tentatively discussed; no definite answers on their feasibility are obtained. The problem of diagnosing the plasma with respect to burn conditions is the topic of chapter 5. The influence on the energy distribution of control actions and the reliability of neutron measurements are discussed, and the question of how to handle sawteeth is briefly revisited. Chapter 6 is a description of process identification and how it could be used for burn control. A particular advantage is that it can be combined with physical

A structural modification of the two dimensional fuel behaviour analysis code FEMAXI-III with high-speed vectorized operation

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki; Ishiguro, Misako; Yamazaki, Takashi; Tokunaga, Yasuo.

1985-02-01

Though the two-dimensional fuel behaviour analysis code FEMAXI-III has been developed by JAERI in form of optimized scalar computer code, the call for more efficient code usage generally arized from the recent trends like high burn-up and load follow operation asks the code into further modification stage. A principal aim of the modification is to transform the already implemented scalar type subroutines into vectorized forms to make the programme structure efficiently run on high-speed vector computers. The effort of such structural modification has been finished on a fair way to success. The benchmarking two tests subsequently performed to examine the effect of the modification led us the following concluding remarks: (1) In the first benchmark test, comparatively high-burned three fuel rods that have been irradiated in HBWR, BWR, and PWR condition are prepared. With respect to all cases, a net computing time consumed in the vectorized FEMAXI is approximately 50 % less than that consumed in the original one. (2) In the second benchmark test, a total of 26 PWR fuel rods that have been irradiated in the burn-up ranges of 13-30 MWd/kgU and subsequently power ramped in R2 reactor, Sweden is prepared. In this case the code is purposed to be used for making an envelop of PCI-failure threshold through 26 times code runs. Before coming to the same conclusion, the vectorized FEMAXI-III consumed a net computing time 18 min., while the original FEMAXI-III consumed a computing time 36 min. respectively. (3) The effects obtained from such structural modification are found to be significantly attributed to saving a net computing time in a mechanical calculation in the vectorized FEMAXI-III code. (author)

Burn mouse models

DEFF Research Database (Denmark)

Calum, Henrik; Høiby, Niels; Moser, Claus

2014-01-01

Severe thermal injury induces immunosuppression, involving all parts of the immune system, especially when large fractions of the total body surface area are affected. An animal model was established to characterize the burn-induced immunosuppression. In our novel mouse model a 6 % third-degree b......Severe thermal injury induces immunosuppression, involving all parts of the immune system, especially when large fractions of the total body surface area are affected. An animal model was established to characterize the burn-induced immunosuppression. In our novel mouse model a 6 % third...... with infected burn wound compared with the burn wound only group. The burn mouse model resembles the clinical situation and provides an opportunity to examine or develop new strategies like new antibiotics and immune therapy, in handling burn wound victims much....

Assessing the response of area burned to changing climate in western boreal North America using a Multivariate Adaptive Regression Splines (MARS) approach

Science.gov (United States)

Michael S. Balshi; A. David McGuire; Paul Duffy; Mike Flannigan; John Walsh; Jerry Melillo

2009-01-01

We developed temporally and spatially explicit relationships between air temperature and fuel moisture codes derived from the Canadian Fire Weather Index System to estimate annual area burned at 2.5o (latitude x longitude) resolution using a Multivariate Adaptive Regression Spline (MARS) approach across Alaska and Canada. Burned area was...

Intraluminal Flagellin Differentially Contributes to Gut Dysbiosis and Systemic Inflammation following Burn Injury.

Directory of Open Access Journals (Sweden)

Logan Grimes

Full Text Available Burn injury is associated with a loss of gut barrier function, resulting in systemic dissemination of gut-derived bacteria and their products. The bacterial protein and TLR5 agonist, flagellin, induces non-specific innate immune responses. Because we detected flagellin in the serum of burn patients, we investigated whether gut-derived flagellin was a primary or secondary contributor to intestinal dysfunction and systemic inflammation following burn injury. The apical surface of polarized human intestinal epithelial cells (IECs, Caco-2BBe, were exposed to 50 or 500 ng of purified flagellin and 1 x 105 of an intestinal E. coli (EC isolate as follows: 1 flagellin added 30 min prior to EC, 2 flagellin and EC added simultaneously, or 3 EC added 30 min prior to flagellin. Our results showed that luminal flagellin and EC modulated each other's biological actions, which influenced their ability to induce basolateral secretion of inflammatory cytokines and subsequent translocation of bacteria and their products. A low dose of flagellin accompanied by an enteric EC in the lumen, tempered inflammation in a dose- and time-dependent manner. However, higher doses of flagellin acted synergistically with EC to induce both intestinal and systemic inflammation that compromised barrier integrity, increasing systemic inflammation following burn injury, a process we have termed flagellemia. In a murine model of burn injury we found that oral gavage of flagellin (1 μg/mouse significantly affected the gut microbiome after burn injury. In these mice, flagellin disseminated out of the intestine into the serum and to distal organs (mesenteric lymph nodes and lungs where it induced secretion of monocyte chemoattractant protein (MCP-1 and CXCL1/KC (mouse equivalent of human IL-8 at 24 and 48h post-burn. Our results illustrated that gut-derived flagellin alone or accompanied by a non-pathogenic enteric EC strain can function as an initiator of luminal and systemic

Coupling the severe accident code SCDAP with the system thermal hydraulic code MARS

Energy Technology Data Exchange (ETDEWEB)

Lee, Young Jin; Chung, Bub Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

2004-07-01

MARS is a best-estimate system thermal hydraulics code with multi-dimensional modeling capability. One of the aims in MARS code development is to make it a multi-functional code system with the analysis capability to cover the entire accident spectrum. For this purpose, MARS code has been coupled with a number of other specialized codes such as CONTEMPT for containment analysis, and MASTER for 3-dimensional kinetics. And in this study, the SCDAP code has been coupled with MARS to endow the MARS code system with severe accident analysis capability. With the SCDAP, MARS code system now has acquired the capability to simulate such severe accident related phenomena as cladding oxidation, melting and slumping of fuel and reactor structures.

Coupling the severe accident code SCDAP with the system thermal hydraulic code MARS

International Nuclear Information System (INIS)

Lee, Young Jin; Chung, Bub Dong

2004-01-01

MARS is a best-estimate system thermal hydraulics code with multi-dimensional modeling capability. One of the aims in MARS code development is to make it a multi-functional code system with the analysis capability to cover the entire accident spectrum. For this purpose, MARS code has been coupled with a number of other specialized codes such as CONTEMPT for containment analysis, and MASTER for 3-dimensional kinetics. And in this study, the SCDAP code has been coupled with MARS to endow the MARS code system with severe accident analysis capability. With the SCDAP, MARS code system now has acquired the capability to simulate such severe accident related phenomena as cladding oxidation, melting and slumping of fuel and reactor structures

Observing the Peripheral Burning of Cigarettes by an Infrared Technique

Directory of Open Access Journals (Sweden)

Liu C

2014-12-01

Full Text Available A modern infrared camera was used to observe the peripheral burning of cigarettes during puffing and smouldering. The computer-controlled infrared system captured thermal images with recording rates up to 50 Hz at 8-bit (256-colour resolution. The response time was less than 0.04 s at ca. 780 °C. The overall performance of the system was superior to most infrared systems used in previously reported investigations. The combined capacity allowed us to capture some faster, smaller high-temperature burning events on the periphery of a cigarette during puffing, which was first described by Egertion et al. in 1963 using an X-ray method. These transient burning events were caused by tobacco shreds near the coal surface experiencing the maximum air influx. The temperature of these transient burning events could be ca. 200 to 250 °C higher than the average peripheral temperature of the cigarette. The likelihood of these high-temperature burning events occurring during smouldering was significantly less. Some other details of the cigarette's combustion were also observed with improved simplicity and clarity.

Studies of a deep burn fuel cycle for the incineration of military plutonium in the GT-MHR using the Monte-Carlo burnup code

International Nuclear Information System (INIS)

Talamo, A.; Gudowski, W.

2004-01-01

The deep burn fuel cycle for the incineration of military plutonium in the GT-MHR is studied using the Monte-Carlo burnup code. The irradiation is DF is so rich in fissile isotopes that the TF cannot guarantee a negative reactive feedback, and the presence of erbium as burnable poison is absolutely necessary for the reactivity safety reasons. At beginning of life (BOL) the fuel composed of DF, consisting of fresh military plutonium, after an irradiation period of three years the fuel is reprocessed into post driver fuel (PDF). The mass flow of the GT-MHR fuelled by military plutonium at the equilibrium of the fuel composition shows that 66% of 239 Pu is burned in three years and 92% in six years. (authors)

Investigation of the burn-up behavior of boron poison rods, placed in a fuel assembly of a pressurized water reactor

International Nuclear Information System (INIS)

Arnold, C.; Lutz, D.C.

1979-09-01

The excess reactivity of a pressurized water reactor is compensated by boron, disolved in the moderator. In addition during the first cycle boron poison rods are placed in fuel assemblies without control rods. The burn-up behavior of a poison rod in a Biblis B fuel assembly is analysed in the present paper. Multigroup spectrum calculations were performed. The influence of critical boron concentration depending from burn-up, the changes of fuel concentration and the concentration of burnable poison were taken into consideration. Furthermore the built-up of rapidly saturating fisson products 135 Xe and 149 Sm was considered. The interaction of these effects are discussed. Spatial influences are emphasized most. Finally two group cross sections were calculated. The results are compared with calculations for a fuel assembly of the same type without burnable poison rods. (orig.) [de

System Design Description for the TMAD Code

International Nuclear Information System (INIS)

Finfrock, S.H.

1995-01-01

This document serves as the System Design Description (SDD) for the TMAD Code System, which includes the TMAD code and the LIBMAKR code. The SDD provides a detailed description of the theory behind the code, and the implementation of that theory. It is essential for anyone who is attempting to review or modify the code or who otherwise needs to understand the internal workings of the code. In addition, this document includes, in Appendix A, the System Requirements Specification for the TMAD System

The slash-and-burn agriculture: a system in transformation

Directory of Open Access Journals (Sweden)

Nelson Novaes Pedroso Júnior

2008-08-01

Full Text Available Slash-and-burn agriculture has been practiced for thousands of years in the forests around the world, especially in the tropics, where it provides for the livelihood of countless poor rural populations. Characterized by an array of techniques based on crop diversification and shifting land use, this cultivation system has on the utilization of forest decomposing vegetation´s energetic capital its main asset. Many studies claim that slash-and-burn agriculture is sustainable only when performed under conditions of low human demographic density and maintenance or even increase of local biodiversity. However, it is growing in the academic literature, as well as in development debates, the concern regarding the role that this system has been playing in the deforestation of the planet´s tropical forests. This process appears to be closely linked to changes in land use patterns (agricultural intensification and urban and rural demographic growth. On the thread of these concerns, this article presents a critical review of the international and national academic literature on slash-and-burn agriculture. Thus, this review intend to draw a broad scenario of the current academic debate on this issue, as well as to identify the main alternatives strategies proposed to maintain or replace this cultivation system.

A Novel Technique to Detect Code for SAC-OCDMA System

Science.gov (United States)

Bharti, Manisha; Kumar, Manoj; Sharma, Ajay K.

2018-04-01

The main task of optical code division multiple access (OCDMA) system is the detection of code used by a user in presence of multiple access interference (MAI). In this paper, new method of detection known as XOR subtraction detection for spectral amplitude coding OCDMA (SAC-OCDMA) based on double weight codes has been proposed and presented. As MAI is the main source of performance deterioration in OCDMA system, therefore, SAC technique is used in this paper to eliminate the effect of MAI up to a large extent. A comparative analysis is then made between the proposed scheme and other conventional detection schemes used like complimentary subtraction detection, AND subtraction detection and NAND subtraction detection. The system performance is characterized by Q-factor, BER and received optical power (ROP) with respect to input laser power and fiber length. The theoretical and simulation investigations reveal that the proposed detection technique provides better quality factor, security and received power in comparison to other conventional techniques. The wide opening of eye in case of proposed technique also proves its robustness.

Work-related burn injuries in Ontario, Canada: a follow-up 10-year retrospective study

Science.gov (United States)

Clouatre, Elsa; Gomez, Manuel; Banfield, Joanne; Jeschke, Marc G

2013-01-01

Work-related burn injuries contribute to a quarter of all burn injuries in USA. In 2009, the provincial Workplace Safety and Insurance Board reported 64,824 work-related injuries that resulted in time-lost, 1188 injuries (2%) were a result of burns. There have been two previous studies performed at a regional burn centre (1984-1990 and 1998-2000) looking at incidence and characteristics of work-related burns. There was no significant change between these two groups. The purpose of this study was to identify the recent pattern of work-related burns from 2001 to 2010 and to compare it to the previous studies. During the study period, 1427 patients were admitted for an acute injury to the regional burn centre. Of these, 330 were due to a work-related incident (23%). The mean age of patients was 40.5±11.9 years, 95% were male. The mean total body surface area burn was 11.9±16.2%. The most common mechanism of burn injury was flame (32.7%) followed by electrical (27%) and scald (19.7%), inhalation injury was present in 4.8% of patients and the mortality was 1.8%. Our study has shown that there has been a significant decrease in the incidence in work-related burns treated at the regional burn centre (23.1%, vs. 28.2% vs. 30.2% pburns have now become the leading cause of injury, there was a significant reduction in inhalation injury (4.8% vs. 23% vs. 14.8%, pburns, improvement in burn care, and that prevention strategies may have been more effective. PMID:23352030

Adult survivors' lived experience of burns and post-burn health: A qualitative analysis.

Science.gov (United States)

Abrams, Thereasa E; Ogletree, Roberta J; Ratnapradipa, Dhitinut; Neumeister, Michael W

2016-02-01

The individual implications of major burns are likely to affect the full spectrum of patients' physical, emotional, psychological, social, environmental, spiritual and vocational health. Yet, not all of the post-burn health implications are inevitably negative. Utilizing a qualitative approach, this heuristic phenomenological study explores the experiences and perceptions early (ages 18-35) and midlife (ages 36-64) adults providing insight for how participants perceived their burns in relationship to their post-burn health. Participants were interviewed using semi-structured interview questions framed around seven domains of health. Interview recordings were transcribed verbatim then coded line by line, identifying dominant categories related to health. Categories were analyzed identifying shared themes among the study sample. Participants were Caucasian, seven males and one female. Mean age at time of interviews was 54.38 and 42.38 at time of burns. Mean time since burns occurred was 9.38 years with a minimum of (20%) total body surface area (TBSA) burns. Qualitative content analysis rendered three emergent health-related categories and associated themes that represented shared meanings within the participant sample. The category of "Physical Health" reflected the theme physical limitations, pain and sensitivity to temperature. Within the category of "Intellectual Health" were themes of insight, goal setting and self-efficacy, optimism and humor and within "Emotional Health" were the themes empathy and gratitude. By exploring subjective experiences and perceptions of health shared through dialog with experienced burned persons, there are opportunities to develop a more complete picture of how holistic health may be affected by major burns that in turn could support future long-term rehabilitative trajectories of early and midlife adult burn patients. Copyright © 2015 Elsevier Ltd and ISBI. All rights reserved.

Modelling of the thermomechanical and physical processes in FR fuel pins using the GERMINAL code

International Nuclear Information System (INIS)

Roche, L.; Pelletier, M.

2000-01-01

In the frame of the R and D on Fast Reactor mixed oxide fuels, CEA/DEC has developed the computer code GERMINAL for studying fuel pin thermal and mechanical behaviour, both during steady-state and incidental conditions, up to high burn-up (25 at%). The first part of this paper is devoted to the description of the main models: fuel evolution (central hole and porosity evolution, Plutonium redistribution, O/M radial profile, transient gas swelling, melting fuel behaviour, minor actinides production), high burn-up models (fission gas, volatile fission products and JOG formation), fuel-cladding heat transfer, fuel-cladding mechanical interaction. The second part gives some examples of calculation results taken from the GERMINAL validation data base (more than 40 experiments from PHENIX, PFR, CABRI reactors), with special emphasis on: local fission gas retention and global release, fuel geometry evolution, radial redistribution of plutonium for high burn-up fuels, solid and annular fuel behaviour during power ramps including fuel melting, helium formation from MA (Am and Np) doped homogeneous fuels. (author)

SPECTRAL AMPLITUDE CODING OCDMA SYSTEMS USING ENHANCED DOUBLE WEIGHT CODE

Directory of Open Access Journals (Sweden)

F.N. HASOON

2006-12-01

Full Text Available A new code structure for spectral amplitude coding optical code division multiple access systems based on double weight (DW code families is proposed. The DW has a fixed weight of two. Enhanced double-weight (EDW code is another variation of a DW code family that can has a variable weight greater than one. The EDW code possesses ideal cross-correlation properties and exists for every natural number n. A much better performance can be provided by using the EDW code compared to the existing code such as Hadamard and Modified Frequency-Hopping (MFH codes. It has been observed that theoretical analysis and simulation for EDW is much better performance compared to Hadamard and Modified Frequency-Hopping (MFH codes.

110 °C range athermalization of wavefront coding infrared imaging systems

Science.gov (United States)

Feng, Bin; Shi, Zelin; Chang, Zheng; Liu, Haizheng; Zhao, Yaohong

2017-09-01

110 °C range athermalization is significant but difficult for designing infrared imaging systems. Our wavefront coding athermalized infrared imaging system adopts an optical phase mask with less manufacturing errors and a decoding method based on shrinkage function. The qualitative experiments prove that our wavefront coding athermalized infrared imaging system has three prominent merits: (1) working well over a temperature range of 110 °C; (2) extending the focal depth up to 15.2 times; (3) achieving a decoded image being approximate to its corresponding in-focus infrared image, with a mean structural similarity index (MSSIM) value greater than 0.85.

Particle and heavy ion transport code system; PHITS

International Nuclear Information System (INIS)

Niita, Koji

2004-01-01

Intermediate and high energy nuclear data are strongly required in design study of many facilities such as accelerator-driven systems, intense pulse spallation neutron sources, and also in medical and space technology. There is, however, few evaluated nuclear data of intermediate and high energy nuclear reactions. Therefore, we have to use some models or systematics for the cross sections, which are essential ingredients of high energy particle and heavy ion transport code to estimate neutron yield, heat deposition and many other quantities of the transport phenomena in materials. We have developed general purpose particle and heavy ion transport Monte Carlo code system, PHITS (Particle and Heavy Ion Transport code System), based on the NMTC/JAM code by the collaboration of Tohoku University, JAERI and RIST. The PHITS has three important ingredients which enable us to calculate (1) high energy nuclear reactions up to 200 GeV, (2) heavy ion collision and its transport in material, (3) low energy neutron transport based on the evaluated nuclear data. In the PHITS, the cross sections of high energy nuclear reactions are obtained by JAM model. JAM (Jet AA Microscopic Transport Model) is a hadronic cascade model, which explicitly treats all established hadronic states including resonances and all hadron-hadron cross sections parametrized based on the resonance model and string model by fitting the available experimental data. The PHITS can describe the transport of heavy ions and their collisions by making use of JQMD and SPAR code. The JQMD (JAERI Quantum Molecular Dynamics) is a simulation code for nucleus nucleus collisions based on the molecular dynamics. The SPAR code is widely used to calculate the stopping powers and ranges for charged particles and heavy ions. The PHITS has included some part of MCNP4C code, by which the transport of low energy neutron, photon and electron based on the evaluated nuclear data can be described. Furthermore, the high energy nuclear

Modification in the FUDA computer code to predict fuel performance at high burnup

Energy Technology Data Exchange (ETDEWEB)

Das, M; Arunakumar, B V; Prasad, P N [Nuclear Power Corp., Mumbai (India)

1997-08-01

The computer code FUDA (FUel Design Analysis) participated in the blind exercises organized by the IAEA CRP (Co-ordinated Research Programme) on FUMEX (Fuel Modelling at Extended Burnup). While the code prediction compared well with the experiments at Halden under various parametric and operating conditions, the fission gas release and fission gas pressure were found to be slightly over-predicted, particularly at high burnups. In view of the results of 6 FUMEX cases, the main models and submodels of the code were reviewed and necessary improvements were made. The new version of the code FUDA MOD 2 is now able to predict fuel performance parameter for burn-ups up to 50000 MWD/TeU. The validation field of the code has been extended to prediction of thorium oxide fuel performance. An analysis of local deformations at pellet interfaces and near the end caps is carried out considering the hourglassing of the pellet by finite element technique. (author). 15 refs, 1 fig.

Modification in the FUDA computer code to predict fuel performance at high burnup

International Nuclear Information System (INIS)

Das, M.; Arunakumar, B.V.; Prasad, P.N.

1997-01-01

The computer code FUDA (FUel Design Analysis) participated in the blind exercises organized by the IAEA CRP (Co-ordinated Research Programme) on FUMEX (Fuel Modelling at Extended Burnup). While the code prediction compared well with the experiments at Halden under various parametric and operating conditions, the fission gas release and fission gas pressure were found to be slightly over-predicted, particularly at high burnups. In view of the results of 6 FUMEX cases, the main models and submodels of the code were reviewed and necessary improvements were made. The new version of the code FUDA MOD 2 is now able to predict fuel performance parameter for burn-ups up to 50000 MWD/TeU. The validation field of the code has been extended to prediction of thorium oxide fuel performance. An analysis of local deformations at pellet interfaces and near the end caps is carried out considering the hourglassing of the pellet by finite element technique. (author). 15 refs, 1 fig

A ranking system for prescribed burn prioritization in Table Mountain National Park, South Africa.

Science.gov (United States)

Cowell, Carly Ruth; Cheney, Chad

2017-04-01

To aid prescribed burn decision making in Table Mountain National Park, in South Africa a priority ranking system was tested. Historically a wildfire suppression strategy was adopted due to wildfires threatening urban areas close to the park, with few prescribed burns conducted. A large percentage of vegetation across the park exceeded the ecological threshold of 15 years. We held a multidisciplinary workshop, to prioritize areas for prescribed burning. Fire Management Blocks were mapped and assessed using the following seven categories: (1) ecological, (2) management, (3) tourism, (4) infrastructure, (5) invasive alien vegetation, (6) wildland-urban interface and (7) heritage. A priority ranking system was used to score each block. The oldest or most threatened vegetation types were not necessarily the top priority blocks. Selected blocks were burnt and burning fewer large blocks proved more effective economically, ecologically and practically due to the limited burning days permitted. The prioritization process was efficient as it could be updated annually following prescribed burns and wildfire incidents. Integration of prescribed burn planning and wildfire suppression strategies resulted in a reduction in operational costs. We recommend protected areas make use of a priority ranking system developed with expert knowledge and stakeholder engagement to determine objective prescribed burn plans. Copyright © 2017 Elsevier Ltd. All rights reserved.

Elements of algebraic coding systems

CERN Document Server

Cardoso da Rocha, Jr, Valdemar

2014-01-01

Elements of Algebraic Coding Systems is an introductory text to algebraic coding theory. In the first chapter, you'll gain inside knowledge of coding fundamentals, which is essential for a deeper understanding of state-of-the-art coding systems. This book is a quick reference for those who are unfamiliar with this topic, as well as for use with specific applications such as cryptography and communication. Linear error-correcting block codes through elementary principles span eleven chapters of the text. Cyclic codes, some finite field algebra, Goppa codes, algebraic decoding algorithms, and applications in public-key cryptography and secret-key cryptography are discussed, including problems and solutions at the end of each chapter. Three appendices cover the Gilbert bound and some related derivations, a derivation of the Mac- Williams' identities based on the probability of undetected error, and two important tools for algebraic decoding-namely, the finite field Fourier transform and the Euclidean algorithm f...

Comparative Analysis of Single and Dual Irradiation Pass of Deep Burn High Temperature Reactor Scenario

International Nuclear Information System (INIS)

Jeong, Chang Joon; Jo, Chang Keun; Noh, Jae Man

2012-01-01

A concept of a deep-burn (DB) of trans uranic (TRU) elements in a high temperature reactor (HTR) has been proposed and studied with a single irradiation pass. However, there is still a significant amount of TRU after burn in an HTR. Therefore, it is necessary to burn more TRU in a fast reactor (FR) with repeated reprocessing such as a pyro-process. In this study, the fuel cycle calculations are performed and the results are compared for a singlepass DB-HHR scenario and a dual-pass sodium-cooled fast reactor (SFR) scenario. For the analysis, front-end and back-end parameters are compared. The calculations were performed by the DANESS (Dynamic Analysis of Nuclear Energy System Strategies), which is an integrated system dynamic fuel cycle analysis code

Code of a Tokamak Fusion Energy Facility ITER

International Nuclear Information System (INIS)

Yasuhide Asada; Kenzo Miya; Kazuhiko Hada; Eisuke Tada

2002-01-01

The technical structural code for ITER (International Thermonuclear Experimental Fusion Reactor) and, as more generic applications, for D-T burning fusion power facilities (hereafter, Fusion Code) should be innovative because of their quite different features of safety and mechanical components from nuclear fission reactors, and the necessity of introducing several new fabrication and examination technologies. Introduction of such newly developed technologies as inspection-free automatic welding into the Fusion Code is rationalized by a pilot application of a new code concept of s ystem-based code for integrity . The code concept means an integration of element technical items necessary for construction, operation and maintenance of mechanical components of fusion power facilities into a single system to attain an optimization of the total margin of these components. Unique and innovative items of the Fusion Code are typically as follows: - Use of non-metals; - Cryogenic application; - New design margins on allowable stresses, and other new design rules; - Use of inspection-free automatic welding, and other newly developed fabrication technologies; - Graded approach of quality assurance standard to cover radiological safety-system components as well as non-safety-system components; - Consideration on replacement components. (authors)

A Novel Criterion for Optimum MultilevelCoding Systems in Mobile Fading Channels

Institute of Scientific and Technical Information of China (English)

YUAN Dongfeng; WANG Chengxiang; YAO Qi; CAO Zhigang

2001-01-01

A novel criterion that is "capac-ity rule" and "mapping rule" for the design of op-timum MLC scheme over mobile fading channels isproposed.According to this theory,the performanceof multilevel coding with multistage decoding schemes(MLC/MSD) in mobile fading channels is investi-gated,in which BCH codes are chosen as componentcodes,and three mapping strategies with 8ASK mod-ulation are used.Numerical results indicate that whencode rates of component codes in MLC scheme are de-signed based on "capacity rule",the performance ofthe system with block partitioning (BP) is optimumfor Rayleigh fading channels,while the performance ofthe system with Ungerboeck partioning (UP) is bestfor AWGN channels.

Application of MCNP code in shielding calculation of minitype fast reactor

International Nuclear Information System (INIS)

He Keyu; Han Weishi

2008-01-01

An accurate shielding calculation model has been set up for the minitype sodium-cooled fast reactor (MFR) based on MCNP code and particular calculation of its primary shielding parameters has been carried out. The results indicate that the photon and neutron flux density of MFR has rapidly fallen to a low-level. The material for the shielding layer outside of main container is primarily of carbon steel, which can be design as a shielding structure satisfying the safety code. The sodium activation in primary circuit is extremely limited and it is simple to shield from. Both the output of helium in reflector and burn up of boron-10 in control rod are very small. These materials can be used for several cycle lives. (authors)

The applicability of ALPHA/PHOENIX/ANC nuclear design code system on Korean standard PWR's

International Nuclear Information System (INIS)

Lee, Kookjong; Choi, Kie-Yong; Lee, Hae-Chan; Roh, Eun-Rae

1996-01-01

For the Korean Standard Nuclear Power Plant (KSNPP) designed based on Combustion Engineering (CE) System 80, the Westinghouse nuclear design code system ALPHA/PHOENIX/ANC was applied to the follow-up design of initial and reload core of KSNPP. The follow-up design results of Yonggwang Unit 3 Cycle 1, 2 and Yonggwang Unit 4 Cycle 1 have shown good agreements with the measured data. The assemblywise power distributions have shown less than 2% average differences and critical boron concentrations have shown less than 20 ppm differences. All the low power physics test parameters are in good agreement. Consequently, APA design code system can be applied to KNSPP cores. (author)

Simulation of water hammer phenomena using the system code ATHLET

Energy Technology Data Exchange (ETDEWEB)

Bratfisch, Christoph; Koch, Marco K. [Bochum Univ. (Germany). Reactor Simulation and Safety Group

2017-07-15

Water Hammer Phenomena can endanger the integrity of structures leading to a possible failure of pipes in nuclear power plants as well as in many industrial applications. These phenomena can arise in nuclear power plants in the course of transients and accidents induced by the start-up of auxiliary feed water systems or emergency core cooling systems in combination with rapid acting valves and pumps. To contribute to further development and validation of the code ATHLET (Analysis of Thermalhydraulics of Leaks and Transients), an experiment performed in the test facility Pilot Plant Pipework (PPP) at Fraunhofer UMSICHT is simulated using the code version ATHLET 3.0A.

Simulation of water hammer phenomena using the system code ATHLET

International Nuclear Information System (INIS)

Bratfisch, Christoph; Koch, Marco K.

2017-01-01

Water Hammer Phenomena can endanger the integrity of structures leading to a possible failure of pipes in nuclear power plants as well as in many industrial applications. These phenomena can arise in nuclear power plants in the course of transients and accidents induced by the start-up of auxiliary feed water systems or emergency core cooling systems in combination with rapid acting valves and pumps. To contribute to further development and validation of the code ATHLET (Analysis of Thermalhydraulics of Leaks and Transients), an experiment performed in the test facility Pilot Plant Pipework (PPP) at Fraunhofer UMSICHT is simulated using the code version ATHLET 3.0A.

An Adaptation of the HELIOS/MASTER Code System to the Analysis of VHTR Cores

International Nuclear Information System (INIS)

Noh, Jae Man; Lee, Hyun Chul; Kim, Kang Seog; Kim, Yong Hee

2006-01-01

KAERI is developing a new computer code system for an analysis of VHTR cores based on the existing HELIOS/MASTER code system which was originally developed for a LWR core analysis. In the VHTR reactor physics, there are several unique neutronic characteristics that cannot be handled easily by the conventional computer code system applied for the LWR core analysis. Typical examples of such characteristics are a double heterogeneity problem due to the particulate fuels, the effects of a spectrum shift and a thermal up-scattering due to the graphite moderator, and a strong fuel/reflector interaction, etc. In order to facilitate an easy treatment of such characteristics, we developed some methodologies for the HELIOS/MASTER code system and tested their applicability to the VHTR core analysis

Measuring device for the distribution of burn-up degree in fuel assembly irradiated in nuclear reactor

International Nuclear Information System (INIS)

Kumanomido, Hironori

1989-01-01

The object of the invention is to measure the distribution of burn-up degree, of fuel assemblies irradiated in a nuclear reactor in a short time and exactly. That is, the device comprises a device main body having substantially the same length as that for the axial length of a fuel assembly and a detector container disposed axially slidably to the main body. A plurality of radiation detectors are arranged at an equi-axial pitch and contained in the container. The container is caused to slide at a pitch equal to the equi-axial distance of the detectors. In the device having thus been constituted, measurement is conducted at least for twice at an axial position on the side of a fuel assembly irradiated in the nuclear reactor and a position caused to slide therefrom by one pitch. Based on the result, the sensitivities between each of the detectors are compared and the relative sensitivity of the radiation detectors is calibrated. Accordingly, the sensitivity between each of the detectors can be calibrated rapidly and easily. As a result, the distribution of the burn-up degree, etc of irradiated fuel assembly can be measured exactly. (K.M.)

Burn-up measurements on nuclear reactor fuels using high performance liquid chromatography

International Nuclear Information System (INIS)

Sivaraman, N.; Subramaniam, S.; Srinivasan, T.G.; Vasudeva Rao, P.R.

2002-01-01

Burn-up measurements on thermal as well as fast reactor fuels were carried out using high performance liquid chromatography (HPLC). A column chromatographic technique using di-(2-ethylhexyl) phosphoric acid (HDEHP) coated column was employed for the isolation of lanthanides from uranium, plutonium and other fission products. Ion-pair HPLC was used for the separation of individual lanthanides. The atom percent fissions were calculated from the concentrations of the lanthanide (neodymium in the case of thermal reactor and lanthanum for the fast reactor fuels) and from uranium and plutonium contents of the dissolver solutions. The HPLC method was also used for determining the fractional fissions from uranium and plutonium for the thermal reactor fuel. (author)

Reduction on high level radioactive waste volume and geological repository footprint with high burn-up and high thermal efficiency of HTGR

Energy Technology Data Exchange (ETDEWEB)

Fukaya, Yuji, E-mail: fukaya.yuji@jaea.go.jp; Nishihara, Tetsuo

2016-10-15

Highlights: • We evaluate the number of canisters and its footprint for HTGR. • We proposed new waste loading method for direct disposal of HTGR. • HTGR can significantly reduce HLW volume compared with LWR. - Abstract: Reduction on volume of High Level radioactive Waste (HLW) and footprint in a geological repository due to high burn-up and high thermal efficiency of High Temperature Gas-cooled Reactor (HTGR) has been investigated. A helium-cooled and graphite-moderated commercial HTGR was designed as a Gas Turbine High Temperature Reactor (GTHTR300), and that has particular features such as significantly high burn-up of approximately 120 GWd/t, high thermal efficiency around 50%, and pin-in-block type fuel. The pin-in-block type fuel was employed to reduce processed graphite volume in reprocessing. By applying the feature, effective waste loading method for direct disposal is proposed in this study. By taking into account these feature, the number of HLW canister generations and its repository footprint are evaluated by burn-up fuel composition, thermal calculation and criticality calculation in repository. As a result, it is found that the number of canisters and its repository footprint per electricity generation can be reduced by 60% compared with Light Water Reactor (LWR) representative case for direct disposal because of the higher burn-up, higher thermal efficiency, less TRU generation, and effective waste loading proposed in this study for HTGR. But, the reduced ratios change to 20% and 50% if the long term durability of LWR canister is guaranteed. For disposal with reprocessing, the number of canisters and its repository footprint per electricity generation can be reduced by 30% compared with LWR because of the 30% higher thermal efficiency of HTGR.

System Based Code: Principal Concept

International Nuclear Information System (INIS)

Yasuhide Asada; Masanori Tashimo; Masahiro Ueta

2002-01-01

This paper introduces a concept of the 'System Based Code' which has initially been proposed by the authors intending to give nuclear industry a leap of progress in the system reliability, performance improvement, and cost reduction. The concept of the System Based Code intends to give a theoretical procedure to optimize the reliability of the system by administrating every related engineering requirement throughout the life of the system from design to decommissioning. (authors)

Quantitative assessment of graded burn wounds using a commercial and research grade laser speckle imaging (LSI) system

Science.gov (United States)

Ponticorvo, A.; Rowland, R.; Yang, B.; Lertsakdadet, B.; Crouzet, C.; Bernal, N.; Choi, B.; Durkin, A. J.

2017-02-01

Burn wounds are often characterized by injury depth, which then dictates wound management strategy. While most superficial burns and full thickness burns can be diagnosed through visual inspection, clinicians experience difficulty with accurate diagnosis of burns that fall between these extremes. Accurately diagnosing burn severity in a timely manner is critical for starting the appropriate treatment plan at the earliest time points to improve patient outcomes. To address this challenge, research groups have studied the use of commercial laser Doppler imaging (LDI) systems to provide objective characterization of burn-wound severity. Despite initial promising findings, LDI systems are not commonplace in part due to long acquisition times that can suffer from artifacts in moving patients. Commercial LDI systems are being phased out in favor of laser speckle imaging (LSI) systems that can provide similar information with faster acquisition speeds. To better understand the accuracy and usefulness of commercial LSI systems in burn-oriented research, we studied the performance of a commercial LSI system in three different sample systems and compared its results to a research-grade LSI system in the same environments. The first sample system involved laboratory measurements of intralipid (1%) flowing through a tissue simulating phantom, the second preclinical measurements in a controlled burn study in which wounds of graded severity were created on a Yorkshire pig, and the third clinical measurements involving a small sample of clinical patients. In addition to the commercial LSI system, a research grade LSI system that was designed and fabricated in our labs was used to quantitatively compare the performance of both systems and also to better understand the "Perfusion Unit" output of commercial systems.

UK regulatory perspective on the application of burn-up credit to the BNFL thorp head end plant

International Nuclear Information System (INIS)

Simister, D.N.; Clemson, P.D.

2003-01-01

In the UK the Health and Safety Executive, which incorporates the Nuclear Installations Inspectorate (NII), is responsible for regulation of safety on nuclear sites. This paper reports progress made in the application and development of a UK regulatory position for assessing licensee's plant safety caes which invoke the use of Burn-up Credit for criticality applications. The NII's principles and strategy for the assessment of this technical area have been developed over a period of time following expressions of interest from UK industry and subsequent involvement in the international collaborations and debate in this area. This experience has now been applied to the first main plant safety case application claiming Burn-up Credit. This case covers the BNFL Thermal Oxide Reprocessing Plant (THORP) dissolver at Sellafield, where dissolved gadolinium neutron poison is used as a criticality control. The case argues for a reduction in gadolinium content by taking credit for the burn-up of input fuel. The UK regulatory process, assessment principles and criteria are briefly outlined, showing the regulatory framework used to review the case. These issues include the fundamental requirement in UK Health and Safety law to demonstrate that risks have been reduced to as low as reasonably practicable (ALARP), the impact on safety margins, compliance and operability procedures, and the need for continuing review. Novel features of methodology, using a ''Residual Enrichment'' and ''Domain Boundary'' approach, were considered and accepted. The underlying validation, both of criticality methodology and isotopic determination, was also reviewed. Compliance was seen to rely heavily on local in-situ measurements of spent fuel used to determine ''Residual Enrichment'' and other parameters, requiring review of the development and basis of the correlations used to underpin the measurement process. Overall, it was concluded that the case as presented was adequate. Gadolinium reduction

Criticality calculations of a generic fuel container for fuel assemblies PWR, by means of the code MCNP

International Nuclear Information System (INIS)

Vargas E, S.; Esquivel E, J.; Ramirez S, J. R.

2013-10-01

The purpose of the concept of burned consideration (Burn-up credit) is determining the capacity of the calculation codes, as well as of the nuclear data associates to predict the isotopic composition and the corresponding neutrons effective multiplication factor in a generic container of spent fuel during some time of relevant storage. The present work has as objective determining this capacity of the calculation code MCNP in the prediction of the neutrons effective multiplication factor for a fuel assemblies arrangement type PWR inside a container of generic storage. The calculations are divided in two parts, the first, in the decay calculations with specified nuclide concentrations by the reference for a pressure water reactor (PWR) with enriched fuel to 4.5% and a discharge burned of 50 GW d/Mtu. The second, in criticality calculations with isotopic compositions dependent of the time for actinides and important fission products, taking 30 time steps, for two actinide groups and fission products. (Author)

Self-Burns in Fars Province, Southern Iran

OpenAIRE

Mohammadi, Ali Akbar; Tohidinik, Hamid Reza; Zardosht, Mitra; Seyed Jafari, Seyed Morteza

2016-01-01

BACKGROUND The alarming incidence of self- burning provoked to set up a multidisciplinary preventive program to decrease the incidence and complications of this harmful issue. This study investigated the incidence and the preventive measures in self-burn in Fars Province, southern Iran. METHODS This study was a longitudinal prospective design on trend of self-inflicted burn injuries in Fars province after setting up a regional multidisciplinary preventive plan (2009-2012). RESULTS From 18862 ...

The symbol coding language for the BUTs processor of in-core reactor control systems

International Nuclear Information System (INIS)

Vorob'ev, D.M.; Golovanov, M.N.; Levin, G.L.; Parfenova, T.K.; Filatov, V.P.

1978-01-01

A symbolic coding language is described; it has been developed for automation of making up programs for in-core control systems. The systems use the ideology of the CAMAC-VECTOR system and include the BUTs-20 processor. The symbolic coding language has been developed as a programming language of the ASSEMBLER type. Operators of instructions and pseudo-instructions, the rules of reading in the text of the source program, and operator record formats are considered

Development of a general coupling interface for the fuel performance code transuranus tested with the reactor dynamic code DYN3D

International Nuclear Information System (INIS)

Holt, L.; Rohde, U.; Seidl, M.; Schubert, A.; Van Uffelen, P.

2013-01-01

Several institutions plan to couple the fuel performance code TRANSURANUS developed by the European Institute for Transuranium Elements with their own codes. One of these codes is the reactor dynamic code DYN3D maintained by the Helmholtz-Zentrum Dresden - Rossendorf. DYN3D was developed originally for VVER type reactors and was extended later to western type reactors. Usually, the fuel rod behavior is modeled in thermal hydraulics and neutronic codes in a simplified manner. The main idea of this coupling is to describe the fuel rod behavior in the frame of core safety analysis in a more detailed way, e.g. including the influence of the high burn-up structure, geometry changes and fission gas release. It allows to take benefit from the improved computational power and software achieved over the last two decades. The coupling interface was developed in a general way from the beginning. Thence it can be easily used also by other codes for a coupling with TRANSURANUS. The user can choose between a one-way as well as a two-way online coupling option. For a one-way online coupling, DYN3D provides only the time-dependent rod power and thermal hydraulics conditions to TRANSURANUS, but the fuel performance code doesn’t transfer any variable back to DYN3D. In a two-way online coupling, TRANSURANUS in addition transfers parameters like fuel temperature and cladding temperature back to DYN3D. This list of variables can be extended easily by geometric and further variables of interest. First results of the code system DYN3D-TRANSURANUS will be presented for a control rod ejection transient in a modern western type reactor. Pre-analyses show already that a detailed fuel rod behavior modeling will influence the thermal hydraulics and thence also the neutronics due to the Doppler reactivity effect of the fuel temperature. The coupled code system has therefore a potential to improve the assessment of safety criteria. The developed code system DYN3D-TRANSURANUS can be used also

Next generation Zero-Code control system UI

CERN Multimedia

CERN. Geneva

2017-01-01

Developing ergonomic user interfaces for control systems is challenging, especially during machine upgrade and commissioning where several small changes may suddenly be required. Zero-code systems, such as *Inspector*, provide agile features for creating and maintaining control system interfaces. More so, these next generation Zero-code systems bring simplicity and uniformity and brake the boundaries between Users and Developers. In this talk we present *Inspector*, a CERN made Zero-code application development system, and we introduce the major differences and advantages of using Zero-code control systems to develop operational UI.

MONTEBURNS 2.0: An Automated, Multi-Step Monte Carlo Burnup Code System

International Nuclear Information System (INIS)

2007-01-01

) Minor corrections to the output file. B - Methods - MONTEBURNS processes input from the user that specifies the system geometry, initial material compositions, feed/removal specifications, and other code-specific parameters. Various results from MCNP, ORIGEN2, and other calculations are then output successively as the code runs. The principle function of MONTEBURNS is to transfer one-group cross-section and flux values from MCNP to ORIGEN2, and then transfer the resulting material compositions (after irradiation and/or decay) from ORIGEN2 back to MCNP in a repeated, cyclic fashion (a simple predictor-corrector method is used during this process). C - Restrictions on the complexity of the problem: The basic requirement of the code is that the user have working versions of PERL, MCNP, and either CINDER90, ORIGEN2.1, or ORIGEN2.2. The code is fairly versatile and allows any number of irradiation (burn steps) to occur, up to 49 materials to be irradiated, and material to be added or removed at each step. More detailed information about limitation is in Section 8.0 of the MONTEBURNS User's Manual (LA-UR-99-4999)

Numerical simulation code for combustion of sodium liquid droplet and its verification

International Nuclear Information System (INIS)

Okano, Yasushi

1997-11-01

The computer programs for sodium leak and burning phenomena had been developed based on mechanistic approach. Direct numerical simulation code for sodium liquid droplet burning had been developed for numerical analysis of droplet combustion in forced convection air flow. Distributions of heat generation and temperature and reaction rate of chemical productions, such as sodium oxide and hydroxide, are calculated and evaluated with using this numerical code. Extended MAC method coupled with a higher-order upwind scheme had been used for combustion simulation of methane-air mixture. In the numerical simulation code for combustion of sodium liquid droplet, chemical reaction model of sodium was connected with the extended MAC method. Combustion of single sodium liquid droplet was simulated in this report for the verification of developed numerical simulation code. The changes of burning rate and reaction product with droplet diameter and inlet wind velocity were investigated. These calculation results were qualitatively and quantitatively conformed to the experimental and calculation observations in combustion engineering. It was confirmed that the numerical simulation code was available for the calculation of sodium liquid droplet burning. (author)

Influence of fuel element burn-up on the power peaking factor in PWR; Vpliv zgorelosti gorivnega elementa na konicne faktorje moci v tlacnovodnem reaktorju

Energy Technology Data Exchange (ETDEWEB)

Ravnik, M; Mele, I [Institut ' Jozef Stefan' , Ljubljana (Yugoslavia); Falkowski, J [Institut energii atomowel, Swierk (Poland)

1988-07-01

Influence of fuel element burn-up distribution on radial power peaking factors is presented for Krsko NPP. The effect is strong for elements loaded in the periphery of the core with large power gradients. Neglecting the burn-up distributions inside fuel elements leads to {+-} 5% error on power peaking factor of the same element and {+-} 2% at other locations in the core. Influence on k is observed due to perturbed leakage from the core and due to redistribution of the importance function of the core. (author)

Development of a method for xenon determination in the microstructure of high burn-up nuclear fuel

International Nuclear Information System (INIS)

Horvath, M. I.

2008-01-01

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, relevant Xe

IFPE/IFA-597.3, centre-line temperature, fission gas release and clad elongation at high burn-up (60-62 MWd/kg)

International Nuclear Information System (INIS)

Turnbull, J.A.

2003-01-01

Description: The fuel segments for the high burn-up integral rod behaviour test IFA-597 were taken from fuel rod 33-25065, which was irradiated in the Ringhals 1 BWR for approximately 12 years. The irradiation of this rod and its sibling rod 33-25046 was performed in two stages. During the first irradiation, 1980 to 1986, the rods were part of Ringhals assembly 6477 and an approximate rod averaged burn-up of 31 MWd/kg UO 2 was reached. The rods were then placed into fuel assembly 9902 for a second period of irradiation from 1986 to 1992. The location of the fuel rods 33-25065 and 33-25046 in this assembly were in positions 9902/D and 9902/E4 respectively. A final rod averaged burn-up of 52 MWd/kg UO 2 was achieved. The burn-up at the location of the Halden segments was estimated as 59 MWd/kg UO 2 , well beyond the formation of High Burn-up Structure (Hobs) formation at the pellet rim. At the rim, the burn-up was estimated as 130 MWd/kg UO 2 . After commercial irradiation, PIE was performed at Studsvik. Inner and outer clad oxide thickness measurements were 42 and 5 microns respectively. The measured cold rod diameter varied between 12.20 and 12.25 mm, thus only a small amount of creep-down had occurred from the original diameter of 12.25 mm. Cold gap measurements were taken by diametral compression of the clad onto the fuel. The stiffness changes twice during these measurements, the first (relocated gap) associated with the onset of pellet fragment movement, the second (compressed gap) when the fragments are together and the pellet is compressed. For these rods, the compressed diametral gap was measured as 30 microns. This is in agreement with the pellet and cladding being in contact during the final irradiation cycle, i.e., at ∼12 kW/m. FGR measurements were made after puncturing and values of 2.5%-3.3% were calculated from the extracted gas. The uncertainty is due to different methods of calculation. Ceramography showed a normal crack pattern and no evidence of

A mean field theory of coded CDMA systems

International Nuclear Information System (INIS)

Yano, Toru; Tanaka, Toshiyuki; Saad, David

2008-01-01

We present a mean field theory of code-division multiple-access (CDMA) systems with error-control coding. On the basis of the relation between the free energy and mutual information, we obtain an analytical expression of the maximum spectral efficiency of the coded CDMA system, from which a mean-field description of the coded CDMA system is provided in terms of a bank of scalar Gaussian channels whose variances in general vary at different code symbol positions. Regular low-density parity-check (LDPC)-coded CDMA systems are also discussed as an example of the coded CDMA systems

A mean field theory of coded CDMA systems

Energy Technology Data Exchange (ETDEWEB)

Yano, Toru [Graduate School of Science and Technology, Keio University, Hiyoshi, Kohoku-ku, Yokohama-shi, Kanagawa 223-8522 (Japan); Tanaka, Toshiyuki [Graduate School of Informatics, Kyoto University, Yoshida Hon-machi, Sakyo-ku, Kyoto-shi, Kyoto 606-8501 (Japan); Saad, David [Neural Computing Research Group, Aston University, Birmingham B4 7ET (United Kingdom)], E-mail: yano@thx.appi.keio.ac.jp

2008-08-15

We present a mean field theory of code-division multiple-access (CDMA) systems with error-control coding. On the basis of the relation between the free energy and mutual information, we obtain an analytical expression of the maximum spectral efficiency of the coded CDMA system, from which a mean-field description of the coded CDMA system is provided in terms of a bank of scalar Gaussian channels whose variances in general vary at different code symbol positions. Regular low-density parity-check (LDPC)-coded CDMA systems are also discussed as an example of the coded CDMA systems.

Electronic nicotine delivery system (ENDS) battery-related burns presenting to US emergency departments, 2016.

Science.gov (United States)

Corey, Catherine G; Chang, Joanne T; Rostron, Brian L

2018-03-05

Currently, an estimated 7.9 million US adults use electronic nicotine delivery systems (ENDS). Although published reports have identified fires and explosions related to use of ENDS since 2009, these reports do not provide national estimates of burn injuries associated with ENDS batteries in the US. We analyzed nationally representative data provided in the National Electronic Injury Surveillance System (NEISS) to estimate the number of US emergency department (ED) visits for burn injuries associated with ENDS batteries. We reviewed the case narrative field to gain additional insights into the circumstances of the burn injury. In 2016, 26 ENDS battery-related burn cases were captured by NEISS, which translates to a national estimate of 1007 (95%CI: 357-1657) injuries presenting in US EDs. Most of the burns were thermal burns (80.4%) and occurred to the upper leg/lower trunk (77.3%). Examination of the case narrative field indicated that at least 20 of the burn injuries occurred while ENDS batteries were in the user's pocket. Our study provides valuable information for understanding the current burden of ENDS battery-related burn injuries treated in US EDs. The nature and circumstances of the injuries suggest these incidents were unintentional and would potentially be prevented through battery design requirements, battery testing standards and public education related to ENDS battery safety.

SASSYS LMFBR systems analysis code

International Nuclear Information System (INIS)

Dunn, F.E.; Prohammer, F.G.

1982-01-01

The SASSYS code provides detailed steady-state and transient thermal-hydraulic analyses of the reactor core, inlet and outlet coolant plenums, primary and intermediate heat-removal systems, steam generators, and emergency shut-down heat removal systems in liquid-metal-cooled fast-breeder reactors (LMFBRs). The main purpose of the code is to analyze the consequences of failures in the shut-down heat-removal system and to determine whether this system can perform its mission adequately even with some of its components inoperable. The code is not plant-specific. It is intended for use with any LMFBR, using either a loop or a pool design, a once-through steam generator or an evaporator-superheater combination, and either a homogeneous core or a heterogeneous core with internal-blanket assemblies

Coupling analysis of deformation and thermal-hydraulics in a FBR fuel pin bundle using BAMBOO and ASFRE-IV Codes

International Nuclear Information System (INIS)

Ito, Masahiro; Imai, Yasutomo; Uwaba, Tomoyuki; Ohshima, Hiroyuki

2004-03-01

The bundle-duct interaction may occur in sodium cooled wire-wrapped FBR fuel subassemblies in high burn-up conditions. JNC has been developing a bundle deformation analysis code BAMBOO (Behavior Analysis code for Mechanical interaction of fuel Bundle under On-power Operation), a thermal hydraulics analysis code ASFRE-IV (Analysis of Sodium Flow in Reactor Elements - ver. IV) and their coupling method as a simulation system for the evaluation on the integrity of deformed FBR fuel pin bundles. In this study, the simulation system was applied to a coupling analysis of deformation and thermal-hydraulics in the fuel pin-bundle under a steady-state condition just after startup for the purpose of the verification of the simulation system. The iterative calculations of deformation and thermal-hydraulics employed in the coupling analysis provided numerically unstable solutions. From the result, it was found that improvement of the coupling algorithm of BAMBOO and ASFRE-IV is necessary to reduce numerical fluctuations and to obtain better convergence by introducing such computational technique as the optimized under-relaxation method. (author)

Feasibility of an Exoskeleton-Based Interactive Video Game System for Upper Extremity Burn Contractures.

Science.gov (United States)

Schneider, Jeffrey C; Ozsecen, Muzaffer Y; Muraoka, Nicholas K; Mancinelli, Chiara; Della Croce, Ugo; Ryan, Colleen M; Bonato, Paolo

2016-05-01

Burn contractures are common and difficult to treat. Measuring continuous joint motion would inform the assessment of contracture interventions; however, it is not standard clinical practice. This study examines use of an interactive gaming system to measure continuous joint motion data. To assess the usability of an exoskeleton-based interactive gaming system in the rehabilitation of upper extremity burn contractures. Feasibility study. Eight subjects with a history of burn injury and upper extremity contractures were recruited from the outpatient clinic of a regional inpatient rehabilitation facility. Subjects used an exoskeleton-based interactive gaming system to play 4 different video games. Continuous joint motion data were collected at the shoulder and elbow during game play. Visual analog scale for engagement, difficulty and comfort. Angular range of motion by subject, joint, and game. The study population had an age of 43 ± 16 (mean ± standard deviation) years and total body surface area burned range of 10%-90%. Subjects reported satisfactory levels of enjoyment, comfort, and difficulty. Continuous joint motion data demonstrated variable characteristics by subject, plane of motion, and game. This study demonstrates the feasibility of use of an exoskeleton-based interactive gaming system in the burn population. Future studies are needed that examine the efficacy of tailoring interactive video games to the specific joint impairments of burn survivors. Copyright © 2016 American Academy of Physical Medicine and Rehabilitation. Published by Elsevier Inc. All rights reserved.

Systems Analysis of a Compact Next Step Burning Plasma Experiment

International Nuclear Information System (INIS)

Jardin, S.C.; Kessel, C.E.; Meade, D.; Neumeyer, C.

2002-01-01

A new burning plasma systems code (BPSC) has been developed for analysis of a next step compact burning plasma experiment with copper-alloy magnet technology. We consider two classes of configurations: Type A, with the toroidal field (TF) coils and ohmic heating (OH) coils unlinked, and Type B, with the TF and OH coils linked. We obtain curves of the minimizing major radius as a function of aspect ratio R(A) for each configuration type for typical parameters. These curves represent, to first order, cost minimizing curves, assuming that device cost is a function of major radius. The Type B curves always lie below the Type A curves for the same physics parameters, indicating that they lead to a more compact design. This follows from that fact that a high fraction of the inner region, r < R-a, contains electrical conductor material. However, the fact that the Type A OH and TF magnets are not linked presents fewer engineering challenges and should lead to a more reliable design. Both the Type A and Type B curves have a minimum in major radius R at a minimizing aspect ratio A typically above 2.8 and at high values of magnetic field B above 10 T. The minimizing A occurs at larger values for longer pulse and higher performance devices. The larger A and higher B design points also have the feature that the ratio of the discharge time to the current redistribution time is largest so that steady-state operation can be more realistically prototyped. A sensitivity study is presented for the baseline Type A configuration showing the dependence of the results on the parameters held fixed for the minimization study

Basic concept of common reactor physics code systems. Final report of working party on common reactor physics code systems (CCS)

International Nuclear Information System (INIS)

2004-03-01

A working party was organized for two years (2001-2002) on common reactor physics code systems under the Research Committee on Reactor Physics of JAERI. This final report is compilation of activity of the working party on common reactor physics code systems during two years. Objectives of the working party is to clarify basic concept of common reactor physics code systems to improve convenience of reactor physics code systems for reactor physics researchers in Japan on their various field of research and development activities. We have held four meetings during 2 years, investigated status of reactor physics code systems and innovative software technologies, and discussed basic concept of common reactor physics code systems. (author)

Self-Burns in Fars Province, Southern Iran

Science.gov (United States)

Mohammadi, Ali Akbar; Tohidinik, Hamid Reza; Zardosht, Mitra; Seyed Jafari, Seyed Morteza

2016-01-01

BACKGROUND The alarming incidence of self- burning provoked to set up a multidisciplinary preventive program to decrease the incidence and complications of this harmful issue. This study investigated the incidence and the preventive measures in self-burn in Fars Province, southern Iran. METHODS This study was a longitudinal prospective design on trend of self-inflicted burn injuries in Fars province after setting up a regional multidisciplinary preventive plan (2009-2012). RESULTS From 18862 admitted patients, 388 (2%) committed self-burning. While the incidence showed a constant decrease in proportion of suicidal cases among all admitted patients (2.5% to 1.6%). The mean age of self-burning victims ranged from 28.3±10.8 to 30.3±11.7 years. The female victims comprised 67.4% of all suicidal burn patients (Female to male ratio: 2.18). The leading causes of suicide commitment were familial conflicts (75.6%) and psychological problems (16.7%) CONCLUSION It is crucial to continue the regional preventive programs and pave the way to set up national, and even international collaborations to alleviate relevant financial, social, cultural and infrastructural difficulties in order to have lower incidence for this dramatic issue. PMID:27308238

The octopus burnup and criticality code system

Energy Technology Data Exchange (ETDEWEB)

Kloosterman, J.L.; Kuijper, J.C. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Leege, P.F.A. de

1996-09-01

The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional geometries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (author)

The OCTOPUS burnup and criticality code system

Energy Technology Data Exchange (ETDEWEB)

Kloosterman, J.L. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Kuijper, J.C. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.

1996-06-01

The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional goemetries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (orig.).

The octopus burnup and criticality code system

International Nuclear Information System (INIS)

Kloosterman, J.L.; Kuijper, J.C.; Leege, P.F.A. de.

1996-01-01

The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional geometries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (author)

The OCTOPUS burnup and criticality code system

International Nuclear Information System (INIS)

Kloosterman, J.L.; Kuijper, J.C.; Leege, P.F.A. de

1996-06-01

The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional goemetries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (orig.)

Numerical solution of the point reactor kinetics equations with fuel burn-up and temperature feedback

International Nuclear Information System (INIS)

Tashakor, S.; Jahanfarnia, G.; Hashemi-Tilehnoee, M.

2010-01-01

Point reactor kinetics equations are solved numerically using one group of delayed neutrons and with fuel burn-up and temperature feedback included. To calculate the fraction of one-group delayed neutrons, a group of differential equations are solved by an implicit time method. Using point reactor kinetics equations, changes in mean neutrons density, temperature, and reactivity are calculated in different times during the reactor operation. The variation of reactivity, temperature, and maximum power with time are compared with the predictions by other methods.

High dynamic range coding imaging system

Science.gov (United States)

Wu, Renfan; Huang, Yifan; Hou, Guangqi

2014-10-01

We present a high dynamic range (HDR) imaging system design scheme based on coded aperture technique. This scheme can help us obtain HDR images which have extended depth of field. We adopt Sparse coding algorithm to design coded patterns. Then we utilize the sensor unit to acquire coded images under different exposure settings. With the guide of the multiple exposure parameters, a series of low dynamic range (LDR) coded images are reconstructed. We use some existing algorithms to fuse and display a HDR image by those LDR images. We build an optical simulation model and get some simulation images to verify the novel system.

Civil plutonium in the world: an estimate by the code REACTOR

International Nuclear Information System (INIS)

Braet, J.; Carchon, R.; Van der Meer, K.

1996-11-01

The computer code REACTOR that was developed by the Belgian Nuclear Research Centre SCK/CEN to study the built-up of plutonium stockpiles in the world is described. The code consists of a central database, containing general information about most commercial civil nuclear facilities. Using this code, an overview is given of the evolution of the nuclear energy production in the world, in the past and the medium term future. The nuclear energy production results in the accumulation of spent fuel stocks, containing vast amounts of energy enclosed in the plutonium. The presence and built-up of large stockpiles of spent fuel and separated plutonium originating from the civil fuel cycle is estimated. In this report several possible scenarios are considered for the use of that plutonium, with the aim of minimizing those stocks. According to the different national policies, scenarios such as open fuel cycle, thermal reactors or fast reactor cycle with the burning of plutonium in fast reactors are envisaged

The characteristics of autonomic nervous system disorders in burning mouth syndrome and Parkinson disease.

Science.gov (United States)

Koszewicz, Magdalena; Mendak, Magdalena; Konopka, Tomasz; Koziorowska-Gawron, Ewa; Budrewicz, Sławomir

2012-01-01

To conduct a clinical electrophysiologic evaluation of autonomic nervous system functions in patients with burning mouth syndrome and Parkinson disease and estimate the type and intensity of the autonomic dysfunction. The study involved 83 subjects-33 with burning mouth syndrome, 20 with Parkinson disease, and 30 controls. The BMS group included 27 women and 6 men (median age, 60.0 years), and the Parkinson disease group included 15 women and 5 men (median age, 66.5 years). In the control group, there were 20 women and 10 men (median age, 59.0 years). All patients were subjected to autonomic nervous system testing. In addition to the Low autonomic disorder questionnaire, heart rate variability (HRV), deep breathing (exhalation/inspiration [E/I] ratio), and sympathetic skin response (SSR) tests were performed in all cases. Parametric and nonparametric tests (ANOVA, Kruskal-Wallis, and Scheffe tests) were used in the statistical analysis. The mean values for HRV and E/I ratios were significantly lower in the burning mouth syndrome and Parkinson disease groups. Significant prolongation of SSR latency in the foot was revealed in both burning mouth syndrome and Parkinson disease patients, and lowering of the SSR amplitude occurred in only the Parkinson disease group. The autonomic questionnaire score was significantly higher in burning mouth syndrome and Parkinson disease patients than in the control subjects, with the Parkinson disease group having the highest scores. In patients with burning mouth syndrome, a significant impairment of both the sympathetic and parasympathetic nervous systems was found but sympathetic/parasympathetic balance was preserved. The incidence and intensity of autonomic nervous system dysfunction was similar in patients with burning mouth syndrome and Parkinson disease, which may suggest some similarity in their pathogeneses.

Concatenated coding systems employing a unit-memory convolutional code and a byte-oriented decoding algorithm

Science.gov (United States)

Lee, L.-N.

1977-01-01

Concatenated coding systems utilizing a convolutional code as the inner code and a Reed-Solomon code as the outer code are considered. In order to obtain very reliable communications over a very noisy channel with relatively modest coding complexity, it is proposed to concatenate a byte-oriented unit-memory convolutional code with an RS outer code whose symbol size is one byte. It is further proposed to utilize a real-time minimal-byte-error probability decoding algorithm, together with feedback from the outer decoder, in the decoder for the inner convolutional code. The performance of the proposed concatenated coding system is studied, and the improvement over conventional concatenated systems due to each additional feature is isolated.

Status of reactor core design code system in COSINE code package

Energy Technology Data Exchange (ETDEWEB)

Chen, Y.; Yu, H.; Liu, Z., E-mail: yuhui@snptc.com.cn [State Nuclear Power Software Development Center, SNPTC, National Energy Key Laboratory of Nuclear Power Software (NEKLS), Beijiing (China)

2014-07-01

For self-reliance, COre and System INtegrated Engine for design and analysis (COSINE) code package is under development in China. In this paper, recent development status of the reactor core design code system (including the lattice physics code and the core simulator) is presented. The well-established theoretical models have been implemented. The preliminary verification results are illustrated. And some special efforts, such as updated theory models and direct data access application, are also made to achieve better software product. (author)

Status of reactor core design code system in COSINE code package

International Nuclear Information System (INIS)

Chen, Y.; Yu, H.; Liu, Z.

2014-01-01

For self-reliance, COre and System INtegrated Engine for design and analysis (COSINE) code package is under development in China. In this paper, recent development status of the reactor core design code system (including the lattice physics code and the core simulator) is presented. The well-established theoretical models have been implemented. The preliminary verification results are illustrated. And some special efforts, such as updated theory models and direct data access application, are also made to achieve better software product. (author)

Modelling of the spent fuel heat-up in the spent fuel pools using one-dimensional system codes and CFD codes

Energy Technology Data Exchange (ETDEWEB)

Grazevicius, Audrius; Kaliatka, Algirdas [Lithuanian Energy Institute, Kaunas (Lithuania). Lab. of Nuclear Installation Safety

2017-07-15

The main functions of spent fuel pools are to remove the residual heat from spent fuel assemblies and to perform the function of biological shielding. In the case of loss of heat removal from spent fuel pool, the fuel rods and pool water temperatures would increase continuously. After the saturated temperature is reached, due to evaporation of water the pool water level would drop, eventually causing the uncover of spent fuel assemblies, fuel overheating and fuel rods failure. This paper presents an analysis of loss of heat removal accident in spent fuel pool of BWR 4 and a comparison of two different modelling approaches. The one-dimensional system thermal-hydraulic computer code RELAP5 and CFD tool ANSYS Fluent were used for the analysis. The results are similar, but the local effects cannot be simulated using a one-dimensional code. The ANSYS Fluent calculation demonstrated that this three-dimensional treatment allows to avoid the need for many one-dimensional modelling assumptions in the pool modelling and enables to reduce the uncertainties associated with natural circulation flow calculation.

Effects of thermal-hydraulic feedback on burnup modeling of the deep burn modular high temperature reactor (DB-MHR)

International Nuclear Information System (INIS)

Bei, Yea; Wen, Wua; Di, Yuna; Stubbins, J.F.; Venneri, F.

2007-01-01

The Deep-Burn concept investigates the use of commercial high temperature gas-cooled reactors such as modular helium reactors (DB-MHR) to transmute spent fuel from light water reactors (LWRs). An essential feature of this technology is the fabrication of spent fuel into TRISO particles with full transuranic composition to achieve very extensive destruction levels (deep-burn) in a one-pass fuel cycle. Due to the strong temperature influence on the cross sections of transuranics, the coupling between temperature and neutronics is very important to be able to simulate realistic operations of the deep burn reactor. In this study, detailed simulations of the DB-MHR operation are performed with a Monte Carlo code system (MCNP-5 + ORIGEN-2.2 + MONTEBURNS-2 for neutronics calculations), POKE code (General Atomics, for thermohydraulics calculations) and NJOY-99 code (for processing nuclear data libraries), called MHRBURNS. Resulting power densities of fuel blocks (from neutronics calculations) are provided as input to the POKE code, which in turn, calculates new temperature distributions. The temperature distributions obtained from POKE are used to update the MCNP input, and NJOY is called to process new nuclear cross sections based on appropriate temperatures. These steps are repeated to calculate the entire burnup performance of the system. In this preliminary study only the feedback on graphite temperature is taken into account. It is observed that the temperature feedback results show a 200 K higher temperature and thus a slight difference in 237 Np and 239 Pu destruction rates, although the overall burnup rates remain the same

Analysis of Delayed Sea Breeze Onset for Fort Ord Prescribed Burning Operations

Science.gov (United States)

2015-12-01

DELAYED SEA BREEZE ONSET FOR FORT ORD PRESCRIBED BURNING OPERATIONS by Dustin D. Hocking December 2015 Thesis Advisor: Wendell Nuss Second...AND DATES COVERED Master’s thesis 4. TITLE AND SUBTITLE ANALYSIS OF DELAYED SEA BREEZE ONSET FOR FORT ORD PRESCRIBED BURNING OPERATIONS 5...release; distribution is unlimited 12b. DISTRIBUTION CODE 13. ABSTRACT (maximum 200 words) The U.S. Army conducts prescribed burns at Fort Ord

On the development of LWR fuel analysis code (1). Analysis of the FEMAXI code and proposal of a new model

International Nuclear Information System (INIS)

Lemehov, Sergei; Suzuki, Motoe

2000-01-01

This report summarizes the review on the modeling features of FEMAXI code and proposal of a new theoretical equation model of clad creep on the basis of irradiation-induced microstructure change. It was pointed out that plutonium build-up in fuel matrix and non-uniform radial power profile at high burn-up affect significantly fuel behavior through the interconnected effects with such phenomena as clad irradiation-induced creep, fission gas release, fuel thermal conductivity degradation, rim porous band formation and associated fuel swelling. Therefore, these combined effects should be properly incorporated into the models of the FEMAXI code so that the code can carry out numerical analysis at the level of accuracy and elaboration that modern experimental data obtained in test reactors have. Also, the proposed new mechanistic clad creep model has a general formalism which allows the model to be flexibly applied for clad behavior analysis under normal operation conditions and power transients as well for Zr-based clad materials by the use of established out-of-pile mechanical properties. The model has been tested against experimental data, while further verification is needed with specific emphasis on power ramps and transients. (author)

Reasons for Distress Among Burn Survivors at 6, 12, and 24 Months Postdischarge: A Burn Injury Model System Investigation.

Science.gov (United States)

Wiechman, Shelley A; McMullen, Kara; Carrougher, Gretchen J; Fauerbach, Jame A; Ryan, Colleen M; Herndon, David N; Holavanahalli, Radha; Gibran, Nicole S; Roaten, Kimberly

2017-12-16

To identify important sources of distress among burn survivors at discharge and 6, 12, and 24 months postinjury, and to examine if the distress related to these sources changed over time. Exploratory. Outpatient burn clinics in 4 sites across the country. Participants who met preestablished criteria for having a major burn injury (N=1009) were enrolled in this multisite study. Participants were given a previously developed list of 12 sources of distress among burn survivors and asked to rate on a 10-point Likert-type scale (0=no distress to 10=high distress) how much distress each of the 12 issues was causing them at the time of each follow-up. The Medical Outcomes Study 12-Item Short-Form Health Survey was administered at each time point as a measure of health-related quality of life. The Satisfaction With Appearance Scale was used to understand the relation between sources of distress and body image. Finally, whether a person returned to work was used to determine the effect of sources of distress on returning to employment. It was encouraging that no symptoms were worsening at 2 years. However, financial concerns and long recovery time are 2 of the highest means at all time points. Pain and sleep disturbance had the biggest effect on ability to return to work. These findings can be used to inform burn-specific interventions and to give survivors an understanding of the temporal trajectory for various causes of distress. In particular, it appears that interventions targeted at sleep disturbance and high pain levels can potentially effect distress over financial concerns by allowing a person to return to work more quickly. Copyright © 2017 American Congress of Rehabilitation Medicine. Published by Elsevier Inc. All rights reserved.

Ice & Fire: the Burning Question

DEFF Research Database (Denmark)

van Gelderen, Laurens; Jomaas, Grunde

2017-01-01

With the Arctic opening up to new shipping routes and increased oil exploration and production due to climate change, the risk of an Arctic oil spill is increasing. Of the classic oil spill response methods (mechanical recovery, dispersants and in-situ burning), in-situ burning is considered...... to be particularly a suitable response method in the Arctic. In-situ burning aims to remove the oil from the marine environment by burning it from the water surface. A recent Ph.D. thesis from the Technical University of Denmark has provided some new insights with respect to the fire science behind this response...

The burnup capabilities of the Deep Burn Modular Helium Reactor analyzed by the Monte Carlo Continuous Energy Code MCB

Energy Technology Data Exchange (ETDEWEB)

Talamo, Alberto E-mail: alby@neutron.kth.se; Gudowski, Waclaw E-mail: wacek@neutron.kth.se; Venneri, Francesco E-mail: venneri@lanl.gov

2004-01-01

We have investigated the waste actinide burnup capabilities of a Gas Turbine Modular Helium Reactor (GT-MHR, similar to the reactor being designed by General Atomics and Minatom for surplus weapons plutonium destruction) with the Monte Carlo Continuous Energy Burnup Code MCB, an extension of MCNP developed at the Royal Institute of Technology in Stockholm and University of Mining and Metallurgy in Krakow. The GT-MHR is a gas-cooled, graphite-moderated reactor, which can be powered with a wide variety of fuels, like thorium, uranium or plutonium. In the present work, the GT-MHR is fueled with the transuranic actinides contained in Light Water Reactors (LWRs) spent fuel for the purpose of destroying them as completely as possible with minimum reliance on multiple reprocessing steps. After uranium extraction from the LWR spent fuel (UREX), the remaining waste actinides, including plutonium are partitioned into two distinct types of fuel for use in the GT-MHR: Driver Fuel (DF) and Transmutation Fuel (TF). The DF supplies the neutrons to maintain the fission chain reaction, whereas the TF emphasizes neutron capture to induce a deep burn transmutation and provide reactivity control by a negative feedback. When used in this mode, the GT-MHR is called Deep Burn Modular Helium Reactor (DB-MHR). Both fuels are contained in a structure of triple isotropic coated layers, TRISO coating, which has been proven to retain fission products up to 1600 deg. C and is expected to remain intact for hundreds of thousands of years after irradiation. Other benefits of this reactor consist of: a well-developed technology, both for the graphite-moderated core and the TRISO structure, a high energy conversion efficiency (about 50%), well established passive safety mechanism and a competitive cost. The destruction of more than 94% of {sup 239}Pu and the other geologically problematic actinide species makes this reactor a valid proposal for the reduction of nuclear waste and the prevention of

The burnup capabilities of the Deep Burn Modular Helium Reactor analyzed by the Monte Carlo Continuous Energy Code MCB

International Nuclear Information System (INIS)

Talamo, Alberto; Gudowski, Waclaw; Venneri, Francesco

2004-01-01

We have investigated the waste actinide burnup capabilities of a Gas Turbine Modular Helium Reactor (GT-MHR, similar to the reactor being designed by General Atomics and Minatom for surplus weapons plutonium destruction) with the Monte Carlo Continuous Energy Burnup Code MCB, an extension of MCNP developed at the Royal Institute of Technology in Stockholm and University of Mining and Metallurgy in Krakow. The GT-MHR is a gas-cooled, graphite-moderated reactor, which can be powered with a wide variety of fuels, like thorium, uranium or plutonium. In the present work, the GT-MHR is fueled with the transuranic actinides contained in Light Water Reactors (LWRs) spent fuel for the purpose of destroying them as completely as possible with minimum reliance on multiple reprocessing steps. After uranium extraction from the LWR spent fuel (UREX), the remaining waste actinides, including plutonium are partitioned into two distinct types of fuel for use in the GT-MHR: Driver Fuel (DF) and Transmutation Fuel (TF). The DF supplies the neutrons to maintain the fission chain reaction, whereas the TF emphasizes neutron capture to induce a deep burn transmutation and provide reactivity control by a negative feedback. When used in this mode, the GT-MHR is called Deep Burn Modular Helium Reactor (DB-MHR). Both fuels are contained in a structure of triple isotropic coated layers, TRISO coating, which has been proven to retain fission products up to 1600 deg. C and is expected to remain intact for hundreds of thousands of years after irradiation. Other benefits of this reactor consist of: a well-developed technology, both for the graphite-moderated core and the TRISO structure, a high energy conversion efficiency (about 50%), well established passive safety mechanism and a competitive cost. The destruction of more than 94% of 239 Pu and the other geologically problematic actinide species makes this reactor a valid proposal for the reduction of nuclear waste and the prevention of

Revised SRAC code system

International Nuclear Information System (INIS)

Tsuchihashi, Keichiro; Ishiguro, Yukio; Kaneko, Kunio; Ido, Masaru.

1986-09-01

Since the publication of JAERI-1285 in 1983 for the preliminary version of the SRAC code system, a number of additions and modifications to the functions have been made to establish an overall neutronics code system. Major points are (1) addition of JENDL-2 version of data library, (2) a direct treatment of doubly heterogeneous effect on resonance absorption, (3) a generalized Dancoff factor, (4) a cell calculation based on the fixed boundary source problem, (5) the corresponding edit required for experimental analysis and reactor design, (6) a perturbation theory calculation for reactivity change, (7) an auxiliary code for core burnup and fuel management, etc. This report is a revision of the users manual which consists of the general description, input data requirements and their explanation, detailed information on usage, mathematics, contents of libraries and sample I/O. (author)

Application of spectral hole burning to the study of in vitro cellular systems

Energy Technology Data Exchange (ETDEWEB)

Milanovich, Nebojsa [Iowa State Univ., Ames, IA (United States)

1999-11-08

Chapter 1 of this thesis describes the various stages of tumor development and a multitude of diagnostic techniques used to detect cancer. Chapter 2 gives an overview of the aspects of hole burning spectroscopy important for its application to the study of cellular systems. Chapter 3 gives general descriptions of cellular organelles, structures, and physical properties that can serve as possible markers for the differentiation of normal and cancerous cells. Also described in Chapter 3 are the principles of cryobiology important for low temperature spectroscopy of cells, characterization of MCF-10F (normal) and MCF-7 (cancer) cells lines which will serve as model systems, and cellular characteristics of aluminum phthalocyanine tetrasulfonate (APT), which was used as the test probe. Chapters 4 and 5 are previously published papers by the author pertaining to the results obtained from the application of hole burning to the study of cellular systems. Chapter 4 presents the first results obtained by spectral hole burning of cellular systems and Chapter 5 gives results for the differentiation of MCF-10F and MCF-7 cells stained with APT by an external applied electric (Stark) field. A general conclusion is presented in Chapter 6. Appendices A and B provide additional characterization of the cell/probe model systems. Appendix A describes the uptake and subcellular distribution of APT in MCF-10F and MCF-7 cells and Appendix B compares the hole burning characteristics of APT in cells when the cells are in suspension and when they are examined while adhering to a glass coverslip. Appendix C presents preliminary results for a novel probe molecule, referred to as a molecular thumbtack, designed by the authors for use in future hole burning applications to cellular systems.

Ignition and fusion burn in fast ignition scheme

International Nuclear Information System (INIS)

Takabe, Hideaki

1998-01-01

The target physics of fast ignition is briefly reviewed by focusing on the ignition and fusion burn in the off-center ignition scheme. By the use of a two dimensional hydrodynamic code with an alpha heating process, the ignition condition is studied. It is shown that the ignition condition of the off-center ignition scheme coincides with that of the the central isochoric model. After the ignition, a nuclear burning wave is seen to burn the cold main fuel with a velocity of 2 - 3 x 10 8 cm/s. The spark energy required for the off-center ignition is 2 - 3 kJ or 10 - 15 kJ for the core density of 400 g/cm 3 or 200 g/cm 3 , respectively. It is demonstrated that a core gain of more than 2,000 is possible for a core energy of 100 kJ with a hot spark energy of 13 kJ. The requirement for the ignition region's heating time is also discussed by modeling a heating source in the 2-D code. (author)

PRELIMINARY STUDY ON APPLICATION OF MAX PLUS ALGEBRA IN DISTRIBUTED STORAGE SYSTEM THROUGH NETWORK CODING

Directory of Open Access Journals (Sweden)

Agus Maman Abadi

2016-04-01

Full Text Available The increasing need in techniques of storing big data presents a new challenge. One way to address this challenge is the use of distributed storage systems. One strategy that implemented in distributed data storage systems is the use of Erasure Code which applied to network coding. The code used in this technique is based on the algebraic structure which is called as vector space. Some studies have also been carried out to create code that is based on other algebraic structures such as module.  In this study, we are going to try to set up a code based on the algebraic structure which is a generalization of the module that is semimodule by utilizing the max operations and sum operations at max plus algebra. The results of this study indicate that the max operation and the addition operation on max plus algebra cannot be used to establish a semimodule code, but by modifying the operation "+" as "min", we get a code based on semimodule. Keywords:   code, distributed storage systems, network coding, semimodule, max plus algebra

Non destructive burn up determination of IEA-R1 reactor fuel elements by gamma-ray spectrometry using a Ge(Li) detector

International Nuclear Information System (INIS)

Madi Filho, T.

1982-01-01

A non destructive determination of burn up of low (IEA-14) and high (IEA-80) activity fuel elements used in the IEA-R1 pool reactor was made from the measured distribution of the Cs-137 gamma-ray activity in these elements. For both series of measurements a 73,7 c.c. Ge(Li) detector was used in 'well collimated' geometry. Where as IEA-14, removed from the reactor some 20 years, showed a gamma-ray spectrum essentially due to Cs-137, IEA-80, with a cooling time of 5 years, showed a more complex spectrum due to the greater number of fission products remaining. The S.I out-of-pool assembly was calibrated using Cs-137 and Co-60 point and Ag-110m plane sources. These measurements provided the necessary constants used to calculate fuel burn-up from measured relative activity distributions of fuel elements. Detailed fuel plate transmission measurements made with the Cs-137 source showed the plates to be highly homogeneous. High activity fuel elements were measured in the S.II in-pool assembly in which the detector was locate on the moveable pool bridge and the test element was positioned immediately below the detector 2.17m below the pool surface. Measurements made in the S.II assembly were normalised with respect to the measured activity of the IEA-14 element. The measured burn up of the IEA-14 and IEA-80 elements obtained in this work is 3.22.10 - 3 gms and 24.44gms. These values may be compared with respective values of 2.63.10 - 3 gms and 61.11gms given by 'total reactor energy/flux distribution' calculations. Calculated errors for the U-235 burn up are 7.4% (IEA-14) and 10.1% (IEA-80). A detailed evaluation of the errors associated with both sets of measurements is given. (Author) [pt

Nuclear fuel and/or fertile material element suitable for non-destructive determination of burn-up

International Nuclear Information System (INIS)

Muench, E.

1976-01-01

The invention refers to a nuclear fuel and/or fertile material element suitable for non-destructive burn-up analysis, where an isotope or a mixture of isotopes capable of being activated is provided for measuring the intensity of radiation emitted from radioactive nuclides, especially the intensity of gamma rays. The half-life of radioactive decay of the isotope or the mixture mentioned above after being activated is sufficiently large compared with the irradiation of the fuel and/or fertile material element in the nuclear reactor. (orig.) [de

Non-burn electric generation: How today's options stack up

International Nuclear Information System (INIS)

1993-01-01

The technical preparedness to generate electricity without burning fuel is dealt with. Nuclear, hydroelectric, solar and wind energy are recommended as the clean options. The aims of energy policy, views upon regulation, technical maturity and commercial preparedness of such variants are discussed. (Z.S.). 4 figs

ETF system code: composition and applications

International Nuclear Information System (INIS)

Reid, R.L.; Wu, K.F.

1980-01-01

A computer code has been developed for application to ETF tokamak system and conceptual design studies. The code determines cost, performance, configuration, and technology requirements as a function of tokamak parameters. The ETF code is structured in a modular fashion in order to allow independent modeling of each major tokamak component. The primary benefit of modularization is that it allows updating of a component module, such as the TF coil module, without disturbing the remainder of the system code as long as the input/output to the modules remains unchanged. The modules may be run independently to perform specific design studies, such as determining the effect of allowable strain on TF coil structural requirements, or the modules may be executed together as a system to determine global effects, such as defining the impact of aspect ratio on the entire tokamak system

Management of 2nd-degree facial burns using the Versajet(®) hydrosurgery system and xenograft: a prospective evaluation of 20 cases.

Science.gov (United States)

Duteille, Franck; Perrot, Pierre

2012-08-01

There is no single therapeutic scheme for the management of intermediary 2nd-degree facial burns, which can cause problems because of their uncertain course. It is preferable to obtain optimal healing of the face in order to avoid functional or cosmetic sequelae. Some practitioners recommend early excision (first week) of these burns, whereas others prefer to wait and perform surgery later (after 2 weeks). The practice in our burns unit is early surgery (from the first week) associated with hydrosurgical excision and application of a biosynthetic dressing (xenograft). A prospective follow-up of 20 cases was carried out to evaluate the efficacy of our protocol. The prospective evaluation was performed with follow-up at 2 weeks and 3, 6 and 12 months. The patients included had intermediary 2nd-degree burns on at least 15% of the face and no life-threatening prognosis. The mean age in our series was 40.5 years (16-72), the mean percentage of burn surface area was 27.75% and the mean percentage of facial burn was 60.75%. Early excision was performed (day 5-10) using the Versajet(®) system, which allows tangential water-dissection. Porcine xenograft (E-Z Derm(®)) was applied immediately afterwards. Patients whose healing process was not complete at 2 weeks were then scheduled to receive a thin autograft. Patients were followed up 2 weeks, 3, 6 and 12 months after discharge. Excision was performed at a mean 7.6 days, and mean initial healing time was 13.4 days. In three cases, a full-thickness skin graft was used, whereas healing occurred in the other patients without further grafts. Two patients had functional sequelae (ectropion) corrected later by repair surgery. The course of healing for the other patients proceeded normally. There is no consensus about the management of intermediate depth 2nd-degree facial burns. We chose to perform early surgery using the Versajet(®) system, which allows fine, precise excision, leaving nearly all of the healthy tissue in place

Advanced thermionic reactor systems design code

International Nuclear Information System (INIS)

Lewis, B.R.; Pawlowski, R.A.; Greek, K.J.; Klein, A.C.

1991-01-01

An overall systems design code is under development to model an advanced in-core thermionic nuclear reactor system for space applications at power levels of 10 to 50 kWe. The design code is written in an object-oriented programming environment that allows the use of a series of design modules, each of which is responsible for the determination of specific system parameters. The code modules include a neutronics and core criticality module, a core thermal hydraulics module, a thermionic fuel element performance module, a radiation shielding module, a module for waste heat transfer and rejection, and modules for power conditioning and control. The neutronics and core criticality module determines critical core size, core lifetime, and shutdown margins using the criticality calculation capability of the Monte Carlo Neutron and Photon Transport Code System (MCNP). The remaining modules utilize results of the MCNP analysis along with FORTRAN programming to predict the overall system performance

Variable-length code construction for incoherent optical CDMA systems

Science.gov (United States)

Lin, Jen-Yung; Jhou, Jhih-Syue; Wen, Jyh-Horng

2007-04-01

The purpose of this study is to investigate the multirate transmission in fiber-optic code-division multiple-access (CDMA) networks. In this article, we propose a variable-length code construction for any existing optical orthogonal code to implement a multirate optical CDMA system (called as the multirate code system). For comparison, a multirate system where the lower-rate user sends each symbol twice is implemented and is called as the repeat code system. The repetition as an error-detection code in an ARQ scheme in the repeat code system is also investigated. Moreover, a parallel approach for the optical CDMA systems, which is proposed by Marić et al., is also compared with other systems proposed in this study. Theoretical analysis shows that the bit error probability of the proposed multirate code system is smaller than other systems, especially when the number of lower-rate users is large. Moreover, if there is at least one lower-rate user in the system, the multirate code system accommodates more users than other systems when the error probability of system is set below 10 -9.

Gas cloud explosions and their effect on nuclear power plant, basic development of explosion codes

International Nuclear Information System (INIS)

Hall, S.F.; Martin, D.; MacKenzie, J.

1985-01-01

The study of factors influencing the pressure and velocity fields produced by the burning of flammable substances has been in progress at SRD for some years. This paper describes an extension of these studies by using existing codes for a parametric survey, and modifying codes to produce more realistic representations of explosions and developing a two dimensional combustion code, FLARE. The one dimensional combustion code, GASEX1, has been used to determine the pressure from a burning gas cloud for a number of different fuels, concentrations and burning velocities. The code was modified so that gas concentrations could be modelled. Results for concentration gradients showed the pressure depended on local conditions and the burning velocity. The two dimensional code, GASEX2, was modified to model the interaction of pressure waves with structures. It was used, with results from GASEX1, to model the interaction of a pressure wave from the combustion of a gas cloud with a rigid structure representing a nuclear power plant. The two dimensional code FLARE has been developed to model the interaction of flames and pressure waves with structures. The code incorporates a simple turbulence model with a turbulence dependent reaction rate. Validation calculations have been carried out for the code. (author)

Parametric study of the behaviour of a pre irradiated BWR fuel rod under conditions of LOCA simulated in the halden in pile test system with the FALCON code

Energy Technology Data Exchange (ETDEWEB)

Khvostov, G.; Zimmermann, M. A. [Laboratory for Reactor Physics and Systems Behaviour, Paul Scherrer Institut, Villigen (Switzerland); Ledergerber, G. [Kernkraftwerk Leibstadt AG, Leibstadt (Switzerland); Kolstad, E. [Institute for Energy Technology - OECD Halden Reactor Project, Halden (Norway); Montgomery, R. O. [Anatech Corporation, San Diego (United States)

2008-10-15

A new LOCA test at Halden was planned as the first experiment within the Halden LOCA program addressing the behaviour of commercially irradiated BWR fuel of medium burn up with burst of the cladding expected to occur at a temperature of about 1050.deg.C, which is essentially higher than in the preceding experiments. The specific measures to be adopted have been suggested based upon a parametric study using the FALCON fuel behaviour code and aimed at an optimized design of the test fuel rod for the given high target cladding temperature of 1150 .deg. C (peak local). The analysis has shown a reasonable agreement with the fundamental experimental findings, such as correlations of NUREG 0630, as well as consistency with the data from Halden LOCA testing available so far. Thus, a general conclusion is drawn about the applicability of the methodology developed at PSI to the analysis of LWR fuel rod behaviour during LOCA, in consideration of the effects of fuel burn up.

Simulation of MHD instability effects on burning plasma transport with ITB in tokamak and helical reactors

International Nuclear Information System (INIS)

Yamazaki, K.; Yamada, I.; Taniguchi, S.; Oishi, T.

2009-01-01

Full text: The high performance plasma behavior is required to realize economic and environmental-friendly fusion reactors compatible with conventional power plant systems. To improve plasma confinement, the formation of internal transport barrier (ITB) is anticipated, and its behavior is analyzed by the simulation code TOTAL (Toroidal Transport Linkage Analysis). This TOTAL code comprises a 2- or 3-dimensional equilibrium and 1-dimensional predictive transport code for both tokamak and helical systems. In the tokamak code TOTAL-T, the external current drive, bootstrap current, sawtooth oscillation, ballooning mode and neoclassical tearing mode (NTM) analyses are included. The steady-state burning plasma operation is achieved by the feedback control of pellet injection fuelling and external heating power control. The impurity dynamics of iron and tungsten is also included in this code. The NTM effects are evaluated using the modified Rutherford Model with the stabilization of the ECCD current drive. The excitation of m=2/n=1 NTM leads to the 20 % reduction in the central temperature in ITER-like reactors. Recently, the external non-resonant helical field application is analyzed and its stabilization properties are evaluated. The pellet injection effects on ITB formation is also clarified in tokamak and helical plasmas. Relationship between sawtooth oscillation and impurity ejection is recently simulated in comparison with experimental data. In this conference, we will show above-stated new results on MHD instability effects on burning plasma transport. (author)

PASC-1, Petten AMPX-II/SCALE-3 Code System for Reactor Neutronics Calculation

International Nuclear Information System (INIS)

Yaoqing, W.; Oppe, J.; Haas, J.B.M. de; Gruppelaar, H.; Slobben, J.

1995-01-01

1 - Description of program or function: The Petten AMPX-II/SCALE-3 Code System PASC-1 is a reactor neutronics calculation programme system consisting of well known IBM-oriented codes, that have been translated into FORTRAN-77, for calculations on a CDC-CYBER computer. Thus, the portability of these codes has been increased. In this system, some AMPX-II and SCALE-3 modules, the one-dimensional transport code ANISN and the 1 to 3-dimensional diffusion code CITATION are linked together on the CDC-CYBER/855 computer. The new cell code XSDRNPM-S and the old XSDRN code are included in the system. Starting from an AMPX fine group library up to CITATION, calculations can be performed for each individual module. Existing AMPX master interface format libraries, such as CSRL-IV, JEF-1, IRI and SCALE-45, and the old XSDRN-formatted libraries such as the COBB library can be used for the calculations. The code system contains the following modules and codes at present: AIM, AJAX, MALOCS, NITAWL-S, REVERT-I, ICE-2, CONVERT, JUAN, OCTAGN, XSDRNPM-S, XSDRN, ANISN and CITATION. The system will be extended with other SCALE modules and transport codes. 2 - Method of solution: The PASC-1 system is based on AMPX-II/SCALE-3 modules. Except for some SCALE-3 modules taken from the SCALIAS package, the original AMPX-II modules were IBM versions written in FORTRAN IV. These modules have been translated into CDC FORTRAN V. In order to test these modules and link them with some codes, some of the sample problem calculations have been performed for the whole PASC-1 system. During these calculations, some FORTRAN-77 errors were found in MALOCS, REVERT, CONVERT and some subroutines of SUBLIB (FORTRAN-77 subroutine library). These errors have been corrected. Because many corrections were made for the REVERT module, it is renamed as REVERT-I (improved version of REVERT). After these corrections, the whole system is running on a CDC-CYBER Computer (NOS-BE operating system). 3 - Restrictions on the

Burnup calculations using Monte Carlo method

International Nuclear Information System (INIS)

Ghosh, Biplab; Degweker, S.B.

2009-01-01

In the recent years, interest in burnup calculations using Monte Carlo methods has gained momentum. Previous burn up codes have used multigroup transport theory based calculations followed by diffusion theory based core calculations for the neutronic portion of codes. The transport theory methods invariably make approximations with regard to treatment of the energy and angle variables involved in scattering, besides approximations related to geometry simplification. Cell homogenisation to produce diffusion, theory parameters adds to these approximations. Moreover, while diffusion theory works for most reactors, it does not produce accurate results in systems that have strong gradients, strong absorbers or large voids. Also, diffusion theory codes are geometry limited (rectangular, hexagonal, cylindrical, and spherical coordinates). Monte Carlo methods are ideal to solve very heterogeneous reactors and/or lattices/assemblies in which considerable burnable poisons are used. The key feature of this approach is that Monte Carlo methods permit essentially 'exact' modeling of all geometrical detail, without resort to ene and spatial homogenization of neutron cross sections. Monte Carlo method would also be better for in Accelerator Driven Systems (ADS) which could have strong gradients due to the external source and a sub-critical assembly. To meet the demand for an accurate burnup code, we have developed a Monte Carlo burnup calculation code system in which Monte Carlo neutron transport code is coupled with a versatile code (McBurn) for calculating the buildup and decay of nuclides in nuclear materials. McBurn is developed from scratch by the authors. In this article we will discuss our effort in developing the continuous energy Monte Carlo burn-up code, McBurn. McBurn is intended for entire reactor core as well as for unit cells and assemblies. Generally, McBurn can do burnup of any geometrical system which can be handled by the underlying Monte Carlo transport code

Study on MAs transmutation of accelerator-driven system sodium-cooled fast reactor loaded with metallic fuel

International Nuclear Information System (INIS)

Han Song; Yang Yongwei

2007-01-01

Through the analysis of the effect of heavy metal actinides on the effective multiplication constant (k eff ) of the core in accelerator-driven system (ADS) sodium-cooled fast reactor loaded with metallic fuel, we gave the method for determining fuel components. the characteristics of minor actinides (MAs) transmutation was analyzed in detail. 3D burn-up code COUPLE, which couples MCNP4c3 and ORIGEN2, was applied to the neutron simulation and burn up calculation. The results of optimized scheme shows that adjusting the proportion of 239 Pu and maintaining the value during the burn-up cycle is an efficient method of designing k eff and keeping stable during the burn-up cycle. Spallation neutrons lead to the neutron spectrum harder at inner core than that at outer core. It is in favor of improving MA's fission cross sections and the capture-to-fission ratio. The total MAs transmutation support ratio 8.3 achieves excellent transmutation effect. For higher flux at inner core leads to obvious differences on transmutation efficiency,only disposing MAs at inner core is in favor of decreasing the loading mass and improving MAs transmutation effect. (authors)

Neutron cross section library production code system for continuous energy Monte Carlo code MVP. LICEM

International Nuclear Information System (INIS)

Mori, Takamasa; Nakagawa, Masayuki; Kaneko, Kunio.

1996-05-01

A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author)

Neutron cross section library production code system for continuous energy Monte Carlo code MVP. LICEM

Energy Technology Data Exchange (ETDEWEB)

Mori, Takamasa; Nakagawa, Masayuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Kunio

1996-05-01

A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author).

Barriers to return to work after burn injuries.

Science.gov (United States)

Esselman, Peter C; Askay, Shelley Wiechman; Carrougher, Gretchen J; Lezotte, Dennis C; Holavanahalli, Radha K; Magyar-Russell, Gina; Fauerbach, James A; Engrav, Loren H

2007-12-01

To identify barriers to return to work after burn injury as identified by the patient. A cohort study with telephone interview up to 1 year. Hospital-based burn centers at 3 national sites. Hospitalized patients (N=154) meeting the American Burn Association criteria for major burn injury, employed at least 20 hours a week at the time of injury, and with access to a telephone after discharge. Patients were contacted via telephone every 2 weeks up to 4 months, then monthly up to 1 year after discharge. A return to work survey was used to identify barriers that prevented patients from returning to work. A graphic rating scale determined the impact of each barrier. By 1 year, 79.7% of patients returned to work. Physical and wound issues were barriers early after discharge. Although physical abilities continued to be a significant barrier up to 1 year, working conditions (temperature, humidity, safety) and psychosocial factors (nightmares, flashbacks, appearance concerns) became important issues in those with long-term disability. The majority of patients return to work after a burn injury. Although physical and work conditions are important barriers, psychosocial issues need to be evaluated and treated to optimize return to work.

Sodium spray and jet fire model development within the CONTAIN-LMR code

International Nuclear Information System (INIS)

Scholtyssek, W.

1993-01-01

An assessment was made of the sodium spray fire model implemented in the CONTAIN code. The original droplet burn model, which was based on the NACOM code, was improved in several aspects, especially concerning evaluation of the droplet burning rate, reaction chemistry and heat balance, spray geometry and droplet motion, and consistency with CONTAIN standards of gas property evaluation. An additional droplet burning model based on a proposal by Krolikowski was made available to include the effect of the chemical equilibrium conditions at the flame temperature. The models were validated against single-droplet burn experiments as well as spray and jet fire experiments. Reasonable agreement was found between the two burn models and experimental data. When the gas temperature in the burning compartment reaches high values, the Krolikowski model seems to be preferable. Critical parameters for spray fire evaluation were found to be the spray characterization, especially the droplet size, which largely determines the burning efficiency, and heat transfer conditions at the interface between the atmosphere and structures, which controls the thermal hydraulic behavior in the burn compartment

Concatenated coding system with iterated sequential inner decoding

DEFF Research Database (Denmark)

Jensen, Ole Riis; Paaske, Erik

1995-01-01

We describe a concatenated coding system with iterated sequential inner decoding. The system uses convolutional codes of very long constraint length and operates on iterations between an inner Fano decoder and an outer Reed-Solomon decoder......We describe a concatenated coding system with iterated sequential inner decoding. The system uses convolutional codes of very long constraint length and operates on iterations between an inner Fano decoder and an outer Reed-Solomon decoder...

Development a minimum data set of the information management system for burns.

Science.gov (United States)

Ahmadi, Maryam; Alipour, Jahanpour; Mohammadi, Ali; Khorami, Farid

2015-08-01

Burns are the most common and destructive injuries in across of the world and especially in developing countries. Nevertheless, a standard tool for collecting the data of burn injury has not been developed yet. The purpose of this study was to develop a minimum data set (MDS) of the information management system for burns in Iran. This descriptive and cross-sectional study was performed in 2014. Data were collected from hospitals affiliated with Hormozgan and Iran University of Medical Sciences and medical documents centers, emergency centers and legal medicine centers located in Bandar Abbas city, in addition to internet access and library. Investigated documents were burn injury records in 2013, and documents that retrieved from the internet, and printed materials. Records were selected randomly based on T20-T29 categories from ICD-10. Data were collected using a checklist. In order to make a consensus about the data elements the decision Delphi technique was applied using a questionnaire. The content validity and reliability of questionnaire were assessed by expert's opinions and test-retest method, respectively. An MDS of burns was developed. This MDS divided into two categories: administrative and clinical with six and 17 section and 161 and 311 data elements respectively. This study showed that comprehensive and uniform data elements about burns do not exist in Iran. Therefore a MDS was developed for burns in Iran. Development of an MDS will result in standardization and effective management of the data through providing uniform and comprehensive data elements for burns. Thus, comparability of the extracted information from different analyses and researches will be possible in various levels. In addition, establishment of policies and prevention and control of burns will be possible, which results in the improvement of the quality of care and containment of costs. Copyright © 2014 Elsevier Ltd and ISBI. All rights reserved.

Comparative calculations on selected two-phase flow phenomena using major PWR system codes

International Nuclear Information System (INIS)

1990-01-01

In 1988 a comparative study on important features and models in six major best estimate thermal hydraulic codes for PWR systems was implemented (Comparison of thermal hydraulic safety codes for PWR Graham, Trotman, London, EUR 11522). It was a limitation of that study that the source codes themselves were not available but the comparison had to be based on the available documentation. In the present study, the source codes were available and the capability of four system codes to predict complex two-phase flow phenomena has been assessed. Two areas of investigation were selected: (a) pressurized spray phenomena; (b) boil-up phenomena in rod bundles. As regards the first area, experimental data obtained in 1972 on the Neptunus Facility (Delft University of Technology) were compared with the results of the calculations using Athlet, Cathare, Relap 5 and TRAC-PT1 and, concerning the second area, the results of two experimental facilities obtained in 1980 and 1985 on Thetis (UKEA) and Pericles (CEA-Grenoble) were considered

Container code recognition in information auto collection system of container inspection

International Nuclear Information System (INIS)

Su Jianping; Chen Zhiqiang; Zhang Li; Gao Wenhuan; Kang Kejun

2003-01-01

Now custom needs electrical application and automatic detection. Container inspection should not only give the image of the goods, but also auto-attain container's code and weight. Its function and track control, information transfer make up the Information Auto Collection system of Container Inspection. Code Recognition is the point. The article is based on model match, the close property of character, and uses it to recognize. Base on checkout rule, design the adjustment arithmetic, form the whole recognition strategy. This strategy can achieve high recognition ratio and robust property

The effect of preexisting respiratory co-morbidities on burn outcomes☆

Science.gov (United States)

Knowlin, Laquanda T.; Stanford, Lindsay B.; Cairns, Bruce A.; Charles, Anthony G.

2018-01-01

Introduction Burns cause physiologic changes in multiple organ systems in the body. Burn mortality is usually attributable to pulmonary complications, which can occur in up to 41% of patients admitted to the hospital after burn. Patients with preexisting comorbidities such as chronic lung diseases may be more susceptible. We therefore sought to examine the impact of preexisting respiratory disease on burn outcomes. Methods A retrospective analysis of patients admitted to a regional burn center from 2002–2012. Independent variables analyzed included basic demographics, burn mechanism, presence of inhalation injury, TBSA, pre-existing comorbidities, smoker status, length of hospital stay, and days of mechanical ventilation. Bivariate analysis was performed and Cox regression modeling using significant variables was utilized to estimate hazard of progression to mechanical ventilation and mortality. Results There were a total of 7640 patients over the study period. Overall survival rate was 96%. 8% (n=672) had a preexisting respiratory disease. Chronic lung disease patients had a higher mortality rate (7%) compared to those without lung disease (4%, pburn. Given the increasing number of Americans with chronic respiratory diseases, there will likely be a greater number of individuals at risk for worse outcomes following burn. PMID:28341260