Reactivity effects of fission product decay in PWRs

International Nuclear Information System (INIS)

Aragones, J.M.; Ahnert, C.

1988-01-01

The purpose of the work reported in this paper is to analyze the effects of fission product chains with radioactive decay on the reactivity in pressurized water reactor (PWR) cores, calculating their accumulation and absorption rates along fuel burnup at continuous operation and after shutdown periods extending from 1 day to a few months. The authors PWR version of the WIMS-D4 code is first used to obtain the individual number densities, absorption rates, and averaged cross sections for every nuclide of the fission product chains with significant decay rates, as a function of fuel burnup at continuous irradiation. Next, by an auxiliary ad hoc code, these data, have been processed together with the required one for fissile nuclides and boron, also taken from WIMS at each burnup step, to calculate the average or effective values relevant for the analysis and the decay and change in overall absorption after several shutdown times. (1) The reactivity effect of fission product decay changes significantly with the shutdown time. The maximum absorption increase by decay is reached in ∼ 10 days' shutdown. (2) The dependence with fuel type, enrichment, and burnup is slight, but the change with previous power density is nearly linear, which might be significant after coast-down in previous cycles. (3) For long shutdown periods, the overall reactivity effect of decay in the three fission product chains considered is much less than if only the samarium peak due to 149 Nd is considered

Fission yields data generation and benchmarks of decay heat estimation of a nuclear fuel

Science.gov (United States)

Gil, Choong-Sup; Kim, Do Heon; Yoo, Jae Kwon; Lee, Jounghwa

2017-09-01

Fission yields data with the ENDF-6 format of 235U, 239Pu, and several actinides dependent on incident neutron energies have been generated using the GEF code. In addition, fission yields data libraries of ORIGEN-S, -ARP modules in the SCALE code, have been generated with the new data. The decay heats by ORIGEN-S using the new fission yields data have been calculated and compared with the measured data for validation in this study. The fission yields data ORIGEN-S libraries based on ENDF/B-VII.1, JEFF-3.1.1, and JENDL/FPY-2011 have also been generated, and decay heats were calculated using the ORIGEN-S libraries for analyses and comparisons.

On Error Analysis of ORIGEN Decay Data Library Based on ENDF/B-VII.1 via Decay Heat Estimation after a Fission Event

Energy Technology Data Exchange (ETDEWEB)

Kim, Do Heon; Gil, Choong-Sup; Lee, Young-Ouk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

The method is strongly dependent on the available nuclear structure data, i.e., fission product yield data and decay data. Consequently, the improvements in the nuclear structure data could have guaranteed more reliable decay heat estimation for short cooling times after fission. The SCALE-6.1.3 code package includes the ENDF/B-VII.0-based fission product yield data and ENDF/B-VII.1-based decay data libraries for the ORIGEN-S code. The generation and validation of the new ORIGEN-S yield data libraries based on the recently available fission product yield data such as ENDF/B-VII.1, JEFF-3.1.1, JENDL/FPY-2011, and JENDL-4.0 have been presented in the previous study. According to the study, the yield data library in the SCALE-6.1.3 could be regarded as the latest one because it resulted in almost the same outcomes as the ENDF/B-VII.1. A research project on the production of the nuclear structure data for decay heat estimation of nuclear fuel has been carried out in Korea Atomic Energy Research Institute (KAERI). The data errors contained in the ORIGEN-S decay data library of SCALE-6.1.3 have been clearly identified by their changing variables. Also, the impacts of the decay data errors have been analyzed by estimating the decay heats for the fission product nuclides and their daughters after {sup 235}U thermal-neutron fission. Although the impacts of decay data errors are quite small, it reminds us the possible importance of decay data when estimating the decay heat for short cooling times after a fission event.

A revised ANS standard for decay heat from fission products

International Nuclear Information System (INIS)

Schrock, V.E.

1978-01-01

The draft ANS 5.1 standard on decay heat was published in 1971 and given minor revision in 1973. Its basis was the best estimate working curve developed by K. Shure in 1961. Liberal uncertainties were assigned to the standard values because of lack of data for short cooling times and large discrepancies among experimental data. Research carried out over the past few years has greatly improved the knowledge of this phenomenon and a major revision of the standard has been completed. Very accurate determination of the decay heat is now possible, expecially within the first 10 4 seconds, where the influence of neutron capture in fission products may be treated as a small correction to the idealized zero capture case. The new standard accounts for differences among fuel nuclides. It covers cooling time to 10 9 seconds, but provides only an ''upper bound'' on the capture correction in the interval 10 4 9 seconds. (author)

Photoproduction data for heating calculations

International Nuclear Information System (INIS)

Van der Marck, Steven C.; Koning, Arjan J.; Rochman, Dimitri

2008-01-01

For irradiations in a materials test reactor, the prediction of the amount of gamma heating in the reactor is important. Only a good predictive calculation will lead to an irradiation in which the specified temperatures are reached. The photons produced by fission product decay are often missing in spectrum calculations for a reactor, but the contribution of the photons can be computed effectively using engineering correlations for the amount of fission product decay and the ensuing photon spectrum. The prompt photons are usually calculated by a spectrum code based on the underlying nuclear data libraries. For most of the important nuclides, the nuclear data libraries contain data for the photon productions rates. However, there are still many nuclides for which the photon production data are missing, and some of these nuclides contribute to gamma heating. In this paper it is estimated what the contributions to heating are from photon production on nuclides such as 236 U, 238 Pu, 135 I, 135 Xe, 147 Pm, 148 Pm, 148m Pm, and 149 Sm. Also, simple arguments are given to judge the effect from photon production on all other (lumped) fission products, and from 28 Al decay. For all these calculations the High Flux Reactor is used as an example. (authors)

[Fission product yields of 60 fissioning reactions]. Final report

International Nuclear Information System (INIS)

Rider, B.F.

1995-01-01

In keeping with the statement of work, I have examined the fission product yields of 60 fissioning reactions. In co-authorship with the UTR (University Technical Representative) Talmadge R. England ''Evaluation and Compilation of Fission Product Yields 1993,'' LA-UR-94-3106(ENDF-349) October, (1994) was published. This is an evaluated set of fission product Yields for use in calculation of decay heat curves with improved accuracy has been prepared. These evaluated yields are based on all known experimental data through 1992. Unmeasured fission product yields are calculated from charge distribution, pairing effects, and isomeric state models developed at Los Alamos National Laboratory. The current evaluation has been distributed as the ENDF/B-VI fission product yield data set

Application of least-squares method to decay heat evaluation

International Nuclear Information System (INIS)

Schmittroth, F.; Schenter, R.E.

1976-01-01

Generalized least-squares methods are applied to decay-heat experiments and summation calculations to arrive at evaluated values and uncertainties for the fission-product decay-heat from the thermal fission of 235 U. Emphasis is placed on a proper treatment of both statistical and correlated uncertainties in the least-squares method

Detailed comparison between decay heat data calculated by the summation method and integral measurements

International Nuclear Information System (INIS)

Rudstam, G.

1979-01-01

The fission product library FPLIB has been used for a calculation of the decay heat effect in nuclear fuel. The results are compared with integral determinations and with results obtained using the ENDF/BIV data base. In the case of the beta part, and also for the total decay heat, the FPLIB-data seem to be superior to the ENDF/BIV-data. The experimental integral data are in many cases reproduced within the combined limits of error of the methods. (author)

TMI-2 decay power: LASL fission-product and actinide decay power calculations for the President's Commission at Three Mile Island

International Nuclear Information System (INIS)

England, T.R.; Wilson, W.B.

1979-10-01

Fission-product and actinide decay heating, gas content, curies, and detailed contributions of the most important nuclide contributors were supplied in a series of letters following requests from the Presidential Commission on the Accident at Three Mile Island. In addition, similar data assuming different irradiation (power) histories were requested for purposes of comparison. This report consolidates the tabular and graphical data supplied and explains its basis

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).

Decay Heat Calculations for Reactors: Development of a Computer Code ADWITA

International Nuclear Information System (INIS)

Raj, Devesh

2015-01-01

Estimation of release of energy (decay heat) over an extended period of time after termination of neutron induced fission is necessary for determining the heat removal requirements when the reactor is shutdown, and for fuel storage and transport facilities as well as for accident studies. A Fuel Cycle Analysis Code, ADWITA (Activation, Decay, Waste Incineration and Transmutation Analysis) which can generate inventory based on irradiation history and calculate radioactivity and decay heat for extended period of cooling, has been written. The method and data involved in Fuel Cycle Analysis Code ADWITA and some results obtained shall also be presented. (author)

Map of calculated radioactivity of fission product, (4)

International Nuclear Information System (INIS)

Takeda, Tsuneo

1978-07-01

The overall radioactivities of fission products depending on irradiation time and cooling time were calculated for 18 different neutron fluxes, which are presented in contour maps and tables. Irradiation condition etc. are the followings: neutron flux (n sub(th)) 1 x 10 12 - 6.8 x 10 14 n/cm 2 /sec, uranium quantity 1 mole (6 x 10 23 atoms, ca. 271 g UO 2 ), U-235 enrichment 2.7%, irradiation time 60. - 6 x 10 7 sec (1 min - 1.9 y), cooling time 0. and 60. - 6 x 10 7 sec (1 min - 1.9 y). The enrichment value represents those for LWRs. To calculate the overall radioactivities, 595 fission product nuclides were introduced. Overall radioactivities calculations were made for 68,000 combinations of irradiation time, cooling time and neutron flux. The many complex decay chains of fission products were treated with CODAC-No.6 computer code. (author)

TMI-2 decay power: LASL fission-product and actinide decay power calculations for the President's Commission at Three Mile Island

Energy Technology Data Exchange (ETDEWEB)

England, T.R.; Wilson, W.B.

1979-10-01

Fission-product and actinide decay heating, gas content, curies, and detailed contributions of the most important nuclide contributors were supplied in a series of letters following requests from the Presidential Commission on the Accident at Three Mile Island. In addition, similar data assuming different irradiation (power) histories were requested for purposes of comparison. This report consolidates the tabular and graphical data supplied and explains its basis.

Easy-to-use application programs for decay heat and delayed neutron calculations on personal computers

Energy Technology Data Exchange (ETDEWEB)

Oyamatsu, Kazuhiro [Nagoya Univ. (Japan)

1998-03-01

Application programs for personal computers are developed to calculate the decay heat power and delayed neutron activity from fission products. The main programs can be used in any computers from personal computers to main frames because their sources are written in Fortran. These programs have user friendly interfaces to be used easily not only for research activities but also for educational purposes. (author)

TMI-2 decay power: LASL fission-product and actinide decay power calculations for the President's commission on the accident at Three Mile Island

International Nuclear Information System (INIS)

England, T.R.; Wilson, W.B.

1980-03-01

Fission-product and actinide decay heating, gas content, curies, and detailed contributions of the most important nuclide contributors were supplied in a series of letters following requests from the Presidential Commission on the Accident at Three Mile Island. In addition, similar data assuming different irradiation (power) histories were requested for purposes of comparison. This report consolidates the tabular and graphical data supplied and explains its basis

Progress in fission product nuclear data

International Nuclear Information System (INIS)

Lammer, M.

1984-09-01

This is the tenth issue of a report series on Fission Product Data, which informs us about all the activities in this field, which are planned, ongoing, or have recently been completed. The types of activities included are measurements, compilations and evaluations of: fission product yields (neutron induced and spontaneous fission), neutron reaction cross sections of fission products, data related to the radioactive decay of fission products, delayed neutron data of fission products, lumped fission product data (decay heat, absorption, etc.). There is also a section with recent references relative to fission product nuclear data

Fission product yield evaluation for the USA evaluated nuclear data files

International Nuclear Information System (INIS)

Rider, B.F.; England, T.R.

1994-01-01

An evaluated set of fission product yields for use in calculation of decay heat curves with improved accuracy has been prepared. These evaluated yields are based on all known experimental data through 1992. Unmeasured fission product yields are calculated from charge distribution, pairing effects, and isomeric state models developed at Los Alamos National Laboratory. The current evaluation has been distributed as the ENDF/B-VI fission product yield data set

Libraries of decay data and fission product yields in the ABBN-93 constant set

International Nuclear Information System (INIS)

Zabrodskaya, S.V.; Nikolaev, M.N.; Tsibulya, A.M.

2001-01-01

This paper describes three new libraries in the Abb. constant set which are essential for calculating radioactivity: basic decay data, radioactive decay photon spectra and fission product yields. (author)

Decay heat rates calculated using ORIGEN-S and CINDER10 with common data libraries

International Nuclear Information System (INIS)

Brady, M.C.; Hermann, O.W.; Beard, C.A.; Bohnhoff, W.J.; England, T.R.

1991-01-01

A set of two benchmark problems were proposed as part of an international comparison of decay heat codes. Problem specifications included explicit fission-yield, decay and capture data libraries to be used in the calculations. This paper describes the results obtained using these common data to perform the benchmark calculations with two popular depletion codes, ORIGEN-S and CINDER10. Short descriptions of the methods used by each of these codes are also presented. Results from other contributors to the international comparison are discussed briefly. This comparison of decay heat codes using common data libraries demonstrates that discrepant results in calculated decay heat rates are the result of differences in the nuclear data input to the codes and not the method of solution. 15 refs., 2 figs., 8 tabs

Delayed neutron spectra and their uncertainties in fission product summation calculations

Energy Technology Data Exchange (ETDEWEB)

Miyazono, T.; Sagisaka, M.; Ohta, H.; Oyamatsu, K.; Tamaki, M. [Nagoya Univ. (Japan)

1997-03-01

Uncertainties in delayed neutron summation calculations are evaluated with ENDF/B-VI for 50 fissioning systems. As the first step, uncertainty calculations are performed for the aggregate delayed neutron activity with the same approximate method as proposed previously for the decay heat uncertainty analyses. Typical uncertainty values are about 6-14% for {sup 238}U(F) and about 13-23% for {sup 243}Am(F) at cooling times 0.1-100 (s). These values are typically 2-3 times larger than those in decay heat at the same cooling times. For aggregate delayed neutron spectra, the uncertainties would be larger than those for the delayed neutron activity because much more information about the nuclear structure is still necessary. (author)

Progress in fission product nuclear data

International Nuclear Information System (INIS)

Lammer, G.

1976-05-01

The purpose of this series is to inform scientists working on Fission Product Nuclear Data, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. This report consists of reproductions of essentially unaltered original contributions which the authors have sent to IAEA/NDS. The types of activities being included in this report are measurements, compilations and evaluations of: fission product yields; neutron cross-section data of fission products; data related to β-, γ-decay of fission products; delayed neutron data; and fission product decay-heat

JNDC nuclear data library of fission products

International Nuclear Information System (INIS)

Tasaka, Kanji; Ihara, Hitoshi; Akiyama, Masatsugu; Yoshida, Tadashi; Matumoto, Zyun-itiro; Nakasima, Ryuzo

1983-10-01

The JNDC (Japanese Nuclear Data Committee) FP (Fission Product) nuclear data library for 1172 fission products is described in this report. The gross theory of beta decay has been used extensively for estimating unknown decay data and also some of known decay data with poor accuracy. The calculated decay powers of fission products using the present library show excellent agreement with the latest measurements at ORNL (Oak Ridge National Laboratory), LANL (Los Alamos National Laboratory) and UTT (University of Tokyo, Tokai) for cooling times shorter than 10 3 s after irradiation. The calculated decay powers by the existing libraries showed systematic deviations at short cooling times; the calculated beta and gamma decay powers after burst fission were smaller than the experimental results for cooling times shorter than 10 s, and in the cooling time range 10 to 10 3 s the beta-decay power was larger than the measured values and the gamma decay power smaller than the measured results. The present JNDC FP nuclear data library resolved these discrepancies in the short cooling time ranges. The decay power of fission products has been calculated for ten fission types and the results have been fitted by an analytical function with 31 exponentials. This permits the easy application of the present results of decay power calculations to a LOCA (Loss-of-Coolant Accident) analysis of a light water reactor and so on. (author)

CINDER, Depletion and Decay Chain Calculation for Fission Products in Thermal Reactors

International Nuclear Information System (INIS)

England, T.R.; Gorrell, T.C.; Hightower, J.H.

2001-01-01

1 - Description of problem or function: CINDER is a four-group, one- point depletion and fission product program based on the evaluation of a general analytical solution of nuclides coupled in any linear sequence of radioactive decays and neutron absorptions in a specified neutron flux spectrum. The desired depletion and fission product chains and all physical data are specified by the problem originator. The program computes individual nuclide number densities, activities, nine energy-group disintegration rates, and macroscopic and barns/fission poisons at each time-step as well as selected summaries of these data. 2 - Method of solution: Time-dependent variations in nuclide cross sections and neutron fluxes are approximated by a user-specified sequential set of values which are considered constant during the duration of the user-specified associated time-increments. When a nuclide concentration is independent of the concentration of any of its progeny, it is possible to resolve the couplings so as to obtain nuclides fed by a single parent. These chains are referred to as linear. 3 - Restrictions on the complexity of the problem: The program is limited to 500 total nuclides formed in up to 240 chains of 20 or fewer nuclides each. Up to 10 nuclides may act as fission product sources, contributing to power, and as many as 99 time-steps of arbitrary length are permitted. All stable nuclides must have a cross section if zero power time-increments are anticipated

Beta-decay and decay heat. Summary report of consultants' meeting

International Nuclear Information System (INIS)

Nicols, A.L.

2006-01-01

Experts on decay data and decay heat calculations participated in a Consultants' Meeting organized at IAEA Headquarters on 12-14 December 2005. Debate focused on the validation of decay heat calculations as a function of cooling time for fuel irradiated in power reactors through comparisons with experimental benchmark data. Both the current understanding and quantification of mean beta and gamma decay energies were reviewed with respect to measurements and the Gross Theory of Beta Decay. Particular emphasis was placed on the known development of total absorption gamma-ray spectroscopy (TAGS), and detailed discussions took place to formulate the measurement requirements for mean beta and gamma data of individual radionuclides. This meeting was organized in cooperation with the OECD/NEA Working Party for Evaluation and Cooperation (WPEC). Proposals and recommendations were made to resolve particular difficulties, and an initial list of fission products was produced for TAGS studies. The discussions, conclusions and recommendations of the meeting are briefly described in this report. (author)

Progress in fission product nuclear data

International Nuclear Information System (INIS)

Lammer, G.

1975-01-01

This is the first issue of a report series on Fission Product Nuclear Data (FPND), published every six months by the Nuclear Data Section (NDS) of the International Atomic Energy Agency (IAEA). Its purpose is to inform scientists working on FPND, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. The types of activities being included in this report are measurements, compilations and evaluations of: fission product yields; neutron cross-section data of fission products; data related to β-, γ-decay of fission products; delayed neutron data; and fission product decay-heat. The present issue includes contributions which were received by NDS before 1 November 1975

Computer program FPIP-REV calculates fission product inventory for U-235 fission

Science.gov (United States)

Brown, W. S.; Call, D. W.

1967-01-01

Computer program calculates fission product inventories and source strengths associated with the operation of U-235 fueled nuclear power reactor. It utilizes a fission-product nuclide library of 254 nuclides, and calculates the time dependent behavior of the fission product nuclides formed by fissioning of U-235.

Consistency among integral measurements of aggregate decay heat power

Energy Technology Data Exchange (ETDEWEB)

Takeuchi, H.; Sagisaka, M.; Oyamatsu, K.; Kukita, Y. [Nagoya Univ. (Japan)

1998-03-01

Persisting discrepancies between summation calculations and integral measurements force us to assume large uncertainties in the recommended decay heat power. In this paper, we develop a hybrid method to calculate the decay heat power of a fissioning system from those of different fissioning systems. Then, this method is applied to examine consistency among measured decay heat powers of {sup 232}Th, {sup 233}U, {sup 235}U, {sup 238}U and {sup 239}Pu at YAYOI. The consistency among the measured values are found to be satisfied for the {beta} component and fairly well for the {gamma} component, except for cooling times longer than 4000 s. (author)

JNDC FP decay data file

International Nuclear Information System (INIS)

Yamamoto, Tohru; Akiyama, Masatsugu

1981-02-01

The decay data file for fission product nuclides (FP DECAY DATA FILE) has been prepared for summation calculation of the decay heat of fission products. The average energies released in β- and γ-transitions have been calculated with computer code PROFP. The calculated results and necessary information have been arranged in tabular form together with the estimated results for 470 nuclides of which decay data are not available experimentally. (author)

The use of averages and other summation quantities in the testing of evaluated fission product yield and decay data. Applications to ENDF/B(IV)

International Nuclear Information System (INIS)

Walker, W.H.

1976-01-01

Averages of some fission product properties can be obtained by multiplying the fission product yield for each fission product by the value of the property (e.g. mass, atomic number, mass defect) for that fission product and summing all significant contributions. These averages can be used to test the reliability of the yield set or provide useful data for reactor calculations. The report gives the derivation of these averages and discusses their application using the ENDF/B(IV) fission product library. The following quantities are treated here: the number of fission products per fission ΣYsub(i); the average mass number and the average number of neutrons per fission; the average atomic number of the stable fission products and the average number of β-decays per fission; the average mass defect of the stable fission products and the total energy release per fission; the average decay energy per fission (beta, gamma and anti-neutrino); the average β-decay energy per fission; individual and group-averaged delayed neutron emission; the total yield for each fission product element. Wherever it is meaningful to do so, a sum is subdivided into its light and heavy mass components. The most significant differences between calculated values based on ENDF/B(IV) and measurements are the β and γ decay energies for 235 U thermal fission and delayed neutron yields for other fissile nuclides, most notably 238 U. (author)

A gamma heating calculation methodology for research reactor application

International Nuclear Information System (INIS)

Lee, Y.K.; David, J.C.; Carcreff, H.

2001-01-01

Gamma heating is an important issue in research reactor operation and fuel safety. Heat deposition in irradiation targets and temperature distribution in irradiation facility should be determined so as to obtain the optimal irradiation conditions. This paper presents a recently developed gamma heating calculation methodology and its application on the research reactors. Based on the TRIPOLI-4 Monte Carlo code under the continuous-energy option, this new calculation methodology was validated against calorimetric measurements realized within a large ex-core irradiation facility of the 70 MWth OSIRIS materials testing reactor (MTR). The contributions from prompt fission neutrons, prompt fission γ-rays, capture γ-rays and inelastic γ-rays to heat deposition were evaluated by a coupled (n, γ) transport calculation. The fission product decay γ-rays were also considered but the activation γ-rays were neglected in this study. (author)

Decay heat uncertainty quantification of MYRRHA

Directory of Open Access Journals (Sweden)

Fiorito Luca

2017-01-01

Full Text Available MYRRHA is a lead-bismuth cooled MOX-fueled accelerator driven system (ADS currently in the design phase at SCK·CEN in Belgium. The correct evaluation of the decay heat and of its uncertainty level is very important for the safety demonstration of the reactor. In the first part of this work we assessed the decay heat released by the MYRRHA core using the ALEPH-2 burnup code. The second part of the study focused on the nuclear data uncertainty and covariance propagation to the MYRRHA decay heat. Radioactive decay data, independent fission yield and cross section uncertainties/covariances were propagated using two nuclear data sampling codes, namely NUDUNA and SANDY. According to the results, 238U cross sections and fission yield data are the largest contributors to the MYRRHA decay heat uncertainty. The calculated uncertainty values are deemed acceptable from the safety point of view as they are well within the available regulatory limits.

The effect of load factor on fission product decay heat from discharged reactor fuel

International Nuclear Information System (INIS)

Davies, B.S.J.

1978-07-01

A sum-of-exponentials expression representing the decay heat power following a burst thermal irradiation of 235 U has been used to investigate the effect of load factor during irradiation on subsequent decay heat production. A sequence of random numbers was used to indicate reactor 'on' and 'off' periods for irradiations which continued for a total of 1500 days at power and were followed by 100 days cooling. It was found that for these conditions decay heat is almost proportional to load factor. Estimates of decay heat uncertainty arising from the random irradiation pattern are also given. (author)

Calculated apparent yields of rare gas fission products

International Nuclear Information System (INIS)

Delucchi, A.A.

1975-01-01

The apparent fission yield of the rare gas fission products from four mass chains is calculated as a function of separation time for six different fissioning systems. A plot of the calculated fission yield along with a one standard deviation error band is given for each rare gas fission product and for each fissioning system. Those parameters in the calculation that were major contributors to the calculated standard deviation at each separation time were identified and the results presented on a separate plot. To extend the usefulness of these calculations as new and better values for the input parameters become available, a third plot was generated for each system which shows how sensitive the derived fission yield is to a change in any given parameter used in the calculation. (U.S.)

ENDF/B fission product decay data

International Nuclear Information System (INIS)

Rose, P.F.; Burrows, T.W.

1976-08-01

The fission product data have been organized by A-chains in order of ascending A from A = 72 to A = 167. The heading page is followed by more detailed information on the individual members of the chain in order of increasing Z and decreasing metastable state. The detailed information for each member includes the ENDF/B-IV File 1 comments and references if available and applicable to the decay data. Following the comments is a decay scheme of the nuclide tabulating the quantities T/sub 1 / 2 /, Q, branching ratio (BR), (E/sub γ/), (E/sub β/), and (E/sub α/). Uncertainties are given if available in the file. Independent fission yields are given, as well as thermal cross sections and resonance integrals as obtained from ENDF/B-IV. All energies listed in this publication are in keV, and all branching ratios (BR) sum to unity. If there are spectra in the decay data file, the decay scheme is followed by tables of photon, particle, and characteristic radiation. For cases in which the multipolarities could be obtained from the file the tables also contain information on x-rays, conversion electrons, and Auger electrons. Associated with the photon and particle radiation tables are the appropriate average energies per decay for each type of radiation, including neutrino radiation

Approximation of the decay of fission and activation product mixtures

International Nuclear Information System (INIS)

Henderson, R.W.

1991-01-01

The decay of the exposure rate from a mixture of fission and activation products is a complex function of time. The exact solution of the problem involves the solution of more than 150 tenth order Bateman equations. An approximation of this function is required for the practical solution of problems involving multiple integrations of this function. Historically this has been a power function, or a series of power functions, of time. The approach selected here has been to approximate the decay with a sum of exponential functions. This produces a continuous, single valued function, that can be made to approximate the given decay scheme to any desired degree of closeness. Further, the integral of the sum is easily calculated over any period. 3 refs

ORIGEN2.1 Cycle Specific Calculation of Krsko Nuclear Power Plant Decay Heat and Core Inventory

International Nuclear Information System (INIS)

Vukovic, J.; Grgic, D.; Konjarek, D.

2010-01-01

This paper presents ORIGEN2.1 computer code calculation of Krsko Nuclear Power Plant core for Cycle 24. The isotopic inventory, core activity and decay heat are calculated in one run for the entire core using explicit depletion and decay of each fuel assembly. Separate pre-ori application which was developed is utilized to prepare corresponding ORIGEN2.1 inputs. This application uses information on core loading pattern to determine fuel assembly specific depletion history using 3D burnup which is obtained from related PARCS computer code calculation. That way both detailed single assembly calculations as well as whole core inventory calculations are possible. Because of the immense output of the ORIGEN2.1, another application called post-ori is used to retrieve and plot any calculated property on the basis of nuclide, element, summary isotope or group of elements for activation products, actinides and fission products segments. As one additional possibility, with the post-ori application it is able to calculate radiotoxicity from calculated ORIGEN2.1 inventory. The results which are obtained using the calculation model of ORIGEN2.1 computer code are successfully compared against corresponding ORIGEN-S computer code results.(author).

Progress in fission product nuclear data. No. 13

International Nuclear Information System (INIS)

Lammer, M.

1990-11-01

This is the 13th issue of a report series published by the Nuclear Data Section of the IAEA. The types of activities included are measurements, compilations and evaluations of: Fission product yields (neutron induced and spontaneous fission), neutron reaction cross-sections of fission products, data related to the radioactive decay of fission products, delayed neutron data of fission products and bumped fission product data (decay heat, absorption, etc.). The first part of the report consists of unaltered original data which the authors have sent to IAEA/NDS. The second part contains some recent references relative to fission product nuclear data, which were not covered by the contributions submitted, and selected papers from conferences. Part 3 contains requirements for further measurements

Volatilization and reaction of fission products in flowing steam

International Nuclear Information System (INIS)

Johnson, I.; Steidl, D.V.; Johnson, C.E.

1985-01-01

The principal risk to the public from nuclear power plants derives from the highly radioactive atoms (fission products) generated as energy is produced in the nuclear fuel. The revolatilization of fission products from reactor system surfaces due to self-heating by radioactive decay has become a complicating factor in the source-term redefinition effort. It has had a major impact on calculations of fission product distributions in accident safety analyses. The focus of this research effort was to investigate the volatilization and transport of fission products and control rod materials in a flowing gaseous steam-hydrogen mixture. Fission product and control rod materials in various combinations were studied including CsI, CsOH, TeO 2 , SrO, Ag, In, Cd and Mn. The vaporization behavior of the deposits were characterized with respect to vaporization rates, chemical species and downstream transport behavior

Progress in fission product nuclear data

International Nuclear Information System (INIS)

Lammer, M.

1981-06-01

This is the seventh issue of a report series on Fission Product Nuclear Data (FPND) which is published by the Nuclear Data Section (NDS) of the International Atomic Energy Agency (IAEA). The purpose of this series is to inform scientists working on FPND, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. The present issue contains also a section with some recent references relative to fission product nuclear data, which were not covered by the contributions submitted. The types of activities being included in this report are measurements, compilations and evaluations of: fission product yields (neutron induced and spontaneous fission); neutron reaction cross sections of fission products; data related to the radioactive decay of fission products; delayed neutron data of fission products; and lumped fission product data (decay heat, absorption etc.). The sixth issue of this series has been published in June 1980 as INDC(NDS)-113/G+P. The present issue includes contributions which were received by NDS between 1 August 1980 and 25 May 1981

Status of the Japanese decay heat standard

International Nuclear Information System (INIS)

Katakura, Jun-ichi

1992-01-01

Fission product decay heat power plays an important role in the safety evaluation of nuclear power plants, especially for the analysis of hypothetical reactor accident scenarios. The ANS-5.1 decay heat standard for safety evaluation issued in 1979 has been used widely, even in Japan. Since the issuance of the standard, several improvements have been made to measurements and summation calculations. Summation calculations, in particular, have improved because of the adoption of theoretically calculated decay energies for nuclides with incomplete decay data. Taking into consideration those improvements, the Atomic Energy Society of Japan (AESJ) organized a research committee on a standard for decay heat power in nuclear reactors in 1987. The committee issued its recommendation after more than 2 yr discussion. After the AESJ recommendation, the Nuclear Safety Commission of Japan also began to discuss whether the recommendation should be included in its regulatory guide. The commission concluded in 1992 that the recommendation should be approved for licensing analysis of reactors if three times the uncertainties attached to the recommendation are included in the analysis. The AESJ recommendation may now be used for the safety evaluation of reactors in Japan in addition to the standards already used, which include ANS-5.1 (1973), General Electric Corporation (GE) curve, and ANS-5.1 (1979)

Progress in fission product nuclear data

International Nuclear Information System (INIS)

Lammer, M.

1983-08-01

This is the ninth issue of a report series on Fission Product Nuclear Data (FPND) which is published by the Nuclear Data Section (NDS) of the International Atomic Energy Agency (IAEA). The purpose of this series is to inform scientists working on FPND, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. The main part of this report consists of unaltered original contributions which the authors have sent to IAEA/NDS. The present issue contains also a section with some recent references relative to fission product nuclear data, which were not covered by the contributions submitted. The types of activities being included in this report are measurements, compilations and evaluations of: Fission product yields (neutron induced and spontaneous fission); Neutron reaction cross sections of fission products; Data related to the radioactive decay of fission products; Delayed neutron data of fission products; and lumped fission product data (decay heat, absorption etc.). The eighth issue of this series has been published in July 1982 as INDC(NDS)-130. The present issue includes contributions which were received by NDS between 1 August 1982 and 25 June 1983

Progress in fission product nuclear data. Information about activities in the field of measurements and compilations/evaluations of fission product nuclear data (FPND)

International Nuclear Information System (INIS)

Lammer, G.

1978-07-01

This is the fourth issue of a report series on Fission Product Nuclear Data (FPND) which is published by the Nuclear Data Section (NDS) of the International Atomic Energy Agency (IAEA). The purpose of this series is to inform scientists working on FPND, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. The main part of this report consists of unaltered original contributions which the authors have sent to IAEA/NDS. The types of activities being included in this report are measurements, compilations and evaluations of: Fission product yields (neutron induced and spontaneous fission); neutron reaction cross sections of fission products; data related to the radioactive decay of fission products; delayed neutron data of fission products; and lumped fission product data (decay heat, absorption etc.)

Status of decay data of fission products

International Nuclear Information System (INIS)

Blachot, J.

1978-01-01

Fission products (F.P.) are neutron rich isotopes ranging from Zn to Tm. The status of decay data of F.P. was described at the Bologna Panel 1973 by Rudstam. Since then, FPND have improved in general, but still much is valid of what Rudstam said about the accuracies of FPND. The lack of decay data for the short lived F.P. has been considerably reduced, and some of the short lived F.P. have now well studied decay data. The present status of decay data is given in this review, which is composed of six sections. In the first one, the principal new facilities used in decay data measurements are reviewed. The second part is devoted to the total decay energy (Q). In the third Section, the half lives are treated. In the fourth and fifth Sections, beta and gamma energies and intensities, and also average values are discussed. Finally, the last Section considers the different files and compilations devoted to the decay of F.P

Calculational tracking of decay heat for FFTF plant

International Nuclear Information System (INIS)

Cillan, T.F.; Carter, L.L.

1985-01-01

A detailed calculational monitoring of decay heat for each assembly on the Fast Flux Test Facility (FFTF) plant is obtained by utilizing a decay heat data base and user friendly computer programs to access the data base. Output includes the time-dependent decay heat for an assembly or a specific set of assemblies, and optional information regarding the curies of activated nuclides along the axial length of the assembly. The decay heat data base is updated periodically, usually at the end of each irradiation cycle. 1 ref., 2 figs

Search for spontaneous fission of 226Ra and systematics of the spontaneous fission, α-decay and cluster decay probabilities

International Nuclear Information System (INIS)

Mikheev, V.L.; Tret'yakova, S.P.; Golovchenko, A.N.; Timofeeva, O.V.; Hussonnois, M.; Le Naour, C.

1998-01-01

The low limit of the 226 Ra spontaneous fission half-life corresponding to T 1/2 ≥ 4 · 10 18 years is measured. The 226 Ra spontaneous fission probability proved to be about 50 times less than the value expected from the known systematics, connecting the ratios of theα-decay and spontaneous fission probabilities with the fissility parameter Z 2 /A. It is shown that the probabilities of spontaneous fission, α-decay and cluster decay can be systematized in the same way according to the difference between the decay products Coulomb energy near the scission point and decay energy Q

Beta and gamma decay heat measurements between 0.1s--50,000s for neutron fission of 235U, 238U and 239Pu. Final report, June 1, 1992--December 31, 1996

International Nuclear Information System (INIS)

Schier, W.A.; Couchell, G.P.

1996-01-01

This is a final reporting on the composition of separate beta and gamma decay heat measurements following neutron fission of 235 U and 238 U and 239 Pu and on cumulative and independent yield measurements of fission products of 235 U and 238 U. What made these studies unique was the very short time of 0.1 s after fission that could be achieved by incorporating the helium jet and tape transport system as the technique for transporting fission fragments from the neutron environment of the fission chamber to the low-background environment of the counting area. This capability allowed for the first time decay heat measurements to extend nearly two decades lower on the logarithmic delay time scale, a region where no comprehensive aggregate decay heat measurements had extended to. This short delay time capability also allowed the measurement of individual fission products with half lives as short as 0.2s. The purpose of such studies was to provide tests both at the aggregate level and at the individual nuclide level of the nation's evaluated nuclear data file associated with fission, ENDF/B-VI. The results of these tests are in general quite encouraging indicating this data base generally predicts correctly the aggregate beta and aggregate gamma decay heat as a function of delay time for 235 U, 238 U and 239 Pu. Agreement with the measured individual nuclide cumulative and independent yields for fission products of 235 U and 238 U was also quite good although the present measurements suggest needed improvements in several individual cases

Estimation of heat transfer and heat source in a molten pool

Energy Technology Data Exchange (ETDEWEB)

Yun, J.I.; Suh, K.Y.; Kang, C.S. [Seoul National Univ., Dept. of Nuclear Engineering (Korea, Republic of)

2001-07-01

Heat transfer and fluid flow in a molten pool are influenced by internal volumetric heat generated from the radioactive decay of fission product species retained in the pool. The pool superheat is determined based on the overall energy balance that equates the heat production rate to the heat loss rate. Decay heat of fission products in the pool was estimated by product of the mass concentration and energy conversion factor of each fission product. For the calculation of heat generation rate in the pool, twenty-nine (29) elements were chosen and classified by their chemical properties. The mass concentration of a fission product is obtained from released fraction and the tabular output of the ORIGEN 2 code. The initial core and pool inventories at each time can also be estimated using ORIGEN 2. The released fraction of each fission product is calculated based on the bubble dynamics and mass transport. Numerical analysis was performed for the TMI-2 accident. The pool is assumed to be a partially filled hemispherical geometry, 1.45 m in radius and 32,700 kg in mass. The change of pool geometry during the numerical calculation was neglected. The peak temperature sizably decreased by about 60 K as the fission products were released from the pool. (author)

Estimation of heat transfer and heat source in a molten pool

International Nuclear Information System (INIS)

Yun, J.I.; Suh, K.Y.; Kang, C.S.

2001-01-01

Heat transfer and fluid flow in a molten pool are influenced by internal volumetric heat generated from the radioactive decay of fission product species retained in the pool. The pool superheat is determined based on the overall energy balance that equates the heat production rate to the heat loss rate. Decay heat of fission products in the pool was estimated by product of the mass concentration and energy conversion factor of each fission product. For the calculation of heat generation rate in the pool, twenty-nine (29) elements were chosen and classified by their chemical properties. The mass concentration of a fission product is obtained from released fraction and the tabular output of the ORIGEN 2 code. The initial core and pool inventories at each time can also be estimated using ORIGEN 2. The released fraction of each fission product is calculated based on the bubble dynamics and mass transport. Numerical analysis was performed for the TMI-2 accident. The pool is assumed to be a partially filled hemispherical geometry, 1.45 m in radius and 32,700 kg in mass. The change of pool geometry during the numerical calculation was neglected. The peak temperature sizably decreased by about 60 K as the fission products were released from the pool. (author)

Progress in fission product nuclear data

International Nuclear Information System (INIS)

Lammer, M.

1982-07-01

This is the eighth issue of a report series on Fission Product Nuclear Data (FPND) which is published by the Nuclear Data Section (NDS) of the International Atomic Energy Agency (IAEA). The purpose of this series is to inform scientists working on FPND, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. The main part of this report consists of unaltered original contributions which the authors have sent to IAEA/NDS. Therefore, the IAEA cannot be held responsible for the information contained nor for any consequences resulting from the use of this information. The present issue contains also a section with some recent references relative to fission product nuclear data, which were not covered by the contributions submitted. The types of activities being included in this report are measurements, compilations and evaluations of: Fission product yields (neutron induced and spontaneous fission); Neutron reaction cross sections of fission products; Data related to the radioactive decay of fission products; Delayed neutron data of fission products; and lumped fission product data (decay heat, absorption etc.). The seventh issue of this series has been published in July 1981 as INDC(NDS)-116. The present issue includes contributions which were received by NDS between 1 August 1981 and 15 June 1982

Fission-product yields for thermal-neutron fission of curium-243

International Nuclear Information System (INIS)

Breederland, D.G.

1982-01-01

Cumulative fission yields for 25 gamma rays emitted during the decay of 23 fission products produced by thermal-neutron fission of 243 Cm have been determined. Using Ge(Li) spectroscopy, 33 successive pulse-height spectra of gamma rays emitted from a 77-ng sample of 243 Cm over a period of approximately two and one-half months were analyzed. Reduction of these spectra resulted in the identification and matching of gamma-ray energies and half-lives to specific radionuclides. Using these results, 23 cumulative fission-product yields were calculated. Only those radionuclides having half-lives between 6 hours and 65 days were observed. Prior to this experiment, no fission-product yields had been recorded for 243 Cm

FPDCYS and FPSPEC: computer programs for calculating fission-product beta and gamma multigroup spectra from ENDF/B-IV data

International Nuclear Information System (INIS)

Stamatelatos, M.G.; England, T.R.

1977-05-01

FPDCYS and FPSPEC are two FORTRAN computer programs used at the Los Alamos Scientific Laboratory (LASL), in conjunction with the CINDER-10 program, for calculating cumulative fission-product beta and/or gamma multigroup spectra in arbitrary energy structures, and for arbitrary neutron irradiation periods and cooling times. FPDCYS processes ENDF/B-IV fission-product decay energy data to generate multigroup beta and gamma spectra from individual ENDF/B-IV fission-product nuclides. FPSPEC further uses these spectra and the corresponding nuclide activities calculated by the CINDER-10 code to produce cumulative beta and gamma spectra in the same energy grids in which FPDCYS generates individual isotope decay spectra. The code system consisting of CINDER-10, FPDCYS, and FPSPEC has been used for comparisons with experimental spectra and continues to be used at LASL for generating spectra in special user-oriented group structures. 3 figures

Study on the calculation method of source term from fission products

International Nuclear Information System (INIS)

Zhou Jing; Gong Quan; Qiu Haifeng

2014-01-01

As a major part of radioactive nuclides, fission products play an important role in nuclear power plant design. The paper analyzes the calculation model of core activity inventory, the model of fission products releasing from the pellets to RCS, the balance model of fission products in RCS, and then proves them by calculation of the typical pressurized water reactor. The model is proved applicable for calculating fission products of pressurized water reactors. (authors)

RSAC-6, Gamma doses, inhalation and ingestion doses, fission products inventory after fission products release

International Nuclear Information System (INIS)

Wenzel, Douglas R.; Schrader, Brad J.

2007-01-01

1 - Description of program or function: RSAC-6 is the latest version of the program RSAC (Radiological Safety Analysis Computer Program). It calculates the consequences of a release of radionuclides to the atmosphere. Using a personal computer, a user can generate a fission product inventory; decay and in-grow the inventory during transport through processes, facilities, and the environment; model the downwind dispersion of the activity; and calculate doses to downwind individuals. Internal dose from the inhalation and ingestion pathways is calculated. External dose from ground surface and plume gamma pathways is calculated. New and exciting updates to the program include the ability to evaluate a release to an enclosed room, resuspension of deposited activity and evaluation of a release up to 1 meter from the release point. Enhanced tools are included for dry deposition, building wake, occupancy factors, respirable fraction, AMAD adjustment, updated and enhanced radionuclide inventory and inclusion of the dose-conversion factors from FOR 11 and 12. 2 - Methods: RSAC6 calculates meteorological dispersion in the atmosphere using Gaussian plume diffusion for Pasquill-Gifford, Hilmeier-Gifford and Markee models. A unique capability is the ability to model Class F fumigation conditions, the meteorological condition that causes the highest ground level concentrations from an elevated release. Doses may be calculated for various pathways including inhalation, ingestion, ground surface, air immersion, water immersion pathways. Dose calculations may be made for either acute or chronic releases. Internal doses (inhalation and ingestion) are calculated using the ICRP-30 model with dose conversion factors from FOR 11. External factors are calculated using FOR 12. 3 - Unusual Features: RSAC6 calculates complete progeny in-growth and decay during all accident phases. The calculation of fission product inventories in particularly useful in the analysis of accidents where the

RIBD-IRT, Isotope Buildup and Isotope Decay from Fission Source

International Nuclear Information System (INIS)

1990-01-01

1 - Description of problem or function: RIBD-IRT calculates isotopic concentrations resulting from two fission sources with normal down- chain decay by beta emission and isomeric transfers and inter-chain coupling resulting from (n,gamma) reactions. Calculations can be made to follow an irradiation history through an unlimited number of step changes of unrestricted duration and variability including shutdown periods, restarts at different power levels and/or any other level changes. In addition, the program permits to track and modify the concentration of individual elements as they decay with time following reactor shutdown. Tracking individual elements enables one to estimate time-dependent source terms for a hypothetical LOCA based on known or postulated fission product release mechanisms. 2 - Method of solution: RIBD-IRT is a grid processor. It organizes the various members described by the fission product library data into a grid with the various linkages established from chain branching data, yield data, and neutron capture cross sections with their branching ratios. Radioactive decay includes not only the simple member-to-member cascade but also the more complex forms where branching may be partially or completely skip one or two intervening members

LOFC fission product release and circulating activity calculations for gas-cooled reactors

International Nuclear Information System (INIS)

Apperson, C.E. Jr.; Carruthers, L.M.; Lee, C.E.

1977-01-01

The inventories of fission products in a gas-cooled reactor under accident and normal steady state conditions are time and temperature dependent. To obtain a reasonable estimate of these inventories it is necessary to consider fuel failure, a temperature dependent variable, and radioactive decay, a time dependent variable. Using arbitrary radioactive decay chains and published fuel failure models for the High Temperature Gas-Cooled Reactor (HTGR), methods have been developed to evaluate the release of fission products during the Loss of Forced Circulation (LOFC) accident and the circulating and plateout fission product inventories during steady state non-accident operation. The LARC-2 model presented here neglects the time delays in the release from the HTGR due to diffusion of fission products from particles in the fuel rod through the graphite matrix. It also neglects the adsorption and evaporation process of metallics at the fuel rod-graphite and graphite-coolant hole interfaces. Any time delay due to the finite time of transport of fission products by convection through the coolant to the outside of the prestressed concrete reactor vessel (PCRV) is also neglected. This model assumes that all fission products released from fuel particles are immediately deposited outside the PCRV with no time delay

Model for fission-product calculations

International Nuclear Information System (INIS)

Smith, A.B.

1984-01-01

Many fission-product cross sections remain unmeasurable thus considerable reliance must be placed upon calculational interpolation and extrapolation from the few available measured cross sections. The vehicle, particularly for the lighter fission products, is the conventional optical-statistical model. The applied goals generally are: capture cross sections to 7 to 10% accuracies and inelastic-scattering cross sections to 25 to 50%. Comparisons of recent evaluations and experimental results indicate that these goals too often are far from being met, particularly in the area of inelastic scattering, and some of the evaluated fission-product cross sections are simply physically unreasonable. It is difficult to avoid the conclusion that the models employed in many of the evaluations are inappropriate and/or inappropriately used. In order to alleviate the above unfortunate situations, a regional optical-statistical (OM) model was sought with the goal of quantitative prediction of the cross sections of the lighter-mass (Z = 30-51) fission products. The first step toward that goal was the establishment of a reliable experimental data base consisting of energy-averaged neutron total and differential-scattering cross sections. The second step was the deduction of a regional model from the experimental data. It was assumed that a spherical OM is appropriate: a reasonable and practical assumption. The resulting OM then was verified against the measured data base. Finally, the physical character of the regional model is examined

Calculating the mass distribution of heavy nucleus fission product by neutrons

International Nuclear Information System (INIS)

Gudkov, A.N.; Koldobskij, A.B.; Kolobashkin, V.M.; Semenova, E.V.

1981-01-01

The technique of calculating the fission product mass yields by neutrons which are necessary for performing nucleus physical calculations in designing nuclear reactor cores is considered. The technique is based on the approximation of fission product mass distribution over the whole mass range by five Gauss functions. New analytical expressions for determining energy weights of used gaussians are proposed. The results of comparison of experimental data with calculated values for fission product mass obtained for reference processes in the capacity of which the fission reactions are chosen: 233 U, 235 U fission by thermal neutrons, 232 Th, 233 U, 235 U, 238 U by fission spectrum neutrons and 14 MeV neutrons and for 232 Th fission reactions by 11 MeV neutrons and 238 U by 7.7 MeV neutrons. On the basis of the analysis of results obtained the conclusion is drawn on a good agreement of fission product mass yield calculation values obtained using recommended values of mass distribution parameters with experimental data [ru

Application of dynamic pseudo fission products and actinides for accurate burnup calculations

Energy Technology Data Exchange (ETDEWEB)

Hoogenboom, J.E.; Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.; Kloosterman, J.L.

1996-09-01

The introduction of pseudo fission products for accurate fine-group spectrum calculations during burnup is discussed. The calculation of the density of the pseudo nuclides is done before each spectrum calculation from the actual densities and their cross sections of all nuclides to be lumped into a pseudo fission product. As there are also many actinides formed in the fuel during its life cycle, a pseudo actinide with fission cross section is also introduced. From a realistic burnup calculation it is demonstrated that only a few fission products and actinides need to be included explicitly in a spectrum calculation. All other fission products and actinides can be accurately represented in the pseudo nuclides. (author)

A brief description of ENDF/B-IV format data for inventory and decay heating calculations

International Nuclear Information System (INIS)

Tobias, A.

1976-07-01

In recent years there has been considerable effort directed towards establishing an international standard format for computerised nuclear data files. At the recent conference on Fission Product Nuclear Data (Bologna, 1973) it was agreed that the ENDF/B format, with certain modifications, be adopted as the standard format for the exchange of such data. A brief description of the basic ENDF/B-IV format of nuclear data files for inventory and decay heat calculations is presented. Although data exchange and inter-comparison will be simple for all files using this format, the data is not generally in a form which can be used directly by inventory codes. One solution to this problem may be for each code to possess a 'translating' routine for rearranging the data into its own format. (author)

An automated system for selective fission product separations; decays of 113-115Pd

International Nuclear Information System (INIS)

Meikrantz, D.H.; Gehrke, R.J.; McIsaac, L.D.; Baker, J.D.; Greenwood, R.C.

1981-01-01

A microcomputer controlled radiochemical separation system has been developed for the isolation and study of fission products with half-lives of approx. >= 10 s. The system is based upon solvent extraction with three centrifugal contactors coupled in series, which provides both rapid and highly efficient separations with large decontamination factors. This automated system was utilized to study the radioactive decays of 113-115 Pd via solvent extraction of the Pd-dimethylglyoxime complex from 252 Cf fission products. As a result of this effort, γ-rays associated with the decay of approx. equal to 90-s sup(113,113m)Pd, 149-s 114 Pd and 47-s 115 Pd have been identified. The isotopic assignments to each of these Pd radioactivities have been confirmed from observation of the growth and decay curves of their respective Ag daughters. In addition, previously unreported Ag γ-rays have been assigned; one to the decay of 69-s 113 Ag, and two to the decay of 19-s 115 Ag. (orig.)

FITPULS: a code for obtaining analytic fits to aggregate fission-product decay-energy spectra

International Nuclear Information System (INIS)

LaBauve, R.J.; George, D.C.; England, T.R.

1980-03-01

The operation and input to the FITPULS code, recently updated to utilize interactive graphics, are described. The code is designed to retrieve data from a library containing aggregate fine-group spectra (150 energy groups) from fission products, collapse the data to few groups (up to 25), and fit the resulting spectra along the cooling time axis with a linear combination of exponential functions. Also given in this report are useful results for aggregate gamma and beta spectra from the decay of fission products released from 235 U irradiated with a pulse (10 -4 s irradiation time) of thermal neutrons. These fits are given in 22 energy groups that are the first 22 groups of the LASL 25-group decay-energy group structure, and the data are expressed both as MeV per fission second and particles per fission second; these pulse functions are readily folded into finite fission histories. 65 figures, 11 tables

Calculation for fission decay from heavy ion reactions at intermediate energies

International Nuclear Information System (INIS)

Blaich, T.; Begemann-Blaich, M.; Fowler, M.M.; Wilhelmy, J.B.; Britt, H.C.; Fields, D.J.; Hansen, L.F.; Namboodiri, M.N.; Sangster, T.C.; Fraenkel, Z.

1992-01-01

A detailed deexcitation calculation is presented for target residues resulting from intermediate-energy heavy ion reactions. The model involves an intranuclear cascade, subsequent fast nucleon emission, and final decay by statistical evaporation including fission. Results are compared to data from bombardments with Fe and Nb projectiles on targets of Ta, Au, and Th at 100 MeV/nucleon. The majority of observable features are reproduced with this simple approach, making obvious the need for involving new physical phenomena associated with multifragmentation or other collective dissipation mechanisms

ELSA: A simplified code for fission product release calculations

International Nuclear Information System (INIS)

Manenc, H.; Notley, M.J.

1996-01-01

During a light water reactor severe accident, fission products are released from the overheated core as it progressively degrades. A new computer module named ELSA is being developed to calculate fission product release. The authors approach is to model the key phenomena, as opposed to more complete mechanistic approaches. Here they present the main features of the module. Different release mechanisms have been identified and are modeled in ELSA, depending on fission product volatility: diffusion seems to govern the release of the highly volatile species if fuel oxidation is properly accounted for, whereas mass transport governs that of lower volatility fission products and fuel volatilization that of the practically involatile species

Influence of high burnup on the decay heat power of spent fuel at long-term storage

International Nuclear Information System (INIS)

Bergelson, B.; Gerasimov, A.; Tikhomirov, G.

2005-01-01

Development and application of advanced fuel with higher burnup is now in practice of NPP with light water reactors in an increasing number of countries. High burnup allows to decrease significantly consumption of uranium. However, spent fuel of this type contains increased amount of high active actinides and fission products in comparison with spent fuel of common-type burnup. Therefore extended time of storage, improved cooling system of the storage facility will be required along with more strong radiation protection during storage, transportation and processing. Calculated data on decay heat power of spent uranium fuel of light water VVER-1000 type reactor are discussed in the paper. Long-term storage of discharged fuel during 100000 years is considered. Calculations were made for burnups of 40-70 MW d/kg. In the initial 50-year period of storage, power of fission products is much higher than that of actinides. Power of gamma-radiation is mainly due to fission products. During subsequent storage power of fission products quickly decreases, the main contribution to the power is given by actinides rather than by fission products. (author)

Fission product model for lattice calculation of high conversion boiling water reactor

International Nuclear Information System (INIS)

Iijima, S.; Yoshida, T.; Yamamoto, T.

1988-01-01

A high precision fission product model for boiling water reactor (BWR) lattice calculation was developed, which consists of 45 nuclides to be treated explicitly and one nonsaturating pseudo nuclide. This model is applied to a high conversion BWR lattice calculation code. From a study based on a three-energy-group calculation of fission product poisoning due to full fission products and explicitly treated nuclides, the multigroup capture cross sections and the effective fission yields of the pseudo nuclide are determined, which do not depend on fuel types or reactor operating conditions for a good approximation. Apart from nuclear data uncertainties, the model and the derived pseudo nuclide constants would predict the fission product reactivity within an error of 0.1% Δk at high burnup

Methodology used to calculate moderator-system heat load at full power and during reactor transients in CANDU reactors

International Nuclear Information System (INIS)

Aydogdu, K.

1998-01-01

Nine components determine the moderator-system heat load during full-power operation and during a reactor power transient in a CANDU reactor. The components that contribute to the total moderator-system heat load at any time consist of the heat generated in the calandria tubes, guide tubes and reactivity mechanisms, moderator and reflector; the heat transferred from calandria shell, the inner tubesheets and the fuel channels; and the heat gained from moderator pumps and heat lost from piping. The contributions from each of these components will vary with time during a reactor transient. The sources of heat that arise from the deposition of nuclear energy can be divided into two categories, viz., a) the neutronic component (which is directly proportional to neutronic power), which includes neutron energy absorption, prompt-fission gamma absorption and capture gamma absorption; and b) the fission-product decay-gamma component, which also varies with time after initiation of the transient. An equation was derived to calculate transient heat loads to the moderator. The equation includes two independent variables that are the neutronic power and fission-product decay-gamma power fractions during the transient and a constant term that represents the heat gained from moderator pumps and heat lost from piping. The calculated heat load in the moderator during steady-state full-power operation for a CANDU 6 reactor was compared with available measurements from the Point Lepreau, Wolsong 1 and Gentilly-2 nuclear generating stations. The calculated and measured values were in reasonably good agreement. (author)

Fission-product releases from a PHWR terminal debris bed

Energy Technology Data Exchange (ETDEWEB)

Brown, M.J.; Bailey, D.G., E-mail: morgan.brown@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

2016-06-15

During an unmitigated severe accident in a pressurized heavy water reactor (PHWR) with horizontal fuel channels, the core may disassemble and relocate to the bottom of the calandria vessel. The resulting heterogeneous in-vessel terminal debris bed (TDB) would likely be quenched by any remaining moderator, and some of the decay heat would be conducted through the calandria vessel shell to the surrounding reactor vault or shield tank water. As the moderator boiled off, the solid debris bed would transform into a more homogeneous molten corium pool located between top and bottom crusts. Until recently, the severe accident code MAAP-CANDU assumed that unreleased volatile and semi-volatile fission products remained in the TDB until after calandria vessel failure, due to low diffusivity through the top crust and the lack of gases or steam to flush released fission products from the debris. However, national and international experimental results indicate this assumption is unlikely; instead, high- and medium-volatility fission products would be released from a molten debris pool, and their volatility and transport should be taken into account in TDB modelling. The resulting change in the distribution of fission products within the reactor and containment, and the associated decay heat, can have significant effects upon the progression of the accident and fission-product releases to the environment. This article describes a postulated PHWR severe accident progression to generate a TDB and the effects of fission-product releases from the terminal debris, using the simple release model in the MAAP-CANDU severe accident code. It also provides insights from various experimental programs related to fission-product releases from core debris, and their applicability to the MAAP-CANDU TDB model. (author)

Decay heat experiment and validation of calculation code systems for fusion reactor

Energy Technology Data Exchange (ETDEWEB)

Maekawa, Fujio; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Wada, Masayuki

1999-10-01

Although accurate estimation of decay heat value is essential for safety analyses of fusion reactors against loss of coolant accidents and so on, no experimental work has been devoted to validating the estimation. Hence, a decay heat measurement experiment was performed as a task (T-339) of ITER/EDA. A new detector, the Whole Energy Absorption Spectrometer (WEAS), was developed for accurate and efficient measurements of decay heat. Decay heat produced in the thirty-two sample materials which were irradiated by 14-MeV neutrons at FNS/JAERI were measured with WEAS for a wide cooling time period from 1 min to 400 days. The data presently obtained were the first experimental decay heat data in the field of fusion. Validity of decay heat calculation codes of ACT4 and CINAC-V4, activation cross section libraries of FENDL/A-2.0 and JENDL Activation File, and decay data was investigated through analyses of the experiment. As a result, several points that should be modified were found in the codes and data. After solving the problems, it was demonstrated that decay heat valued calculated for most of samples were in good agreement with the experimental data. Especially for stainless steel 316 and copper, which were important materials for ITER, decay heat could be predicted with accuracy of {+-}10%. (author)

Decay heat experiment and validation of calculation code systems for fusion reactor

International Nuclear Information System (INIS)

Maekawa, Fujio; Ikeda, Yujiro; Wada, Masayuki

1999-10-01

Although accurate estimation of decay heat value is essential for safety analyses of fusion reactors against loss of coolant accidents and so on, no experimental work has been devoted to validating the estimation. Hence, a decay heat measurement experiment was performed as a task (T-339) of ITER/EDA. A new detector, the Whole Energy Absorption Spectrometer (WEAS), was developed for accurate and efficient measurements of decay heat. Decay heat produced in the thirty-two sample materials which were irradiated by 14-MeV neutrons at FNS/JAERI were measured with WEAS for a wide cooling time period from 1 min to 400 days. The data presently obtained were the first experimental decay heat data in the field of fusion. Validity of decay heat calculation codes of ACT4 and CINAC-V4, activation cross section libraries of FENDL/A-2.0 and JENDL Activation File, and decay data was investigated through analyses of the experiment. As a result, several points that should be modified were found in the codes and data. After solving the problems, it was demonstrated that decay heat valued calculated for most of samples were in good agreement with the experimental data. Especially for stainless steel 316 and copper, which were important materials for ITER, decay heat could be predicted with accuracy of ±10%. (author)

Development and application of the PBMR fission product release calculation model

International Nuclear Information System (INIS)

Merwe, J.J. van der; Clifford, I.

2008-01-01

At PBMR, long-lived fission product release from spherical fuel spheres is calculated using the German legacy software product GETTER. GETTER is a good tool when performing calculations for fuel spheres under controlled operating conditions, including irradiation tests and post-irradiation heat-up experiments. It has proved itself as a versatile reactor analysis tool, but is rather cumbersome when used for accident and sensitivity analysis. Developments in depressurized loss of forced cooling (DLOFC) accident analysis using GETTER led to the creation of FIssion Product RElease under accident (X) conditions (FIPREX), and later FIPREX-GETTER. FIPREX-GETTER is designed as a wrapper around GETTER so that calculations can be carried out for large numbers of fuel spheres with design and operating parameters that can be stochastically varied. This allows full Monte Carlo sensitivity analyses to be performed for representative cores containing many fuel spheres. The development process and application of FIPREX-GETTER in reactor analysis at PBMR is explained and the requirements for future developments of the code are discussed. Results are presented for a sample PBMR core design under normal operating conditions as well as a suite of design-base accident events, illustrating the functionality of FIPREX-GETTER. Monte Carlo sensitivity analysis principles are explained and presented for each calculation type. The plan and current status of verification and validation (V and V) is described. This is an important and necessary process for all software and calculation model development at PBMR

Energy distribution of antineutrinos originating from the decay of fission products in a nuclear reactor

International Nuclear Information System (INIS)

Rudstam, G.; Aleklett, K.

1979-01-01

The energy spectrum of antineutrinos around a nuclear reactor has been derived by summing contributions from individual fission products. The resulting spectrum is weaker at energies above approx. 8 MeV than earlier published antineutrino spectra. The reason may be connected to the strong feeding of high-lying daughter states in the beta decay of fission products with high disintegration energies

User's manual for computer code RIBD-II, a fission product inventory code

International Nuclear Information System (INIS)

Marr, D.R.

1975-01-01

The computer code RIBD-II is used to calculate inventories, activities, decay powers, and energy releases for the fission products generated in a fuel irradiation. Changes from the earlier RIBD code are: the expansion to include up to 850 fission product isotopes, input in the user-oriented NAMELIST format, and run-time choice of fuels from an extensively enlarged library of nuclear data. The library that is included in the code package contains yield data for 818 fission product isotopes for each of fourteen different fissionable isotopes, together with fission product transmutation cross sections for fast and thermal systems. Calculational algorithms are little changed from those in RIBD. (U.S.)

Microscopic beta and gamma data for decay-heat needs

International Nuclear Information System (INIS)

Dickens, J.K.

1983-01-01

Microscopic beta and gamma data for decay-heat needs are defined as absolute-intensity spectral distributions of beta and gamma rays following radioactive decay of radionuclides created by, or following, the fission process. Four well-known evaluated data files, namely the US ENDF/B-V, the UK UKFPDD-2, the French BDN (for fission products), and the Japanese JNDC Nuclear Data Library, are reviewed. Comments regarding the analyses of experimental data (particularly gamma-ray data) are given; the need for complete beta-ray spectral measurements is emphasized. Suggestions on goals for near-term future experimental measurements are presented. 34 references

Measurement of the fission yields of selected prompt and decay fission product gamma-rays of spontaneously fissioning 252Cf and 244Cm

International Nuclear Information System (INIS)

Reber, E.L.; Gehrke, R.J.; Aryaeinejad, R.; Hartwell, J.K.

2005-01-01

Gamma-ray spectrometry measurements have been made of the fission yields of selected γ -rays emitted by the spontaneously fissioning isotopes 252 Cf and 244 Cm. The measured γ-rays were selected based on their relative abundance in the spectrum and their freedom from interference or, in a few instances, ease of interference correction. From these data and the cumulative and independent yield data of England and Rider, those γ-rays that are primarily produced by radioactive decay, as opposed to direct yield, were converted into the decays per spontaneous fission expressed in percent and compared to cumulative yield values of England and Rider. For those γ-rays whose production is dominated by direct (independent) yield, the ratio of γ-rays per spontaneous fission is reported. The γ-ray yield can be compared to the independent yield values of England and Rider when 100% of the direct feeding passes through the γ-ray. In those cases where both cumulative and independent yields contribute to the observed γ-ray emission rate, a direct comparison is not possible but a method to quantify the contribution from each is proposed. (author)

Jeff-3 and decay heat calculations

International Nuclear Information System (INIS)

Huynh, T.D.

2009-07-01

The decay heat power, i.e. the residual heat generated by irradiated nuclear fuels, is a significant parameter to define the power of a reactor. A good evaluation of this power depends both on the accuracy of the processing algorithm and on the quality of the physical data used. This report describes the steps carried out, ranging from tests of consistency to the validation by calculations - experiments comparisons, allowing to choose the validated nuclear data. We have compared the Jeff-3 evaluation (only the file 8 containing decay data) with the Jeff-2.2 and Endf/B7.O evaluations through the computation of residual power. It appears that the residual powers computed by the DARWIN code from Jeff-3.1.1 data for short times agree more with experimental data. There is a slight discrepancy (∼ 2%) between Jeff-3.1 and Jeff-3.1.1 on the total residual power computed for PWR UO 2 fuel. For long decay times the discrepancy is more significant between Jeff-3.1.1 and Jeff-2 on the computation of detailed residual powers because some prevailing isotopes have more formation channels taken into account in Jeff-3 and Jeff-3.1.1 than in Jeff-2

Current status of decay heat measurements, evaluations, and needs

International Nuclear Information System (INIS)

Dickens, J.K.

1986-01-01

Over a decade ago serious concern over possible consequences of a loss-of-coolant accident in a commercial light-water reactor prompted support of several experiments designed specifically to measure the latent energy of beta-ray and gamma-ray emanations from fission products for thermal reactors. This latent energy was termed Decay Heat. At about the same time the American Nuclear Society convened a working group to develop a standard for use in computing decay heat in real reactor environs primarily for regulatory requirements. This working group combined the new experimental results and best evaluated data into a standard which was approved by the ANS and by the ANSI. The primary work since then has been (a) on improvements to computational efforts and (b) experimental measurements for fast reactors. In addition, the need for decay-heat data has been extended well beyond the time regime of a loss-of-coolant accident; new concerns involve, for example, away-from-reactor shipments and storage. The efficacy of the ANS standard for these longer time regimes has been a subject of study with generally positive results. However, a specific problem, namely, the consequences of fission-product neutron capture, remains contentious. Satisfactory resolution of this problem merits a high priority. 31 refs

RSAC, Gamma Doses, Inhalation and Ingestion Doses, Fission Products Inventory after Fission Products Release

International Nuclear Information System (INIS)

Richardson, L.C.

1967-01-01

1 - Description of problem or function: RSAC generates a fission product inventory from a given set of reactor operating conditions and then computes the external gamma dose, the deposition gamma dose, and the inhalation-ingestion dose to critical body organs as a result of exposure to these fission products. Program output includes reactor operating history, fission product inventory, dosages, and ingestion parameters. 2 - Method of solution: The fission product inventory generated by the reactor operating conditions and the inventory remaining at various times after release are computed using the equations of W. Rubinson in Journal of Chemical Physics, Vol. 17, pages 542-547, June 1949. The external gamma dose and the deposition gamma dose are calculated by determining disintegration rates as a function of space and time, then integrating using Hermite's numerical techniques for the spatial dependence. The inhalation-ingestion dose is determined by the type and quantity of activity inhaled and the biological rate of decay following inhalation. These quantities are integrated with respect to time to obtain the dosage. The ingestion dose is related to the inhalation dose by an input constant

A Covariance Generation Methodology for Fission Product Yields

Directory of Open Access Journals (Sweden)

Terranova N.

2016-01-01

Full Text Available Recent safety and economical concerns for modern nuclear reactor applications have fed an outstanding interest in basic nuclear data evaluation improvement and completion. It has been immediately clear that the accuracy of our predictive simulation models was strongly affected by our knowledge on input data. Therefore strong efforts have been made to improve nuclear data and to generate complete and reliable uncertainty information able to yield proper uncertainty propagation on integral reactor parameters. Since in modern nuclear data banks (such as JEFF-3.1.1 and ENDF/BVII.1 no correlations for fission yields are given, in the present work we propose a covariance generation methodology for fission product yields. The main goal is to reproduce the existing European library and to add covariance information to allow proper uncertainty propagation in depletion and decay heat calculations. To do so, we adopted the Generalized Least Square Method (GLSM implemented in CONRAD (COde for Nuclear Reaction Analysis and Data assimilation, developed at CEA-Cadarache. Theoretical values employed in the Bayesian parameter adjustment are delivered thanks to a convolution of different models, representing several quantities in fission yield calculations: the Brosa fission modes for pre-neutron mass distribution, a simplified Gaussian model for prompt neutron emission probability, theWahl systematics for charge distribution and the Madland-England model for the isomeric ratio. Some results will be presented for the thermal fission of U-235, Pu-239 and Pu-241.

Decay heat power of spent nuclear fuel of power reactors with high burnup at long-term storage

Directory of Open Access Journals (Sweden)

Ternovykh Mikhail

2017-01-01

Full Text Available Decay heat power of actinides and fission products from spent nuclear fuel of power VVER-1000 type reactors at long-term storage is calculated. Two modes of storage are considered: mode in which single portion of actinides or fission products is loaded in storage facility, and mode in which actinides or fission products from spent fuel of one VVER reactor are added every year in storage facility during 30 years and then accumulated nuclides are stored without addition new nuclides. Two values of fuel burnup 40 and 70 MW·d/kg are considered for the mode of storage of single fuel unloading. For the mode of accumulation of spent fuel with subsequent storage, one value of burnup of 70 MW·d/kg is considered. Very long time of storage 105 years accepted in calculations allows to simulate final geological disposal of radioactive wastes. Heat power of fission products decreases quickly after 50-100 years of storage. The power of actinides decreases very slow. In passing from 40 to 70 MW·d/kg, power of actinides increases due to accumulation of higher fraction of 244Cm. These data are important in the back end of fuel cycle when improved cooling system of the storage facility will be required along with stronger radiation protection during storage, transportation and processing.

Decay heat power of spent nuclear fuel of power reactors with high burnup at long-term storage

Science.gov (United States)

Ternovykh, Mikhail; Tikhomirov, Georgy; Saldikov, Ivan; Gerasimov, Alexander

2017-09-01

Decay heat power of actinides and fission products from spent nuclear fuel of power VVER-1000 type reactors at long-term storage is calculated. Two modes of storage are considered: mode in which single portion of actinides or fission products is loaded in storage facility, and mode in which actinides or fission products from spent fuel of one VVER reactor are added every year in storage facility during 30 years and then accumulated nuclides are stored without addition new nuclides. Two values of fuel burnup 40 and 70 MW·d/kg are considered for the mode of storage of single fuel unloading. For the mode of accumulation of spent fuel with subsequent storage, one value of burnup of 70 MW·d/kg is considered. Very long time of storage 105 years accepted in calculations allows to simulate final geological disposal of radioactive wastes. Heat power of fission products decreases quickly after 50-100 years of storage. The power of actinides decreases very slow. In passing from 40 to 70 MW·d/kg, power of actinides increases due to accumulation of higher fraction of 244Cm. These data are important in the back end of fuel cycle when improved cooling system of the storage facility will be required along with stronger radiation protection during storage, transportation and processing.

Fission product inventory calculation by a CASMO/ORIGEN coupling program

Energy Technology Data Exchange (ETDEWEB)

Kim, Do Heon; Kim, Jong Kyung [Hanyang University, Seoul (Korea, Republic of); Choi, Hang Bok; Roh, Gyu Hong; Jung, In Ha [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

A CASMO/ORIGEN coupling utility program was developed to predict the composition of all the fission products in spent PWR fuels. The coupling program reads the CASMO output file, modifies the ORIGEN cross section library and reconstructs the ORIGEN input file at each depletion step. In ORIGEN, the burnup equation is solved for actinides and fission products based on the fission reaction rates and depletion flux of CASMO. A sample calculation has been performed using a 14 x 14 PWR fuel assembly and the results are given in this paper. 3 refs., 1 fig., 1 tab. (Author)

Fission product inventory calculation by a CASMO/ORIGEN coupling program

Energy Technology Data Exchange (ETDEWEB)

Kim, Do Heon; Kim, Jong Kyung [Hanyang University, Seoul (Korea, Republic of); Choi, Hang Bok; Roh, Gyu Hong; Jung, In Ha [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-12-31

A CASMO/ORIGEN coupling utility program was developed to predict the composition of all the fission products in spent PWR fuels. The coupling program reads the CASMO output file, modifies the ORIGEN cross section library and reconstructs the ORIGEN input file at each depletion step. In ORIGEN, the burnup equation is solved for actinides and fission products based on the fission reaction rates and depletion flux of CASMO. A sample calculation has been performed using a 14 x 14 PWR fuel assembly and the results are given in this paper. 3 refs., 1 fig., 1 tab. (Author)

Nuclear decay data measurements at the INEL ISOL facility

International Nuclear Information System (INIS)

Greenwood, R.C.; Helmer, R.G.; Putnam, M.H.; Struttmann, D.A.; Watts, K.D.

1991-01-01

In recent years, the use of the mass separation technique coupled on-line to a source of fission product nuclides has provided a wealth of new information on the nuclear decay properties of such nuclides. In addition to their relevance in basic studies of nuclear properties of neutron-rich nuclei, the fission product nuclides as a group, because of their intimate link with energy production in fission reactors, occupy a unique position in the field of applied nuclear decay data. Further, in addition to their critical role in nuclear reactor technology (decay heat source term, environmental concerns, etc.), such data have important applications in astrophysical calculations involving the rapid neutron capture process (r-process) of elemental synthesis in stellar environments. The scope of the nuclear decay data measurements being undertaken using the Idaho National Engineering Laboratory's (INEL) isotope separation on-line (ISOL) facility is focused on a systematic study of the gross nuclear decay properties of short-lived fission product isotopes, i.e., ground-state half-lives, beta-decay energies and beta-decay feeding (or beta-strength) distributions. In this paper, the authors discuss the results of new measurements of beta-decay energies and feeding distributions

Realistic fission models, new beta-decay half-lives and the r-process in neutron star mergers

International Nuclear Information System (INIS)

Shibagaki, S.; Kajino, T.; Chiba, S.; Lorusso, G.; Nishimura, S.; Mathews, G. J.

2014-01-01

Almost half of heavy nuclei beyond iron are considered to be produced by rapid neutron capture process (r-process). This process occurs in the neutron-rich environment such as core-collapse supernovae or neutron star mergers, but the main production site is still unknown. In the r-process of neutron star mergers, nuclear fission reactions play an important role. Also beta-decay half-lives of magic nuclei are crucial for the r-process. We have carried out r-process nucleosynthesis calculations based upon new theoretical estimates of fission fragment distributions and new beta-decay half-lives for N=82 nuclei measured at RIBF-RIKEN. We investigate the effect of nuclear fission on abundance patterns in the matter ejected from neutron star mergers with two different fission fragment mass distributions. We also discuss how the new experimental beta-decay half-lives affect the r-process

Fission product release in conditions of a spent fuel pool severe accident

International Nuclear Information System (INIS)

Ohai, Dumitru

2007-01-01

Full text: Depending on the residence time, fuel burnup, and fuel rack configuration, there may be sufficient decay heat for the fuel clad to heat up, swell, and burst in case of a loss of pool water. Initiating event categories can be: loss of offsite power from events initiated by severe weather, internal fire, loss of pool cooling, loss of coolant inventory, seismic event, aircraft impact, tornado, missile attack. The breach in the clad releases the radioactive gases present in the gap between the fuel and clad, what is called 'gap release'. If the fuel continues to heat up, the zirconium clad will reach the point of rapid oxidation in air. This reaction of zirconium and air, or zirconium and steam is exothermic. The energy released from the reaction, combined with the fuel's decay energy, can cause the reaction to become self-sustaining and ignite the zirconium. The increase in heat from the oxidation reaction can also raise the temperature in adjacent fuel assemblies and propagate the oxidation reaction. Simultaneously, the sintered UO 2 pellets resulting from pins destroying are oxidized. Due to the self-disintegration of pellets by oxidation, fission gases and low volatile fission products are released. The release rate, the chemical nature and the amount of fission products depend on powder granulation distribution and environmental conditions. The zirconium burning and pellets self-disintegration will result in a significant release of spent fuel fission products that will be dispersed from the reactor site. (author)

WAD, a program to calculate the heat produced by alpha decay

International Nuclear Information System (INIS)

Jarvis, R.G.; Bretzlaff, C.I.

1982-09-01

The FORTRAN program WAD (Watts from Alpha Decay) deals with the alpha and beta decay chains to be encountered in advanced fuel cycles for CANDU reactors. The data library covers all necessary alpha-emitting and beta-emitting nuclides and the program calculates the heat produced by alpha decay. Any permissible chain can be constructed very simply

Production of Fission Product 99Mo using High-Enriched Uranium Plates in Polish Nuclear Research Reactor MARIA: Technology and Neutronic Analysis

Directory of Open Access Journals (Sweden)

Jaroszewicz Janusz

2014-07-01

Full Text Available The main objective of 235U irradiation is to obtain the 99mTc isotope, which is widely used in the domain of medical diagnostics. The decisive factor determining its availability, despite its short lifetime, is a reaction of radioactive decay of 99Mo into 99mTc. One of the possible sources of molybdenum can be achieved in course of the 235U fission reaction. The paper presents activities and the calculation results obtained upon the feasibility study on irradiation of 235U targets for production of 99Mo in the MARIA research reactor. Neutronic calculations and analyses were performed to estimate the fission products activity for uranium plates irradiated in the reactor. Results of dummy targets irradiation as well as irradiation uranium plates have been presented. The new technology obtaining 99Mo is based on irradiation of high-enriched uranium plates in standard reactor fuel channel and calculation of the current fission power generation. Measurements of temperatures and the coolant flow in the molybdenum installation carried out in reactor SAREMA system give online information about the current fission power generated in uranium targets. The corrective factors were taken into account as the heat generation from gamma radiation from neighbouring fuel elements as well as heat exchange between channels and the reactor pool. The factors were determined by calibration measurements conducted with aluminium mock-up of uranium plates. Calculations of fuel channel by means of REBUS code with fine mesh structure and libraries calculated by means of WIMS-ANL code were performed.

Derivation of decay heat benchmarks for U235 and Pu239 by a least squares fit to measured data

International Nuclear Information System (INIS)

Tobias, A.

1989-05-01

A least squares technique used by previous authors has been applied to an extended set of available decay heat measurements for both U235 and Pu239 to yield simultaneous fits to the corresponding beta, gamma and total decay heat. The analysis takes account of both systematic and statistical uncertainties, including correlations, via calculations which use covariance matrices constructed for the measured data. The results of the analysis are given in the form of beta, gamma and total decay heat estimates following fission pulses and a range of irradiation times in both U235 and Pu239. These decay heat estimates are considered to form a consistent set of benchmarks for use in the assessment of summation calculations. (author)

Beta decay of the fission product 125Sb and a new complete evaluation of absolute gamma ray transition intensities

Science.gov (United States)

Rajput, M. U.; Ali, N.; Hussain, S.; Mujahid, S. A.; MacMahon, D.

2012-04-01

The radionuclide 125Sb is a long-lived fission product, which decays to 125Te by negative beta emission with a half-life of 1008 day. The beta decay is followed by the emission of several gamma radiations, ranging from low to medium energy, that can suitably be used for high-resolution detector calibrations, decay heat calculations and in many other applications. In this work, the beta decay of 125Sb has been studied in detail. The complete published experimental data of relative gamma ray intensities in the beta decay of the radionuclide 125Sb has been compiled. The consistency analysis was performed and discrepancies found at several gamma ray energies. Evaluation of the discrepant data was carried out using Normalized Residual and RAJEVAL methods. The decay scheme balance was carried out using beta branching ratios, internal conversion coefficients, populating and depopulating gamma transitions to 125Te levels. The work has resulted in the consistent conversion factor equal to 29.59(13) %, and determined a new evaluated set of the absolute gamma ray emission probabilities. The work has also shown 22.99% of the delayed intensity fraction as outgoing from the 58 d isomeric 144 keV energy level and 77.01% of the prompt intensity fraction reaching to the ground state from the other excited states. The results are discussed and compared with previous evaluations. The present work includes additional experimental data sets which were not included in the previous evaluations. A new set of recommended relative and absolute gamma ray emission probabilities is presented.

Beta decay of the fission product 125Sb and a new complete evaluation of absolute gamma ray transition intensities

International Nuclear Information System (INIS)

Rajput, M.U.; Ali, N.; Hussain, S.; Mujahid, S.A.; MacMahon, D.

2012-01-01

The radionuclide 125 Sb is a long-lived fission product, which decays to 125 Te by negative beta emission with a half-life of 1008 day. The beta decay is followed by the emission of several gamma radiations, ranging from low to medium energy, that can suitably be used for high-resolution detector calibrations, decay heat calculations and in many other applications. In this work, the beta decay of 125 Sb has been studied in detail. The complete published experimental data of relative gamma ray intensities in the beta decay of the radionuclide 125 Sb has been compiled. The consistency analysis was performed and discrepancies found at several gamma ray energies. Evaluation of the discrepant data was carried out using Normalized Residual and RAJEVAL methods. The decay scheme balance was carried out using beta branching ratios, internal conversion coefficients, populating and depopulating gamma transitions to 125 Te levels. The work has resulted in the consistent conversion factor equal to 29.59(13) %, and determined a new evaluated set of the absolute gamma ray emission probabilities. The work has also shown 22.99% of the delayed intensity fraction as outgoing from the 58 d isomeric 144 keV energy level and 77.01% of the prompt intensity fraction reaching to the ground state from the other excited states. The results are discussed and compared with previous evaluations. The present work includes additional experimental data sets which were not included in the previous evaluations. A new set of recommended relative and absolute gamma ray emission probabilities is presented.

Decay heat measurement on fusion reactor materials and validation of calculation code system

Energy Technology Data Exchange (ETDEWEB)

Maekawa, Fujio; Ikeda, Yujiro; Wada, Masayuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-03-01

Decay heat rates for 32 fusion reactor relevant materials irradiated with 14-MeV neutrons were measured for the cooling time period between 1 minute and 400 days. With using the experimental data base, validity of decay heat calculation systems for fusion reactors were investigated. (author)

Analysis of fission product release from HTGR core during transient temperature excursion

International Nuclear Information System (INIS)

Saito, Takao; Yamatoya, Naotoshi; Onuma, Mamoru

1978-01-01

The computer program ''FRANC'' was developed to calculate the release activity of fission products from a high-temperature gas cooled reactor (HTGR) core during transient temperature excursions such as a hypothetical loss of forced circulation combined with design basis depressurization. The program utilizes a segmented cylindrical core spatial model with the associated values of the prior fuel irradiation history and temperature conditions. The fission product transport and decay chain behavior is expressed by a set of differential equations. This set of equations describes the entire core inventory of fission products by means of calculated parameters based on the detailed spatial core conditions. The program determines the time-dependent amounts of fission product nuclides escaping from the core into the coolant. Coded in Continuous System Simulation Language (CSSL) with double precision, FRANC showed appropriate results for both short- and long-lived fission product nuclides. The sample calculation conducted by applying the program to a large HTGR indicated that it would take about one hour for noble gases and volatile nuclides to be released to the coolant, and several hours for metalic nuclides. (auth.)

JEFF-3T. Decay data and fission yield libraries

International Nuclear Information System (INIS)

Bersillon, O.; Blachot, J.; Dean, C.J.; Mills, R.W.; Nichols, A.L.; Nouri, A.

2002-01-01

Comprehensive decay-data and fission-yield libraries provide important input to a wide range of nuclear physics codes for nuclear applications. A new initiative has begun under the auspices of the NEA/OECD to generate improved data sets that will constitute the JEFF-3 libraries in ENDF-6 format, primarily for nuclear power, fuel reprocessing and waste management needs. Various sources of decay data have been accessed in order to assemble these files: NUBASE, ENSDF, UKPADD-6 and UKHEDD-2. Efforts have also focused on the evaluation of decay data for a number of important short-lived fission products, so that artificial adjustments to some of the relevant decay data and fission yields are not required to accommodate a previous lack of such data. Fission yields were adopted from UK evaluations recently undertaken to create the UKFY3 library. Decay-data files for 3 755 nuclides have been prepared, including sets of data for the stable nuclides (i.e. mass, natural abundance, spin and parity). Problems in the assignment of ENDF material numbers were addressed, while format and consistency tests were made using CHECKR and FIZCON, respectively. The assembly processes are discussed and reviewed, and the contents of the JEFF-3T starter libraries are described. (author)

Estimation of delayed neutron emission probability by using the gross theory of nuclear β-decay

International Nuclear Information System (INIS)

Tachibana, Takahiro

1999-01-01

The delayed neutron emission probabilities (P n -values) of fission products are necessary in the study of reactor physics; e.g. in the calculation of total delayed neutron yields and in the summation calculation of decay heat. In this report, the P n -values estimated by the gross theory for some fission products are compared with experiment, and it is found that, on the average, the semi-gross theory somewhat underestimates the experimental P n -values. A modification of the β-decay strength function is briefly discussed to get more reasonable P n -values. (author)

Application of a Bayesian/generalised least-squares method to generate correlations between independent neutron fission yield data

International Nuclear Information System (INIS)

Fiorito, L.; Diez, C.; Cabellos, O.; Stankovskiy, A.; Van den Eynde, G.; Labeau, P.E.

2014-01-01

Fission product yields are fundamental parameters for several nuclear engineering calculations and in particular for burn-up/activation problems. The impact of their uncertainties was widely studied in the past and evaluations were released, although still incomplete. Recently, the nuclear community expressed the need for full fission yield covariance matrices to produce inventory calculation results that take into account the complete uncertainty data. In this work, we studied and applied a Bayesian/generalised least-squares method for covariance generation, and compared the generated uncertainties to the original data stored in the JEFF-3.1.2 library. Then, we focused on the effect of fission yield covariance information on fission pulse decay heat results for thermal fission of 235 U. Calculations were carried out using different codes (ACAB and ALEPH-2) after introducing the new covariance values. Results were compared with those obtained with the uncertainty data currently provided by the library. The uncertainty quantification was performed with the Monte Carlo sampling technique. Indeed, correlations between fission yields strongly affect the statistics of decay heat. (authors)

SOURCES-3A: A code for calculating (α, n), spontaneous fission, and delayed neutron sources and spectra

International Nuclear Information System (INIS)

Perry, R.T.; Wilson, W.B.; Charlton, W.S.

1998-04-01

In many systems, it is imperative to have accurate knowledge of all significant sources of neutrons due to the decay of radionuclides. These sources can include neutrons resulting from the spontaneous fission of actinides, the interaction of actinide decay α-particles in (α,n) reactions with low- or medium-Z nuclides, and/or delayed neutrons from the fission products of actinides. Numerous systems exist in which these neutron sources could be important. These include, but are not limited to, clean and spent nuclear fuel (UO 2 , ThO 2 , MOX, etc.), enrichment plant operations (UF 6 , PuF 4 , etc.), waste tank studies, waste products in borosilicate glass or glass-ceramic mixtures, and weapons-grade plutonium in storage containers. SOURCES-3A is a computer code that determines neutron production rates and spectra from (α,n) reactions, spontaneous fission, and delayed neutron emission due to the decay of radionuclides in homogeneous media (i.e., a mixture of α-emitting source material and low-Z target material) and in interface problems (i.e., a slab of α-emitting source material in contact with a slab of low-Z target material). The code is also capable of calculating the neutron production rates due to (α,n) reactions induced by a monoenergetic beam of α-particles incident on a slab of target material. Spontaneous fission spectra are calculated with evaluated half-life, spontaneous fission branching, and Watt spectrum parameters for 43 actinides. The (α,n) spectra are calculated using an assumed isotropic angular distribution in the center-of-mass system with a library of 89 nuclide decay α-particle spectra, 24 sets of measured and/or evaluated (α,n) cross sections and product nuclide level branching fractions, and functional α-particle stopping cross sections for Z < 106. The delayed neutron spectra are taken from an evaluated library of 105 precursors. The code outputs the magnitude and spectra of the resultant neutron source. It also provides an

Calculated leaching of certain fission products from a cylinder of French glass

International Nuclear Information System (INIS)

Blomqvist, G.

1977-07-01

The probable total leaching of the most important fission products and actinides have been tabulated for a cylinder of French HLW glass with approximately 9 percent fission products. The calculations cover the period between 30 and 10000 years after removal from the reactor. The cylinder is of the type planned for the introduction of the HLW into Swedish crystalline rocks. All the components are supposed to have the same leach rate. The calculations also include the probable thickness of eroded glass layer/year. (author)

Calculation of Beta Decay Half-Lives and Delayed Neutron Branching Ratio of Fission Fragments with Skyrme-QRPA

Directory of Open Access Journals (Sweden)

Minato Futoshi

2016-01-01

Full Text Available Nuclear β-decay and delayed neutron (DN emission is important for the r-process nucleosynthesis after the freeze-out, and stable and safe operation of nuclear reactors. Even though radioactive beam facilities have enabled us to measure β-decay and branching ratio of neutron-rich nuclei apart from the stability line in the nuclear chart, there are still a lot of nuclei which one cannot investigate experimentally. In particular, information on DN is rather scarce than that of T1/2. To predict T1/2 and the branching ratios of DN for next JENDL decay data, we have developed a method which comprises the quasiparticle-random-phase-approximation (QRPA and the Hauser-Feshbach statistical model (HFSM. In this work, we calculate fission fragments with T1/2 ≤ 50 sec. We obtain the rms deviation from experimental half-life of 3:71. Although the result is still worse than GT2 which has been adopted in JENDL decay data, DN spectra are newly calculated. We also discuss further subjects to be done in future for improving the present approach and making next generation of JENDL decay data.

Equilibrium Temperature Profiles within Fission Product Waste Forms

Energy Technology Data Exchange (ETDEWEB)

Kaminski, Michael D. [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-10-01

We studied waste form strategies for advanced fuel cycle schemes. Several options were considered for three waste streams with the following fission products: cesium and strontium, transition metals, and lanthanides. These three waste streams may be combined or disposed separately. The decay of several isotopes will generate heat that must be accommodated by the waste form, and this heat will affect the waste loadings. To help make an informed decision on the best option, we present computational data on the equilibrium temperature of glass waste forms containing a combination of these three streams.

Fission cross-section calculations and the multi-modal fission model

International Nuclear Information System (INIS)

Hambsch, F.J.

2004-01-01

New, self consistent, neutron-induced reaction cross section calculations for 235,238 U, 237 Np have been performed. The statistical model code STATIS was improved to take into account the multimodality of the fission process. The three most dominant fission modes, the two asymmetric standards I (S1) and standard II (S2) modes and the symmetric superlong (SL) mode have been taken into account. De-convoluted fission cross sections for those modes for 235,238 U(n,f) and 237 Np(n,f) based on experimental branching ratios, were calculated for the first time up to the second chance fission threshold. For 235 U(n,f), the calculations being made up to 28 MeV incident neutron energy, higher fission chances have been considered. This implied the need for additional calculations for the neighbouring isotopes. As a side product also mass yield distributions could be calculated at energies hitherto not accessible by experiment. Experimental validation of the predictions is being envisaged

Measurement of Fission Product Yields from Fast-Neutron Fission

Science.gov (United States)

Arnold, C. W.; Bond, E. M.; Bredeweg, T. A.; Fowler, M. M.; Moody, W. A.; Rusev, G.; Vieira, D. J.; Wilhelmy, J. B.; Becker, J. A.; Henderson, R.; Kenneally, J.; Macri, R.; McNabb, D.; Ryan, C.; Sheets, S.; Stoyer, M. A.; Tonchev, A. P.; Bhatia, C.; Bhike, M.; Fallin, B.; Gooden, M. E.; Howell, C. R.; Kelley, J. H.; Tornow, W.

2014-09-01

One of the aims of the Stockpile Stewardship Program is a reduction of the uncertainties on fission data used for analyzing nuclear test data [1,2]. Fission products such as 147Nd are convenient for determining fission yields because of their relatively high yield per fission (about 2%) and long half-life (10.98 days). A scientific program for measuring fission product yields from 235U,238U and 239Pu targets as a function of bombarding neutron energy (0.1 to 15 MeV) is currently underway using monoenergetic neutron beams produced at the 10 MV Tandem Accelerator at TUNL. Dual-fission chambers are used to determine the rate of fission in targets during activation. Activated targets are counted in highly shielded HPGe detectors over a period of several weeks to identify decaying fission products. To date, data have been collected at neutron bombarding energies 4.6, 9.0, 14.5 and 14.8 MeV. Experimental methods and data reduction techniques are discussed, and some preliminary results are presented.

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

International Nuclear Information System (INIS)

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

1999-01-01

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

Yields of fission products produced by thermal-neutron fission of 245Cm

International Nuclear Information System (INIS)

Dickens, J.K.; McConnell, J.W.

1981-01-01

Absolute yields have been determined for 105 gamma rays emitted in the decay of 95 fission products representing 54 mass chains created during thermal-neutron fission of 245 Cm. These results include 17 mass chains for which no prior yield data exist. Using a Ge(Li) detector, spectra were obtained of gamma rays between 30 sec and 0.3 yr after very short irradiations of thermal neutrons on a 1 μg sample of 245 Cm. On the basis of measured gamma-ray yields and known nuclear data, total chain mass yields and relative uncertainties were obtained for 51 masses between 84 and 156. The absolute overall normalization uncertainty is 239 Pu and for 252 Cf(s.f.); the influences of the closed shells Z=50, N=82 are not as marked as for thermal-neutron fission of 239 Pu but much more apparent than for 252 Cf(s.f.). Information on the charge distribution along several isobaric mass chains was obtained by determining fractional yields for 12 fission products. The charge distribution width parameter, based upon data for the heavy masses, A=128 to 140, is independent of mass to within the uncertainties of the measurements. Gamma-ray assignments were made for decay of short-lived fission products for which absolute gamma-ray transition probabilities are either not known or in doubt. Absolute gamma-ray transition probabilities were determined as (51 +- 8)% for the 374-keV gamma ray from decay of 110 Rh, (35 +- 7)% for the 1096-keV gamma ray from decay of 133 Sb, and (21.2 +- 1.2)% for the 255-keV gamma ray from decay of 142 Ba

Filtered thermal neutron captured cross sections measurements and decay heat calculations

International Nuclear Information System (INIS)

Pham Ngoc Son; Vuong Huu Tan

2015-01-01

Recently, a pure thermal neutron beam has been developed for neutron capture measurements based on the horizontal channel No.2 of the research reactor at the Nuclear Research Institute, Dalat. The original reactor neutron spectrum is transmitted through an optimal composition of Bi and Si single crystals for delivering a thermal neutron beam with Cadmium ratio (R ed ) of 420 and neutron flux (Φ th ) of 1.6*10 6 n/cm 2 .s. This thermal neutron beam has been applied for measurements of capture cross sections for nuclide of 51 V, by the activation method relative to the standard reaction 197 Au(n,γ) 198 Au. In addition to the activities of neutron capture cross sections measurements, the study on nuclear decay heat calculations has been also considered to be developed at the Institute. Some results on calculation procedure and decay heat values calculated with update nuclear database for 235 U are introduced in this report. (author)

Correlation of recent fission product release data

International Nuclear Information System (INIS)

Kress, T.S.; Lorenz, R.A.; Nakamura, T.; Osborne, M.F.

1989-01-01

For the calculation of source terms associated with severe accidents, it is necessary to model the release of fission products from fuel as it heats and melts. Perhaps the most definitive model for fission product release is that of the FASTGRASS computer code developed at Argonne National Laboratory. There is persuasive evidence that these processes, as well as additional chemical and gas phase mass transport processes, are important in the release of fission products from fuel. Nevertheless, it has been found convenient to have simplified fission product release correlations that may not be as definitive as models like FASTGRASS but which attempt in some simple way to capture the essence of the mechanisms. One of the most widely used such correlation is called CORSOR-M which is the present fission product/aerosol release model used in the NRC Source Term Code Package. CORSOR has been criticized as having too much uncertainty in the calculated releases and as not accurately reproducing some experimental data. It is currently believed that these discrepancies between CORSOR and the more recent data have resulted because of the better time resolution of the more recent data compared to the data base that went into the CORSOR correlation. This document discusses a simple correlational model for use in connection with NUREG risk uncertainty exercises. 8 refs., 4 figs., 1 tab

Extension of hybrid micro-depletion model for decay heat calculation in the DYN3D code

International Nuclear Information System (INIS)

Bilodid, Yurii; Fridman, Emil; Shwageraus, E.

2017-01-01

This work extends the hybrid micro-depletion methodology, recently implemented in DYN3D, to the decay heat calculation by accounting explicitly for the heat contribution from the decay of each nuclide in the fuel.

Extension of hybrid micro-depletion model for decay heat calculation in the DYN3D code

Energy Technology Data Exchange (ETDEWEB)

Bilodid, Yurii; Fridman, Emil [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety; Kotlyar, D. [Georgia Institute of Technology, Atlanta, GA (United States); Shwageraus, E. [Cambridge Univ. (United Kingdom)

2017-06-01

This work extends the hybrid micro-depletion methodology, recently implemented in DYN3D, to the decay heat calculation by accounting explicitly for the heat contribution from the decay of each nuclide in the fuel.

Fission decay properties of nuclear giant multipole resonances

International Nuclear Information System (INIS)

Dias, H.; Arruda Neto, J.D.T.; Hussein, M.S.; Carlson, B.V.

1986-05-01

The statistical fission decay properties of the giant dipole, quadrupole and monopole resonances in 236 U are investigated with the aid of the Hauser-Feshbach model. It is found, contrary to several recent claims, that the GQR fission decay probability is as large as that of the GDR, at energies higher than the fission barrier. At energies close to the f.b., the GQR fission probability is found to be appreciably larger than that of the GDR. The GMR fission probability follows closely that of the GQR. (Author) [pt

Integral decay-heat measurements and comparisons to ENDF/B--IV and V

International Nuclear Information System (INIS)

England, T.R.; Schenter, R.E.; Schmittroth, F.

Results from recent integral decay-power experiments are presented and compared with summation calculations. The experiments include the decay power following thermal fission of 233 U, 235 U, and 239 Pu. The summation calculations use ENDF/B-IV decay data and yields from Versions IV and V. Limited comparisons of experimental β and γ spectra with summation calculations using ENDF/B-IV are included. Generalized least-squares methods are applied to the recent 235 U and 239 Pu decay-power experiments and summation calculations to arrive at evaluated values and uncertainties. Results for 235 U imply uncertainties less than 2% (1 sigma) for the ''infinite'' exposure case for all cooling times greater than 10 seconds. The uncertainties for 239 Pu are larger. Accurate analytical representations of the decay power are presented for 235 , 238 U, and 239 Pu for use in light-water reactors and as the nominal values in the new ANS 5.1 Draft Standard (1978). Comparisons of the nominal values with ENDF/B-IV and the 1973 ANS Draft Standard in current use are included. Gas content, important to decay-heat experiments, and absorption effects on decay power are reviewed. 37 figures, 8 tables

Experimental validation of decay heat calculation codes and associated nuclear data libraries for fusion energy

International Nuclear Information System (INIS)

Maekawa, Fujio; Wada, Masayuki; Ikeda, Yujiro

2001-01-01

Validity of decay heat calculations for safety designs of fusion reactors was investigated by using decay heat experimental data on thirty-two fusion reactor relevant materials obtained at the 14-MeV neutron source facility of FNS in JAERI. Calculation codes developed in Japan, ACT4 and CINAC version 4, and nuclear data bases such as JENDL/Act-96, FENDL/A-2.0 and Lib90 were used for the calculation. Although several corrections in algorithms for both the calculation codes were needed, it was shown by comparing calculated results with the experimental data that most of activation cross sections and decay data were adequate. In cases of type 316 stainless steel and copper which were important for ITER, prediction accuracy of decay heat within ±10% was confirmed. However, it was pointed out that there were some problems in parts of data such as improper activation cross sections, e,g., the 92 Mo(n, 2n) 91g Mo reaction in FENDL, and lack of activation cross section data, e.g., the 138 Ba(n, 2n) 137m Ba reaction in JENDL. Modifications of cross section data were recommended for 19 reactions in JENDL and FENDL. It was also pointed out that X-ray and conversion electron energies should be included in decay data. (author)

Experimental validation of decay heat calculation codes and associated nuclear data libraries for fusion energy

Energy Technology Data Exchange (ETDEWEB)

Maekawa, Fujio; Wada, Masayuki; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-01-01

Validity of decay heat calculations for safety designs of fusion reactors was investigated by using decay heat experimental data on thirty-two fusion reactor relevant materials obtained at the 14-MeV neutron source facility of FNS in JAERI. Calculation codes developed in Japan, ACT4 and CINAC version 4, and nuclear data bases such as JENDL/Act-96, FENDL/A-2.0 and Lib90 were used for the calculation. Although several corrections in algorithms for both the calculation codes were needed, it was shown by comparing calculated results with the experimental data that most of activation cross sections and decay data were adequate. In cases of type 316 stainless steel and copper which were important for ITER, prediction accuracy of decay heat within {+-}10% was confirmed. However, it was pointed out that there were some problems in parts of data such as improper activation cross sections, e,g., the {sup 92}Mo(n, 2n){sup 91g}Mo reaction in FENDL, and lack of activation cross section data, e.g., the {sup 138}Ba(n, 2n){sup 137m}Ba reaction in JENDL. Modifications of cross section data were recommended for 19 reactions in JENDL and FENDL. It was also pointed out that X-ray and conversion electron energies should be included in decay data. (author)

Implementation of a Thermodynamic Solver within a Computer Program for Calculating Fission-Product Release Fractions

Science.gov (United States)

Barber, Duncan Henry

During some postulated accidents at nuclear power stations, fuel cooling may be impaired. In such cases, the fuel heats up and the subsequent increased fission-gas release from the fuel to the gap may result in fuel sheath failure. After fuel sheath failure, the barrier between the coolant and the fuel pellets is lost or impaired, gases and vapours from the fuel-to-sheath gap and other open voids in the fuel pellets can be vented. Gases and steam from the coolant can enter the broken fuel sheath and interact with the fuel pellet surfaces and the fission-product inclusion on the fuel surface (including material at the surface of the fuel matrix). The chemistry of this interaction is an important mechanism to model in order to assess fission-product releases from fuel. Starting in 1995, the computer program SOURCE 2.0 was developed by the Canadian nuclear industry to model fission-product release from fuel during such accidents. SOURCE 2.0 has employed an early thermochemical model of irradiated uranium dioxide fuel developed at the Royal Military College of Canada. To overcome the limitations of computers of that time, the implementation of the RMC model employed lookup tables to pre-calculated equilibrium conditions. In the intervening years, the RMC model has been improved, the power of computers has increased significantly, and thermodynamic subroutine libraries have become available. This thesis is the result of extensive work based on these three factors. A prototype computer program (referred to as SC11) has been developed that uses a thermodynamic subroutine library to calculate thermodynamic equilibria using Gibbs energy minimization. The Gibbs energy minimization requires the system temperature (T) and pressure (P), and the inventory of chemical elements (n) in the system. In order to calculate the inventory of chemical elements in the fuel, the list of nuclides and nuclear isomers modelled in SC11 had to be expanded from the list used by SOURCE 2.0. A

BIPAL - a data library for computing the burnup of fissionable isotopes and products of their decay

International Nuclear Information System (INIS)

Kralovcova, E.; Hep, J.; Valenta, V.

1978-01-01

The BIPAL databank contains data on 100 heavy metal isotopes starting with 206 Tl and finishing with 253 Es. Four are stable, the others are unstable. The following data are currently stored in the databank: the serial number and name of isotopes, decay modes and, for stable isotopes, the isotopic abundance (%), numbers of P decays and Q captures, numbers of corresponding final products, branching ratios, half-lives and their units, decay constants, thermal neutron captures, and fission cross sections, and other data (mainly alpha, beta and gamma intensities). The description of data and a printout of the BIPAL library are presented. (J.B.)

Impact of the total absorption gamma-ray spectroscopy on FP decay heat calculations

International Nuclear Information System (INIS)

Yoshida, Tadashi; Tachibana, Takahiro; Katakura, Jun-ichi

2004-01-01

We calculated the average β- and γ-ray energies, E β and E γ , for 44 short-lived isotopes of Rb, Sr, Y, Cs, Ba, La, Ce, Pr, Nd, Pm, Sm and Eu from the data by Greenwood et al, who measured the β-feed in the decay of these nuclides using the total absorption γ-ray spectrometer. These E β and E γ were incorporated into the decay files from JENDL, JEF2.2 and ENDF-B/VI, and the decay heats were calculated. The results were compared with the integral measurements by the University of Tokyo, ORNL and Lowell. In the case of JENDL, where the correction for the so-called Pandemonium effect is applied on the basis of the gross theory, the very good agreement is no longer maintained. The γ-ray component is overestimated in the cooling time range from 3 to 300 seconds, suggesting a kind of an over-correction as for the Pandemonium effect. We have to evaluate both the applicability of the TAGS results and the correction method itself in order to generate a more consistent data basis for decay heat summation calculations. (author)

Filtered thermal neutron captured cross-sections measurements and decay heat calculations

International Nuclear Information System (INIS)

Son, Pham Ngoc; Tan, Vuong Huu

2014-01-01

Recently, a pure thermal neutron beam has been developed for neutron capture measurements based on the horizontal channel No.2 of the research reactor at the Nuclear Research Institute, Dalat. The original reactor neutron spectrum is transmitted through an optimal composition of Bi and Si single crystals for delivering a thermal neutron beam with Cadmium ratio (R cd ) of 420 and neutron flux (Φ th ) of 1.6x10 6 n/cm 2 .s. This thermal neutron beam has been applied for measurements of capture cross-sections for nuclide of 51 V, 55 Mn, 180 Hf and 186 W by the activation method relative to the standard reaction 197 Au(n,g) 198 Au. In addition to the activities of neutron capture cross-sections measurements, the study on nuclear decay heat calculations has been also considered to be developed at the Institute. Some results on calculation procedure and decay heat values calculated with update nuclear database for 235 U, 238 U, 239 Pu and 232 Th are introduced in this report. (author)

Production of molten UO2 pools by internal heating: apparatus and preliminary experimental heat transfer results

International Nuclear Information System (INIS)

Chasanov, M.G.; Gunther, W.H.; Baker, L. Jr.

1977-01-01

The capability for removal of heat from a pool of molten fuel under postaccident conditions is an important consideration in liquid-metal fast breeder reactor safety analysis. No experimental data for pool heat transfer from molten UO 2 under conditions simulating internal heat generation by fission product decay have been reported previously in the literature. An apparatus to provide such data was developed and used to investigate heat transfer from pools containing up to 7.5 kg of UO 2 ; the internal heat generation rates and pool depths attained cover most of the ranges of interest for postaccident heat removal analysis. It was also observed in these studies that the presence of simulated fission products corresponding to approximately 150,000 kW-day/kg burnup had no significant effect on the observed heat transfer

Time evolution of the fission-decay width under the influence of dissipation

International Nuclear Information System (INIS)

Jurado, B.; Schmidt, K.H.; Benlliure, J.

2002-12-01

Different analytical approximations to the time-dependent fission-decay width used to extract the influence of dissipation on the fission process are critically examined. Calculations with a new, highly realistic analytical approximation to the exact solution of the Fokker-Planck equation sheds doubts on previous conclusions on the dissipation strength made on the basis of less realistic approximations. (orig.)

Progress in fission product nuclear data. No. 14

International Nuclear Information System (INIS)

Lammer, M.

1994-06-01

This is the 14th issue of a report series on Fission Product Nuclear Data published by the Nuclear Data Section of the IAEA. The types of activities included are measurements, compilations and evaluations of fission product yields, neutron reaction cross sections of fission products, data related to the radioactive decay of fission products, delayed neutron data from neutron induced and spontaneous fission, lumped fission product data. The first part of the report consists of unaltered original contributions which the authors have sent to IAEA/NDS. The second part contains some recent references relative to fission product nuclear data, which were not covered by the contributions submitted, and selected papers from conferences. The third part contains requirements for further measurements

Tables and figures from JNDC Nuclear Data Library of fission products, version 2

International Nuclear Information System (INIS)

Ihara, Hitoshi

1989-11-01

The content of JNDC (Japanese Nuclear Data Committee) FP (Fission Product) Nuclear Data Library version 2 for 1227 fission products is presented in the form of tables and figures. The library is inclusive of evaluated decay data such as decay constant, Q-value, average energies of beta, gamma and internal conversion electron, spin-parity, branching ratio of each decay mode and fission yield. The neutron capture cross-sections are also contained for 166 nuclides. The mass number of the fission product nuclides ranges from A = 66 to A = 172. (author)

Estimation of shutdown heat generation rates in GHARR-1 due to ...

African Journals Online (AJOL)

Fission products decay power and residual fission power generated after shutdown of Ghana Research Reactor-1 (GHARR-1) by reactivity insertion accident were estimated by solution of the decay and residual heat equations. A Matlab program code was developed to simulate the heat generation rates by fission product ...

Transport of fission products in matrix and graphite

International Nuclear Information System (INIS)

Hoinkis, E.

1983-06-01

In the past years new experimental methods were applied to or developed for the investigation of fission product transport in graphitic materials and to characterization of the materials. Models for fission product transport and computer codes for the calculation of core release rates were improved. Many data became available from analysis of concentration profiles in HTR-fuel elements. New work on the effect on diffusion of graphite corrosion, fast neutron flux and fluence, heat treatment, chemical interactions and helium pressure was reported on recently or was in progress in several laboratories. It seemed to be the right time to discuss the status of transport of metallic fission products in general, and in particular the relationship between structural and transport properties. Following a suggestion a Colloquium was organized at the HMI Berlin. Interdisciplinary discussions were stimulated by only inviting a limited number of participants who work in different fields of graphite and fission product transport research. (orig./RW)

Code ACTIVE for calculation of the transmutation, induced activity and decay heat in neutron irradiation

International Nuclear Information System (INIS)

Ioki, Kimihiro; Harada, Yuhei; Asami, Naoto.

1976-03-01

The computer code ACTIVE has been prepared for calculation of the transmutation rate, induced activity and decay heat. Calculations are carried out with activation chain and spatial distribution of neutron energy spectrum. The spatial distribution of secondary gamma-ray source due to the unstable nuclides is also obtainable. Special attension is paid to the short life decays. (auth.)

Fission product releases at severe LWR accident conditions: ORNL/CEA measurements versus calculations

Energy Technology Data Exchange (ETDEWEB)

Andre, B.; Ducros, G.; Leveque, J.P. [CEA Centre d`Etudes de Grenoble, 38 (France). Dept. de Thermohydraulique et de Physique; Osborne, M.F.; Lorenz, R.A. [Oak Ridge National Lab., TN (United States); Maro, D. [CEA Centre d`Etudes de Fontenay-aux-Roses, 92 (France). Dept. de Protection de l`Environnement et des Installations

1995-12-31

Experimental programs in the United States and France have followed similar paths in supplying much of the data needed to analyze severe accidents. Both the HI/VI program, conducted at the Oak Ridge National Laboratory (ORNL) under the sponsorship of the U. S. Nuclear Regulatory Commission (NRC), and the HEVA/VERCORS program, supported by IPSN-Commissariat a l`Energie Atomique (CEA) and carried out at the Centre d`Etudes Nucleaires de Grenoble, have studied fission product release from light water reactor (LWR) fuel samples during test sequences representative of severe accidents. Recognizing that more accurate data, i.e., a better defined source term, could reduce the safety margins included in the rather conservative source terms originating from WASH-1400, the primary objective of these programs has been to improve the data base concerning fission product release and behavior at high temperatures. To facilitate the comparison, a model based on fission product diffusion mechanisms that was developed at ORNL and adapted with CEA experimental data is proposed. This CEA model is compared with the ORNL experimental data in a blind test. The two experimental programs used similar techniques in out-of-pile studies. Highly irradiated fuel samples were heated in radiofrequency induction furnaces to very high temperatures (up to 2700 K at ORNL and 2750 K at CEA) in oxidizing (H{sub 2}O), reducing (H{sub 2}) or mixed (H{sub 2}O+H{sub 2}) environments. The experimental parameters, which were chosen from calculated accident scenarios, did not duplicate specific accidents, but rather emphasized careful control of test conditions to facilitate extrapolation of the results to a wide variety of accident situations. This paper presents a broad and consistent database from ORNL and CEA release results obtained independently since the early 1980`S. A comparison of CORSOR and CORSOR Booth calculations, currently used in safety analysis, and the experimental results is presented and

The new isotope 270110 and its decay products 266Hs and 262Sg

International Nuclear Information System (INIS)

Hofmann, S.; Hessberger, F.P.; Ackermann, D.

2000-11-01

The even-even nucleus 270 110 was synthesized using the reaction 64 Ni + 207 Pb. A total of eight α-decay chains was measured during an irradiation time of seven days. Decay data were obtained for the ground-state and a high spin K isomer. The new nuclei 266 Hs and 262 Sg were identified as daughter products after α decay. Spontaneous fission of 262 Sg terminates the decay chain. The measured data are in agreement with calculations using the macroscopic-microscopic model and with self-consistent HFB calculations with Skyrme-Sly4 interaction. (orig.)

Device for measuring fission product density

International Nuclear Information System (INIS)

Kaneda, Mitsunori.

1980-01-01

Purpose: To determine the fission product density of xenon or the like and enable measurement of real time of fission product density in a reactor by calculating the disintegration and annihilation of the fission product on the basis of neutron detected output. Constitution: The neutron flux in a reactor is detected by a detector, and applied to first and second density calculators. Second fission product density signal of xenon or the like outputted from first device is again inputted to the device to form an annihilation signal due to disintegration to determine the present density of the second fission product of xenon or the like corresponding to the decrease of the neutron due to the poison of xeron or the like. Similarly, second device determines the first fission product density of iodine or the like. (Sekiya, K.)

Decay heat of 235U fission products by beta- and gamma-ray spectrometry

International Nuclear Information System (INIS)

Dickens, J.K.; Love, T.A.; McConnell, J.W.; Peelle, R.W.

1976-09-01

The fast-rabbit facilities of the ORRR were used to irradiate 1- to 10-μg samples of 235 U for 1, 10, and 100 s. Released power is observed using nuclear spectroscopy to permit separate observations of emitted β and γ spectra in successive time intervals. The spectra were integrated over energy to obtain total decay heat and the β- and γ-ray results are summed together. 10 fig, 2 tables

Heat Transfer Characteristics of SiC-coated Heat Pipe for Passive Decay Heat Removal

International Nuclear Information System (INIS)

Kim, Kyung Mo; Kim, In Guk; Jeong, Yeong Shin; Bang, In Cheol

2014-01-01

The main concern with the Fukushima accident was the failure of active and passive core cooling systems. The main function of existing passive decay heat removal systems is feeding additional coolant to the reactor core. Thus, an established emergency core cooling system (ECCS) cannot operate properly because of impossible depressurization under the station blackout (SBO) condition. Therefore, a new concept for passive decay heat removal system is required. In this study, an innovative hybrid control rod concept is considered for passive in-core decay heat removal that differs from the existing direct vessel injection core cooling system and passive auxiliary feedwater system (PAFS). The heat transfer between the evaporator and condenser sections occurs by phase change of the working fluid and capillary action induced by wick structures installed on the inner wall of the heat pipe. In this study, a hybrid control rod is developed to take the roles of both neutron absorption and heat removal by combining the functions of a heat pipe and control rod. Previous studies on enhancing the heat removal capacity of heat pipes used nanofluids, self-rewetting fluids, various wick structures and condensers. Many studies have examined the thermal performances of heat pipes using various nanofluids. They concluded that the enhanced thermal performance of the heat pipe using nanofluids is due to nanoparticle deposition on the wick structures. Thus, the wick structure of heat pipes has been modified by nanoparticle deposition to enhance the heat removal capacity. However, previous studies used relatively small heat pipes and narrow ranges of heat loads. The environment of a nuclear reactor is very specific, and the decay heat produced by fission products after shutdown is relatively large. Thus, this study tested a large-scale heat pipe over a wide range of power. The concept of a hybrid heat pipe for an advanced in-core decay heat removal system was introduced for complete

Heat Transfer Characteristics of SiC-coated Heat Pipe for Passive Decay Heat Removal

Energy Technology Data Exchange (ETDEWEB)

Kim, Kyung Mo; Kim, In Guk; Jeong, Yeong Shin; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

2014-10-15

The main concern with the Fukushima accident was the failure of active and passive core cooling systems. The main function of existing passive decay heat removal systems is feeding additional coolant to the reactor core. Thus, an established emergency core cooling system (ECCS) cannot operate properly because of impossible depressurization under the station blackout (SBO) condition. Therefore, a new concept for passive decay heat removal system is required. In this study, an innovative hybrid control rod concept is considered for passive in-core decay heat removal that differs from the existing direct vessel injection core cooling system and passive auxiliary feedwater system (PAFS). The heat transfer between the evaporator and condenser sections occurs by phase change of the working fluid and capillary action induced by wick structures installed on the inner wall of the heat pipe. In this study, a hybrid control rod is developed to take the roles of both neutron absorption and heat removal by combining the functions of a heat pipe and control rod. Previous studies on enhancing the heat removal capacity of heat pipes used nanofluids, self-rewetting fluids, various wick structures and condensers. Many studies have examined the thermal performances of heat pipes using various nanofluids. They concluded that the enhanced thermal performance of the heat pipe using nanofluids is due to nanoparticle deposition on the wick structures. Thus, the wick structure of heat pipes has been modified by nanoparticle deposition to enhance the heat removal capacity. However, previous studies used relatively small heat pipes and narrow ranges of heat loads. The environment of a nuclear reactor is very specific, and the decay heat produced by fission products after shutdown is relatively large. Thus, this study tested a large-scale heat pipe over a wide range of power. The concept of a hybrid heat pipe for an advanced in-core decay heat removal system was introduced for complete

Fission properties of superheavy nuclei for r -process calculations

Science.gov (United States)

Giuliani, Samuel A.; Martínez-Pinedo, Gabriel; Robledo, Luis M.

2018-03-01

We computed a new set of static fission properties suited for r -process calculations. The potential energy surfaces and collective inertias of 3640 nuclei in the superheavy region are obtained from self-consistent mean-field calculations using the Barcelona-Catania-Paris-Madrid energy density functional. The fission path is computed as a function of the quadrupole moment by minimizing the potential energy and exploring octupole and hexadecapole deformations. The spontaneous fission lifetimes are evaluated employing different schemes for the collective inertias and vibrational energy corrections. This allows us to explore the sensitivity of the lifetimes to those quantities together with the collective ground-state energy along the superheavy landscape. We computed neutron-induced stellar reaction rates relevant for r -process nucleosynthesis using the Hauser-Feshbach statistical approach and study the impact of collective inertias. The competition between different reaction channels including neutron-induced rates, spontaneous fission, and α decay is discussed for typical r -process conditions.

Fission-product energy release for times following thermal-neutron fission of 235U between 2 and 14000 seconds

International Nuclear Information System (INIS)

Dickens, J.K.; Emery, J.F.; Love, T.A.; McConnell, J.W.; Northcutt, K.J.; Peelle, R.W.; Weaver, H.

1977-10-01

Fission-product decay energy-releases rates were measured for thermal-neutron fission of 235 U. Samples of mass 1 to 10 μg were irradiated for 1 to 100 sec by use of the fast pneumatic-tube facility at the Oak Ridge Research Reactor. The resulting beta- and gamma-ray emissions were counted for times-after-fission between 2 and 14,000 seconds. The data were obtained for beta and gamma rays separately as spectral distributions, N(E/sub γ/) vs E/sub γ/ and N(E/sub beta/) vs E/sub β/. For the gamma-ray data the spectra were obtained by using a NaI detector, while for the beta-ray data the spectra were obtained by using an NE-110 detector with an anticoincidence mantle. The raw data were unfolded to provide spectral distributions of modest resolution. These were integrated over E/sub γ/ and E/sub β/ to provide total yield and energy integrals as a function of time after fission. Results are low compared to the present 1973 ANS Decay-heat standard. A complete description of the experimental apparatus and data-reduction techniques is presented. The final integral data are given in tabular and graphical form and are compared with published data. 41 figures, 13 tables

Fission product data for thermal reactors. Final report. Part I. A data set for EPRI-CINDER using ENDF/B-IV

International Nuclear Information System (INIS)

England, T.R.; Wilson, W.B.; Stamatelatos, M.G.

1976-12-01

A four-group fission-product neutron absorption library, appropriate for use in thermal reactors, is described. All decay parameters are taken from ENDF/B-IV. The absorption cross sections are also processed from ENDF/B-IV files, first into a 154-group set and subsequently collapsed into the 4-group set described in this report. The decay and cross section data were used to form 84 linear chains in the CINDER code format. These chains contain all significant fission products having half-lives exceeding 4 hours--a total of 186 nuclides. A 12-chain set containing one pseudo-chain for use in spatial depletion calculations is described. This set accurately reproduces the aggregate absorption buildup of the 84 chains. This report describes the chains and processed data, results of comparison calculations for various fuels, and a comparison of calculated temporal fission-product absorption buildup with corresponding results from a long-term fuel irradiation and cooling integral experiment

A revised calculational model for fission

Energy Technology Data Exchange (ETDEWEB)

Atchison, F

1998-09-01

A semi-empirical parametrization has been developed to calculate the fission contribution to evaporative de-excitation of nuclei with a very wide range of charge, mass and excitation-energy and also the nuclear states of the scission products. The calculational model reproduces measured values (cross-sections, mass distributions, etc.) for a wide range of fissioning systems: Nuclei from Ta to Cf, interactions involving nucleons up to medium energy and light ions. (author)

Half-lives for proton emission, alpha decay, cluster radioactivity, and cold fission processes calculated in a unified theoretical framework

International Nuclear Information System (INIS)

Duarte, S.B.; Tavares, O.A.P.; Guzman, F.; Dimarco, A.; Garcia, F.; Goncalves, M.

2002-01-01

Half-life values of spontaneous nuclear decay processes are presented in the framework of the Effective Liquid Drop Model (ELDM) using the combination of varying mass asymmetry shape description for the mass transfer with Werner-Wheeler's inertia coefficient V MAS /WW. The calculated half-lives of ground-state to ground-state transitions for the proton emission, alpha decay, cluster radioactivity, and cold fission processes are compared with experimental data. Results have shown that the ELDM is a very efficient model to describe these different decay processes in a same, unified theoretical framework. A Table listing the predicted half-life values, τ c is presented for all possible cases of spontaneous nuclear break-up such that -7.30 10 τ c [S] 10 (τ/τ c ) > -17.0, where τ is the total half-life of the parent nucleus. (author)

NEACRP thermal fission product benchmark

International Nuclear Information System (INIS)

Halsall, M.J.; Taubman, C.J.

1989-09-01

The objective of the thermal fission product benchmark was to compare the range of fission product data in use at the present time. A simple homogeneous problem was set with 200 atoms H/1 atom U235, to be burnt up to 1000 days and then decay for 1000 days. The problem was repeated with 200 atoms H/1 atom Pu239, 20 atoms H/1 atom U235 and 20 atoms H/1 atom Pu239. There were ten participants and the submissions received are detailed in this report. (author)

On the safety of conceptual fusion-fission hybrid reactors

International Nuclear Information System (INIS)

Kastenberg, W.E.; Okrent, D.; Badham, V.; Caspi, S.; Chan, C.K.; Ferrell, W.J.; Frederking, T.H.K.; Grzesik, J.; Lee, J.Y.; McKone, T.E.; Pomraning, G.C.; Ullman, A.Z.; Ting, T.D.; Kim, Y.I.

1979-01-01

A preliminary examination of some potential safety questions for conceptual fusion-fission hybrid reactors is presented in this paper. The study and subsequent analysis was largely based upon one design, a conceptual mirror fusion-fission reactor, operating on the deuterium-tritium plasma fusion fuel cycle and the uranium-plutonium fission fuel cycle. The major potential hazards were found to be: (a) fission products, (b) actinide elements, (c) induced radioactivity, and (d) tritium. As a result of these studies, it appears that highly reliable and even redundent decay heat removal must be provided. Loss of the ability to remove decay heat results in melting of fuel, with ultimate release of fission products and actinides to the containment. In addition, the studies indicate that blankets can be designed which will remain subcritical under extensive changes in both composition and geometry. Magnet safety and the effects of magnetic fields on thermal parameters were also considered. (Auth.)

Calculation of vapor pressure of fission product fluorides and oxyfluorides

International Nuclear Information System (INIS)

Roux, J.P.

1976-03-01

The equilibrium diagrams of the condensed phases - solid and liquid - and vapor phase are collected for the principal fluorides and oxyfluorides of fission product elements (atomic number from 30 to 66). These diagrams are used more particularly in fuel reprocessing by fluoride volatility process. Calculations and curves (vapor pressure in function of temperature) are processed using a computer program given in this report [fr

ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data

International Nuclear Information System (INIS)

Palmiotti, G.

2011-01-01

The ENDF/B-VII.1 library is our latest recommended evaluated nuclear data file for use in nuclear science and technology applications, and incorporates advances made in the five years since the release of ENDF/B-VII.0. These advances focus on neutron cross sections, covariances, fission product yields and decay data, and represent work by the US Cross Section Evaluation Working Group (CSEWG) in nuclear data evaluation that utilizes developments in nuclear theory, modeling, simulation, and experiment. The principal advances in the new library are: (1) An increase in the breadth of neutron reaction cross section coverage, extending from 393 nuclides to 418 nuclides; (2) Covariance uncertainty data for 185 of the most important nuclides, as documented in companion papers in this edition; (3) R-matrix analyses of neutron reactions on light nuclei, including isotopes of He, Li, and Be; (4) Resonance parameter analyses at lower energies and statistical high energy reactions at higher energies for isotopes of F, Cl, K, Ti, V, Mn, Cr, Ni, Zr and W; (5) Modifications to thermal neutron reactions on fission products (isotopes of Mo, Tc, Rh, Ag, Cs, Nd, Sm, Eu) and neutron absorber materials (Cd, Gd); (6) Improved minor actinide evaluations for isotopes of U, Np, Pu, and Am (we are not making changes to the major actinides 235,238U and 239Pu at this point, except for delayed neutron data, and instead we intend to update them after a further period of research in experiment and theory), and our adoption of JENDL-4.0 evaluations for isotopes of Cm, Bk, Cf, Es, Fm, and some other minor actinides; (7) Fission energy release evaluations; (8) Fission product yield advances for fission-spectrum neutrons and 14 MeV neutrons incident on 239Pu; and (9) A new Decay Data sublibrary. Integral validation testing of the ENDF/B-VII.1 library is provided for a variety of quantities: For nuclear criticality, the VII.1 library maintains the generally-good performance seen for VII.0 for a wide

Decay heat and activity of the structural materials of the fuel and blanket assemblies of the second and third core of KNK II

International Nuclear Information System (INIS)

Winterhagen, D.

1986-06-01

The decay heat and activity caused by structural materials have been calculated for the fuel assemblies of KNK II (second and third core) with a residence time of 720 equivalent full-power days (efpd) and the blanket assemblies with 1880 efpd. The values are given for the different zones of the assemblies (head, active zone, fission gas plenum, foot and stellite area) for decay times from 1 to 20 years. For decay times beyond 2 years more than 80 % of the decay heat are caused by the Co60-decay, more than 60 % of which result from the stellite in the foot area [de

ENDF/B-5 Fission Products Library 1979

International Nuclear Information System (INIS)

Schwerer, O.; Lemmel, H.D.

1981-10-01

This document summarizes contents and documentation of the 1979 version of the Fission Products File of the ENDF/B Library maintained by the National Nuclear Data Center (NNDC) at the Brookhaven National Laboratory, USA. This file contains numerical neutron reaction data and decay data for 877 fission product nuclides. The entire file or selective retrievals from it can be obtained on magnetic tape from the IAEA Nuclear Data Section. (author)

Fission product yields

International Nuclear Information System (INIS)

Valenta, V.; Hep, J.

1978-01-01

Data are summed up necessary for determining the yields of individual fission products from different fissionable nuclides. Fractional independent yields, cumulative and isobaric yields are presented here for the thermal fission of 235 U, 239 Pu, 241 Pu and for fast fission (approximately 1 MeV) of 235 U, 238 U, 239 Pu, 241 Pu; these values are included into the 5th version of the YIELDS library, supplementing the BIBFP library. A comparison is made of experimental data and possible improvements of calculational methods are suggested. (author)

Fission-product source terms

International Nuclear Information System (INIS)

Lorenz, R.A.

1981-01-01

This presentation consists of a review of fission-product source terms for light water reactor (LWR) fuel. A source term is the quantity of fission products released under specified conditions that can be used to calculate the consequences of the release. The source term usually defines release from breached fuel-rod cladding but could also describe release from the primary coolant system, the reactor containment shell, or the site boundary. The source term would be different for each locality, and the chemical and physical forms of the fission products could also differ

Propagation of nuclear data uncertainties in fuel cycle calculations using Monte-Carlo technique

International Nuclear Information System (INIS)

Diez, C.J.; Cabellos, O.; Martinez, J.S.

2011-01-01

Nowadays, the knowledge of uncertainty propagation in depletion calculations is a critical issue because of the safety and economical performance of fuel cycles. Response magnitudes such as decay heat, radiotoxicity and isotopic inventory and their uncertainties should be known to handle spent fuel in present fuel cycles (e.g. high burnup fuel programme) and furthermore in new fuel cycles designs (e.g. fast breeder reactors and ADS). To deal with this task, there are different error propagation techniques, deterministic (adjoint/forward sensitivity analysis) and stochastic (Monte-Carlo technique) to evaluate the error in response magnitudes due to nuclear data uncertainties. In our previous works, cross-section uncertainties were propagated using a Monte-Carlo technique to calculate the uncertainty of response magnitudes such as decay heat and neutron emission. Also, the propagation of decay data, fission yield and cross-section uncertainties was performed, but only isotopic composition was the response magnitude calculated. Following the previous technique, the nuclear data uncertainties are taken into account and propagated to response magnitudes, decay heat and radiotoxicity. These uncertainties are assessed during cooling time. To evaluate this Monte-Carlo technique, two different applications are performed. First, a fission pulse decay heat calculation is carried out to check the Monte-Carlo technique, using decay data and fission yields uncertainties. Then, the results, experimental data and reference calculation (JEFF Report20), are compared. Second, we assess the impact of basic nuclear data (activation cross-section, decay data and fission yields) uncertainties on relevant fuel cycle parameters (decay heat and radiotoxicity) for a conceptual design of a modular European Facility for Industrial Transmutation (EFIT) fuel cycle. After identifying which time steps have higher uncertainties, an assessment of which uncertainties have more relevance is performed

Update and evaluation of decay data for spent nuclear fuel analyses

Science.gov (United States)

Simeonov, Teodosi; Wemple, Charles

2017-09-01

Studsvik's approach to spent nuclear fuel analyses combines isotopic concentrations and multi-group cross-sections, calculated by the CASMO5 or HELIOS2 lattice transport codes, with core irradiation history data from the SIMULATE5 reactor core simulator and tabulated isotopic decay data. These data sources are used and processed by the code SNF to predict spent nuclear fuel characteristics. Recent advances in the generation procedure for the SNF decay data are presented. The SNF decay data includes basic data, such as decay constants, atomic masses and nuclide transmutation chains; radiation emission spectra for photons from radioactive decay, alpha-n reactions, bremsstrahlung, and spontaneous fission, electrons and alpha particles from radioactive decay, and neutrons from radioactive decay, spontaneous fission, and alpha-n reactions; decay heat production; and electro-atomic interaction data for bremsstrahlung production. These data are compiled from fundamental (ENDF, ENSDF, TENDL) and processed (ESTAR) sources for nearly 3700 nuclides. A rigorous evaluation procedure of internal consistency checks and comparisons to measurements and benchmarks, and code-to-code verifications is performed at the individual isotope level and using integral characteristics on a fuel assembly level (e.g., decay heat, radioactivity, neutron and gamma sources). Significant challenges are presented by the scope and complexity of the data processing, a dearth of relevant detailed measurements, and reliance on theoretical models for some data.

Update and evaluation of decay data for spent nuclear fuel analyses

Directory of Open Access Journals (Sweden)

Simeonov Teodosi

2017-01-01

Full Text Available Studsvik’s approach to spent nuclear fuel analyses combines isotopic concentrations and multi-group cross-sections, calculated by the CASMO5 or HELIOS2 lattice transport codes, with core irradiation history data from the SIMULATE5 reactor core simulator and tabulated isotopic decay data. These data sources are used and processed by the code SNF to predict spent nuclear fuel characteristics. Recent advances in the generation procedure for the SNF decay data are presented. The SNF decay data includes basic data, such as decay constants, atomic masses and nuclide transmutation chains; radiation emission spectra for photons from radioactive decay, alpha-n reactions, bremsstrahlung, and spontaneous fission, electrons and alpha particles from radioactive decay, and neutrons from radioactive decay, spontaneous fission, and alpha-n reactions; decay heat production; and electro-atomic interaction data for bremsstrahlung production. These data are compiled from fundamental (ENDF, ENSDF, TENDL and processed (ESTAR sources for nearly 3700 nuclides. A rigorous evaluation procedure of internal consistency checks and comparisons to measurements and benchmarks, and code-to-code verifications is performed at the individual isotope level and using integral characteristics on a fuel assembly level (e.g., decay heat, radioactivity, neutron and gamma sources. Significant challenges are presented by the scope and complexity of the data processing, a dearth of relevant detailed measurements, and reliance on theoretical models for some data.

Measurements of fission product yield in the neutron-induced fission of 238U with average energies of 9.35 MeV and 12.52 MeV

Science.gov (United States)

Mukerji, Sadhana; Krishnani, Pritam Das; Shivashankar, Byrapura Siddaramaiah; Mulik, Vikas Kaluram; Suryanarayana, Saraswatula Venkat; Naik, Haladhara; Goswami, Ashok

2014-07-01

The yields of various fission products in the neutron-induced fission of 238U with the flux-weightedaveraged neutron energies of 9.35 MeV and 12.52 MeV were determined by using an off-line gammaray spectroscopic technique. The neutrons were generated using the 7Li(p, n) reaction at Bhabha Atomic Research Centre-Tata Institute of Fundamental Research Pelletron facility, Mumbai. The gamma- ray activities of the fission products were counted in a highly-shielded HPGe detector over a period of several weeks to identify the decaying fission products. At both the neutron energies, the fission-yield values are reported for twelve fission product. The results obtained from the present work have been compared with the similar data for mono-energetic neutrons of comparable energy from the literature and are found to be in good agreement. The peak-to-valley (P/V) ratios were calculated from the fission-yield data and were found to decreases for neutron energy from 9.35 to 12.52 MeV, which indicates the role of excitation energy. The effect of the nuclear structure on the fission product-yield is discussed.

Half-lives for proton emission, alpha decay, cluster radioactivity, and cold fission processes calculated in a unified theoretical framework

Energy Technology Data Exchange (ETDEWEB)

Duarte, S.B.; Tavares, O.A.P.; Guzman, F.; Dimarco, A. [Centro Brasileiro de Pesquisas Fisicas (CBPF), Rio de Janeiro, RJ (Brazil); Garcia, F. [Sao Paulo Univ., SP (Brazil). Inst. de Fisica; Universidade Estadual de Santa Cruz, Ilheus, BA (Brazil). Dept. de Ciencias Exatas e Tecnologicas; Rodriguez, O. [Sao Paulo Univ., SP (Brazil). Inst. de Fisica; Instituto Superior de Ciencias e Tecnologia Nucleares, La Habana (Cuba); Goncalves, M. [Instituto de Radioprotecao e Dosimetria (IRD), Rio de Janeiro, RJ (Brazil)

2002-01-01

Half-life values of spontaneous nuclear decay processes are presented in the framework of the Effective Liquid Drop Model (ELDM) using the combination of varying mass asymmetry shape description for the mass transfer with Werner-Wheeler's inertia coefficient V{sub MAS}/WW. The calculated half-lives of ground-state to ground-state transitions for the proton emission, alpha decay, cluster radioactivity, and cold fission processes are compared with experimental data. Results have shown that the ELDM is a very efficient model to describe these different decay processes in a same, unified theoretical framework. A Table listing the predicted half-life values, {tau}{sub c} is presented for all possible cases of spontaneous nuclear break-up such that -7.30 <{approx_equal} log{sub 10} {tau}{sub c} [S] <{approx_equal} 27.50 and log {sub 10}({tau}/{tau}{sub c}) > -17.0, where {tau} is the total half-life of the parent nucleus. (author)

JENDL FP decay data file 2000 and the beta-decay theory

International Nuclear Information System (INIS)

Yoshida, Tadashi; Katakura, Jun Ichi; Tachibana, Takahiro

2002-01-01

JENDL FP Decay Data File 2000 has been developed as one of the special purpose files of the Japanese Evaluated Nuclear Data Library (JENDL), which constitutes a versatile nuclear data basis for science and technology. In the format of ENDF-6 this file includes the decay data for 1087 unstable fission product (FP) nuclides and 142 stable nuclides as their daughters. The primary purpose of this file is to use in the summation calculation of FP decay heat, which plays a critical role in nuclear safety analysis; the loss-of-coolant accident analysis of reactors, for example. The data for a given nuclide are its decay modes, the Q value, the branching ratios, the average energies released in the form of beta- and gamma-rays per decay, and their spectral data. The primary source of the decay data adopted here is the ENSDF (Evaluated Nuclear Structure Data File). The data in ENSDF, however, cover only the measured values. The data of the short-lived nuclides, which are essential for the decay heat calculations at short cooling times, are often fully lacking or incomplete even if they exist. This is mainly because of their short half-life nature. For such nuclides a theoretical model calculation is applied in order to fill the gaps between the true and the experimentally known decay schemes. In practice we have to predict the average decay energies and the spectral data for a lot of short-lived FPs by use of beta-decay theories. Thus the beta-decay theory plays a very important role in generating the FP decay data file

ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data

Energy Technology Data Exchange (ETDEWEB)

Chadwick, M. B. [Los Alamos National Laboratory (LANL); Herman, Micheal W [Brookhaven National Laboratory (BNL); Oblozinsky, Pavel [Brookhaven National Laboratory (BNL); Dunn, Michael E [ORNL; Danon, Y. [Rensselaer Polytechnic Institute (RPI); Kahler, A. [Los Alamos National Laboratory (LANL); Smith, Donald L. [Argonne National Laboratory (ANL); Pritychenko, B [Brookhaven National Laboratory (BNL); Arbanas, Goran [ORNL; Arcilla, r [Brookhaven National Laboratory (BNL); Brewer, R [Los Alamos National Laboratory (LANL); Brown, D A [Brookhaven National Laboratory (BNL); Capote, R. [International Atomic Energy Agency (IAEA); Carlson, A. D. [National Institute of Standards and Technology (NIST); Cho, Y S [Korea Atomic Energy Research Institute; Derrien, Herve [ORNL; Guber, Klaus H [ORNL; Hale, G. M. [Los Alamos National Laboratory (LANL); Hoblit, S [Brookhaven National Laboratory (BNL); Holloway, Shannon T. [Los Alamos National Laboratory (LANL); Johnson, T D [Brookhaven National Laboratory (BNL); Kawano, T. [Los Alamos National Laboratory (LANL); Kiedrowski, B C [Los Alamos National Laboratory (LANL); Kim, H [Korea Atomic Energy Research Institute; Kunieda, S [Los Alamos National Laboratory (LANL); Larson, Nancy M [ORNL; Leal, Luiz C [ORNL; Lestone, J P [Los Alamos National Laboratory (LANL); Little, R C [Los Alamos National Laboratory (LANL); Mccutchan, E A [Brookhaven National Laboratory (BNL); Macfarlane, R E [Los Alamos National Laboratory (LANL); MacInnes, M [Los Alamos National Laboratory (LANL); Matton, C M [Lawrence Livermore National Laboratory (LLNL); Mcknight, R D [Argonne National Laboratory (ANL); Mughabghab, S F [Brookhaven National Laboratory (BNL); Nobre, G P [Brookhaven National Laboratory (BNL); Palmiotti, G [Idaho National Laboratory (INL); Palumbo, A [Brookhaven National Laboratory (BNL); Pigni, Marco T [ORNL; Pronyaev, V. G. [Institute of Physics and Power Engineering (IPPE), Obninsk, Russia; Sayer, Royce O [ORNL; Sonzogni, A A [Brookhaven National Laboratory (BNL); Summers, N C [Lawrence Livermore National Laboratory (LLNL); Talou, P [Los Alamos National Laboratory (LANL); Thompson, I J [Lawrence Livermore National Laboratory (LLNL); Trkov, A. [Jozef Stefan Institute, Slovenia; Vogt, R L [Lawrence Livermore National Laboratory (LLNL); Van der Marck, S S [Nucl Res & Consultancy Grp, Petten, Netherlands; Wallner, A [University of Vienna, Austria; White, M C [Los Alamos National Laboratory (LANL); Wiarda, Dorothea [ORNL; Young, P C [Los Alamos National Laboratory (LANL)

2011-01-01

The ENDF/B-VII.1 library is our latest recommended evaluated nuclear data file for use in nuclear science and technology applications, and incorporates advances made in the five years since the release of ENDF/B-VII.0. These advances focus on neutron cross sections, covariances, fission product yields and decay data, and represent work by the US Cross Section Evaluation Working Group (CSEWG) in nuclear data evaluation that utilizes developments in nuclear theory, modeling, simulation, and experiment. The principal advances in the new library are: (1) An increase in the breadth of neutron reaction cross section coverage, extending from 393 nuclides to 423 nuclides; (2) Covariance uncertainty data for 190 of the most important nuclides, as documented in companion papers in this edition; (3) R-matrix analyses of neutron reactions on light nuclei, including isotopes of He; Li, and Be; (4) Resonance parameter analyses at lower energies and statistical high energy reactions for isotopes of Cl; K; Ti, V, Mn, Cr, Ni, Zr and W; (5) Modifications to thermal neutron reactions on fission products (isotopes of Mo, Tc, Rh, Ag, Cs, Nd, Sm, Eu) and neutron absorber materials (Cd, Gd); (6) Improved minor actinide evaluations for isotopes of U, Np, Pu, and Am (we are not making changes to the major actinides (235,238)U and (239)Pu at this point, except for delayed neutron data and covariances, and instead we intend to update them after a further period of research in experiment and theory), and our adoption of JENDL-4.0 evaluations for isotopes of Cm, Bk, Cf, Es; Fm; and some other minor actinides; (7) Fission energy release evaluations; (8) Fission product yield advances for fission-spectrum neutrons and 14 MeV neutrons incident on (239)Pu; and (9) A new decay data sublibrary. Integral validation testing of the ENDF/B-VII.1 library is provided for a variety of quantities: For nuclear criticality, the VII.1 library maintains the generally-good performance seen for VII.0 for a wide

Study of short-lived fission products with the aid of an isotope separator connected to reactor R2-0

International Nuclear Information System (INIS)

Rudstam, G.

1976-01-01

This report constitutes a final report on project 74-3289 together with a preliminary report for project 75-3332. These projects have been included in the budget years 1974/75 and 1975/76 as a contribution to the operating costs of reactor R2-0 at Studsvik. The reactor was used for experimental studies on short-lived fission products with OSIRIS isotope-separator equipment. The scientific programme is very broad. It comprises, in the first place, characterisation of fission products (a study of their excitation levels, measurement of decay properties such as half-life and emission of delayed neutrons, determination of neutron energy spectrum, determination of total decay energy, etc.). An important application of this field of research is the determination of decay heat in nuclear fuel. The programme thus comprises research of a fundamental character and applied research. (H.E.G.)

Fission neutron multiplicity calculations

International Nuclear Information System (INIS)

Maerten, H.; Ruben, A.; Seeliger, D.

1991-01-01

A model for calculating neutron multiplicities in nuclear fission is presented. It is based on the solution of the energy partition problem as function of mass asymmetry within a phenomenological approach including temperature-dependent microscopic energies. Nuclear structure effects on fragment de-excitation, which influence neutron multiplicities, are discussed. Temperature effects on microscopic energy play an important role in induced fission reactions. Calculated results are presented for various fission reactions induced by neutrons. Data cover the incident energy range 0-20 MeV, i.e. multiple chance fission is considered. (author). 28 refs, 13 figs

Fission products detection in irradiated TRIGA fuel by means of gamma spectroscopy and MCNP calculation.

Science.gov (United States)

Cagnazzo, M; Borio di Tigliole, A; Böck, H; Villa, M

2018-05-01

Aim of this work was the detection of fission products activity distribution along the axial dimension of irradiated fuel elements (FEs) at the TRIGA Mark II research reactor of the Technische Universität (TU) Wien. The activity distribution was measured by means of a customized fuel gamma scanning device, which includes a vertical lifting system to move the fuel rod along its vertical axis. For each investigated FE, a gamma spectrum measurement was performed along the vertical axis, with steps of 1 cm, in order to determine the axial distribution of the fission products. After the fuel elements underwent a relatively short cooling down period, different fission products were detected. The activity concentration was determined by calibrating the gamma detector with a standard calibration source of known activity and by MCNP6 simulations for the evaluation of self-absorption and geometric effects. Given the specific TRIGA fuel composition, a correction procedure is developed and used in this work for the measurement of the fission product Zr 95 . This measurement campaign is part of a more extended project aiming at the modelling of the TU Wien TRIGA reactor by means of different calculation codes (MCNP6, Serpent): the experimental results presented in this paper will be subsequently used for the benchmark of the models developed with the calculation codes. Copyright © 2018 Elsevier Ltd. All rights reserved.

Modification of the fission product inventory program FISPIN

International Nuclear Information System (INIS)

Thomas, R.B.

1977-05-01

The fission product inventory program FISPIN calculates inventories of fission products, actinides and activation products, during and after irradiation in a nuclear reactor, estimates also being given for heat output and radioactive activity of the isotopes. The program has been developed further by making provision for the simulation of fuel reprocessing and in providing subroutines to make the program compatible with nuclear data in a slightly modified ENDF/B4 format. Continuous development of FISPIN over the years has however involved many program alterations and additions, and this has resulted in a generally untidy and cumbersome program. An attempt has therefore been made to improve the basic structure of the program. The subject is dealt with under the following headings: modularisation, direct access data, override facility, selective output, flowcharts, summary. (U.K.)

Measurements of fission product yield in the neutron-induced fission of {sup 238}U with average energies of 9.35 MeV and 12.52 MeV

Energy Technology Data Exchange (ETDEWEB)

Mukerji, Sadhana; Krishnani, Pritam Das; Suryanarayana, Saraswatula Venkat; Naik, Haladhara; Goswami, Ashok [Bhabha Atomic Research Centre, Mumbai (India); Shivashankar, Byrapura Siddaramaiah [Manipal University, Manipal (India); Mulik, Vikas Kaluram [University of Pune, Pune (India)

2014-07-15

The yields of various fission products in the neutron-induced fission of {sup 238}U with the flux-weighted averaged neutron energies of 9.35 MeV and 12.52 MeV were determined by using an off-line gamma ray spectroscopic technique. The neutrons were generated using the {sup 7}Li(p, n) reaction at Bhabha Atomic Research Centre-Tata Institute of Fundamental Research Pelletron facility, Mumbai. The gamma- ray activities of the fission products were counted in a highly-shielded HPGe detector over a period of several weeks to identify the decaying fission products. At both the neutron energies, the fission-yield values are reported for twelve fission product. The results obtained from the present work have been compared with the similar data for mono-energetic neutrons of comparable energy from the literature and are found to be in good agreement. The peak-to-valley (P/V) ratios were calculated from the fission-yield data and were found to decreases for neutron energy from 9.35 to 12.52 MeV, which indicates the role of excitation energy. The effect of the nuclear structure on the fission product-yield is discussed.

Activation calculation and environmental safety analysis for fusion experimental breeder (FEB)

Energy Technology Data Exchange (ETDEWEB)

Kaiming, Feng [Southwest Inst. of Physics, Leshan, SC (China)

1996-04-01

An activation calculation code FDKR and decay chain data library AFDCDLIB are used to calculate the radioactivity, decay heat, dose rate and biological hazard potential (BHP) form activation products, actinides and fission products in a Fusion Experiment Breeder (FEB). The code and library are introduced briefly, and calculation results and decay curves of related hazards after one year operation with 150 MW fusion power are given. The total radioactivity inventory, decay heat and BHP are 5.74 x 10{sup 20} Bq, 8.34 MW and 4.08 x 10{sup 8} km{sup 3} of air, respectively, at shutdown. Results obtained show that the first wall of FEB can meet the nuclear waste disposal criteria for the NRC 10 CFR61 Class C after a few weeks from shutdown. The inventory of important actinides for the fuel reprocessing, such as {sup 232}U and {sup 237}Np were also calculated. It was shown that their concentrations do not excess the limit value of environmental safety required. (9 refs., 4 figs., 9 tabs.).

Use of ELOCA.Mk5 to calculate transient fission product release from CANDU fuel elements

International Nuclear Information System (INIS)

Walker, J.R.; de Vaal, J.W.; Arimescu, V.I.; McGrady, T.G.; Wong, C.

1992-04-01

A change in fuel element power output, or a change in heat transfer conditions, will result in an immediate change in the temperature distribution in a fuel element. The temperature distribution change will be accompanied by concomitant changes in fuel stress distribution that lead, in turn, to a release of fission products to the fuel-to-sheath gap. It is important to know the inventory of fission products in the fuel-to-sheath gap, because this inventory is a major component of the source term for many postulated reactor accidents. ELOCA.Mk5 is a FORTRAN-77 computer code that has been developed to estimate transient releases to the fuel-to-sheath gap in CANDU reactors. ELOCA.Mk5 is an integration of the FREEDOM fission product release model into the ELOCA fuel element thermo-mechanical code. The integration of FREEDOM into ELOCA allows ELOCA.Mk5 to model the feedback mechanisms between the fission product release and the thermo-mechanical response of the fuel element. This paper describes the physical model, gives details of the ELOCA.Mkt code, and describes the validation of the model. We demonstrate that the model gives good agreement with experimental results for both steady state and transient conditions

Modeling Fission Product Sorption in Graphite Structures

International Nuclear Information System (INIS)

Szlufarska, Izabela; Morgan, Dane; Allen, Todd

2013-01-01

The goal of this project is to determine changes in adsorption and desorption of fission products to/from nuclear-grade graphite in response to a changing chemical environment. First, the project team will employ principle calculations and thermodynamic analysis to predict stability of fission products on graphite in the presence of structural defects commonly observed in very high-temperature reactor (VHTR) graphites. Desorption rates will be determined as a function of partial pressure of oxygen and iodine, relative humidity, and temperature. They will then carry out experimental characterization to determine the statistical distribution of structural features. This structural information will yield distributions of binding sites to be used as an input for a sorption model. Sorption isotherms calculated under this project will contribute to understanding of the physical bases of the source terms that are used in higher-level codes that model fission product transport and retention in graphite. The project will include the following tasks: Perform structural characterization of the VHTR graphite to determine crystallographic phases, defect structures and their distribution, volume fraction of coke, and amount of sp2 versus sp3 bonding. This information will be used as guidance for ab initio modeling and as input for sorptivity models; Perform ab initio calculations of binding energies to determine stability of fission products on the different sorption sites present in nuclear graphite microstructures. The project will use density functional theory (DFT) methods to calculate binding energies in vacuum and in oxidizing environments. The team will also calculate stability of iodine complexes with fission products on graphite sorption sites; Model graphite sorption isotherms to quantify concentration of fission products in graphite. The binding energies will be combined with a Langmuir isotherm statistical model to predict the sorbed concentration of fission products

ENDF/B-5 Fission Products Library. Rev. 2

International Nuclear Information System (INIS)

Schwerer, O.; Pronyaev, V.G.; Lemmel, H.D.

1984-07-01

This document summarizes contents and documentation of the 1984 version of the Fission Products Nuclear Data File of the ENDF/B-5 Library (Rev. 2) maintained by the National Nuclear Data Center (NNDC) at the Brookhaven National Laboratory, USA. This file contains numerical neutron reaction data and decay data for 877 fission product nuclides. The entire file or selective retrievals from it can be obtained on magnetic tape from the IAEA Nuclear Data Section. (author)

Beta-delayed fission and neutron emission calculations for the actinide cosmochronometers

International Nuclear Information System (INIS)

Meyer, B.S.; Howard, W.M.; Mathews, G.J.; Takahashi, K.; Moeller, P.; Leander, G.A.

1989-01-01

The Gamow-Teller beta-strength distributions for 19 neutron-rich nuclei, including ten of interest for the production of the actinide cosmochronometers, are computed microscopically with a code that treats nuclear deformation explicitly. The strength distributions are then used to calculate the beta-delayed fission, neutron emission, and gamma deexcitation probabilities for these nuclei. Fission is treated both in the complete damping and WKB approximations for penetrabilities through the nuclear potential-energy surface. The resulting fission probabilities differ by factors of 2 to 3 or more from the results of previous calculations using microscopically computed beta-strength distributions around the region of greatest interest for production of the cosmochronometers. The indications are that a consistent treatment of nuclear deformation, fission barriers, and beta-strength functions is important in the calculation of delayed fission probabilities and the production of the actinide cosmochronometers. Since we show that the results are very sensitive to relatively small changes in model assumptions, large chronometric ages for the Galaxy based upon high beta-delayed fission probabilities derived from an inconsistent set of nuclear data calculations must be considered quite uncertain

An investigation of fission models for high-energy radiation transport calculations

International Nuclear Information System (INIS)

Armstrong, T.W.; Cloth, P.; Filges, D.; Neef, R.D.

1983-07-01

An investigation of high-energy fission models for use in the HETC code has been made. The validation work has been directed checking the accuracy of the high-energy radiation transport computer code HETC to investigate the appropriate model for routine calculations, particularly for spallation neutron source applications. Model calculations are given in terms of neutron production, fission fragment energy release, and residual nuclei production for high-energy protons incident on thin uranium targets. The effect of the fission models on neutron production from thick uranium targets is also shown. (orig.)

ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data

Energy Technology Data Exchange (ETDEWEB)

Chadwick, M.B.; Herman, M.; Author(s): Chadwick,M.B.; Herman,M.; Oblozinsky,P.; Dunn,M.E.; Danon,Y.; Kahler,A.C.; Smith,D.L.; Pritychenko,B.; Arbanas,G.; Arcilla,R.; Brewer,R.; Brown,D.A.; Capote,R.; Carlson,A.D.; Cho,Y.S.; Derrien,H.; Guber,K.; Hale,G.M.; Hoblit,S.; Holloway,S.: Johnson,T.D.; Kawano,T.; Kiedrowski,B.C.; Kim,H.; Kunieda,S.; Larson,N.M.; Leal,L.; Lestone,J.P.; Little,R.C.; McCutchan,E.A.; MacFarlane,R.E.; MacInnes,M.; Mattoon,C.M.; McKnight,R.D.; Mughabghab,S.F.; Nobre,G.P.A.; Palmiotti,G.; Palumbo,A.; Pigni,M.T.; Pronyaev,V.G.; Sayer,R.O.; Sonzogni,A.A.; Summers,N.C.; Talou,P.; Thompson,I.J.; Trkov,A.; Vogt,R.L.; van der Marck,S.C.; Wallner,A.; White,M.C.; Wiarda,D.; Young,P.G.

2011-12-01

The ENDF/B-VII.1 library is our latest recommended evaluated nuclear data file for use in nuclear science and technology applications, and incorporates advances made in the five years since the release of ENDF/B-VII.0. These advances focus on neutron cross sections, covariances, fission product yields and decay data, and represent work by the US Cross Section Evaluation Working Group (CSEWG) in nuclear data evaluation that utilizes developments in nuclear theory, modeling, simulation, and experiment. The principal advances in the new library are: (1) An increase in the breadth of neutron reaction cross section coverage, extending from 393 nuclides to 423 nuclides; (2) Covariance uncertainty data for 190 of the most important nuclides, as documented in companion papers in this edition; (3) R-matrix analyses of neutron reactions on light nuclei, including isotopes of He, Li, and Be; (4) Resonance parameter analyses at lower energies and statistical high energy reactions for isotopes of Cl, K, Ti, V, Mn, Cr, Ni, Zr and W; (5) Modifications to thermal neutron reactions on fission products (isotopes of Mo, Tc, Rh, Ag, Cs, Nd, Sm, Eu) and neutron absorber materials (Cd, Gd); (6) Improved minor actinide evaluations for isotopes of U, Np, Pu, and Am (we are not making changes to the major actinides {sup 235,238}U and {sup 239}Pu at this point, except for delayed neutron data and covariances, and instead we intend to update them after a further period of research in experiment and theory), and our adoption of JENDL-4.0 evaluations for isotopes of Cm, Bk, Cf, Es, Fm, and some other minor actinides; (7) Fission energy release evaluations; (8) Fission product yield advances for fission-spectrum neutrons and 14 MeV neutrons incident on {sup 239}Pu; and (9) A new decay data sublibrary. Integral validation testing of the ENDF/B-VII.1 library is provided for a variety of quantities: For nuclear criticality, the VII.1 library maintains the generally-good performance seen for VII.0

Study of the desintegration of short-life fission products. Application to the mass distribution in the fission of 238U and 233U induced by 14MeV neutrons

International Nuclear Information System (INIS)

Cavallini, Pierre.

1975-01-01

Nuclear spectrometry of short-life fission products was investigated, together with direct applications to the study of mass and charge distribution in fission reactions. It is shown that, by choosing judiciously the target in which the fission product is created and owing to the differences in stabilities and evaporation temperatures of the compounds obtained, it is possible to separate some elements. For example, niobium was separated by heating after irradiation of a mixture of UC and RuCl 3 , and sublimation in a tube with temperature gradient. It was thus possible to study the 99 Nb isotope. Other classical chemical separation processes were used for yttrium and strontium. The half-lifes beta and gamma spectra, decay schemes of 93 Sr, 94 Y and 95 Y were studied. It was shown how to obtain mass distribution in fission using a nondestructive gamma analysis method. As an application, results obtained in the fission of 233 U and 238 U at 14 MeV are given [fr

Dynamical decay of nuclei at high temperature: competition between particle emission and fission decay

International Nuclear Information System (INIS)

Delagrange, H.; Gregoire, C.; Scheuter, F.; Abe, Y.

1985-06-01

A generalized diffusion equation is propounded to follow the time evolution of an excited nucleus towards fission including along the particle decay. This theoretical model is built in order to try to analyse the anomalous behaviour of particle emission observed in many experimental data for heavy-ion induced reactions. Some calculations for the systems 194 Hg, 170 Yb and 248 Cf are presented. A possible extension of this generalized formalism is suggested to deal more consistently with the experimental data. 52 refs. 10 figs.

Equilibrium fission model calculations

International Nuclear Information System (INIS)

Beckerman, M.; Blann, M.

1976-01-01

In order to aid in understanding the systematics of heavy ion fission and fission-like reactions in terms of the target-projectile system, bombarding energy and angular momentum, fission widths are calculated using an angular momentum dependent extension of the Bohr-Wheeler theory and particle emission widths using angular momentum coupling

Microscopic description of the competition between spontaneous fission and α -decay in neutron-rich Ra, U and Pu nuclei

International Nuclear Information System (INIS)

Rodríguez-Guzmán, R; Robledo, L M

2017-01-01

Constrained mean-field calculations, based on the Gogny-D1M energy density functional, have been carried out to describe fission in Ra, U and Pu nuclei with neutron number 144 ≤ N ≤ 176. Fission paths, collective masses and zero-point quantum vibrational and rotational corrections are used to compute the spontaneous fission half-lives. We also pay attention to isomeric states along the considered fission paths. Alpha decay half-lives have also been computed using a parametrization of the Viola-Seaborg formula. Though there exists a strong variance of the predicted fission rates with respect to the details involved in their computation a robust trend is obtained indicating, that with increasing neutron number fission dominates over α -decay. Our results also suggest that a dynamical treatment of pairing correlations is required within the microscopic studies of the fission process in heavy nuclear systems. (paper)

Map of calculated radioactivity of fission product, 3

International Nuclear Information System (INIS)

Takeda, Tsuneo

1977-02-01

In this work, the radioactivities of fission products were calculated and summarized in contour maps and tables depending on irradiation and cooling times. The irradiation condition and other parameters used for the present calculation are shown in the followings. Neutron flux (N sub(th)): 3x10 13 n/sec/cm 2 Atom number of uranium: 1 mole (6x10 23 , ca. 271 gUO 2 ) Enrichment of U-235: 2.7% Range of irradiation time: 60-6x10 7 sec (ca. 1.9 y) Range of cooling time: 60-6x10 7 sec (ca. 1.9 y). Values of the neutron flux and the enrichment treated here are representative for common LWRs. The maps and tables of 560 nuclides are divided and compiled into the following three volumes. Vol. I: Maps of radioactivity of overall total, element total and each nuclide (Ni - Zr), Vol. II: Maps of radioactivity of each nuclide (Nb - Sb), Vol. III: Maps of radioactivity of each nuclide (Te - Tm). (auth.)

Map of calculated radioactivity of fission product, (1)

International Nuclear Information System (INIS)

Takeda, Tsuneo

1977-02-01

In this work, the radioactivities of fission products were calculated and summarized in contour maps and tables depending on irradiation and cooling times. The irradiation condition and other parameters used for the present calculation are shown in the followings. Neutron flux (N sub(th)): 3x10 13 n/sec/cm 2 Atom number of uranium: 1 mole (6x10 23 , ca. 271 gUO 2 ) Enrichment of U-235: 2.7% Range of irradiation time: 60-6x10 7 sec (ca. 1.9 y) Range of cooling time: 60-6x10 7 sec (ca. 1.9 y). Values of the neutron flux and the enrichment treated here are representative for common LWRs. The maps and tables of 560 nuclides are divided and compiled into the following three volumes. Vol. I Maps of radioactivity of overall total, element total and each nuclide (Ni - Zr) Vol. II Maps of radioactivity of each nuclide (Nb - Sb) Vol. III Maps of radioactivity of each nuclide (Te - Tm) (auth.)

Map of calculated radioactivity of fission product, 2

International Nuclear Information System (INIS)

Takeda, Tsuneo

1977-02-01

In this work, the radioactivities of fission products were calculated and summarized in contour maps and tables depending on irradiation and cooling times. The irradiation condition and other parameters used for the present calculation are shown in the followings. Neutron flux (N sub(th)): 3x10 13 n/sec/cm 2 Atom number of uranium: 1 mole (6x10 23 , ca. 271 gUO 2 ) Enrichment of U-235: 2.7% Range of irradiation time: 60-6x10 7 sec (ca. 1.9 y) Range of cooling time: 60-6x10 7 sec (ca. 1.9 y). Values of the neutron flux and the enrichment treated here are representative for common LWRs. The maps and tables of 560 nuclides are divided and compiled into the following three volumes. Vol. I: Maps of radioactivity of overall total, element total and each nuclide (Ni - Zr), Vol. II: Maps of radioactivity of each nuclide (Nb - Sb), Vol. III: Maps of radioactivity of each nuclide (Te - Tm). (auth.)

Post shut-down decay heat removal from nuclear reactor core by natural convection loops in sodium pool

Energy Technology Data Exchange (ETDEWEB)

Rajamani, A. [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Sundararajan, T., E-mail: tsundar@iitm.ac.in [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Prasad, B.V.S.S.S. [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Parthasarathy, U.; Velusamy, K. [Nuclear Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India)

2016-05-15

Highlights: • Transient simulations are performed for a worst case scenario of station black-out. • Inter-wrapper flow between various sub-assemblies reduces peak core temperature. • Various natural convection paths limits fuel clad temperatures below critical level. - Abstract: The 500 MWe Indian pool type Prototype Fast Breeder Reactor (PFBR) has a passive core cooling system, known as the Safety Grade Decay Heat Removal System (SGDHRS) which aids to remove decay heat after shut down phase. Immediately after reactor shut down the fission products in the core continue to generate heat due to beta decay which exponentially decreases with time. In the event of a complete station blackout, the coolant pump system may not be available and the safety grade decay heat removal system transports the decay heat from the core and dissipates it safely to the atmosphere. Apart from SGDHRS, various natural convection loops in the sodium pool carry the heat away from the core and deposit it temporarily in the sodium pool. The buoyancy driven flow through the small inter-wrapper gaps (known as inter-wrapper flow) between fuel subassemblies plays an important role in carrying the decay heat from the sub-assemblies to the hot sodium pool, immediately after reactor shut down. This paper presents the transient prediction of flow and temperature evolution in the reactor subassemblies and the sodium pool, coupled with the safety grade decay heat removal system. It is shown that with a properly sized decay heat exchanger based on liquid sodium and air chimney stacks, the post shutdown decay heat can be safely dissipated to atmospheric air passively.

Activity inventories and decay heat calculations for a DEMO with HCPB and HCLL blanket modules

International Nuclear Information System (INIS)

Stankunas, Gediminas; Tidikas, Andrius; Pereslavstev, Pavel; Catalán, Juan; García, Raquel; Ogando, Francisco; Fischer, Ulrich

2016-01-01

Highlights: • The afterheat and activity inventories were calculated for Eurofer steel which is the reference structural material for DEMO. • The decay heat for the HCPB DEMO was found to be larger than for the HCLL both for short and longer cooling times. • The comparison calculations were performed for a single outboard blanket module of the HCLL DEMO assuming High-Temperature Ferritic–Martensitic (HT-FM) steel and SS-316 (LN) as structural material. - Abstract: Activation inventories, decay heat and radiation doses are important nuclear quantities which need to be assessed on a reliable basis for the safe operation of a fusion nuclear power reactor. The afterheat and activity inventories were shown to be dominated by the Eurofer steel which is the reference structural material for DEMO. The decay heat for the HCPB DEMO was found to be larger than for the HCLL both for short (a few days) and longer (more than a year) cooling times. As for the alternative steels, the induced radioactivity was turned out to be lowest for the SS-316 until about 200 years after shut-down. Afterwards, the activity level of SS-316 steel was found to be the highest. For these times, the activity of both Eurofer and the HT-FM steel is about one order of magnitude lower.

Activity inventories and decay heat calculations for a DEMO with HCPB and HCLL blanket modules

Energy Technology Data Exchange (ETDEWEB)

Stankunas, Gediminas, E-mail: gediminas.stankunas@lei.lt [Lithuanian Energy Institute, Laboratory of Nuclear Installation Safety, Breslaujos Str. 3, LT-44403 Kaunas (Lithuania); Tidikas, Andrius [Lithuanian Energy Institute, Laboratory of Nuclear Installation Safety, Breslaujos Str. 3, LT-44403 Kaunas (Lithuania); Pereslavstev, Pavel [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Catalán, Juan; García, Raquel; Ogando, Francisco [Departamento de Ingeniería Energética, UNED, 28040 Madrid (Spain); Fischer, Ulrich [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

2016-11-01

Highlights: • The afterheat and activity inventories were calculated for Eurofer steel which is the reference structural material for DEMO. • The decay heat for the HCPB DEMO was found to be larger than for the HCLL both for short and longer cooling times. • The comparison calculations were performed for a single outboard blanket module of the HCLL DEMO assuming High-Temperature Ferritic–Martensitic (HT-FM) steel and SS-316 (LN) as structural material. - Abstract: Activation inventories, decay heat and radiation doses are important nuclear quantities which need to be assessed on a reliable basis for the safe operation of a fusion nuclear power reactor. The afterheat and activity inventories were shown to be dominated by the Eurofer steel which is the reference structural material for DEMO. The decay heat for the HCPB DEMO was found to be larger than for the HCLL both for short (a few days) and longer (more than a year) cooling times. As for the alternative steels, the induced radioactivity was turned out to be lowest for the SS-316 until about 200 years after shut-down. Afterwards, the activity level of SS-316 steel was found to be the highest. For these times, the activity of both Eurofer and the HT-FM steel is about one order of magnitude lower.

Fission product retention in TRISO coated UO2 particle fuels subjected to HTR simulated core heating tests

International Nuclear Information System (INIS)

Baldwin, C.A.; Kania, M.J.

1991-01-01

Results of the examination and analysis of 25,730 individual microspheres from spherical fuel elements HFR-K3/1 and HFR-K3/3 are reported. The parent spheres were irradiated in excess of end-of-life exposure and subsequently subjected to simulated core heating tests in a special high-temperature furnace at Forschungszentrum, Juelich, GmbH (KFA). Following the heating tests, the spheres were electrolytically deconsolidated to obtain unbounded fuel particles for Irradiated Microsphere Gamma Analyzer (IMGA) analysis. For sphere HFR-K3/1, which was heated for 500 h at 1600 deg. C, only four particles were identified as having released fission products. The remaining particles from the sphere showed no statistical evidence of fission product release. Scanning Electron Microscopy (SEM) examination showed that three of the defect particles had large sections of the TRISO coating missing, while the fourth appeared normal. For sphere HFR-K3/3, which was heated for 100 h at 1800 deg. C, the IMGA data revealed that fission product release (cesium) from individual particles was significant and that there was large particle-to-particle variation in retention capabilities. Individual particle release (cesium) averaged ten times the KFA-measured integral spherical fuel element release value. In addition, the bimodal distribution of the individual particle data indicated that two distinct modes of failure at fuel temperatures of 1800 deg. C and above may exist. (author). 6 refs, 6 figs, 4 tabs

Fission product retention in TRISCO coated UO2 particle fuels subjected to HTR simulated core heating tests

International Nuclear Information System (INIS)

Baldwin, C.A.; Kania, M.J.

1990-11-01

Results of the examination and analysis of 25,730 individual microspheres from spherical fuel elements HFR-K3/1 and HFR-K3/3 are reported. The parent spheres were irradiated in excess of end-of-life exposure and subsequently subjected to simulated core heating tests in a special high-temperature furnace at Forschungszentrum, Juelich, GmbH (KFA). Following the heating tests, the spheres were electrolytically deconsolidated to obtain unbonded fuel particles for Irradiated Microsphere Gamma Analyzer (IMGA) analysis. For sphere HFR-K3/1, which was heated for 500 h at 1600 degree C, only four particles were identified as having released fission products. The remaining particles from the sphere showed no statistical evidence of fission product release. Scanning Electron Microscopy (SEM) examination showed that three of the defect particles had large sections of the TRISO coating missing, while the fourth appeared normal. For sphere HFR-K3/3, which was heated for 100 h at 1800 degree C, the IMGA data revealed that fission product release (cesium) from individual particles was significant and that there was large particle-to-particle variation in retention capabilities. Individual particle release (cesium) averaged ten times the KFA-measured integral spherical fuel element release value. In addition, the bimodal distribution of the individual particle data indicated that two distinct modes of failure at fuel temperatures of 1800 degree C and above may exist. 6 refs., 6 figs., 4 tabs

Yields of fission products produced by thermal-neutron fission of 229Th

International Nuclear Information System (INIS)

Dickens, J.K.; McConnell, J.W.

1983-01-01

Absolute yields have been determined for 47 gamma rays emitted in the decay of 37 fission products representing 25 mass chains created during thermal-neutron fission of 229 Th. Using a Ge(Li) detector, spectra were obtained of gamma rays emitted between 15 min and 0.4 yr after very short irradiations by thermal neutrons of a 15-μg sample of 229 Th. On the basis of measured gamma-ray yields and known nuclear data, yields for cumulative production of 37 fission products were deduced. The absolute overall normalization uncertainty is 235 U, we postulate a simple functional dependence sigma = sigma(Z/sub p/), and using this dependence obtain values of Z/sub p/(A) for 15 mass chains created during fission of 229 Th. Values of Z/sub p/(A) were estimated for other mass chains based upon results of a recent study of Z/sub p/(A). Charge distributions determined using the deduced mass distribution and the deduced sets of Z/sub p/(A) and sigma(Z/sub p/) are in very good agreement with recent measurements, exhibiting a pronounced even-odd effect in elemental yields. These results may be used to predict unmeasured yields for 229 Th fission

Thermophysical modeling of volatile fission product release from a debris pool

International Nuclear Information System (INIS)

Yun, J. I.; Suh, K. Y.; Kang, C. S.

1999-01-01

A model is described for fission product release from the debris pool in the lower plenum of the reactor pressure vessel. In the pool, turbulent natural convection flow is formed due to homogeneous internal heat generation. Using the best-known correlations, heat transfer at the curved bottom and the top of the pool may be calculated. Volatile fission product gases in the pool nucleate and diffuse to bubbles. Both the homogeneous nucleation and heterogeneous nucleation are considered. The bubble nucleation, growth, coalescence and loss due to rise is modeled pursuant to bubble dynamics. If the pressure and temperature of the pool are very high, homogeneous nucleation that accounts for effect of decrease in the pool pressure can occur. The effect of the bubble-to-pool interfacial tension and the pool pressure on the nucleation rate is investigated in this work

CACA-2: revised version of CACA-a heavy isotope and fission-product concentration calculational code for experimental irradiation capsules

International Nuclear Information System (INIS)

Allen, E.J.

1976-02-01

A computer program is described which calculates nuclide concentration histories, power or neutron flux histories, burnups, and fission-product birthrates for fueled experimental capsules subjected to neutron irradiations. Seventeen heavy nuclides in the chain from 232 Th to 242 Pu and a user-specified number of fission products are treated. A fourth-order Runge-Kutta calculational method solves the differential equations for nuclide concentrations as a function of time. For a particular problem, a user-specified number of fuel regions may be treated. A fuel region is described by volume, length, and specific irradiation history. A number of initial fuel compositions may be specified for each fuel region. The irradiation history for each fuel region can be divided into time intervals, and a constant power density or a time-dependent neutron flux is specified for each time interval. Also, an independent cross-section set may be selected for each time interval in each irradiation history. The fission-product birthrates for the first composition of each fuel region are summed to give the total fission-product birthrates for the problem

Systematics of Fission-Product Yields

International Nuclear Information System (INIS)

Wahl, A.C.

2002-01-01

Empirical equations representing systematics of fission-product yields have been derived from experimental data. The systematics give some insight into nuclear-structure effects on yields, and the equations allow estimation of yields from fission of any nuclide with atomic number Z F = 90 thru 98, mass number A F = 230 thru 252, and precursor excitation energy (projectile kinetic plus binding energies) PE = 0 thru ∼200 MeV--the ranges of these quantities for the fissioning nuclei investigated. Calculations can be made with the computer program CYFP. Estimates of uncertainties in the yield estimates are given by equations, also in CYFP, and range from ∼ 15% for the highest yield values to several orders of magnitude for very small yield values. A summation method is used to calculate weighted average parameter values for fast-neutron (∼ fission spectrum) induced fission reactions

Systematics of Fission-Product Yields

Energy Technology Data Exchange (ETDEWEB)

A.C. Wahl

2002-05-01

Empirical equations representing systematics of fission-product yields have been derived from experimental data. The systematics give some insight into nuclear-structure effects on yields, and the equations allow estimation of yields from fission of any nuclide with atomic number Z{sub F} = 90 thru 98, mass number A{sub F} = 230 thru 252, and precursor excitation energy (projectile kinetic plus binding energies) PE = 0 thru {approx}200 MeV--the ranges of these quantities for the fissioning nuclei investigated. Calculations can be made with the computer program CYFP. Estimates of uncertainties in the yield estimates are given by equations, also in CYFP, and range from {approx} 15% for the highest yield values to several orders of magnitude for very small yield values. A summation method is used to calculate weighted average parameter values for fast-neutron ({approx} fission spectrum) induced fission reactions.

Comparison of decay and yield data between JNDC2 and ENDF/B-VI

Energy Technology Data Exchange (ETDEWEB)

Oyamatsu, K.; Sagosaka, M.; Miyazono, T. [Nagoya Univ. (Japan)

1997-03-01

This work is intended to be our first step to solve disagreements of the decay heat powers between measurements and summation calculations. We examine differences between nuclear data libraries to complement our uncertainty evaluation of the decay heat summation calculations only with ENDF/B-VI. The comparison is made mainly between JNDC2 and ENDF/B-VI while JEF2.2 decay data is also discussed. In this study, we propose and use a simple method which is an analogue of the overlap integral of two wave functions in quantum mechanics. As the first step, we compare the whole input nuclear data for the summation calculations as a whole. We find a slight difference of the fission yields especially for high-energy neutron induced fissions between JNDC2 and ENDF/B-VI. As for the decay energies, JNDC2, ENDF/B-VI are quite similar while JEF2.2 is found significantly different from these two libraries. We find substantial differences in the decay constant values among the three libraries. As the second step, we calculate the decay heat powers with FPGS90 using JNDC2 and ENDF/B-VI. The total decay heat powers with the two libraries differ by more than 10% at short cooling times while they agree well on the average at cooling times longer that 100 (s). We also point out nuclides whose contributions are significantly different between the two libraries even though the total decay heats agree well. These nuclides may cause some problems in predicting aggregate spectra of {beta} and {gamma} rays as well as delayed neutrons, and are to be reviewed in the future revision of decay and yield data. (author)

A high temperature heating device for the study of fission product release from nuclear fuel

International Nuclear Information System (INIS)

Svedkauskaite-Le Gore, Jolanta; Kivel, Niko; Guenther-Leopold, Ines

2010-01-01

At the Paul Scherrer Institute a high temperature inductive heating furnace, which can heat fuel samples up to 2300 deg. C, has been developed in order to study the release of fission products. The furnace can be directly connected to an inductively coupled plasma mass spectrometer for online monitoring of the released elements and does not require their trapping before measurement. This paper describes the design of the inductive heating furnace, discusses its operating parameters, limitations and illustrates foreseen applications. (authors)

Contribution of short-lived nuclides to decay heat

International Nuclear Information System (INIS)

Katakura, Jun-ichi

1987-01-01

Comments are made on the calculation of decay heat, centering on evaluation of average decay energy. It is difficult to obtain sufficiently useful decay diagrams of short lived nucleides. High-energy levels are often missing in inferior decay diagrams, leading to an overestimation of the intensity of beta-rays at low-energy levels. Such an overestimation or underestimation due to the inferiority of a decay diagram is referred to as pandemonium effect. The pandemonium effect can be assessed by means of the ratio of the measured energy of the highest level of the daughter nuclide to the Q β -value of the beta-decay. When a satisfactory decay diagram cannot be obtained, the average decay energy has to be estimated by theoretical calculation. The gross theory for beta-decay proposed by Yamada and Takahashi is employed for the calculation. To carry out the calculation according to this theory, it is required to determine the value for the parameter Q 00 , the lowest energy of the daughter nuclide that meets the selection rule for beta-decay. Currently, Q 00 to be used for this purpose is estimated from data on the energy of the lowest level found in a decay diagram, even if it is inferior. Some examples of calculation of decay heat using the average beta- or gamma-ray energy are shown and compared with measurements. (author)

Angular momentum distribution of primary fission fragments by measurement of the relative yield of isomeric fission products

International Nuclear Information System (INIS)

Dornhoefer, H.

1980-01-01

The fission products 132 I and 136 I produced in the fission reactions 238 U(α,f) and 238 U(d,f) were spectroscoped using a gas transport system. Thereby was taken advantage of the fact that at the transport with pure helium without aerosols only iodine activities were collected in a membrane filter. The relative independent yields of the isomeric fission products of 132 I and 136 I were determined for different excitation energies. Thereby was taken advantage of the fact that the transport yield of the gas transport system for 136 I directly produced from the fission was greater than for iodine indirectly produced by β-decay. (orig./HSI) [de

Cluster decay analysis and related structure effects of fissionable ...

Indian Academy of Sciences (India)

2015-08-01

Aug 1, 2015 ... Collective clusterization approach of dynamical cluster decay model (DCM) has been ... fusion–fission process resulting in the emission of symmetric and/or ... represents the relative separation distance between two fragments or clusters ... decay constant λ or decay half-life T1/2 is defined as λ = (ln 2/T1/2) ...

An evaluation of neutron and gamme heating in fission product isotopes

International Nuclear Information System (INIS)

Leal, L.C.; Hill, R.N.; Khalil, H.S.

1993-01-01

The accurate prediction of the energy deposition rate in fast reactors, particularly in blanket and nonfueled regions, requires explicit treatment of gamma photon transport. Such an explicit treatment is part of the coupled neutron-photon heating method in use at Argonne National Laboratory, (ANL). In applying this procedure, three approximations are made in connection with the modeling of fission products (FPs): 1. The contribution of the FP neutron interactions to the gamma source is neglected. 2. In computing the macroscopic gamma interaction cross sections, the FPs are either neglected or simulated with an element (usually molybdenum) representative of an open-quotes averageclose quotes FP. 3. The heating contribution of the FP is neglected by use of zero FP kerma factors

Development of an integrated fission product release and transport code for spatially resolved full-core calculations of V/HTRs

International Nuclear Information System (INIS)

Xhonneux, Andre; Allelein, Hans-Josef

2014-01-01

The computer codes FRESCO-I, FRESCO-II, PANAMA and SPATRA developed at Forschungszentrum Jülich in Germany in the early 1980s are essential tools to predict the fission product release from spherical fuel elements and the TRISO fuel performance, respectively, under given normal or accidental conditions. These codes are able to calculate a conservative estimation of the source term, i.e. quantity and duration of radionuclide release. Recently, these codes have been reversed engineered, modernized (FORTRAN 95/2003) and combined to form a consistent code named STACY (Source Term Analysis Code System). STACY will later become a module of the V/HTR Code Package (HCP). In addition, further improvements have been implemented to enable more detailed calculations. For example the distinct temperature profile along the pebble radius is now taken into account and coated particle failure rates can be calculated under normal operating conditions. In addition, the absolute fission product release of an V/HTR pebble bed core can be calculated by using the newly developed burnup code Topological Nuclide Transformation (TNT) replacing the former rudimentary approach. As a new functionality, spatially resolved fission product release calculations for normal operating conditions as well as accident conditions can be performed. In case of a full-core calculation, a large number of individual pebbles which follow a random path through the reactor core can be simulated. The history of the individual pebble is recorded, too. Main input data such as spatially resolved neutron fluxes and fluid dynamics data are provided by the VSOP code. Capabilities of the FRESCO-I and SPATRA code which allow for the simulation of the redistribution of fission products within the primary circuit and the deposition of fission products on graphitic and metallic surfaces are also available in STACY. In this paper, details of the STACY model and first results for its application to the 200 MW(th) HTR

Code Development on Fission Product Behavior under Severe Accident-Validation of Aerosol Sedimentation

International Nuclear Information System (INIS)

Ha, Kwang Soon; Kim, Sung Il; Jang, Jin Sung; Kim, Dong Ha

2016-01-01

aerosol decay masses between the developed fission product module and MELCOR code. These discrepancies come from model differences of the aerosol sedimentation and steam condensation. The fission product module will be updated on the species release from the gap between the fuel and cladding, decay heat evaluation, aerosol size tracking, etc.

Code Development on Fission Product Behavior under Severe Accident-Validation of Aerosol Sedimentation

Energy Technology Data Exchange (ETDEWEB)

Ha, Kwang Soon; Kim, Sung Il; Jang, Jin Sung; Kim, Dong Ha [KAERI, Daejeon (Korea, Republic of)

2016-05-15

aerosol decay masses between the developed fission product module and MELCOR code. These discrepancies come from model differences of the aerosol sedimentation and steam condensation. The fission product module will be updated on the species release from the gap between the fuel and cladding, decay heat evaluation, aerosol size tracking, etc.

Yields of products from thermal-neutron induced fission of 235U

International Nuclear Information System (INIS)

Rudstam, G.; Aagaard, P.; Zwicky, H.U.

1985-01-01

Methods for fission yield determinations at an ISOL-system connected to a nuclear reactor have been developed. The present report contains detailed descriptions both of the experimental techniques and of the method used to correct the experimental yields for the decay of short-lived nuclear species in the delay between production and measurement. The methods have been applied to the determination of the fission yields of 40 fission products including 2 isometric pairs in the light mass region and those of 99 fission products including 25 isometric pairs or triplets in the heavy mass region. For 64 cases this is the first determination published. (author)

Calculation of the Fission Product Release for the HTR-10 based on its Operation History

International Nuclear Information System (INIS)

Xhonneux, A.; Druska, C.; Struth, S.; Allelein, H.-J.

2014-01-01

Since the first criticality of the HTR-10 test reactor in 2000, a rather complex operation history was performed. As the HTR-10 is the only pebble bed reactor in operation today delivering experimental data for HTR simulation codes, an attempt was made to simulate the whole reactor operation up to the presence. Special emphasis was put on the fission product release behaviour as it is an important safety aspect of such a reactor. The operation history has to be simulated with respect to the neutronics, fluid mechanics and depletion to get a detailed knowledge about the time-dependent nuclide inventory. In this paper we report about such a simulation with VSOP 99/11 and our new fission product release code STACY. While STACY (Source Term Analysis Code System) so far was able to calculate the fission product release rates in case of an equilibrium core and during transients, it now can also be applied to running-in-phases. This coupling demonstrates a first step towards an HCP Prototype. Based on the published power histogram of the HTR-10 and additional information about the fuel loading and shuffling, a coupled neutronics, fluid dynamics and depletion calculation was performed. Special emphasis was put on the complex fuel-shuffling scheme within both VSOP and STACY. The simulations have shown that the HTR-10 up to now generated about 2580 MWd while reshuffling the core about 2.3 times. Within this paper, STACY results for the equilibrium core will be compared with FRESCO-II results being published by INET. Compared to these release rates, which are based on a few user defined life histories, in this new approach the fission product release rates of Ag-110m, Cs-137, Sr-90 and I-131 have been simulated for about 4000 tracer pebbles with STACY. For the calculation of the HTR-10 operation history time-dependent release rates are being presented as well. (author)

Migration of fission products in UO2. Final report

International Nuclear Information System (INIS)

Prussin, S.G.; Olander, D.R.

1995-01-01

Results of an experimental and calculational effort to examine the fundamental mechanisms of fission product migration in and release from polycrystalline uranium dioxide are reported. The experiments were designed to provide diffusion parameters for the representative fission products tellurium, iodine, xenon, molybdenum and ruthenium under both reducing and oxidizing conditions. The calculational effort applied a new model of fission product release from reactor fuel that incorporates grain growth as well as grain boundary and lattice diffusion

Development of limiting decay heat values

International Nuclear Information System (INIS)

Khotylev, V.A.; Thompson, J.W.; Gibb, R.A.

1999-01-01

A number of tools are used in the assessment of decay heat during an outage of the CANDU-6. Currently, the technical basis for all of these tools is 'CANDU Channel Decay Power', Reference 1. The methods used in that document were limited to channel decay powers. However, for most outage support analysis, decay heat limits are based on bundle heats. Since the production of that document in 1977, new versions of codes, and updates of general-purpose and CANDU-specific libraries have become available. These tools and libraries have both a more formal technical basis than Reference 1, and also a more formal validation base. Using these tools it is now possible to derive decay heat with more specific input parameters, such as fuel composition, heat per unit of fuel, and irradiation history, and to assign systematically derived uncertainty allowances to such decay heat values. In particular, we sought to examine a broad range of likely bundle histories, and thus establish a set of limiting bundle decay beat values, that could serve as a bounding envelope for use in Nuclear Safety Analysis. (author)

Fission product release during MCCI. CEC nuclear safety program: MCCI project

Energy Technology Data Exchange (ETDEWEB)

Cenerino, G [CEA Centre d` Etudes Nucleaires de Fontenay-aux-Roses, 92 (France). Inst. de Protection et de Surete Nucleaire; Cordfunke, E H.P. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Hunterlaar, M E [Netherlands Energy Research Foundation (ECN), Petten (Netherlands)

1995-01-01

The assessment of the consequences of severe accidents in nuclear reactors involving molten core-concrete interactions (MCCls) requires estimates of the quantities and physicochemical forms of the radioactive species released from the melt into the cavity atmosphere. Such estimates in turn require a detailed knowledge of the complex chemical interactions which would occur between the fission products, fuel and the components of the core structural materials and the concrete. In recent years, effort has been put into the thermodynamic characterization of these processes. The results of such studies are important for predicting several aspects of MCCls, including: 1. The release of species by vaporization; 2. the extent of concrete penetration: a. The melt solidus and liquidus temperatures, which in turn affect the heat transfer processes and hence tile predictions of the melt temperature and the onset of solidification, b. the amounts of the solid and liquid phases and the respective compositions, which determines the viscosity of the melt, and c. the composition of the crust formed following the addition of water to quench the interaction. d. the distribution of fission products among metallic and oxidic phases. This SOAR is devoted to thermochemical calculations in the context of MCCI where most fission products and the metallic components of the melt are transferred into an oxidic form sooner or later. Calculations on fission product release from a molten pool without MCCI are underway in the source term project of the CEC-RCA. The following conditions have to be taken into account in order to be able to perform reliable thermodynamic calculations. (orig./HP).

Fission product release during MCCI. CEC nuclear safety program: MCCI project

International Nuclear Information System (INIS)

Cenerino, G.; Hunterlaar, M.E.

1995-01-01

The assessment of the consequences of severe accidents in nuclear reactors involving molten core-concrete interactions (MCCls) requires estimates of the quantities and physicochemical forms of the radioactive species released from the melt into the cavity atmosphere. Such estimates in turn require a detailed knowledge of the complex chemical interactions which would occur between the fission products, fuel and the components of the core structural materials and the concrete. In recent years, effort has been put into the thermodynamic characterization of these processes. The results of such studies are important for predicting several aspects of MCCls, including: 1. The release of species by vaporization; 2. the extent of concrete penetration: a. The melt solidus and liquidus temperatures, which in turn affect the heat transfer processes and hence tile predictions of the melt temperature and the onset of solidification, b. the amounts of the solid and liquid phases and the respective compositions, which determines the viscosity of the melt, and c. the composition of the crust formed following the addition of water to quench the interaction. d. the distribution of fission products among metallic and oxidic phases. This SOAR is devoted to thermochemical calculations in the context of MCCI where most fission products and the metallic components of the melt are transferred into an oxidic form sooner or later. Calculations on fission product release from a molten pool without MCCI are underway in the source term project of the CEC-RCA. The following conditions have to be taken into account in order to be able to perform reliable thermodynamic calculations. (orig./HP)

LEAF: a computer program to calculate fission product release from a reactor containment building for arbitrary radioactive decay chains

International Nuclear Information System (INIS)

Lee, C.E.; Apperson, C.E. Jr.; Foley, J.E.

1976-10-01

The report describes an analytic containment building model that is used for calculating the leakage into the environment of each isotope of an arbitrary radioactive decay chain. The model accounts for the source, the buildup, the decay, the cleanup, and the leakage of isotopes that are gas-borne inside the containment building

Delayed β ray spectrum of 235U fission fragments

International Nuclear Information System (INIS)

Pascholati, P.R.

1973-01-01

The time-dependent electron spectra of fission fragments from the thermal-neutron-induced fission of 235 U are calculated. The Gross theory of nuclear beta decay is used to obtain the decay constant and individual electron spectra. The mean energy per fission carried by the electrons and the number of electrons per fission are also calculated. Comparison of these calculated spectra to experimental ones shows good agreements. (Author) [pt

Comparison of yield and decay data among JNDC2, ENDF/B-VI and JEF2.2

Energy Technology Data Exchange (ETDEWEB)

Oyamatsu, Kazuhiro; Sagisaka, Mitsuyuki; Miyazono, Toshimitsu [Nagoya Univ. (Japan)

1997-03-01

Fission yields and decay data for fission product summation calculations are compared among JNDC2 and ENDF/B-VI and JEF2.2. Special attention is paid to the summation calculation of the total delayed neutrons per fission because it requires the data of the most unstable nuclides among all fission products. The cumulative fission yields of delayed neutron precursors are found to be appreciably different among the libraries even though values of the independent fission yields and the total number of delayed neutrons are chosen to be in fair agreement with each other. This suggests that there still exist large uncertainties in delayed neutron emission probabilities (or decay chains) for the precursors far from the stability line. (author)

Neutron emission and fragment yield in high-energy fission

International Nuclear Information System (INIS)

Grudzevich, O. T.; Klinov, D. A.

2013-01-01

The KRIS special library of spectra and emission probabilities in the decays of 1500 nuclei excited up to energies between 150 and 250 MeV was developed for correctly taking into account the decay of highly excited nuclei appearing as fission fragments. The emission of neutrons, protons, and photons was taken into account. Neutron emission fromprimary fragments was found to have a substantial effect on the formation of yields of postneutron nuclei. The library was tested by comparing the calculated and measured yields of products originating from the fission of nuclei that was induced by high-energy protons. The method for calculating these yields was tested on the basis of experimental data on the thermal-neutroninduced fission of 235 U nuclei

Decay and fission of the oriented nuclei

CERN Document Server

Kadmenskij, S G

2002-01-01

The fragment angular distributions for binary decay of oriented spherical and deformed nuclei with taking into account the correct transformational properties of wave functions under time inversion have been investigated. It has been shown that for description of fragment angular distributions the adiabatic approximation for collective rotational nuclear degrees of freedom is not correct. It has been demonstrated that this approximation is valid for description of spontaneous and induced low-energy nuclear fission. The dependence of partial fission widths on the orientation of the internal axes spins, projections of spins, and relative angular moments of fission fragments has been analyzed. It has been shown that the adiabatic approximation results in coherent interference of wave functions of fragments relative movement. This interference forms fragments the universal angular distributions of fission fragments for oriented nuclei. For these distributions the deviations from A. Bohr's formula have been invest...

Calculation of the nuclear fission data based on the framework of the QMD + SDM

International Nuclear Information System (INIS)

Rong Jian; Iwamoto, O.; Fukahori, T.

2002-01-01

The quantum molecular dynamics (QMD), statistical decay model (SDM) and the statistical fission theory were used to analyze the mass distribution of the fission products, the prompt fission neutron spectrum (x(E)) and the prompt fission neutron multiplicities (ν-bar pf (E)) caused by the intermediate energy nucleon-induced fission. The semi-empirical formula of energy level density parameter used in the statistical process was also studied. Very few adjustable parameters were included in the present method. By some physical analysis, it can be thought that the present results are reasonable. The x(E) and ν-bar pf (E) can be obtained in the intermediate energy region by the present method

Methodology and application of the WIMS-D4M fission product data

International Nuclear Information System (INIS)

Mo, S.C.

1995-01-01

The WIMS-D4 code has been modified (WIMS-D4m) to generate burn-up dependent microscopic cross sections for use in full core depletion calculations. The calculation of neutron absorption by fission products can be obtained from a reduced fission-product-chain model that includes the 135 Xe and 149 Sm chains, and a lumped fission product to account for the absorption by fission products not explicitly treated. Burn-up calculations were performed for the ANS MEU core using WIMS and EPRI-CELL cross sections. The calculated eigenvalues and material loadings are in good agreements

Cement As a Waste Form for Nuclear Fission Products: The Case of (90)Sr and Its Daughters.

Science.gov (United States)

Dezerald, Lucile; Kohanoff, Jorge J; Correa, Alfredo A; Caro, Alfredo; Pellenq, Roland J-M; Ulm, Franz J; Saúl, Andrés

2015-11-17

One of the main challenges faced by the nuclear industry is the long-term confinement of nuclear waste. Because it is inexpensive and easy to manufacture, cement is the material of choice to store large volumes of radioactive materials, in particular the low-level medium-lived fission products. It is therefore of utmost importance to assess the chemical and structural stability of cement containing radioactive species. Here, we use ab initio calculations based on density functional theory (DFT) to study the effects of (90)Sr insertion and decay in C-S-H (calcium-silicate-hydrate) in order to test the ability of cement to trap and hold this radioactive fission product and to investigate the consequences of its β-decay on the cement paste structure. We show that (90)Sr is stable when it substitutes the Ca(2+) ions in C-S-H, and so is its daughter nucleus (90)Y after β-decay. Interestingly, (90)Zr, daughter of (90)Y and final product in the decay sequence, is found to be unstable compared to the bulk phase of the element at zero K but stable when compared to the solvated ion in water. Therefore, cement appears as a suitable waste form for (90)Sr storage.

Calculation and Evaluation of Fission Yields and Capture Cross Sections Leading to the Production of Therapeutic Radionuclide by Means of Nuclear Reactors

International Nuclear Information System (INIS)

Sublet, J.C.

2009-01-01

Much progress has been made in nuclear medicine that involves the use of radionuclides for both diagnosis and therapy. Because of this qualitative and quantitative growth, the adoption of a set of established radionuclides for various applications, the methods of nuclide production need to be addressed and consideration given to other, emerging radionuclides that are judged to be developing in importance. The methods involved are characterized by the transmutation of isotopes by neutron-induced reactions and decays. Therefore, newly evaluated cross sections, fission yields and decay characteristics of relevance to the reactor production of those therapeutic radionuclides have been reviewed. Considerations of the decay schemes of all the nuclides involved are also included. (author)

Preparation of a primary target for the production of fission products in a nuclear reactor

International Nuclear Information System (INIS)

Arino, H.; Cosolito, F.J.; George, K.D.; Thornton, A.K.

1976-01-01

A primary target for the production of fission products in a nuclear reactor, such as uranium or plutonium fission products, is comprised of an enclosed, cylindrical vessel, preferably comprised of stainless steel, having a thin, continuous, uniform layer of fissionable material, integrally bonded to its inner walls and a port permitting access to the interior of the vessel. A process is also provided for depositing uranium material on to the inner walls of the vessel. Upon irradiation of the target with neutrons from a nuclear reactor, radioactive fission products, such as molybdenum-99, are formed, and thereafter separated from the target by the introduction of an acidic solution through the port to dissolve the irradiated inner layer. The irradiation and dissolution are thus effected in the same vessel without the necessity of transferring the fissionable material and fission products to a separate chemical reactor. Subsequently, the desired isotopes are extracted and purified. Molybdenum-99 decays to technetium-99m which is a valuable medical diagnostic radioisotope. 3 claims, 1 drawing figure

Fission products collecting devices

International Nuclear Information System (INIS)

Matsumoto, Hiroshi

1979-01-01

Purpose: To enable fission products trap with no contamination to coolants and cover gas by the provision of a fission products trap above the upper part of a nuclear power plant. Constitution: Upon fuel failures in a reactor core, nuclear fission products leak into coolants and move along the flow of the coolants to the coolants above the reactor core. The fission products are collected in a trap container and guided along a pipeline into fission products detector. The fission products detector monitors the concentration of the fission products and opens the downstream valve of the detector when a predetermined concentration of the fission products is detected to introduce the fission products into a waste gas processing device and release them through the exhaust pipe. (Seki, T.)

Transport properties of fission product vapors

International Nuclear Information System (INIS)

Im, K.H.; Ahluwalia, R.K.

1983-07-01

Kinetic theory of gases is used to calculate the transport properties of fission product vapors in a steam and hydrogen environment. Provided in tabular form is diffusivity of steam and hydrogen, viscosity and thermal conductivity of the gaseous mixture, and diffusivity of cesium iodide, cesium hydroxide, diatomic tellurium and tellurium dioxide. These transport properties are required in determining the thermal-hydraulics of and fission product transport in light water reactors

Characteristics of fission product release from a molten pool

International Nuclear Information System (INIS)

Yun, J.I.; Suh, K.Y.; Kang, C.S.

2001-01-01

The volatile fission products are released from the debris pool, while the less volatile fission products tend to remain as condensed phases because of their low vapor pressure. The release of noble gases and the volatile fission products is dominated by bubble dynamics. The release of the less volatile fission products from the pool can be analyzed based on mass transport through a liquid with the convection flow. The physico-numerical models were orchestrated from existing submodels in various disciplines of engineering to estimate the released fraction of fission products from a molten pool. It was assumed that the pool has partially filled hemispherical geometry. For the high pool pressure, the diameter of the bubbles at detachment was calculated utilizing the Cole and Shulman correlation with the effect of system pressure. Sensitivity analyses were performed and results of the numerical calculations were compared with analysis results for the TMI-2 accident. (author)

Deposition of aerosols formed by HCDA due to decay heat transport in inner containment atmospheres

International Nuclear Information System (INIS)

Vate, J.F. van de

1976-01-01

Coupling of decay heat transfer by aerosol-laden inner containment atmospheres with aerosol deposition from such atmospheres leads to useful and simple models for calculation of the time dependence of the aerosol mass concentration. Special attention is given to thermophoretic deposition (dry case) and condensation followed by gravitational deposition (wet case). Attractive features of the models are: 1) coagulation can be omitted and therefore complicated and doubtful calculations on coagulation are avoided, 2) material and particle size of the aerosol are not important for the aerosol decay rate, 3) the aerosol decay rate is related to the decay heat production which is known function of time, and the relevant part of it must be assessed usually for other purposes as well. (orig.) [de

Fission-product SiC reaction in HTGR fuel

International Nuclear Information System (INIS)

Montgomery, F.

1981-01-01

The primary barrier to release of fission product from any of the fuel types into the primary circuit of the HTGR are the coatings on the fuel particles. Both pyrolytic carbon and silicon carbide coatings are very effective in retaining fission gases under normal operating conditions. One of the possible performance limitations which has been observed in irradiation tests of TRISO fuel is chemical interaction of the SiC layer with fission products. This reaction reduces the thickness of the SiC layer in TRISO particles and can lead to release of fission products from the particles if the SiC layer is completely penetrated. The experimental section of this report describes the results of work at General Atomic concerning the reaction of fission products with silicon carbide. The discussion section describes data obtained by various laboratories and includes (1) a description of the fission products which have been found to react with SiC; (2) a description of the kinetics of silicon carbide thinning caused by fission product reaction during out-of-pile thermal gradient heating and the application of these kinetics to in-pile irradiation; and (3) a comparison of silicon carbide thinning in LEU and HEU fuels

Vitrification processes for fission product solutions

International Nuclear Information System (INIS)

Bonniaud, R.; Jouan, A.; Moncouyoux, J.P.; Sombret, C.

1982-10-01

The different processes for fission product vitrification in the world are reviewed. Continuous or discontinuous processes, induction or arc heating, in can melting or casting, tests with radioactive or simulated wastes and industrial realizations are described [fr

Cluster expression in fission and fusion in high-dimensional macroscopic-microscopic calculations

International Nuclear Information System (INIS)

Iwamoto, Akira; Ichikawa, Takatoshi; Moller, Peter; Sierk, Arnold J.

2004-01-01

We discuss the relation between the fission-fusion potential-energy surfaces of very heavy nuclei and the formation process of these nuclei in cold-fusion reactions. In the potential-energy surfaces, we find a pronounced valley structure, with one valley corresponding to the cold-fusion reaction, the other to fission. As the touching point is approached in the cold-fusion entrance channel, an instability towards dynamical deformation of the projectile occurs, which enhances the fusion cross section. These two 'cluster effects' enhance the production of superheavy nuclei in cold-fusion reactions, in addition to the effect of the low compound-system excitation energy in these reactions. Heavy-ion fusion reactions have been used extensively to synthesize heavy elements beyond actinide nuclei. In order to proceed further in this direction, we need to understand the formation process more precisely, not just the decay process. The dynamics of the formation process are considerably more complex than the dynamics necessary to interpret the spontaneous-fission decay of heavy elements. However, before implementing a full dynamical description it is useful to understand the basic properties of the potential-energy landscape encountered in the initial stages of the collision. The collision process and entrance-channel landscape can conveniently be separated into two parts, namely the early-stage separated system before touching and the late-stage composite system after touching. The transition between these two stages is particularly important, but not very well understood until now. To understand better the transition between the two stages we analyze here in detail the potential energy landscape or 'collision surface' of the system both outside and inside the touching configuration of the target and projectile. In Sec. 2, we discuss calculated five-dimensional potential-energy landscapes inside touching and identify major features. In Sec. 3, we present calculated

Fission-product yields for thermal-neutron fission of 243Cm determined from measurements with a high-resolution low-energy germanium gamma-ray detector

International Nuclear Information System (INIS)

Merriman, L.D.

1984-04-01

Cumulative fission-product yields have been determined for 13 gamma rays emitted during the decay of 12 fission products created by thermal-neutron fission of 243 Cm. A high-resolution low-energy germanium detector was used to measure the pulse-height spectra of gamma rays emitted from a 77-nanogram sample of 243 Cm after the sample had been irradiated by thermal neutrons. Analysis of the data resulted in the identification and matching of gamma-ray energies and half-lives to individual radioisotopes. From these results, 12 cumulative fission product yields were deduced for radionuclides with half-lives between 4.2 min and 84.2 min. 7 references

HTR fuel modelling with the ATLAS code. Thermal mechanical behaviour and fission product release assessment

International Nuclear Information System (INIS)

Guillermier, Pierre; Daniel, Lucile; Gauthier, Laurent

2009-01-01

To support AREVA NP in its design on HTR reactor and its HTR fuel R and D program, the Commissariat a l'Energie Atomique developed the ATLAS code (Advanced Thermal mechanicaL Analysis Software) with the objectives: - to quantify, with a statistical approach, the failed particle fraction and fission product release of a HTR fuel core under normal and accidental conditions (compact or pebble design). - to simulate irradiation tests or benchmark in order to compare measurements or others code results with ATLAS evaluation. These two objectives aim at qualifying the code in order to predict fuel behaviour and to design fuel according to core performance and safety requirements. A statistical calculation uses numerous deterministic calculations. The finite element method is used for these deterministic calculations, in order to be able to choose among three types of meshes, depending on what must be simulated: - One-dimensional calculation of one single particle, for intact particles or particles with fully debonded layers. - Two-dimensional calculations of one single particle, in the case of particles which are cracked, partially debonded or shaped in various ways. - Three-dimensional calculations of a whole compact slice, in order to simulate the interactions between the particles, the thermal gradient and the transport of fission products up to the coolant. - Some calculations of a whole pebble, using homogenization methods are being studied. The temperatures, displacements, stresses, strains and fission product concentrations are calculated on each mesh of the model. Statistical calculations are done using these results, taking into account ceramic failure mode, but also fabrication tolerances and material property uncertainties, variations of the loads (fluence, temperature, burn-up) and core data parameters. The statistical method used in ATLAS is the importance sampling. The model of migration of long-lived fission products in the coated particle and more

Analysis of effects of updated decay and fission yield data on ORIGEN 2 results

International Nuclear Information System (INIS)

Daniel, P.R.

1993-01-01

Work has been performed to improve the accuracy of ORIGEN2 results by updating both the decay library and the fission yield data in the cross-section library. This effort was performed under the auspices of Oak Ridge National Laboratory (ORNL) to ensure that ORIGEN2 uses the most up-to-date data. The impact of the new data was then quantitatively evaluated by solving a set of standard light water reactor (LWR) problems solved with ORIGEN2. The ORIGEN code, developed at ORNL in the late 1960's, is a point depletion code used to determine the composition and characteristics of spent fuel. The results from calculations performed with the code often form the basis for the study and design of reprocessing plants, spent-fuel shipping casks, waste treatment systems, and disposal facilities. The decay data were updated using data from ENDF/B-VI; fission yield data were updated using data from ENDF/B-V. The impact of these new data was then evaluated

Nuclear decay data for dosimetry calculation. Revised data of ICRP Publication 38

International Nuclear Information System (INIS)

Endo, Akira; Yamaguchi, Yasuhiro

2005-02-01

New nuclear decay data used for dose calculation have been compiled for 1034 radionuclides, which are significant in medical, environmental and occupational exposures. The decay data were assembled from decay data sets of the Evaluated Nuclear Structure Data File (ENSDF), the latest version as of 2003. Basic nuclear properties in the ENSDF that are particularly important for calculating energies and intensities of radiations were examined and updated by referring to UNBASE2003/AME2003, the database for nuclear and decay properties of nuclides. In addition, modification of incomplete ENSDF was done for their format errors, level schemes, normalization records, and so on. The energies and intensities of emitted radiations by the nuclear decay and the subsequent atomic process were computed from the ENSDF using the computer code EDISTR04. EDISTR04 is an enhanced version of EDISTR used for assembling ICRP Publication 38 (ICRP38), and incorporates updates of atomic data and computation methods for calculating atomic radiations and spontaneous fission radiations. Quality assurance of the compiled data has been made by comparisons with various experimental data and decay databases prepared from different computer codes and data libraries. A package of the data files, called DECDC2 (Nuclear DECay Data for Dosimetry Calculation, Version 2), will succeed ICRP38 that has been used extensively in dose calculation and will be utilized in various fields. (author)

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)

Decay and Transmutation of Nuclides

CERN Document Server

Aarnio, Pertti A

1999-01-01

We present a computer code DeTra which solves analytically the Bateman equations governing the decay, build-up and transmutation of radionuclides. The complexity of the chains and the number of nuclides are not limited. The nuclide production terms considered include transmutation of the nuclides inside the chain, external production, and fission. Time dependent calculations are possible since all the production terms can be re-defined for each irradiation step. The number of irradiation steps and output times is unlimited. DeTra is thus able to solve any decay and transmutation problem as long as the nuclear data i.e. decay data and production rates, or cross sections, are known.

Calculated model of radioactive fission and corrosion product accumulation and distribution in a fast reactor sodium coolant circuit

International Nuclear Information System (INIS)

Kizin, V.D.; Konyashov, V.V.

1987-01-01

A simple calculation procedure of radioactive products accumulation and distribution in a primary circuit has been developed on the basis of experimental investigations at the BOR-60 reactor. Common knowledge on the impurity products transfer at the liquid-solid and liquid-gas phase boundary is taken. Use is made of the typical in reactor physics relationships for the description of the products transition to the equipment surfaces, of fission products release, metal corrosion and others. Satisfactory agreement of the calculation data with the experimental ones has been obtained. (orig.)

Studies on the separation of rare earth elements and the nuclear decay properties of short lived rare-earth nuclides in U-235 fission products

International Nuclear Information System (INIS)

Ohyoshi, Akira; Ohyoshi, Emiko.

1980-01-01

The effect of a complex-forming agent, with which rare earths consecutively form the complexes, on the separation of a pair of adjacent rare earths by electromigration has been investigated. The relation between the separation factor for two complexes and the ligand-ion concentration was examined in the separation of La-Ce and Ce-Pr pairs with nitrilotriacetic acid. Rare earths were able to be isolated rapidly at the optimum ligand-ion concentration in lower one, and this method was applied to study the nuclear decay properties of the short lived isotopes of La, Ce, Pr, Nd and Yt formed in the fission of U-235. This method permits the direct measurement of the decay of La-144 without the interference from the radiation of other fission products. The gamma-ray spectrum of La-144 was measured with a high resolution Ge(Li) detector, and the gamma-transition was observed. From the decay plots of two strong photopeaks, the half-life of La-144 was determined. In the case of Ce fraction, the photopeaks assigned to respective isotopes were observed. In the studies on the decay properties of Pr-148 and Pr-149, the decay plot of the strong photopeak showed good linearity, and the accurate half-life of Pr-148 was determined. Similarly, the half-life of Pr-149 was longer than the previously reported value. (Kako, I.)

HLW disposal by fission reactors; calculation of trans-mutation rate and recycle

International Nuclear Information System (INIS)

Mulyanto

1997-01-01

Transmutation of MA (Minor actinide) and LLFPS (long-lived fission products) into stable nuclide or short-lived isotopes by fission reactors seem to become an alternative technology for HLW disposal. in this study, transmutation rate and recycle calculation were developed in order to evaluate transmutation characteristics of MA and LLFPs in the fission reactors. inventory of MA and LLFPs in the transmutation reactors were determined by solving of criticality equation with 1-D cylindrical geometry of multigroup diffusion equations at the beginning of cycle (BOC). transmutation rate and burn-up was determined by solving of depletion equation. inventory of MA and LLFPs was calculated for 40 years recycle. From this study, it was concluded that characteristics of MA and LLFPs in the transmutation reactors can be evaluated by recycle calculation. by calculation of transmutation rate, performance of fission reactor for transmutation of MA or LLFPs can be discussed

Calculations of thermal-reactor spent-fuel nuclide inventories and comparisons with measurements

International Nuclear Information System (INIS)

Wilson, W.B.; LaBauve, R.J.; England, T.R.

1982-01-01

Comparisons with integral measurements have demonstrated the accuracy of CINDER codes and libraries in calculating aggregate fission-product properties, including neutron absorption, decay power, and decay spectra. CINDER calculations have, alternatively, been used to supplement measured integral data describing fission-product decay power and decay spectra. Because of the incorporation of the extensive actinide library and the use of ENDF/B-V data, it is desirable to compare the inventory of individual nuclides obtained from tandem EPRI-CELL/CINDER-2 calculations with those determined in documented benchmark inventory measurements of spent reactor fuel. The development of the popular 148 Nd burnup measurement procedure is outlined, and areas of uncertainty in it and lack of clarity in its interpretation are indicated. Six inventory samples of varying quality and completeness are examined. The power histories used in the calculations have been listed for other users

Needs and accuracy requirements for fission product nuclear data in the physics design of power reactor cores

International Nuclear Information System (INIS)

Rowlands, J.L.

1978-01-01

The fission product nuclear data accuracy requirements for fast and thermal reactor core performance predictions were reviewed by Tyror at the Bologna FPND Meeting. The status of the data was assessed at the Meeting and it was concluded that the requirements of thermal reactors were largely met, and the yield data requirements of fast reactors, but not the cross section requirements, were met. However, the World Request List for Nuclear Data (WRENDA) contains a number of requests for fission product capture cross sections in the energy range of interest for thermal reactors. Recent reports indicate that the fast reactor reactivity requirements might have been met by integral measurements made in zero power critical assemblies. However, there are requests for the differential cross sections of the individual isotopes to be determined in addition to the integral data requirements. The fast reactor requirements are reviewed, taking into account some more recent studies of the effects of fission products. The sodium void reactivity effect depends on the fission product cross sections in a different way to the fission product reactivity effect in a normal core. This requirement might call for different types of measurement. There is currently an interest in high burnup fuel cycles and alternative fuel cycles. These might require more accurate fission product data, data for individual isotopes and data for capture products. Recent calculations of the time dependence of fission product reactivity effects show that this is dependent upon the data set used and there are significant uncertainties. Some recent thermal reactor studies on approximations in the treatment of decay chains and the importance of xenon and samarium poisoning are also summarized. (author)

Delayed fission

Energy Technology Data Exchange (ETDEWEB)

Hatsukawa, Yuichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1997-07-01

Delayed fission is a nuclear decay process that couples {beta} decay and fission. In the delayed fission process, a parent nucleus undergoes {beta} decay and thereby populates excited states in the daughter. If these states are of energies comparable to or greater than the fission barrier of the daughter, then fission may compete with other decay modes of the excited states in the daughter. In this paper, mechanism and some experiments of the delayed fission will be discussed. (author)

ORIGEN-S: scale system module to calculate fuel depletion, actinide transmutation, fission product buildup and decay, and associated radiation source terms

International Nuclear Information System (INIS)

Anon.

1982-01-01

ORIGEN-S computes time-dependent concentrations and source terms of a large number of isotopes, which are simultaneously generated or depleted through neutronic transmutation, fission, radioactive decay, input feet rates and physical or chemical removal rates. The calculations may pertain to fuel irradiation within nuclear reactors, or the storage, management, transportation or subsequent chemical processing of removed fuel elements. The matrix exponential expansion model of the ORIGIN code is unaltered in ORIGEN-S. Essentially all features of ORIGEN were retained, expanded or supplemented within new computations. The primary objective of ORIGEN-S, as requested by the Nuclear Regulatory Commission, is that the calculations may utilize the multi-energy group cross sections from any currently processed standardized ENDF/B data base. This purpose has been implemented through the prior execution of codes within either the SCALE System or the AMPX System, developed at the Oak Ridge National Laboratory. These codes compute flux-weighted cross sections, simulating conditions within any given reactor fuel assembly, and convert the data into a library that can be input to ORIGEN-S. Time-dependent libraries may be produced, reflecting fuel composition variations during irradiation. Presented in the document are: detailed and condensed input instructions, model theory, features available, range of applicability, brief subroutine descriptions, sample input, and I/O requirements. Presently the code is operable on IBM 360/370 computers and may be converted for CDC computers. ORIGEN-S is a functional module in the SCALE System and will be one of the modules invoked in the SAS2 Control Module, presently being developed, or may be applied as a stand alone program. It can be used in nuclear reactor and processing plant design studies, radiation safety analyses, and environmental assessments

Decay Power Calculation for Safety Analysis of Innovative Reactor Systems

International Nuclear Information System (INIS)

Shwageraus, E.; Fridman, E.

2008-01-01

In this work, we verified the decay heat calculation capabilities of BGCore computer code system developed recently at Ben-Gurion University. Decay power was calculated for a typical UO 2 fuel in Pressurized Water Reactor environment using BGCore code and using procedure prescribed by the ANS/ANSI-2005 standard. Very good agreement between the two methods was obtained. Once BGCore calculation capabilities were verified, we calculated decay power as a function of time after shutdown for various reactors with innovative fuels, for which no standard procedure is currently available. Notable differences were observed for decay power of the advanced reactors as compared with conventional UO 2 LWR. The observed differences suggest that the design of new reactors safety systems must be based on corresponding decay power curves for each individual case in order to assure the desired performance of such systems. (authors)

Nuclear fission and neutron-induced fission cross-sections

Energy Technology Data Exchange (ETDEWEB)

James, G.D.; Lynn, J.E.; Michaudon, A.; Rowlands, J.; de Saussure, G.

1981-01-01

A general presentation of current knowledge of the fission process is given with emphasis on the low energy fission of actinide nuclei and neutron induced fission. The need for and the required accuracy of fission cross section data in nuclear energy programs are discussed. A summary is given of the steps involved in fission cross section measurement and the range of available techniques. Methods of fission detection are described with emphasis on energy dependent changed and detector efficiency. Examples of cross section measurements are given and data reduction is discussed. The calculation of fission cross sections is discussed and relevant nuclear theory including the formation and decay of compound nuclei and energy level density is introduced. A description of a practical computation of fission cross sections is given.

FAKIR: a user-friendly standard for decay heat and activity calculation of LWR fuel

International Nuclear Information System (INIS)

Pretesacque, P.; Nimal, J.C.; Huynh, T.D.; Zachar, M.

1993-01-01

The shipping casks owned by the transporters and the unloading and storage facilities are subjected by their design safety report to decay heat and activity limits. It is the responsibility of the consignor or the consignee to check the compliance of the fuel assemblies to the shipped or stored with regard to these limiting safety parameters. Considering the diversity of the parties involved in the transport and storage cycle, a standardization has become necessary. This has been achieved by the FAKIR code. The FAKIR development started in 1984 in collaboration between COGEMA, CEA-SERMA and NTL. Its main specifications were to be a user-friendly code, to use the contractual data given in the COGEMA transport and reprocessing sheet 1 as input, and to over-estimate decay heat and activity. Originally based on computerizable standards such as ANSI or USNRC, the FAKIR equations and data libraries are now based on the fully qualified PEPIN/APOLLO calculation codes. FAKIR is applicable to all patterns of irradiation histories, with burn up from 1000 MWd/TeU to 70.000 MWd/TeU and cooling times from 1 second to 100 years. (J.P.N.)

Yields of fission products produced by thermal-neutron fission of 249Cf

International Nuclear Information System (INIS)

Dickens, J.K.; McConnell, J.W.

1981-01-01

Absolute yields have been determined for 107 gamma rays emitted in the decay of 97 fission products representing 54 mass chains created during thermal-neutron fission of 249 Cf. These results include 14 mass chains for which no prior yield data exist. Using a Ge(Li) detector, spectra were obtained of gamma rays emanating from a 0.4 μg sample of 249 Cf between 45 s and 0.4 yr after very short irradiations of the 249 Cf by thermal neutrons. On the basis of measured gamma-ray yields and known nuclear data, total chain mass yields and relative uncertainties were obtained for 51 masses between 89 and 156. The absolute overall normalization uncertainty is approx.8%. The measured A-chain cumulative yields make up 77% of the total light mass (A 249 Cf

Compilation of nuclear decay data used for dose calculation. Revised data for radionuclides listed in ICRP Publication 38

International Nuclear Information System (INIS)

Endo, Akira; Yamaguchi, Yasuhiro

2001-03-01

New nuclear decay data used for dose calculation have been compiled for 817 radionuclides that are listed in ICRP Publication 38 (Publ. 38) and for 6 additional isomers. The decay data were prepared using decay data sets from the Evaluated Nuclear Structure Data File (ENSDF), the latest version in August 1997. Basic nuclear properties in the decay data sets that are particularly important for calculating energies and intensities of emissions were examined and updated by referring to NUBASE, the database for nuclear and decay properties of nuclides. The reviewed and updated data were half-life, decay mode and its branching ratio, spin and parity of the ground and isomeric states, excitation energy of isomers, and Q value. In addition, possible revisions of partial and incomplete decay data sets were done for their format and syntax errors, level schemes, normalization records, and so on. After that, the decay data sets were processed by EDISTR in order to compute the energies and intensities of α particles, β particles, γ rays, internal conversion electrons, X rays, and Auger electrons emitted in nuclear transformation. For spontaneously fissioning nuclides, the average energies and intensities of neutrons, fission fragments, prompt γ rays, delayed γ rays, and β particles were also calculated. The compiled data were prepared in two different types of format: Publ. 38 and NUCDECAY formats. Comparison of the compiled decay data with those in Publ. 38 was also presented. The decay data will be widely used for internal and external dose calculations in radiation protection and will be beneficial to a future revision of ICRP Publ. 38. (author)

Compilation of nuclear decay data used for dose calculation. Revised data for radionuclides listed in ICRP Publication 38

Energy Technology Data Exchange (ETDEWEB)

Endo, Akira; Yamaguchi, Yasuhiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-03-01

New nuclear decay data used for dose calculation have been compiled for 817 radionuclides that are listed in ICRP Publication 38 (Publ. 38) and for 6 additional isomers. The decay data were prepared using decay data sets from the Evaluated Nuclear Structure Data File (ENSDF), the latest version in August 1997. Basic nuclear properties in the decay data sets that are particularly important for calculating energies and intensities of emissions were examined and updated by referring to NUBASE, the database for nuclear and decay properties of nuclides. The reviewed and updated data were half-life, decay mode and its branching ratio, spin and parity of the ground and isomeric states, excitation energy of isomers, and Q value. In addition, possible revisions of partial and incomplete decay data sets were done for their format and syntax errors, level schemes, normalization records, and so on. After that, the decay data sets were processed by EDISTR in order to compute the energies and intensities of {alpha} particles, {beta} particles, {gamma} rays, internal conversion electrons, X rays, and Auger electrons emitted in nuclear transformation. For spontaneously fissioning nuclides, the average energies and intensities of neutrons, fission fragments, prompt {gamma} rays, delayed {gamma} rays, and {beta} particles were also calculated. The compiled data were prepared in two different types of format: Publ. 38 and NUCDECAY formats. Comparison of the compiled decay data with those in Publ. 38 was also presented. The decay data will be widely used for internal and external dose calculations in radiation protection and will be beneficial to a future revision of ICRP Publ. 38. (author)

Evaluation of fission-product gases in program GAPCON series and FREG-3 to estimate the gap heat transfer coefficient

International Nuclear Information System (INIS)

Ohki, Naohisa; Harayama, Yasuo; Takeda, Tsuneo; Izumi, Fumio.

1977-12-01

In safety evaluation of a fuel rod, estimation of the stored energy in the fuel rod is indispensable. For this estimation, the temperature distribution in the fuel rod is calculated. Most important in determination of the temperature distribution is the gap heat transfer coefficient (gap conductance) between pellet surface and cladding inner surface. Under fuel rod operating condition, the mixed gas in the gap is composed of He, Xe and Kr. He is initial seald gas. Xe and Kr are fission-product gases, of which the quantities depend on the fuel burn-up. In program GAPCON series (GAPCON and GAPCON-THERMAL-1 and -2) and FREG-3, these quantities are given as a function of the irradiation time, power rating and neutron flux in estimation of the thermal conductivity of the mixed gas. The methods of calculating the quantities of Xe and Kr in the programs have been examined. Input of the neutron flux which influences F.P. gas production rates is better than the determination from the fuel-rod power rating. (auth.)

Refinements to temperature calculations of spent fuel assemblies when in a stagnant gas environment

International Nuclear Information System (INIS)

Rhodes, C.A.; Haire, M.J.

1984-01-01

Undesirably high temperatures are possible in irradiated fuel assemblies because of the radioactive decay of fission products formed while in the reactor. The COXPRO computer code has been used for some time to calculate temperatures in spent fuel when the fuel is suspended in a stagnant gas environment. This code assumed radiation to be the only mode of heat dissipation within the fuel pin bundle. Refinements have been made to include conduction as well as radiation heat transfer within this code. Comparison of calculated and measured temperatures in four separate and independent tests indicate that maximum fuel assembly temperatures can be predicted to within about 6%. 2 references, 5 figures

Decay Power Calculation for Safety Analysis of Innovative Reactor Systems

Energy Technology Data Exchange (ETDEWEB)

Shwageraus, E.; Fridman, E. [Department of Nuclear Engineering, Ben-Gurion University of the Negev Beer-Sheva 84105 (Israel)

2008-07-01

In this work, we verified the decay heat calculation capabilities of BGCore computer code system developed recently at Ben-Gurion University. Decay power was calculated for a typical UO{sub 2} fuel in Pressurized Water Reactor environment using BGCore code and using procedure prescribed by the ANS/ANSI-2005 standard. Very good agreement between the two methods was obtained. Once BGCore calculation capabilities were verified, we calculated decay power as a function of time after shutdown for various reactors with innovative fuels, for which no standard procedure is currently available. Notable differences were observed for decay power of the advanced reactors as compared with conventional UO{sub 2} LWR. The observed differences suggest that the design of new reactors safety systems must be based on corresponding decay power curves for each individual case in order to assure the desired performance of such systems. (authors)

Recoil release of fission products from nuclear fuel

International Nuclear Information System (INIS)

Wise, C.

1985-01-01

An analytical approximation is developed for calculating recoil release from nuclear fuel into gas filled interspaces. This expression is evaluated for a number of interspace geometries and shown to be generally accurate to within about 10% by comparison with numerical calculations. The results are applied to situations of physical interest and it is demonstrated that recoil can be important when modelling fission product release from low temperature CAGR pin failures. Furthermore, recoil can contribute significantly in experiments on low temperature fission product release, particularly where oxidation enhancement of this release is measured by exposing the fuel to CO 2 . The calculations presented here are one way of allowing for this, other methods are suggested. (orig.)

An Evaluation of a Fission Product Inventory for CANDU Fuels

International Nuclear Information System (INIS)

Jung, Jong Yeob; Park, Joo Hwan

2007-01-01

Fission products are released by two processes when a single channel accident occurs. One is a 'prompt release' and the other is a 'delayed release'. Prompt release assumes that the gap inventory of the fuel elements is released by a fuel element failure at the time of an accident. Delayed release assumes that the inventories within the grain or at the grain boundary are released after a accident due to a diffusion through grains, an oxidation of the fuel and an interaction between the fuel and the Zircaloy sheath. Therefore, the calculation of a fission product inventory and its distribution in a fuel during a normal operating is the starting point for the assessment of a fission product release for single channel accidents. In this report, the fission product inventories and their distributions within s fuel under a normal operating condition are evaluated for three types of CANDU fuels such as the 37 element fuel, CANFLEX-NU and CANFLEX-RU fuel bundles in the 'limiting channel'. To accomplish the above mentioned purposes, the basic power histories for each type of CANDU fuel were produced and the fission product inventories were calculated by using the ELESTRES code

The file of evaluated decay data in ENDF/B

International Nuclear Information System (INIS)

Reich, C.W.

1991-01-01

One important application of nuclear decay data is the Evaluated Nuclear Data File/B (ENDF/B), the base of evaluated nuclear data used in reactor research and technology activities within the United States. The decay data in the Activation File (158 nuclides) and the Actinide File (108 nuclides) excellently represent the current status of this information. In particular, the half-lives and gamma and alpha emission probabilities, quantities that are so important for many applications, of the actinide nuclides represent a significant improvement over those in ENDF/B-V because of the inclusion of data produced by an International Atomic Energy Agency Coordinated Research Program. The Fission Product File contains experimental decay data on ∼510 nuclides, which is essentially all for which a meaningful number of data are available. For the first time, delayed-neutron spectra for the precursor nuclides are included. Some hint of problems in the fission product data base is provided by the gamma decay heat following a burst irradiation of 239 Pu

On the calculation of multi-group fission spectrum vectors

International Nuclear Information System (INIS)

Mueller, E.Z.

1984-05-01

In this report, the problem of calculating fission spectrum vectors in a consistent manner is formulated. The practical implications of using fission spectrum vectors in multi-group transport calculations are also addressed. The significance of the weighting spectra used for the calculation of fission spectrum vectors is illustrated for the case of a simple neutronic assembly

Modelling of fission product release behavior from HTR spherical fuel elements under accident conditions

International Nuclear Information System (INIS)

Verfondern, K.; Mueller, D.

1991-01-01

Computer codes for modelling the fission product release behavior of spherical fuel elements for High Temperature Reactors (HTR) have been developed for the purpose of being used in risk analyses for HTRs. An important part of the validation and verification procedure for these calculation models is the theoretical investigation of accident simulation experiments which have been conducted in the KueFA test facility in the Hot Cells at KFA. The paper gives a presentation of the basic modeling and the calculational results of fission product release from modern German HTR fuel elements in the temperature range 1600-1800 deg. C using the TRISO coated particle failure model PANAMA and the diffusion model FRESCO. Measurements of the transient release behavior for cesium and strontium and of their concentration profiles after heating have provided informations about diffusion data in the important retention barriers of the fuel: silicon carbide and matrix graphite. It could be shown that the diffusion coefficients of both cesium and strontium in silicon carbide can significantly be reduced using a factor in the range of 0.02 - 0.15 compared to older HTR fuel. Also in the development of fuel element graphite, a tendency towards lower diffusion coefficients for both nuclides can be derived. Special heating tests focussing on the fission gases and iodine release from the matrix contamination have been evaluated to derive corresponding effective diffusion data for iodine in fuel element graphite which are more realistic than the iodine transport data used so far. Finally, a prediction of krypton and cesium release from spherical fuel elements under heating conditions will be given for fuel elements which at present are irradiated in the FRJ2, Juelich, and which are intended to be heated at 1600/1800 deg. C in the KueFA furnace in near future. (author). 7 refs, 11 figs

Fission decay properties of ultra neutron-rich uranium isotopes

Indian Academy of Sciences (India)

in the chain of neutron-rich uranium isotopes is examined here. The neutron ... mean field theory, this nucleus is shown to undergo exotic decay mode of thermal neu- .... For 250U with a fission barrier of 4.3 MeV [5], we obtain the value of.

Chemical reactions of fission products with ethylene using the gas jet technique

International Nuclear Information System (INIS)

Contis, E.T.; Rengan, Krish; Griffin, Henry C.

1994-01-01

An understanding of the nature of the chemical reactions taking place between fission products and their carrier gases, and the designing of a fast separation procedure were the purposes of this investigation. Chemical reactions of short-lived (less than one minute half-life) fission products with carrier gases lead to various chemical species which can be separated in the gas phase. The Gas Jet Facility at the Ford Nuclear Reactor was used to study the yields of volatile selenium and bromine fission products of 235 U using a semi-automatic batch solvent extraction technique. Heptane and water were used as organic and inorganic solvents. A carrier gas mixture of ethylene to pre-purified nitrogen (1 : 3) was used to sweep the fission products from the target to the chemistry area for analysis. The results indicated that the volatile selenium products generated by the interaction of selenium fission fragments with ethylene were predominantly organic in nature (84%), possibly organoselenides. The selenium values were used to resolve the fractions of the bromine nuclides, which come from two major sources, viz., directly from fission and from the beta-decay of selenium. The data showed that the fractions of independent bromine fission products in the organic phase were much lower compared to selenium; the bromine values range from 10 to 22% and varied with mass number. Results indicated that the bromine products were inorganic in nature, as possibly hydrogen chloride. ((orig.))

A proposed Regulatory Guide basis for spent fuel decay heat

International Nuclear Information System (INIS)

Hermann, O.W.; Parks, C.V.; Renier, J.P.

1991-01-01

A proposed revision to Regulatory Guide 3.54, ''Spent Fuel Heat Generation in an Independent Spent Fuel Storage Installation'' has been developed for the US Nuclear Regulatory Commission. The proposed revision includes a data base of decay heat rates calculated as a function of burnup, specific power, cooling time, initial fuel 235 U enrichment and assembly type (i.e., PWR or BWR). Validation of the calculational method was done by comparison with existing measured decay heat rates. Procedures for proper use of the data base, adjustment formulae accounting for effects due to differences in operating history and initial enrichment, and a defensible safety factor were derived. 15 refs., 6 tabs

Study on decay of rare earth nuclei produced by fission

Energy Technology Data Exchange (ETDEWEB)

Kawade, Kiyoshi; Yamamoto, Hiroshi; Shibata, Michihiro; Asai, Masato [Nagoya Univ. (Japan); Tsukada, Kazuaki; Osa, Akihiko; Shinohara, Nobuo; Iimura, Hideki

1996-01-01

JAERI-ISOL utilizes charge particle induced fission by proton and heavy proton produced by the tandem type accelerator (JAERI). To study the decay mechanism and nuclei structure of neutron and excess nuclei produced by actinoid fission, JAERI-ISOL was improved by developing the multilayer target tank. So that, the intensity of mass separated ion beam increased enough to use. New 76.6 KeV {gamma}-ray with about 10s of half life was found in the preliminary experiment. (S.Y.)

Study of the Fission Decay of Heavy Hypernuclei

CERN Multimedia

2002-01-01

The purpose of the original experiment PS177 was to produce heavy hypernuclei using the annihilation at rest of antiprotons in heavy targets, and to measure their lifetime. \\\\ \\\\ Lambda hyperons can be produced, within a nucleus, in a 2-step process: p@*~@A~K&bar.K~+~X; &bar.KN~@A~@L@p; or in a direct 3-body interaction: @*NN~@A~K|+@L. In the first case, the kinematical conditions favour recoilless lambda with, consequently, a higher probability of attachment to the nucleus. In a heavy nucleus the lambda-hyperon decays weakly according to: @LN~@A~NN, and the &prop.170~MeV energy released induces fission.\\\\ \\\\ The identification of the hypernuclei and their lifetime measurements were performed through the detection of delayed fission using the recoil-distance-method (suitable for lifetimes in the expected region @=10|-|1|0s). The fission fragments were detected by parallel-plate avalanche counters. \\\\ \\\\ The new proposal aims at i) increasing the accuracy of the measured lifetimes, ii) having a str...

Release of fission products from irradiated aluminide fuel at high temperature

International Nuclear Information System (INIS)

Shibata, Toshikazu; Kanda, Keiji; Mishima, Kaichiro; Tamai, Tadaharu; Hayashi, Masatoshi; Snelgrove, James L.; Stahl, David; Matos, James E.; Travelli, Armando; Case, F. Neil; Posey, John C.

1983-01-01

Irradiated uranium aluminide fuel plates of 40% U-235 enrichment were heated for the determination of fission products released under flowing helium gas at temperatures up to and higher than the melting point of fuel cladding material. The release of fission products from the fuel plate at temperature below 500 deg. C was found negligible. The first rapid release of fission products was observed with the occurrence of blistering at 561±1 deg. C on the plates. The next release at 585. C might be caused by melting of the cladding material of 6061-Al alloy. The last release of fission product gases was occurred at the eutectic temperature of 640 deg. C of U-Al x . The released material was mostly xenon, but small amounts of iodine and cesium were observed. (author)

Release of fission products from irradiated aluminide fuel at high temperature

International Nuclear Information System (INIS)

Shibata, T.; Kanda, K.; Mishima, K.

1982-01-01

Irradiated uranium aluminide fuel plates of 40% U-235 enrichment were heated for the determination of fission products released under flowing helium gas at temperatures up to and higher than the melting point of fuel-cladding material. The release of fission products from the fuel plate at temperature below 500 0 C was found negligible. The firist rapid release of fission products was observed with the occurrence of blistering at 561 +- 1 0 C on the plates. The next release at 585 0 C might be caused by melting of the cladding material of 6061-Al alloy. The last release of fission product gases was occurred at the eutectic temperature of 640 0 C of U-Al/sub x/. The released material was mostly xenon, but small amounts of iodine and cesium were observed

CRBRP decay heat removal systems

International Nuclear Information System (INIS)

Hottel, R.E.; Louison, R.; Boardman, C.E.; Kiley, M.J.

1977-01-01

The Decay Heat Removal Systems for the Clinch River Breeder Reactor Plant (CRBRP) are designed to adequately remove sensible and decay heat from the reactor following normal shutdown, operational occurrences, and postulated accidents on both a short term and a long term basis. The Decay Heat Removal Systems are composed of the Main Heat Transport System, the Main Condenser and Feedwater System, the Steam Generator Auxiliary Heat Removal System (SGAHRS), and the Direct Heat Removal Service (DHRS). The overall design of the CRBRP Decay Heat Removal Systems and the operation under normal and off-normal conditions is examined. The redundancies of the system design, such as the four decay heat removal paths, the emergency diesel power supplies, and the auxiliary feedwater pumps, and the diversities of the design such as forced circulation/natural circulation and AC Power/DC Power are presented. In addition to overall design and system capabilities, the detailed designs for the Protected Air Cooled Condensers (PACC) and the Air Blast Heat Exchangers (ABHX) are presented

Reactions of newly formed fission products in the gas phase

International Nuclear Information System (INIS)

Strickert, R.G.

1976-01-01

A dynamic gas-flow system was constructed which stopped fission products in the gas phase and rapidly separated (in less than 2 sec) volatile compounds from non-volatile ones. The filter assembly designed and used was shown to stop essentially all non-volatile fission products. Between 5 percent and 20 percent of tellurium fission-product isotopes reacted with several hydrocarbon gases to form volatile compounds, which passed through the filter. With carbon monoxide gas, volatile tellurium compound(s) (probably TeCO) were also formed with similar efficiencies. The upper limits for the yields of volatile compounds formed between CO and tin and antimony fission products were shown to be less than 0.3 percent, so tellurium nuclides, not their precursors, reacted with CO. It was found that CO reacted preferentially with independently produced tellurium atoms; the reaction efficiency of beta-produced atoms was only 27 +- 3 percent of that of the independently formed atoms. The selectivity, which was independent of the over-all reaction efficiency, was shown to be due to reaction of independently formed atoms in the gas phase. The gas phase reactions are believed to occur mainly at thermal energies because of the independence of the yield upon argon moderator mole-fraction (up to 80 percent). It was shown in some experiments that about one-half of the TeCO decomposed in passing through a filter and that an appreciable fraction (approximately 20 percent) of the tellurium atoms deposited on the filter reacted agin with CO. Other tellurium atoms on the filter surface (those formed by beta decay and those formed independently but not reacting in the gas phase) also reacted with CO, but probably somewhat less efficiently than atoms formed by TeCO decomposition. No evidence was found for formation of TeCO as a direct result of beta-decay

Simulating fission product transients via the history-based local-parameter methodology

International Nuclear Information System (INIS)

Jenkins, D.A.; Rouben, B.; Salvatore, M.

1993-01-01

This paper describes the fission-product-calculation capacity of the history-based local-parameter methodology for evaluating lattice properties for use in core-tracking calculations in CANDU reactors. In addition to taking into account the individual past history of each bundles flux/power level, fuel temperature, and coolant density and temperature that the bundle has seen during its stay in the core, the latest refinement of the history-based method provides the capability of fission-product-drivers. It allows the bundle-specific concentrations of the three basic groups of saturating fission products to be calculated in steady state or following a power transient, including long shutdowns. The new capability is illustrated by simulating the startup period following a typical long-shutdown, starting from a snapshot in the Point Lepreau operating history. 9 refs., 7 tabs

The coupled kinetics of grain growth and fission product behavior in nuclear fuel under degraded-core accident conditions

International Nuclear Information System (INIS)

Rest, J.

1985-01-01

The theoretical FASTGRASS-VFP model has been used in the interpretation of fission gas, iodine, and cesium release from (1) irradiated high-burnup LWR fuel in a flowing steam atmosphere during high-temperature, in-cell heating tests (performed at Oak Ridge National Laboratory) and (2) trace-irratiated LWR fuel during severe-fuel-damage (SFD) tests (performed in the PBF reactor in Idaho). A theory of grain boundary sweeping of gas bubbles has been included within the FASTGRASS-VFP formalism. This theory considers the interaction between the moving grain boundary and two distinct size classes of bubbles, those on grain faces and on grain edges, and provides a means of determining whether gas bubbles are caught up and moved along by a moving grain boundary or whether the grain boundary is only temporarily retarded by the bubbles and then breaks away. In addition, as FASTGRASS-VFP provides for a mechanistic calculation of intra- and intergranular fission product behavior, the coupled calculation between fission gas behavior and grain growth is kinetically comprehensive. Results of the analyses demonstrate that intragranular fission product behavior during both types of tests can be interpreted in terms of a grain-growth/grain-boundary-sweeping mechanism that enhances the flow of fission products from within the grains to the grain boundaries. The effect of fuel oxidation by steam on fission product and grain growth behavior is also considered. The FASTGRASS-VFP predictions, measured release rates from the above tests, and previously published release rates are compared and differences between fission product behavior in trace-irradiated and in high-burnup fuel are highlighted. (orig.)

Decay Curves and Half-Lives of Gamma-Emitting States from a Study of Prompt Fission Gamma Radiation

Energy Technology Data Exchange (ETDEWEB)

Albinsson, H [Chalmers Univ. of Technology, Goeteborg (SE)

1971-04-15

Measurements were made on the time distributions of the prompt gamma radiation emitted from fragments in the thermal-neutron induced fission of 235U. The gamma radiation emitted during different time intervals after the fission event was studied with the help of a collimator, the position of which was changed along the path of the fragments. In this way decay curves were obtained from which half-lives could be estimated. Time components with half-lives of 7.5, 18 and 60 ps were found and their relative intensities were calculated. Half-lives and associated intensities are in good agreement with earlier data from uranium and californium fission. Problems involved in this type of study are discussed. The collimator technique has proved to be effective for determination of half lives down to less than 10 ps

Behaviour of short-lived fission products within operating UO2 fuel elements

International Nuclear Information System (INIS)

Hastings, I.J.; Hunt, C.E.L.; Lipsett, J.J.

1983-01-01

We have carried out experiments using a ''sweep gas'' technique to determine the behaviour of short-lived fission products within operating, intact UO 2 fuel elements. The Zircaloy-4-clad elements were 500 mm long and contained fuel of density 10.65-10.71 Mg/m 3 . A He-2% H 2 carrier gas swept gaseous or volatile fission products out of the operating fuel element past a gamma spectrometer for measurement. In tests at linear powers of 45 and 60 kW/m to maximum burnups of 70 MW.h/kg U, the species measured directly at the spectrometer were generally the short-lived xenons and kryptons. We did not observe iodine or bromine during normal operation. However, we have deduced the behaviour of I-133 and I-135 from the decay of Xe-133 and Xe-135 during reactor shutdowns. Plots of R/B (released/born) against lambda (decay constant) or effective lambda for all isotopes observed at 45 and 60 kW/m show that a line of slope -0.5, corresponding with diffusion kinetics, is a good fit to the measured xenon and krypton data. Our inferred release of iodine fits the same line. From this we can extrapolate to an R/B for I-131 of about 5x10 -3 . The ANS 5.4 release correlation gives calculated results in good agreement with our measurements. (author)

Compilation of nuclear decay data used for dose calculations. Data for radionuclides not listed in ICRP publication 38

Energy Technology Data Exchange (ETDEWEB)

Endo, Akira; Yamaguchi, Yasuhiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Tamura, Tsutomu

1999-07-01

Nuclear decay data used for dose calculations were compiled for 162 nuclides with half-lives greater than or equal to 10 min that are not listed in ICRP Publication 38 (Publ. 38) and their 28 daughter nuclides. Additional 14 nuclides that are considered to be important in fusion reactor facilities were also included. The data were compiled using decay data sets of the Evaluated Nuclear Structure Data File (ENSDF), the latest version in August 1997. Investigations of the data sets were performed to check their consistency by referring to recent literature and NUBASE, the database for nuclear and decay properties of nuclides, and by using the utility programs of ENSDF. Possible revisions of the data sets were made for their format and syntax errors, level schemes, normalization records, and so on. The revised data sets were processed by EDISTR in order to calculate the energies and intensities of {alpha} particles, {beta} particles, {gamma} rays including annihilation photons, internal conversion electrons, X rays, and Auger electrons emitted in nuclear transformations of the radionuclides. For spontaneously fissioning nuclides, the average energies and intensities of neutrons, fission fragments, prompt {gamma} rays, delayed {gamma} rays, and {beta} particles were also calculated. The compiled data were presented in two types of format; Publ. 38 and NUCDECAY formats. This report provides the decay data in the Publ. 38 format along with decay scheme drawings. The data will be widely used for internal and external dose calculations in radiation protection. (author)

Trapping technology for gaseous fission products from voloxidation process

International Nuclear Information System (INIS)

Shin, Jin Myeong; Park, J. J.; Park, G. I.; Jung, I. H.; Lee, H. H.; Kim, G. H.; Yang, M. S.

2005-05-01

The objective of this report is to review the different technologies for trapping the gaseous wastes containing Cs, Ru, Tc, 14 C, Kr, Xe, I and 3 H from a voloxidation process. Based on literature reviews and KAERI's experimental results on the gaseous fission products trapping, appropriate trapping method for each fission product has been selected considering process reliability, simplicity, decontamination factor, availability, and disposal. Specifically, the most promising trapping method for each fission product has been proposed for the development of the INL off-gas trapping system. A fly ash filter is proposed as a trapping media for a cesium trapping unit. In addition, a calcium filter is proposed as a trapping media for ruthenium, technetium, and 14 C trapping unit. In case of I trapping unit, AgX is proposed. For Kr and Xe, adsorption on solid is proposed. SDBC (Styrene Divinyl Benzene Copolymer) is also proposed as a conversion media to HTO for 3 H. This report will be used as a useful means for analyzing the known trapping technologies and help selecting the appropriate trapping methods for trapping volatile and semi-volatile fission products, long-lived fission products, and major heat sources generated from a voloxidation process. It can also be used to design an off-gas treatment system

Passive decay heat removal from the core region

International Nuclear Information System (INIS)

Hichen, E.F.; Jaegers, H.

2002-01-01

The decay heat in commercial Light Water Reactors is commonly removed by active and redundant safety systems supported by emergency power. For advanced power plant designs passive safety systems using a natural circulation mode are proposed: several designs are discussed. New experimental data gained with the NOKO and PANDA facilities as well as operational data from the Dodewaard Nuclear Power Plant are presented and compared with new calculations by different codes. In summary, the effectiveness of these passive decay heat removal systems have been demonstrated: original geometries and materials and for the NOKO facility and the Dodewaard Reactor typical thermal-hydraulic inlet and boundary conditions have been used. With several codes a good agreement between calculations and experimental data was achieved. (author)

Detector for gaseous nuclear fission products

International Nuclear Information System (INIS)

Kobayashi, Yoshihiro; Kubo, Katsumi.

1979-01-01

Purpose: To facilitate the fabrication of a precipitator type detector, as well as improve the reliability. Constitution: Gas to be measured flown in an anode is stored in a gas processing system. By applying a voltage between the anode and the cathode, if positively charged Rb or Cs which is the daughter products of gaseous fission products are present in the gas to be measured, the daughter products are successively deposited electrostatically to the cathode. The daughter products issue beta-rays and gamma-rays to ionize the argon gas at the anode, whereby ionizing current flows between both of the electrodes. Pulses are generated from the ionizing current, and presence or absence, as well as the amount of the gaseous fission products are determined by the value recorded for the number of the pulses to thereby detect failures in the nuclear fuel elements. After the completion of the detection, the inside of the anode is evacuated and the cathode is heated to evaporate and discharge the daughter products externally. This eliminates the effects of the former detection to the succeeding detection. (Moriyama, K.)

Fission product release from fuel of water-cooled reactors

International Nuclear Information System (INIS)

Strupczewski, A.; Marks, P.; Klisinska, M.

1997-01-01

The report contains a review of theoretical models and experimental works of gaseous and volatile fission products from uranium dioxide fuel. The experimental results of activity release at low burnup and the model of fission gas behaviour at initial stage of fuel operational cycle are presented. Empirical models as well as measured results of transient fission products release rate in the temperature up to UO 2 melting point, with consideration of their chemical reactions with fuel and cladding, are collected. The theoretical and experimental data were used for calculations of gaseous and volatile fission products release, especially iodine and caesium, to the gas volume of WWER-1000 and WWER-440 type fuel rods at low and high burnup and their further release from defected rods at the assumed loss-of-coolant accident. (author)

Radiation characteristics of spent nuclear fuel at accumulation in long-term storage

International Nuclear Information System (INIS)

Bergelson, Boris R.; Gerasimov, Aleksander S.

1999-01-01

Time dependence of a decay heat power and radiotoxicity of a single spent nuclear fuel unloading of VVER-1000 reactors at its storage or the same characteristics in accumulation mode with annual addition of spent nuclear fuel in long-term storage are investigated. At calculations of decay heat power, the contributions of alpha-, beta-, and gamma- irradiations were taken into account, at calculations of a radiotoxicity - maximum permissible activity of nuclides in air and in water were taken into account. It is determined that at accumulation less than 100 years, the main contribution to decay heat power is given by fission products, at further storage the power is determined in greater degree by actinides. The radiotoxicity of actinides by air is rich greater than that of fission products - more than 50 times in beginning of a storage and by 2-3 orders of magnitude after 100 and more years. A radiotoxicity of fission products by water at accumulation less than 20 years is a little bit more than actinides, at further accumulation the contribution of fission products decreases. At time of accumulation 100 years, the fission products give the contribution in total radiotoxicity about 40%, at time 1000 years - about 7%. (author)

Determination of the fission products yields, lanthanide and yttrium, in the fission of 238U with neutrons of fission spectra

International Nuclear Information System (INIS)

Nicoli, I.G.

1981-06-01

A radiochemical investigation is performed to measure the cumulative fission product yields of several lantanides and yttrium nuclides in the 238 U by fission neutron spectra. Natural and depleted uranium are irradiated under the same experimental conditions in order to find a way to subtract the contribution of the 235 U fission. 235 U percentage in the natural uranium was 3.5 times higher than in the depleted uranium. Uranium oxides samples are irradiated inside the core of the Argonaut Reactor, at the Instituto de Engenharia Nuclear, and the lantanides and yttrium are chemically separated. The fission products gamma activities were detected, counted and analysed in a system constituted by a high resolution Ge(Li) detector, 4096 multichannel analyser and a PDP-11 computer. Cumulative yields for fission products with half-lives between 1 to 33 hours are measured: 93 Y, 141 La, 142 La, 143 Ce and 149 Nd. The chain total yields are calculated. The cumulative fission yields measured for 93 Y, 141 La, 142 La, 143 Ce and 149 Nd are 4,49%, 4,54%, 4,95%, 4,16% and 1,37% respectively and they are in good agreement with the values found in the literature. (Author) [pt

Monte Carlo simulation of fission yields, kinetic energy, fission neutron spectrum and decay γ-ray spectrum for 232Th(n,f) reaction induced by 3H(d,n) 4He neutron source

International Nuclear Information System (INIS)

Zheng Wei; Zeen Yao; Changlin Lan; Yan Yan; Yunjian Shi; Siqi Yan; Jie Wang; Junrun Wang; Jingen Chen; Chinese Academy of Sciences, Shanghai

2015-01-01

Monte Carlo transport code Geant4 has been successfully utilised to study of neutron-induced fission reaction for 232 Th in the transport neutrons generated from 3 H(d,n) 4 He neutron source. The purpose of this work is to examine the applicability of Monte Carlo simulations for the computation of fission reaction process. For this, Monte Carlo simulates and calculates the characteristics of fission reaction process of 232 Th(n,f), such as the fission yields distribution, kinetic energy distribution, fission neutron spectrum and decay γ-ray spectrum. This is the first time to simulate the process of neutron-induced fission reaction using Geant4 code. Typical computational results of neutron-induced fission reaction of 232 Th(n,f) reaction are presented. The computational results are compared with the previous experimental data and evaluated nuclear data to confirm the certain physical process model in Geant4 of scientific rationality. (author)

A passive decay heat removal system for LWRs based on air cooling

Energy Technology Data Exchange (ETDEWEB)

Mochizuki, Hiroyasu, E-mail: mochizki@u-fukui.ac.jp [Research Institute of Nuclear Engineering, University of Fukui, 1-2-4 Kanawa-cho, Tsuruga, Fukui 914-0055 (Japan); Yano, Takahiro [Graduate School of Engineering, University of Fukui, 1-2-4 Kanawa-cho, Tsuruga, Fukui 914-0055 (Japan)

2015-05-15

Highlights: • A passive decay heat removal system for LWRs is discussed. • An air cooler model which condenses steam is developed. • The decay heat can be removed by air coolers with forced convection. • The dimensions of the air cooler are proposed. - Abstract: The present paper describes the capability of an air cooling system (ACS) to remove decay heat from a core of LWR such as an advanced boiling water reactor (ABWR) and a pressurized water reactor (PWR). The motivation of the present research is the Fukushima severe accident (SA) on 11 March 2011. Since emergency cooling systems using electricity were not available due to station blackout (SBO) and malfunctions, many engineers might understand that water cooling was not completely reliable. Therefore, a passive decay heat removal (DHR) system would be proposed in order to prevent such an SA under the conditions of an SBO event. The plant behaviors during the SBO are calculated using the system code NETFLOW++ for the ABWR and PWR with the ACS. Two types of air coolers (ACs) are applied for the ABWR, i.e., a steam condensing air cooler (SCAC) of which intake for heat transfer tubes is provided in the steam region, and single-phase type of which intake is provided in the water region. The DHR characteristics are calculated under the conditions of the forced air circulation and also the natural air convection. As a result of the calculations, the decay heat can be removed safely by the reasonably sized ACS when heat transfer tubes are cooled with the forced air circulation. The heat removal rate per one finned heat transfer tube is evaluated as a function of air flow rate. The heat removal rate increases as a function of the air flow rate.

Calculated fission properties of the heaviest elements

International Nuclear Information System (INIS)

Moeller, P.; Nix, J.R.; Swiatecki, W.J.

1986-09-01

A quantitative calculation is presented that shows where high-kinetic-energy symmetric fission occurs and why it is associated with a sudden and large decrease in fission half-lives. The study is based on calculations of potential-energy surfaces in the macroscopic-microscopic model and a semi-empirical model for the nuclear inertia. For the macroscopic part a Yukawa-plus-exponential model is used and for the microscopic part a folded-Yukawa single-particle potential is used. The three-quadratic-surface parameterization generates shapes for which the potential-energy surfaces are calculated. The use of this parameterization and the use of the finite-range macroscopic model allows for the study of two touching spheres and similar shapes. The results of the calculations in terms of potential-energy surfaces and fission half-lives are presented for heavy even nuclei. The surfaces are displayed in the form of contour diagrams as functions of two moments of the shape. 53 refs., 15 figs., 1 tab

The ratio between the decay heat output and activity content of discharged magnox fuel

International Nuclear Information System (INIS)

Davies, B.S.J.

1977-01-01

Values of the ratio between activity and heat production rate have been calculated for magnox fuel irradiated to 3500 and 8000 MWd.Te -1 and for cooling times of 100, 200 and 500 days. Results are expressed in terms of both MeV.decay -1 and MCi.KW -1 . The results indicate that: for these irradiation and cooling conditions 21 nuclides account for over 99% of the total activity; the calculated values show only small variations with burn-up and cooling time, although the mean energy per decay does fall slightly at 500 days cooling: so for many purposes a median value of 0.63 MeV.decay -1 (0.27 MCi.MW -1 ) may be used; the calculated values have standard deviations ranging from 2.6% at 100 days cooling to 9% at 500 days cooling. (author)

Calculation of fission gases internal pressure in nuclear fuel rods

International Nuclear Information System (INIS)

Vasconcelos Santana, M. de.

1981-12-01

Models concerning the principal phenomena, particularly thermal expansion, fuel swelling, densification, reestructuring, relocation, mechanical strain, fission gas production and release, direct or indirectly important to calculate the internal pressure in nuclear fuel rods were analysed and selected. Through these analyses a computer code was developed to calculate fuel pin internal pressure evolution. Three different models were utilized to calculate the internal pressure in order to select the best and the most conservative estimate. (Author) [pt

Preliminary decay heat calculations for the fuel loaded irradiation loop device of the RMB multipurpose Brazilian reactor

Energy Technology Data Exchange (ETDEWEB)

Campolina, Daniel; Costa, Antonio Carlos L. da; Andrade, Edison P., E-mail: campolina@cdtn.br, E-mail: aclp@cdtn.br, E-mail: epa@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (SETRE/CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores

2017-07-01

The structuring project of the Brazilian Multipurpose Reactor (RMB) is responsible for meeting the capacity to develop and test materials and nuclear fuel for the Brazilian Nuclear Program. An irradiation test device (Loop) capable of performing fuel test for power reactor rods is being conceived for RMB reflector. In this work preliminary neutronic calculations have been carried out in order to determine parameters to the cooling system of the Loop basic design. The heat released as a result of radioactive decay of fuel samples was calculated using ORIGEN-ARP and it resulted less than 200 W after 1 hour of irradiation interruption. (author)

Trapping technology for gaseous fission products from voloxidation process

Energy Technology Data Exchange (ETDEWEB)

Shin, Jin Myeong; Park, J. J.; Park, G. I.; Jung, I. H.; Lee, H. H.; Kim, G. H.; Yang, M. S

2005-05-15

The objective of this report is to review the different technologies for trapping the gaseous wastes containing Cs, Ru, Tc, {sup 14}C, Kr, Xe, I and {sup 3}H from a voloxidation process. Based on literature reviews and KAERI's experimental results on the gaseous fission products trapping, appropriate trapping method for each fission product has been selected considering process reliability, simplicity, decontamination factor, availability, and disposal. Specifically, the most promising trapping method for each fission product has been proposed for the development of the INL off-gas trapping system. A fly ash filter is proposed as a trapping media for a cesium trapping unit. In addition, a calcium filter is proposed as a trapping media for ruthenium, technetium, and {sup 14}C trapping unit. In case of I trapping unit, AgX is proposed. For Kr and Xe, adsorption on solid is proposed. SDBC (Styrene Divinyl Benzene Copolymer) is also proposed as a conversion media to HTO for {sup 3}H. This report will be used as a useful means for analyzing the known trapping technologies and help selecting the appropriate trapping methods for trapping volatile and semi-volatile fission products, long-lived fission products, and major heat sources generated from a voloxidation process. It can also be used to design an off-gas treatment system.

Nondestructive analysis of the RA fuel burnup, Calculation of the gamma activity ratio of fission products in the fuel - program QU0C1

International Nuclear Information System (INIS)

Bulovic, V.F.

1973-01-01

The γ radiation of RA reactor fuel element was measured under precisely defined measuring conditions. The spectrum was analysed by spectrometer with semiconductor Ge(Li) detector. The gamma counting rate in the fuel spectrum is defined as a function of fission product activity, gamma energy and yield, fuel thickness and additional absorbers, dimensions of the gamma collimator. Activity ratio of two fission products is defined as a function of counting rate peaks and part of the mentioned quantities. Four options for calculating the activities for fission products are discussed. Three of them are covered by the QU0C1 code written in FORTRAN for the CDC 3600 computer. The code is included in this report [sr

Development of nuclear decay data library JDDL, and nuclear generation and decay calculation code COMRAD

International Nuclear Information System (INIS)

Naito, Yoshitaka; Ihara, Hitoshi; Katakura, Jun-ichi; Hara, Toshiharu.

1986-08-01

For safety evaluation of nuclear fuel facilities, a nuclear decay data library named JDDL and a computer code COMRAD have been developed to calculate isotopic composition of each nuclide, radiation source intensity, energy spectrum of γ-ray and neutron, and decay heat of spent fuel. JDDL has been produced mainly from the evaluated nuclear data file ENSDF to use new nuclear data. To supplement the data file for short life nuclides, the JNDC data set were also used which had been evaluated by Japan Nuclear Data Committee. Using these data, calculations became possible from short period to long period after irradiation. (author)

FREVAP-6, Metal Fission Products Release from HTGR Fuel Elements

International Nuclear Information System (INIS)

Pierce, V.H.

2005-01-01

1 - Description of problem or function: The FREVAP type of code for estimating the release of longer-lived metallic fission products from HTGR fuel elements has been developed to take into account the combined effects of the retention of metallic fission products by fuel particles and the rather strong absorption of these fission products by the graphite of the fuel elements. Release calculations are made on the basis that the loss of fission product nuclides such as strontium, cesium, and barium is determined by their evaporation from the graphite surfaces and their transpiration induced by the flowing helium coolant. The code is devised so that changes of fission rate (fuel element power), fuel temperature, and graphite temperature may be incorporated into the calculation. Temperature is quite important in determining release because, in general, both release from fuel particles and loss by evaporation (transpiration) vary exponentially with the reciprocal of the absolute temperature. NESC0301/02: This version differs from the previous one in the following points: The source and output files were converted from BCD to ASCII coding. 2 - Method of solution: A problem is defined as having a one-dimensional segment made up of three parts - (1) the fission product source (fuel particles) in series with, (2) a non-source and absorption part (element graphite) and (3) a surface for evaporation to the coolant (graphite-helium interface). More than one segment may be connected (possibly segments stacked axially) by way of the coolant. At any given segment, a continuity equation is solved assuming equilibrium between the source term, absorption term, evaporation at coolant interface and the partial pressure of the fission product isotope in the coolant. 3 - Restrictions on the complexity of the problem - Maxima of: 5 isotopes; 10 time intervals for time-dependent variable; 49 segments (times number of isotopes); 5 different output print time-steps

Statistical model calculations with a double-humped fission barrier GIVAB computer code

International Nuclear Information System (INIS)

Delagrange, H.; Gilat, J.

1979-01-01

Neutron and gamma emission probabilities and fission probabilities are computed, taking into account the special feature of the actinide fission barriers with two maxima. Spectra and cross sections are directly deduced from these probabilities. Populations of both wells are followed step by step. For each initial E and J, decay rates are computed and normalized in order to obtain the de-excitation probabilities imposed by the two-humped fission barrier

Yields of products from thermal-neutron induced fission of 235U

International Nuclear Information System (INIS)

Aagaard, P.; Rudstam, G.; Zwicky, H.U.

1982-01-01

Methods for fission yield determinations at an ISOL-system connected nuclear reactor have been developed. The present report contains detailed descriptions both of the experimental techniques and of the method used to correct the experimental yields for the decay of short-lived nuclear species in the delay between production and measurement. (Authors)

Heated uranium tetrafluoride target system to release non-rare gas fission products for the TRISTAN isotope separator

International Nuclear Information System (INIS)

Gill, R.L.

1977-10-01

Off-line experiments indicated that fluorides of As, Se, Br, Kr, Zr, Nb, Mo, Tc, Ru, Sb, Te, I and Xe could be volatilized, but except for Br, Kr, I and Xe, none of these elements were observed after mass separation in the on-line experiments. The results of the on-line experiments indicated a very low level of hydride contamination at ambient temperature and consequently, uranium tetrafluoride replaced uranyl stearate as the primary gaseous fission product target. Possible reasons for the failure of the heated target system to yield non-rare gas activities are discussed and suggestions for designing a new heated target system are presented

New calculation for the neutron-induced fission cross section of 233Pa between 1.0 and 3.0 MeV

International Nuclear Information System (INIS)

Mesa, J.; Deppman, A.; Likhachev, V.P.; Arruda-Neto, J.D.T.; Manso, M.V.; Garcia, C.E.; Rodriguez, O.; Guzman, F.; Garcia, F.

2003-01-01

The 233 Pa(n,f) cross section, a key ingredient for fast reactors and accelerators driven systems, was measured recently with relatively good accuracy [F. Tovesson et al., Phys. Rev. Lett. 88, 062502 (2002)]. The results are at strong variance with accepted evaluations and an existing indirect experiment. This circumstance led us to perform a quite detailed and complete evaluation of the 233 Pa(n,f) cross section between 1.0 and 3.0 MeV, where use of our newly developed routines for the parametrization of the nuclear surface and the calculation of deformation parameters and level densities (including low-energy discrete levels) were made. The results show good quantitative and excellent qualitative agreement with the experimental direct data obtained by Tovesson et al. [F. Tovesson et al., Phys. Rev. Lett. 88, 062502 (2002)]. Additionally, our methodology opens new possibilities for the analysis of subthreshold fission and above threshold second-chance fission for both 233 Pa and its decay product 233 U, as well as other strategically important fissionable nuclides

The nuclear fission

International Nuclear Information System (INIS)

Fiorentino, J.

1983-01-01

The nuclear fission process considering initially the formation of compound nucleus and finishing with radioactive decay of fission products is studied. The process is divided in three parts which consist of the events associated to the nucleus of intermediate transitional state, the scission configuration, and the phenomenum of post scission. (M.C.K.) [pt

SACHET, Dynamic Fission Products Inventory in PWR Multiple Compartment System

International Nuclear Information System (INIS)

Kodaira, Hideki

1990-01-01

1 - Description of program or function: SACHET evaluates the dynamic fission product inventories in the multiple compartment system of pressurized water reactor (PWR) plants. 2 - Method of solution: SACHET utilizes a matrix of fission product core inventory which is previously calculated by the ORIGEN code. 3 - Restrictions on the complexity of the problem: Liquid wastes such as chemical waste and detergent waste are not included

Calculation of CWKB envelope in boson and fermion productions

International Nuclear Information System (INIS)

Biswas, S.; Chowdhury, I.

2007-01-01

We present the calculation of envelope of boson and of both low-and high-mass fermion production at the end of inflation when the coherently oscillating inflations decay into bosons and fermions. We consider three different models of inflation and use CWKB technique to calculate the envelope to understand the structure of resonance band formation. We observe that though low-mass fermion production is not effective in preheating because of Pauli blocking, it is quite probable for high-mass fermion to take part in pre heating. (author)

Sensitivity analysis of the effect of various key parameters on fission product concentration (mass number 120 to 126)

International Nuclear Information System (INIS)

Sola, A.

1978-01-01

An analytical sensitivity analysis has been made of the effect of various parameters on the evaluation of fission product concentration. Such parameters include cross sections, decay constants, branching ratios, fission yields, flux and time. The formulae are applied to isotopes of the Tin, Antimony and Tellurium series. The agreement between analytically obtained data and that derived from a computer evaluated model is good, suggesting that the analytical representation includes all the important parameters useful to the evaluation of the fission product concentrations

Sensibility analysis of the effect of various key parameters on fission product concentration (Mass Number 133 to 138)

International Nuclear Information System (INIS)

Sola, A.

1978-01-01

An analytical sensitivity analysis has been made of the effect of various parameters on the evaluation of fission product concentration. Such parameters include cross-sections, decay constants, branching ratios, fission yields, flux and time. The formulae are applied to isotopes of the iodine, xenon, caesium and barium series. The agreement between analytically obtained data and that derived from a computer-evaluated model is good, suggesting that the analytical representation includes all the important parameters useful to the evaluation of the fission product concentrations

Final report: Accelerated beta decay for disposal of fission fragment wastes

International Nuclear Information System (INIS)

Reiss, Howard R.

2000-01-01

The fundamental theory of the interaction of intense, low-frequency electromagnetic fields with certain radioactive nuclei has been fully formulated. The nuclei are of the type that exists in high-level radioactive wastes that are end products of the production of energy from nuclear fission. The basic physical mechanisms that underlie the coupling of the applied field to the nucleus have been identified. Both the basic theory and numerical predictions that stem from it support the conclusion that high-level radioactive wastes can be disposed of by substantially accelerating the rate of radioactive decay. Some old experiments on the acceleration of this type of radioactivity, with results that were not understood at the time, have been re-examined. Their interpretation is now clear, and the experiments are found to be in agreement with the theory

Fission Product Release Behavior of Individual Coated Fuel Particles for High-Temperature Gas-Cooled Reactors

International Nuclear Information System (INIS)

Minato, Kazuo; Sawa, Kazuhiro; Koya, Toshio; Tomita, Takeshi; Ishikawa, Akiyoshi; Baldwin, Charles A.; Gabbard, William Alexander; Malone, Charlie M.

2000-01-01

Postirradiation heating tests of TRISO-coated UO 2 particles at 1700 and 1800degC were performed to understand fission product release behavior at accident temperatures. The inventory measurements of the individual particles were carried out before and after the heating tests with gamma-ray spectrometry to study the behavior of the individual particles. The time-dependent release behavior of 85 Kr, 110m Ag, 134 Cs, 137 Cs, and 154 Eu were obtained with on-line measurements of fission gas release and intermittent measurements of metallic fission product release during the heating tests. The inventory measurements of the individual particles revealed that fission product release behavior of the individual particles was not uniform, and large particle-to-particle variations in the release behavior of 110m Ag, 134 Cs, 137 Cs, and 154 Eu were found. X-ray microradiography and ceramography showed that the variations could not be explained by only the presence or absence of cracks in the SiC coating layer. The SiC degradation may have been related to the variations

Metastable decay of photoionized niobium clusters: Evaporation vs fission fragmentation

International Nuclear Information System (INIS)

Cole, S.K.; Liu, K.; Riley, S.J.

1986-01-01

The metastable decay of photoionized niobium clusters (Nb/sub n/ + ) has been observed in a newly constructed cluster beam machine. The decay manifests itself in the time-of-flight (TOF) mass spectrum as an asymmetric broadening of daughter ion peaks. Pulsed ion extraction has been used to measure the decay rate constants and to establish the mechanism of the fragmentation, evaporation and/or fission of the photoionized clusters. It is found that within the experimental time window evaporation dominates for the smaller clusters (n 6 sec -1 . The average kinetic energy release is also determined and is found to be on the order of 5 MeV. 8 refs., 3 figs., 1 tab

Fission product behaviour in severe accidents

International Nuclear Information System (INIS)

Jokiniemi, J.; Auvinen, A.; Maekynen, J.; Valmari, T.

1998-01-01

The understanding of fission product (FP) behaviour in severe accidents is important for source term assessment and accident mitigation measures. For example in accident management the operator needs to know the effect of different actions on the behaviour and release of fission products. At VTT fission product behaviour have been studied in different national and international projects. In this presentation the results of projects in EU funded 4th framework programme Nuclear Fission Safety 1994-1998 are reported. The projects are: fission product vapour/aerosol chemistry in the primary circuit (FI4SCT960020), aerosol physics in containment (FI4SCT950016), revaporisation of test samples from Phebus fission products (FI4SCT960019) and assessment of models for fission product revaporisation (FI4SCT960044). Also results from the national project 'aerosol experiments in the Victoria facility' funded by IVO PE and VTT Energy are reported

Ex-vessel water-level and fission-product monitoring for LWR

International Nuclear Information System (INIS)

DeVolpi, A.; Markoff, D.

1988-01-01

Given that the need for direct measurement of reactor coolant inventory under operational or abnormal conditions remains unsatisfied, a high-energy gamma-ray detection system is described for ex-vessel monitoring. The system has been modeled to predict response in a PWR, and the model has been validated with a LOFT LOCA sequence. The apparatus, situated outside the pressure vessel, would give relative water level and density over the entire vessel height and distinguish differing levels in the downcomer and core. It would also have significant sensitivity after power shutdown because of high-energy gamma rays from photoneutron capture, the photoneutrons being the result of fission-product decay in the core. Fission-products released to the coolant and accumulated in the top of a PWR vessel would also be theoretically detectable

Theoretical Description of the Fission Process

International Nuclear Information System (INIS)

Nazarewicz, Witold

2009-01-01

Advanced theoretical methods and high-performance computers may finally unlock the secrets of nuclear fission, a fundamental nuclear decay that is of great relevance to society. In this work, we studied the phenomenon of spontaneous fission using the symmetry-unrestricted nuclear density functional theory (DFT). Our results show that many observed properties of fissioning nuclei can be explained in terms of pathways in multidimensional collective space corresponding to different geometries of fission products. From the calculated collective potential and collective mass, we estimated spontaneous fission half-lives, and good agreement with experimental data was found. We also predicted a new phenomenon of trimodal spontaneous fission for some transfermium isotopes. Our calculations demonstrate that fission barriers of excited superheavy nuclei vary rapidly with particle number, pointing to the importance of shell effects even at large excitation energies. The results are consistent with recent experiments where superheavy elements were created by bombarding an actinide target with 48-calcium; yet even at high excitation energies, sizable fission barriers remained. Not only does this reveal clues about the conditions for creating new elements, it also provides a wider context for understanding other types of fission. Understanding of the fission process is crucial for many areas of science and technology. Fission governs existence of many transuranium elements, including the predicted long-lived superheavy species. In nuclear astrophysics, fission influences the formation of heavy elements on the final stages of the r-process in a very high neutron density environment. Fission applications are numerous. Improved understanding of the fission process will enable scientists to enhance the safety and reliability of the nation's nuclear stockpile and nuclear reactors. The deployment of a fleet of safe and efficient advanced reactors, which will also minimize radiotoxic

Uncertainties in source term calculations generated by the ORIGEN2 computer code for Hanford Production Reactors

International Nuclear Information System (INIS)

Heeb, C.M.

1991-03-01

The ORIGEN2 computer code is the primary calculational tool for computing isotopic source terms for the Hanford Environmental Dose Reconstruction (HEDR) Project. The ORIGEN2 code computes the amounts of radionuclides that are created or remain in spent nuclear fuel after neutron irradiation and radioactive decay have occurred as a result of nuclear reactor operation. ORIGEN2 was chosen as the primary code for these calculations because it is widely used and accepted by the nuclear industry, both in the United States and the rest of the world. Its comprehensive library of over 1,600 nuclides includes any possible isotope of interest to the HEDR Project. It is important to evaluate the uncertainties expected from use of ORIGEN2 in the HEDR Project because these uncertainties may have a pivotal impact on the final accuracy and credibility of the results of the project. There are three primary sources of uncertainty in an ORIGEN2 calculation: basic nuclear data uncertainty in neutron cross sections, radioactive decay constants, energy per fission, and fission product yields; calculational uncertainty due to input data; and code uncertainties (i.e., numerical approximations, and neutron spectrum-averaged cross-section values from the code library). 15 refs., 5 figs., 5 tabs

A passive decay-heat removal system for an ABWR based on air cooling

Energy Technology Data Exchange (ETDEWEB)

Mochizuki, Hiroyasu, E-mail: mochizki@u-fukui.ac.jp [Research Institute of Nuclear Engineering, University of Fukui, 1-2-4 Kanawa-cho, Tsuruga, Fukui 914-0055 (Japan); Yano, Takahiro [School of Engineering, University of Fukui, 1-2-4 Kanawa-cho, Tsuruga, Fukui 914-0055 (Japan)

2017-01-15

Highlights: • A passive decay heat removal system for an ABWR is discussed using combined system of the reactor and an air cooler. • Effect of number of pass of the finned heat transfer tubes on heat removal is investigated. • The decay heat can be removed by air coolers with natural convection. • Two types of air cooler are evaluated, i.e., steam condensing and water cooling types. • Measures how to improve the heat removal rate and to make compact air cooler are discussed. - Abstract: This paper describes the capability of an air cooling system (ACS) operated under natural convection conditions to remove decay heat from the core of an Advanced Boiling Water Reactor (ABWR). The motivation of the present research is the Fukushima Severe Accident (SA). The plant suffered damages due to the tsunami and entered a state of Station Blackout (SBO) during which seawater cooling was not available. To prevent this kind of situation, we proposed a passive decay heat removal system (DHRS) in the previous study. The plant behavior during the SBO was calculated using the system code NETFLOW++ assuming an ABWR with the ACS. However, decay heat removal under an air natural convection was difficult. In the present study, a countermeasure to increase heat removal rate is proposed and plant transients with the ACS are calculated under natural convection conditions. The key issue is decreasing pressure drop over the tube banks in order to increase air flow rate. The results of the calculations indicate that the decay heat can be removed by the air natural convection after safety relief valves are actuated many times during a day. Duct height and heat transfer tube arrangement of the AC are discussed in order to design a compact and efficient AC for the natural convection mode. As a result, a 4-pass heat transfer tubes with 2-row staggered arrangement is the candidate of the AC for the DHRS under the air natural convection conditions. The heat removal rate is re-evaluated as

Prediction of fission product and actinide levels in irradiated fuel and cladding

International Nuclear Information System (INIS)

Burstall, R.F.; Thornton, D.E.J.

1977-01-01

The production of radioactive isotopes and their subsequent decay is of crucial importance in the nuclear industry, dominating the shield design of chemical reprocessing plants, transport flasks and waste disposal facilities which account for a large part of the capital investment in a nuclear programme. The isotopes are also important in studies of reactor shielding. The computation of the level and behavior of such nuclides has been practiced for many years in countries with nuclear industries, with ever-increasing sophistication in methods of calculation and in improving the accuracy of the basic nuclide data. Calculation is usually made for three groups of nuclides, the actinides or transuranics, the fission products, and nuclides present in the cladding. The currently accepted computer code within the UKAEA for such calculations is FISPIN. This code calculates activities for all the above groups either separately or in combination. As well as individual nuclide concentrations and activities integral information is produced. The paper describes the methods of calculation. The code has been compared with other codes which have a similar function, and it is concluded that the only significant differences are those associated with data. A number of different data sets, to a large degree independent, have been compared using the code, and the paper describes some of the results obtained

Potential for large-scale uses for fission-product Xenon

International Nuclear Information System (INIS)

Rohrmann, C.A.

1983-03-01

Of all fission products in spent, low-enrichment-uranium power-reactor fuels, xenon is produced in the highest yield - nearly one cubic meter, STP, per metric ton. In aged fuels which may be considered for processing in the US, radioactive xenon isotopes approach the lowest limits of detection. The separation from accompanying radioactive 85 Kr is the essential problem; however, this is state-of-the-art technology which has been demonstrated on the pilot scale to yield xenon with pico-curie levels of 85 Kr contamination. If needed for special applications, such levels could be further reduced. Environmental considerations require the isolation of essentially all fission-product krypton during fuel processing. Economic restraints assure that the bulk of this krypton will need to be separated from the much-more-voluminous xenon fraction of the total amount of fission gas. Xenon may thus be discarded or made available for uses at probably very low cost. In contrast with many other fission products which have unique radioactive characteristics which make them useful as sources of heat, gamma and x-rays, and luminescence - as well as for medicinal diagnostics and therapeutics - fission-product xenon differs from naturally occurring xenon only in its isotopic composition which gives it a slightly hgiher atomic weight, because of the much higher concentrations of the 134 Xe and 136 Xe isotopes. Therefore, fission-product xenon can most likely find uses in applications which already exist but which can not be exploited most beneficially because of the high cost and scarcity of natural xenon. Unique uses would probably include applications in improved incandescent light illumination in place of krypton and in human anesthesia

Prediction of fission mass-yield distributions based on cross section calculations

International Nuclear Information System (INIS)

Hambsch, F.-J.; G.Vladuca; Tudora, Anabella; Oberstedt, S.; Ruskov, I.

2005-01-01

For the first time, fission mass-yield distributions have been predicted based on an extended statistical model for fission cross section calculations. In this model, the concept of the multi-modality of the fission process has been incorporated. The three most dominant fission modes, the two asymmetric standard I (S1) and standard II (S2) modes and the symmetric superlong (SL) mode are taken into account. De-convoluted fission cross sections for S1, S2 and SL modes for 235,238 U(n, f) and 237 Np(n, f), based on experimental branching ratios, were calculated for the first time in the incident neutron energy range from 0.01 to 5.5 MeV providing good agreement with the experimental fission cross section data. The branching ratios obtained from the modal fission cross section calculations have been used to deduce the corresponding fission yield distributions, including mean values also for incident neutron energies hitherto not accessible to experiment

Radiation characteristics of spent fuel of heavy-water research reactor during long-term storage

International Nuclear Information System (INIS)

Gerasimov, A.S.; Kiselev, G.V.; Myrtsymova, L.A.; Zaritskaya, T.S.

2002-01-01

Decay heat power and radiotoxicity by water of actinides and fission products from spent fuel of heavy-water research reactor RA were calculated for period of storage during 300000 years. Three variants of fuel enrichment by 235 U were considered: 2%, 21%, and 80%. The mass of 235 U in one fuel element was supposed to be the same for all variants of enrichment. The decay heat power of fission products in initial period is about 20 times higher than that of actinides. Decay heat power and radiotoxicity of actinides do not practically decrease during long period of time as they are determined by nuclides with very long half-life periods. (author)

New Burnup Calculation System for Fusion-Fission Hybrid System

International Nuclear Information System (INIS)

Isao Murata; Shoichi Shido; Masayuki Matsunaka; Keitaro Kondo; Hiroyuki Miyamaru

2006-01-01

Investigation of nuclear waste incineration has positively been carried out worldwide from the standpoint of environmental issues. Some candidates such as ADS, FBR are under discussion for possible incineration technology. Fusion reactor is one of such technologies, because it supplies a neutron-rich and volumetric irradiation field, and in addition the energy is higher than nuclear reactor. However, it is still hard to realize fusion reactor right now, as well known. An idea of combination of fusion and fission concepts, so-called fusion-fission hybrid system, was thus proposed for the nuclear waste incineration. Even for a relatively lower plasma condition, neutrons can be well multiplied by fission in the nuclear fuel, tritium is thus bred so as to attain its self-sufficiency, enough energy multiplication is then expected and moreover nuclear waste incineration is possible. In the present study, to realize it as soon as possible with the presently proven technology, i.e., using ITER model with the achieved plasma condition of JT60 in JAEA, Japan, a new calculation system for fusion-fission hybrid reactor including transport by MCNP and burnup by ORIGEN has been developed for the precise prediction of the neutronics performance. The author's group already has such a calculation system developed by them. But it had a problem that the cross section libraries in ORIGEN did not have a cross section library, which is suitable specifically for fusion-fission hybrid reactors. So far, those for FBR were approximately used instead in the analysis. In the present study, exact derivation of the collapsed cross section for ORIGEN has been investigated, which means it is directly evaluated from calculated track length by MCNP and point-wise nuclear data in the evaluated nuclear data file like JENDL-3.3. The system realizes several-cycle calculation one time, each of which consists of MCNP criticality calculation, MCNP fixed source calculation with a 3-dimensional precise

Shielding calculation of a hot cell for the processing of fission products

International Nuclear Information System (INIS)

Rocha, A.C.S. da; Pina, J.L.S. de; Silva, J.J.G. da.

1986-12-01

A dose rate estimation is made for an operator of a lead wall, fission products processing hot cell, in a distance of 50 cm from the emission source, at Brazilian Institute of Nuclear Engineering (IEN). (L.C.J.A.)

Use of dwell time concept in fission product inventory assessment for CANDU reactors

International Nuclear Information System (INIS)

Bae, C.J.; Choi, J.H.; Hwang, H.R.; Seo, J.T.

2003-01-01

A realistic approach in calculating the initial fission product inventory within the CANFLEX-NU fuel has been assessed for its applicability to the single channel event safety analysis for CANDU reactors. This approach is based on the dwell time concept in which the accident is assumed to occur at the dwell time when the summation of fission product inventory for all isotopes becomes largest. However, in the current conservative analysis, the maximum total inventory and the corresponding gap inventory for each isotope are used as the initial fission product inventories regardless of the accident initiation time. The fission product inventory analysis has been performed using ELESTRES code considering power histories and burnup of the fuel bundles in the limiting channel. The analysis results showed that the total fission product inventory is found to be largest at 20% dwell time. Therefore, the fission product inventory at 20% dwell time can be used as the initial condition for the single channel event for the CANDU 6 reactors. (author)

Fission Product Release Behavior of Individual Coated Fuel Particles for High-Temperature Gas-Cooled Reactors

Energy Technology Data Exchange (ETDEWEB)

Minato, Kazuo [Japan Atomic Energy Research Institute (Japan); Sawa, Kazuhiro [Japan Atomic Energy Research Institute (Japan); Koya, Toshio [Japan Atomic Energy Research Institute (Japan); Tomita, Takeshi [Japan Atomic Energy Research Institute (Japan); Ishikawa, Akiyoshi [Japan Atomic Energy Research Institute (Japan); Baldwin, Charles A; Gabbard, William Alexander [Oak Ridge National Laboratory (United States); Malone, Charlie M [Oak Ridge National Laboratory (United States)

2000-07-15

Postirradiation heating tests of TRISO-coated UO{sub 2} particles at 1700 and 1800degC were performed to understand fission product release behavior at accident temperatures. The inventory measurements of the individual particles were carried out before and after the heating tests with gamma-ray spectrometry to study the behavior of the individual particles. The time-dependent release behavior of {sup 85}Kr, {sup 110m}Ag, {sup 134}Cs, {sup 137}Cs, and {sup 154}Eu were obtained with on-line measurements of fission gas release and intermittent measurements of metallic fission product release during the heating tests. The inventory measurements of the individual particles revealed that fission product release behavior of the individual particles was not uniform, and large particle-to-particle variations in the release behavior of {sup 110m}Ag, {sup 134}Cs, {sup 137}Cs, and {sup 154}Eu were found. X-ray microradiography and ceramography showed that the variations could not be explained by only the presence or absence of cracks in the SiC coating layer. The SiC degradation may have been related to the variations.

Decay heat and gamma dose-rate prediction capability in spent LWR fuel

International Nuclear Information System (INIS)

Neely, G.J.; Schmittroth, F.

1982-08-01

The ORIGEN2 code was established as a valid means to predict decay heat from LWR spent fuel assemblies for decay times up to 10,000 year. Calculational uncertainties ranged from 8.6% to a maximum of 16% at 2.5 years and 300 years cooling time, respectively. The calculational uncertainties at 2.5 years cooling time are supported by experiment. Major sources of uncertainty at the 2.5 year cooling time were identifed as irradiation history (5.7%) and nuclear data together with calculational methods (6.3%). The QAD shielding code was established as a valid means to predict interior and exterior gamma dose rates of spent LWR fuel assemblies. A calculational/measurement comparison was done on two assemblies with different irradiation histories and supports a 35% calculational uncertainty at the 1.8 and 3.0 year decay times studied. Uncertainties at longer times are expected to increase, but not significantly, due to an increased contribution from the actinides whose inventories are assigned a higher uncertainty. The uncertainty in decay heat rises to a maximum of 16% due to actinide uncertainties. A previous study was made of the neutron emission rate from a typical Turkey Point Unit 3, Region 4 spent fuel assembly at 5 years decay time. A conservative estimate of the neutron dose rate at the assembly surface was less than 0.5 rem/hr

Feasibility study of chabazite absorber tube utilization in online absorption of released gaseous fission products and substitution of burnable absorber rods with chabazite absorber tubes in VVER-1000 reactor series

International Nuclear Information System (INIS)

Rahmani, Yashar

2017-01-01

Highlights: • Chabazite tubes are used for online removal of the released gaseous fission products. • The feasibility of using chabazite tubes instead of burnable absorber rods was studied. • A computational cycle was designed using the WIMSD5-B, CITATION-LDI2 and WERL codes. • In modeling fission gas release, the Weisman, Booth, Mason and T.S. models were used. • By this method, it is possible to increase cycle length and enhance heat transfer. - Abstract: As gaseous fission products, e.g. xenon and krypton have adverse effects such as reducing the rate of heat transfer in fuel rods and adding negative reactivity to the reactor core, the present manuscript was dedicated to development of a novel method for improving these defects. In the proposed method, chabazite absorber tubes were used for online removal of the released gaseous fission products from gaseous gap spaces. Moreover, in this research, feasibility of using chabazite absorber tubes instead of burnable absorber rods was examined. To perform the required modeling and calculations to successfully meet the mentioned objectives, a thermo-neutronic computational cycle was designed using the coupling of WIMSD5-B and CITATION-LDI2 codes in the neutronic section and the WERL code in the thermo-hydraulic calculations. In addition, in modeling the release process of gaseous fission products, the Weisman, Booth, Mason, and T.S. models were examined. It is worth mentioning that in this research, calculations and modeling procedures were based on the first cycle of Bushehr’s VVER-1000 reactor to study the feasibility of the proposed solution. The obtained results revealed that with application of the proposed method in this research, it is possible to increase cycle length, improve safety thresholds, and enhance heat transfer in the core of nuclear reactors.

Fission energy of uranium isotopes and transuranium elements

International Nuclear Information System (INIS)

Nemirovskij, P.Eh.; Manevich, L.G.

1981-01-01

A comparison is made between the prompt fission energy, Esub(pr), calculated from the mass and binding energy spectrum and the Esub(pr) value obtained from the experimental data on the kinetic energy of fragments, the energy of prompt neutrons and prompt γ-quanta. Basing on the data on β-decay chains of fission fragments, the energies of neutrinos, γ-quanta and β-electrons are obtained, which permits to calculate the actual energy released during fission. The calculations are performed for thermal neutron-induced fission, fast-neutron induced fission and for fission after bombardment with 14 MeV neutrons. The available experimental data on the fission fragment kinetic energy, prompt γ-quanta energy and fission neutron energy are presented. The comparison of the Esub(pr) values obtained experimentally for the thermal-neutron-induced fission with the calculated Esub(pr) value shows that for 233 U, 239 Pu, 241 Pu the agreement is rather favourable. For 235 U the agreement is within the error limits. As to the Esub(pr) values for the fast-neutron-induced fission, the agreement between the calculated and experimental data for all nuclides is quite good

Proceedings of the specialists' meeting on physics and engineering of fission and spallation, 1989

International Nuclear Information System (INIS)

Nakagome, Yoshihiro

1990-07-01

The third meeting was held on August 1, and the fourth meeting was held on December 12, 1989. The reports of the international conferences on 50 years research on nuclear fission in Germany and USA, and the reports on the nuclear data of fission-produced nuclei for evaluating reactor decay heat, the atomic mass formula considering proton-neutron interaction and unstable nuclei, research on short life fission fragments by on-line isotope separation process, the reactor physics on waste annihilation disposal and fuel breeding with an accelerator, the double differential cross section of back neutrons in nuclear spallation reaction, measurement of fission cross section and fission neutron spectra with fast neutrons, U-235 fission spectra by unfolding activation foil data and production mechanisms of intermediate mass fragments from hot nuclei-emission of complex and fission fragments for 84 Kr+ 27 Al at 10.6 MeV/u were made. (K.I.)

On fission product retention in the core of the low powered high temperature reactor under accident conditions

International Nuclear Information System (INIS)

Bastek, H.

1984-01-01

In the core of the high temperature reactor the fuel element and the coated particles contained herein provide the safest enclosure for fission products. The complex process of fission product transport out of the particle kernel, through the particle coating and within the fuel element graphite is described in a simplified form by the Fick's diffusion. The effective diffusion coefficient is used for calculation. Starting from the existing ideas of fission product transport five burn-up and temperature-dependent diffusion coefficients for Cesium in (Th,U)O 2 -kernels are derived in this study. The results have been gained from several fuel element radiation experiments in recent years, which showed extreme variation in regard to burn-up, temperature cycle, neutron flux and operation time. Cs-137 release measurements from single particle kernels were present from all the experiments. Furthermore, annealing tests of AVR-fuel elements were analyzed. Heat-temperatur and heating-time, the fuel element burn-up in the AVR-reactor, as well as the measured Cs-137 inventory of the fuel elements before and after annealing, are included in the investigation as essential parameters. With the aid of the derived diffusion coeffizients and already present data sets the Cs-137 release of fuel elements into a small reactor core is investigated under unrestricted core heat-up. While the released Cs-137 is derived mainly from defective particles at accident temperatures up to 1600 0 C, the main part diffuses through the particle coating at higher accident temperatures. (orig./HP) [de

Preparation of lumped fission product (FP) cross sections for a multigroup library

International Nuclear Information System (INIS)

Ono, S.; Corcuera, R.P.

1984-01-01

A method for the calculation of lumped Fission Product (FP) cross sections has been developed. The group constants fo each nuclide are generated by NJOY code, based on ENDF/B-V data. In this first version, cross section of 28 nuclides are lumped for typical characteristics of Binary Breeder Reactor (BBR). One energy group calculations are made for a 1000 MWe fast reactor to verify the influence of burnup, number of FP and fuel composition on the lumped fission product cross sections. (Author) [pt

Fission product detection

International Nuclear Information System (INIS)

Liatard, E.; Akrouf, S.; Bruandet, J.F

1987-01-01

The response of photovoltaic cells to heavy ions and fission products have been tested on beam. Their main advantages are their extremely low price, their low sensitivity to energetic light ions with respect to fission products, and the possibility to cut and fit them together to any shape without dead zone. The time output signals of a charge sensitive preamplifier connected to these cells allows fast coincidences. A resolution of 12ns (F.W.H.M.) have been measured between two cells [fr

Extraction chromatography of fission products

International Nuclear Information System (INIS)

Bonnevie-Svendsen, M.; Goon, K.

1978-01-01

Various cases of using extraction chromatography during analysis of fission products are reviewed. The use of the extraction chromatography method is considered while analysing reprocessed products of nuclear fuel for quantitative radiochemical analysis and control of fission product and actinoide separation during extraction and their chemical state in production solutions. The method is used to obtain pure fractions of typical burnup monitors (neodymium, molybdenum, cerium, cesium, europium, lanthanides) during determination of nuclear fuel burnup degree. While studying the nature of nuclear reactions the method is used to separate quickly short-life isotopes, to purify β-radiator fractions before measuring their half-life periods, to enrich isotopes forming with low output during fission. Examples of using extraction chromatography are given to separate long half-life or stable fission products from spent solutions, to control environment object contamination

Fission cross section calculations for 209Bi target nucleus based on fission reaction models in high energy regions

Directory of Open Access Journals (Sweden)

Kaplan Abdullah

2015-01-01

Full Text Available Implementation of projects of new generation nuclear power plants requires the solving of material science and technological issues in developing of reactor materials. Melts of heavy metals (Pb, Bi and Pb-Bi due to their nuclear and thermophysical properties, are the candidate coolants for fast reactors and accelerator-driven systems (ADS. In this study, α, γ, p, n and 3He induced fission cross section calculations for 209Bi target nucleus at high-energy regions for (α,f, (γ,f, (p,f, (n,f and (3He,f reactions have been investigated using different fission reaction models. Mamdouh Table, Sierk, Rotating Liquid Drop and Fission Path models of theoretical fission barriers of TALYS 1.6 code have been used for the fission cross section calculations. The calculated results have been compared with the experimental data taken from the EXFOR database. TALYS 1.6 Sierk model calculations exhibit generally good agreement with the experimental measurements for all reactions used in this study.

Library of data for fission products

International Nuclear Information System (INIS)

Blachot, Jean; Devillers, Christian; Tourreil, Roland de; Nimal, Bernadette; Fiche, Charles; Noel, J.-P.

1975-10-01

This is the fourth version of the CEA fission products nuclear data library. The third one has been previously published in CEA-N--1526. Data for 635 nuclides ranging from mass A=71 up to A=170 are arranged in increasing order of atomic number. Data are presented in two tables: the first one gives for each nuclide, the half-life, the Q-values and branching ratios for the various decay modes, the energies and intensities of the β - , β + and isomeric transitions and of gamma rays; the second one gives an ordered list of all gamma ray energies, with associated nuclide, half-life and intensity. Bibliographic references and, for most of the data, uncertainties are provided [fr

Freedom: a transient fission-product release model for radioactive and stable species

International Nuclear Information System (INIS)

Macdonald, L.D.; Lewis, B.J.; Iglesias, F.C.

1989-05-01

A microstructure-dependent fission-gas release and swelling model (FREEDOM) has been developed for UO 2 fuel. The model describes the transient release behaviour for both the radioactive and stable fission-product species. The model can be applied over the full range of operating conditions, as well as for accident conditions that result in high fuel temperatures. The model accounts for lattice diffusion and grain-boundary sweeping of fusion products to the grain boundaries, where the fission gases accumulate in grain-face bubbles as a result of vacancy diffusion. Release of fission-gas to the free void of the fuel element occurs through the interlinkage of bubbles and cracks on the grain boundaries. This treatment also accounts for radioactive chain decay and neutron-induced transmutation effects. These phenomena are described by mass balance equations which are numerically solved using a moving-boundary, finite-element method with mesh refinement. The effects of grain-face bubbles on fuel swelling and fuel thermal conductivity are included in the ELESIM fuel performance code. FREEDOM has an accuracy of better than 1% when assessed against an analytic solution for diffusional release. The code is being evaluated against a fuel performance database for stable gas release, and against sweep-gas and in-cell fission-product release experiments at Chalk River for active species

A method for rapid estimation of internal dose to members of the public from inhalation of mixed fission products (based on the ICRP 1994 human respiratory tract model for radiological protection)

International Nuclear Information System (INIS)

Hou Jieli

1999-01-01

Based on the computing principle given in ICRP-30, a method had been given by the author for fast estimating internal dose from an intake of mixed fission products after nuclear accident. Following the ICRP-66 Human respiratory tract model published in 1994, the method was reconstructed. The doses of 1 Bq intake of mixed fission products (its AMAD = 1 μm, decay rate coefficient n = 0.2∼2.0) during the period of 1∼15 d after an accident were calculated. It is lower slightly based on ICRP 1994 respiratory tract model than that based on ICRP-30 model

Excitation of giant resonances through inelastic scattering of 170 at 84 MeV/u. Fission decay of giant resonances

International Nuclear Information System (INIS)

Cabot, C.; Barrette, J.; Mark, S.K.; Turcotte, R.; Xing, J.; Van der Woude, A.; Van Den Berg, A. M.

1991-01-01

Inelastic scattering of 84 MeV/u 17 0 projectiles have been used to excite the giant resonances (GR) in various nuclei ranging from A=60 to A=232. For the isoscalar giant quadrupole resonance (ISGQR), the energy and width of the resonance, as well as the EWSR obtained from the measured cross sections, are in agreement with the known systematics for A>40. The observed GMR strengths are close to 100% EWRS and are consistent with other recent experimental results using heavy ion projectiles. These results lead to a somewhat different picture than that provided by previous studies using light projectiles. Strength is also observed at high excitation energy. The analysis of these resonances is in progress. Our study of the fission decay of GR in 232 Th leads to a somewhat different conclusion than previously deduced from data obtained with light ion projectiles, where no evidence for the fission decay of the ISGQR has been found. In the present work, due to the very good peak-to-continuum ratio, a structure is observed in the fission coincidence spectrum around 10 MeV which can be attributed to the fission decay of giant resonances. The measured fission probability is consistent with a statistical decay of the ISGQR. 10 figs

Recent improvements in the calculation of prompt fission neutron spectra: Preliminary results

International Nuclear Information System (INIS)

Madland, D.G.; LaBauve, R.J.; Nix, J.R.

1989-01-01

We consider three topics in the refinement and improvement of our original calculations of prompt fission neutron spectra. These are an improved calculation of the prompt fission neutron spectrum N(E) from the spontaneous fission of 252 Cf, a complete calculation of the prompt fission neutron spectrum matrix N(E,E n ) from the neutron-induced fission of 235 U, at incident neutron energies ranging from 0 to 15 MeV, and an assessment of the scission neutron component of the prompt fission neutron spectrum. Preliminary results will be presented and compared with experimental measurements and an evaluation. A suggestion is made for new integral cross section measurements. (author). 45 refs, 12 figs, 1 tab

Behaviour of fission products in PWR primary coolant and defected fuel rods evaluation

International Nuclear Information System (INIS)

Bourgeois, P.; Stora, J.P.

1979-01-01

The activity surveillance of the PWR primary coolant by γ spectometry gives some informations on fuel failures. The activity of different nuclides e.g. Xenons, Kryptons, Iodines, can be correlated with the number of the defected fuel rods. Therefore the precharacterization with eventually a prelocalization of the related fuel assemblies direct the sipping-test and allows a saving of time during refueling. A model is proposed to calculate the number of the defected rods from the activity measurements of the primary coolant. A semi-empirical model of the release of the fission products has been built from the activity measurements of the primary coolant in a 900 MWe PWR. This model allows to calculate the number of the defected rods and also a typical parameter of the mean damage. Fission product release is described by three stages: release from uranium dioxide, transport across the gas gap and behaviour in the primary coolant. The model of release from the oxide considers a diffusion process in the grains with trapping. The release then occurs either directly to free surfaces or with a delay due to a transit into closed porosity of the oxide. The amount released is the same for iodine and rare gas. With the gas gap transit is associated a transport time and a probability of trapping for the iodines. In the primary coolant the purification and the radioactive decay are considered. (orig.)

A model for the release of low-volatility fission products in oxidizing conditions

International Nuclear Information System (INIS)

Cox, D.S.; Hunt, C.E.L.; Liu, Z.; Keller, N.A.; Barrand, R.D.; O'Connor, R.F.

1991-07-01

A thermodynamic and kinetic model has been developed for calculating low-volatility fission-product releases from UO 2 at high temperatures in oxidizing conditions. Volatilization of the UO 2 matrix is assumed to be the rate controlling process. Oxidation kinetics of the UO 2 are modelled by either interfacial rate control, gas phase oxidant transport control, or solid-state diffusion of oxygen. The vapour pressure of UO 3 in equilibrium with the oxidizing fuel is calculated from thermodynamic data, and volatilization rates are determined using a model for forced convective mass transport. Low-volatility fission-product releases are calculated from the volume of vapourized fuel. Model calculations are conservative compared to experimental data for Zr, La, Ce and Nb fission-product releases from irradiated UO 2 exposed to air at 1973-2350 K. The implications of this conservatism are discussed in terms of possible rate control by processes other than convective mass transport of UO 3 . Coefficients for effective surface area (based on experimental data) and for heterogeneous rate controlling reaction kinetics are introduced to facilitate agreement between calculations and the experimental data.

Investigation of short-living fission products from the spontaneous fission of Cf-252

International Nuclear Information System (INIS)

Klonk, H.

1976-01-01

In this paper, a method of separating and measuring fission products of Cf-252 is presented. The measurement was achieved by means of γ-spectrometry and thus provides a quantitative analysis with a good separation of the fission products with respect to both atomic number Z and mass number A. The separation of the fission products from the fission source was achieved by means of solid traps. An automatic changing apparatus made it possible to keep irradiation and measuring times short, so even very short-lived fission products could be registered. The quantitative evaluation of primary fission products was made possible by correction according to Bateman equations. With that, the yields of single nuclides and the dispersion of charge can be determined. (orig./WL) [de

Calculated nuclide production yields in relativistic collisions of fissile nuclei

Energy Technology Data Exchange (ETDEWEB)

Benlliure, J.; Schmidt, K.H. [Gesellschaft fuer Schwerionenforschung mbH, Darmstadt (Germany); Grewe, A.; Jong, M. de [Technische Univ. Darmstadt (Germany). Inst. fuer Kernphysik; Zhdanov, S. [AN Kazakhskoj SSR, Alma-Ata (USSR). Inst. Yadernoj Fiziki

1997-11-01

A model calculation is presented which predicts the complex nuclide distribution resulting from peripheral relativistic heavy-ion collisions involving fissile nuclei. The model is based on a modern version of the abrasion-ablation model which describes the formation of excited prefragments due to the nuclear collisions and their consecutive decay. The competition between the evaporation of different light particles and fission is computed with an evaporation code which takes dissipative effects and the emission of intermediate-mass fragments into account. The nuclide distribution resulting from fission processes is treated by a semiempirical description which includes the excitation-energy dependent influence of nuclear shell effects and pairing correlations. The calculations of collisions between {sup 238}U and different reaction partners reveal that a huge number of isotopes of all elements up to uranium is produced. The complex nuclide distribution shows the characteristics of fragmentation, mass-asymmetric low-energy fission and mass-symmetric high-energy fission. The yields of the different components for different reaction partners are studied. Consequences for technical applications are discussed. (orig.)

Analysis of the WCLL European demo blanket concept in terms of activation and decay heat after exposure to neutron irradiation

Directory of Open Access Journals (Sweden)

Stankunas Gediminas

2017-01-01

Full Text Available This comparative paper describes the activation and decay heat calculations for water-cooled lithium-lead performed part of the EURO fusion WPSAE programme and specifications in comparison to other European DEMO blanket concepts on the basis of using a three-dimensional neutronics calculation model. Results are provided for a range of decay times of interest for maintenance activities, safety and waste management assessments. The study revealed that water-cooled lithium-lead has the highest total decay heat at longer decay times in comparison to the helium-cooled design which has the lowest total decay heat. In addition, major nuclides were identified for water-cooled lithium-lead in W armour, Eurofer, and LiPb. In addition, great attention has been dedicated to the analysis of the decay heat and activity both from the different water-cooled lithium-lead blanket modules for the entire reactor and from each water-cooled lithium-lead blanket module separately. The neutron induced activation and decay heat at shutdown were calculated by the FISPACT code, using the neutron flux densities and spectra that were provided by the preceding MCNP neutron transport calculations.

Calculated secondary yields for proton broadband using DECAY TURTLE

International Nuclear Information System (INIS)

Sondgeroth, A.

1995-02-01

The calculations for the yields were done by Al Sondgeroth and Anthony Malensek. The authors used the DECAY deck called PBSEC E.DAT from the CMS DECKS library. After obtaining the run modes and calibration modes from the liaison physicist, they made individual decay runs, using DECAY TURTLE from the CMS libraries and a production spectrum subroutine which was modified by Anthony, for each particle and decay mode for all particle types coming out of the target box. Results were weighted according to branching ratios for particles with more than one decay mode. The production spectra were produced assuming beryllium as the target. The optional deuterium target available to broadband will produce slightly higher yields. It should be noted that they did not include pion yields from klong decays because they could not simulate three body decays. Pions from klongs would add a very small fraction to the total yield

Calculation of the fast multiplication factor by the fission matrix method

International Nuclear Information System (INIS)

Naumov, V.A.; Rozin, S.G.; Ehl'perin, T.I.

1976-01-01

A variation of the Monte Carlo method to calculate an effective breeding factor of a nuclear reactor is described. The evaluation procedure of reactivity perturbations by the Monte Carlo method in the first order perturbation theory is considered. The method consists in reducing an integral neutron transport equation to a set of linear algebraic equations. The coefficients of this set are elements of a fission matrix. The fission matrix being a Grin function of the neutron transport equation, is evaluated by the Monte Carlo method. In the program realizing the suggested algorithm, the game for initial neutron energy of a fission spectrum and then for the region of neutron birth, ΔVsub(f)sup(i)has been played in proportion to the product of Σsub(f)sup(i)ΔVsub(f)sup(i), where Σsub(f)sup(i) is a macroscopic cross section in the region numbered at the birth energy. Further iterations of a space distribution of neutrons in the system are performed by the generation method. In the adopted scheme of simulation of neutron histories the emission of secondary neutrons is controlled by weights; it occurs at every collision and not only in the end on the history. The breeding factor is calculated simultaneously with the space distribution of neutron worth in the system relative to the fission process and neutron flux. Efficiency of the described procedure has been tested on the calculation of the breeding factor for the Godiva assembly, simulating a fast reactor with a hard spectrum. A high accuracy of calculations at moderate number of zones in the core and reasonable statistics has been stated

The potential for large scale uses for fission product xenon

International Nuclear Information System (INIS)

Rohrmann, C.A.

1983-01-01

Of all fission products in spent, low enrichment, uranium, power reactor fuels xenon is produced in the highest yield - nearly one cubic meter, STP, per metric ton. In aged fuels which may be considered for processing in the U.S. radioactive xenon isotopes approach the lowest limits of detection. The separation from accompanying radioactive 85 Kr is the essential problem; however, this is state of the art technology which has been demonstrated on the pilot scale to yield xenon with pico-curie levels of 85 Kr contamination. If needed for special applications, such levels could be further reduced. Environmental considerations require the isolation of essentially all fission product krypton during fuel processing. Economic restraints assure that the bulk of this krypton will need to be separated from the much more voluminous xenon fraction of the total amount of fission gas. Xenon may thus be discarded or made available for uses at probably very low cost. In contrast with many other fission products which have unique radioactive characteristics which make them useful as sources of heat, gamma and x-rays and luminescence as well as for medicinal diagnostics and therapeutics fission product xenon differs from naturally occurring xenon only in its isotopic composition which gives it a slightly higher atomic weight, because of the much higher concentrations of the 134 X and 136 Xe isotopes. Therefore, fission product xenon can most likely find uses in applications which already exist but which can not be exploited most beneficially because of the high cost and scarcity of natural xenon. Unique uses would probably include applications in improved incandescent light illumination in place of krypton and in human anesthesia

Alpha decay calculations with a new formula

International Nuclear Information System (INIS)

Akrawy, D T; Poenaru, D N

2017-01-01

A new semi-empirical formula for calculations of α  decay half-lives is presented. It was derived from the Royer relationship by introducing new parameters which are fixed by fit to a set of experimental data. We are using three sets: set A with 130 e–e (even–even), 119 e–o (even–odd), 109 o–e, and 96 o–o, set B with 188 e–e, 147 e–o, 131 o–e and 114 o–o, and set C with 136 e–e, 84 e–o, 76 o–e and 48 o–o alpha emitters. A comparison of results obtained with the new formula (newF) and the following well known relationships: semiempirical relationship based on fission theory (semFIS), analytical superasymmetric fission (ASAF) model and universal formula (UNIV) made in terms of rms standard deviation. We also introduced a weighted mean value of this quantity, allowing us to compare the global properties of a given model. For set B the order of the four models is the following: semFIS, UNIV, newF and ASAF. Nevertheless for even–even alpha emitters, UNIV gives the second best result after semFIS, and for odd–even parents the second is newF. Despite its simplicity in comparison with semFIS, newF, presented in this article, behaves quite well, competing with the other well known relationships. (paper)

Remarks on the fission barriers of super-heavy nuclei

Energy Technology Data Exchange (ETDEWEB)

Hofmann, S. [GSI Helmholtzzentrum fuer Schwerionenforschung, Darmstadt (Germany); Goethe-Universitaet Frankfurt, Institut fuer Physik, Frankfurt (Germany); Heinz, S.; Mann, R.; Maurer, J.; Muenzenberg, G.; Barth, W.; Dahl, L.; Kindler, B.; Kojouharov, I.; Lang, R.; Lommel, B.; Runke, J.; Scheidenberger, C.; Tinschert, K. [GSI Helmholtzzentrum fuer Schwerionenforschung, Darmstadt (Germany); Antalic, S. [Comenius University, Department of Nuclear Physics and Biophysics, Bratislava (Slovakia); Eberhardt, K.; Thoerle-Pospiech, P.; Trautmann, N. [Johannes Gutenberg-Universitaet Mainz, Mainz (Germany); Grzywacz, R. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); University of Tennessee, Knoxville, TN (United States); Hamilton, J.H. [Vanderbilt University, Department of Physics and Astronomy, Nashville, TN (United States); Henderson, R.A.; Kenneally, J.M.; Moody, K.J.; Shaughnessy, D.A.; Stoyer, M.A. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Miernik, K. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); University of Warsaw, Warsaw (Poland); Miller, D. [University of Tennessee, Knoxville, TN (United States); Morita, K. [RIKEN Nishina Center for Accelerator-Based Science, Wako, Saitama (Japan); Nishio, K. [Japan Atomic Energy Agency, Tokai, Ibaraki (Japan); Popeko, A.G.; Yeremin, A.V. [Joint Institute for Nuclear Research, Dubna (Russian Federation); Roberto, J.B.; Rykaczewski, K.P. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Uusitalo, J. [University of Jyvaeskylae, Department of Physics, Jyvaeskylae (Finland)

2016-04-15

Shell-correction energies of super-heavy nuclei are approximated by using Q{sub α} values of measured decay chains. Five decay chains were analyzed, which start at the isotopes {sup 285}Fl, {sup 294}118, {sup 291}Lv, {sup 292}Lv and {sup 293}Lv. The data are compared with predictions of macroscopic-microscopic models. Fission barriers are estimated that can be used to eliminate uncertainties in partial fission half-lives and in calculations of evaporation-residue cross-sections. In that calculations, fission probability of the compound nucleus is a major factor contributing to the total cross-section. The data also provide constraints on the cross-sections of capture and quasi-fission in the entrance channel of the fusion reaction. Arguments are presented that fusion reactions for synthesis of isotopes of elements 118 and 120 may have higher cross-sections than assumed so far. (orig.)

Fission Product Library and Resource

Energy Technology Data Exchange (ETDEWEB)

Burke, J. T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Padgett, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2016-09-29

Fission product yields can be extracted from an irradiated sample by performing gamma ray spectroscopy on the whole sample post irradiation. There are several pitfalls to avoid when trying to determine a specific isotope's fission product yield.

Contingency of alpha decay in 287-306120 isotopes of SHE

International Nuclear Information System (INIS)

Carmel Vigila Bai, G.M.; Umai Parvathiy, J.

2014-01-01

In recent years the synthesis and identification of super heavy nuclei has a particular attention in the field of nuclear physics. Many theoretical calculations have been done to study the properties of even-Z Super heavy elements (SHE). Durate et al. applied the effective liquid drop model to predict the alpha decay, cluster emission and cold fission half-life values of nuclei on the region of super heavy elements, defined by 155 ≤ N ≤ 220 and 110 ≤ Z ≤ 135. In the case of super heavy elements spontaneous fission and alpha decay are the main decay modes. Super heavy nuclei which have relatively small alpha decay half times compared to spontaneous fission half lives will survive fission and thus can be detected in the laboratory through α-decay. The present paper aims to predict possibility of alpha decay in the element Z = 120 isotopes using CYE model and the spontaneous fission half lives are computed using the phenomenological formula

Extensions to COGEND for ENDF/B-V output of spontaneous fission decay data

International Nuclear Information System (INIS)

Tobias, A.

1978-06-01

The computer code COGEND, used to produce ENDF/B-IV or -V format nuclear decay scheme data, has been modified in order to extend its range of application. Details are given of the additional facilities which permit the handling of spontaneous fission decay data including any associated continuous spectra. In order to accommodate these additional features it is necessary to increase the core region by 4 kilobytes. (author)

JEF-2.2 radioactive decay data

International Nuclear Information System (INIS)

1994-08-01

This work deals with the JEF-2.2 radioactive decay data and is divided into four tables. The first table presents the origin of the JEF-2.2 radioactive decay data and subsequent modifications. The second one is a summary of the JEF-2.2 radioactive decay data file. The third one describes the JEF-2.2 fission products and the main decay and fission yield data. The last one consists of the main decay parameters from the JEF-2.2, ENDF/B-VI and JNDC-2.0 libraries. (O.L.). 100 figs., 4 tabs