Paired replacement fuel assemblies for BWR-type reactor

International Nuclear Information System (INIS)

Oguchi, Kazushige.

1997-01-01

There are disposed a large-diameter water rod constituting a non-boiling region at a central portion and paired replacement fuel assemblies for two streams having the same average enrichment degree and different amount of burnable poisons. The paired replacement fuel assemblies comprise a first fuel assembly having a less amount of burnable poisons and a second fuel assembly having a larger amount of burnable poisons. A number of burnable poison-containing fuel rods in adjacent with the large diameter water rod is increased in the second fuel assembly than the first fuel assembly. Then, the poison of the paired replacement fuel assemblies for the BWR type reactor can be annihilated simultaneously at the final stage of the cycle. Accordingly, fuels for a BWR type reactor excellent in economical property and safety and facilitating the design of the replacement reactor core can be obtained. (N.H.)

Study of behavior on bonding and failure mode of pressurized and doped BWR fuel rod

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki

1992-03-01

The study of transient behavior on the bonding and the failure mode was made using the pressurized/doped 8 x 8 BWR type fuel rod. The dopant was mullite minerals consisted mainly of silicon and aluminum up to 1.5 w/o. Pressurization of the fuel rod with pure helium was made to the magnitude about 0.6 MPa. As a reference, the non-pressurized/non-doped 8 x 8 BWR fuel rod and the pressurized/7 x 7 BWR fuel rod up to 0.6 MPa were prepared. Magnitude of energy deposition given to the tested fuel rods was 248, 253, and 269 cal/g·fuel, respectively. Obtained results from the pulse irradiation in NSRR are as follows. (1) It was found from the experiment that alternation of the fuel design by the adoption of pressurization up to 0.6 MPa and the use of wider gap up to 0.38 mm could avoid the dopant BWR fuel from the overall bonding. The failure mode of the present dopant fuel was revealed to be the melt combined with rupture. (2) The time of fuel failure of the pressurized/doped 8 x 8 BWR fuel defected by the melt/rupture mode is of order of two times shorter than that of the pressurized/ 7 x 7 BWR defected by the rupture mode. Failure threshold of the pressurized/doped 8 x 8 BWR BWR tended to be lower than that of non-pressurized/non-doped 8 x 8 BWR one. Cracked area of the pressurized/doped 8 x 8 BWR was more wider and magnitude of oxidation at the place is relatively larger than the other tested fuels. (3) Failure mode of the non-pressurized/ 8 x 8 BWR fuel rod was the melt/brittle accompanied with a significant bonding at failed location. While, failure mode of the pressurized/ 7 x 7 BWR fuel rod was the cladding rupture accompanied with a large ballooning. No bonding at failed location of the latter was observed. (author)

Behavior of small-sized BWR fuel under reactivity initiated accident conditions

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki; Fujishiro, Toshio; Horiki, Oichiro; Chen Dianshan; Takeuchi, Kiyoshi.

1992-01-01

The present work was performed on this small-sized BWR fuel, where Zr liner and rod prepressurization were taken as experimental parameters. Experiment was done under simulated reactivity initiated accident (RIA) conditions at Nuclear Safety Research Reactor (NSRR) belonged to Japan Atomic Energy Research Institute (JAERI). Major remarks obtained are as follows: (1) Three different types of the fuel rods consisted of (a) Zr lined/pressurized (0.65MPa), (b) Zr lined/non-pressurized and (c) non-Zr lined/pressurized (o.65MPa) were used, respectively. Failure thresholds of these were not less than that (260 cal/g·fuel) described in Japanese RIA Licensing Guideline. Small-sized BWR and conventional 8 x 8 BWR fuels were considered to be in almost the same level in failure threshold. Failure modes of the three were (a) cladding melt/brittle, (b) cladding melt/brittle and (c) rupture by large ballooning, respectively. (2) The magnitude of pressure pulse at fuel fragmentation was also studied by lined/pressurized and non-lined/pressurized fuels. Above the energy deposition of 370 cal/g·fuel, mechanical energy (or pressure) was found to be released from these fragmented fuels. No measurable difference was, however, observed between the tested fuels and NSRR standard (and conventional 8 x 8 BWR) fuels. (3) It is worthy of mentioning that Zr liner tended to prevent the cladding from large ballooning. Non-lined/pressurized fuel tended to cause wrinkle deformation at cladding. Hence, cladding external was notched much by the wrinkles. (4) Time to fuel failure measured from the tested BWR fuels (pressurization < 0.6MPA) was longer than that measured from PWR fuels (pressurization < 3.2MPa). The magnitude of the former was of the order of 3 ∼ 6s, while that of the latter was < 1s. (J.P.N.)

Phenomenology of BWR fuel assembly degradation

Science.gov (United States)

Kurata, Masaki; Barrachin, Marc; Haste, Tim; Steinbrueck, Martin

2018-03-01

Severe accidents occurred at the Fukushima-Daiichi Nuclear Power Station (FDNPS) which required an immediate re-examination of fuel degradation phenomenology. The present paper reviews the updated knowledge on the phenomenology of the fuel degradation, focusing mainly on the BWR fuel assembly degradation at the macroscopic scale and that of the individual interactions at the meso-scale. Oxidation of boron carbide (B4C) control rods potentially generates far larger amounts of heat and hydrogen under BWR accident conditions. All integral tests with B4C control rods or control blades have shown early failure, liquefaction, relocation and oxidation of B4C starting at temperatures around 1250 °C, well below the significant interaction temperatures of UO2-Zry. These interactions or reactions potentially influence the progress of fuel degradation in the early phase. The steam-starved conditions, which are being discussed as a likely scenario at the FDNPS accident, highly influence the individual interactions and potentially lead the fuel degradation in non-prototypical directions. The detailed phenomenology of individual interactions and their influence on the transient and on the late phase of the severe accidents are also discussed.

Delivering high performance BWR fuel reliably

International Nuclear Information System (INIS)

Schardt, J.F.

1998-01-01

Utilities are under intense pressure to reduce their production costs in order to compete in the increasingly deregulated marketplace. They need fuel, which can deliver high performance to meet demanding operating strategies. GE's latest BWR fuel design, GE14, provides that high performance capability. GE's product introduction process assures that this performance will be delivered reliably, with little risk to the utility. (author)

An optimized BWR fuel lattice for improved fuel utilization

International Nuclear Information System (INIS)

Bernander, O.; Helmersson, S.; Schoen, C.G.

1984-01-01

Optimization of the BWR fuel lattice has evolved into the water cross concept, termed ''SVEA'', whereby the improved moderation within bundles augments reactivity and thus improves fuel cycle economy. The novel design introduces into the assembly a cruciform and double-walled partition containing nonboiling water, thus forming four subchannels, each of which holds a 4x4 fuel rod bundle. In Scandinavian BWRs - for which commercial SVEA reloads are now scheduled - the reactivity gain is well exploited without adverse impact in other respects. In effect, the water cross design improves both mechanical and thermal-hydraulic performance. Increased average burnup is also promoted through achieving flatter local power distributions. The fuel utilization savings are in the order of 10%, depending on the basis of comparison, e.g. choice of discharge burnup and lattice type. This paper reviews the design considerations and the fuel utilization benefits of the water cross fuel for non-Scandinavian BWRs which have somewhat different core design parameters relative to ASEA-ATOM reactors. For one design proposal, comparisons are made with current standard 8x8 fuel rod bundles as well as with 9x9 type fuel in reactors with symmetric or asymmetric inter-assembly water gaps. The effect on reactivity coefficients and shutdown margin are estimated and an assessment is made of thermal-hydraulic properties. Consideration is also given to a novel and advantageous way of including mixed-oxide fuel in BWR reloads. (author)

BWR fuel experience with zinc injection

International Nuclear Information System (INIS)

Levin, H.A.; Garcia, S.E.

1995-01-01

In 1982 a correlation between low primary recirculation system dose rates in BWR's and the presence of ionic zinc in reactor water was identified. The source of the zinc was primarily from Admiralty brass condensers. Plants with brass condensers are called ''natural zinc'' plants. Brass condensers were also a source of copper that was implicated in crude induced localized corrosion (CILC) fuel failures. In 1986 the first BWR intentionally injected zinc for the benefits of dose rate control. Although zinc alone was never implicated in fuel degradation of failures, a comprehensive fuel surveillance program was initiated to monitor fuel performance. Currently there are 14 plants that are injecting zinc. Six of these plants are also on hydrogen water chemistry. This paper describes the effect on both Zircaloy corrosion and the cruding characteristics as a result of these changes in water chemistry. Fuel rod corrosion was found to be independent of the specific water chemistry of the plants. The corrosion behavior was the same with the additions of zinc alone or zinc plus hydrogen and well within the operating experience for fuel without either of these additions. No change was observed in the amounts of crude deposited on the fuel rods, both for the adherent and loosely held deposits. One of the effects of the zinc addition was the trend to form more of the zinc rich iron spinel in the fuel deposits rather than the hematite deposits that are predominantly formed with non additive water chemistry

Characteristics of hydride precipitation and reorientation in spent-fuel cladding

International Nuclear Information System (INIS)

Chung, H. M.; Strain, R. V.; Billone, M. C.

2000-01-01

The morphology, number density, orientation, distribution, and crystallographic aspects of Zr hydrides in Zircaloy fuel cladding play important roles in fuel performance during all phases before and after discharge from the reactor, i.e., during normal operation, transient and accident situations in the reactor, temporary storage in a dry cask, and permanent storage in a waste repository. In the past, partly because of experimental difficulties, hydriding behavior in irradiated fuel cladding has been investigated mostly by optical microscopy (OM). In the present study, fundamental metallurgical and crystallographic characteristics of hydride precipitation and reorientation were investigated on the microscopic level by combined techniques of OM and transmission electron and scanning electron microscopy (TEM and SEM) of spent-fuel claddings discharged from several boiling and pressurized water reactors (BWRs and PWRs). Defueled sections of standard and Zr-lined Zircaloy-2 fuel claddings, irradiated to fluences of ∼3.3 x 10 21 n cm -2 and ∼9.2 x 10 21 n cm -2 (E > 1 MeV), respectively, were obtained from spent fuel rods discharged from two BWRs. Sections of standard and low-tin Zircaloy-4 claddings, irradiated to fluences of ∼4.4 x 10 21 n cm -2 , ∼5.9 x 10 21 n cm -2 , and ∼9.6 x 10 21 n cm -2 (E > 1 MeV) in three PWRs, were also obtained. Microstructural characteristics of hydrides were analyzed in as-irradiated condition and after gas-pressurization-burst or expanding-mandrel tests at 292-325 C in Ar for some of the spent-fuel claddings. Analyses were also conducted of hydride habit plane, morphology, and reorientation characteristics on unirradiated Zircaloy-4 cladding that contained dense radial hydrides. Reoriented hydrides in the slowly cooled unirradiated cladding were produced by expanding-mandrel loading

Delivering high performance BWR fuel reliably

Energy Technology Data Exchange (ETDEWEB)

Schardt, J.F. [GE Nuclear Energy, Wilmington, NC (United States)

1998-07-01

Utilities are under intense pressure to reduce their production costs in order to compete in the increasingly deregulated marketplace. They need fuel, which can deliver high performance to meet demanding operating strategies. GE's latest BWR fuel design, GE14, provides that high performance capability. GE's product introduction process assures that this performance will be delivered reliably, with little risk to the utility. (author)

Boiling transition phenomenon in BWR fuel assemblies effect of fuel spacer shape on critical power

International Nuclear Information System (INIS)

Yamamoto, Yasushi; Morooka, Shin-ichi; Mitsutake, Toru; Yokobori, Seiichi; Kimura, Jiro.

1996-01-01

A thorough understanding of the thermal-hydraulic phenomena near fuel spacer is necessary for the accurate prediction of the critical power of BWR fuel assemblies, and is thus essential for effective developments of a new BWR fuel assembly. The main purpose of this study is to develop an accurate method for predicting the effect of spacer shapes on critical power. Tests have been conducted under actual BWR operating conditions, using an annulus flow channel consisting of a heated rod and circular-tube channel, and BWR simulated 4x4 rod bundles with heater rods unheated just upsteam of spacer. The effect of spacer shapes on critical power was predicted analytically based on the droplet deposition rate estimation. The droplet deposition rate for different spacer shapes was calculated using a single-phase flow model. The prediction results were compared with the test results for the annulus flow channel using ring-type spacers. Analytical results of critical power agreed with measured critical power from point of the effects of changes in the rod-spacer clearance and the spacer thickness on critical power. (author)

Secondary hydriding of defected zircaloy-clad fuel rods

International Nuclear Information System (INIS)

Olander, D.R.; Vaknin, S.

1993-01-01

The phenomenon of secondary hydriding in LWR fuel rods is critically reviewed. The current understanding of the process is summarized with emphasis on the sources of hydrogen in the rod provided by chemical reaction of water (steam) introduced via a primary defect in the cladding. As often noted in the literature, the role of hydrogen peroxide produced by steam radiolysis is to provide sources of hydrogen by cladding and fuel oxidation that are absent without fission-fragment irradiation of the gas. Quantitative description of the evolution of the chemical state inside the fuel rod is achieved by combining the chemical kinetics of the reactions between the gas and the fuel and cladding with the transport by diffusion of components of the gas in the gap. The chemistry-gas transport model provides the framework into which therate constants of the reactions between the gases in the gap and the fuel and cladding are incorporated. The output of the model calculation is the H 2 0/H 2 ratio in the gas and the degree of claddingand fuel oxidation as functions of distance from the primary defect. This output, when combined with a criterion for the onset of massive hydriding of the cladding, can provide a prediction of the time and location of a potential secondary hydriding failure. The chemistry-gas transport model is the starting point for mechanical and H-in-Zr migration analyses intended to determine the nature of the cladding failure caused by the development of the massive hydride on the inner wall

FEASIBILITY OF RECYCLING PLUTONIUM AND MINOR ACTINIDES IN LIGHT WATER REACTORS USING HYDRIDE FUEL

International Nuclear Information System (INIS)

Greenspan, Ehud; Todreas, Neil; Taiwo, Temitope

2009-01-01

The objective of this DOE NERI program sponsored project was to assess the feasibility of improving the plutonium (Pu) and minor actinide (MA) recycling capabilities of pressurized water reactors (PWRs) by using hydride instead of oxide fuels. There are four general parts to this assessment: (1) Identifying promising hydride fuel assembly designs for recycling Pu and MAs in PWRs; (2) Performing a comprehensive systems analysis that compares the fuel cycle characteristics of Pu and MA recycling in PWRs using the promising hydride fuel assembly designs identified in Part 1 versus using oxide fuel assembly designs; (3) Conducting a safety analysis to assess the likelihood of licensing hydride fuel assembly designs; and (4) Assessing the compatibility of hydride fuel with cladding materials and water under typical PWR operating conditions Hydride fuel was found to offer promising transmutation characteristics and is recommended for further examination as a possible preferred option for recycling plutonium in PWRs

FEASIBILITY OF RECYCLING PLUTONIUM AND MINOR ACTINIDES IN LIGHT WATER REACTORS USING HYDRIDE FUEL

Energy Technology Data Exchange (ETDEWEB)

Greenspan, Ehud; Todreas, Neil; Taiwo, Temitope

2009-03-10

The objective of this DOE NERI program sponsored project was to assess the feasibility of improving the plutonium (Pu) and minor actinide (MA) recycling capabilities of pressurized water reactors (PWRs) by using hydride instead of oxide fuels. There are four general parts to this assessment: 1) Identifying promising hydride fuel assembly designs for recycling Pu and MAs in PWRs 2) Performing a comprehensive systems analysis that compares the fuel cycle characteristics of Pu and MA recycling in PWRs using the promising hydride fuel assembly designs identified in Part 1 versus using oxide fuel assembly designs 3) Conducting a safety analysis to assess the likelihood of licensing hydride fuel assembly designs 4) Assessing the compatibility of hydride fuel with cladding materials and water under typical PWR operating conditions Hydride fuel was found to offer promising transmutation characteristics and is recommended for further examination as a possible preferred option for recycling plutonium in PWRs.

Transmutation of minor actinide using BWR fueled mixed oxide

International Nuclear Information System (INIS)

Susilo, Jati

2000-01-01

Nuclear spent fuel recycle has a strategic importance in the aspect of nuclear fuel economy and prevention of its spread-out. One among other application of recycle is to produce mixed oxide fuel (Mo) namely mixed Plutonium and uranium oxide. As for decreasing the burden of nuclear high level waste (HLW) treatment, transmutation of minor actinide (MA) that has very long half life will be carried out by conversion technique in nuclear reactor. The purpose of this study was to know influence of transition fuel cell regarding the percent weight of transmutation MA in the BWR fueled MOX. Calculation of cell BWR was used SRAC computer code, with assume that the reactor in equilibrium. The percent weight of transmutation MA to be optimum by increasing the discharge burn-up of nuclear fuel, raising ratio of moderator to fuel volume (Vm/Vf), and loading MA with percent weight about 3%-6% and also reducing amount of percent weight Pu in MOX fuel. For mixed fuel standard reactor, reactivity value were obtained between about -50pcm ∼ -230pcm for void coefficient and -1.8pcm ∼ -2.6pcm for fuel temperature coefficient

BWR Fuel Assemblies Physics Analysis Utilizing 3D MCNP Modeling

International Nuclear Information System (INIS)

Chiang, Ren-Tai; Williams, John B.; Folk, Ken S.

2008-01-01

MCNP is used to model a partially controlled BWR fresh fuel four assemblies (2x2) system for better understanding BWR fuel behavior and for benchmarking production codes. The impact of the GE14 plenum regions on axial power distribution is observed by comparing against the GE13 axial power distribution, in which the GE14 relative power is lower than the GE13 relative power at the 15. node and at the 16. node due to presence of the plenum regions in GE14 fuel in these two nodes. The segmented rod power distribution study indicates that the azimuthally dependent power distribution is very significant for the fuel rods next to the water gap in the uncontrolled portion. (authors)

BWR Fuel Assemblies Physics Analysis Utilizing 3D MCNP Modeling

Energy Technology Data Exchange (ETDEWEB)

Chiang, Ren-Tai [University of Florida, Gainesville, Florida 32611 (United States); Williams, John B.; Folk, Ken S. [Southern Nuclear Company, Birmingham, Alabama 35242 (United States)

2008-07-01

MCNP is used to model a partially controlled BWR fresh fuel four assemblies (2x2) system for better understanding BWR fuel behavior and for benchmarking production codes. The impact of the GE14 plenum regions on axial power distribution is observed by comparing against the GE13 axial power distribution, in which the GE14 relative power is lower than the GE13 relative power at the 15. node and at the 16. node due to presence of the plenum regions in GE14 fuel in these two nodes. The segmented rod power distribution study indicates that the azimuthally dependent power distribution is very significant for the fuel rods next to the water gap in the uncontrolled portion. (authors)

Nanoindentation measurements of the mechanical properties of zirconium matrix and hydrides in unirradiated pre-hydrided nuclear fuel cladding

International Nuclear Information System (INIS)

Rico, A.; Martin-Rengel, M.A.; Ruiz-Hervias, J.; Rodriguez, J.; Gomez-Sanchez, F.J.

2014-01-01

It is well known that the mechanical properties of the nuclear fuel cladding may be affected by the presence of hydrides. The average mechanical properties of hydrided cladding have been extensively investigated from a macroscopic point of view. In addition, the mechanical and fracture properties of bulk hydride samples fabricated from zirconium plates have also been reported. In this paper, Young’s modulus, hardness and yield stress are measured for each phase, namely zirconium hydrides and matrix, of pre-hydrided nuclear fuel cladding. To this end, nanoindentation tests were performed on ZIRLO samples in as-received state, on a hydride blister and in samples with 150 and 1200 ppm of hydrogen homogeneously distributed along the hoop direction of the cladding. The results show that the measured mechanical properties of the zirconium hydrides and ZIRLO matrix (Young’s modulus, hardness and yield stress) are rather similar. From the experimental data, the hydride volume fraction in the cladding samples with 150 and 1200 ppm was estimated and the average mechanical properties were calculated by means of the rule of mixtures. These values were compared with those obtained from ring compression tests. Good agreement between the results obtained by both methods was found

Control in fabrication of PWR and BWR type reactor fuel elements

International Nuclear Information System (INIS)

Gorskij, V.V.

1981-01-01

Both destructive and non-destructive testing methods now in use in fabrication of BWR and PWR type reactor fuel elements at foreign plants are reviewed. Technological procedures applied in fabrication of fuel elements and fuel assemblies are described. Major attention is paid to radiographic, ultrasonic, metallographic, visual and autoclavic testings. A correspondence of the methods applied to the ASTM standards is discussed. The most part of the countries are concluded the apply similar testing methods enabling one to reliably evaluate the quality of primary materials and fabricated fuel elements and thus meeting the demands to contemporary PWR and BWR type reactor fuel elements. Practically all fuel element and pipe fabrication plants in Western Europe, Asia and America use the ASTM standards as the basis for the quality contr [ru

Irradiation effects on thermal properties of LWR hydride fuel

Energy Technology Data Exchange (ETDEWEB)

Terrani, Kurt, E-mail: terrani@berkeley.edu [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States); Balooch, Mehdi [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States); Carpenter, David; Kohse, Gordon [Massachusetts Institute of Technology, 138 Albany St., Cambridge, MA 02139 (United States); Keiser, Dennis; Meyer, Mitchell [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Olander, Donald [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States)

2017-04-01

Three hydride mini-fuel rods were fabricated and irradiated at the MIT nuclear reactor with a maximum burnup of 0.31% FIMA or ∼5 MWd/kgU equivalent oxide fuel burnup. Fuel rods consisted of uranium-zirconium hydride (U (30 wt%)ZrH{sub 1.6}) pellets clad inside a LWR Zircaloy-2 tubing. The gap between the fuel and the cladding was filled with lead-bismuth eutectic alloy to eliminate the gas gap and the large temperature drop across it. Each mini-fuel rod was instrumented with two thermocouples with tips that are axially located halfway through the fuel centerline and cladding surface. In-pile temperature measurements enabled calculation of thermal conductivity in this fuel as a function of temperature and burnup. In-pile thermal conductivity at the beginning of test agreed well with out-of-pile measurements on unirradiated fuel and decreased rapidly with burnup.

Power ramp tests of BWR-MOX fuels

International Nuclear Information System (INIS)

Asahi, K.; Oguma, M.; Higuchi, S.; Kamimua, K.; Shirai, Y.; Bodart, S.; Mertens, L.

1996-01-01

Power ramp test of BWR-MOX and UO 2 fuel rods base irradiated up to about 60 GWd/t in Dodewaard reactor have been conducted in BR2 reactor in the framework of the international DOMO programme. The MOX pellets were provided by BN (MIMAS process) and PNC (MH method). The MOX fuel rods with Zr-liner and non-liner cladding and the UO 2 fuel rods with Zr-liner cladding remained intact during the stepwise power ramp tests to about 600 W/cm, even at about 60 GWd/t

IFPE/IFA-432, Fission Gas Release, Mechanical Interaction BWR Fuel Rods, Halden

International Nuclear Information System (INIS)

Turnbull, J.A.

1996-01-01

Description: It contains data from experiments that have been performed at the IFE/OECD Halden Reactor Project, available for use in fuel performance studies. It covers experiments on thermal performance, fission product release, clad properties and pellet clad mechanical interaction. It includes also experimental data relevant to high burn-up behaviour. IFA-432: Measurements of fuel temperature response, fission gas release and mechanical interaction on BWR-type fuel rods up to high burn-ups. The assembly featured several variations in rod design parameters, including fuel type, fuel/cladding gap size, fill gas composition (He and Xe) and fuel stability. It contained 6 BWR-type fuel rods with fuel centre thermocouples at two horizontal planes, rods were also equipped with pressure transducers and cladding extensometers. Only data from 6 rods are compiled here

BWR SFAT, gross-defect verification of spent BWR fuel. Final report on Task FIN A563 on the Finnish Support Programme to IAEA Safeguards including BWR SFAT User Manual

International Nuclear Information System (INIS)

Tarvainen, M.; Paakkunainen, M.; Tiitta, A.; Sarparanta, K.

1994-04-01

A measurement instrument called Spent Fuel Attribute Tester, SFAT, has been designed, fabricated and taken into use by the IAEA in gross defect verification of spent BWR fuel assemblies. The equipment consists of an underwater measurement head connected with cables to a control unit on the bridge of the fuel handling machine as well as to a PMCA for measurement of the gamma spectra. The BWR SFAT is optimized for the AFR interim storage, TVO KPA-STORE, of the TVO Power Company in Olkiluoto, Finland. It has a shape and it is moved like a fuel assembly using the fuel handling machine. No fuel movements are needed. Spent fuel specific radiation from the fission product 137 Cs at the gamma-ray energy of 662 keV is detected above the assemblies in the storage rack using a NaI(Tl) detector. In the design and in licensing the requirements of the IAEA, operator and the safety authority have been taken into account. The BWR SFAT allows modifications for other LWR fuel types with minor changes. The work has been carried out under the task FIN A 563 of the Finnish Support Programme to IAEA Safeguards. (orig.) (9 refs., 22 figs.)

BWR 9 X 9 Fuel Assembly Thermal-Hydraulic Tests (2): Hydraulic Vibration Test

International Nuclear Information System (INIS)

Yoshiaki Tsukuda; Katsuichiro Kamimura; Toshiitsu Hattori; Akira Tanabe; Noboru Saito; Masahiko Warashina; Yuji Nishino

2002-01-01

Nuclear Power Engineering Corporation (NUPEC) conducted thermal-hydraulic projects for verification of thermal-hydraulic design reliability for BWR high-burnup 8 x 8 and 9 x 9 fuel assemblies, entrusted by the Ministry of Economy, Trade and Industry (METI). As a part of the NUPEC thermal-hydraulic projects, hydraulic vibration tests using full-scale test assemblies simulating 9 x 9 fuel assemblies were carried out to evaluate BWR fuel integrity. The test data were applied to development of a new correlation for the estimation of fuel rod vibration amplitude. (authors)

Development of alternative materials for BWR fuel springs

International Nuclear Information System (INIS)

Uruma, Y.; Osato, T.; Yamazaki, K.

2002-01-01

Major sources of radioactivity introduced into reactor water of BWR were estimated fuel crud and in-core materials (especially, fuel springs). Fuel springs are used for fixation of fuel cladding tubes with spacer grid. Those are small parts (total length is only within 25 mm) and so many numbers are loaded simultaneously and then total surfaces area are calculated up to about 200 m 2 . Fuel springs are located under high radiation field and high oxidative environment. Conventional fuel spring is made of alloy-X750 which is one of nickel-based alloy and is reported to show relatively higher corrosion release rate. 58 Co and 60 Co will be released directly into reactor water from intensely radio-activated fuel springs surface and increase radioactivity concentrations in primary coolant. Corrosion release control from fuel springs is an important technical item and a development of alternative material instead of alloy-X750 for fuel spring is a key subject to achieve ultra low man-rem exposure BWR plant. In present work, alloy-X718 which started usage for PWR fuel springs and stainless steel type 316L which has many mechanical property data are picked up for alternative materials and compared their corrosion behaviors with conventional material. Corrosion experiment was conducted under vapor-water two phases flow which is simulated fuel cladding surface boiling condition. After exposure, corrosion film formed under corrosion test was analyzed in detail and corrosion film amount and corrosion release amount are estimated among three materials. (authors)

Study on thermal performance and margins of BWR fuel elements

International Nuclear Information System (INIS)

Stosic, Zoran

1999-01-01

This paper contributes to developing a methodology of predicting and analyzing thermal performance and margins of Boiling Water Reactor (BWR) fuel assemblies under conditions of reaching high quality Boiling Crisis and subsequent post-dryout thermal hydraulics causing temperature excursion of fuel cladding. Operational margins against dryout and potential for increasing fuel performance with appropriate benefits are discussed. The philosophy of modeling with its special topics are demonstrated on the HECHAN (HEated CHannel ANalyzer) model as the state-of-art for thermal-hydraulics analysis of BWR fuel assemblies in pre- and post-dryout two-phase flow regimes. The scope of further work either being or has to be performed concerning implementation of new physical aspects, including domain extension of HECHAN model applications to the Pressurized Water Reactors (PWRs), is discussed. Finally, a comprehensive overview of the literature dealing with development of the model is given. (author)

Manufacturing technology and process for BWR fuel

International Nuclear Information System (INIS)

Kato, Shigeru

1996-01-01

Following recent advanced technologies, processes and requests of the design changes of BWR fuel, Nuclear Fuel Industries, Ltd. (NFI) has upgraded the manufacturing technology and honed its own skills to complete its brand-new automated facility in Tokai in the latter half of 1980's. The plant uses various forms of automation throughout the manufacturing process: the acceptance of uranium dioxide powder, pelletizing, fuel rod assembling, fuel bundle assembling and shipment. All processes are well computerized and linked together to establish the integrated control system with three levels of Production and Quality Control, Process Control and Process Automation. This multi-level system plays an important role in the quality assurance system which generates the highest quality of fuels and other benefits. (author)

Computational fluid dynamics modeling of two-phase flow in a BWR fuel assembly

International Nuclear Information System (INIS)

Andrey Ioilev; Maskhud Samigulin; Vasily Ustinenko; Simon Lo; Adrian Tentner

2005-01-01

Full text of publication follows: The goal of this project is to develop an advanced Computational Fluid Dynamics (CFD) computer code (CFD-BWR) that allows the detailed analysis of the two-phase flow and heat transfer phenomena in a Boiling Water Reactor (BWR) fuel bundle under various operating conditions. This code will include more fundamental physical models than the current generation of sub-channel codes and advanced numerical algorithms for improved computational accuracy, robustness, and speed. It is highly desirable to understand the detailed two-phase flow phenomena inside a BWR fuel bundle. These phenomena include coolant phase changes and multiple flow regimes which directly influence the coolant interaction with fuel assembly and, ultimately, the reactor performance. Traditionally, the best analysis tools for the analysis of two-phase flow phenomena inside the BWR fuel assembly have been the sub-channel codes. However, the resolution of these codes is still too coarse for analyzing the detailed intra-assembly flow patterns, such as flow around a spacer element. Recent progress in Computational Fluid Dynamics (CFD), coupled with the rapidly increasing computational power of massively parallel computers, shows promising potential for the fine-mesh, detailed simulation of fuel assembly two-phase flow phenomena. However, the phenomenological models available in the commercial CFD programs are not as advanced as those currently being used in the sub-channel codes used in the nuclear industry. In particular, there are no models currently available which are able to reliably predict the nature of the flow regimes, and use the appropriate sub-models for those flow regimes. The CFD-BWR code is being developed as a customized module built on the foundation of the commercial CFD Code STAR-CD which provides general two-phase flow modeling capabilities. The paper describes the model development strategy which has been adopted by the development team for the

BWROPT: A multi-cycle BWR fuel cycle optimization code

Energy Technology Data Exchange (ETDEWEB)

Ottinger, Keith E.; Maldonado, G. Ivan, E-mail: Ivan.Maldonado@utk.edu

2015-09-15

Highlights: • A multi-cycle BWR fuel cycle optimization algorithm is presented. • New fuel inventory and core loading pattern determination. • The parallel simulated annealing algorithm was used for the optimization. • Variable sampling probabilities were compared to constant sampling probabilities. - Abstract: A new computer code for performing BWR in-core and out-of-core fuel cycle optimization for multiple cycles simultaneously has been developed. Parallel simulated annealing (PSA) is used to optimize the new fuel inventory and placement of new and reload fuel for each cycle considered. Several algorithm improvements were implemented and evaluated. The most significant of these are variable sampling probabilities and sampling new fuel types from an ordered array. A heuristic control rod pattern (CRP) search algorithm was also implemented, which is useful for single CRP determinations, however, this feature requires significant computational resources and is currently not practical for use in a full multi-cycle optimization. The PSA algorithm was demonstrated to be capable of significant objective function reduction and finding candidate loading patterns without constraint violations. The use of variable sampling probabilities was shown to reduce runtime while producing better results compared to using constant sampling probabilities. Sampling new fuel types from an ordered array was shown to have a mixed effect compared to random new fuel type sampling, whereby using both random and ordered sampling produced better results but required longer runtimes.

Protecting AREVA ATRIUM™ BWR fuel from debris fretting failure

International Nuclear Information System (INIS)

Cole, Steven E.; Garner, Norman L.; Lippert, Hans-Joachim; Graebert, Rüdiger; Mollard, Pierre; Hahn, Gregory C.

2014-01-01

Historically, debris fretting has been the leading cause of fuel rod failure in BWR fuel assemblies, costing the industry millions of dollars in lost generation and negatively impacting the working area of plant site personnel. In this paper the focus will be on recent BWR fuel product innovation designed to eliminate debris related failures. Experience feedback from more than three decades of operation history with non-line-of-sight FUELGUARD™ lower tie plate debris filters will be presented. The development and relative effectiveness of successive generations of filtration technology will be discussed. It will be shown that modern, state of the art debris filters are an effective defense against debris fretting failure. Protective measures extend beyond inlet nozzle debris filters. The comprehensive debris resistance features built into AREVA’s newest fuel design, the ATRIUM™ 11, reduce the overall risk of debris entrapment as well as providing a degree of protection from debris that may fall down on the fuel assembly from above, e.g., during refueling operations. The positive recent experience in a debris sensitive plant will be discussed showing that the combination of advanced fuel technology and a robust foreign material exclusion program at the reactor site can eliminate the debris fretting failure mechanism. (author)

The mechanical structure of the SVEA BWR fuel

International Nuclear Information System (INIS)

Nylund, O.; Johansson, A.; Junkrans, S.

1985-01-01

The SVEA BWR fuel assembly design is characterized by a double-wall cruciform internal structure forming an internal water gap and dividing the assembly into 4 subbundles. The effect is a favourable distribution of fuel and moderator, a minimum amount of structural material in active core, a combination of structural stability and flexibility for minimum control rod friction in reduced gaps and a reduced creep deformation of the fuel assembly. The results of a laboratory test program confirm the much lower friction force obtained with the SVEA fuel assemblies while withdrawing and inserting the control rod. (RF)

Advanced chemical hydride-based hydrogen generation/storage system for fuel cell vehicles

Energy Technology Data Exchange (ETDEWEB)

Breault, R.W.; Rolfe, J. [Thermo Power Corp., Waltham, MA (United States)

1998-08-01

Because of the inherent advantages of high efficiency, environmental acceptability, and high modularity, fuel cells are potentially attractive power supplies. Worldwide concerns over clean environments have revitalized research efforts on developing fuel cell vehicles (FCV). As a result of intensive research efforts, most of the subsystem technology for FCV`s are currently well established. These include: high power density PEM fuel cells, control systems, thermal management technology, and secondary power sources for hybrid operation. For mobile applications, however, supply of hydrogen or fuel for fuel cell operation poses a significant logistic problem. To supply high purity hydrogen for FCV operation, Thermo Power`s Advanced Technology Group is developing an advanced hydrogen storage technology. In this approach, a metal hydride/organic slurry is used as the hydrogen carrier and storage media. At the point of use, high purity hydrogen will be produced by reacting the metal hydride/organic slurry with water. In addition, Thermo Power has conceived the paths for recovery and regeneration of the spent hydride (practically metal hydroxide). The fluid-like nature of the spent hydride/organic slurry will provide a unique opportunity for pumping, transporting, and storing these materials. The final product of the program will be a user-friendly and relatively high energy storage density hydrogen supply system for fuel cell operation. In addition, the spent hydride can relatively easily be collected at the pumping station and regenerated utilizing renewable sources, such as biomass, natural, or coal, at the central processing plants. Therefore, the entire process will be economically favorable and environmentally friendly.

Manufacturing and investigation of U-Mo LEU fuel granules by hydride-dehydride processing

International Nuclear Information System (INIS)

Stetskiy, Y.A.; Trifonov, Y.I.; Mitrofanov, A.V.; Samarin, V.I.

2002-01-01

Investigations of hydride-dehydride processing for comminution of U-Mo alloys with Mo content in the range 1.9/9.2% have been performed. Some regularities of the process as a function of Mo content have been determined as well as some parameters elaborated. Hydride-dehydride processing has been shown to provide necessary phase and chemical compositions of U-Mo fuel granules to be used in disperse fuel elements for research reactors. Pin type disperse mini-fuel elements for irradiation tests in the loop of 'MIR' reactor (Dmitrovgrad) have been fabricated using U-Mo LEU fuel granules obtained by hydride-dehydride processing. Irradiation tests of these mini-fuel elements loaded to 4 g U tot /cm 3 are planned to start by the end of this year. (author)

Prediction of droplet deposition around BWR fuel spacer by FEM flow analysis

International Nuclear Information System (INIS)

Yamamoto, Yasushi; Morooka, Shinichi

1997-01-01

The critical power of the BWR fuel assembly has been remarkably increased. That increase mainly depends on the improvement of the spacer which keeps fixed gaps between fuel rods. So far, these improvements have been carried out on the basis of what developers consider to be appropriate and the results of mockup tests of the BWR fuel assembly. However, continued reliance on these approaches for the development of a higher performance fuel assembly will prove time-consuming and costly. Therefore, it is hoped that the spacer effects for the critical power can be investigated by computer simulation, and it is significantly important to develop the critical power prediction method. Direct calculation of the two-phase flow in a BWR fuel channel s still difficult. Accordingly, a new method for predicting the critical power was proposed. Our method consists of CFD (computer fluid dynamics) code based on the single-phase flow analysis method and the subchannel analysis code. To verify our method, the critical power predictions for various spacer geometries were performed. The predicted results of the critical power were compared with the experimental data. The result of the comparison showed a good agreement and the applicability of our method for various spacer geometries. (author)

Internal hydriding in irradiated defected Zircaloy fuel rods: A review (LWBR Development Program)

International Nuclear Information System (INIS)

Clayton, J.C.

1987-10-01

Although not a problem in recent commercial power reactors, including the Shippingport Light Water Breeder Reactor, internal hydriding of Zircaloy cladding was a persistent cause of gross cladding failures during the 1960s. It occurred in the fuel rods of water-cooled nuclear power reactors that had a small cladding defect. This report summarizes the experimental findings, causes, mechanisms, and methods of minimizing internal hydriding in defected Zircaloy-clad fuel rods. Irradiation test data on the different types of defected fuel rods, intentionally fabricated defected and in-pile operationally defected rods, are compared. Significant factors affecting internal hydriding in defected Zircaloy-clad fuel rods (defect hole size, internal and external sources of hydrogen, Zircaloy cladding surface properties, nickel alloy contamination of Zircaloy, the effect of heat flux and fluence) are discussed. Pertinent in-pile and out-of-pile test results from Bettis and other laboratories are used as a data base in constructing a qualitative model which explains hydrogen generation and distribution in Zircaloy cladding of defected water-cooled reactor fuel rods. Techniques for minimizing internal hydride failures in Zircaloy-clad fuel rods are evaluated

AREVA 10x10 BWR fuel experience feedback and on going upgrading

International Nuclear Information System (INIS)

Lippert, Hans Joachim; Rentmeister, Thomas; Garner, Norman; Tandy, Jay; Mollard, Pierre

2008-01-01

Established with engineering and manufacturing operations in the US and Europe, AREVA NP has been and is supplying nuclear fuel assemblies and associated core components to boiling water reactors worldwide, representing today more than 63 000 fuel assemblies. The evolution of BWR fuel rod arrays from early 6x6 designs to the 10x10 designs first introduced in the mid 1990's yielded significant improvements in thermal mechanical operating limits, critical power level, cold shutdown margin, discharge burnup, as well as other key operational capabilities. Since first delivered in 1992, ATRIUM T M 1 0 fuel assemblies have now been supplied to a total of 32 BWR plants in the US, Europe, and Asia resulting in an operating experience over 20 000 fuel assemblies. This article presents in detail the operational experience consolidated by these more than 20 000 ATRIUM T M 1 0 BWR assemblies already supplied to utilities. Within the different 10x10 fuel assemblies available, the Fuel Assembly design is chosen and tailored to the operating strategies of each reactor. Among them, the latest versions of ATRIUM T M a re ATRIUM T M 1 0XP and ATRIUM T M 1 0XM fuel assemblies which have been delivered to several utilities worldwide. The article details key aspects of ATRIUM T M 1 0 fuel assemblies in terms of reliability and performance. Special attention is paid to key proven features, ULTRAFLOW T M s pacer grids, the use of part length fuel rods (PLFRs) and their geometrical optimization, water channel and load chain, upgraded features available for inclusion with most advanced designs. Regular upgrading of the product has been made possible thanks to a continuous improvement process with the aim of further upgrading BWR fuel assembly performance and reliability. Regarding thermal mechanical behavior of fuel rods, chromia (Cr2O3) doped fuel pellets, described in Reference 1, well illustrate this improvement strategy to reduce fission gas release, increase power thresholds for PCI

BWR fuel clad behaviour following LOCA

International Nuclear Information System (INIS)

Chaudhry, S.M.; Vyas, K.N.; Dinesh Babu, R.

1996-01-01

Flow and pressure through the fuel coolant channel reduce rapidly following a loss of coolant accident. Due to stored energy and decay heat, fuel and cladding temperatures rise rapidly. Increase in clad temperature causes deterioration of mechanical properties of clad material. This coupled with increase of pressure inside the cladding due to accumulation of fission gases and de-pressurization of coolant causes the cladding to balloon. This phenomenon is important as it can reduce or completely block the flow passages in a fuel assembly causing reduction of emergency coolant flow. Behaviour of a BWR clad is analyzed in a design basis LOCA. Fuel and clad temperatures following a LOCA are calculated. Fission gas release and pressure is estimated using well established models. An elasto-plastic analysis of clad tube is carried out to determine plastic strains and corresponding deformations using finite-element technique. Analysis of neighbouring pins gives an estimate of flow areas available for emergency coolant flow. (author). 7 refs, 6 figs, 3 tabs

A deformation and thermodynamic model for hydride precipitation kinetics in spent fuel cladding

International Nuclear Information System (INIS)

Stout, R.B.

1989-10-01

Hydrogen is contained in the Zircaloy cladding of spent fuel rods from nuclear reactors. All the spent fuel rods placed in a nuclear waste repository will have a temperature history that decreases toward ambient; and as a result, most all of the hydrogen in the Zircaloy will eventually precipitate as zirconium hydride platelets. A model for the density of hydride platelets is a necessary sub-part for predicting Zircaloy cladding failure rate in a nuclear waste repository. A model is developed to describe statistically the hydride platelet density, and the density function includes the orientation as a physical attribute. The model applies concepts from statistical mechanics to derive probable deformation and thermodynamic functionals for cladding material response that depend explicitly on the hydride platelet density function. From this model, hydride precipitation kinetics depend on a thermodynamic potential for hydride density change and on the inner product of a stress tensor and a tensor measure for the incremental volume change due to hydride platelets. The development of a failure response model for Zircaloy cladding exposed to the expected conditions in a nuclear waste repository is supported by the US DOE Yucca Mountain Project. 19 refs., 3 figs

Results of modeling advanced BWR fuel designs using CASMO-4

International Nuclear Information System (INIS)

Knott, D.; Edenius, M.

1996-01-01

Advanced BWR fuel designs from General Electric, Siemens and ABB-Atom have been analyzed using CASMO-4 and compared against fission rate distributions and control rod worths from MCNP. Included in the analysis were fuel storage rack configurations and proposed mixed oxide (MOX) designs. Results are also presented from several cycles of SIMULATE-3 core follow analysis, using nodal data generated by CASMO-4, for cycles in transition from 8x8 designs to advanced fuel designs. (author)

Design criteria for confidence in the manufacture of BWR fuel rods

International Nuclear Information System (INIS)

Anantharaman, K.; Basu, S.; Anand, A.K.; Mehta, S.K.

Based on the experience of fuel manufacture for BWR type reactors in India, the parameters which need stringent quality control, are discussed. The design specifications of the fuel rods as well as the cladding material and tubes are reported. The defect mechanisms to be taken into account and the fuel failure in reference to the variation of mechanical properties of the cladding are also described. (K.B.)

Optimization of axial enrichment and gadolinia distributions for BWR fuel under control rod programming, (2)

International Nuclear Information System (INIS)

Hida, Kazuki; Yoshioka, Ritsuo

1992-01-01

A method has been developed for optimizing the axial enrichment and gadolinia distributions for the reload BWR fuel under control rod programming. The problem was to minimize the enrichment requirement subject to the criticality and axial power peaking constraints. The optimization technique was based on the successive linear programming method, each linear programming problem being solved by a goal programming algorithm. A rapid and practically accurate core neutronics model, named the modified one-dimensional core model, was developed to describe the batch-averaged burnup behavior of the reload fuel. A core burnup simulation algorithm, employing a burnup-power-void iteration, was also developed to calculate the rigorous equilibrium cycle performance. This method was applied to the optimization of axial two- and 24-region fuels for demonstrative purposes. The optimal solutions for both fuels have proved the optimality of what is called burnup shape optimization spectral shift. For the two-region fuel with a practical power peaking of 1.4, the enrichment distribution was nearly uniform, because a bottom-peaked burnup shape flattens the axial power shape. Optimization of the 24-region fuel has shown a potential improvement in BWR fuel cycle economics, which will guide future advancement in BWR fuel designs. (author)

Technology developments for Japanese BWR MOX fuel utilization

International Nuclear Information System (INIS)

Oguma, M.; Mochida, T.; Nomata, T.; Asahi, K.

1997-01-01

The Long-Term Program for Research, Development and Utilization of Nuclear Energy established by the Atomic Energy Commission of Japan asserts that Japan will promote systematic utilization of MOX fuel in LWRs. Based on this Japanese nuclear energy policy, we have been pushing development of MOX fuel technology aimed at future full scale utilization of this fuel in BWRs. In this paper, the main R and D topics are described from three subject areas, MOX core and fuel design, MOX fuel irradiation behaviour, and MOX fuel fabrication technology. For the first area, we explain the compatibility of MOX fuel with UO 2 core, the feasibility of the full MOX core, and the adaptability of MOX design methods based on a mock-up criticality experiment. In the second, we outline the Tsuruga MOX irradiation program and the DOMO program, and suggest that MOX fuel behaviour is comparable to ordinary BWR UO 2 fuel behaviour. In the third, we examine the development of a fully automated MOX bundle assembling apparatus and its features. (author). 14 refs, 11 figs, 3 tabs

Composition and Distribution of Tramp Uranium Contamination on BWR and PWR Fuel Rods

International Nuclear Information System (INIS)

Schienbein, Marcel; Zeh, Peter; Hurtado, Antonio; Rosskamp, Matthias; Mailand, Irene; Bolz, Michael

2012-09-01

In a joint research project of VGB and AREVA NP GmbH the behaviour of alpha nuclides in nuclear power plants with light water reactors has been investigated. Understanding the source and the behaviour of alpha nuclides is of big importance for planning radiation protection measures for outages and upcoming dismantling projects. Previous publications have shown the correlation between plant specific alpha contamination of the core and the so called 'tramp fuel' or 'tramp uranium' level which is linked to the defect history of fuel assemblies and accordingly the amount of previously washed out fuel from defective fuel rods. The methodology of tramp fuel estimation is based on fission product concentrations in reactor coolant but also needs a good knowledge of tramp fuel composition and in-core distribution on the outer surface of fuel rods itself. Sampling campaigns of CRUD deposits of irradiated fuel assemblies in different NPPs were performed. CRUD analyses including nuclide specific alpha analysis have shown systematic differences between BWR and PWR plants. Those data combined with literature results of fuel pellet investigations led to model improvements showing that a main part of fission products is caused by fission of Pu-239 an activation product of U-238. CRUD investigations also gave a better picture of the in-core composition and distribution of the tramp uranium contamination. It was shown that the tramp uranium distribution in PWR plants is time dependent. Even new fuel assemblies will be notably contaminated after only one cycle of operation. For PWR applies the following logic: the higher the local power the higher the contamination. With increasing burnup the local rod power usually decreases leading to decreasing tramp uranium contamination on the fuel rod surface. This is not applicable for tramp uranium contamination in BWR. CRUD contamination (including the tramp fuel deposits) is much more fixed and is constantly increasing

BWR - Spent Fuel Transport and Storage with the TNTM9/4 and TNTM24BH Casks

International Nuclear Information System (INIS)

Wattez, L.; Marguerat, Y.; Hoesli, C.

2006-01-01

The Swiss Nuclear Utilities have started in 2001 to store spent fuel in dry metallic dual-purpose casks at ZWILAG, the Swiss interim storage facility. BKW FMB Energy Ltd., the Muehleberg Nuclear Power Plant owner, is involved in this process and has elected to store its BWR spent fuel in a new high capacity dual-purpose cask, the TNeTeM24BH from the COGEMA Logistics/TRANSNUCLEAR TN TM 24 family. The Muehleberg BWR spent fuels are transported by road in a medium size shuttle transport cask and then transferred to a heavy transport/storage cask (dry transfer) in the hot cell of ZWILAG site. For that purpose, COGEMA Logistics designed and supplied: - Two shuttle casks, TN TM 9/4, mainly devoted to transport of spent fuel from Muehleberg NPP to ZWILAG. Licensed according to IAEA 1996, the TN TM 9/4 is a 40 ton transport cask, for 7 BWR high bum-up spent fuel assemblies. - A series of new high capacity dual-purpose casks, TN TM 24BH, holding 69 BWR spent fuels. Two transport campaigns took place in 2003 and 2004. For each campaign, ten TN TM 9/4 round trips are performed, and one TN TM 24BH is loaded. 5 additional TN TM 24BH are being manufactured for BKW, and the next transport campaigns are scheduled from 2006. The TN TM 24BH high capacity dual purpose cask and the TN TM 9/4 transport cask characteristics and capabilities will then be detailed. (authors)

Fuel assemblies for BWR type reactors

International Nuclear Information System (INIS)

Ishizuka, Takao.

1981-01-01

Purpose: To enable effective failed fuel detection by the provision of water rod formed with a connecting section connected to a warmed water feed pipe of a sipping device at the lower portion and with a warmed water jetting port in the lower portion in a fuel assembly of a BWR type reactor to thereby carry out rapid sipping. Constitution: Fuel rods and water rods are contained in the channel box of a fuel assembly, and the water rod is provided at its upper portion with a connecting section connected to the warmed water feed pipe of the sipping device and formed at its lower portion with a warmed water jetting port for jetting warmed water fed from the warmed water feed pipe. Upon detection of failed fuels, the reactor operation is shut down and the reactor core is immersed in water. The cover for the reactor container is removed and the cap of the sipping device is inserted to connect the warmed water feed pipe to the connecting section of the water rod. Then, warmed water is fed to the water rod and jetted out from the warmed water jetting port to cause convection and unify the water of the channel box in a short time. Thereafter, specimen is sampled and analyzed for the detection of failed fuels. (Moriyama, K.)

High fidelity analysis of BWR fuel assembly with COBRA-TF/PARCS and trace codes

International Nuclear Information System (INIS)

Abarca, A.; Miro, R.; Barrachina, T.; Verdu, G.; Soler, A.

2013-01-01

The growing importance of detailed reactor core and fuel assembly description for light water reactors (LWRs) as well as the sub-channel safety analysis requires high fidelity models and coupled neutronic/thermalhydraulic codes. Hand in hand with advances in the computer technology, the nuclear safety analysis is beginning to use a more detailed thermal hydraulics and neutronics. Previously, a PWR core and a 16 by 16 fuel assembly models were developed to test and validate our COBRA-TF/PARCS v2.7 (CTF/PARCS) coupled code. In this work, a comparison of the modeling and simulation advantages and disadvantages of modern 10 by 10 BWR fuel assembly with CTF/PARCS and TRACE codes has been done. The objective of the comparison is making known the main advantages of using the sub-channel codes to perform high resolution nuclear safety analysis. The sub-channel codes, like CTF, permits obtain accurate predictions, in two flow regime, of the thermalhydraulic parameters important to safety with high local resolution. The modeled BWR fuel assembly has 91 fuel rods (81 full length and 10 partial length fuel rods) and a big square central water rod. This assembly has been modeled with high level of detail with CTF code and using the BWR modeling parameters provided by TRACE. The same neutronic PARCS's model has been used for the simulation with both codes. To compare the codes a coupled steady state has be performed. (author)

gamma-Zr-Hydride Precipitate in Irradiated Massive delta- Zr-Hydride

DEFF Research Database (Denmark)

Warren, M. R.; Bhattacharya, D. K.

1975-01-01

During examination of A Zircaloy-2-clad fuel pin, which had been part of a test fuel assembly in a boiling water reactor, several regions of severe internal hydriding were noticed in the upper-plenum end of the pin. Examination of similar fuel pins has shown that hydride of this type is caused by...... to irradiation-induced swelling....

MELCOR 1.8.2 assessment: The DF-4 BWR Damaged Fuel experiment

International Nuclear Information System (INIS)

Tautges, T.J.

1993-10-01

MELCOR is a fully integrated, engineering-level computer code being developed at Sandia National Laboratories for the USNRC, that models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRs. As a part of an ongoing assessment, program, MELCOR has been used to model the ACRR in-pile DF-4 Damaged Fuel experiment. DF-4 provided data for early phase melt progression in BWR fuel assemblies, particularly for phenomena associated with eutectic interactions in the BWR control blade and zircaloy oxidation in the canister and cladding. MELCOR provided good agreement with experimental data in the key areas of eutectic material behavior and canister and cladding oxidation. Several shortcomings associated with the MELCOR modeling of BWR geometries were found and corrected. Twenty-five sensitivity studies were performed on COR, HS and CVH parameters. These studies showed that the new MELCOR eutectics model played an important role in predicting control blade behavior. These studies revealed slight time step dependence and no machine dependencies. Comparisons made with the results from four best-estimate codes showed that MELCOR did as well as these codes in matching DF-4 experimental data

PWR and BWR spent fuel assembly gamma spectra measurements

Energy Technology Data Exchange (ETDEWEB)

Vaccaro, S. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Tobin, S.J.; Favalli, A. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Hu, J. [Oak Ridge National Laboratory, Oak Ridge (United States); Schwalbach, P. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company (SKB) (Sweden); Trellue, H.; Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States)

2016-10-11

A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of {sup 137}Cs, {sup 154}Eu, and {sup 134}Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

Infinite fuel element simulation of pin power distributions and control blade history in a BWR fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Li, J.; Nuenighoff, K.; Allelein, H.J. [Forschungszentrum Juelich GmbH (DE). Inst. fuer Energie- und Klimaforschung (IEK), Sicherheitsforschung und Reaktortechnik (IEK-6)

2011-07-01

Pellet-Cladding Interaction (PCI) is a well known effect in fuel pins. One possible reason for PCI-effects could be local power excursions in the fuel pins, which can led to a rupture of the fuel cladding tube. From a reactor safety point of view this has to be considered as a violence of the barrier principal in order to retain fission products in the fuel pins. This paper focuses on the pin power distributions in a 2D infinite lattice of a BWR fuel element. Lots of studies related PCI effect can be found in the literature. In this compact, coupled neutronic depletion calculations taking the control history effect into account are described. Depletion calculations of an infinite fuel element of a BWR were carried out with controlled, uncontrolled and temporarily controlled scenarios. Later ones are needed to describe the control blade history (CBH) effect. A Monte-Carlo approach is mandatory to simulate the neutron physics. The VESTA code was applied to couple the Monte-Carlo-Code MCNP(X) with the burnup code ORIGEN. Additionally, CASMO-4 is also employed to verify the method of simulation results from VESTA. The cross sections for Monte Carlo and burn-up calculations are derived from ENDF/B-VII.0. (orig.)

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

International Nuclear Information System (INIS)

Morimoto, Yuichi; Okumura, Keisuke

1992-05-01

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

Siemens Nuclear Power Corporation experience with BWR and PWR fuels

International Nuclear Information System (INIS)

Reparaz, A.; Smith, M.H.; Stephens, L.G.

1992-01-01

The large data base of fuel performance parameters available to Siemens Nuclear Power Corporation (SNP), and the excellent track record of innovation and fuel reliability accumulated over the last twenty-three years, allows SNP to have a clear insight on the characteristics of future developments in the area of fuel design. Following is a description of some of SNP's recent design innovations to prevent failures and to extend burnup capabilities. A goal paramount to the design and manufacture of BWR and PWR fuel is that of zero defects from any case during its operation in the reactor. Progress has already been made in achieving this goal. This paper summarized the cumulative failure rate of SNP fuel rod through January 1992

Thermal coupling of a high temperature PEM fuel cell with a complex hydride tank

DEFF Research Database (Denmark)

Pfeifer, P.; Wall, C.; Jensen, Jens Oluf

2009-01-01

the possibilities of a thermal coupling of a high temperature PEM fuel cell operating at 160-200 degrees C. The starting temperatures and temperature hold-times before starting fuel cell operation, the heat transfer characteristics of the hydride storage tanks, system temperature, fuel cell electrical power......Sodium alanate doped with cerium catalyst has been proven to have fast kinetics for hydrogen ab- and de-sorption as well as a high gravimetric storage density around 5 wt%. The kinetics of hydrogen sorption can be improved by preparing the alanate as nanocrystalline material. However, the second...... decomposition step, i.e. the decomposition of the hexahydride to sodium hydride and aluminium which refers to 1.8 wt% hydrogen is supposed to happen above 110 degrees C. The discharge of the material is thus limited by the level of heat supplied to the hydride storage tank. Therefore, we evaluated...

Basic evaluation on nuclear characteristics of BWR high burnup MOX fuel and core

International Nuclear Information System (INIS)

Nagano, M.; Sakurai, S.; Yamaguchi, H.

1997-01-01

MOX fuel will be used in existing commercial BWR cores as a part of reload fuels with equivalent operability, safety and economy to UO 2 fuel in Japan. The design concept should be compatible with UO 2 fuel design. High burnup UO 2 fuels are being developed and commercialized step by step. The MOX fuel planned to be introduced in around year 2000 will use the same hardware as UO 2 8 x 8 array fuel developed for a second step of UO 2 high burnup fuel. The target discharge exposure of this MOX fuel is about 33 GWd/t. And the loading fraction of MOX fuel is approximately one-third in an equilibrium core. On the other hand, it becomes necessary to minimize a number of MOX fuels and plants utilizing MOX fuel, mainly due to the fuel economy, handling cost and inspection cost in site. For the above reasons, it needed to developed a high burnup MOX fuel containing much Pu and a core with a large amount of MOX fuels. The purpose of this study is to evaluate basic nuclear fuel and core characteristics of BWR high burnup MOX fuel with batch average exposure of about 39.5 GWd/t using 9 x 9 array fuel. The loading fraction of MOX fuel in the core is within a range of about 50% to 100%. Also the influence of Pu isotopic composition fluctuations and Pu-241 decay upon nuclear characteristics are studied. (author). 3 refs, 5 figs, 3 tabs

Fuel assemblies for use in BWR type reactors

International Nuclear Information System (INIS)

Hirukawa, Koji.

1987-01-01

Purpose: To moderate the peak configuration of the burnup degree change curve for the infinite multiplication factor by applying an improvement to the arrangement of fuel rods. Constitution: In a fuel assembly for a BWR type reactor comprising a plurality of fuel rods and water rods arranged in a square lattice, fuel rods containing burnable poisons are arranged at four corners at the second and the third layers from the outside of the square lattice arrangement. Among them, the Cd poison effect in the burnable poison incorporated fuel rods disposed at the second layer is somewhat greater at the initial burning stage and then rapidly decreased along with burning. While on the other hand, the poison effect of the burnable poison-incorporated fuel rods at the third layer is smaller than that at the second layer at the initial burning stage and the reduction in the poison effect due to burning is somewhat more moderate. Since these fuel rods are in adjacent with each other, they interfere to each other and also provide an effect of moderating the burning of the burnable poisons. (Takahashi, M.)

Experimental data report for Test TS-2 reactivity initiated accident test in NSRR with pre-irradiated BWR fuel rod

International Nuclear Information System (INIS)

Nakamura, Takehiko; Yoshinaga, Makio; Sobajima, Makoto; Fujishiro, Toshio; Kobayashi, Shinsho; Yamahara, Takeshi; Sukegawa, Tomohide; Kikuchi, Teruo

1993-02-01

This report presents experimental data for Test TS-2 which was the second test in a series of Reactivity Initiated Accident (RIA) condition test using pre-irradiated BWR fuel rods, performed at the Nuclear Safety Research Reactor (NSRR) in February, 1990. Test fuel rod used in the Test TS-2 was a short sized BWR (7x7) type rod which was fabricated from a commercial rod irradiated at Tsuruga Unit 1 power reactor. The fuel had an initial enrichment of 2.79% and a burnup of 21.3Gwd/tU (bundle average). A pulse irradiation of the test fuel rod was performed under a cooling condition of stagnant water at atmospheric pressure and at ambient temperature which simulated a BWR's cold start-up RIA event. The energy deposition of the fuel rod in this test was evaluated to be 72±5cal/g·fuel (66±5cal/g·fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, transient behavior of the test rod during the pulse irradiation, and, results of pre and post pulse irradiation examinations are described in this report. (author)

Optimization of BWR fuel lattice enrichment and gadolinia distribution using genetic algorithms and knowledge

International Nuclear Information System (INIS)

Martin-del-Campo, Cecilia; Francois, Juan Luis; Carmona, Roberto; Oropeza, Ivonne P.

2007-01-01

An optimization methodology based on the Genetic Algorithms (GA) method was developed for the design of radial enrichment and gadolinia distributions for boiling water reactor (BWR) fuel lattices. The optimization algorithm was linked to the HELIOS code to evaluate the neutronic parameters included in the objective function. The goal is to search for a fuel lattice with the lowest average enrichment, which satisfy a reactivity target, a local power peaking factor (PPF), lower than a limit value, and an average gadolinia concentration target. The methodology was applied to the design of a 10 x 10 fuel lattice, which can be used in fuel assemblies currently used in the two BWRs operating at Mexico. The optimization process showed an excellent performance because it found forty lattice designs in which the worst one has a better neutronic performance than the reference lattice design. The main contribution of this study is the development of an efficient procedure for BWR fuel lattice design, using GA with an objective function (OF) which saves computing time because it does not require lattice burnup calculations

Analysis of Core Physics Experiments on Irradiated BWR MOX Fuel in REBUS Program

International Nuclear Information System (INIS)

Yamamoto, Toru; Ando, Yoshihira; Hayashi, Yamato

2008-01-01

As part of analyses of experimental data of a critical core containing a irradiated BWR MOX test bundle in the REBUS program, depletion calculations was performed for the BWR MOX fuel assemblies from that the MOX test rods were selected by using a general purpose neutronics code system SRAC. The core analyses were carried out using SRAC and a continuous energy Monte Carlo code MVP. The calculated k eff s were compared with those of the core containing a fresh MOX fuel bundle in the program. The SRAC-diffusion calculation underestimates k eff s of the both cores by 1.0 to 1.3 %dk and the k eff s of MVP are 1.001. The difference in k eff between the irradiated BWR MOX test bundle core and the fresh MOX one is 0.4 %dk in the SRAC-diffusion calculation and 0.0 %dk in the MVP calculation. The calculated fission rate distributions are in good agreement with the measurement in the SRAC-diffusion and MVP calculations. The calculated neutron flux distributions are also in good agreement with the measurement. The calculated burnup reactivity in the both calculations well reproduce the measurements. (authors)

BWR 90 and BWR 90+: Two advanced BWR design generations from ABB

International Nuclear Information System (INIS)

Haukeland, S.; Ivung, B.; Pedersen, T.

1999-01-01

ABB has two evolutionary advanced light water reactors available today - the BWR 90 boiling water reactor and the System 80+ pressurised water reactor. The BWR 90 is based on the design, construction, commissioning and operation of the BWR 75 plants. The operation experience of the six plants of this advanced design has been very good. The average annual energy availability is above 90%, and total power generation costs have been low. When developing the BWR 90 specific changes were introduced to a reference design, to adapt to technological progress, new safety requirements and to achieve cost savings. The thermal power rating of BWR 90 is 3800 MWth (providing a nominal 1374 MWe net), slightly higher than that of the reference plant ABB Atom has taken advantage of margins gained using a new generation of its SVEA fuel to attain this power rating without major design modifications. The BWR 90 design was completed and offered to the TVO utility in Finland in 1991, as one of the contenders for the fifth Finnish nuclear power plant project. Hence, the design is available today for deployment in new plant projects. Utility views were incorporated through co-operation with the Finnish utility TVO, owner and operator of the two Olkiluoto plants of BWR 75 design. A review against the European Utility Requirement (EUR) set of requirements has been performed, since the design, in 1997, was selected by the EUR Steering Committee to be the first BWR to be evaluated against the EUR documents. The review work was completed in 1998. It will be the subject of an 'EUR Volume 3 Subset for BWR 90' document. ABB is continuing its BWR development work with an 'evolutionary' design called BWR 90+, which aims at developing the BWR as a competitive option for the anticipated revival of the market for new nuclear plants beyond the turn of the century, as well as feeding ideas and inputs to the continuous modernisation efforts at operating plants. The development is performed by ABB Atom

Connection between end plates and rods in a BWR fuel element

International Nuclear Information System (INIS)

Cali', G.P.

1975-01-01

The problem of the connection between the end plates and the rods of a BWR fuel element is analytically formulated. The behaviour of the springs coupling the rods with the upper plate is analyzed with particular detail since the deformation of these springs affects the forces at the interface of the fuel element structure components. A tool is given to design the springs according to some considerations regarding the mechanical strength of the interacting components as well as the influence of the possible geometrical unevennes of the system that can arise during the fuel element lifetime. (Cali', G.P.)

Solution of a benchmark set problems for BWR and PWR reactors with UO2 and MOX fuels using CASMO-4

International Nuclear Information System (INIS)

Martinez F, M.A.; Valle G, E. del; Alonso V, G.

2007-01-01

In this work some of the results for a group of benchmark problems of light water reactors that allow to study the physics of the fuels of these reactors are presented. These benchmark problems were proposed by Akio Yamamoto and collaborators in 2002 and they include two fuel types; uranium dioxide (UO 2 ) and mixed oxides (MOX). The range of problems that its cover embraces three different configurations: unitary cell for a fuel bar, fuel assemble of PWR and fuel assemble of BWR what allows to carry out an understanding analysis of the problems related with the fuel performance of new generation in light water reactors with high burnt. Also these benchmark problems help to understand the fuel administration in core of a BWR like of a PWR. The calculations were carried out with CMS (of their initials in English Core Management Software), particularly with CASMO-4 that is a code designed to carry out analysis of fuels burnt of fuel bars cells as well as fuel assemblies as much for PWR as for BWR and that it is part in turn of the CMS code. (Author)

BWR type nuclear reactors

International Nuclear Information System (INIS)

Yamamoto, Toru.

1987-01-01

Purpose: To obtain reactor core characteristics with less changes in the excess reactivity due to fuel burnup even when the operation period varies. Constitution: In a BWR type reactor where fuel assemblies containing fuel rods incorporated with burnable poisons are arranged, the fuel assemblies are grouped into first fuel assemblies and second fuel assemblies. Then, the number of fuel rods incorporated with burnable poisons within the first fuel assemblies is made greater than that of the second fuel rods, while the concentration of the burnable poisons in the fuel rods incorporated with the burnable poisons in the first fuel assemblies is made lower than that of the fuel rods incorporated with the burnable poisons in the second fuel assemblies. In the BWR type reactor constituted in this way, the reactor core characteristics can be improved by changing the ratio between the first fuel assemblies and the second fuel assemblies charged to the reactor core, thereby decreasing the changes in the burnup of the excess reactivity. (Kamimura, M.)

A Study on the Radial Hydride Assisted Delayed Hydride Cracking of Zircaloy

Energy Technology Data Exchange (ETDEWEB)

Jeong, Jin-Ho; Lee, Ji-Min; Kim, Yong-Soo [Hanyang University, Seoul (Korea, Republic of)

2015-05-15

Extensive studies have been done on understanding of DHC(Delayed hydride cracking) phenomenon since several zirconium alloy pressure tubes failed in nuclear reactor in the 1970s. Recently, long-term dry storage strategy has been considered seriously in order to manage spent nuclear fuel in Korea and other countries around the world. Consequentially, many researches have been investigated the degradation mechanisms which will threaten the spent fuel integrity during dry storage and showed that hydrogen related phenomenon such as hydride reorientation and DHC are the critical factors. Especially, DHC is the direct cracking mechanism which can cause not only a through-wall defect but also a radiation leak to the environment. In addition, DHC can be enhanced by radial hydride as reported by Kim who demonstrate that radial hydrides clearly act as crack linkage path. This phenomenon is known as the radial hydride assisted DHC (RHA-DHC). Therefore, study on DHC is essential to ensure the safety of spent fuel. Finite element analysis will be carried out for the stress gradient evaluation around notch tip. A variation in thermal cycle which leads to change in hydrogen solid solution trajectory may be required. If the radial hydride precipitates at notch tip, we will investigate what conditions should be met. Ultimately, we will suggest the regulation criteria for long-term dry storage of spent nuclear fuel.

Hydrogen storage systems based on magnesium hydride: from laboratory tests to fuel cell integration

Science.gov (United States)

de Rango, P.; Marty, P.; Fruchart, D.

2016-02-01

The paper reviews the state of the art of hydrogen storage systems based on magnesium hydride, emphasizing the role of thermal management, whose effectiveness depends on the effective thermal conductivity of the hydride, but also depends of other limiting factors such as wall contact resistance and convective exchanges with the heat transfer fluid. For daily cycles, the use of phase change material to store the heat of reaction appears to be the most effective solution. The integration with fuel cells (1 kWe proton exchange membrane fuel cell and solid oxide fuel cell) highlights the dynamic behaviour of these systems, which is related to the thermodynamic properties of MgH2. This allows for "self-adaptive" systems that do not require control of the hydrogen flow rate at the inlet of the fuel cell.

BWR 90: The ABB advanced BWR design

International Nuclear Information System (INIS)

Haukeland, S.; Ivung, B.; Pedersen, T.

1999-01-01

ABB has two evolutionary advanced fight water reactors available today - the BWR 90 boiling water reactor and the System 80+ pressurised water reactor. The BWR 90 is based on the design, construction, commissioning and operation of the BWR 75 plants. The operation experience of the six plants of this advanced design has been very good. The average annual energy availability is above 90%, and the total power generation costs have been low. In the development of BWR 90 specific changes were introduced to the reference design, to adapt to technological progress, new safety requirements and to achieve cost savings. The thermal power rating of BWR 90 is 3800 MWth (providing a nominal 1374 MWe net), slightly higher dim that of the reference plant ABB Atom has taken advantage of margins gained using a new generation of its SVEA fuel to attain this power rating without major design modifications. The BWR 90 design was completed and offered to the TVO utility in Finland in 1991, as one of the contenders for the fifth Finnish nuclear power plant project. Thus, the design is available today for deployment in new plant projects. Utility views were incorporated through co-operation with the Finnish utility TVO, owner and operator of the two Olkiluoto plants of BWR 75 design. A review against the European Utility Requirement (EUR) set of requirements has been performed, since the design, in 1997, was selected by the EUR Steering Committee to be the first BWR to be evaluated against the EUR documents. The work is scheduled for completion in 1998. It will be the subject of an 'EUR Volume 3 Subset for BWR 90' document. ABB is continuing its BWR development work with the 'evolutionary' design BWR 90+. The primary design goal is to develop the BWR as a competitive option for the anticipated revival of the market for new nuclear plants beyond the turn of the century, as well as feeding ideas and inputs to the continuous modernisation efforts at operating plants. The development is

Design study of Thorium-232 and Protactinium-231 based fuel for long life BWR

Energy Technology Data Exchange (ETDEWEB)

Trianti, N.; Su' ud, Z.; Riyana, E. S. [Nuclear Physics and Biophysics Research Division Department of Physics - Institut Teknologi Bandung (ITB) Jalan Ganeca 10 Bandung 40132 (Indonesia)

2012-06-06

A preliminary design study for the utilization of thorium added with {sup 231}Pa based fuel on BWR type reactor has been performed. In the previous research utilization of fuel based Thorium-232 and Uranium-233 show 10 years operation time with maximum excess-reactivity about 4.075% dk/k. To increase reactor operation time and reduce excess-reactivity below 1% dk/k, Protactinium (Pa-231) is used as Burnable Poison. Protactinium-231 has very interesting neutronic properties, which enable the core to reduce initial excess-reactivity and simultaneously increase production of {sup 233}U to {sup 231}Pa in burn-up process. Optimizations of the content of {sup 231}Pa in the core enables the BWR core to sustain long period of operation time with reasonable burn-up reactivity swing. Based on the optimization of fuel element composition (Th and Pa) in various moderation ratio we can get reactor core with longer operation time, 20 {approx} 30 years operation without fuel shuffling or refuelling, with average power densities maximum of about 35 watt/cc, and maximum excess-reactivity 0.56% dk/k.

A neutronic assessment of the new Spherical Cermets Fuel concept for the BWR-PB reactor

International Nuclear Information System (INIS)

Benchrif, A.; Chetaine, A.; Amsil, H.; Bounakhla, M.

2010-01-01

The tri-structural-isotopic (TRISO) fuel directly cooled by boiling light water is used in the boiling water reactor with pebble-bed coated particles (BWR-PB). At the lower coolant temperature, the TRISO fuel particles demonstrate an unacceptable irradiation swelling in the silicon carbide coating layer during a fuel cycle. So, the objectives of this paper, on the one hand is to evaluate some neutronic parameters of a new fuel concept, Spherical Cermets Fuel (SCF), for a BWR-PB reactor. On the other hand, to assess the fact of SCF fuel concept on the fuel assembly lifetime and the burn-up characteristic. All the parameters as well as Infinite Multiplication Factor, Spectrum Index, Instantaneous Conversion Ratio and Neutron Energy Spectrum was calculated then compared for the TRISO and the SCF fuel concept. It can be seen from the assessment of fuel assembly burn-up characteristics that the normalised neutron spectra of all the assembly's parts pointed out a thermal spectrum for the SCF fuel assembly's parts than the TRISO one. The SCF fuel element increase the assembly life time about 6.1 EFPY corresponding 8000 MWd/t. So, the fuel assembly can be operated for a reasonably long period without outside refuelling. The difference in the assembly lifetime might leads to SCF fuel concept adopted, because the geometry and concept of TRISO fuel particles are wholly different to SCF ones. (author)

Investigation of 3H and 14C inventory and distribution in spent BWR fuel rods

International Nuclear Information System (INIS)

Bleier, A.; Beuerle, M.; Neeb, K.H.

1984-10-01

In order to obtain reliable data for fuel reprocessing and waste disposal, the T and C-14 inventory, distribution and behaviour was investigated on a typical LWR fuel rod discharged from a BWR plant. The results showed that 50 ± 5% of the T generated in the fuel is present in the cladding after reactor operation. The remainder of the T stays with the fuel. Related to the reactor power the total T inventory corresponds to a T production rate of 19 000 Ci/GW e . a. The C-14 built up in the fuel represents approximately 60% of the C-14 inventory of the BWR fuel rod. The remaining part of C-14 (about 40%) experimentally determined by this analysis for the first time is generated in the cladding. From the total C-14 inventory a C-14 production rate of 17,5 Ci/GW e . a can be calculated. The fill gas contains only negligible fractions of both nuclides. The results obtained in this program are generally in good agreement with the data of theoretical estimates and with results of earlier investigations on PWR fuel rods. (orig.) [de

Effect of zinc injection on BWR fuel cladding corrosion. Pt. 1. Study on an accelerated corrosion condition to evaluate corrosion resistance of zircaloy-2 fuel cladding

International Nuclear Information System (INIS)

Kawamura, Hirotaka; Kanbe, Hiromu; Furuya, Masahiro

2002-01-01

Japanese BWR utilities have a plan to apply zinc injection to the primary coolant in order to reduce radioactivity accumulation on the structure. Prior to applying the zinc injection to BWR plants, it is necessary to evaluate the effect of zinc injection on corrosion resistance of fuel cladding. The objective of this report was to examine the accelerated corrosion condition for evaluation of BWR fuel cladding corrosion resistance under non-irradiated conditions, as the first step of a zinc injection evaluation study. A heat transfer corrosion test facility, in which a two phase flow condition could be achieved, was designed and constructed. The effects of heat flux, void fraction and solution temperature on BWR fuel cladding corrosion resistance were quantitatively investigated. The main findings were as follows. (1) In situ measurements using high speed camera and a void sensor together with one dimensional two phase flow analysis results showed that a two phase flow simulated BWR core condition can be obtained in the corrosion test facility. (2) The heat transfer corrosion test results showed that the thickness of the zirconium oxide layer increased with increasing solution temperature and was independent of heat flux and void fraction. The corrosion accelerating factor was about 2.5 times in the case of a temperature increase from 288degC to 350degC. (author)

Subchannel analysis of a critical power test, using simulated BWR 8x8 fuel assembly

International Nuclear Information System (INIS)

Mitsutake, T.; Terasaka, H.; Yoshimura, K.; Oishi, M.; Inoue, A.; Akiyama, M.

1990-01-01

Critical power predictions have been compared with the critical power test data obtained in simulated BWR 8x8 fuel rod assemblies. Two analytical methods for the critical power prediction in rod assemblies are used in the prediction, which are the subchannel analysis using the COBRA/BWR subchannel computer code with empirical critical heat flux (CHF) correlations and the liquid film dryout estimation using the CRIPP-3F 'multi-fluid' computer code. Improvements in both the analytical methods were made for spacer effect modeling, though they were specific for application to the current BWR rod assembly type. In general a reasonable agreement was obtained, though comparisons, between the prediction and the obtained test data. (orig.)

Experimental data report for test TS-3 Reactivity Initiated Accident test in the NSRR with pre-irradiated BWR fuel rod

International Nuclear Information System (INIS)

Nakamura, Takehiko; Yoshinaga, Makio; Fujishiro, Toshio; Kobayashi, Shinsho; Yamahara, Takeshi; Sukegawa, Tomohide; Kikuchi, Teruo; Sobajima, Makoto.

1993-09-01

This report presents experimental data for Test TS-3 which was the third test in a series of Reactivity Initiated Accident (RIA) tests using pre-irradiated BWR fuel rods, performed in the Nuclear Safety Research Reactor (NSRR) in September, 1990. Test fuel rod used in the Test TS-3 was a short-sized BWR (7 x 7) type rod which was re-fabricated from a commercial rod irradiated in the Tsuruga Unit 1 power reactor of Japan Atomic Power Co. The fuel had an initial enrichment of 2.79 % and a burnup of 26 Gwd/tU. A pulse irradiation of the test fuel rod was performed under a cooling condition of stagnant water at atmospheric pressure and at ambient temperature which simulated a BWR's cold start-up RIA event. The energy deposition of the fuel rod in this test was evaluated to be 94 ± 4 cal/g · fuel (88 ± 4 cal/g · fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, transient behavior of the test rod during the pulse irradiation, and results of pre-pulse and post-pulse irradiation examinations are described in this report. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-01-01

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

3D modeling of missing pellet surface defects in BWR fuel

Energy Technology Data Exchange (ETDEWEB)

Spencer, B.W., E-mail: Benjamin.Spencer@inl.gov; Williamson, R.L.; Stafford, D.S.; Novascone, S.R.; Hales, J.D.; Pastore, G.

2016-10-15

Highlights: • A global/local analysis procedure for missing pellet surface defects is proposed. • This is applied to defective BWR fuel under blade withdrawal and high power ramp conditions. • Sensitivity of the cladding response to key model parameters is studied. - Abstract: One of the important roles of cladding in light water reactor fuel rods is to prevent the release of fission products. To that end, it is essential that the cladding maintain its integrity under a variety of thermal and mechanical loading conditions. Local geometric irregularities in fuel pellets caused by manufacturing defects known as missing pellet surfaces (MPS) can in some circumstances lead to elevated cladding stresses that are sufficiently high to cause cladding failure. Accurate modeling of these defects can help prevent these types of failures. The BISON nuclear fuel performance code developed at Idaho National Laboratory can be used to simulate the global thermo-mechanical fuel rod behavior, as well as the local response of regions of interest, in either 2D or 3D. In either case, a full set of models to represent the thermal and mechanical properties of the fuel, cladding and plenum gas is employed. A procedure for coupling 2D full-length fuel rod models to detailed 3D models of the region of the rod containing a MPS defect is detailed here. The global and local model each contain appropriate physics and behavior models for nuclear fuel. This procedure is demonstrated on a simulation of a boiling water reactor (BWR) fuel rod containing a pellet with an MPS defect, subjected to a variety of transient events, including a control blade withdrawal and a ramp to high power. The importance of modeling the local defect using a 3D model is highlighted by comparing 3D and 2D representations of the defective pellet region. Parametric studies demonstrate the effects of the choice of gaseous swelling model and of the depth and geometry of the MPS defect on the response of the cladding

Impact of modeling Choices on Inventory and In-Cask Criticality Calculations for Forsmark 3 BWR Spent Fuel

International Nuclear Information System (INIS)

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

2015-01-01

Simulation of boiling water reactor (BWR) fuel depletion poses a challenge for nuclide inventory validation and nuclear criticality safety analyses. This challenge is due to the complex operating conditions and assembly design heterogeneities that characterize these nuclear systems. Fuel depletion simulations and in-cask criticality calculations are affected by (1) completeness of design information, (2) variability of operating conditions needed for modeling purposes, and (3) possible modeling choices. These effects must be identified, quantified, and ranked according to their significance. This paper presents an investigation of BWR fuel depletion using a complete set of actual design specifications and detailed operational data available for five operating cycles of the Swedish BWR Forsmark 3 reactor. The data includes detailed axial profiles of power, burnup, and void fraction in a very fine temporal mesh for a GE14 (10x10) fuel assembly. The specifications of this case can be used to assess the impacts of different modeling choices on inventory prediction and in-cask criticality, specifically regarding the key parameters that drive inventory and reactivity throughout fuel burnup. This study focused on the effects of the fidelity with which power history and void fraction distributions are modeled. The corresponding sensitivity of the reactivity in storage configurations is assessed, and the impacts of modeling choices on decay heat and inventory are addressed.

Calculation device for fuel power history in BWR type reactors

International Nuclear Information System (INIS)

Sakagami, Masaharu.

1980-01-01

Purpose: To enable calculations for power history and various variants of power change in the power history of fuels in a BWR type reactor or the like. Constitution: The outputs of the process computation for the nuclear reactor by a process computer are stored and the reactor core power distribution is judged from the calculated values for the reactor core power distribution based on the stored data. Data such as for thermal power, core flow rate, control rod position and power distribution are recorded where the changes in the power distribution exceed a predetermined amount, and data such as for thermal power and core flow rate are recorded where the changes are within the level of the predetermined amount, as effective data excluding unnecessary data. Accordingly, the recorded data are taken out as required and the fuel power history and the various variants in the fuel power are calculated and determined in a calculation device for fuel power history and variants for fuel power fluctuation. (Furukawa, Y.)

BWR AXIAL PROFILE

International Nuclear Information System (INIS)

Huffer, J.

2004-01-01

The purpose of this calculation is to develop axial profiles for estimating the axial variation in burnup of a boiling water reactor (BWR) assembly spent nuclear fuel (SNF) given the average burnup of an assembly. A discharged fuel assembly typically exhibits higher burnup in the center and lower burnup at the ends of the assembly. Criticality safety analyses taking credit for SNF burnup must account for axially varying burnup relative to calculations based on uniformly distributed assembly average burnup due to the under-burned tips. Thus, accounting for axially varying burnup in criticality analyses is also referred to as accounting for the ''end effect'' reactivity. The magnitude of the reactivity change due to ''end effect'' is dependent on the initial assembly enrichment, the assembly average burnup, and the particular axial profile characterizing the burnup distribution. The set of bounding axial profiles should incorporate multiple BWR core designs and provide statistical confidence (95 percent confidence that 95 percent of the population is bound by the profile) that end nodes are conservatively represented. The profiles should also conserve the overall burnup of the fuel assembly. More background on BWR axial profiles is provided in Attachment I

Sphere-pac versus pellet UO2 fuel in de Dodewaard BWR

International Nuclear Information System (INIS)

Linde, A. van der.

1989-04-01

Comparative testing of UO 2 sphere-pac and pellet fuel rods under LWR conditions has been jointly performed by the Netherlands Utilities Research Centre (KEMA) in Arnhem, the Netherlands Energy Research Foundation (ECN) at Petten and the Netherlands Joint Nuclear Power Utility (GKN) at Dodewaard. This final report summarizes the highlights of this 1968-1988 program with strong emphasis on the fuel rods irradiated in the Dodewaard BWR. The conclusion reached is that under normal LWR conditions sphere-pac UO 2 in LWR fuel rods offers better resistance against stress corrosion cracking of the cladding, but that under fast, single step, power ramping conditions pellet UO 2 in LWR fuel rods has a better resistance against hoop stress failure of the cladding. 128 figs., 36 refs., 19 tabs

EPRI BWR Water Chemistry Guidelines Revision

International Nuclear Information System (INIS)

Garcia, Susan E.; Giannelli, Joseph F.

2014-01-01

BWRVIP-190: BWR Water Chemistry Guidelines – 2008 Revision has been revised. The revision committee consisted of U.S. and non-U.S. utilities (members of the BWR Vessel and Internals Protection (BWRVIP) Mitigation Committee), reactor system manufacturers, fuel suppliers, and EPRI and industry experts. The revised document, BWRVIP-190 Revision 1, was completely reformatted into two volumes, with a simplified presentation of water chemistry control, diagnostic and good practice parameters in Volume 1 and the technical bases in Volume 2, to facilitate use. The revision was developed in parallel and in coordination with preparation of the Fuel Reliability Guidelines Revision 1: BWR Fuel Cladding Crud and Corrosion. Guidance is included for plants operating under normal water chemistry (NWC), moderate hydrogen water chemistry (HWC-M), and noble metal application (GE-Hitachi NobleChem™) plus hydrogen injection. Volume 1 includes significant changes to BWR feedwater and reactor water chemistry control parameters to provide increased assurance of intergranular stress corrosion cracking (IGSCC) mitigation of reactor materials and fuel reliability during all plant conditions, including cold shutdown (≤200°F (93°C)), startup/hot standby (>200°F (93°C) and ≤ 10%) and power operation (>10% power). Action Level values for chloride and sulfate have been tightened to minimize environmentally assisted cracking (EAC) of all wetted surfaces, including those not protected by hydrogen injection, with or without noble metals. Chemistry control guidance has been enhanced to minimize shutdown radiation fields by clarifying targets for depleted zinc oxide (DZO) injection while meeting requirements for fuel reliability. Improved tabular presentations of parameter values explicitly indicate levels at which actions are to be taken and required sampling frequencies. Volume 2 provides the technical bases for BWR water chemistry control for control of EAC, flow accelerated corrosion

Neutron physical aspects of the storage of BWR fuel elements

International Nuclear Information System (INIS)

Woloch, F.; Sdouz, G.; Suda, M.

1980-01-01

For the storage of BWR fuel elements in a high density fuel rack using boronated steel absorbers and in a fuel rack with a larger pitch without absorber, criticality calculations are performed. The cooling water density is varied for the storage without absorbers. For the selected pitches of 16.5 cm for the high density fuel rack and 25 cm for the fuel rack without absorber respectively the ksub(infinitely) values of 0.933 and 0.748 are obtained. The dependence of the results on different calculational methods and on the influence of the variation of three important design parameters, i.e. of the concentration of boron, of the thickness of the boronated steel and of the watergap is investigated for the high density fuel rack. The average isothermal temperature coefficient is obtained for the high density fuel rack as -4.5 x 10 -40 sup(0)C -1 and as approx. 2.0 x 10 -40 sup(0)C -1 for the fuel rack without absorbers. For both ways of storage the aspects of safety of the results are discussed thoroughly. (orig.) 891 RW/orig. 892 CKA [de

K-infinite trends with burnup, enrichment, and cooling time for BWR fuel assemblies

International Nuclear Information System (INIS)

Broadhead, B.L.

1998-08-01

This report documents the work performed by ORNL for the Yucca Mountain project (YMP) M and O contractor, Framatome Cogema Fuels. The goal of this work was to obtain k inf values for infinite arrays of flooded boiling-water-reactor (BWR) fuel assemblies as a function of various burnup/enrichment and cooling-time combinations. These scenarios simulate expected limiting criticality loading conditions (for a given assembly type) for drift emplacements in a repository. Upon consultation with the YMP staff, a Quad Cities BWR fuel assembly was selected as a baseline assembly. This design consists of seven axial enrichment zones, three of which contain natural uranium oxide. No attempt was made to find a bounding or even typical assembly design due to the wide variety in fuel assembly designs necessary for consideration. The current work concentrates on establishing a baseline analysis, along with a small number of sensitivity studies which can be expected later if desired. As a result of similar studies of this nature, several effects are known to be important in the determination of the final k inf for spent fuel in a cask-like geometry. For a given enrichment there is an optimal burnup: for lower burnups, excess energy (and corresponding excess reactivity) is present in the fuel assembly; for larger burnups, the assembly is overburned and essentially driven by neighboring fuel assemblies. The majority of the burnup/enrichment scenarios included in this study were for some near-optimum burnup/enrichment combinations as determined from Energy Information Administration (EIA) data. Several calculations were performed for under- and over-burned fuel to show these effects

Fundamentals of boiling water reactor (BWR)

International Nuclear Information System (INIS)

Bozzola, S.

1982-01-01

These lectures on fundamentals of BWR reactor physics are a synthesis of known and established concepts. These lectures are intended to be a comprehensive (even though descriptive in nature) presentation, which would give the basis for a fair understanding of power operation, fuel cycle and safety aspects of the boiling water reactor. The fundamentals of BWR reactor physics are oriented to design and operation. In the first lecture general description of BWR is presented, with emphasis on the reactor physics aspects. A survey of methods applied in fuel and core design and operation is presented in the second lecture in order to indicate the main features of the calculational tools. The third and fourth lectures are devoted to review of BWR design bases, reactivity requirements, reactivity and power control, fuel loading patterns. Moreover, operating limits are reviewed, as the actual limits during power operation and constraints for reactor physics analyses (design and operation). The basic elements of core management are also presented. The constraints on control rod movements during the achieving of criticality and low power operation are illustrated in the fifth lecture. Some considerations on plant transient analyses are also presented in the fifth lecture, in order to show the impact between core and fuel performance and plant/system performance. The last (sixth) lecture is devoted to the open vessel testing during the startup of a commercial BWR. A control rod calibration is also illustrated. (author)

PWR and BWR light water reactor systems in the USA and their fuel cycle

International Nuclear Information System (INIS)

Crawford, W.D.

1977-01-01

Light water reactor operating experience in the USA can be considered to date from the choice of the pressurized water reactor (PWR) for use in the naval reactor program and the subsequent construction and operation of the nuclear power plant at Shippingport, Pennsylvania in 1957. The development of the boiling water reactor (BWR) in 1954 and its selection for the plant at Dresden, Illinois in 1959 established this concept as the other major reactor type in the US nuclear power program. The subsequent growth profile is presented, leading to 31 PWR's and 23 BWR's currently in operation as well as to plants in the planning and construction phase. A significant operating record has been accumulated concerning the availability of each of these reactor types as determined by: (1) outage for refueling, (2) component reliability, (3) maintenance requirements, and (4) retrofitting required by government regulation. In addition, the use and performance of BWR's and PWR's in meeting system load requirements is discussed. The growing concern regarding possible terrorist activities and other potential threats has resulted in systems and procedures designed to assure effective safeguards at nuclear power installations. Safeguards measures currently in place are described. Environmental effects of operating plants are subject to both radiological and non-radiological monitoring to verify that results are within the limits established in the licensing process. The operating results achieved and the types of modifications that have been required of operating plants by the Nuclear Regulatory Commission are reviewed. The PWR and BWR Fuel Cycle is examined in terms of: (1) fuel burnup experience and prospects for improvement, (2) the status and outlook for natural uranium resources, (3) enrichment capacity, (4) reprocessing and recycle, and the interrelationships among the latter three factors. High level waste management currently involving on-site storage of spent fuel is discussed

Taking burnup credit for interim storage and transportation system for BWR fuels

International Nuclear Information System (INIS)

Yoshioka, Ken-ichi; Ando, Y.; Kumanomido, H.; Sasaki, T.; Mitsuhashi, I.; Ueda, M.

2001-01-01

In order to establish a realistic burnup credit design system, a calculation system has been developed for determining isotope compositions, burnup, and criticality. The calculation system consists of several modules such as TGBLA, ORIGEN, CITATION, MCNP, and KENO. The TGBLA code is a fuel design code for LWR fuels developed in TOSHIBA Corporation. A compact measurement system for a fuel assembly has been being developed to meet requirements for the burnup determination, the neutron emission-rate evaluation, and the nuclear materials management. For a spent MOX fuel, a neutron emission rate measurement method has been being developed. The system consists of Cd-Te detectors and / or fission chambers. Some model calculations were carried out for the latest design BWR fuels. The effect of taking burnup credit for a transportation cask is shown. (authors)

Fuel rod response to BWR power oscillations during anticipated transient without scram

International Nuclear Information System (INIS)

Cunningham, M.; Scott, H.

1998-01-01

The US NRC is examining fuel behaviour during a postulated BWR anticipated transient without scram (ATWS) with power oscillations to determine if current regulatory criteria are adequate. Currently, the 280 cal/g limit for RIAs is used to show that coolable geometry is maintained and pressure pulses are avoided during ATWSs. Two specific questions have now been raised about the continued use of the 280 cal/g value. First, this value was derived from energy deposition values whereas the regulatory requirements are written in terms of fuel enthalpy. The second is that fuel rod rupture with fuel dispersal has been observed in RIA tests with high bum-up fuel rods having energy deposition values well below the current limit. However, the BWR ATWS power oscillation transient is slower than a RIA power pulse, thus reducing the likelihood of failure. Therefore questions about the adequacy of the 280 cal/g limit do not necessarily imply unacceptable fuel damage occurring during such power oscillations and there is no immediate safety concern. The reported analysis, using the FRAPTRAN transient fuel rod analysis code, was thus undertaken to determine if further investigation might be appropriate and with the intention of starting some discussions about the issue. There was a comment that a limit of 100 cal/g fuel enthalpy had been mentioned following the scoping calculations but that perhaps enthalpy was not the main concern in an ATWS. It was also observed that cladding stresses are lower than in all RIA. The question was what really is the main concern. It was replied that the main concern was a question of maintaining a coolable geometry i.e. not loosing fuel particles out of the rod. And it was agreed that enthalpy may not be the important issue, rather that it previously had been used as the parameter and so had been considered. Confirmation of this presently being an evaluation and not a regulatory concern was sought and provided, it being pointed out that the NRC

Delayed hydride cracking of Zircaloy-4 fuel cladding

International Nuclear Information System (INIS)

Pizarro, Luis M.; Fernandez, Silvia; Lafont, Claudio; Mizrahi, Rafael; Haddad, Roberto

2007-01-01

Crack propagation rates, grown by the delayed hydride cracking mechanism, were measured in Zircaloy-4 fuel cladding, according to a Coordinated Research Project (CRP) sponsored by the International Atomic Energy Agency (IAEA). During the first stage of the program a Round Robin Testing was performed on fuel cladding samples provided by Studsvik (Sweden), of the type used in PWR reactors. Crack growth in the axial direction is obtained through the specially developed 'pin load testing' (PLT) device. In these tests, crack propagation rates were determined at 250 C degrees on several samples of the material described above, obtaining a mean value of about 2.5 x 10 -8 m s -1 . The results were analyzed and compared satisfactorily with those obtained by the other laboratories participating in the CRP. At the present moment, similar tests on CANDU and Atucha I type fuel cladding are being performed. It is thought that the obtained results will give valuable information concerning the analysis of possible failures affecting fuel cladding under reactor operation. (author) [es

Finite element analysis of BWR fuel channel buckling during a seismic event

International Nuclear Information System (INIS)

Kinoshita, Mika; Iwamoto, Yuji; Ledford, Kevin; Cantonwine, Paul

2014-01-01

This paper documents the predicted response of three BWR fuel channel designs in bending using a typical moment profile for GNF fuel designs. The bending performance of the fuel channel is predicted using ANSYS, a finite element modeling tool. Specifically, linear and non-linear buckling analyses were performed to determine the onset of elastic buckling, which causes a wavy structure on the compression face in bending that might also increase channel – control blade friction, and to determine to onset of channel collapse, which causes permanent deformation and would inhibit control rod insertion. The three channel designs considered in this paper are the 0.080 inch uniform channel, the 0.100 inch uniform channel and the 0.120 inch uniform channel at the beginning of fuel life (BOL) and at the end of fuel life (EOL). (author)

Evaluation of U-Zr hydride fuel for a thorium fuel cycle in an RTR concept

Energy Technology Data Exchange (ETDEWEB)

Lee, Kyung Taek; Cho, Nam Zin [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1999-12-31

In this paper, we performed a design study of a thorium fueled reactor according to the design concept of the Radkowsky Thorium Reactor (RTR) and evaluated its overall performance. To enhance its performance and alleviate its problems, we introduced a new metallic uranium fuel, uranium-zirconium hydride (U-ZrH{sub 1.6}), as a seed fuel. For comparison, typical ABB/CE-type PWR based on SYSTEM 80+and standard RTR-type thorium reactor were also studied. From the results of performance analysis, we could ascertain advantages of RTR-type thorium fueled reactor in proliferation resistance, fuel cycle economics, and back-end fuel cycle. Also, we found that enhancement of proliferation resistance and safer operating conditions may be achieved by using the U-ZrH{sub 1.6} fuel in the seed region without additional penalties in comparison with the standard RTR`s U-Zr fuel. 6 refs., 2 figs., 6 tabs. (Author)

Evaluation of U-Zr hydride fuel for a thorium fuel cycle in an RTR concept

Energy Technology Data Exchange (ETDEWEB)

Lee, Kyung Taek; Cho, Nam Zin [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1998-12-31

In this paper, we performed a design study of a thorium fueled reactor according to the design concept of the Radkowsky Thorium Reactor (RTR) and evaluated its overall performance. To enhance its performance and alleviate its problems, we introduced a new metallic uranium fuel, uranium-zirconium hydride (U-ZrH{sub 1.6}), as a seed fuel. For comparison, typical ABB/CE-type PWR based on SYSTEM 80+and standard RTR-type thorium reactor were also studied. From the results of performance analysis, we could ascertain advantages of RTR-type thorium fueled reactor in proliferation resistance, fuel cycle economics, and back-end fuel cycle. Also, we found that enhancement of proliferation resistance and safer operating conditions may be achieved by using the U-ZrH{sub 1.6} fuel in the seed region without additional penalties in comparison with the standard RTR`s U-Zr fuel. 6 refs., 2 figs., 6 tabs. (Author)

Studies on the fission products behavior during dissolution process of BWR spent fuel

International Nuclear Information System (INIS)

Sato, K.; Nakai, E.; Kobayashi, Y.

1987-01-01

In order to obtain basic data on fission products behavior in connection with the head end process of fuel reprocessing, especially to obtain better understanding on undissolved residues, small scale dissolution studies were performed by using BWR spent fuel rods which were irradiated as monitoring fuel rods under the monitoring program for LWR fuel assembly performance entitled PROVING TEST ON RELIABILITY OF FUEL ASSEMBLY . The Zircaloy-2 claddings and the fuel pellets were subjected individually to the following studies on 1) release of fission products during dissolution process, 2) characterization of undissolved residues, and 3) analysis of the claddings. This paper presents comprehensive descriptions of the fission products behavior during dissolution process, based on detailed and through PIE conducted by JNFS under the sponsorship of MITI (Ministry of International Trade and Industry)

Azimuthally anisotropic hydride lens structures in Zircaloy 4 nuclear fuel cladding: High-resolution neutron radiography imaging and BISON finite element analysis

Science.gov (United States)

Lin, Jun-Li; Zhong, Weicheng; Bilheux, Hassina Z.; Heuser, Brent J.

2017-12-01

High-resolution neutron radiography has been used to image bulk circumferential hydride lens particles in unirradiated Zircaloy 4 tubing cross section specimens. Zircaloy 4 is a common light water nuclear reactor (LWR) fuel cladding; hydrogen pickup, hydride formation, and the concomitant effect on the mechanical response are important for LWR applications. Ring cross section specimens with three hydrogen concentrations (460, 950, and 2830 parts per million by weight) and an as-received reference specimen were imaged. Azimuthally anisotropic hydride lens particles were observed at 950 and 2830 wppm. The BISON finite element analysis nuclear fuel performance code was used to model the system elastic response induced by hydride volumetric dilatation. The compressive hoop stress within the lens structure becomes azimuthally anisotropic at high hydrogen concentrations or high hydride phase fraction. This compressive stress anisotropy matches the observed lens anisotropy, implicating the effect of stress on hydride formation as the cause of the observed lens azimuthal asymmetry. The cause and effect relation between compressive stress and hydride lens anisotropy represents an indirect validation of a key BISON output, the evolved hoop stress associated with hydride formation.

Experimental data report for Test TS-1 Reactivity Initiated Accident Test in NSRR with pre-irradiated BWR fuel rod

International Nuclear Information System (INIS)

Nakamura, Takehiko; Yoshinaga, Makio; Sobajima, Makoto; Fujishiro, Toshio; Horiki, Ohichiro; Yamahara, Takeshi; Ichihashi, Yoshinori; Kikuchi, Teruo

1992-01-01

This report presents experimental data for Test TS-1 which was the first in a series of tests, simulating Reactivity Initiated Accident (RIA) conditions using pre-irradiated BWR fuel rods, performed in the Nuclear Safety Research Reactor (NSRR) in October, 1989. Test fuel rod used in the Test TS-1 was a short-sized BWR (7 x 7) type rod which was fabricated from a commercial rod provided from Tsuruga Unit 1 power reactor. The fuel had an initial enrichment of 2.79 % and burnup of 21.3 GWd/t (bundle average). Pulse irradiation was performed at a condition of stagnant water cooling, atmospheric pressure and ambient temperature using a newly developed double container-type capsule. Energy deposition of the rod in this test was evaluated to be about 61 cal/g·fuel (55 cal/g·fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, fuel burnup measurements, transient behavior of the test rod during pulse irradiation and results of post pulse irradiation examinations are contained in this report. (author)

BWR simulation in a stationary state for the evaluation of fuel cell design

International Nuclear Information System (INIS)

Montes T, J. L.; Ortiz S, J. J.; Perusquia del C, R.; Castillo M, A.

2014-10-01

In this paper the simulation of a BWR in order to evaluate the performance of a set of fuel assemblies under stationary state in three dimensions (3-D) is presented. 15 cases selected from a database containing a total of 18225 cases are evaluated. The main selection criteria were based on the results of the design phase of the power cells in two dimensions (2-D) and 3-D initial study. In 2-D studies the parameters that were used to qualify and select the designs were basically the local power peaking factor and neutron multiplication factor of each fuel cell. In the initial 3-D study variables that defined the quality of results, and from which the selection was realized, are the margins to thermal limits of reactor operation and the value of the effective multiplication factor at the end of cycle operation. From the 2-D and 3-D results of the studies described a second 3-D study was realized, where the optimizations of the fuel reload pattern was carried out. The results presented in this paper correspond to this second 3-D study. It was found that the designs of the fuel cell they had a similar behavior to those provided by the fuel supplier of reference BWR. Particularly it noted the impact of reload pattern on the cold shut down margin. An estimate of the operation costs of reference cycle analyzed with each one designed reload batch was also performed. As a result a positive difference (gain) up to 10,347 M/US D was found. (Author)

BWR spent fuel storage cask performance test. Volume 1. Cask handling experience and decay heat, heat transfer, and shielding data

International Nuclear Information System (INIS)

McKinnon, M.A.; Doman, J.W.; Tanner, J.E.; Guenther, R.J.; Creer, J.M.; King, C.E.

1986-02-01

This report documents a heat transfer and shielding performance test conducted on a Ridihalgh, Eggers and Associates REA 2023 boiling water reactor (BWR) spent fuel storage cask. The testing effort consisted of three parts: pretest preparations, performance testing, and post-test activities. Pretest preparations included conducting cask handling dry runs and characterizing BWR spent fuel assemblies from Nebraska Public Power District's Cooper Nuclear Station. The performance test matrix included 14 runs consisting of two loadings, two cask orientations, and three backfill environments. Post-test activities included calorimetry and axial radiation scans of selected fuel assemblies, in-basin sipping of each assembly, crud collection, video and photographic scans, and decontamination of the cask interior and exterior

Experience using individually supplied heater rods in critical power testing of advanced BWR fuel

Energy Technology Data Exchange (ETDEWEB)

Majed, M.; Morback, G.; Wiman, P. [ABB Atom AB, Vasteras (Sweden)] [and others

1995-09-01

The ABB Atom FRIGG loop located in Vasteras Sweden has during the last six years given a large experience of critical power measurements for BWR fuel designs using indirectly heated rods with individual power supply. The loop was built in the sixties and designed for maximum 100 bar pressure. Testing up to the mid eighties was performed with directly heated rods using a 9 MW, 80 kA power supply. Providing test data to develop critical power correlations for BWR fuel assemblies requires testing with many radial power distributions over the full range of hydraulic conditions. Indirectly heated rods give large advantages for the testing procedure, particularly convenient for variation of individual rod power. A test method being used at Stern Laboratories (formerly Westinghouse Canada) since the early sixties, allows one fuel assembly to simulate all required radial power distributions. This technique requires reliable indirectly heated rods with independently controlled power supplies and uses insulated electric fuel rod simulators with built-in instrumentation. The FRIGG loop was adapted to this system in 1987. A 4MW power supply with 10 individual units was then installed, and has since been used for testing 24 and 25 rod bundles simulating one subbundle of SVEA-96/100 type fuel assemblies. The experience with the system is very good, as being presented, and it is selected also for a planned upgrading of the facility to 15 MW.

BWR core melt progression phenomena: Experimental analyses

International Nuclear Information System (INIS)

Ott, L.J.

1992-01-01

In the BWR Core Melt in Progression Phenomena Program, experimental results concerning severe fuel damage and core melt progression in BWR core geometry are used to evaluate existing models of the governing phenomena. These include control blade eutectic liquefaction and the subsequent relocation and attack on the channel box structure; oxidation heating and hydrogen generation; Zircaloy melting and relocation; and the continuing oxidation of zirconium with metallic blockage formation. Integral data have been obtained from the BWR DF-4 experiment in the ACRR and from BWR tests in the German CORA exreactor fuel-damage test facility. Additional integral data will be obtained from new CORA BWR test, the full-length FLHT-6 BWR test in the NRU test reactor, and the new program of exreactor experiments at Sandia National Laboratories (SNL) on metallic melt relocation and blockage formation. an essential part of this activity is interpretation and use of the results of the BWR tests. The Oak Ridge National Laboratory (ORNL) has developed experiment-specific models for analysis of the BWR experiments; to date, these models have permitted far more precise analyses of the conditions in these experiments than has previously been available. These analyses have provided a basis for more accurate interpretation of the phenomena that the experiments are intended to investigate. The results of posttest analyses of BWR experiments are discussed and significant findings from these analyses are explained. The ORNL control blade/canister models with materials interaction, relocation and blockage models are currently being implemented in SCDAP/RELAP5 as an optional structural component

Investigations on the thermal-hydraulics of a natural circulation cooled BWR fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Kok, H.V.; Hagen, T.H.J.J. van der; Mudde, R.F. [Delft Univ. of Technology (Netherlands)

1995-09-01

A scaled natural circulation loop facility has been built after the Dodewaard Boiling Water Reactor, which is the only operating natural circulation cooled BWR in the world. The loop comprises one fuel assembly, a riser with a downcomer and a condenser with a cooling system. Freon-12 is used as a scaling liquid. This paper reports on the first measurements done with this facility. Quantities like the circulation flow, carry-under and the void-fraction have been measured as a function of power, pressure, liquid level, riser length, condensate temperature and friction factors. The behavior of the circulation flow can be understood by considering the driving force. Special attention has been paid to the carry-under, which has been shown to have a very important impact on the dynamics of a natural circulation cooled BWR.

A practical optimization procedure for radial BWR fuel lattice design using tabu search with a multiobjective function

International Nuclear Information System (INIS)

Francois, J.L.; Martin-del-Campo, C.; Francois, R.; Morales, L.B.

2003-01-01

An optimization procedure based on the tabu search (TS) method was developed for the design of radial enrichment and gadolinia distributions for boiling water reactor (BWR) fuel lattices. The procedure was coded in a computing system in which the optimization code uses the tabu search method to select potential solutions and the HELIOS code to evaluate them. The goal of the procedure is to search for an optimal fuel utilization, looking for a lattice with minimum average enrichment, with minimum deviation of reactivity targets and with a local power peaking factor (PPF) lower than a limit value. Time-dependent-depletion (TDD) effects were considered in the optimization process. The additive utility function method was used to convert the multiobjective optimization problem into a single objective problem. A strategy to reduce the computing time employed by the optimization was developed and is explained in this paper. An example is presented for a 10x10 fuel lattice with 10 different fuel compositions. The main contribution of this study is the development of a practical TDD optimization procedure for BWR fuel lattice design, using TS with a multiobjective function, and a strategy to economize computing time

On the effect of temperature on the threshold stress intensity factor of delayed hydride cracking in light water reactor fuel cladding

Directory of Open Access Journals (Sweden)

Anna-Maria Alvarez Holston

2017-06-01

Full Text Available Delayed hydride cracking (DHC was first observed in pressure tubes in Canadian CANDU reactors. In light water reactors, DHC was not observed until the late 1990s in high-burnup boiling water reactor (BWR fuel cladding. In recent years, the focus on DHC has resurfaced in light of the increased interest in the cladding integrity during interim conditions. In principle, all spent fuel in the wet pools has sufficient hydrogen content for DHC to operate below 300°C. It is therefore of importance to establish the critical parameters for DHC to operate. This work studies the threshold stress intensity factor (KIH to initiate DHC as a function of temperature in Zry-4 for temperatures between 227°C and 315°C. The experimental technique used in this study was the pin-loading testing technique. To determine the KIH, an unloading method was used where the load was successively reduced in a stepwise manner until no cracking was observed during 24 hours. The results showed that there was moderate temperature behavior at lower temperatures. Around 300°C, there was a sharp increase in KIH indicating the upper temperature limit for DHC. The value for KIH at 227°C was determined to be 2.6 ± 0.3 MPa √m.

On the effect of temperature on the threshold stress intensity factor of delayed hydride cracking in light water reactor fuel cladding

Energy Technology Data Exchange (ETDEWEB)

Holston, Anna-MariaAlvarez; Stjarnsater, Johan [Studsvik Nuclear AB, Nykoping (Sweden)

2017-06-15

Delayed hydride cracking (DHC) was first observed in pressure tubes in Canadian CANDU reactors. In light water reactors, DHC was not observed until the late 1990s in high-burnup boiling water reactor (BWR) fuel cladding. In recent years, the focus on DHC has resurfaced in light of the increased interest in the cladding integrity during interim conditions. In principle, all spent fuel in the wet pools has sufficient hydrogen content for DHC to operate below 300°C. It is therefore of importance to establish the critical parameters for DHC to operate. This work studies the threshold stress intensity factor (K{sub IH}) to initiate DHC as a function of temperature in Zry-4 for temperatures between 227°C and 315°C. The experimental technique used in this study was the pin-loading testing technique. To determine the K{sub IH}, an unloading method was used where the load was successively reduced in a stepwise manner until no cracking was observed during 24 hours. The results showed that there was moderate temperature behavior at lower temperatures. Around 300°C, there was a sharp increase in K{sub IH} indicating the upper temperature limit for DHC. The value for K{sub IH} at 227°C was determined to be 2.6 ± 0.3 MPa √m.

Preliminary study on characteristics of equilibrium thorium fuel cycle of BWR

International Nuclear Information System (INIS)

Waris, A.; Kurniadi, R.; Su'ud, Z.; Permana, S.

2007-01-01

One of the main objectives behind the transuranium recycling ideas is not merely to utilize natural resource that is uranium much more efficiently, but to reduce the environmental impact of the radio-toxicity of the nuclear spent fuel. Beside uranium resource, there is thorium which has three times abundance compared to that of uranium which can be utilized as nuclear fuel. On top of that thorium is believed to have less radio-toxicity of spent fuel since its produce smaller amount of higher actinides compared to that of uranium. However, the studies on the thorium utilization in nuclear reactor in particular in light water reactors (LWR) are not performed intensively yet. Therefore, the aim of the present study is to evaluate the characteristics of thorium fuel cycle in LWR, especially boiling water reactor (BWR). To conduct the comprehensive investigations we have employed the equilibrium burnup model (1-3). The equilibrium burnup model is an alternative powerful method since its can handle all possible generated nuclides in any nuclear system. Moreover, this method is a simple time independent method. Hence the equilibrium burnup method could be very useful for evaluating and forecasting the characteristics of any nuclear fuel cycle, even the strange one, e.g. all nuclides are confined in the reactor1). We have employed 1368 nuclides in the equilibrium burnup calculation where 129 of them are heavy metals (HMs). This burnup code then is coupled with SRAC cell calculation code by using PIJ module to compose an equilibrium-cell burnup code. For cell calculation, 26 HMs, 66 fission products (FPs) and one pseudo FP have been utilized. The JENDL 3.2 library has been used in this study. References: 1. A. Waris and H. Sekimoto, 'Characteristics of several equilibrium fuel cycles of PWR', J. Nucl. Sci. Technol., 38, p.517-526, 2001 2. A. Waris, H. Sekimoto, and G. Kastchiev, Influence of Moderator-to-Fuel Volume Ratio on Pu and MA Recycling in Equilibrium Fuel Cycles of

Fuel cycle and waste management. 2. Design of a BWR Core with Over-moderated MOX Fuel Assemblies

International Nuclear Information System (INIS)

Francois, J.L.; Del Campo, C. Martin

2001-01-01

The use of uranium-plutonium mixed-oxide (MOX) fuel in light water reactors is a current practice in several countries. Generally one-third of the reactor core is loaded with MOX fuel assemblies, and the other two-thirds is loaded with uranium assemblies. Nevertheless, the plutonium utilization could be more effective if the full core could be loaded with MOX fuel. In this work, the design of a boiling water reactor (BWR) core fully loaded with over-moderated MOX fuel designs was investigated. In previous work, the design of over-moderated BWR MOX fuel assemblies based on a 10 x 10 lattice was presented; these designs improve the neutron spectrum and the plutonium consumption rate, compared with standard MOX assemblies. To increase the moderator-to-fuel ratio (MFR), two approaches were followed. In the first approach, 8 or 12 fuel rods were replaced by water rods in the 10x10 assembly, which increased the MFR from 1.9 to 2.2 and 2.4, respectively. These designs are called MOX-8WR and MOX-12WR, respectively, in this paper. In the second approach, an 11 x 11 lattice with 24 water rods (11 x 11-24WR) was designed, which is a design with a number of active fuel rods (88) very close to the standard MOX assembly (91). The fuel rod diameter is smaller to preserve the assembly dimensions, and in this last case, the MFR is 2.4. The calculations were performed with the CM-PRESTO three-dimensional steady-state simulator. The nuclear data banks were generated with the HELIOS system, and they were processed by TABGEN to produce tables of nuclear cross sections depending on burnup, void, and exposure weighted void (void history), which are used by CM-PRESTO. One base reload pattern was designed for a BWR/5 rated at 1931 MW(thermal), to be used with the different over-moderated assembly designs. The reload pattern has 112 fresh fuel assemblies (FFAs) out of a total of 444 fuel assemblies and was simulated during 20 cycles with the Haling strategy, until an equilibrium cycle of

Feasibility studies of computed tomography in partial defect detection of spent BWR fuel

International Nuclear Information System (INIS)

Levai, F.; Tikkinen, J.; Tarvainen, M.; Arlt, R.

1990-10-01

Feasibility studies were made for tomographic reconstruction of a cross-sectional activity distribution of a spent nuclear fuel assembly. The purpose was to determine the number of fuel rods (pins) and localize the positisons where pins are missing. The activity distribution map showing the locations of fuel rods in the assembly was reconstructed. The theoretical part of this work consists of simulation of image reconstruction based on theoretically calculated data from a reference assembly model. Evaluation of different image reconstruction techniques was made. Measurements were made in real facility conditions. Gamma radiation from an irradiated 8 x 8 - 1 BWR fuel assembly was measured through a narrow custom made collimator from different angles and positions. The measured data set was used as projections for reconstructing the activity profile of the assembly in cross-sectional plane

Utility experience with BWR-PSMS

International Nuclear Information System (INIS)

Bond, G.R.

1986-01-01

The BWR Power Shape Monitoring System (BWR-PSMS) has proven to be an effective and versatile tool for core monitoring. GPU Nuclear Corporation's (GPUN) Oyster Creek plant has been involved in the PSMS development since its inception, having been selected by EPRI as the initial demonstration site. Beginning with Cycle 10, Oyster Creek has been applying the BWR-PSMS as the primary core monitoring tool. Although the system has been in operation at Oyster Creek for the past several cycles, this is the first time the PSMS was used to monitor compliance to the plant technical specifications, to guide adherence to vendore fuel maneuvering recommendations and to develop data for certain performance records such as fuel burnup, isotopic accounting, etc. This paper will discuss the bases for the decision to apply PSMS as the fundamental core monitoring system, the experience in implementing the PSMS in this mode, activities currently underway or planned related to PSMS, and potential future extensions and applications of PSMS at Oyster Creek

Computational fluid dynamics modeling of two-phase flow in a BWR fuel assembly. Final CRADA Report

International Nuclear Information System (INIS)

Tentner, A.

2009-01-01

A direct numerical simulation capability for two-phase flows with heat transfer in complex geometries can considerably reduce the hardware development cycle, facilitate the optimization and reduce the costs of testing of various industrial facilities, such as nuclear power plants, steam generators, steam condensers, liquid cooling systems, heat exchangers, distillers, and boilers. Specifically, the phenomena occurring in a two-phase coolant flow in a BWR (Boiling Water Reactor) fuel assembly include coolant phase changes and multiple flow regimes which directly influence the coolant interaction with fuel assembly and, ultimately, the reactor performance. Traditionally, the best analysis tools for this purpose of two-phase flow phenomena inside the BWR fuel assembly have been the sub-channel codes. However, the resolution of these codes is too coarse for analyzing the detailed intra-assembly flow patterns, such as flow around a spacer element. Advanced CFD (Computational Fluid Dynamics) codes provide a potential for detailed 3D simulations of coolant flow inside a fuel assembly, including flow around a spacer element using more fundamental physical models of flow regimes and phase interactions than sub-channel codes. Such models can extend the code applicability to a wider range of situations, which is highly important for increasing the efficiency and to prevent accidents.

Optimization of fuel reloads for a BWR using the ant colony system

International Nuclear Information System (INIS)

Esquivel E, J.; Ortiz S, J. J.

2009-10-01

In this work some results obtained during the development of optimization systems are presented, which are employees for the fuel reload design in a BWR. The systems use the ant colony optimization technique. As first instance, a system is developed that was adapted at travel salesman problem applied for the 32 state capitals of Mexican Republic. The purpose of this implementation is that a similarity exists with the design of fuel reload, since the two problems are of combinatorial optimization with decision variables that have similarity between both. The system was coupled to simulator SIMULATE-3, obtaining good results when being applied to an operation cycle in equilibrium for reactors of nuclear power plant of Laguna Verde. (Author)

Artificial intelligence applied to fuel management in BWR type reactors

International Nuclear Information System (INIS)

Ortiz S, J.J.

1998-01-01

In this work two techniques of artificial intelligence, neural networks and genetic algorithms were applied to a practical problem of nuclear fuel management; the determination of the optimal fuel reload for a BWR type reactor. This is an important problem in the design of the operation cycle of the reactor. As a result of the application of these techniques, comparable or even better reloads proposals than those given by expert companies in the subject were obtained. Additionally, two other simpler problems in reactor physics were solved: the determination of the axial power profile and the prediction of the value of some variables of interest at the end of the operation cycle of the reactor. Neural networks and genetic algorithms have been applied to solve many problems of engineering because of their versatility but they have been rarely used in the area of fuel management. The results obtained in this thesis indicates the convenience of undertaking further work on this area and suggest the application of these techniques of artificial intelligence to the solution of other problems in nuclear reactor physics. (Author)

BWR spent fuel transport and storage system for KKL: TN trademark 52L, TN trademark 97L, TN trademark 24 BHL

International Nuclear Information System (INIS)

Sicard, D.; Verdier, A.; Monsigny, P.A.

2004-01-01

The LEIBSTADT (KKL) nuclear power plant in Switzerland has opted to ship spent fuel to a central facility called ZWILAG for interim storage. In the mid-nineties, COGEMA LOGISTICS was contracted by KKL for the supply of the TN trademark a52L and TN trademark 97L transport and storage casks for BWR fuel types. In 2003, KKL also ordered from COGEMA LOGISTICS the supply of six TNae24 BHL transport and storage casks. This paper shows how all the three cask designs have responded to the KKL needs to ship and store BWR spent fuel. In addition, it highlights the already significant operational feedback of the TN trademark 52L and TN trademark 97L casks by the KKL and ZWILAG operators

Development of a fuzzy logic method to build objective functions in optimization problems: application to BWR fuel lattice design

International Nuclear Information System (INIS)

Martin-del-Campo, C.; Francois, J.L.; Barragan, A.M.; Palomera, M.A.

2005-01-01

In this paper we develop a methodology based on the use of the Fuzzy Logic technique to build multi-objective functions to be used in optimization processes applied to in-core nuclear fuel management. As an example, we selected the problem of determining optimal radial fuel enrichment and gadolinia distributions in a typical 'Boiling Water Reactor (BWR)' fuel lattice. The methodology is based on the use of the mathematical capability of Fuzzy Logic to model nonlinear functions of arbitrary complexity. The utility of Fuzzy Logic is to map an input space into an output space, and the primary mechanism for doing this is a list of if-then statements called rules. The rules refer to variables and adjectives that describe those variables and, the Fuzzy Logic technique interprets the values in the input vectors and, based on the set of rules assigns values to the output vector. The methodology was developed for the radial optimization of a BWR lattice where the optimization algorithm employed is Tabu Search. The global objective is to find the optimal distribution of enrichments and burnable poison concentrations in a 10*10 BWR lattice. In order to do that, a fuzzy control inference system was developed using the Fuzzy Logic Toolbox of Matlab and it has been linked to the Tabu Search optimization process. Results show that Tabu Search combined with Fuzzy Logic performs very well, obtaining lattices with optimal fuel utilization. (authors)

Modeling and validation of a mechanistic tool (MEFISTO) for the prediction of critical power in BWR fuel assemblies

International Nuclear Information System (INIS)

Adamsson, Carl; Le Corre, Jean-Marie

2011-01-01

Highlights: → The MEFISTO code efficiently and accurately predicts the dryout event in a BWR fuel bundle, using a mechanistic model. → A hybrid approach between a fast and robust sub-channel analysis and a three-field two-phase analysis is adopted. → MEFISTO modeling approach, calibration, CPU usage, sensitivity, trend analysis and performance evaluation are presented. → The calibration parameters and process were carefully selected to preserve the mechanistic nature of the code. → The code dryout prediction performance is near the level of fuel-specific empirical dryout correlations. - Abstract: Westinghouse is currently developing the MEFISTO code with the main goal to achieve fast, robust, practical and reliable prediction of steady-state dryout Critical Power in Boiling Water Reactor (BWR) fuel bundle based on a mechanistic approach. A computationally efficient simulation scheme was used to achieve this goal, where the code resolves all relevant field (drop, steam and multi-film) mass balance equations, within the annular flow region, at the sub-channel level while relying on a fast and robust two-phase (liquid/steam) sub-channel solution to provide the cross-flow information. The MEFISTO code can hence provide highly detailed solution of the multi-film flow in BWR fuel bundle while enhancing flexibility and reducing the computer time by an order of magnitude as compared to a standard three-field sub-channel analysis approach. Models for the numerical computation of the one-dimensional field flowrate distributions in an open channel (e.g. a sub-channel), including the numerical treatment of field cross-flows, part-length rods, spacers grids and post-dryout conditions are presented in this paper. The MEFISTO code is then applied to dryout prediction in BWR fuel bundle using VIPRE-W as a fast and robust two-phase sub-channel driver code. The dryout power is numerically predicted by iterating on the bundle power so that the minimum film flowrate in the

NSRR experiment with un-irradiated uranium-zirconium hydride fuel. Design, fabrication process and inspection data of test fuel rod

Energy Technology Data Exchange (ETDEWEB)

Sasajima, Hideo; Fuketa, Toyoshi; Ishijima, Kiyomi; Kuroha, Hiroshi; Ikeda, Yoshikazu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Aizawa, Keiichi

1998-08-01

An experiment plan is progressing in the Nuclear Safety Research Reactor (NSRR) to perform pulse-irradiation with uranium-zirconium hydride (U-ZrH{sub x}) fuel. This fuel is widely used in the training research and isotope production reactor of GA (TRIGA). The objectives of the experiment are to determine the fuel rod failure threshold and to investigate fuel behavior under simulated reactivity initiated accident (RIA) conditions. This report summarizes design, fabrication process and inspection data of the test fuel rods before pulse-irradiation. The experiment with U-ZrH{sub x} fuel will realize precise safety evaluation, and improve the TRIGA reactor performance. The data to be obtained in this program will also contribute development of next-generation TRIGA reactor and its safety evaluation. (author)

Verification of a BWR code package by gamma scan measurements

International Nuclear Information System (INIS)

Nakajima, Tsuyoshi; Iwamoto, Tatsuya; Kumanomido, Hironori

1996-01-01

High-burnup 8 x 8 fuel with a large central water rod (called step 2 fuel) has been recently introduced to the latest Japanese boiling water reactor (BWR) plants. Lanthanum-140 gamma intensity is almost directly related to nodal powers. By gamma scan measurement, the axial distribution of 140 La in the exposed fuel was measured at the end of cycle (EOC) 1 and was compared with the calculation by a BWR code package TGBLA/LOGOS. The multienrichment fuel-type core (MEC) design was adopted for the initial cycle core of the plants. The MEC design contains three different enrichment types of fuels to simulate the equilibrium cycles, achieve much higher discharge exposure, and save fuel cycle cost, and the low-enrichment fuels are loaded in periphery and in control cells. Such MEC design could be a challenge to the BWR design methods because of the large spectrum mismatch among the fuel assemblies of the different enrichments. The aforementioned comparison has shown that the accuracy of the TGBLA/LOGOS code package is satisfactory

Obtention of the constitutive equation of hydride blisters in fuel cladding from nanoindentation tests

Energy Technology Data Exchange (ETDEWEB)

Martin Rengel, M.A., E-mail: mamartin.rengel@upm.es [E.T.S.I. Caminos, Canales y Puertos, Universidad Politécnica de Madrid, c/ Profesor Aranguren, 3, E-28040 Madrid (Spain); Gomez, F.J., E-mail: javier.gomez@amsimulation.com [Advanced Material Simulation, AMS, Bilbao (Spain); Rico, A., E-mail: alvaro.rico@urjc.es [DIMME, Universidad Rey Juan Carlos, Mostoles (Spain); Ruiz-Hervias, J. [E.T.S.I. Caminos, Canales y Puertos, Universidad Politécnica de Madrid, c/ Profesor Aranguren, 3, E-28040 Madrid (Spain); Rodriguez, J. [DIMME, Universidad Rey Juan Carlos, Mostoles (Spain)

2017-04-15

It is well known that the presence of hydrides in nuclear fuel cladding may reduce its mechanical and fracture properties. This situation may be worsened as a consequence of the formation of hydride blisters. These blisters are zones with an extremely high hydrogen concentration and they are usually associated to the oxide spalling which may occur at the outer surface of the cladding. In this work, a method which allows us to reproduce, in a reliable way, hydride blisters in the laboratory has been devised. Depth-sensing indentation tests with a spherical indenter were conducted on a hydride blister produced in the laboratory with the aim of measuring its mechanical behaviour. The plastic stress-strain curve of the hydride blister was calculated for first time by combining depth-sensing indentation tests results with an iterative algorithm using finite element simulations. The algorithm employed reduces, in each iteration, the differences between the numerical and the experimental results by modifying the stress-strain curve. In this way, an almost perfect adjustment of the experimental data was achieved after several iterations. The calculation of the constitutive equation of the blister from nanoindentation tests, may involve a lack of uniqueness. To evaluate it, a method based on the optimization of parameters of analytical equations has been proposed in this paper. An estimation of the error which involves this method is also provided.

BWR chemistry control status: a summary of industry chemistry status relative to the BWR water chemistry guidelines

International Nuclear Information System (INIS)

Garcia, S.E.; Giannelli, J.F.; Jarvis, M.L.

2010-01-01

The EPRI Boiling Water Reactor (BWR) Water Chemistry Guidelines were revised and issued in October 2008. The 2008 Revision of the Guidelines continues to focus on intergranular stress corrosion cracking (IGSCC), which can limit the service life of susceptible materials and components exposed to water chemistry environments. The 2008 Revision also places increased emphasis on fuel performance and meeting the industry goal of zero fuel failures by 2010. As an industry consensus document, the Guidelines were created to provide proactive water chemistry control strategies for mitigating IGSCC, maintaining fuel integrity and controlling radiation fields. The Guidelines provide a technically-based framework for an effective BWR water chemistry program. This paper provides an overview of industry experience relative to the Guidelines. Over the past few years, many BWR units have implemented noble metal chemical application technologies either during plant hot or cold shutdown or at normal power operating conditions. This paper explores plant experience with optimized water chemistry, implementation of various additive chemistries such as noble metal application and zinc addition, and compliance with the Guidelines recommendations. Depleted zinc oxide addition has been broadly applied across the BWR fleet since the 1980s. The guidance for zinc addition has been revised in the Guidelines to reflect concerns with fuel performance. While zinc addition is a successful method for shutdown dose rate control, concerns still exist for high zinc deposition on fuel surfaces, especially when feedwater iron is elevated and as fuel cores are being driven to provide maximum power output over longer fuel cycles. Recent plant experience has shown that the utilization of online noble metal application and continued zinc addition may provide additional benefits for radiation control. Dose rate experiences at plants utilizing the online noble metal application technology and zinc addition

BWR chemistry control status: a summary of industry chemistry status relative to the BWR water chemistry guidelines

Energy Technology Data Exchange (ETDEWEB)

Garcia, S.E., E-mail: sgarcia@epri.com [Electric Power Research Inst. (EPRI), Palo Alto, California (United States); Giannelli, J.F.; Jarvis, M.L., E-mail: jgiannelli@finetech.com [Finetech, Inc., Parsippany, NJ (United States)

2010-07-01

The EPRI Boiling Water Reactor (BWR) Water Chemistry Guidelines were revised and issued in October 2008. The 2008 Revision of the Guidelines continues to focus on intergranular stress corrosion cracking (IGSCC), which can limit the service life of susceptible materials and components exposed to water chemistry environments. The 2008 Revision also places increased emphasis on fuel performance and meeting the industry goal of zero fuel failures by 2010. As an industry consensus document, the Guidelines were created to provide proactive water chemistry control strategies for mitigating IGSCC, maintaining fuel integrity and controlling radiation fields. The Guidelines provide a technically-based framework for an effective BWR water chemistry program. This paper provides an overview of industry experience relative to the Guidelines. Over the past few years, many BWR units have implemented noble metal chemical application technologies either during plant hot or cold shutdown or at normal power operating conditions. This paper explores plant experience with optimized water chemistry, implementation of various additive chemistries such as noble metal application and zinc addition, and compliance with the Guidelines recommendations. Depleted zinc oxide addition has been broadly applied across the BWR fleet since the 1980s. The guidance for zinc addition has been revised in the Guidelines to reflect concerns with fuel performance. While zinc addition is a successful method for shutdown dose rate control, concerns still exist for high zinc deposition on fuel surfaces, especially when feedwater iron is elevated and as fuel cores are being driven to provide maximum power output over longer fuel cycles. Recent plant experience has shown that the utilization of online noble metal application and continued zinc addition may provide additional benefits for radiation control. Dose rate experiences at plants utilizing the online noble metal application technology and zinc addition

Axial profiles of burned and fraction of holes for calculations of criticality with credit for BWR fuel burning

International Nuclear Information System (INIS)

Casado Sanchez, C.; Rubio Oviedo, P.

2014-01-01

This paper presents a method to define surround profiles of burning and fraction of holes suited for use in applications of credit to burning of BWR fuel from results obtained with the module STARBUCS of SCALE. (Author)

Calibration of the TVO spent BWR reference fuel assembly

International Nuclear Information System (INIS)

Tarvainen, M.; Baecklin, A.; Haakanson, A.

1992-02-01

In 1989 the Support Programmes of Finland (FSP) and Sweden (SSP) initiated a joint task to cross calibrate the burnup of the IAEA spent BWR reference fuel assembly at the TVO AFR storage facility (TVO KPA-STORE) in Finland. The reference assembly, kept separately under the IAEA seal, is used for verification measurements of spent fuel by GBUV method (SG-NDA-38). The cross calibration was performed by establishing a calibration curve, 244 Cm neutron rate versus burnup, using passive neutron assay (PNA) measurements. The declared burnup of the reference assembly was compared with the burnup value deduced from the calibration curve. A calibration line was also established by using the GBUV method with the aid of high resolution gamma ray spectrometry (HRGS). Normalization between the two different facilities was performed using sealed neutron and gamma calibration sources. The results of the passive neutron assay show consistency, better than 1 %, between the declared mean burnup of the reference assembly and the burnup deduced from the calibration curve. The corresponding consistency is within +-2 % for the HRGS measurements

Automatic refueling platform and CRD remote handling device for BWR plant

International Nuclear Information System (INIS)

Kato, Hiroaki; Takagi, Kaoru

1978-01-01

In BWR plants, machines for replacing fuel assemblies and control rod drives are usually operated directly by personnel. An automatic refueling platform and a CRD remote handling device aiming at radiation exposure reduction and handling perfectness are described, which are already used in BWR plants. Automation of the former is achieved in transporting fuel assemblies between a reactor pressure vessel and a fuel storage pool, shuffling fuel assemblies in a reactor core and moving fuel assemblies in a fuel storage pool. In the latter, replacement of CRDs is nearly all performed remotely. (Mori, K.)

Identification and characterization of a new Zirconium hydride

International Nuclear Information System (INIS)

Zhao, Z.

2007-01-01

In order to control the integrity of the fuel clad, alloy of zirconium, it is necessary to predict the behavior of zirconium hydrides in the environment (temperature, stress...), at a microscopic scale. A characterization study by TEM of hydrides has been realized. It shows little hydrides about 500 nm, in hydride Zircaloy 4. Then a more detailed study identified a new hydride phase presented in this paper. (A.L.B.)

Observations of crud deposits, corrosion and erosion of BWR and PWR fuel

International Nuclear Information System (INIS)

Bairiot, H.

1983-01-01

The BWR experience is limited to one reactor but the PWR experience covers a wide range of successive generations of power plants (7 in total). The systems are described and their water chemistry briefly commented. Some R and D performed on the effects of the operating regimes (steady state and transients) are summarized. Observations made by pool-side inspections and postirradiation examinations of fuel are outlined concerning water chemistry effects (crud deposits and corrosion) and ''mechanical'' coolant-cladding interaction (chip deposits and baffle jetting). (author)

Advanced BWR core component designs and the implications for SFD analysis

International Nuclear Information System (INIS)

Ott, L.J.

1997-01-01

Prior to the DF-4 boiling water reactor (BWR) severe fuel damage (SFD) experiment conducted at the Sandia National Laboratories in 1986, no experimental data base existed for guidance in modeling core component behavior under postulated severe accident conditions in commercial BWRs. This paper will present the lessons learned from the DF-4 experiment (and subsequent German CORA BWR SFD tests) and the impact on core models in the current generation of SFD codes. The DF-4 and CORA BWR test assemblies were modeled on the core component designs circa 1985; that is, the 8 x 8 fuel assembly with two water rods and a cruciform control blade constructed of B 4 C-filled tubelets. Within the past ten years, the state-of-the-art with respect to BWR core component development has out-distanced the current SFD experimental data base and SFD code capabilities. For example, modern BWR control blade design includes hafnium at the tips and top of each control blade wing for longer blade operating lifetimes; also water rods have been replaced by larger water channels for better neutronics economy; and fuel assemblies now contain partial-length fuel rods, again for better neutronics economy. This paper will also discuss the implications of these advanced fuel assembly and core component designs on severe accident progression and on the current SFD code capabilities

Development of CFD analysis method based on droplet tracking model for BWR fuel assemblies

International Nuclear Information System (INIS)

Onishi, Yoichi; Minato, Akihiko; Ichikawa, Ryoko; Mashara, Yasuhiro

2011-01-01

It is well known that the minimum critical power ratio (MCPR) of the boiling water reactor (BWR) fuel assembly depends on the spacer grid type. Recently, improvement of the critical power is being studied by using a spacer grid with mixing devices attaching various types of flow deflectors. In order to predict the critical power of the improved BWR fuel assembly, we have developed an analysis method based on the consideration of detailed thermal-hydraulic mechanism of annular mist flow regime in the subchannels for an arbitrary spacer type. The proposed method is based on a computational fluid dynamics (CFD) model with a droplet tracking model for analyzing the vapor-phase turbulent flow in which droplets are transported in the subchannels of the BWR fuel assembly. We adopted the general-purpose CFD software Advance/FrontFlow/red (AFFr) as the base code, which is a commercial software package created as a part of Japanese national project. AFFr employs a three-dimensional (3D) unstructured grid system for application to complex geometries. First, AFFr was applied to single-phase flows of gas in the present paper. The calculated results were compared with experiments using a round cellular spacer in one subchannel to investigate the influence of the choice of turbulence model. The analyses using the large eddy simulation (LES) and re-normalisation group (RNG) k-ε models were carried out. The results of both the LES and RNG k-ε models show that calculations of velocity distribution and velocity fluctuation distribution in the spacer downstream reproduce the experimental results qualitatively. However, the velocity distribution analyzed by the LES model is better than that by the RNG k-ε model. The velocity fluctuation near the fuel rod, which is important for droplet deposition to the rod, is also simulated well by the LES model. Then, to examine the effect of the spacer shape on the analytical result, the gas flow analyses with the RNG k-ε model were performed

Effect of high burn-up and MOX fuel on reprocessing, vitrification and disposal of PWR and BWR spent fuels based on accurate burn-up calculation

Energy Technology Data Exchange (ETDEWEB)

Yoshikawa, T.; Iwasaki, T.; Wada, K. [Tohoku Univ., Graduate School of Engineering, Dept. of Quantum Science and Energy Engineering, Sendai 980-8579 (Japan); Suyama, K. [Japan Atomic Energy Agency, Shirakata-Shirane 2-4, Naka-gun, Ibaraki-ken 319-1195 (Japan)

2006-07-01

To examine the procedures of the reprocessing, the vitrification and the geologic disposal, precise burn-up calculation for high burn-up and MOX fuels has been performed for not only PWR but also BWR by using SWAT and SWAT2 codes which are the integrated bum-up calculation code systems combined with the bum-up calculation code, ORIGEN2, and the transport calculation code, SRAC (the collision probability method) or MVP (the continuous energy Monte Carlo method), respectively. The calculation results shows that all of the evaluated items (heat generation and concentrations of Mo and Pt) largely increase and those significantly effect to the current procedures of the vitrification and the geologic disposal. The calculation result by SWAT2 confirms that the bundle calculation is required for BWR to be discussed about those effects in details, especially for the MOX fuel. (authors)

FEMAXI-7 analysis on behavior of medium and high burnup BWR fuels during base-irradiation and power ramp

Energy Technology Data Exchange (ETDEWEB)

Ogiyanagi, Jin, E-mail: ohgiyanagi.jin@jaea.go.jp [Japan Atomic Energy Agency, 2-4 Shirane, Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Hanawa, Satoshi; Suzuki, Motoe; Nagase, Fumihisa [Japan Atomic Energy Agency, 2-4 Shirane, Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan)

2012-12-15

Highlights: Black-Right-Pointing-Pointer Two power ramp experiments of BWR fuels were analyzed by FEMAXI-7 code. Black-Right-Pointing-Pointer Calculated FGR and cladding deformation showed reasonable agreement with PIE data. Black-Right-Pointing-Pointer High temperature FGR could be predicted by the enhanced Turnbull FG diffusion constant. Black-Right-Pointing-Pointer Local PCMI model in the code could reasonably predict cladding ridging deformation. - Abstract: Irradiation behavior of medium and high burnup BWR fuels during base-irradiation and subsequent power ramp test is analyzed by a fuel performance code FEMAXI-7. The code has a 1.5-D cylindrical geometry (4 axial segments) to have a coupled solution of thermal analysis and FEM mechanical analysis. Two kinds of target fuels are selected; one was subjected to a power ramp test in the DR3 reactor at RISO after the base-irradiation in a commercial BWR, and the other was subjected to the power ramp test in the DR3 reactor after the base-irradiation in the Halden boiling water reactor. The calculated values such as fission gas release after the base-irradiation and a cladding diameter profile before and after the ramp test show a reasonable agreement with measured data. In addition, the calculated ridging deformation of the cladding before and after the ramp test, which is obtained by using a local pellet-cladding mechanical interaction (PCMI) analysis geometry in FEMAXI-7, is compared with the measured data, and it is found that the FEMAXI-7 code is applicable to the local PCMI analysis of medium and high burnup rods under normal operation and power ramp conditions.

Metal hydride store for hydrogen supply and cooling of fuel cell vehicles; Metallhydridspeicher zur Wasserstoffversorgung und Kuehlung von Brennstoffzellenfahrzeugen

Energy Technology Data Exchange (ETDEWEB)

Wenger, David

2009-07-01

In the context of the author's work, a compact, dynamic metal hydride store was developed which in addition to storing hydrogen can also support the thermomanagement of fuel cell vehicles in extreme situations. The requirements were identified using a semiphysical model of a fuel cell vehicle, and a store was dimensioned accordingly. Additionally, a metal hydride store model was developed on the basis of the balance equations. The model was validated by experiments on a specially designed and constructed store. The simulations enable the optimisation of the store geometry and the prediction of its efficiency in a given operating cycle. (orig.)

BWR-spent fuel transport and storage with the TN trademark 9/4 and TN trademark 24BH casks

International Nuclear Information System (INIS)

Wattez, L.; Marguerat, Y.; Hoesli, C.

2004-01-01

The Swiss Nuclear Utilities have started in 2001 to store spent fuel in dry metallic dual-purpose casks in ZWILAG, the Swiss interim storage facility. BKW FMB Energy Ltd., as Muehleberg Nuclear Power Plant owner, is involved in this process and has selected to store its spent fuel, a new high capacity dual-purpose cask, the TN trademark 24BH. For the transport in a medium size cask, COGEMA LOGISTICS has developed a new cask, the TN trademark 9/4, to replace the NTL9 cask, which performed numerous transports of BWR spent fuel in the past decades. Licensed IAEA 1996, the TN trademark 9/4 is a 40 ton transport cask, for 7 BWR high burn-up spent fuel assemblies. The spent fuel assemblies can be transferred in the ZWILAG hot cell in the TN trademark 24BH cask. The first use of these casks took place in 2003. Ten TN trademark 9/4 transports were performed, and one TN trademark 24BH was loaded. After a brief presentation of the operational aspects, the paper will focus on the TN trademark 24BH high capacity dual purpose cask, the TN trademark 9/4 transport cask and describe in detail their characteristics and possibilities

Large bundle BWR test CORA-18: Test results

International Nuclear Information System (INIS)

Hagen, S.; Hofmann, P.; Noack, V.; Sepold, L.; Schanz, G.; Schumacher, G.

1998-04-01

The CORA out-of-pile experiments are part of the international Severe Fuel Damage (SFD) Program. They were performed to provide information on the damage progression of Light Water Reactor (LWR) fuel elements in Loss-of-coolant Accidents in the temperature range 1200 C to 2400 C. CORA-18 was the large BWR bundle test corresponding to the PWR test CORA-7. It should investigate if there exists an influence of the BWR bundle size on the fuel damage behaviour. Therefore, the standard-type BWR CORA bundle with 18 fuel rod simulators was replaced by a large bundle with two additional surrounding rows of 30 rods (48 rods total). Power input and steam flow were increased proportionally to the number of fuel rod simulators to give the same initial heat-up rate of about 1 K/s as in the smaller bundles. Emphasis was put on the initial phase of the damage progression. More information on the chemical composition of initial and intermediate interaction products and their relocation behaviour should be obtained. Therefore, power and steam input were terminated after the onset of the temperature escalation. (orig.) [de

Thermohydraulic analysis of BWR and PWR spent fuel assemblies contained within square canisters

International Nuclear Information System (INIS)

Wiles, L.E.; McCann, R.A.

1981-09-01

This report presents the results of several thermohydraulic simulations of spent fuel assembly/canister configurations performed in support of a program investigating the feasibility of storing spent nuclear fuel assemblies in canisters that would be stored in an air environment. Eleven thermohydraulic simulations were performed. Five simulations were performed using a single BWR fuel assembly/canister design. The various cases were defined by changing the canister spacing and the heat generation rate of the fuel assembly. For each simulation a steady-state thermohydraulic solution was achieved for the region inside the canister. Similarly, six simulations were performed for a single PWR fuel assembly/canister design. The square fuel rod arrays were contained in square canisters which would permit closer packing of the canisters in a storage facility. However, closer packing of the canisters would result in higher fuel temperatures which would possibly have an adverse impact on fuel integrity. Thus, the most important aspect of the analysis was to define the peak fuel assembly temperatures for each case. These results are presented along with various temperature profiles, heat flux distributions, and air velocity profiles within the canister. 48 figures, 4 tables

Thermal analyses for the spend fuel pool of Taiwan BWR plants during the loss of cooling accident

Energy Technology Data Exchange (ETDEWEB)

Chen, B-Y.; Yeh, C-L.; Wei, W-C.; Chen, Y-S., E-mail: onepicemine@iner.gov.tw, E-mail: clinyeh@iner.gov.tw, E-mail: hn150456@iner.gov.tw, E-mail: yschen@iner.gov.tw [Inst. of Nuclear Energy Research, Longtan Township, Taoyuan County, Taiwan (China)

2014-07-01

After the Fukushima nuclear accident, the safety of the spent fuel pool has become an important concern. In this study, thermal analysis of the spent fuel pool under a loss of cooling accident is performed. The BWR spent fuel pools in Taiwan are investigated, including the Chinshan, Kuosheng, and Lungmen plants. The transient pool temperature and level behaviors are calculated based on lumped energy balance. After the pool level drops below the top of the fuel, the peak cladding temperature is predicted by the Computational Fluid Dynamics (CFD) analysis. The influence to the cladding temperature of the uniform and checkboard fuel loading patterns is also investigated. (author)

Enhancing BWR proliferation resistance fuel with minor actinides

Science.gov (United States)

Chang, Gray S.

2009-03-01

To reduce spent fuel for storage and enhance the proliferation resistance for the intermediate-term, there are two major approaches (a) increase the discharged spent fuel burnup in the advanced light water reactor- LWR (Gen-III Plus), which not only can reduce the spent fuel for storage, but also increase the 238Pu isotopes ratio to enhance the proliferation resistance, and (b) use of transuranic nuclides ( 237Np and 241Am) in the high burnup fuel, which can drastically increase the proliferation resistance isotope ratio of 238Pu/Pu. For future advanced nuclear systems, minor actinides (MA) are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. As a result, MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. In the study, a typical boiling water reactor (BWR) fuel unit lattice cell model with UO 2 fuel pins will be used to investigate the effectiveness of minor actinide reduction approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance in the intermediate-term goal for future nuclear energy systems. To account for the water coolant density variation from the bottom (0.76 g/cm 3) to the top (0.35 g/cm 3) of the core, the axial coolant channel and fuel pin were divided to 24 nodes. The MA transmutation characteristics at different elevations were compared and their impact on neutronics criticality discussed. The concept of MARA, which involves the use of transuranic nuclides ( 237Np and/or 241Am), significantly increases the 238Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in

BWR Radiation Assessment and Control Program: assessment and control of BWR radiation fields. Volume 1. Executive summary

International Nuclear Information System (INIS)

Anstine, L.D.

1983-05-01

This report covers work on the BWR Radiation Assessment and Control (BRAC) Program from 1978 to 1982. The major activities during this report period were assessment of the radiation-level trends in BWRs, evaluation of the effects of forward-pumped heater drains on BWR water quality, installation and operation of a corrosion-product deposition loop in an operating BWR, and analyzation of fuel-deposit samples from two BWRs. Radiation fields were found to be controlled by cobalt-60 and to vary from as low as 50 mr/hr to as high as 800 mr/hr on the recirculation-system piping. Detailed information on BWR corrosion films and system deposits is presented in the report. Additionally, the results of an oxygen-injection experiment and recontamination monitoring studies are provided

U-8 wt %Mo and 7 wt %Mo alloys powder obtained by an hydride-de hydride process

International Nuclear Information System (INIS)

Balart, Silvia N.; Bruzzoni, Pablo; Granovsky, Marta S.; Gribaudo, Luis M. J.; Hermida, Jorge D.; Ovejero, Jose; Rubiolo, Gerardo H.; Vicente, Eduardo E.

2000-01-01

Uranium-molybdenum alloys are been tested as a component in high-density LEU dispersion fuels with very good performances. These alloys need to be transformed to powder due to the manufacturing requirements of the fuels. One method to convert ductile alloys into powder is the hydride-de hydride process, which takes advantage of the ability of the U-α phase to transform to UH 3 : a brittle and relatively low-density compound. U-Mo alloys around 7 and 8 wt % Mo were melted and heat treated at different temperature ranges in order to partially convert γ -phase to α -phase. Subsequent hydriding transforms this α -phase to UH 3 . The volume change associated to the hydride formation embrittled the material which ends up in a powdered alloy. Results of the optical metallography, scanning electron microscopy, X-ray diffraction during different steps of the process are shown. (author)

Evaluation of the radial design of fuel cells in an operation cycle of a BWR reactor

International Nuclear Information System (INIS)

Gonzalez C, J.; Martin del Campo M, C.

2003-01-01

This work is continuation of one previous in the one that the application of the optimization technique called Tabu search to the radial design of fuel cells of boiling water reactors (BWR, Boiling Water Reactor) is presented. The objective function used in the optimization process only include neutron parameters (k-infinite and peak of radial power) considering the cell at infinite media. It was obtained to reduce the cell average enrichment completing the characteristics of reactivity of an original cell. The objective of the present work is to validate the objective function that was used for the radial design of the fuel cell (test cell), analyzing the operation of a one cycle of the reactor in which fuels have been fresh recharged that contain an axial area with the nuclear database of the cell designed instead of the original cell. For it is simulated it with Cm-Presto the cycle 10 of the reactor operation of the Unit 1 of the Nuclear Power station of Laguna Verde (U1-CNLV). For the cycle evaluation its were applied so much the simulation with the Haling strategy, as the simulation of the one cycle with control rod patterns and they were evaluated the energy generation and several power limits and reactivity that are used as design parameters in fuel reloads of BWR reactors. The results at level of an operation cycle of the reactor, show that the objective function used in the optimization and radial design of the cell is adequate and that it can induce to one good use of the fuel. (Author)

Analysis of effects of pellet-cladding bonding on trapping of the released fission gases in high burnup KKL BWR fuels

Energy Technology Data Exchange (ETDEWEB)

Brankov, Vladimir [Laboratory for Reactor Physics and Systems Behaviour at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Swiss Federal Institute of Technology Lausanne (EPFL), Route Cantonale, 1015 Lausanne (Switzerland); Khvostov, Grigori; Mikityuk, Konstantin [Laboratory for Reactor Physics and Systems Behaviour at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Pautz, Andreas [Laboratory for Reactor Physics and Systems Behaviour at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Swiss Federal Institute of Technology Lausanne (EPFL), Route Cantonale, 1015 Lausanne (Switzerland); Restani, Renato; Abolhassani, Sousan [Laboratory for Nuclear Materials at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Ledergerber, Guido [Kernkraftwerk Leibstadt, 5325 Leibstadt (Switzerland); Wiesenack, Wolfgang [Institutt for Energiteknikk - OECD Halden Reactor Project, Os Allé 5, 1777 Halden (Norway)

2016-08-15

Highlights: • Explanation for the scatter in measured fission gas release in high-BU BWR fuel rods. • Partial fuel-clad bond layer formation in high-BU BWR fuel. • Hypothesis for fission gas trapping facilitated by the pellet-cladding bond layer. • Correlation between burnup asymmetry and the quantity of trapped fission gas. • Implications of the trapped FG in LOCA transient. - Abstract: The first part of the paper presents results of a numerical analysis of the fuel behavior during base irradiation in the Kernkraftwerk Leibstadt Boiling Water Reactor (KKL BWR) using EPRI’s FALCON code coupled to GRSW-A – an advanced model for fuel swelling and fission gas release. Post-irradiation examinations conducted at the Paul Scherrer Institute’s (PSI) hot laboratory gave evidence of a distinct circumferential non-uniformity of local burnup at pellet surfaces. For several fuel samples, intact pellet-cladding bonding areas on the high burnup sides of the pellets at high burnup above ∼70 MWd/kgU were observed. It is hypothesized that a part of the fission gases, which are expected to be released by those areas, can be trapped and do not reach the rod plenum. In this paper, a simple approach to modeling of fission gas trapping is employed which reveals a potential correlation between the position of the rod within the fuel assembly (and therefore the degree of circumferential burnup non-uniformity) and the degree of fission gas trapping. A model is suggested to correlate the amount of locally trapped gas with the integral of the local contact pressure and the degree of circumferential burnup non-uniformity. The model is calibrated with available measurements of FGR from rod puncturing at the level of the plenums. In future work, the hypothesis about the axial distribution of trapped fission gas will be extrapolated to the Loss-Of-Coolant Accident (LOCA) analysis as an attempt to explain the fission gas release observed in some samples fabricated from

A probabilistic analysis of PWR and BWR fuel rod performance using the code CASINO-SLEUTH

International Nuclear Information System (INIS)

Bull, A.J.

1987-01-01

This paper presents a brief description of the Monte Carlo and response surface techniques used in the code, and a probabilistic analysis of fuel rod performance in PWR and BWR applications. The analysis shows that fission gas release predictions are very sensitive to changes in certain of the code's inputs, identifies the most dominant input parameters and compares their effects in the two cases. (orig./HP)

Metal hydride and pyrophoric fuel additives for dicyclopentadiene based hybrid propellants

Science.gov (United States)

Shark, Steven C.

The purpose of this study is to investigate the use of reactive energetic fuel additives that have the potential to increase the combustion performance of hybrid rocket propellants in terms of solid fuel regression rate and combustion efficiency. Additives that can augment the combustion flame zone in a hybrid rocket motor by means of increased energy feedback to the fuel grain surface are of great interest. Metal hydrides have large volumetric hydrogen densities, which gives these materials high performance potential as fuel additives in terms of specifc impulse. The excess hydrogen and corresponding base metal may also cause an increase in the hybrid rocket solid fuel regression rate. Pyrophoric additives also have potential to increase the solid fuel regression rate by reacting more readily near the burning fuel surface providing rapid energy feedback. An experimental performance evaluation of metal hydride fuel additives for hybrid rocket motor propulsion systems is examined in this study. Hypergolic ignition droplet tests and an accelerated aging study revealed the protection capabilities of Dicyclopentadiene (DCPD) as a fuel binder, and the ability for unaided ignition. Static hybrid rocket motor experiments were conducted using DCPD as the fuel. Sodium borohydride (NabH4) and aluminum hydride (AlH3) were examined as fuel additives. Ninety percent rocket grade hydrogen peroxide (RGHP) was used as the oxidizer. In this study, the sensitivity of solid fuel regression rate and characteristic velocity (C*) efficiency to total fuel grain port mass flux and particle loading is examined. These results were compared to HTPB combustion performance as a baseline. Chamber pressure histories revealed steady motor operation in most tests, with reduced ignition delays when using NabH4 as a fuel additive. The addition of NabH4 and AlH3 produced up to a 47% and 85% increase in regression rate over neat DCPD, respectively. For all test conditions examined C* efficiency ranges

Field test and evaluation of the IAEA coincidence collar for the measurement of unirradiated BWR fuel assemblies

International Nuclear Information System (INIS)

Menlove, H.O.; Keddar, A.

1982-12-01

The neutron coincidence counter has been field tested and evaluated for the measurement of boiling-water-reactor (BWR) fuel assemblies at the ASEA-ATOM Fuel Fabrication Facility. The system measures the 235 U content per unit length of full fuel assemblies using neutron interrogation and coincidence counting. The 238 U content is measured in the passive mode without the AmLi neutron interrogatioin source. The field tests included both standard production movable fuel rods to investigate enrichment and absorber variations. Results gave a response standard deviation of 0.9% for the active case and 2.1% for the passive case in 1000-s measurement times. 10 figures, 2 tables

BWR Refill-Reflood Program, Task 4.7 - model development: TRAC-BWR component models

International Nuclear Information System (INIS)

Cheung, Y.K.; Parameswaran, V.; Shaug, J.C.

1983-09-01

TRAC (Transient Reactor Analysis Code) is a computer code for best-estimate analysis for the thermal hydraulic conditions in a reactor system. The development and assessment of the BWR component models developed under the Refill/Reflood Program that are necessary to structure a BWR-version of TRAC are described in this report. These component models are the jet pump, steam separator, steam dryer, two-phase level tracking model, and upper-plenum mixing model. These models have been implemented into TRAC-B02. Also a single-channel option has been developed for individual fuel-channel analysis following a system-response calculation

Strategies of operation cycles in BWR type reactors

International Nuclear Information System (INIS)

Molina, D.; Sendino, F.

1996-01-01

The article analyzes the operation cycles in BWR type reactors. The cycle size of operation is the consequence on the optimization process of the costs with the technical characteristics of nuclear fuel and the characteristics of demand and production. The authors analyze the cases of Garona NP and Cofrentes NP, both with BWR reactors. (Author)

Application of gadolinia credit to cask transportation of BWR-STEP3 SFAs

International Nuclear Information System (INIS)

Kikuchi, Tsukasa; Mitsuhashi, Ishi; Ito, Dai-ichiro; Nakamura, Yu

2003-01-01

Instead of the fresh-fuel assumption, the application of gadolinia credit to cask transportation of BWR SFAs is studied. Its efficacy for BWR-STEP2 SFAs had already been estimated. This paper reports on the application of gadolinia credit to cask transportation of BWR-STEP3 SFAs. (author)

BWR Assembly Optimization for Minor Actinide Recycling

Energy Technology Data Exchange (ETDEWEB)

G. Ivan Maldonado; John M. Christenson; J.P. Renier; T.F. Marcille; J. Casal

2010-03-22

The Primary objective of the proposed project is to apply and extend the latest advancements in LWR fuel management optimization to the design of advanced boiling water reactor (BWR) fuel assemblies specifically for the recycling of minor actinides (MAs).

A BWR 24-month cycle analysis using multicycle techniques

International Nuclear Information System (INIS)

Hartley, K.D.

1993-01-01

Boiling water reactor (BWR) fuel cycle design analyses have become increasingly challenging in the past several years. As utilities continue to seek improved capacity factors, reduced power generation costs, and reduced outage costs, longer cycle lengths and fuel design optimization become important considerations. Accurate multicycle analysis techniques are necessary to determine the viability of fuel designs and cycle operating strategies to meet reactor operating requirements, e.g., meet thermal and reactivity margin constraints, while minimizing overall fuel cycle costs. Siemens Power Corporation (SPC), Nuclear Division, has successfully employed multi-cycle analysis techniques with realistic rodded cycle depletions to demonstrate equilibrium fuel cycle performance in 24-month cycles. Analyses have been performed by a BWR/5 reactor, at both rated and uprated power conditions

An analysis of fuel performance cycle 20 of BWR unit 2

International Nuclear Information System (INIS)

Hemantha Rao, G.V.S.; Prasad, P.N.; Jayaraj, R.N.

2008-01-01

Nuclear Fuel Complex (NFC), an industrial unit of the Department of Atomic Energy (DAE), Government of India manufactures and supplies fuel assemblies to the two Boiling Water Reactors (BWR) at Tarapur Atomic Power Station (TAPS 1 and 2) in India which were commissioned on turnkey collaboration with GE, USA. Each fuel assembly has 36 fuel elements arranged in 6x6 square configuration. Each fuel assembly contains UO 2 pellets of different enrichments. Several improvements have been carried out over the years in the manufacture of fuel assemblies. These changes have helped in improving the fuel performance considerably. During cycle 20, the unit 2 was operating at 506/153 MWth/MWe (95.47% of rated thermal power of 530MWth) prior to shut down for refueling outage. In core sipping was completed within two days. Five leakers were identified during in core sipping. The average leaky assembly's exposure was 16,098.4 MWD/T. The minimum value of a leaky assembly's exposure was 8,591 MWD/T. Out of five assemblies, four assemblies had seen two cycles of exposure and were due for discharge. One assembly had seen single cycle. Trend of chemistry parameters for the last four cycles were within tech spec limits. Similarly trend of physics parameters for the fuel assemblies for the last cycles were also within design/tech spec limits. There were no fuel failures in the previous cycles 18 and 19. The manufacturing and QA details of the five assemblies show no deviations from the procedures and the trends are normal and within specified limits. This paper discusses the analysis of fuel failures in detail

Fuel loading and control rod patterns optimization in a BWR using tabu search

International Nuclear Information System (INIS)

Castillo, Alejandro; Ortiz, Juan Jose; Montes, Jose Luis; Perusquia, Raul

2007-01-01

This paper presents the QuinalliBT system, a new approach to solve fuel loading and control rod patterns optimization problem in a coupled way. This system involves three different optimization stages; in the first one, a seed fuel loading using the Haling principle is designed. In the second stage, the corresponding control rod pattern for the previous fuel loading is obtained. Finally, in the last stage, a new fuel loading is created, starting from the previous fuel loading and using the corresponding set of optimized control rod patterns. For each stage, a different objective function is considered. In order to obtain the decision parameters used in those functions, the CM-PRESTO 3D steady-state reactor core simulator was used. Second and third stages are repeated until an appropriate fuel loading and its control rod pattern are obtained, or a stop criterion is achieved. In all stages, the tabu search optimization technique was used. The QuinalliBT system was tested and applied to a real BWR operation cycle. It was found that the value for k eff obtained by QuinalliBT was 0.0024 Δk/k greater than that of the reference cycle

Nuclear fuel activity with minor actinides after their useful life in a BWR

International Nuclear Information System (INIS)

Martinez C, E.; Ramirez S, J. R.; Alonso V, G.

2016-09-01

Nuclear fuel used in nuclear power reactors has a life cycle, in which it provides energy, at the end of this cycle is withdrawn from the reactor core. This used fuel is known as spent nuclear fuel, a strong problem with this fuel is that when the fuel was irradiated in a nuclear reactor it leaves with an activity of approximately 1.229 x 10 15 Bq. The aim of the transmutation of actinides from spent nuclear fuel is to reduce the activity of high level waste that must be stored in geological repositories and the lifetime of high level waste; these two achievements would reduce the number of necessary repositories, as well as the duration of storage. The present work is aimed at evaluating the activity of a nuclear fuel in which radioactive actinides could be recycled to remove most of the radioactive material, first establishing a reference of actinides production in the standard nuclear fuel of uranium at end of its burning in a BWR, and a fuel rod design containing 6% of actinides in an uranium matrix from the enrichment tails is proposed, then 4 standard uranium fuel rods are replaced by 4 actinide bars to evaluate the production and transmutation of the same, finally the reduction of actinide activity in the fuel is evaluated. (Author)

Fuel loading method to exchangeable reactor core of BWR type reactor and its core

International Nuclear Information System (INIS)

Koguchi, Kazushige.

1995-01-01

In a fuel loading method for an exchangeable reactor core of a BWR type reactor, at least two kinds of fresh fuel assemblies having different reactivities between axial upper and lower portions are preliminarily prepared, and upon taking out fuel assemblies of advanced combustion and loading the fresh fuel assemblies dispersingly, they are disposed so as to attain a predetermined axial power distribution in the reactor. At least two kinds of fresh fuel assemblies have a content of burnable poisons different between the axial upper portion and lower portions. In addition, reactivity characteristics are made different at a region higher than the central boundary and a region lower than the central boundary which is set within a range of about 6/24 to 16/24 from the lower portion of the fuel effective length. There can be attained axial power distribution as desired such as easy optimization of the axial power distribution, high flexibility, and flexible flattening of the power distribution, and it requires no special change in view of the design and has a good economical property. (N.H.)

Advanced methods for BWR transient and stability analysis

Energy Technology Data Exchange (ETDEWEB)

Schmidt, A; Wehle, F; Opel, S; Velten, R [AREVA, AREVA NP, Erlangen (Germany)

2008-07-01

The design of advanced Boiling Water Reactor (BWR) fuel assemblies and cores is governed by the basic requirement of safe, reliable and flexible reactor operation with optimal fuel utilization. AREVA NP's comprehensive steady state and transient BWR methodology allows the designer to respond quickly and effectively to customer needs. AREVA NP uses S-RELAP5/RAMONA as the appropriate methodology for the representation of the entire plant. The 3D neutron kinetics and thermal-hydraulics code has been developed for the prediction of system, fuel and core behavior and provides additional margins for normal operation and transients. Of major importance is the extensive validation of the methodology. The validation is based on measurements at AREVA NP's test facilities, and comparison of the predictions with a great wealth of measured data gathered from BWR plants during many years of operation. Three of the main fields of interest are stability analysis, operational transients and reactivity initiated accidents (RIAs). The introduced 3D methodology for operational transients shows significant margin regarding the operational limit of critical power ratio, which has been approved by the German licensing authority. Regarding BWR stability a large number of measurements at different plants under various conditions have been performed and successfully post-calculated with RAMONA. This is the basis of reliable pre-calculations of the locations of regional and core-wide stability boundaries. (authors)

Advanced methods for BWR transient and stability analysis

International Nuclear Information System (INIS)

Schmidt, A.; Wehle, F.; Opel, S.; Velten, R.

2008-01-01

The design of advanced Boiling Water Reactor (BWR) fuel assemblies and cores is governed by the basic requirement of safe, reliable and flexible reactor operation with optimal fuel utilization. AREVA NP's comprehensive steady state and transient BWR methodology allows the designer to respond quickly and effectively to customer needs. AREVA NP uses S-RELAP5/RAMONA as the appropriate methodology for the representation of the entire plant. The 3D neutron kinetics and thermal-hydraulics code has been developed for the prediction of system, fuel and core behavior and provides additional margins for normal operation and transients. Of major importance is the extensive validation of the methodology. The validation is based on measurements at AREVA NP's test facilities, and comparison of the predictions with a great wealth of measured data gathered from BWR plants during many years of operation. Three of the main fields of interest are stability analysis, operational transients and reactivity initiated accidents (RIAs). The introduced 3D methodology for operational transients shows significant margin regarding the operational limit of critical power ratio, which has been approved by the German licensing authority. Regarding BWR stability a large number of measurements at different plants under various conditions have been performed and successfully post-calculated with RAMONA. This is the basis of reliable pre-calculations of the locations of regional and core-wide stability boundaries. (authors)

Parallel channel effects under BWR LOCA conditions

International Nuclear Information System (INIS)

Suzuki, H.; Hatamiya, S.; Murase, M.

1988-01-01

Due to parallel channel effects, different flow patterns such as liquid down-flow and gas up-flow appear simultaneously in fuel bundles of a BWR core during postulated LOCAs. Applying the parallel channel effects to the fuel bundle, water drain tubes with a restricted bottom end have been developed in order to mitigate counter-current flow limiting and to increase the falling water flow rate at the upper tie plate. The upper tie plate with water drain tubes is an especially effective means of increasing the safety margin of a reactor with narrow gaps between fuel rods and high steam velocity at the upper tie plate. The characteristics of the water drain tubes have been experimentally investigated using a small-scaled steam-water system simulating a BWR core. Then, their effect on the fuel cladding temperature was evaluated using the LOCA analysis program SAFER. (orig.)

Behavior and failure of fresh, hydrided and irradiated Zircaloy-4 fuel claddings under RIA conditions

International Nuclear Information System (INIS)

Le Saux, M.

2008-01-01

The purpose of this study is to characterize and simulate the mechanical behaviour and failure of fresh, hydrided and irradiated (in pressurized water reactors) cold-worked stress relieved Zircaloy-4 fuel claddings under reactivity initiated accident conditions. A model is proposed to describe the anisotropic viscoplastic mechanical behavior of the material as a function of temperature (from 20 C up to 1100 C), strain rate (from 3.10 -4 s -1 up to 5 s -1 ), fluence (from 0 up to 1026 n.m -2 ) and irradiation conditions. Axial tensile, hoop tensile, expansion due to compression and hoop plane strain tensile tests are performed at 25 C, 350 C and 480 C in order to analyse the anisotropic plastic and failure properties of the non-irradiated material hydrided up to 1200 ppm. Material strength and strain hardening depend on temperature and hydrogen in solid solution and precipitated hydride contents. Plastic anisotropy is not significantly modified by hydrogen. The material is embrittled by hydrides at room temperature. The plastic strain that leads to hydride cracking decreases with increasing hydrogen content. The material ductility, which increases with increasing temperature, is not deteriorated by hydrogen at 350 C and 480 C. Macroscopic fracture modes and damage mechanisms depend on specimen geometry, temperature and hydrogen content. A Gurson type model is finally proposed to describe both the anisotropic viscoplastic behavior and the ductile fracture of the material as a function of temperature and hydrogen content. (author) [fr

Evaluation of Thermal Creep and Hydride Re-orientation Properties of High Burnup Spent Fuel Cladding under Long Term Dry storage

Energy Technology Data Exchange (ETDEWEB)

Kamimura, K [JNES (Japan)

2012-07-01

In Japan, spent fuels will be reprocessed as recyclable energy source at a reprocessing plant. The first commercial plant is under-constructing and will start operation in 2008. It is necessary that spent fuels should be stored in the independent interim storage facilities (ISF) until reprocessing. Utilities plan the operation of the first ISF in 2010. JNES has a mission to support the safety body by researching the data of technical standard and regulation. Investigating of spent fuel integrity during long term dry storage is one of them. The objectives are: 1) Evaluation of the effects of material design changes on creep properties of high burnup spent fuel cladding; 2) Evaluation of the effects of alloy elements and texture of irradiated Zircaloy on hydride re-orientation properties and the effects of radial hydrides on cladding mechanical properties; 3) Evaluation of the effects of temperature on irradiation hardening recovery.

BWR stability analysis

International Nuclear Information System (INIS)

Valtonen, K.

1990-01-01

The objective of this study has been to examine TVO-I oscillation incident, which occured in February 22.1987 and to find out safety implications of oscillations in ATWS incidents. Calculations have been performed with RAMONA-3B and TRAB codes. RAMONA-3B is a BWR transient analysis code with three-dimencional neutron kinetics and nonequilibrium, nonhomogeneous thermal hydraulics. TRAB code is a one-dimencional BWR transient code which uses methods similar to RAMONA-3B. The results have shown that both codes are capable of analyzing of the oscillation incidents. Both out-of-phase and in-phase oscillations are possible. If the reactor scram fails (ATWS) during oscillations the severe fuel failures are always possible and the reactor core may exceed the prompt criticality

Artificial intelligence applied to fuel management in BWR type reactors; Inteligencia artificial aplicada a la administracion de combustible en reactores BWR

Energy Technology Data Exchange (ETDEWEB)

Ortiz S, J.J

1998-10-01

In this work two techniques of artificial intelligence, neural networks and genetic algorithms were applied to a practical problem of nuclear fuel management; the determination of the optimal fuel reload for a BWR type reactor. This is an important problem in the design of the operation cycle of the reactor. As a result of the application of these techniques, comparable or even better reloads proposals than those given by expert companies in the subject were obtained. Additionally, two other simpler problems in reactor physics were solved: the determination of the axial power profile and the prediction of the value of some variables of interest at the end of the operation cycle of the reactor. Neural networks and genetic algorithms have been applied to solve many problems of engineering because of their versatility but they have been rarely used in the area of fuel management. The results obtained in this thesis indicates the convenience of undertaking further work on this area and suggest the application of these techniques of artificial intelligence to the solution of other problems in nuclear reactor physics. (Author)

Liquid films and droplet deposition in a BWR fuel element

International Nuclear Information System (INIS)

Damsohn, M.

2011-01-01

In the upper part of boiling water reactors (BWR) the flow regime is dominated by a steam-water droplet flow with liquid films on the nuclear fuel rod, the so called (wispy) annular flow regime. The film thickness and liquid flow rate distribution around the fuel rod play an important role especially in regard to so called dryout, which is the main phenomenon limiting the thermal power of a fuel assembly. The deposition of droplets in the liquid film is important, because this process sustains the liquid film and delays dryout. Functional spacers with different vane shapes have been used in recent decades to enhance droplet deposition and thus create more favorable conditions for heat removal. In this thesis the behavior of liquid films and droplet deposition in the annular flow regime in BWR bundles is addressed by experiments in an adiabatic flow at nearly ambient pressure. The experimental setup consists of a vertical channel with the cross-section resembling a pair of neighboring subchannels of a fuel rod bundle. Within this double subchannel an annular flow is established with a gas-water mixture. The impact of functional spacers on the annular flow behavior is studied closely. Parameter variations comprise gas and liquid flow rates, gas density and spacer shape. The setup is instrumented with a newly developed liquid film sensor that measures the electrical conductance between electrodes flush to the wall with high temporal and spatial resolution. Advanced post-processing methods are used to investigate the dynamic behavior of liquid films and droplet deposition. The topic is also assessed numerically by means of single-phase Reynolds-Averaged-Navier-Stokes CFD simulations of the flow in the gas core. For this the commercial code STAR-CCM+ is used coupled with additional models for the liquid film distribution and droplet motion. The results of the experiments show that the liquid film is quite evenly distributed around the circumference of the fuel rods. The

Determination and microscopic study of incipient defects in irradiated power reactor fuel rods. Final report

International Nuclear Information System (INIS)

Pasupathi, V.; Perrin, J.S.; Roberts, E.

1978-05-01

This report presents the results of nondestructive and destructive examinations carried out on the Point Beach-1 (PWR) and Dresden-3 (BWR) candidate fuel rods selected for the study of pellet-clad interaction (PCI) induced incipient defects. In addition, the report includes results of examination of sections from Oskarshamn-1 (BWR) fuel rods. Eddy current examination of Point Beach-1 rods showed indications of possible incipient defects in the fuel rods. The profilometry and the gamma scan data also indicated that the source of the eddy current indications may be incipient defects. No failed rods or rods with incipient failure were found in the sample from Point Beach-1. Despite the lack of success in finding incipient defects and filed rods, the mechanism for fuel rod failures in Point Beach-1 is postulated to be PCI-related, with high startup rates and fuel handling being the key elements. Nine out of the 10 candidate fuel rods from Dresden-3 (BWR) were failed, and all the failed rods had leaked water so that the initial mechanism was observed. Examination of clad inner surfaces of the specimens from failed and unfailed rods showed fuel deposits of widely varying appearance. The deposits were found to contain uranium, cesium, and tellurium. Transmission electron microscopy of clad specimens showed evidence of microscopic strain. Metallographic examination of fuel pellets from the peak transient power location showed extensive grain boundary separation and axial movement of the fuel indicative of rapid release of fission products. Examination of Oskarshamn clad specimens did not show any stress corrosion crack (SCC) type defects. The defects found in the examinations appear to be related to secondary hydriding. The clad inner surface of the Oskarshamn specimens also showed uranium-rich deposits of varying features

BWR simulation in a stationary state for the evaluation of fuel cell design; Simulacion de un reactor BWR en estado estacionario para la evaluacion del diseno de celdas de combustible

Energy Technology Data Exchange (ETDEWEB)

Montes T, J. L.; Ortiz S, J. J.; Perusquia del C, R.; Castillo M, A., E-mail: joseluis.montes@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2014-10-15

In this paper the simulation of a BWR in order to evaluate the performance of a set of fuel assemblies under stationary state in three dimensions (3-D) is presented. 15 cases selected from a database containing a total of 18225 cases are evaluated. The main selection criteria were based on the results of the design phase of the power cells in two dimensions (2-D) and 3-D initial study. In 2-D studies the parameters that were used to qualify and select the designs were basically the local power peaking factor and neutron multiplication factor of each fuel cell. In the initial 3-D study variables that defined the quality of results, and from which the selection was realized, are the margins to thermal limits of reactor operation and the value of the effective multiplication factor at the end of cycle operation. From the 2-D and 3-D results of the studies described a second 3-D study was realized, where the optimizations of the fuel reload pattern was carried out. The results presented in this paper correspond to this second 3-D study. It was found that the designs of the fuel cell they had a similar behavior to those provided by the fuel supplier of reference BWR. Particularly it noted the impact of reload pattern on the cold shut down margin. An estimate of the operation costs of reference cycle analyzed with each one designed reload batch was also performed. As a result a positive difference (gain) up to 10,347 M/US D was found. (Author)

BUTREN-RC an hybrid system for the recharges optimization of nuclear fuels in a BWR

International Nuclear Information System (INIS)

Ortiz S, J.J.; Castillo M, J.A.; Valle G, E. del

2004-01-01

The obtained results with the hybrid system BUTREN-RC are presented that obtains recharges of nuclear fuel for a BWR type reactor. The system has implemented the methods of optimization heuristic taboo search and neural networks. The optimization it carried out with the technique of taboo search, and the neural networks, previously trained, were used to predict the behavior of the recharges of fuel, in substitution of commercial codes of reactor simulation. The obtained recharges of nuclear fuel correspond to 5 different operation cycles of the Laguna Verde Nuclear Power plant, Veracruz in Mexico. The obtained results were compared with the designs of this cycles. The energy gain with the recharges of fuel proposals is of approximately 4.5% with respect to those of design. The time of compute consumed it was considerably smaller that when a commercial code for reactor simulation is used. (Author)

Fast, quantitative, and nondestructive evaluation of hydrided LWR fuel cladding by small angle incoherent neutron scattering of hydrogen

Energy Technology Data Exchange (ETDEWEB)

Yan, Y.; Qian, S.; Littrell, K.; Parish, C.M. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Plummer, L.K. [University of Oregon, Eugene, OR 97403 (United States)

2015-05-15

A nondestructive neutron scattering method to precisely measure the uptake of hydrogen and the distribution of hydride precipitates in light water reactor (LWR) fuel cladding was developed. Zircaloy-4 cladding used in commercial LWRs was used to produce hydrided specimens. The hydriding apparatus consists of a closed stainless-steel vessel that contains Zr alloy specimens and hydrogen gas. Following hydrogen charging, the hydrogen content of the hydrided specimens was measured using the vacuum hot extraction method, by which the samples with desired hydrogen concentrations were selected for the neutron study. Optical microscopy shows that our hydriding procedure results in uniform distribution of circumferential hydrides across the wall thickness. Small angle neutron incoherent scattering was performed in the High Flux Isotope Reactor at Oak Ridge National Laboratory. Our study demonstrates that the hydrogen in commercial Zircaloy-4 cladding can be measured very accurately in minutes by this nondestructive method over a wide range of hydrogen concentrations from a very small amount (≈20 ppm) to over 1000 ppm. The hydrogen distribution in a tube sample was obtained by scaling the neutron scattering rate with a factor determined by a calibration process using standard, destructive direct chemical analysis methods on the specimens. This scale factor can be used in future tests with unknown hydrogen concentrations, thus providing a nondestructive method for determining absolute hydrogen concentrations.

Mechanical interaction between fuel pins and assemblies during LOCA in BWR

International Nuclear Information System (INIS)

Jonsson, T.

1978-10-01

The size of the rod elongation by oxidation is so large that deformation of a standard BWR fuel element with tie rods in the outer row will surely occur during a LOCA transient typical for BWRs with external pumps. Available data does not however show whether this deformation will occur early in the transient or during the cooling. Combined effects of thermal expansion of zircaloy and expansion due to oxidation and dissolution of oxygen can be expected to be large enough to cause rod bowing early in a LOCA transient. It is however not impossible that observed residual expansion of zircaloy tubes to a dominating extent are caused through expansion of zirconium oxide during cool-down. Length measurements of zircaloy tubes during a transient are desirable. (author)

Rapid hydrogen charging on metal hydride negative electrode of Fuel Cell/Battery (FCB) systems

Energy Technology Data Exchange (ETDEWEB)

Choi, Bokkyu; Lee, Sunmook; Kawai, Hiroyuki; Fushimi, Chihiro; Tsutsumi, Atsushi [Collaborative Research Center for Energy Engineering, Institute of Industrial Science, The University of Tokyo, 4-6-1 Komaba, Meguro-ku, Tokyo 153-8505 (Japan)

2009-02-15

The characteristics of rapid gaseous H{sub 2} charging/electrochemical discharging of the metal hydride negative electrode were investigated for the application in Fuel Cell/Battery (FCB) systems. They were evaluated with the H{sub 2} gas absorption, followed by the subsequent electrochemical discharging in the electrolyte solution (6M KOH). Then, the cyclability of charge-discharge was also examined. It was observed that more than 70% of the theoretical capacity was charged within 10 min with 0.3 MPa and 0.5 MPa of the initial H{sub 2} pressures. The electrochemical discharge curve showed that more than 86% of the absorbed H{sub 2} was discharged. Furthermore, the cycled charge-discharge process indicated that the H{sub 2} gas charge and electrochemical discharge process is an effective way to rapidly charge and activate the metal hydride without degeneration. (author)

Development of neural network simulating power distribution of a BWR fuel bundle

International Nuclear Information System (INIS)

Tanabe, A.; Yamamoto, T.; Shinfuku, K.; Nakamae, T.

1992-01-01

A neural network model is developed to simulate the precise nuclear physics analysis program code for quick scoping survey calculations. The relation between enrichment and local power distribution of BWR fuel bundles was learned using two layers neural network (ENET). A new model is to introduce burnable neutron absorber (Gadolinia), added to several fuel rods to decrease initial reactivity of fresh bundle. The 2nd stages three layers neural network (GNET) is added on the 1st stage network ENET. GNET studies the local distribution difference caused by Gadolinia. Using this method, it becomes possible to survey of the gradients of sigmoid functions and back propagation constants with reasonable time. Using 99 learning patterns of zero burnup, good error convergence curve is obtained after many trials. This neural network model is able to simulate no learned cases fairly as well as the learned cases. Computer time of this neural network model is about 100 times faster than a precise analysis model. (author)

BWR condensate filtration studies

International Nuclear Information System (INIS)

Wilson, J.A.; Pasricha, A.; Rekart, T.E.

1993-09-01

Poor removal of particulate corrosion products (especially iron) from condensate is one of the major problems in BWR systems. The presence of activated corrosion products creates ''hot spots'' and increases piping dose rates. Also, fuel efficiency is reduced and the risk of fuel failure is increased by the deposit of corrosion products on the fuel. Because of these concerns, current EPRI guidelines call for a maximum of 2 ppb of iron in the reactor feedwater with a level of 0.5 ppb being especially desirable. It has become clear that conventional deep bed resins are incapable of meeting these levels. While installation of prefilter systems is an option, it would be more economical for plants with naked deep beds to find an improved bead resin for use in existing systems. BWR condensate filtration technologies are being tested on a condensate side stream at Hope Creek Nuclear Generating Station. After two years of testing, hollow fiber filters (HFF) and fiber matrix filters (FMF), and low crosslink cation resin, all provide acceptable results. The results are presented for pressure drop, filtration efficiency, and water quality measurements. The costs are compared for backwashable non-precoat HFF and FMF. Results are also presented for full deep bed vessel tests of the low crosslink cation resin

Delayed hydride cracking of zirconium alloy fuel cladding

International Nuclear Information System (INIS)

2010-10-01

This report describes the work performed in a coordinated research project on Hydrogen and Hydride Degradation of the Mechanical and Physical Properties of Zirconium Alloys. It is the second in the series. In 2005-2009 that work was extended within a new CRP called Delayed Hydride Cracking in Zirconium Alloy Fuel Cladding. The project consisted of adding hydrogen to samples of Zircaloy-4 claddings representing light water reactors (LWRs), CANDU and Atucha, and measuring the rates of delayed hydride cracking (DHC) under specified conditions. The project was overseen by a supervisory group of experts in the field who provided advice and assistance to participants as required. All of the research work undertaken as part of the CRP is described in this report, which includes details of the experimental procedures that led to a consistent set of data for LWR cladding. The participants and many of their co-workers in the laboratories involved in the CRP contributed results and material used in this report, which compiles the results, their analysis, discussions of their interpretation and conclusions and recommendations for future work. The research was coordinated by an advisor and by representatives in three laboratories in industrialized Member States. Besides the basic goal to transfer the technology of the testing technique from an experienced laboratory to those unfamiliar with the methods, the CRP was set up to harmonize the experimental procedures to produce consistent sets of data, both within a single laboratory and between different laboratories. From the first part of this project it was demonstrated that by following a standard set of experimental protocols, consistent results could be obtained. Thus, experimental vagaries were minimized by careful attention to detail of microstructure, temperature history and stress state in the samples. The underlying idea for the test programme was set out at the end of the first part of the project on pressure tubes. The

Axial profiles of burned and fraction of holes for calculations of criticality with credit for BWR fuel burning; Perfiles axiales de quemado y fraccion de huecos para calculos de criticidad con credito al quemado para combustible BWR

Energy Technology Data Exchange (ETDEWEB)

Casado Sanchez, C.; Rubio Oviedo, P.

2014-07-01

This paper presents a method to define surround profiles of burning and fraction of holes suited for use in applications of credit to burning of BWR fuel from results obtained with the module STARBUCS of SCALE. (Author)

Development of neural network for analysis of local power distributions in BWR fuel bundles

International Nuclear Information System (INIS)

Tanabe, Akira; Yamamoto, Toru; Shinfuku, Kimihiro; Nakamae, Takuji.

1993-01-01

A neural network model has been developed to learn the local power distributions in a BWR fuel bundle. A two layers neural network with total 128 elements is used for this model. The neural network learns 33 cases of local power peaking factors of fuel rods with given enrichment distribution as the teacher signals, which were calculated by a fuel bundle nuclear analysis code based on precise physical models. This neural network model studied well the teacher signals within 1 % error. It is also able to calculate the local power distributions within several % error for the different enrichment distributions from the teacher signals when the average enrichment is close to 2 %. This neural network is simple and the computing speed of this model is 300 times faster than that of the precise nuclear analysis code. This model was applied to survey the enrichment distribution to meet a target local power distribution in a fuel bundle, and the enrichment distribution with flat power shape are obtained within short computing time. (author)

BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

Energy Technology Data Exchange (ETDEWEB)

Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mertyurek, Ugur [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-10-01

utilized or referenced, justification has been provided as to why the data can be utilized for BWR fuel.

Implement of MOX fuel assemblies in the design of the fuel reload for a BWR; Implemento de ensambles de combustible MOX en el diseno de la recarga de combustible para un BWR

Energy Technology Data Exchange (ETDEWEB)

Enriquez C, P.; Ramirez S, J. R.; Alonso V, G.; Palacios H, J. C., E-mail: pastor.enriquez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2011-11-15

At the present time the use of mixed oxides as nuclear fuel is a technology that has been implemented in mixed reloads of fuel for light water reactors. Due to the plutonium production in power reactors, is necessary to realize a study that presents the plutonium use like nuclear fuel. In this work a study is presented that has been carried out on the design of a fuel assembly with MOX to be proposed in the supply of a fuel reload. The fissile relationship of uranium to plutonium is presented for the design of the MOX assembly starting from plutonium recovered in the reprocessing of spent fuel and the comparison of the behavior of the infinite multiplication factor is presented and of the local power peak factor, parameters of great importance in the fuel assemblies design. The study object is a fuel assembly 10 x 10 GNF2 type for a boiling water reactor. The design of the fuel reload pattern giving fuel assemblies with MOX, so the comparison of the behavior of the stop margin for a fuel reload with UO{sub 2} and a mixed reload, implementing 12 and 16 fuel assemblies with MOX are presented. The results show that the implement of fuel assemblies with MOX in a BWR is possible, but this type of fuels creates new problems that are necessary to study with more detail. In the development of this work the calculus tools were the codes: INTREPIN-3, CASMO-4, CMSLINK and SIMULATE-3. (Author)

Hydrogen storage in the form of metal hydrides

Science.gov (United States)

Zwanziger, M. G.; Santana, C. C.; Santos, S. C.

1984-01-01

Reversible reactions between hydrogen and such materials as iron/titanium and magnesium/ nickel alloy may provide a means for storing hydrogen fuel. A demonstration model of an iron/titanium hydride storage bed is described. Hydrogen from the hydride storage bed powers a converted gasoline electric generator.

Modelling the gas transport and chemical processes related to clad oxidation and hydriding

Energy Technology Data Exchange (ETDEWEB)

Montgomery, R O; Rashid, Y R [ANATECH Research Corp., San Diego, CA (United States)

1997-08-01

Models are developed for the gas transport and chemical processes associated with the ingress of steam into a LWR fuel rod through a small defect. These models are used to determine the cladding regions in a defective fuel rod which are susceptible to massive hydriding and the creation of sunburst hydrides. The brittle nature of zirconium hydrides (ZrH{sub 2}) in these susceptible regions produces weak spots in the cladding which can act as initiation sites for cladding cracks under certain cladding stress conditions caused by fuel cladding mechanical interaction. The modeling of the axial gas transport is based on gaseous bimolar diffusion coupled with convective mass transport using the mass continuity equation. Hydrogen production is considered from steam reaction with cladding inner surface, fission products and internal components. Eventually, the production of hydrogen and its diffusion along the length results in high hydrogen concentration in locations remote from the primary defect. Under these conditions, the hydrogen can attack the cladding inner surface and breakdown the protective ZrO{sub 2} layer locally, initiating massive localized hydriding leading to sunburst hydride. The developed hydrogen evolution model is combined with a general purpose fuel behavior program to integrate the effects of power and burnup into the hydriding kinetics. Only in this manner can the behavior of a defected fuel rod be modeled to determine the conditions the result in fuel rod degradation. (author). 14 refs, 6 figs.

Development of RBWR (Resource-renewable BWR) for environmental burden reduction of radioactive wastes

International Nuclear Information System (INIS)

Hino, Tetsushi; Ohtsuka, Masaya; Moriya, Kumiaki; Matsuura, Masayoshi

2014-01-01

Accumulation of long-life transuranium elements produced as by-products with uranium fuel burning became an issue of nuclear power. Hitachi had been developing the reactor with transuranium elements burning as fuels based on BWR type reactors successfully used as commercial reactors: RBWR (Resource-renewable BWR). Efficient transmutation and fissioning of transuranium elements needed adjustment of in-core neutron energy spectra distribution better for nuclear reaction of transuranium elements. Taking advantage of characteristics of BWR type reactors with neutron spectra hardening more easily adjustable than other type of reactors, multiple recycling and fissioning transuranium elements as fuels could make environmental burden reduction of radioactive wastes and efficient use of resources compatible. This article described the concept and history of RBWR and showed its specifications and reactor core characteristics. (T. Tanaka)

Influence of hydrides orientation on strain, damage and failure of hydrided zircaloy-4

International Nuclear Information System (INIS)

Racine, A.

2005-09-01

In pressurized water reactors of nuclear power plants, fuel pellets are contained in cladding tubes, made of Zirconium alloy, for instance Zircaloy-4. During their life in the primary water of the reactor (155 bars, 300 C), cladding tubes are oxidized and consequently hydrided. A part of the hydrogen given off precipitates as Zirconium hydrides in the bulk material and embrittles the material. This embrittlement depends on many parameters, among which hydrogen content and orientation of hydrides with respect to the applied stress. This investigation is devoted to the influence of the orientation of hydrides with respect to the applied stress on strain, damage and failure mechanisms. Macroscopic and SEM in-situ ring tensile tests are performed on cladding tube material (unirradiated cold worked stress-relieved Zircaloy-4) hydrided with about 200 and 500 wppm hydrogen, and with different main hydrides orientation: either parallel or perpendicular to the circumferential tensile direction. We get the mechanical response of the material as a function of hydride orientation and hydrogen content and we investigate the deformation, damage and failure mechanisms. In both cases, digital image correlation techniques are used to estimate local and global strain distributions. Neither the tensile stress-strain response nor the global and local strain modes are significantly affected by hydrogen content or hydride orientation, but the failure modes are strongly modified. Indeed, only 200 wppm radial hydrides embrittle Zy-4: sample fail in the elastic domain at about 350 MPa before strain bands could develop; whereas in other cases sample reach at least 750 MPa before necking and final failure, in ductile or brittle mode. To model this particular heterogeneous material behavior, a non-coupled damage approach which takes into account the anisotropic distribution of the hydrides is proposed. Its parameters are identified from the macroscopic strain field measurements and a

Simplified distributed parameters BWR dynamic model for transient and stability analysis

International Nuclear Information System (INIS)

Espinosa-Paredes, Gilberto; Nunez-Carrera, Alejandro; Vazquez-Rodriguez, Alejandro

2006-01-01

This paper describes a simplified model to perform transient and linear stability analysis for a typical boiling water reactor (BWR). The simplified transient model was based in lumped and distributed parameters approximations, which includes vessel dome and the downcomer, recirculation loops, neutron process, fuel pin temperature distribution, lower and upper plenums reactor core and pressure and level controls. The stability was determined by studying the linearized versions of the equations representing the BWR system in the frequency domain. Numerical examples are used to illustrate the wide application of the simplified BWR model. We concluded that this simplified model describes properly the dynamic of a BWR and can be used for safety analysis or as a first approach in the design of an advanced BWR

Design and optimization of a fuel reload of BWR with plutonium and minor actinides

International Nuclear Information System (INIS)

Guzman A, J. R.; Francois L, J. L.; Martin del Campo M, C.; Palomera P, M. A.

2008-01-01

In this work is designed and optimized a pattern of fuel reload of a boiling water reactor (BWR), whose fuel is compound of uranium coming from the enrichment lines, plutonium and minor actinides (neptunium, americium, curium); obtained of the spent fuel recycling of reactors type BWR. This work is divided in two stages: in the first stage a reload pattern designs with and equilibrium cycle is reached, where the reload lot is invariant cycle to cycle. This reload pattern is gotten adjusting the plutonium content of the assembly for to reach the length of the wished cycle. Furthermore, it is necessary to increase the concentration of boron-10 in the control rods and to introduce gadolinium in some fuel rods of the assembly, in order to satisfy the margin approach of out. Some reactor parameters are presented: the axial profile of power average of the reactor core, and the axial and radial distribution of the fraction of holes, for the one reload pattern in balance. For the design of reload pattern codes HELIOS and CM-PRESTO are used. In the second stage an optimization technique based on genetic algorithms is used, along with certain obtained heuristic rules of the engineer experience, with the intention of optimizing the reload pattern obtained in the first stage. The objective function looks for to maximize the length of the reactor cycle, at the same time as that they are satisfied their limits related to the power and the reactor reactivity. Certain heuristic rules are applied in order to satisfy the recommendations of the fuel management: the strategy of the control cells core, the strategy of reload pattern of low leakage, and the symmetry of a quarter of nucleus. For the evaluation of the parameters that take part in the objective function it simulates the reactor using code CM-PRESTO. Using the technique of optimization of the genetic algorithms an energy of the cycle of 10834.5 MW d/tHM is obtained, which represents 5.5% of extra energy with respect to the

FY 2016 Status Report: CIRFT Testing on Spent Nuclear Fuels and Hydride Reorientation Study

Energy Technology Data Exchange (ETDEWEB)

Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Yan, Yong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Bevard, Bruce B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Scaglione, John M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division

2016-08-04

This report provides a detailed description of the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) testing conducted on spent nuclear fuel (SNF) rods in FY 2016, including hydride reorientation test results. Contact-based measurement, or three-LVDT-based curvature measurement, of SNF rods has proven to be quite reliable in CIRFT testing. However, how the linear variable differential transformer (LVDT) head contacts the SNF rod may have a significant effect on the curvature measurement, depending on the magnitude and direction of rod curvature. To correct such contact/curvature issues, sensor spacing, defined as the amount of separation between the three LVDT probes, is a critical measurement that can be used to calculate rod curvature once the deflections are obtained. Recently developed CIRFT data analyses procedures were integrated into FY 2016 CIRFT testing results for the curvature measurements. The variations in fatigue life are provided in terms of moment, equivalent stress, curvature, and equivalent strain for the tested SNFs. The equivalent stress plot collapsed the data points from all of the SNFs into a single zone. A detailed examination revealed that, at same stress level, fatigue lives display a descending order as follows: H. B. Robinson Nuclear Power Station (HBR), Limerick Nuclear Power Station (LMK), mixed uranium-plutonium oxide (MOX). If looking at the strain, then LMK fuel has a slightly longer fatigue life than HBR fuel, but the difference is subtle. The knee point of endurance limit in the curve of moment and curvature or equivalent quantities is more clearly defined for LMK and HBR fuels. The treatment affects the fatigue life of specimens. Both a drop of 12 in. and radial hydride treatment (RHT) have a negative impact on fatigue life. The effect of thermal annealing on MOX fuel rods was relatively small at higher amplitude but became significant at low amplitude of moment. Thermal annealing tended to extend the fatigue life of

Standard for assessment of fuel integrity under anticipated operational occurrences in BWR power plant:2002

International Nuclear Information System (INIS)

Mishima, Kaichiro; Suzuki, Riichiro; Komura, Seiichi; Kudo, Yoshiro; Yamanaka, Akihiro; Oomizu, Satoru; Kitamura, Hideya; Nagata, Yoshifumi

2003-01-01

To secure fuel integrity, a Light Water Reactor (LWR) core is designed so that no boiling transition (BT) should take place in fuel assemblies and excessive rise in fuel cladding temperature due to deteriorated that transfer should be avoided in Anticipated Operational Occurrences (AOO). In some AOO in a Boiling Water Reactor (BWR), however, the rise in reactor power could be limited by SCRAM or void reactivity effect. Recent studies have provided accumulated knowledge that even if BT takes place in fuel assemblies, the rise in fuel cladding temperature could be so small that it will not threat to fuel integrity, as long as the BT condition terminates within a short period of time. In addition, appropriate methods have been developed to evaluate the cladding temperature during dryout. This standard provides requirements in the assessment of fuel integrity under AOO in which limited-BT condition is temporarily reached and the propriety of reusing a fuel assembly that has experienced limited-BT condition. The standard has been approved by the Atomic Energy Society of Japan following deliberation by impartial members for two and half years. It is now expected that this standard will provide an effective measure for the rational expansion of fuel design and operational margin. (author)

Optimization of fuel reloads for a BWR using the ant colony system; Optimizacion de recargas de combustible para un BWR usando el sistema de colonia de hormigas

Energy Technology Data Exchange (ETDEWEB)

Esquivel E, J. [Universidad Autonoma del Estado de Mexico, Facultad de Ingenieria, Cerro de Coatepec s/n, Ciudad Universitaria, 50110 Toluca, Estado de Mexico (Mexico); Ortiz S, J. J. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)], e-mail: jaime.es.jaime@gmail.com

2009-10-15

In this work some results obtained during the development of optimization systems are presented, which are employees for the fuel reload design in a BWR. The systems use the ant colony optimization technique. As first instance, a system is developed that was adapted at travel salesman problem applied for the 32 state capitals of Mexican Republic. The purpose of this implementation is that a similarity exists with the design of fuel reload, since the two problems are of combinatorial optimization with decision variables that have similarity between both. The system was coupled to simulator SIMULATE-3, obtaining good results when being applied to an operation cycle in equilibrium for reactors of nuclear power plant of Laguna Verde. (Author)

Prediction of the local power factor in BWR fuel cells by means of a multilayer neural network

International Nuclear Information System (INIS)

Montes, J.L.; Ortiz, J.J.; Perusquia C, R.; Francois, J.L.; Martin del Campo M, C.

2007-01-01

To the beginning of a new operation cycle in a BWR reactor the reactivity of this it increases by means of the introduction of fresh fuel, the one denominated reload fuel. The problem of the definition of the characteristics of this reload fuel represents a combinatory optimization problem that requires significantly a great quantity of CPU time for their determination. This situation has motivated to study the possibility to substitute the Helios code, the one which is used to generate the new cells of the reload fuel parameters, by an artificial neuronal network, with the purpose of predicting the parameters of the fuel reload cell of a BWR reactor. In this work the results of the one training of a multilayer neuronal net that can predict the local power factor (LPPF) in such fuel cells are presented. The prediction of the LPPF is carried out in those condition of beginning of the life of the cell (0.0 MWD/T, to 40% of holes in the one moderator, temperature of 793 K in the fuel and a moderator temperature of 560 K. The cells considered in the present study consist of an arrangement of 10x10 bars, of those which 92 contains U 235 , some of these bars also contain a concentration of Gd 2 O 3 and 8 of them contain only water. The axial location inside the one assembles of recharge of these cells it is exactly up of the cells that contain natural uranium in the base of the reactor core. The training of the neuronal net is carried out by means of a retro-propagation algorithm that uses a space of training formed starting from previous evaluations of cells by means of the Helios code. They are also presented the results of the application of the neuronal net found for the prediction of the LPPF of some cells used in the real operation of the Unit One of the Laguna Verde Nuclear Power station. (Author)

Seismic evaluation of BWR spent fuel storage racks using actual damping by vibration test in water

International Nuclear Information System (INIS)

Yamasaki, Hiroto; Iwakura, Shigeyoshi; Imaoka, Tetsuo; Okumura, Kazue; Orita, Syuichi; Namita, Yoshio

2010-01-01

Damping value for BWR spent fuel storage racks has been used 1 percent damping, which is applied to welded steel structures in air as defined JEAG4601. However, it is considered that the actual damping is higher than that of the above mentioned, because of its underwater installation. This report shows the actual damping value of the Check Arrayed Rack by vibration test in water and Evaluation by the analysis of rack using actual damping. (author)

Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

International Nuclear Information System (INIS)

Vaibhaw, Kumar; Rao, S.V.R.; Jha, S.K.; Saibaba, N.; Jayaraj, R.N.

2008-01-01

Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (∼300 deg. C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation (F n ) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process

Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

Science.gov (United States)

Vaibhaw, Kumar; Rao, S. V. R.; Jha, S. K.; Saibaba, N.; Jayaraj, R. N.

2008-12-01

Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (˜300 °C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation ( F n) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process.

Siemens Nuclear Power Corporation methods development for BWR/PWR reactor licensing

International Nuclear Information System (INIS)

Pruitt, D.W.

1992-01-01

This presentation addresses the Siemens Nuclear Power Corporation (SNP) perspective on the primary forces driving methods development in the nuclear industry. These forces are fuel design, computational environment and industry requirement evolution. The first segment of the discussion presents the SNP experience base. SNP develops, manufactures and licenses both BWR and PWR reload fuel. A review of this experience base highlights the accelerating rate at which new fuel designs are being introduced into the nuclear industry. The application of advanced BWR lattice geometries provides an example of fuel design trends. The second aspect of the presentation is the rapid evolution of the computing environment. The final subject in the presentation is the impact of industry requirements on code or methods development

The effect of stress state on zirconium hydride reorientation

Science.gov (United States)

Cinbiz, Mahmut Nedim

Prior to storage in a dry-cask facility, spent nuclear fuel must undergo a vacuum drying cycle during which the spent fuel rods are heated up to elevated temperatures of ≤ 400°C to remove moisture the canisters within the cask. As temperature increases during heating, some of the hydride particles within the cladding dissolve while the internal gas pressure in fuel rods increases generating multi-axial hoop and axial stresses in the closed-end thin-walled cladding tubes. As cool-down starts, the hydrogen in solid solution precipitates as hydride platelets, and if the multiaxial stresses are sufficiently large, the precipitating hydrides reorient from their initial circumferential orientation to radial orientation. Radial hydrides can severely embrittle the spent nuclear fuel cladding at low temperature in response to hoop stress loading. Because the cladding can experience a range of stress states during the thermo-mechanical treatment induced during vacuum drying, this study has investigated the effect of stress state on the process of hydride reorientation during controlled thermo-mechanical treatments utilizing the combination of in situ X-ray diffraction and novel mechanical testing analyzed by the combination of metallography and finite element analysis. The study used cold worked and stress relieved Zircaloy-4 sheet containing approx. 180 wt. ppm hydrogen as its material basis. The failure behavior of this material containing radial hydrides was also studied over a range of temperatures. Finally, samples from reactor-irradiated cladding tubes were examined by X-ray diffraction using synchrotron radiation. To reveal the stress state effect on hydride reorientation, the critical threshold stress to reorient hydrides was determined by designing novel mechanical test samples which produce a range of stress states from uniaxial to "near-equibiaxial" tension when a load is applied. The threshold stress was determined after thermo-mechanical treatments by

Development and Assessment of CFD Models Including a Supplemental Program Code for Analyzing Buoyancy-Driven Flows Through BWR Fuel Assemblies in SFP Complete LOCA Scenarios

Science.gov (United States)

Artnak, Edward Joseph, III

This work seeks to illustrate the potential benefits afforded by implementing aspects of fluid dynamics, especially the latest computational fluid dynamics (CFD) modeling approach, through numerical experimentation and the traditional discipline of physical experimentation to improve the calibration of the severe reactor accident analysis code, MELCOR, in one of several spent fuel pool (SFP) complete loss-ofcoolant accident (LOCA) scenarios. While the scope of experimental work performed by Sandia National Laboratories (SNL) extends well beyond that which is reasonably addressed by our allotted resources and computational time in accordance with initial project allocations to complete the report, these simulated case trials produced a significant array of supplementary high-fidelity solutions and hydraulic flow-field data in support of SNL research objectives. Results contained herein show FLUENT CFD model representations of a 9x9 BWR fuel assembly in conditions corresponding to a complete loss-of-coolant accident scenario. In addition to the CFD model developments, a MATLAB based controlvolume model was constructed to independently assess the 9x9 BWR fuel assembly under similar accident scenarios. The data produced from this work show that FLUENT CFD models are capable of resolving complex flow fields within a BWR fuel assembly in the realm of buoyancy-induced mass flow rates and that characteristic hydraulic parameters from such CFD simulations (or physical experiments) are reasonably employed in corresponding constitutive correlations for developing simplified numerical models of comparable solution accuracy.

Hydride effect on crack instability of Zircaloy cladding

Energy Technology Data Exchange (ETDEWEB)

Tseng, Che-Chung, E-mail: cctseng@iner.gov.tw [Institute of Nuclear Energy Research, No. 1000, Wunhua Road, Jiaan Village, Lungtan, Township, Taoyuan County 32546, Taiwan (China); Sun, Ming-Hung [Institute of Nuclear Energy Research, No. 1000, Wunhua Road, Jiaan Village, Lungtan, Township, Taoyuan County 32546, Taiwan (China); Chao, Ching-Kong [Department of Mechanical Engineering, National Taiwan University of Science and Technology, 43 Keelung Road, Section 4, Taipei 106, Taiwan (China)

2014-04-01

Highlights: • Radial hydrides near the crack tip had a significant effect on crack propagation. • For radial hydrides off the crack line vertically, the effect on crack propagation was notably reduced. • The longer hydride platelet resulted in a remarkable effect on crack propagation. • A long split in the radial hydride precipitate would enhance crack propagation. • The presence of circumferential hydride among radial hydrides may play an important role in crack propagation. - Abstract: A methodology was proposed to investigate the effect of hydride on the crack propagation in fuel cladding. The analysis was modeled based on an outside-in crack with radial hydrides located near its crack tip. The finite element method was used in the calculation; both stress intensity factor K{sub I} and J integral were applied to evaluate the crack stability. The parameters employed in the analysis included the location of radial hydride, hydride dimensions, number of hydrides, and the presence of circumferential hydride, etc. According to our study, the effective distance between a radial hydride and the assumed cladding surface crack for the enhancement of crack propagation proved to be no greater than 0.06 mm. For a hydride not on the crack line, it would induce a relatively minor effect on crack propagation if the vertical distance was beyond 0.05 mm. However, a longer hydride precipitate as well as double radial hydrides could have a remarkable effect on crack propagation. A combined effect of radial and circumferential hydrides was also discussed.

Influence of uranium hydride oxidation on uranium metal behaviour

International Nuclear Information System (INIS)

Patel, N.; Hambley, D.; Clarke, S.A.; Simpson, K.

2013-01-01

This work addresses concerns that the rapid, exothermic oxidation of active uranium hydride in air could stimulate an exothermic reaction (burning) involving any adjacent uranium metal, so as to increase the potential hazard arising from a hydride reaction. The effect of the thermal reaction of active uranium hydride, especially in contact with uranium metal, does not increase in proportion with hydride mass, particularly when considering large quantities of hydride. Whether uranium metal continues to burn in the long term is a function of the uranium metal and its surroundings. The source of the initial heat input to the uranium, if sufficient to cause ignition, is not important. Sustained burning of uranium requires the rate of heat generation to be sufficient to offset the total rate of heat loss so as to maintain an elevated temperature. For dense uranium, this is very difficult to achieve in naturally occurring circumstances. Areas of the uranium surface can lose heat but not generate heat. Heat can be lost by conduction, through contact with other materials, and by convection and radiation, e.g. from areas where the uranium surface is covered with a layer of oxidised material, such as burned-out hydride or from fuel cladding. These rates of heat loss are highly significant in relation to the rate of heat generation by sustained oxidation of uranium in air. Finite volume modelling has been used to examine the behaviour of a magnesium-clad uranium metal fuel element within a bottle surrounded by other un-bottled fuel elements. In the event that the bottle is breached, suddenly, in air, it can be concluded that the bulk uranium metal oxidation reaction will not reach a self-sustaining level and the mass of uranium oxidised will likely to be small in relation to mass of uranium hydride oxidised. (authors)

Influence of uranium hydride oxidation on uranium metal behaviour

Energy Technology Data Exchange (ETDEWEB)

Patel, N.; Hambley, D. [National Nuclear Laboratory (United Kingdom); Clarke, S.A. [Sellafield Ltd (United Kingdom); Simpson, K.

2013-07-01

This work addresses concerns that the rapid, exothermic oxidation of active uranium hydride in air could stimulate an exothermic reaction (burning) involving any adjacent uranium metal, so as to increase the potential hazard arising from a hydride reaction. The effect of the thermal reaction of active uranium hydride, especially in contact with uranium metal, does not increase in proportion with hydride mass, particularly when considering large quantities of hydride. Whether uranium metal continues to burn in the long term is a function of the uranium metal and its surroundings. The source of the initial heat input to the uranium, if sufficient to cause ignition, is not important. Sustained burning of uranium requires the rate of heat generation to be sufficient to offset the total rate of heat loss so as to maintain an elevated temperature. For dense uranium, this is very difficult to achieve in naturally occurring circumstances. Areas of the uranium surface can lose heat but not generate heat. Heat can be lost by conduction, through contact with other materials, and by convection and radiation, e.g. from areas where the uranium surface is covered with a layer of oxidised material, such as burned-out hydride or from fuel cladding. These rates of heat loss are highly significant in relation to the rate of heat generation by sustained oxidation of uranium in air. Finite volume modelling has been used to examine the behaviour of a magnesium-clad uranium metal fuel element within a bottle surrounded by other un-bottled fuel elements. In the event that the bottle is breached, suddenly, in air, it can be concluded that the bulk uranium metal oxidation reaction will not reach a self-sustaining level and the mass of uranium oxidised will likely to be small in relation to mass of uranium hydride oxidised. (authors)

Optimization of fuel cells for BWR based in Tabu modified search

International Nuclear Information System (INIS)

Martin del Campo M, C.; Francois L, J.L.; Palomera P, M.A.

2004-01-01

The advances in the development of a computational system for the design and optimization of cells for assemble of fuel of Boiling Water Reactors (BWR) are presented. The method of optimization is based on the technique of Tabu Search (Tabu Search, TS) implemented in progressive stages designed to accelerate the search and to reduce the time used in the process of optimization. It was programed an algorithm to create the first solution. Also for to diversify the generation of random numbers, required by the technical TS, it was used the Makoto Matsumoto function obtaining excellent results. The objective function has been coded in such a way that can adapt to optimize different parameters like they can be the enrichment average or the peak factor of radial power. The neutronic evaluation of the cells is carried out in a fine way by means of the HELIOS simulator. In the work the main characteristics of the system are described and an application example is presented to the design of a cell of 10x10 bars of fuel with 10 different enrichment compositions and gadolinium content. (Author)

Evaluation of the radial design of fuel cells in an operation cycle of a BWR reactor; Evaluacion del diseno radial de celdas de combustible en un ciclo de operacion de un reactor BWR

Energy Technology Data Exchange (ETDEWEB)

Gonzalez C, J.; Martin del Campo M, C. [Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Facultad de Ingenieria, UNAM, Paseo Cuauhnahuac 8532, Jiutepec, Morelos (Mexico)]. e-mail: jgco@ver.megared.net.mx

2003-07-01

This work is continuation of one previous in the one that the application of the optimization technique called Tabu search to the radial design of fuel cells of boiling water reactors (BWR, Boiling Water Reactor) is presented. The objective function used in the optimization process only include neutron parameters (k-infinite and peak of radial power) considering the cell at infinite media. It was obtained to reduce the cell average enrichment completing the characteristics of reactivity of an original cell. The objective of the present work is to validate the objective function that was used for the radial design of the fuel cell (test cell), analyzing the operation of a one cycle of the reactor in which fuels have been fresh recharged that contain an axial area with the nuclear database of the cell designed instead of the original cell. For it is simulated it with Cm-Presto the cycle 10 of the reactor operation of the Unit 1 of the Nuclear Power station of Laguna Verde (U1-CNLV). For the cycle evaluation its were applied so much the simulation with the Haling strategy, as the simulation of the one cycle with control rod patterns and they were evaluated the energy generation and several power limits and reactivity that are used as design parameters in fuel reloads of BWR reactors. The results at level of an operation cycle of the reactor, show that the objective function used in the optimization and radial design of the cell is adequate and that it can induce to one good use of the fuel. (Author)

Optimization of fuel cells for BWR using Path Re linking and flexible strategies of solution

International Nuclear Information System (INIS)

Castillo M, J. A.; Ortiz S, J. J.; Torres V, M.; Perusquia del Cueto, R.

2009-10-01

In this work are presented the obtained preliminary results to design nuclear fuel cells for boiling water reactors (BWR) using new strategies. To carry out the cells design some of the used rules in the fuel administration were discarded and other were implemented. The above-mentioned with the idea of making a comparative analysis between the used rules and those implemented here, under the hypothesis that it can be possible to design nuclear fuel cells without using all the used rules and executing the security restrictions that are imposed in these cases. To evaluate the quality of the obtained cells it was taken into account the power pick factor and the infinite multiplication factor, in the same sense, to evaluate the proposed configurations and to obtain the mentioned parameters was used the CASMO-4 code. To optimize the design it is uses the combinatorial optimization technique named Path Re linking and the Dispersed Search as local search method. The preliminary results show that it is possible to implement new strategies for the cells design of nuclear fuel following new rules. (Author)

The BWR Hybrid 4 control rod

International Nuclear Information System (INIS)

Gross, H.; Fuchs, H.P.; Lippert, H.J.; Dambietz, W.

1988-01-01

The service life of BWR control rods designed in the past has been unsatisfactory. The main reason was irradiation assisted stress corrosion cracking of B 4 C rods caused by external swelling of the B 4 C powder. By this reason KWU developed an improved BWR control rod (Hybrid 4 control rod) with extended service life and increased control rod worth. It also allows the procedure for replacing and rearranging fuel assemblies to be considerably simplified. A complete set of Hydbrid 4 control rods is expected to last throughout the service life of a plant (assumption: ca. 40 years) if an appropriate control rod reshuffling management program is used. (orig.)

Review of consequences of uranium hydride formation in N-Reactor fuel elements stored in the K-Basins

Energy Technology Data Exchange (ETDEWEB)

Weber, J.W.

1994-09-28

The 105-K Basins on the Hanford site are used to store uranium fuel elements and assemblies irradiated in and discharged from N Reactor. The storage cylinders in KW Basin are known to have some broken N reactor fuel elements in which the exposed uranium is slowly reacting chemically with water in the cylinder. The products of these reactions are uranium oxide, hydrogen, and potentially some uranium hydride. The purpose of this report is to document the results f the latest review of potential, but highly unlikely accidents postulated to occur as closed cylinders containing N reactor fuel assemblies are opened under water in the KW basin and as a fuel assembly is raised from the basin in a shipping cask for transportation to the 327 Building for examination as part of the SNF Characterization Program. The postulated accidents reviews in this report are considered to bound all potential releases of radioactivity and hydrogen. These postulated accidents are: (1) opening and refill of a cylinder containing significant amounts of hydrogen and uranium hydride; and (2) draining of the single element can be used to keep the fuel element submerged in water after the cask containing the can and element is lifted from the KW Basin. Analysis shows the release of radioactivity to the site boundary is significantly less than that allowed by the K Basin Safety Evaluation. Analysis further shows there would be no damage to the K Basin structure nor would there be injury to personnel for credible events.

Understanding hydride formation in Zr-1Nb alloy through microstructural characterization

International Nuclear Information System (INIS)

Neogy, S.; Srivastava, D.; Tewari, R.; Singh, R.N.; Dey, G.K.; De, P.K.; Banerjee, S.

2003-07-01

In this study the experimental results of hydride formation and their microstructure evolution in Zr-1Nb alloy is presented. This Zr-1Nb binary alloy and other Zr-1 Nb based ternary and quaternary alloys are being used as fuel tube materials and have the potential for meeting the requirement of high burn up fuel. Hydriding of Zr-1Nb alloy having a microstructure comprising equiaxed α grains and a uniform distribution of spherical particles of the β phase has been carried out in this study. The specimens have been hydrided by gaseous charging method to different hydrogen levels. The microstructures of hydrided samples were examined as a function of hydrogen content. The formation of δ hydride in slow cooled specimens and formation of γ hydride in rapidly cooled specimens has been studied with their morphology, habit plane and orientation relationship with the α matrix in view. The habit planes of either type of hydride phase has been determined and compared with those observed in other Zr-Nb alloys. The orientation relationship between the α matrix and the δ hydride was found to be the following: (0001) α // (111) δ and [1120] α // [110] δ . The orientation relationship between the α matrix and the γ hydride was of the following type: (0001) α // (111) γ and [1120] α // [110] γ . The internal structure of both types of hydride has been examined. The effect of the presence of the spherical β phase particles in the a matrix on the growth of the hydride plates has been investigated. (author)

SCORPIO-BWR: status and future plans

International Nuclear Information System (INIS)

Porsmyr, Jan; Bodal, Terje; Beere, William H.

2004-01-01

Full text: During the years from 2000 to 2003 a joint project has been performed by IFE, Halden and TEPCO Systems Corporation, Japan, to develop a core monitoring system for BWRs based on the their existing core monitoring system TiARA and the SCORPIO framework. It has been emphasised to develop a reliable, flexible, adaptable and user-friendly system, which is easy to maintain. Therefore, a rather general framework (SCORPIO Framework) has been used which facilitates easy software modifications as well as adding/ replacing physics modules. The software modules is integrated in the SCORPIO framework using the Software Bus as the communication tool and with the Picasso UIMS tool for MMI. The SCORPIO-BWR version is developed on a Windows-PC platform. The SCORPIO-BWR version provides all functions, which are necessary for all analyses and operations performed on a BWR plant and comprises functions for on-line core monitoring, predictive analysis and core management with interfaces to plant instrumentation and physics codes. Functions for system initialisation and maintenance are also included. A SCORPIO-BWR version adapted for ABWR was installed in TEPSYS facilities in Tokyo in January 2003, where the final acceptance tests were carried out and accepted. The ABWR version of the system is now in the verification and validation phase. In the period from April 2003 until March 2004 a project for realizing an offline-version of SCORPIO-BWR system, which supports the offline tasks of BWR in-core fuel management for ABWR and BWR-5 type of reactors, was developed. The offline-version of the SCORPIO-BWR system for ABWR and BWR-5 type of reactors was installed at TEPSYS in March 2003, where the final acceptance tests were carried out and accepted. Plans for the next version of this system is to study the possibility of adapting SCORPIO-BWR to work with 'mobile technology'. This means that it should be possible to access and display information from the SCORPIO-BWR system on a

SCORPIO-BWR: status and future plans

Energy Technology Data Exchange (ETDEWEB)

Porsmyr, Jan; Bodal, Terje; Beere, William H. (and others)

2004-07-01

Full text: During the years from 2000 to 2003 a joint project has been performed by IFE, Halden and TEPCO Systems Corporation, Japan, to develop a core monitoring system for BWRs based on the their existing core monitoring system TiARA and the SCORPIO framework. It has been emphasised to develop a reliable, flexible, adaptable and user-friendly system, which is easy to maintain. Therefore, a rather general framework (SCORPIO Framework) has been used which facilitates easy software modifications as well as adding/ replacing physics modules. The software modules is integrated in the SCORPIO framework using the Software Bus as the communication tool and with the Picasso UIMS tool for MMI. The SCORPIO-BWR version is developed on a Windows-PC platform. The SCORPIO-BWR version provides all functions, which are necessary for all analyses and operations performed on a BWR plant and comprises functions for on-line core monitoring, predictive analysis and core management with interfaces to plant instrumentation and physics codes. Functions for system initialisation and maintenance are also included. A SCORPIO-BWR version adapted for ABWR was installed in TEPSYS facilities in Tokyo in January 2003, where the final acceptance tests were carried out and accepted. The ABWR version of the system is now in the verification and validation phase. In the period from April 2003 until March 2004 a project for realizing an offline-version of SCORPIO-BWR system, which supports the offline tasks of BWR in-core fuel management for ABWR and BWR-5 type of reactors, was developed. The offline-version of the SCORPIO-BWR system for ABWR and BWR-5 type of reactors was installed at TEPSYS in March 2003, where the final acceptance tests were carried out and accepted. Plans for the next version of this system is to study the possibility of adapting SCORPIO-BWR to work with 'mobile technology'. This means that it should be possible to access and display information from the SCORPIO-BWR

Sodium-based hydrides for thermal energy applications

Science.gov (United States)

Sheppard, D. A.; Humphries, T. D.; Buckley, C. E.

2016-04-01

Concentrating solar-thermal power (CSP) with thermal energy storage (TES) represents an attractive alternative to conventional fossil fuels for base-load power generation. Sodium alanate (NaAlH4) is a well-known sodium-based complex metal hydride but, more recently, high-temperature sodium-based complex metal hydrides have been considered for TES. This review considers the current state of the art for NaH, NaMgH3- x F x , Na-based transition metal hydrides, NaBH4 and Na3AlH6 for TES and heat pumping applications. These metal hydrides have a number of advantages over other classes of heat storage materials such as high thermal energy storage capacity, low volume, relatively low cost and a wide range of operating temperatures (100 °C to more than 650 °C). Potential safety issues associated with the use of high-temperature sodium-based hydrides are also addressed.

Fundamental experiments on hydride reorientation in zircaloy

Science.gov (United States)

Colas, Kimberly B.

reoriented hydride fraction and connectivity increase with number of cycles which could lead to more dangerous microstructure for storage of spent fuel. Pre-existing cracks were also found to affect hydride connectivity and morphology which directly impacts DHC and fuel integrity. (Abstract shortened by UMI.).

BWR Assembly Optimization for Minor Actinide Recycling

International Nuclear Information System (INIS)

Maldonado, G. Ivan; Christenson, John M.; Renier, J.P.; Marcille, T.F.; Casal, J.

2010-01-01

The Primary objective of the proposed project is to apply and extend the latest advancements in LWR fuel management optimization to the design of advanced boiling water reactor (BWR) fuel assemblies specifically for the recycling of minor actinides (MAs). A top-level objective of the Advanced Fuel Cycle Systems Analysis program element of the DOE NERI program is to investigate spent fuel treatment and recycling options for current light water reactors (LWRs). Accordingly, this project targets to expand the traditional scope of nuclear fuel management optimization into the following two complementary specific objectives: (1) To develop a direct coupling between the pin-by-pin within-bundle loading control variables and core-wide (bundle-by-bundle) optimization objectives, (2) to extend the methodology developed to explicitly encompass control variables, objectives, and constraints designed to maximize minor actinide incineration in BWR bundles and cycles. The first specific objective is projected to 'uncover' dormant thermal margin made available by employing additional degrees of freedom within the optimization process, while the addition of minor actinides is expected to 'consume' some of the uncovered thermal margin. Therefore, a key underlying goal of this project is to effectively invest some of the uncovered thermal margin into achieving the primary objective.

Determination of hydrogen in zirconium hydride and uranium-zirconium hydride by inert gas exraction-gravimetric method

International Nuclear Information System (INIS)

Hoshino, Akira; Iso, Shuichi

1976-01-01

An inert gas extraction-gravimetric method has been applied to the determination of hydrogen in zirconium hydride and uranium-zirconium hydride which are used as neutron moderator and fuel of nuclear safety research reactor (NSRR), respectively. The sample in a graphite-enclosed quartz crucible is heated inductively to 1200 0 C for 20 min in a helium stream. Hydrogen liberated from the sample is oxidized to water by copper(I) oxide-copper(II) oxide at 400 0 C, and the water is determined gravimetrically by absorption in anhydrone. The extraction curves of hydrogen for zirconium hydride and uranium-zirconium hydride samples are shown in Figs. 2 and 3. Hydrogen in the samples is extracted quantitatively by heating at (1000 -- 1250) 0 C for (10 -- 40) min. Recoveries of hydrogen in the case of zirconium hydride were examined as follows: a weighed zirconium rod (5 phi x 6 mm, hydrogen -5 Torr. After the chamber was filled with purified hydrogen to 200 Torr, the rod was heated to 400 0 C for 15 h, and again weighed to determine the increase in weight. Hydrogen in the rod was then determined by the proposed method. The results are in excellent agreement with the increase in weight as shown in Table 1. Analytical results of hydrogen in zirconium hydride samples and an uranium-zirconium hydride sample are shown in Table 2. (auth.)

Experimental study of defect power reactor fuel. Final report

International Nuclear Information System (INIS)

Forsyth, R.S.; Jonsson, T.

1982-01-01

Two BWR fuel rods, one intact and one defect, with the same manufacturing and irradiation data have been examined in a comparative study. The defect rod has been irradiated in a defect condition during approximately one reactor cycle and has consequently some secondary defects. The defect rod has two penetrating defects at a distance of about 1.5 meters from each other. Comparison with the intact rod shows a large Cs loss from the defect rod, especially between the cladding defects, where the loss is measured to about 30 %. The leachibility in deionized water is higher for Cs, U and Cm for fuel from the defect rod. The leaching results are more complex for Sr-90, Pu and Am. The fuel in the defect rod has undergone a change of structure with gain growth and formation of oriented fuel structure. The cladding of the defect rod is hydrided locally in some parts of the lower part of the rod and furthermore over a more extended region near the end of the rod. (Authors)

Compact modular BWR (CM-BWR)

International Nuclear Information System (INIS)

Fennern, Larry; Boardman, Charles; Carroll, Douglas G.; Hida, Takahiko

2003-01-01

A preliminary assessment has shown that a small 350 MWe BWR reactor can be placed within a close fitting steel containment vessel that is 7.1 meters inside diameter. This allows the technology and manufacturing capability currently used to fabricate large ABWR reactor vessels to be used to provide a factory fabricated containment vessel for a 350 MWe BWR. When a close fitted steel containment is combined with a passive closed loop isolation condenser system and a natural circulating reactor system that contains a large water inventory, primary system leaks cannot uncover the core. This eliminates many of the safety systems needed in response to a LOCA that are common to large, conventional plant designs including. Emergency Core Flooding, Automatic Depressurization System, Active Residual Heat Removal, Safety Related Auxiliary Cooling, Safety Related Diesel Generators, Hydrogen Re-Combiners, Ex-vessel Core Retention and Cooling. By fabricating the containment in a factory and eliminating most of the conventional safety systems, the construction schedule is shortened and the capital cost reduced to levels that would not otherwise be possible for a relatively small modular BWR. This makes the CM-BWR a candidate for applications where smaller incremental power additions are desired relative to a large ALWR or where the local infrastructure is not able to accommodate a conventional ALWR plant rated at 1350 MWe or more. This paper presents a preliminary design description of a Compact Modular BWR (CM-BWR) whose design features dramatically reduce the size and cost of the reactor building and associated safety systems. (author)

Effects of δ-hydride precipitation at a crack tip on crack propagation in delayed hydride cracking of Zircaloy-2

Energy Technology Data Exchange (ETDEWEB)

Kubo, T., E-mail: kubo@nfd.co.jp [Nippon Nuclear Fuel Development Co., Ltd., 2163 Narita-cho, Oarai-machi, Ibaraki 311-1313 (Japan); Kobayashi, Y. [M.O.X. Co., Ltd., 1828-520 Hirasu-cho, Mito, Ibaraki 311-0853 (Japan)

2013-08-15

Highlights: • Steady state crack velocity of delayed hydride cracking in Zircaloy-2 was analyzed. • A large stress peak is induced at an end of hydride by volume expansion of hydride. • Hydrogen diffuses to the stress peak, thereby accelerating steady hydride growth. • Crack velocity was estimated from the calculated hydrogen flux into the stress peak. • There was good agreement between calculation results and experimental data. -- Abstract: Delayed hydride cracking (DHC) of Zircaloy-2 is one possible mechanism for the failure of boiling water reactor fuel rods in ramp tests at high burnup. Analyses were made for hydrogen diffusion around a crack tip to estimate the crack velocity of DHC in zirconium alloys, placing importance on effects of precipitation of δ-hydride. The stress distribution around the crack tip is significantly altered by precipitation of hydride, which was strictly analyzed using a finite element computer code. Then, stress-driven hydrogen diffusion under the altered stress distribution was analyzed by a differential method. Overlapping of external stress and hydride precipitation at a crack tip induces two stress peaks; one at a crack tip and the other at the front end of the hydride precipitate. Since the latter is larger than the former, more hydrogen diffuses to the front end of the hydride precipitate, thereby accelerating hydride growth compared with that in the absence of the hydride. These results indicated that, after hydride was formed in front of the crack tip, it grew almost steadily accompanying the interaction of hydrogen diffusion, hydride growth and the stress alteration by hydride precipitation. Finally, crack velocity was estimated from the calculated hydrogen flux into the crack tip as a function of temperature, stress intensity factor and material strength. There was qualitatively good agreement between calculation results and experimental data.

Leak-before-break assessment of RBMK-1500 fuel channel in case of delayed hydride cracking

International Nuclear Information System (INIS)

Klimasauskas, A.; Grybenas, A.; Makarevicius, V.; Nedzinskas, L.; Levinskas, R.; Kiselev, V.

2003-01-01

One of the factors determining remaining lifetime of Zr-2.5% Nb fuel channel (FC) is the amount of hydrogen dissolved during corrosion process. When the concentration of hydrogen exceeds the terminal solid solubility limit zirconium hydrides are precipitated. As a result form necessary conditions for delayed hydride cracking (DHC). Data from the RBMK-1500 fuel channel tubes (removed from service) shows that hydrogen in some cases distributes unevenly and hydrogen concentration can differ several times between individual FC tubes or separate zones of the same tube and possibly, can reach dangerous levels in the future. Consequently, lacking statistical research data, it is difficult to forecast increase of hydrogen concentration and formation of DHC. So it is important to verify if under the most unfavorable situation leak before break condition will be satisfied in the case of DHC. To estimate possible DHC rates in RBMK 1500 FC pressure tubes experiments were done in the following order: hydriding of the Zr-2.5Nb pressure tube material to the required hydrogen concentration; hydrogen analysis; machining of specimens, fatigue crack formation in the axial direction, DHC testing; average crack length measurement and DHC velocity calculation. During the tests in average DHC values were determined at 283, 250 and 144 degC (with hydrogen concentrations correspondingly 76, 54 and 27 ppm). The fracture resistance dependence from hydrogen concentration was measured at 20 degC. To calculate leak through the postulated flaw, statistical distribution of DHC surface irregularity was determined. Leak before break analysis was carried out according to requirements of RBMK 1500 regulatory documents. J integral and crack opening were calculated using finite element method. Loading of the FC was determined using RELAP5 code. Critical crack length was calculated using R6 and J-integral methods. Coolant flow rate through the postulated crack was estimated using SQUIRT software

Automatic determination of BWR fuel loading patterns based on K.E. technique with core physics simulation

International Nuclear Information System (INIS)

Ikehara, T.; Tsuiki, M.; Takeshita, T.

1990-01-01

On the basis oof a computerized search method, a prototype for a fuel loading pattern expert system has been developed to support designers in core design for BWRs. The method was implemented by coupling rules and core physics simulators into an inference engine to establish an automated generate-and-test cycle. A search control mechanism, which prunes paths to be searched and selects appropriate rules through the interaction with the user, was also introduced to accomplish an effective search. The constraints in BWR core design are: (1) cycle length more than L, (2) core shutdown margin more than S, and (3) thermal margin more than T. Here L, S, and T are the specified minimum values. In this system, individual rules contain the manipulation to improve the core shutdown margin explicitly. Other items were taken into account only implicitly. Several applications to the test cases were carried out. It was found that the results were comparable with those obtained by human expert engineers. Broad applicability of the present method in the BWR core design domain was proved

Nanostructured, complex hydride systems for hydrogen generation

Directory of Open Access Journals (Sweden)

Robert A. Varin

2015-02-01

Full Text Available Complex hydride systems for hydrogen (H2 generation for supplying fuel cells are being reviewed. In the first group, the hydride systems that are capable of generating H2 through a mechanical dehydrogenation phenomenon at the ambient temperature are discussed. There are few quite diverse systems in this group such as lithium alanate (LiAlH4 with the following additives: nanoiron (n-Fe, lithium amide (LiNH2 (a hydride/hydride system and manganese chloride MnCl2 (a hydride/halide system. Another hydride/hydride system consists of lithium amide (LiNH2 and magnesium hydride (MgH2, and finally, there is a LiBH4-FeCl2 (hydride/halide system. These hydride systems are capable of releasing from ~4 to 7 wt.% H2 at the ambient temperature during a reasonably short duration of ball milling. The second group encompasses systems that generate H2 at slightly elevated temperature (up to 100 °C. In this group lithium alanate (LiAlH4 ball milled with the nano-Fe and nano-TiN/TiC/ZrC additives is a prominent system that can relatively quickly generate up to 7 wt.% H2 at 100 °C. The other hydride is manganese borohydride (Mn(BH42 obtained by mechano-chemical activation synthesis (MCAS. In a ball milled (2LiBH4 + MnCl2 nanocomposite, Mn(BH42 co-existing with LiCl can desorb ~4.5 wt.% H2 at 100 °C within a reasonable duration of dehydrogenation. Practical application aspects of hydride systems for H2 generation/storage are also briefly discussed.

Design and axial optimization of nuclear fuel for BWR reactors

International Nuclear Information System (INIS)

Garcia V, M.A.

2006-01-01

In the present thesis, the modifications made to the axial optimization system based on Tabu Search (BT) for the axial design of BWR fuel type are presented, developed previously in the Nuclear Engineering Group of the UNAM Engineering Faculty. With the modifications what is mainly looked is to consider the particular characteristics of the mechanical design of the GE12 fuel type, used at the moment in the Laguna Verde Nucleo electric Central (CNLV) and that it considers the fuel bars of partial longitude. The information obtained in this thesis will allow to plan nuclear fuel reloads with the best conditions to operate in a certain cycle guaranteeing a better yield and use in the fuel burnt, additionally people in charge in the reload planning will be favored with the changes carried out to the system for the design and axial optimization of nuclear fuel, which facilitate their handling and it reduces their execution time. This thesis this developed in five chapters that are understood in the following way in general: Chapter 1: It approaches the basic concepts of the nuclear energy, it describes the physical and chemical composition of the atoms as well as that of the uranium isotopes, the handling of the uranium isotope by means of the nuclear fission until arriving to the operation of the nuclear reactors. Chapter 2: The nuclear fuel cycle is described, the methods for its extraction, its conversion and its enrichment to arrive to the stages of the nuclear fuel management used in the reactors are described. Beginning by the radial design, the axial design and the core design of the nuclear reactor related with the fuel assemblies design. Chapter 3: the optimization methods of nuclear fuel previously used are exposed among those that are: the genetic algorithms method, the search methods based on heuristic rules and the application of the tabu search method, which was used for the development of this thesis. Chapter 4: In this part the used methodology to the

Modeling of the water gap in BWR fuel elements using SCALE/TRITON; Modellierung des Wasserspalts bei SWR-BE mit SCALE/TRITON

Energy Technology Data Exchange (ETDEWEB)

Tittelbach, S.; Chernykh, M. [WTI Wissenschaftlich-Technische Ingenieurberatung GmbH, Juelich (Germany)

2012-11-01

The authors show that an adequate modeling of the water gap in BWR fuel element models using the code TRITON requires an explicit consideration of the Dancoff factors. The analysis of three modeling options reveals that considering the moderating effects of the water gap coolant for the peripheral fuel elements the resulting deviations of the U-235 and Pu-239 concentrations are significantly reduced. The increased temporal calculation efforts are justified with respect to the burnup credits for criticality safety analyses.

A study on the criticality search of transuranium recycling BWR core by adjusting supplied fuel composition in equilibrium state

International Nuclear Information System (INIS)

Seino, Takeshi; Sekimoto, Hiroshi

1998-01-01

There have been some difficulties in carrying out an extensive evaluation of the equilibrium state of Light Water Reactor (LWR) recycling operations keeping their fixed criticality condition using conventional design codes because of the complexity of their calculation model for practical fuel and core design and because of a large amount of calculation time. This study presents an efficient approach to secure the criticality in an equilibrium cycle by adjusting a supplied fuel composition. The criticality search is performed by the use of fuel importance obtained from the equation adjoint to a continuously fuel supplied core burnup equation. Using this method, some numerical analyses were carried out in order to evaluate the mixed oxide (MOX) fuel composition of equilibrium Boiling Water Reactor (BWR) cores satisfying the criticality requirement. The results showed the comprehensive and quantitative characteristics on the equilibrium cores confining transuraniums for different MOX fuel loading fractions and irradiating conditions

BWR fuel performance under advanced water chemistry conditions – a delicate journey towards zero fuel failures – a review

International Nuclear Information System (INIS)

Hettiarachchi, S.

2015-01-01

Boiling Water Reactors (BWRs) have undergone a variety of chemistry evolutions over the past few decades as a result of the need to control stress corrosion cracking of reactor internals, radiation fields and personnel exposure. Some of the advanced chemistry changes include hydrogen addition, zinc addition, iron reduction using better filtration technologies, and more recently noble metal chemical addition to many of the modern day operating BWRs. These water chemistry evolutions have resulted in changes in the crud distribution on fuel cladding material, Co-60 levels and the Rod oxide thickness (ROXI) measurements using the conventional eddy current techniques. A limited number of Post-Irradiation Examinations (PIE) of fuel rods that exhibited elevated oxide thickness using eddy current techniques showed that the actual oxide thickness by metallography is much lower. The difference in these observations is attributed to the changing magnetic properties of the crud affecting the rod oxide thickness measurement by the eddy current technique. This paper will review and summarize the BWR fuel cladding performance under these advanced and improved water chemistry conditions and how these changes have affected the goal to reach zero fuel failures. The paper will also provide a brief summary of some of the results of hot cell PIE, results of crud composition evaluation, crud spallation, oxide thickness measurements, hydrogen content in the cladding and some fuel failure observations. (author) Key Words: Boiling Water Reactor, Fuel Performance, Hydrogen Addition, Zinc Addition, Noble Metal Chemical Addition, Zero Leakers

HYDRIDE-RELATED DEGRADATION OF SNF CLADDING UNDER REPOSITORY CONDITIONS

International Nuclear Information System (INIS)

McCoy, K.

2000-01-01

The purpose and scope of this analysis/model report is to analyze the degradation of commercial spent nuclear fuel (CSNF) cladding under repository conditions by the hydride-related metallurgical processes, such as delayed hydride cracking (DHC), hydride reorientation and hydrogen embrittlement, thereby providing a better understanding of the degradation process and clarifying which aspects of the process are known and which need further evaluation and investigation. The intended use is as an input to a more general analysis of cladding degradation

LAPUR5 BWR stability analysis in Kuosheng nuclear power plant

International Nuclear Information System (INIS)

Kunlung Wu; Chunkuan Shih; Wang, J.R.; Kao, L.S.

2005-01-01

Full text of publication follows: Unstable oscillation of a nuclear power reactor core is one of the main reasons that causes minor core damage. Stability analysis needs to be performed to predict the potential problem as early as possible and to prevent core instability events from happening. Nuclear Regulatory Commission (NRC) requests all BWR licensees to examine each core reload and to impose operating limitations, as appropriate, to ensure compliance with GDC 10 and 12. GDC 10 requires that the reactor core be designed with appropriate margin to assure that specified acceptable fuel design limits will not be exceeded during any condition of normal operation, including anticipated operational occurrences. GDC 12 requires assurance that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are either not possible or can be reliably and readily detected and suppressed. Therefore, the core instability is directly related to the fuel design limits. The core and channel DR (decay ratio) calculation are commonly performed to determine system's stability when new fuel designs are introduced in the core. In order to establish the independent analysis technology for BWR licensees and verifications, the Institute of Nuclear Energy Research (INER) has obtained agreement from NRC and implemented the 'Methodology and Procedure for Calculation of Core and Channel Decay Ratios with LAPUR', which was developed by the IBERINCO in 2001. LAPUR5 uses a multi-nodal description of the neutron dynamics, together with a distributed parameter model of the core thermal hydrodynamics to produce a space-dependent representation of the dynamics of a BWR in the frequency domain for small perturbations around a steady state condition. From the output of LAPUR5, the following results are obtained: global core decay ratio, out-of phase core decay ratio, and channel decay ratio. They are key parameters in the determination of BWR core stability

Anisotropic Azimuthal Power and Temperature distribution on FuelRod. Impact on Hydride Distribution

Energy Technology Data Exchange (ETDEWEB)

Motta, Arthur [Pennsylvania State Univ., State College, PA (United States); Ivanov, Kostadin [Pennsylvania State Univ., State College, PA (United States); Arramova, Maria [Pennsylvania State Univ., State College, PA (United States); Hales, Jason [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-04-29

The degradation of the zirconium cladding may limit nuclear fuel performance. In the high temperature environment of a reactor, the zirconium in the cladding corrodes, releasing hydrogen in the process. Some of this hydrogen is absorbed by the cladding in a highly inhomogeneous manner. The distribution of the absorbed hydrogen is extremely sensitive to temperature and stress concentration gradients. The absorbed hydrogen tends to concentrate near lower temperatures. This hydrogen absorption and hydride formation can cause cladding failure. This project set out to improve the hydrogen distribution prediction capabilities of the BISON fuel performance code. The project was split into two primary sections, first was the use of a high fidelity multi-physics coupling to accurately predict temperature gradients as a function of r, θ , and z, and the second was to use experimental data to create an analytical hydrogen precipitation model. The Penn State version of thermal hydraulics code COBRA-TF (CTF) was successfully coupled to the DeCART neutronics code. This coupled system was verified by testing and validated by comparison to FRAPCON data. The hydrogen diffusion and precipitation experiments successfully calculated the heat of transport and precipitation rate constant values to be used within the hydrogen model in BISON. These values can only be determined experimentally. These values were successfully implemented in precipitation, diffusion and dissolution kernels that were implemented in the BISON code. The coupled output was fed into BISON models and the hydrogen and hydride distributions behaved as expected. Simulations were conducted in the radial, axial and azimuthal directions to showcase the full capabilities of the hydrogen model.

Power oscillations in BWR reactors

International Nuclear Information System (INIS)

Espinosa P, G.

2002-01-01

One of the main problems in the operation of BWR type reactors is the instability in power that these could present. One type of oscillations and that is the objective of this work is the named density wave, which is attributed to the thermohydraulic processes that take place in the reactor core. From the beginnings of the development of BWR reactors, the stability of these has been an important aspect in their design, due to its possible consequences on the fuel integrity. The reactor core operates in two phase flow conditions and it is observed that under certain power and flow conditions, power instabilities appear. Studying this type of phenomena is complex, due to that a reactor core is constituted approximately by 27,000 fuel bars with different distributions of power and flow. The phenomena that cause the instability in BWR reactors continue being matter of scientific study. In the literature mainly in nuclear subject, it can be observed that exist different methods and approximations for studying this type of phenomena, nevertheless, their results are focused to establish safety limits in the reactor operation, instead of studying in depth of the knowledge about. Also in this line sense of the reactor data analysis, the oscillations characteristic frequencies are obtained for trying to establish if the power is growing or decreasing. In addition to that before mentioned in this paper it is presented a rigorous study applying the volumetric average method, for obtaining the vacuum waves propagation velocities and its possible connection with the power oscillations. (Author)

Radial optimization of a BWR fuel cell using genetic algorithms

International Nuclear Information System (INIS)

Martin del Campo M, C.; Carmona H, R.; Oropeza C, I.P.

2006-01-01

The development of the application of the Genetic Algorithms (GA) to the optimization of the radial distribution of enrichment in a cell of fuel of a BWR (Boiling Water Reactor) is presented. The optimization process it was ties to the HELIOS simulator, which is a transport code of neutron simulation of fuel cells that has been validated for the calculation of nuclear banks for BWRs. With heterogeneous radial designs can improve the radial distribution of the power, for what the radial design of fuel has a strong influence in the global design of fuel recharges. The optimum radial distribution of fuel bars is looked for with different enrichments of U 235 and contents of consumable poison. For it is necessary to define the representation of the solution, the objective function and the implementation of the specific optimization process to the solution of the problem. The optimization process it was coded in 'C' language, it was automated the creation of the entrances to the simulator, the execution of the simulator and the extraction, in the exit of the simulator, of the parameters that intervene in the objective function. The objective function includes four parameters: average enrichment of the cell, average gadolinia concentration of the cell, peak factor of radial power and k-infinite multiplication factor. To be able to calculate the parameters that intervene in the objective function, the one evaluation process of GA was ties to the HELIOS code executed in a Compaq Alpha workstation. It was applied to the design of a fuel cell of 10 x 10 that it can be employee in the fuel assemble designs that are used at the moment in the Laguna Verde Nucleo electric Central. Its were considered 10 different fuel compositions which four contain gadolinia. Three heuristic rules that consist in prohibiting the placement of bars with gadolinia in the ends of the cell, to place the compositions with the smallest enrichment in the corners of the cell and to fix the placement of

Fuel safety research 2001

Energy Technology Data Exchange (ETDEWEB)

Uetsuka, Hiroshi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2002-11-01

The Fuel Safety Research Laboratory is in charge of research activity which covers almost research items related to fuel safety of water reactor in JAERI. Various types of experimental and analytical researches are being conducted by using some unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and the Reactor Fuel Examination Facility (RFEF) of JAERI. The research to confirm the safety of high burn-up fuel and MOX fuel under accident conditions is the most important item among them. The laboratory consists of following five research groups corresponding to each research fields; Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). Research group of fuel behavior analysis (FEMAXI group). Research group of radionuclides release and transport behavior from irradiated fuel under severe accident conditions (VEGA group). The research conducted in the year 2001 produced many important data and information. They are, for example, the fuel behavior data under BWR power oscillation conditions in the NSRR, the data on failure-bearing capability of hydrided cladding under LOCA conditions and the FP release data at very high temperature in steam which simulate the reactor core condition during severe accidents. This report summarizes the outline of research activities and major outcomes of the research executed in 2001 in the Fuel Safety Research Laboratory. (author)

Microstructure and orientation of hydrides in zircaloy-4 tubes for CAREM 25 fuel rods

International Nuclear Information System (INIS)

Juarez, G; Flores, A; Bianchi, D; Vizcaino, P

2012-01-01

The aim of the present work was giving support to the development of Zircaloy-4 fuel claddings for the CAREM 25 reactor through microstructural and mechanical properties studies along the manufacturing process. The manufacturing route was defined in 4 cold rolling stages and two thermal treatments, one at the middle and one after the last rolling stage. The first two rolling stages were performed in FAESA and the last two in PPFAECNEA using the rolling machine HPTR 8-15. The reference values for the evaluation were those indicated in the technical specification CAREM25 F ET-3-B0610. In this context, four tubes were received from FAESA. To these tubes microstructural and hydride orientation studies were performed to characterize the material in each step performed in PPFAE. The material received from FAESA has a grain size of 6 um, which fulfills the specification requirements for the final product (<11.2 um), so that, the final rolling stages (3rd and 4th) refine the grain until 3 um finished in the microstructure. Hydride morphology evolved from a random orientation in the recrystallized material or radial in the stress relieved material to the circumferential orientation requested in the specification. The final thermal treatment did not modify this morphology. The PD. (Marshall parameter) produced a value of 48, indicating that only 4% of the hydrides are oriented between 48 o and 90 o , that is, far from the circumferential direction (author)

Reproduction in laboratory of the morphology distribution and orientation of hydrides in different stages fuel cycle; Reproduccion en laboratorio de la morfologia, distribucion y orientacion de hidruros en distintas etapas del ciclo de combustible

Energy Technology Data Exchange (ETDEWEB)

Martin-Rengel, M. A.; Gomez, F. J.; Ruiz-Hervias, J.

2013-07-01

In this paper, the experimental techniques employed to reproduce in the laboratory the distribution, morphology and orientation of the hydrides during the different steps of the nuclear fuel cycle are reported. A cathodic charging technique was employed to produce ZIRLO cladding samples with an homogeneous distribution of hydrides and concentrations of 150, 250, 500, 1200 and 2000 ppm of hydrogen. The treatments developed to produce radial hydride reorientation, hydride blisters and a peripheral rim of hydrides are described.

Development of transmutation technologies of radioactive waste by actinoid hydride

International Nuclear Information System (INIS)

Konashi, Kenji; Matsui, Hideki; Yamawaki, Michio

2001-01-01

Two waste treatment methods, geological disposal and transmutation, have been studied. The transmutation method changes long-lived radioactive nuclides to short-lived one or stabilizes them by nuclear transformation. The transmutation by actinoid hydride is exactly alike that transformation method from actinoid disposal waste to Pu fuel. For this object, OMEGA project is processing now. The transmutation is difficult by two causes such as large amount of long-lived radioactive nuclides and not enough development of control technologies of nuclear reaction except atomic reactor. The transmutation using actinoid hydride has merits that the amount of actinoid charged in the target increases and the effect of thermal neutrons on fuel decreases depending on homogeneous transmutation velocity in the target. Development of stable actinoid hydride under the conditions of reactor temperature and irradiation environment is important. The experimental results of U-ZrH 1.6 are shown in this paper. The irradiation experiment using Th hydride has been proceeding. (S.Y.)

Non-Fourier Vernotte-Cattaneo numerical model for heat conduction in a BWR fuel rod

Energy Technology Data Exchange (ETDEWEB)

Espinosa-Martinez, E.G.; Vazquez-Rodriguez, A.; Varela-Ham, J.R.; Espinosa-Paredes, G., E-mail: gepe@xanum.uam.mx [Universidad Autonoma Metropolitana, Area de Ingenieria en Recursos Energeticos, Iztapalapa (Mexico)

2014-07-01

A fuel rod mathematical model based on transient heat conduction as constitutive Non-Fourier law for Light Water Reactors (LWRs) transient analysis is presented. The structure of the fuel pellet is affected due to high temperatures and irradiation, which eventually produce fracture or cracks. In principle the fractures are saturated of gas. Then, the Fourier law of the heat conduction is not strictly applicable to describe these phenomena, where the physical properties such as thermal conductivity, heat capacity and density correspond to a heterogeneous material due to gas, and therefore the thermal diffusion process due to molecular transport in the fuel pellet is affected. From the point of view of nuclear reactor safety analysis, the heat transfer from the fuel to the coolant is crucial and superheating of the wall can cause the cladding failure. In the classical theory of diffusion, the Fourier law of heat conduction is used to describe the relation between the heat flux vector and the temperature gradient assuming that the heat propagation speeds are infinite. The Non-Fourier approach presented in this work eliminates the assumption of an infinite thermal wave speed, therefore time-dependent heat sources were considered in the fuel rod heat transfer model. The numerical experiments in a BWR, show that the Non-Fourier approach is crucial in the pressurization transients such as turbine trip and reactor isolation. (author)

Non-Fourier Vernotte-Cattaneo numerical model for heat conduction in a BWR fuel rod

International Nuclear Information System (INIS)

Espinosa-Martinez, E.G.; Vazquez-Rodriguez, A.; Varela-Ham, J.R.; Espinosa-Paredes, G.

2014-01-01

A fuel rod mathematical model based on transient heat conduction as constitutive Non-Fourier law for Light Water Reactors (LWRs) transient analysis is presented. The structure of the fuel pellet is affected due to high temperatures and irradiation, which eventually produce fracture or cracks. In principle the fractures are saturated of gas. Then, the Fourier law of the heat conduction is not strictly applicable to describe these phenomena, where the physical properties such as thermal conductivity, heat capacity and density correspond to a heterogeneous material due to gas, and therefore the thermal diffusion process due to molecular transport in the fuel pellet is affected. From the point of view of nuclear reactor safety analysis, the heat transfer from the fuel to the coolant is crucial and superheating of the wall can cause the cladding failure. In the classical theory of diffusion, the Fourier law of heat conduction is used to describe the relation between the heat flux vector and the temperature gradient assuming that the heat propagation speeds are infinite. The Non-Fourier approach presented in this work eliminates the assumption of an infinite thermal wave speed, therefore time-dependent heat sources were considered in the fuel rod heat transfer model. The numerical experiments in a BWR, show that the Non-Fourier approach is crucial in the pressurization transients such as turbine trip and reactor isolation. (author)

Fuel safety research 2000

Energy Technology Data Exchange (ETDEWEB)

Uetsuka, Hiroshi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-03-01

In April 1999, the Fuel Safety Research Laboratory was newly established as a part of reorganization of the Nuclear Safety Research Center, JAERI. The new laboratory was organized by combining three pre-existing laboratories, Reactivity Accident Laboratory, Fuel Reliability Laboratory, and a part of Severe Accident Research Laboratory. The Fuel Safety Research Laboratory becomes to be in charge of all fuel safety research in JAERI. Various experimental and analytical researches are conducted in the laboratory by using the unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and hot cells in JAERI. The laboratory consists of following five research groups corresponding to each research fields; (a) Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). (b) Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). (c) Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). (d) Research group of fuel behavior analysis (FEMAXI group). (e) Research group of FP release/transport behavior from irradiated fuel (VEGA group). The research activities in year 2000 produced many important data and information. They are, for example, failure of high burnup BWR fuel rod under RIA conditions, data on the behavior of hydrided Zircaloy cladding under LOCA conditions and FP release data from VEGA experiments at very high temperature/pressure condition. This report summarizes the outline of research activities and major outcomes of the research executed in 2000 in the Fuel Safety Research Laboratory. (author)

Control chart analysis of data regarding 0.2% yield strength (YS) and percent total circumferential elongation (%TCE) for zircaloy clad tubes for PHWR and BWR fuels

International Nuclear Information System (INIS)

Yadav, M.B.; Singh, Hari; Vaidyanathan, S.; Sood, D.D.; Raghavan, S.V.; Bandyopadhyay, A.K.; Kulkarni, P.G.

1992-01-01

Zircaloy cladding tubes for PHWR and BWR fuels are manufactured and tested at Nuclear Fuel Complex (NFC), Hyderabad. Atomic Fuels Division is carrying out the quality assurance of the fuels on behalf of Nuclear Power Corporation (NPC). In this paper an attempt has been made to assess whether the quality of the clad tubes has met the requirements specified for the two mechanical properties of the tubes namely 0.2% yield strength and percent total circumferential elongation using control chart technique. For this purpose data for about 100 lots in each case were used. Process means and process standard deviations for these properties and the control limits for the corresponding control charts were estimated. The main findings are: (i) In case of PHWR tubes the production quality level with respect to 0.2% YS is higher, while that in case of %TCE is lower causing rejection of lots. On the other hand in the case of BWR tubes the production quality levels with respect to both the properties are higher than the required one. (ii) With respect to 0.2% YS, in case of BWR tubes a change in the pattern of distribution is detected beyond the lot serial no.47. However in case of PHWR tubes, though the data falls into two groups, no such pattern is seen. A modification in the acceptance/rejection criterion of the lot has been suggested. It is also pointed out that to have a correct picture of the total variation it is necessary to study the within tube variation. (author). 4 figs, 2 tabs

Novel modular natural circulation BWR design and safety evaluation

International Nuclear Information System (INIS)

Ishii, Mamoru; Shi, Shanbin; Yang, Won Sik; Wu, Zeyun; Rassame, Somboon; Liu, Yang

2015-01-01

Highlights: • Introduction of BWR-type natural circulation small modular reactor preliminary design (NMR-50). • Design of long fuel cycle length for the NMR-50. • Design of double passive safety systems for the NMR-50. • RELAP5 analyses of design basis accidents for the NMR-50. - Abstract: The Purdue NMR (Novel Modular Reactor) represents a BWR-type small modular reactor with a significantly reduced reactor pressure vessel (RPV) height. Specifically, it has one third the height of a conventional BWR RPV with an electrical output of 50 MWe. The preliminary design of the NMR-50 including reactor, fuel cycle, and safety systems is described and discussed. The improved neutronics design of the NMR-50 extends the fuel cycle length up to 10 years. The NMR-50 is designed with double passive engineering safety system, which is intended to withstand a prolonged station black out with loss of ultimate heat sink accident such as experienced at Fukushima. In order to evaluate the safety features of the NMR-50, two representative design basis accidents, i.e. main steam line break (MSLB) and bottom drain line break (BDLB), are simulated by using the best-estimate thermal–hydraulic code RELAP5. The RPV water inventory, containment pressure, and the performance of engineering safety systems are investigated for about 33 h after the initiation of the accidents

A study on the criticality search of transuranium recycling BWR core by adjusting supplied fuel composition in equilibrium state

International Nuclear Information System (INIS)

Seino, Takeshi; Sekimoto, Hiroshi

1997-01-01

There have been some difficulties in carrying out an extensive evaluation of the equilibrium state of Light Water Reactor (LWR) recycling operations keeping their fixed criticality condition using conventional design codes, because of the complexity of their calculational model for practical fuel and core design and because of a large amount of calculation time. This study presents an efficient approach to secure the criticality in an equilibrium cycle by adjusting a supplied fuel composition. The criticality search is performed by the use of fuel importance obtained from the equilibrium adjoint to a continuously fuel supplied core burnup equation. Using this method, some numerical analyses were carried out in order to evaluate the mixed oxide (MOX) fuel composition of equilibrium Boiling Water Reactor (BWR) cores satisfying the criticality requirement. The results showed the comprehensive and quantitative characteristics on the equilibrium cores confining transuranium for different MOX fuel loading fractions and irradiating conditions. (author)

BUTREN-RC an hybrid system for the recharges optimization of nuclear fuels in a BWR; BUTREN-RC un sistema hibrido para la optimizacion de recargas de combustible nuclear en un BWR

Energy Technology Data Exchange (ETDEWEB)

Ortiz S, J.J.; Castillo M, J.A. [ININ, Carretera Mexico-Toluca Km. 36.5, 52045 Estado de Mexico (Mexico); Valle G, E. del [IPN, ESFM, 07738 Mexico D.F. (Mexico)

2004-07-01

The obtained results with the hybrid system BUTREN-RC are presented that obtains recharges of nuclear fuel for a BWR type reactor. The system has implemented the methods of optimization heuristic taboo search and neural networks. The optimization it carried out with the technique of taboo search, and the neural networks, previously trained, were used to predict the behavior of the recharges of fuel, in substitution of commercial codes of reactor simulation. The obtained recharges of nuclear fuel correspond to 5 different operation cycles of the Laguna Verde Nuclear Power plant, Veracruz in Mexico. The obtained results were compared with the designs of this cycles. The energy gain with the recharges of fuel proposals is of approximately 4.5% with respect to those of design. The time of compute consumed it was considerably smaller that when a commercial code for reactor simulation is used. (Author)

Powder production of U-Mo alloy, HMD process (Hydriding- Milling- Dehydriding)

Energy Technology Data Exchange (ETDEWEB)

Pasqualini, E. E.; Garcia, J.H.; Lopez, M.; Cabanillas, E.; Adelfang, P. [Dept. Combustibles Nucleares. Comision Nacional de Energia Atomica, Av. Gral. Paz 1499, 1650 Buenos Aires (Argentina)

2002-07-01

Uranium-molybdenum (U-Mo) alloys can be hydrided massively in metastable {gamma} (gamma) phase. The brittle hydride can be milled and dehydrided to acquire the desired size distributions needed for dispersion nuclear fuels. The developments of the different steps of this process called hydriding-milling- dehydriding (HMD Process) are described. Powder production scales for industrial fabrication is easily achieved with conventional equipment, small man-power and low investment. (author)

Powder production of U-Mo alloy, HMD process (Hydriding- Milling- Dehydriding)

International Nuclear Information System (INIS)

Pasqualini, E. E.; Garcia, J.H.; Lopez, M.; Cabanillas, E.; Adelfang, P.

2002-01-01

Uranium-molybdenum (U-Mo) alloys can be hydrided massively in metastable γ (gamma) phase. The brittle hydride can be milled and dehydrided to acquire the desired size distributions needed for dispersion nuclear fuels. The developments of the different steps of this process called hydriding-milling- dehydriding (HMD Process) are described. Powder production scales for industrial fabrication is easily achieved with conventional equipment, small man-power and low investment. (author)

Nuclear fuel activity with minor actinides after their useful life in a BWR; Actividad del combustible nuclear con actinidos menores despues de su vida util en un reactor BWR

Energy Technology Data Exchange (ETDEWEB)

Martinez C, E.; Ramirez S, J. R.; Alonso V, G., E-mail: eduardo.martinez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2016-09-15

Nuclear fuel used in nuclear power reactors has a life cycle, in which it provides energy, at the end of this cycle is withdrawn from the reactor core. This used fuel is known as spent nuclear fuel, a strong problem with this fuel is that when the fuel was irradiated in a nuclear reactor it leaves with an activity of approximately 1.229 x 10{sup 15} Bq. The aim of the transmutation of actinides from spent nuclear fuel is to reduce the activity of high level waste that must be stored in geological repositories and the lifetime of high level waste; these two achievements would reduce the number of necessary repositories, as well as the duration of storage. The present work is aimed at evaluating the activity of a nuclear fuel in which radioactive actinides could be recycled to remove most of the radioactive material, first establishing a reference of actinides production in the standard nuclear fuel of uranium at end of its burning in a BWR, and a fuel rod design containing 6% of actinides in an uranium matrix from the enrichment tails is proposed, then 4 standard uranium fuel rods are replaced by 4 actinide bars to evaluate the production and transmutation of the same, finally the reduction of actinide activity in the fuel is evaluated. (Author)

Development of a detailed BWR core thermal-hydraulic analysis method based on the Japanese post-BT standard using a best-estimate code

International Nuclear Information System (INIS)

Ono, H.; Mototani, A.; Kawamura, S.; Abe, N.; Takeuchi, Y.

2004-01-01

The post-BT standard is a new fuel integrity standard or the Atomic Energy Society of Japan that allows temporary boiling transition condition in the evaluation for BWR anticipated operational occurrences. For application of the post-BT standard to BWR anticipated operational occurrences evaluation, it is important to identify which fuel assemblies and which axial, radial positions of fuel rods have temporarily experienced the post-BT condition and to evaluates how high the fuel cladding temperature rise was and how long the dryout duration continued. Therefore, whole bundle simulation, in which each fuel assembly is simulated independently by one thermal-hydraulic component, is considered to be an effective analytical method. In the present study, a best-estimate thermal-hydraulic code, TRACG02, has been modified to extend it predictive capability by implementing the post-BT evaluation model such as the post-BT heat transfer correlation and rewetting correlation and enlarging the number of components used for BWR plant simulation. Based on new evaluation methods, BWR core thermal-hydraulic behavior has been analyzed for typical anticipated operational occurrence conditions. The location where boiling transition occurs and the severity of fuel assembly in the case of boiling transition conditions such as fuel cladding temperature, which are important factors in determining whether the reuse of the fuel assembly can be permitted, were well predicted by the proposed evaluation method. In summary, a new evaluation method for a detailed BWR core thermal-hydraulic analysis based on the post-BT standard of the Atomic Energy Society of Japan has been developed and applied to the evaluation of the post-BT standard during the actual BWR plant anticipated operational occurrences. (author)

Advances in BWR water chemistry

International Nuclear Information System (INIS)

Garcia, Susan E.; Giannelli, Joseph F.; Jarvis, Mary L.

2012-09-01

This paper reviews recent advances in Boiling Water Reactor (BWR) water chemistry control with examples of plant experiences at U.S. designed BWRs. Water chemistry advances provide some of the most effective methods for mitigating materials degradation, reducing fuel performance concerns and lowering radiation fields. Mitigation of stress corrosion cracking (SCC) of materials remains a high priority and improved techniques that have been demonstrated in BWRs will be reviewed, specifically hydrogen injection combined with noble metal chemical addition (NMCA) and the newer on-line noble metal application process (OLNC). Hydrogen injection performance, an important part of SCC mitigation, will also be reviewed for the BWR fleet, highlighting system improvements that have enabled earlier injection of hydrogen including the potential for hydrogen injection during plant startup. Water chemistry has been significantly improved by the application of pre-filtration and optimized use of ion exchange resins in the CP (condensate polishing) and reactor water cleanup (RWCU) systems. EPRI has monitored and supported water treatment improvements to meet water chemistry goals as outlined in the EPRI BWR Water Chemistry Guidelines, particularly those for SCC mitigation of reactor internals and piping, minimization of fuel risk due to corrosion and crud deposits and chemistry control for radiation field reduction. In recent years, a significant reduction has occurred in feedwater corrosion product input, particularly iron. A large percentage of plants are now reporting <0.1 ppb feedwater iron. The impacts to plant operation and chemistry of lower feedwater iron will be explored. Depleted zinc addition is widely practiced across the fleet and the enhanced focus on radiation reduction continues to emphasize the importance of controlling radiation source term. In addition, shutdown chemistry control is necessary to avoid excessive release of activated corrosion products from fuel

Development of underwater high-definition camera for the confirmation test of core configuration and visual examination of BWR fuel

International Nuclear Information System (INIS)

Watanabe, Masato; Tuji, Kenji; Ito, Keisuke

2010-01-01

The purpose of this study is to develop underwater High-Definition camera for the confirmation test of core configuration and visual examination of BWR fuels in order to reduce the time of these tests and total cost regarding to purchase and maintenance. The prototype model of the camera was developed and examined in real use condition in spent fuel pool at HAMAOKA-2 and 4. The examination showed that the ability of prototype model was either equaling or surpassing to conventional product expect for resistance to radiation. The camera supposes to be used in the dose rate condition of under about 10 Gy/h. (author)

Optimization of analysis best-estimate of a fuel element BWR with Code STAR-CCM+; Optimizacion del analisis best-estimate de un elemento combustible BWR con el codigo STAR-CCM+

Energy Technology Data Exchange (ETDEWEB)

Morgado Canada, E.; Concejal Barmejo, A.; Jimenez Varas, G.; Solar Martinez, A.

2014-07-01

The objective of the project is the evaluation of the code STAR-CCM +, as well as the establishment of guidelines and standardized procedures for the discretization of the area of study and the selection of physical models suitable for the simulation of BWR fuel. For this purpose several of BFBT experiments have simulated [1] provide a data base for the development of experiments for measuring distribution of fractions of holes to changes in power in order to find the most appropriate models for the simulation of the problem. (Author)

Simulations of ex-vessel fuel coolant interactions in a Nordic BWR using MC3D code

International Nuclear Information System (INIS)

Thakre, S.; Ma, W.

2013-08-01

Nordic Boiling Water Reactors (BWRs) employ a drywell cavity flooding technique as a nuclear severe accident management strategy. In case of core melt accident where the reactor pressure vessel will fail and the melt will eject from the lower head and fall into a water pool, may be in the form of a continuous jet. It is assumed that the melt jet will fragment, quench and form a coolable debris bed into the water pool. The melt interaction with a water pool may cause an energetic steam explosion which creates a potential risk towards the integrity of containment, leading to fission products release into the atmosphere. The results of the APRI-7 project suggest that the significant damage to containment structures by steam explosion cannot be ruled according to the state-of-the-art knowledge about corresponding accident scenario. In the follow-up project APRI-8 (2012-2016) one of the goals of the KTH research is to resolve the steam explosion energetics (SEE) issue, developing a risk-oriented framework for quantifying conditional threats to containment integrity for a Nordic type BWR. The present study deals with the premixing and explosion phase calculations of a Nordic BWR dry cavity, using MC3D, a multiphase CFD code for fuel coolant interactions. The main goal of the study is the assessment of pressure buildup in the cavity and the impact loading on the side walls. The conditions for the calculations are used from the SERENA-II BWR case exercise. The other objective was to do the sensitivity analysis of the parameters in modeling of fuel coolant interactions, which can help to reduce uncertainty in assessment of steam explosion energetics. The results show that the amount of liquid melt droplets in the water (region of void<0.6) is maximum even before reaching the jet at the bottom. In the explosion phase, maximum pressure is attained at the bottom and the maximum impulse on the wall is at the bottom of the wall. The analysis is carried out using two different

Simulations of ex-vessel fuel coolant interactions in a Nordic BWR using MC3D code

Energy Technology Data Exchange (ETDEWEB)

Thakre, S.; Ma, W. [Royal Institute of Technology, KTH. Div. of Nuclear Power Safety, Stockholm (Sweden)

2013-08-15

Nordic Boiling Water Reactors (BWRs) employ a drywell cavity flooding technique as a nuclear severe accident management strategy. In case of core melt accident where the reactor pressure vessel will fail and the melt will eject from the lower head and fall into a water pool, may be in the form of a continuous jet. It is assumed that the melt jet will fragment, quench and form a coolable debris bed into the water pool. The melt interaction with a water pool may cause an energetic steam explosion which creates a potential risk towards the integrity of containment, leading to fission products release into the atmosphere. The results of the APRI-7 project suggest that the significant damage to containment structures by steam explosion cannot be ruled according to the state-of-the-art knowledge about corresponding accident scenario. In the follow-up project APRI-8 (2012-2016) one of the goals of the KTH research is to resolve the steam explosion energetics (SEE) issue, developing a risk-oriented framework for quantifying conditional threats to containment integrity for a Nordic type BWR. The present study deals with the premixing and explosion phase calculations of a Nordic BWR dry cavity, using MC3D, a multiphase CFD code for fuel coolant interactions. The main goal of the study is the assessment of pressure buildup in the cavity and the impact loading on the side walls. The conditions for the calculations are used from the SERENA-II BWR case exercise. The other objective was to do the sensitivity analysis of the parameters in modeling of fuel coolant interactions, which can help to reduce uncertainty in assessment of steam explosion energetics. The results show that the amount of liquid melt droplets in the water (region of void<0.6) is maximum even before reaching the jet at the bottom. In the explosion phase, maximum pressure is attained at the bottom and the maximum impulse on the wall is at the bottom of the wall. The analysis is carried out using two different

Actinides record, power calculations and activity for present isotopes in the spent fuel of a BWR

International Nuclear Information System (INIS)

Enriquez C, P.; Ramirez S, J. R.; Lucatero, M. A.

2012-10-01

The administration of spent fuel is one of the more important stages of the nuclear fuel cycle, and this has become a problem of supreme importance in countries that possess nuclear reactors. Due to this in this work, the study on the actinides record and present fission products to the discharge of the irradiated fuel in a light water reactor type BWR is shown, to quantify the power and activity that emit to the discharge and during the cooling time. The analysis was realized on a fuel assembly type 10 x 10 with an enrichment average of 3.69 wt % in U-235 and the assembly simulation assumes four cycles of operation of 18 months each one and presents an exposition of 47 G Wd/Tm to the discharge. The module OrigenArp of the Scale 6 code is the computation tool used for the assembly simulation and to obtain the results on the actinides record presents to the fuel discharge. The study covers the following points: a) Obtaining of the plutonium vector used in the fuel production of mixed oxides, and b) Power calculation and activity for present actinides to the discharge. The results presented in this work, correspond at the same time immediate of discharge (0 years) and to a cooling stage in the irradiated fuel pool (5 years). (Author)

Fission gas release and pellet microstructure change of high burnup BWR fuel

International Nuclear Information System (INIS)

Itagaki, N.; Ohira, K.; Tsuda, K.; Fischer, G.; Ota, T.

1998-01-01

UO 2 fuel, with and without Gadolinium, irradiated for three, five, and six irradiation cycles up to about 60 GWd/t pellet burnup in a commercial BWR were studied. The fission gas release and the rim effect were investigated by the puncture test and gas analysis method, OM (optical microscope), SEM (scanning electron microscope), and EPMA (electron probe microanalyzer). The fission gas release rate of the fuel rods irradiated up to six cycles was below a few percent; there was no tendency for the fission gas release to increase abruptly with burnup. On the other hand, microstructure changes were revealed by OM and SEM examination at the rim position with burnup increase. Fission gas was found depleted at both the rim position and the pellet center region using EPMA. There was no correlation between the fission gas release measured by the puncture test and the fission gas depletion at the rim position using EPMA. However, the depletion of fission gas in the center region had good correlation with the fission gas release rate determined by the puncture test. In addition, because the burnup is very large at the rim position of high burnup fuel and also due to the fission rate of the produced Pu, the Xe/Kr ratio at the rim position of high burnup fuel is close to the value of the fission yield of Pu. The Xe/Kr ratio determined by the gas analysis after the puncture test was equivalent to the fuel average but not to the pellet rim position. From the results, it was concluded that fission gas at the rim position was released from the UO 2 matrix in high burnup, however, most of this released fission gas was held in the porous structure and not released from the pellet to the free volume. (author)

Evaluation of burnup characteristics and energy deposition during NSRR pulse irradiation tests on irradiated BWR fuels

International Nuclear Information System (INIS)

Nakamura, Takehiko; Yoshinaga, Makio

2000-11-01

Pulse irradiation tests of irradiated fuel are performed in the Nuclear Safety Research Reactor (NSRR) to investigate the fuel behavior under Reactivity Initiated Accident Conditions (RIA). The severity of the RIA is represented by energy deposition or peak fuel enthalpy during the power excursion. In case of the irradiated fuel tests, the energy deposition varies depending both on the amounts and distribution of residual fissile and neutron absorbing fission products generated during the base irradiation. Thus, proper fuel burnup characterization, especially for low enriched commercial fuels, is important, because plutonium (Pu) takes a large part of fissile and its generation depends on the neutron spectrum during the base irradiation. Fuel burnup calculations were conducted with ORIGEN2, RODBURN and SWAT codes for the BWR fuels tested in the NSRR. The calculation results were compared with the measured isotope concentrations and used for the NSRR neutron calculations to evaluate energy depositions of the test fuel. The comparison of the code calculations and the measurements revealed that the neutron spectrum change due to difference in void fraction altered Pu generation and energy deposition in the NSRR tests considerably. With the properly evaluated neutron spectrum, the combined burnup and NSRR neutron calculation gave reasonably good evaluation of the energy deposition. The calculations provided radial distributions of the fission product accumulation during the base irradiation and power distribution during the NSRR pulse irradiation, which were important for the evaluation of both burnup characteristics and fission gas release behavior. (author)

Equipment for nondestructive testing of the PWR and BWR spept fUel elements and assemblies in the NPP storage pools

International Nuclear Information System (INIS)

Gorskij, V.V.

1983-01-01

Design features are considered of units for nondestructive testing of spent fUel elements and fuel assemblies (FA) in the storage pools of NPP with the PWR and BWR reactors. Units for remote viewing control of fuel element cans and FA, for direct measurements of their geometrical dimensions, for FA leak-testing, fuel element can nondestructive testing and gamma scanning, for measuring gaseous fission product pressure and fuel element free volume are described along with units for complex checking of fuel element and FA parameters. The units for nondestructive testing of spent fuel elements and EA are shown to differ both in their designs and a number of checked parameters of fuel elements and FA. The remote viewing and those for measuring the basic FA parameters are most generally employed. Units for complex testing of multiple fuel element parameters, designed in the last few years, are intended for operation with FA disassembled partially or fully and are characteristic of a high degree of computer measuring automation both for the process control and data processing

Appraisal of BWR plutonium burners for energy centers

International Nuclear Information System (INIS)

Williamson, H.E.

1976-01-01

The design of BWR cores with plutonium loadings beyond the self-generation recycle (SGR) level is investigated with regard to their possible role as plutonium burners in a nuclear energy center. Alternative plutonium burner approaches are also examined including the substitution of thorium for uranium as fertile material in the BWR and the use of a high-temperature gas reactor (HTGR) as a plutonium burner. Effects on core design, fuel cycle facility requirements, economics, and actinide residues are considered. Differences in net fissile material consumption among the various plutonium-burning systems examined were small in comparison to uncertainties in HTGR, thorium cycle, and high plutonium-loaded LWR technology. Variation in the actinide content of high-level wastes is not likely to be a significant factor in determining the feasibility of alternate systems of plutonium utilization. It was found that after 10,000 years the toxicity of actinide high-level wastes from the plutonium-burning fuel cycles was less than would have existed if the processed natural ores had not been used for nuclear fuel. The implications of plutonium burning and possible future fuel cycle options on uranium resource conservation are examined in the framework of current ERDA estimates of minable uranium resources

Predicting formation enthalpies of metal hydrides

Energy Technology Data Exchange (ETDEWEB)

Andreasen, A.

2004-12-01

In order for the hydrogen based society viz. a society in which hydrogen is the primary energy carrier to become realizable an efficient way of storing hydrogen is required. For this purpose metal hydrides are serious candidates. Metal hydrides are formed by chemical reaction between hydrogen and metal and for the stable hydrides this is associated with release of heat ({delta}H{sub f} ). The more thermodynamically stable the hydride, the larger {delta}H{sub f}, and the higher temperature is needed in order to desorp hydrogen (reverse reaction) and vice versa. For practical application the temperature needed for desorption should not be too high i.e. {delta}H{sub f} should not be too large. If hydrogen desorption is to be possible below 100 deg C (which is the ultimate goal if hydrogen storage in metal hydrides should be used in conjunction with a PEM fuel cell), {delta}H{sub f} should not exceed -48 kJ/mol. Until recently only intermetallic metal hydrides with a storage capacity less than 2 wt.% H{sub 2} have met this criterion. However, discovering reversible hydrogen storage in complex metal hydrides such as NaAlH{sub 4} (5.5 wt. % reversible hydrogen capacity) have revealed a new group of potential candiates. However, still many combination of elements from the periodic table are yet to be explored. Since experimental determination of thermodynamic properties of the vast combinations of elements is tedious it may be advantagous to have a predictive tool for this task. In this report different ways of predicting {delta}H{sub f} for binary and ternary metal hydrides are reviewed. Main focus will be on how well these methods perform numerically i.e. how well experimental results are resembled by the model. The theoretical background of the different methods is only briefly reviewed. (au)

Preparation of beryllium hydride

International Nuclear Information System (INIS)

Bergeron, C.R.; Baker, R.W.

1975-01-01

Beryllium hydride of high bulk density, suitable for use as a component of high-energy fuels, is prepared by the pyrolysis, in solution in an inert solvent, of a ditertiary-alkyl beryllium. An agitator introduces mechanical energy into the reaction system, during the pyrolysis, at the rate of 0.002 to 0.30 horsepower per gallon of reaction mixture. (U.S.)

Characterization of a U-Mo alloy subjected to direct hydriding of the gamma phase

International Nuclear Information System (INIS)

Balart, Silvia N.; Bruzzoni, Pablo; Granovsky, Marta S.

2003-01-01

The Reduced Enrichment for Research and Test Reactors (RERTR) program has imposed the need to develop plate-type fuel elements based on high density uranium compounds, such as U-Mo alloys. One of the steps in the fabrication of the fuel elements is the pulverization of the fissile material. In the case of the U-Mo alloys, the pulverization can be accomplished through hydriding - dehydriding. Two alternative methods of the hydriding-dehydriding process, namely the selective hydriding in alpha phase (HS-alpha) and the massive hydriding in gamma phase (HM-gamma) are currently being studied at the Comision Nacional de Energia Atomica. The HM-gamma method was reproduced at laboratory scale starting from a U-7 wt % Mo alloy. The hydrided and dehydrided materials were characterized using metallographic techniques, scanning electron microscopy, energy dispersive X-ray analysis and X-ray diffraction. These results are compared with previous results of the HS-alpha method. (author)

Optimization of fuel cells for BWR based in Tabu modified search; Optimizacion de celdas de combustible para BWR basada en busqueda Tabu modificada

Energy Technology Data Exchange (ETDEWEB)

Martin del Campo M, C.; Francois L, J.L. [Facultad de Ingenieria, UNAM, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico); Palomera P, M.A. [Facultad de Ingenieria, UNAM, Posgrado en Ingenieria en Computacion, Circuito exterior s/n, Ciudad Universitaria, Mexico, D.F. (Mexico)]. e-mail: cmcm@fi-b.unam.mx

2004-07-01

The advances in the development of a computational system for the design and optimization of cells for assemble of fuel of Boiling Water Reactors (BWR) are presented. The method of optimization is based on the technique of Tabu Search (Tabu Search, TS) implemented in progressive stages designed to accelerate the search and to reduce the time used in the process of optimization. It was programed an algorithm to create the first solution. Also for to diversify the generation of random numbers, required by the technical TS, it was used the Makoto Matsumoto function obtaining excellent results. The objective function has been coded in such a way that can adapt to optimize different parameters like they can be the enrichment average or the peak factor of radial power. The neutronic evaluation of the cells is carried out in a fine way by means of the HELIOS simulator. In the work the main characteristics of the system are described and an application example is presented to the design of a cell of 10x10 bars of fuel with 10 different enrichment compositions and gadolinium content. (Author)

Evaluation of the thermal-mechanical performance of fuel rods of a BWR during a power ramp using the FUELSIM code

International Nuclear Information System (INIS)

Pantoja C, R.

2010-01-01

To avoid the risk to environment due to release of radioactive material, because of occurrence of an accident, it is the priority of the design and performance of the diverse systems of safety of a commercial nuclear power plant. The safety of nuclear power plants requires, therefore, monitoring those parameters having some direct or indirect effect on safety. The thermal limits are values set for those parameters considered having most impact on the safe operation of a nuclear power reactor. Some thermal limits monitoring requires the thermal-mechanical analysis of the rods containing the nuclear fuel. The fuel rod thermal-mechanical behavior under irradiation is a complex process in which there exists a great deal of interrelated physical and chemical phenomena, so that the fuel rod performance analysis in the core of a nuclear power reactor is generally accomplished by using computer codes, which integrate several of the phenomena that are expected to occur during the lifetime of the fuel rod in the core. The main application of the thermal-mechanical analysis codes is the prediction of occurrence of conditions and/or phenomena that could lead to the deterioration or even mechanical failure of the fuel rod cladding, as, for example, the pellet-cladding interaction. In the operation of a nuclear power reactor, fuel preconditioning operations refer to the operational procedures employed to reduce the fuel rod failure probability due to fuel-cladding interaction, specially during reactor startup. Preconditioning simulations are therefore necessary to determine in advance limit values for the power that can be generated in a fuel rod, and thus avoiding any rod damage. In this work, a first analysis of the thermal-mechanical performance of typical fuel rods used in nuclear reactors of the type BWR 5/6, as those two nuclear reactors in Laguna Verde, Veracruz, is performed. This study includes two types of fuel rods: one from a fuel assembly design with an array 8 x 8

Chemical Hydride Slurry for Hydrogen Production and Storage

Energy Technology Data Exchange (ETDEWEB)

McClaine, Andrew W

2008-09-30

University have demonstrated the technical viability of the process and have provided data for the cost analyses that have been performed. We also concluded that a carbothermic process could also produce magnesium at acceptable costs. The use of slurry as a medium to carry chemical hydrides has been shown during this project to offer significant advantages for storing, delivering, and distributing hydrogen: • Magnesium hydride slurry is stable for months and pumpable. • The oils of the slurry minimize the contact of oxygen and moisture in the air with the metal hydride in the slurry. Thus reactive chemicals, such as lithium hydride, can be handled safely in the air when encased in the oils of the slurry. • Though magnesium hydride offers an additional safety feature of not reacting readily with water at room temperatures, it does react readily with water at temperatures above the boiling point of water. Thus when hydrogen is needed, the slurry and water are heated until the reaction begins, then the reaction energy provides heat for more slurry and water to be heated. • The reaction system can be relatively small and light and the slurry can be stored in conventional liquid fuel tanks. When transported and stored, the conventional liquid fuel infrastructure can be used. • The particular metal hydride of interest in this project, magnesium hydride, forms benign byproducts, magnesium hydroxide (“Milk of Magnesia”) and magnesium oxide. • We have estimated that a magnesium hydride slurry system (including the mixer device and tanks) could meet the DOE 2010 energy density goals. During the investigation of hydriding techniques, we learned that magnesium hydride in a slurry can also be cycled in a rechargeable fashion. Thus, magnesium hydride slurry can act either as a chemical hydride storage medium or as a rechargeable hydride storage system. Hydrogen can be stored and delivered and then stored again thus significantly reducing the cost of storing and delivering

Complex metal hydrides for hydrogen, thermal and electrochemical energy storage

DEFF Research Database (Denmark)

Møller, Kasper T.; Sheppard, Drew; Ravnsbæk, Dorthe B.

2017-01-01

field. This review illustrates that complex metal hydrides may store hydrogen in the solid state, act as novel battery materials, both as electrolytes and electrode materials, or store solar heat in a more efficient manner as compared to traditional heat storage materials. Furthermore, it is highlighted...... how complex metal hydrides may act in an integrated setup with a fuel cell. This review focuses on the unique properties of light element complex metal hydrides mainly based on boron, nitrogen and aluminum, e.g., metal borohydrides and metal alanates. Our hope is that this review can provide new...

MCTP, a code for the thermo-mechanical analysis of a fuel rod of BWR type reactors (Neutron part)

International Nuclear Information System (INIS)

Hernandez L, H.; Ortiz V, J.

2003-01-01

In the National Institute of Nuclear Research of Mexico a code for the thermo-mechanical analysis of the fuel rods of the BWR type reactors of the Nucleo electric Central of Laguna Verde is developed. The code solves the diffusion equation in cylindrical coordinates with several energy groups. The code, likewise, calculates the temperature distribution and power distribution in those fuel rods. The code is denominated Multi groups With Temperatures and Power (MCTP). In the code, the energy with which the fission neutrons are emitted it is divided in six groups. They are also considered the produced perturbations by the changes in the temperatures of the materials that constitute the fuel rods, the content of fission products, the uranium consumption and in its case the gadolinium, as well as the plutonium production. In this work there are present preliminary results obtained with the code, using data of operation of the Nucleo electric Central of Laguna Verde. (Author)

Examination of minor actinide annihilation by BWR core

International Nuclear Information System (INIS)

Hida, Kazuki

1995-01-01

From the viewpoint of reducing burden for disposing high level waste generated from spent fuel, the examination of recycling minor actinide (MA) to reactors and reducing its accumulation has been advanced. In this study, the possibility of annihilation in the case of recycling it to a BWR was examined. The main MAs are 237 Np, 241 Am, 243 Am, 242 Cm, and 244 Cm. However, as for Cm isotopes, the half life is short, the amount of generation is small, and the rate of neutron emission is high, therefore, those are disposed as waste, and 237 Np, 241 Am and 243 Am were taken as the objects of recycling. In order to grasp the basic characteristics in the case of recycling MAs to a BWR, MAs were added to UO 2 fuel, MOX fuel and HCR fuel and burned, and the nuclear conversion characteristics were examined. As the result, it was found that they were converted to short half life nuclides, and as the neutron spectra were softer, the rate of annihilation was higher. In the case of recycling MAs by concentrating to a specific reactor, reactivity loss, the degree of uranium enrichment required for compensating reactivity, and the rate of MA annihilation were calculated. Based on these data, the MA recycling system was set up, and the rate of MA annihilation was evaluated. This is reported. (K.I.)

Impact analysis of modifying the composition of the nuclear fuel of a BWR with beryllium oxide

International Nuclear Information System (INIS)

Gallardo V, J. M.; Morales S, J. B.

2013-10-01

The beryllium oxide (Be O) presents excellent physical properties, especially its high thermal conductivity that contrasts clearly with that of the uranium dioxide (UO 2 ) used at the present as fuel in a great number of nuclear plants. The present work models a nuclear reactor cooled by light water in boiling with two external recirculation loops (BWR/5) using the code for the transitory analysis and postulated accidents Trac-B F1, implementing a UO 2 mixture and different fractions of Be O, with the objective of improving the thermal conductivity of the fuel. The numeric results and the realized analyses indicate that when adding a fraction in volume of 10% the central temperature decreases in 30.4% in stationary state, while during the large break loss of coolant accident the peak cladding temperature diminishes in 7%. Although the real interaction of the mixture has not been determined experimentally, the obtained results are promising. (Author)

Surface modification of a proton exchange membrane and hydrogen storage in a metal hydride for fuel cells

Science.gov (United States)

Andrews, Lisa

promising option. Effective hydrogen storage methods must be used as sources of available hydrogen. One possibility is to use hydrogen stored in a solid chemical compound such as magnesium hydride. The kinetics of hydrogen release from the hydrolysis of magnesium hydride with 2 wt% acetic acid was examined. The hydrogen produced was supplied to a fuel cell and the amount of hydrogen consumed by the fuel cell was determined. Carbon nanotubes also can play a role in energy sources and as components in fuel cells. VUV photo-oxidized single walled carbon nanotubes (SWNT) paper was grafted with polyacrylic acid and analyzed using XPS.

Radial optimization of a BWR fuel cell using genetic algorithms; Optimizacion radial de una celda de combustible BWR usando algoritmos geneticos

Energy Technology Data Exchange (ETDEWEB)

Martin del Campo M, C.; Carmona H, R.; Oropeza C, I.P. [UNAM, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)]. e-mail: cmcm@fi-b.unam.mx

2006-07-01

The development of the application of the Genetic Algorithms (GA) to the optimization of the radial distribution of enrichment in a cell of fuel of a BWR (Boiling Water Reactor) is presented. The optimization process it was ties to the HELIOS simulator, which is a transport code of neutron simulation of fuel cells that has been validated for the calculation of nuclear banks for BWRs. With heterogeneous radial designs can improve the radial distribution of the power, for what the radial design of fuel has a strong influence in the global design of fuel recharges. The optimum radial distribution of fuel bars is looked for with different enrichments of U{sup 235} and contents of consumable poison. For it is necessary to define the representation of the solution, the objective function and the implementation of the specific optimization process to the solution of the problem. The optimization process it was coded in 'C' language, it was automated the creation of the entrances to the simulator, the execution of the simulator and the extraction, in the exit of the simulator, of the parameters that intervene in the objective function. The objective function includes four parameters: average enrichment of the cell, average gadolinia concentration of the cell, peak factor of radial power and k-infinite multiplication factor. To be able to calculate the parameters that intervene in the objective function, the one evaluation process of GA was ties to the HELIOS code executed in a Compaq Alpha workstation. It was applied to the design of a fuel cell of 10 x 10 that it can be employee in the fuel assemble designs that are used at the moment in the Laguna Verde Nucleo electric Central. Its were considered 10 different fuel compositions which four contain gadolinia. Three heuristic rules that consist in prohibiting the placement of bars with gadolinia in the ends of the cell, to place the compositions with the smallest enrichment in the corners of the cell and to fix

Cross-sections for homogenized BWR fuel elements in 2d-diffusion theory by 1d-transport calculations

International Nuclear Information System (INIS)

Ambrosius, G.

1980-01-01

Leakage has a large influence on the thermal spectrum in a fuel rod cell of a BWR and originates: a) from rods with different absorptions and; b) from the different distances to the water gaps. Due to reason a) Gd-rods are treated together with a ring of the homogenized eight nearest neighbours. The often used definition of homogenized cross-sections as the ratio of the integrated reaction rate to the integrated flux proved to be inadequate. This homogenization method is exact as far as the flux is constant over the boundary and as the leakag e during calculating the homogenized cross-sections is similar to that during application. With respect to the condition b) a 1d-transport calculation for the whole fuel element with rings or slabs of homogenized fuel rod cells is performed. With the definition above the flux distribution is that of the fluxes in the moderator regions. The spectrum within each fuel rod cell which includes the leakage is calculated by superimposing at each energy on the flux distribution in the cell the flux at the cell position from the bundle calculation. Changes in the flux ratio between fuel and moderator due to the leakage are taken into account in a final few group 2d-diffusion calculation with fuel and (moderator + cladding) taken separately

Coretran/Vipre assembly critical power assessment against Nupec BWR experiments

International Nuclear Information System (INIS)

Aounallah, Y.

2001-01-01

This study has been performed, in the framework of the STARS project, to assess CORETRAN-01/VIPRE-02 code capability to predict critical heat flux conditions for BWR fuel assemblies. The assessment is based on comparisons of the code results with the NUPEC steady-state critical power measurements on full-scale assemblies tested under a range of flow conditions. Two assembly types were considered, the standard BWR 8 x 8 and the so-called ''high-burnup'' assembly, similar to GE-10. Code modelling options that have a significant impact on the results have been identified, along with code limitations. (author)

BWR type reactor core

International Nuclear Information System (INIS)

Tatemichi, Shin-ichiro.

1981-01-01

Purpose: To eliminate the variation in the power distribution of a BWR type reactor core in the axial direction even if the flow rate is increased or decreased by providing a difference in the void coefficient between the upper part and the lower parts of the reactor core, and increasing the void coefficient at the lower part of the reactor core. Constitution: The void coefficient of the lower region from the center to the lower part along the axial direction of a nuclear fuel assembly is increased to decrease the dependence on the flow rate of the axial power distribution of the nuclear fuel assembly. That is, a water/fuel ratio is varied, the water in non-boiled region is increased or the neutron spectrum is varied so as to vary the void coefficient. In order to exemplify it, the rate of the internal pellets of the fuel rod of the nuclear fuel assembly or the shape of the channel box is varied. Accordingly, the power does not considerably vary even if the flow rate is altered since the power is varied in the power operation. (Yoshihara, H.)

Development of Hydrogen Storage Tank Systems Based on Complex Metal Hydrides

Directory of Open Access Journals (Sweden)

Morten B. Ley

2015-09-01

Full Text Available This review describes recent research in the development of tank systems based on complex metal hydrides for thermolysis and hydrolysis. Commercial applications using complex metal hydrides are limited, especially for thermolysis-based systems where so far only demonstration projects have been performed. Hydrolysis-based systems find their way in space, naval, military and defense applications due to their compatibility with proton exchange membrane (PEM fuel cells. Tank design, modeling, and development for thermolysis and hydrolysis systems as well as commercial applications of hydrolysis systems are described in more detail in this review. For thermolysis, mostly sodium aluminum hydride containing tanks were developed, and only a few examples with nitrides, ammonia borane and alane. For hydrolysis, sodium borohydride was the preferred material whereas ammonia borane found less popularity. Recycling of the sodium borohydride spent fuel remains an important part for their commercial viability.

Development of Hydrogen Storage Tank Systems Based on Complex Metal Hydrides

Science.gov (United States)

Ley, Morten B.; Meggouh, Mariem; Moury, Romain; Peinecke, Kateryna; Felderhoff, Michael

2015-01-01

This review describes recent research in the development of tank systems based on complex metal hydrides for thermolysis and hydrolysis. Commercial applications using complex metal hydrides are limited, especially for thermolysis-based systems where so far only demonstration projects have been performed. Hydrolysis-based systems find their way in space, naval, military and defense applications due to their compatibility with proton exchange membrane (PEM) fuel cells. Tank design, modeling, and development for thermolysis and hydrolysis systems as well as commercial applications of hydrolysis systems are described in more detail in this review. For thermolysis, mostly sodium aluminum hydride containing tanks were developed, and only a few examples with nitrides, ammonia borane and alane. For hydrolysis, sodium borohydride was the preferred material whereas ammonia borane found less popularity. Recycling of the sodium borohydride spent fuel remains an important part for their commercial viability. PMID:28793541

Electrolytic hydriding and hydride distribution in zircaloy-4

International Nuclear Information System (INIS)

Gomes, M.H.L.

1974-01-01

A study has been made of the electrolytic hydriding of zircaloy-4 in the range 20-80 0 C, for reaction times from 5 to 30 hours, and the effect of potential, pH and dissolved oxygen has been investigated. The hydriding reaction was more sensitive to time and temperature conditions than to the electrochemical variables. It has been shown that a controlled introduction of hydrides in zircaloy is feasible. Hydrides were found to be plate like shaped and distributed mainly along grain-boundaries. It has been shown that hydriding kinetics do not follow a simple law but may be described by a Johnson-Mehl empirical equation. On the basis of this equation an activation energy of 9.400 cal/mol has been determined, which is close to the activation energy for diffusion of hydrogen in the hydride. (author)

U-8 wt %Mo and 7 wt %Mo alloys powder obtained by an hydride-de hydride process; Obtencion de polvo de aleaciones U-8% Mo y U-7% Mo (en peso) mediante hidruracion

Energy Technology Data Exchange (ETDEWEB)

Balart, Silvia N; Bruzzoni, Pablo; Granovsky, Marta S; Gribaudo, Luis M.J.; Hermida, Jorge D; Ovejero, Jose; Rubiolo, Gerardo H; Vicente, Eduardo E [Comision Nacional de Energia Atomica, General San Martin (Argentina). Dept. de Materiales

2000-07-01

Uranium-molybdenum alloys are been tested as a component in high-density LEU dispersion fuels with very good performances. These alloys need to be transformed to powder due to the manufacturing requirements of the fuels. One method to convert ductile alloys into powder is the hydride-de hydride process, which takes advantage of the ability of the U-{alpha} phase to transform to UH{sub 3}: a brittle and relatively low-density compound. U-Mo alloys around 7 and 8 wt % Mo were melted and heat treated at different temperature ranges in order to partially convert {gamma} -phase to {alpha} -phase. Subsequent hydriding transforms this {alpha} -phase to UH{sub 3}. The volume change associated to the hydride formation embrittled the material which ends up in a powdered alloy. Results of the optical metallography, scanning electron microscopy, X-ray diffraction during different steps of the process are shown. (author)

Flux and power distributions in BWR multi-bundle fuel arrays

International Nuclear Information System (INIS)

Cheng, H.S.

1976-02-01

Multi-bundle calculations have been performed in order to shed some light on an abnormal TIP trace recently discovered in a BWR/3. Transport theory was employed to perform the calculations with ENDF/B-IV data. The results indicate that a strong variation of the TIP reading does exist along the narrow water gap of a BWR due to the steep gradient of the thermal neutron flux; the maxima occurring at the intersections of the water gaps and the minima in between. Using this characteristic behavior of the TIP reading, together with the observed normal TIP trace, the abnormal behavior of the affected TIP trace exhibiting three peaks along the channel was roughly simulated. The calculations confirmed that the observed TIP trace anomaly was caused by the severe bending of the affected instrument tube as was actually discovered. The effect of hot water intrusion into the TIP guide tube, as well as that of loading the new 8 x 8 reload bundles, was also evaluated

Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

International Nuclear Information System (INIS)

Greneche, D.

2014-01-01

This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

Draft of M2 Report on Integration of the Hybrid Hydride Model into INL's MBM Framework for Review

Energy Technology Data Exchange (ETDEWEB)

Tikare, Veena; Weck, Philippe F.; Schultz, Peter Andrew; Clark, Blythe; Glazoff, Michael V.; Homer, Eric R.

2014-07-01

This report documents the development, demonstration and validation of a mesoscale, microstructural evolution model for simulation of zirconium hydride {delta}-ZrH{sub 1.5} precipitation in the cladding of used nuclear fuels that may occur during long-term dry storage. While the Zr-based claddings are manufactured free of any hydrogen, they absorb hydrogen during service, in the reactor by a process commonly termed ‘hydrogen pick-up’. The precipitation and growth of zirconium hydrides during dry storage is one of the most likely fuel rod integrity failure mechanisms either by embrittlement or delayed hydride cracking of the cladding. While the phenomenon is well documented and identified as a potential key failure mechanism during long-term dry storage (NUREG/CR-7116), the ability to actually predict the formation of hydrides is poor. The model being documented in this work is a computational capability for the prediction of hydride formation in different claddings of used nuclear fuels. This work supports the Used Fuel Disposition Research and Development Campaign in assessing the structural engineering performance of the cladding during and after long-term dry storage. This document demonstrates a basic hydride precipitation model that is built on a recently developed hybrid Potts-phase field model that combines elements of Potts-Monte Carlo and the phase-field models. The model capabilities are demonstrated along with the incorporation of the starting microstructure, thermodynamics of the Zr-H system and the hydride formation mechanism.

Tritium processing using metal hydrides

International Nuclear Information System (INIS)

Mallett, M.W.

1986-01-01

E.I. duPont de Nemours and Company is commissioned by the US Department of Energy to operate the Savannah River Plant and Laboratory. The primary purpose of the plant is to produce radioactive materials for national defense. In keeping with current technology, new processes for the production of tritium are being developed. Three main objectives of this new technology are to ease the processing of, ease the storage of, and to reduce the operating costs of the tritium production facility. Research has indicated that the use of metal hydrides offers a viable solution towards satisfying these objectives. The Hydrogen and Fuels Technology Division has the responsibility to conduct research in support of the tritium production process. Metal hydride technology and its use in the storage and transportation of hydrogen will be reviewed

Some new techniques in tritium gas handling as applied to metal hydride synthesis

International Nuclear Information System (INIS)

Nasise, J.E.

1988-01-01

A state-of-the-art tritium Hydriding Synthesis System (HSS) was designed and built to replace the existing system within the Tritium Salt Facility (TSF) at the Los Alamos National Laboratory. This new hydriding system utilizes unique fast-cycling 7.9 mole uranium beds (47.5g of T at 100% loading) and novel gas circulating hydriding furnaces. Tritium system components discussed include fast-cycling uranium beds, circulating gas hydriding furnaces, valves, storage volumes, manifolds, gas transfer pumps, and graphic display and control consoles. Many of the tritium handling and processing techniques incorporated into this system are directly applicable to today's fusion fuel loops. 12 refs., 7 figs

Comminution of the U-10Mo by hydriding cycles innovative process

Energy Technology Data Exchange (ETDEWEB)

Faeda, Kelly C.M.; Santos, Ana Maria M. dos; Paula, Joao B. de; Pereira, Edilson M.; Pedrosa, Tercio A.; Lameiras, Fernando S.; Ferraz, Wilmar B., E-mail: ferrazw@cdtn.br, E-mail: kelly.faeda@prof.una.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2013-07-01

The research, test and producing radioisotopes compact reactors were developed with the use of high levels of enriched fuel of approximately 90% of the fissile isotope U-235. Since the 80s', a policy under the context of international program RERTR (Reduced Enrichment for Research and Test Reactors) encourages the fuel replacement of the high enriched fuel by the low one of about 20 % U-235. One way to compensate the substitution for the low enrichment fuel is to employ high density metal uranium alloys. The fabrication of compact reactor fuel uses the metal matrix dispersion and, for this, uranium alloys are used in the form of powders. Despite the high densities, the metallic uranium based alloys are ductile and therefore difficult to be comminuted. Among the different comminution processes, the hydriding-dehydriding process has proved most advantageous, primarily due to their relative simplicity of processing and low manufacturing cost. In this paper, we present the results of the development of the U-10Mo alloy comminution process by the hydriding-dehydriding method on a laboratory scale. Samples of the alloy were subjected to different hydriding cycle numbers in order to verify its influence in relation to the particle size distribution of powders. Powders of different particle sizes were obtained and characterized by the physical and morphological characteristics by optical microscopy, scanning electron microscopy and X ray diffraction. The obtained results are evaluated and discussed. (author)

Coretran/Vipre assembly critical power assessment against Nupec BWR experiments

Energy Technology Data Exchange (ETDEWEB)

Aounallah, Y. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

2001-07-01

This study has been performed, in the framework of the STARS project, to assess CORETRAN-01/VIPRE-02 code capability to predict critical heat flux conditions for BWR fuel assemblies. The assessment is based on comparisons of the code results with the NUPEC steady-state critical power measurements on full-scale assemblies tested under a range of flow conditions. Two assembly types were considered, the standard BWR 8 x 8 and the so-called ''high-burnup'' assembly, similar to GE-10. Code modelling options that have a significant impact on the results have been identified, along with code limitations. (author)

BWR and PWR chemistry operating experience and perspectives

International Nuclear Information System (INIS)

Fruzzetti, K.; Garcia, S.; Lynch, N.; Reid, R.

2014-01-01

It is well recognized that proper control of water chemistry plays a critical role in ensuring the safe and reliable operation of Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). State-of-the-art water chemistry programs reduce general and localized corrosion of reactor coolant system, steam cycle equipment, and fuel cladding materials; ensure continued integrity of cycle components; and reduce radiation fields. Once a particular nuclear plant component has been installed or plant system constructed, proper water chemistry provides a global tool to mitigate materials degradation problems, thereby reducing the need for costly repairs or replacements. Recognizing the importance of proper chemistry control and the value in understanding the relationship between chemistry guidance and actual operating experience, EPRI continues to collect, monitor, and evaluate operating data from BWRs and PWRs around the world. More than 900 cycles of valuable BWR and PWR operating chemistry data has been collected, including online, startup and shutdown chemistry data over more than 10 years (> 20 years for BWRs). This paper will provide an overview of current trends in BWR and PWR chemistry, focusing on plants in the U.S.. Important chemistry parameters will be highlighted and discussed in the context of the EPRI Water Chemistry Guidelines requirements (i.e., those parameters considered to be of key importance as related to the major goals identified in the EPRI Guidelines: materials integrity; fuel integrity; and minimizing plant radiation fields). Perspectives will be provided in light of recent industry initiatives and changes in the EPRI BWR and PWR Water Chemistry Guidelines. (author)

ABB high burnup fuel

International Nuclear Information System (INIS)

Andersson, S.; Helmersson, S.; Nilsson, S.; Jourdain, P.; Karlsson, L.; Limback, M.; Garde, A.M.

1999-01-01

Fuel designed and fabricated by ABB is now operating in 40 PWRs and BWRs in Europe, the United States and Korea. An excellent fuel reliability track record has been established. High burnups are proven for both PWR and BWR. Thermal margin improving features and advanced burnable absorber concepts enable the utilities to adopt demanding duty cycles to meet new economic objectives. In particular we note the excellent reliability record of ABB PWR fuel equipped with Guardian TM debris filter proven to meet the 6 rod-cycles fuel failure goal, and the out-standing operating record of the SVEA 10 x 10 fuel, where ABB is the only vendor to date with batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of PWR and BWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its utility customers. This paper provides an overview of recent fuel performance and reliability experience at ABB. Selected development and validation activities for PWR and BWR fuel are presented, for which the ABB test facilities in Windsor (TF-2 loop, mechanical test laboratory) and Vaesteras (FRIGG, BURE) are essential. (authors)

Metal Hydride Compression

Energy Technology Data Exchange (ETDEWEB)

Johnson, Terry A. [Sandia National Lab. (SNL-CA), Livermore, CA (United States); Bowman, Robert [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Smith, Barton [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Anovitz, Lawrence [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jensen, Craig [Hawaii Hydrogen Carriers LLC, Honolulu, HI (United States)

2017-07-01

Conventional hydrogen compressors often contribute over half of the cost of hydrogen stations, have poor reliability, and have insufficient flow rates for a mature FCEV market. Fatigue associated with their moving parts including cracking of diaphragms and failure of seal leads to failure in conventional compressors, which is exacerbated by the repeated starts and stops expected at fueling stations. Furthermore, the conventional lubrication of these compressors with oil is generally unacceptable at fueling stations due to potential fuel contamination. Metal hydride (MH) technology offers a very good alternative to both conventional (mechanical) and newly developed (electrochemical, ionic liquid pistons) methods of hydrogen compression. Advantages of MH compression include simplicity in design and operation, absence of moving parts, compactness, safety and reliability, and the possibility to utilize waste industrial heat to power the compressor. Beyond conventional H2 supplies of pipelines or tanker trucks, another attractive scenario is the on-site generating, pressuring and delivering pure H₂ at pressure (≥ 875 bar) for refueling vehicles at electrolysis, wind, or solar generating production facilities in distributed locations that are too remote or widely distributed for cost effective bulk transport. MH hydrogen compression utilizes a reversible heat-driven interaction of a hydride-forming metal alloy with hydrogen gas to form the MH phase and is a promising process for hydrogen energy applications [1,2]. To deliver hydrogen continuously, each stage of the compressor must consist of multiple MH beds with synchronized hydrogenation & dehydrogenation cycles. Multistage pressurization allows achievement of greater compression ratios using reduced temperature swings compared to single stage compressors. The objectives of this project are to investigate and demonstrate on a laboratory scale a two-stage MH hydrogen (H₂) gas compressor with a

Comparative analysis of results between CASMO, MCNP and Serpent for a suite of Benchmark problems on BWR reactors; Analisis comparativo de resultados entre CASMO, MCNP y SERPENT para una suite de problemas Benchmark en reactores BWR

Energy Technology Data Exchange (ETDEWEB)

Xolocostli M, J. V.; Vargas E, S.; Gomez T, A. M. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Reyes F, M. del C.; Del Valle G, E., E-mail: vicente.xolocostli@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, UP - Adolfo Lopez Mateos, Edif. 9, 07738 Mexico D. F. (Mexico)

2014-10-15

In this paper a comparison is made in analyzing the suite of Benchmark problems for reactors type BWR between CASMO-4, MCNP6 and Serpent code. The Benchmark problem consists of two different geometries: a fuel cell of a pin and assembly type BWR. To facilitate the study of reactors physics in the fuel pin their nuclear characteristics are provided to detail, such as burnt dependence, the reactivity of selected nuclide, etc. With respect to the fuel assembly, the presented results are regarding to infinite multiplication factor for burning different steps and different vacuum conditions. Making the analysis of this set of Benchmark problems provides comprehensive test problems for the next fuels generation of BWR reactors with high extended burned. It is important to note that when making this comparison the purpose is to validate the methodologies used in modeling for different operating conditions, if the case is of other BWR assembly. The results will be within a range with some uncertainty, considering that does not depend on code that is used. Escuela Superior de Fisica y Matematicas of Instituto Politecnico Nacional (IPN (Mexico) has accumulated some experience in using Serpent, due to the potential of this code over other commercial codes such as CASMO and MCNP. The obtained results for the infinite multiplication factor are encouraging and motivate the studies to continue with the generation of the X S of a core to a next step a respective nuclear data library is constructed and this can be used by codes developed as part of the development project of the Mexican Analysis Platform of Nuclear Reactors AZTLAN. (Author)

Obtaining zircaloy powder through hydriding

International Nuclear Information System (INIS)

Dupim, Ivaldete da Silva; Moreira, Joao M.L.

2009-01-01

Zirconium alloys are good options for the metal matrix in dispersion fuels for power reactors due to their low thermal neutron absorption cross-section, good corrosion resistance, good mechanical strength and high thermal conductivity. A necessary step for obtaining such fuels is producing Zr alloy powder for the metal matrix composite material. This article presents results from the Zircaloy-4 hydrogenation tests with the purpose to embrittle the alloy as a first step for comminuting. Several hydrogenation tests were performed and studied through thermogravimetric analysis. They included H 2 pressures of 25 and 50 kPa and temperatures ranging between from 20 to 670 deg C. X-ray diffraction analysis showed in the hydrogenated samples the predominant presence of ZrH 2 and some ZrO 2 . Some kinetics parameters for the Zircaloy-4 hydrogenation reaction were obtained: the time required to reach the equilibrium state at the dwell temperature was about 100 minutes; the hydrogenation rate during the heating process from 20 to 670 deg C was about 21 mg/h, and at constant temperature of 670 deg C, the hydride rate was about 1.15 mg/h. The hydrogenation rate is largest during the heating process and most of it occurs during this period. After hydrogenated, the samples could easily be comminuted indicating that this is a possible technology to obtain Zircaloy powder. The results show that only few minutes of hydrogenation are necessary to reach the hydride levels required for comminuting the Zircaloy. The final hydride stoichiometry was between 2.7 and 2.8 H for each Zr atom in the sample (author)

Advanced chemistry management system to optimize BWR chemistry control

International Nuclear Information System (INIS)

Maeda, K.; Nagasawa, K.

2002-01-01

BWR plant chemistry control has close relationships among nuclear safety, component reliability, radiation field management and fuel integrity. Advanced technology is required to improve chemistry control [1,3,6,7,10,11]. Toshiba has developed TACMAN (Toshiba Advanced Chemistry Management system) to support BWR chemistry control. The TACMAN has been developed as response to utilities' years of requirements to keep plant operation safety, reliability and cost benefit. The advanced technology built into the TACMAN allows utilities to make efficient chemistry control and to keep cost benefit. TACMAN is currently being used in response to the needs for tools those plant chemists and engineers could use to optimize and identify plant chemistry conditions continuously. If an incipient condition or anomaly is detected at early stage, root causes evaluation and immediate countermeasures can be provided. Especially, the expert system brings numerous and competitive advantages not only to improve plant chemistry reliability but also to standardize and systematize know-how, empirical knowledge and technologies in BWR chemistry This paper shows detail functions of TACMAN and practical results to evaluate actual plant. (authors)

The JAERI code system for evaluation of BWR ECCS performance

International Nuclear Information System (INIS)

Kohsaka, Atsuo; Akimoto, Masayuki; Asahi, Yoshiro; Abe, Kiyoharu; Muramatsu, Ken; Araya, Fumimasa; Sato, Kazuo

1982-12-01

Development of respective computer code system of BWR and PWR for evaluation of ECCS has been conducted since 1973 considering the differences of the reactor cooling system, core structure and ECCS. The first version of the BWR code system, of which developmental work started earlier than that of the PWR, has been completed. The BWR code system is designed to provide computational tools to analyze all phases of LOCAs and to evaluate the performance of the ECCS including an ''Evaluation Model (EM)'' feature in compliance with the requirements of the current Japanese Evaluation Guideline of ECCS. The BWR code system could be used for licensing purpose, i.e. for ECCS performance evaluation or audit calculations to cross-examine the methods and results of applicants or vendors. The BWR code system presented in this report comprises several computer codes, each of which analyzes a particular phase of a LOCA or a system blowdown depending on a range of LOCAs, i.e. large and small breaks in a variety of locations in the reactor system. The system includes ALARM-B1, HYDY-B1 and THYDE-B1 for analysis of the system blowdown for various break sizes, THYDE-B-REFLOOD for analysis of the reflood phase and SCORCH-B2 for the calculation of the fuel assembl hot plane temperature. When the multiple codes are used to analyze a broad range of LOCA as stated above, it is very important to evaluate the adequacy and consistency between the codes used to cover an entire break spectrum. The system consistency together with the system performance are discussed for a large commercial BWR. (author)

Combined on-board hydride slurry storage and reactor system and process for hydrogen-powered vehicles and devices

Science.gov (United States)

Brooks, Kriston P; Holladay, Jamelyn D; Simmons, Kevin L; Herling, Darrell R

2014-11-18

An on-board hydride storage system and process are described. The system includes a slurry storage system that includes a slurry reactor and a variable concentration slurry. In one preferred configuration, the storage system stores a slurry containing a hydride storage material in a carrier fluid at a first concentration of hydride solids. The slurry reactor receives the slurry containing a second concentration of the hydride storage material and releases hydrogen as a fuel to hydrogen-power devices and vehicles.

Fission product model for BWR analysis with improved accuracy in high burnup

International Nuclear Information System (INIS)

Ikehara, Tadashi; Yamamoto, Munenari; Ando, Yoshihira

1998-01-01

A new fission product (FP) chain model has been studied to be used in a BWR lattice calculation. In attempting to establish the model, two requirements, i.e. the accuracy in predicting burnup reactivity and the easiness in practical application, are simultaneously considered. The resultant FP model consists of 81 explicit FP nuclides and two lumped pseudo nuclides having the absorption cross sections independent of burnup history and fuel composition. For the verification, extensive numerical tests covering over a wide range of operational conditions and fuel compositions have been carried out. The results indicate that the estimated errors in burnup reactivity are within 0.1%Δk for exposures up to 100GWd/t. It is concluded that the present model can offer a high degree of accuracy for FP representation in BWR lattice calculation. (author)

Hydride precipitation crack propagation in zircaloy cladding during a decreasing temperature history

International Nuclear Information System (INIS)

Stout, R.B.

2001-01-01

An assessment of safety, design, and cost tradeoff issues for short (ten to fifty years) and longer (fifty to hundreds of years) interim dry storage of spent nuclear fuel in Zircaloy rods shall address potential failures of the Zircaloy cladding caused by the precipitation response of zirconium hydride platelets. To perform such assessment analyses rigorously and conservatively will be necessarily complex and difficult. For Zircaloy cladding, a model for zirconium hydride induced crack propagation velocity was developed for a decreasing temperature field and for hydrogen, temperature, and stress dependent diffusive transport of hydrogen to a generic hydride platelet at a crack tip. The development of the quasi-steady model is based on extensions of existing models for hydride precipitation kinetics for an isolated hydride platelet at a crack tip. An instability analysis model of hydride-crack growth was developed using existing concepts in a kinematic equation for crack propagation at a constant thermodynamic crack potential subject to brittle fracture conditions. At the time an instability is initiated, the crack propagation is no longer limited by hydride growth rate kinetics, but is then limited by stress rates. The model for slow hydride-crack growth will be further evaluated using existing available data. (authors)

Hydride precipitation crack propagation in zircaloy cladding during a decreasing temperature history

Energy Technology Data Exchange (ETDEWEB)

Stout, R.B. [California Univ., Livermore, CA (United States). Lawrence Livermore National Lab

2001-07-01

An assessment of safety, design, and cost tradeoff issues for short (ten to fifty years) and longer (fifty to hundreds of years) interim dry storage of spent nuclear fuel in Zircaloy rods shall address potential failures of the Zircaloy cladding caused by the precipitation response of zirconium hydride platelets. To perform such assessment analyses rigorously and conservatively will be necessarily complex and difficult. For Zircaloy cladding, a model for zirconium hydride induced crack propagation velocity was developed for a decreasing temperature field and for hydrogen, temperature, and stress dependent diffusive transport of hydrogen to a generic hydride platelet at a crack tip. The development of the quasi-steady model is based on extensions of existing models for hydride precipitation kinetics for an isolated hydride platelet at a crack tip. An instability analysis model of hydride-crack growth was developed using existing concepts in a kinematic equation for crack propagation at a constant thermodynamic crack potential subject to brittle fracture conditions. At the time an instability is initiated, the crack propagation is no longer limited by hydride growth rate kinetics, but is then limited by stress rates. The model for slow hydride-crack growth will be further evaluated using existing available data. (authors)

The Effect of Peak Temperatures and Hoop Stresses on Hydride Reorientations of Zirconium Alloy Cladding Tubes under Interim Dry Storage Condition

International Nuclear Information System (INIS)

Cha, Hyun Jin; Jang, Ki Nam; Kim, Kyu Tae

2016-01-01

In this study, the effect of peak temperatures and hoop tensile stresses on hydride reorientation in cladding was investigated. It was shown that the 250ppm-H specimens generated larger radial hydride fractions and longer radial hydrides than the 500ppm-H ones. The precipitated hydride in radial direction severely degrades mechanical properties of spent fuel rod. Hydride reorientation is related to cladding material, cladding temperature, hydrogen contents, thermal cycling, hoop stress and cooling rate. US NRC established the regulation on cladding temperature during the dry storage, which is the maximum fuel cladding temperature should not exceed 400 .deg. C for all fuel burnups under normal conditions of storage. However, if it is proved that the best estimate cladding hoop stress is equal to or less than 90MPa for the temperature limit proposed, a higher short-term temperature limit is allowed for low burnup fuel. In this study, 250ppm and 500ppm hydrogen-charged Zr-Nb alloy cladding tubes were selected to evaluate the effect of peak temperatures and hoop tensile stresses on the hydride reorientation during the dry storage. In order to evaluate threshold stresses in relation to various peak temperatures, four peak temperatures of 250, 300, 350, and 400 .deg. C and three tensile hoop stresses of 80, 100, 120MPa were selected.

Hydrogen metal hydride storage with integrated catalytic recombiner for mobile application

Energy Technology Data Exchange (ETDEWEB)

Marinescu-Pasoi, L.; Behrens, U.; Langer, G.; Gramatte, W.; Rastogi, A.K.; Schmitt, R.E. (Battelle-Institut e.V., Frankfurt am Main (DE). Dept. of Energy Technology)

1991-01-01

A novel, thermodynamically efficient device is under development at Battelle in Frankfurt, by which the range of hydrogen-driven cars with a metal hydride tank might be roughly doubled. The device makes use of the properties of metal hydrides, combined with catalytic combustion. Its development is funded by the Hessian Ministry of Economic Affairs and Technology; it is to be completed by the end of 1990. High-temperature hydrides (HTH) have about three times the storage capacity of low temperature hydrides (LTH), but require relatively large amounts of heat at high temperatures to release the hydrogen. The exhaust heat from combustion-engine-driven vehicles is insufficient for this, and vehicles with electric (fuel cell) drive produce practically no exhaust heat at all. The Battelle-developed device is a combination of an HTH storage cell, an LTH storage cell and a catalyst. (author).

BWR Spent Nuclear Fuel Interfacial Bonding Efficiency Study

Energy Technology Data Exchange (ETDEWEB)

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

2015-04-30

The objective of this project is to perform a systematic study of spent nuclear fuel (SNF, also known as “used nuclear fuel” [UNF]) integrity under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL) in August 2013. Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmark tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. The clad of the HBR fuels was made of Zircaloy-4. Testing was continued in fiscal year (FY) 2014 using Department of Energy (DOE) funds. Additional CIRFT testing was conducted on three HBR rods; two specimens failed, and one specimen was tested to over 2.23 × 10⁷ cycles without failing. The data analysis on all the HBR SNF rods demonstrated that it is necessary to characterize the fatigue life of the SNF rods in terms of (1) the curvature amplitude and (2) the maximum absolute of curvature extremes. The maximum extremes are significant because they signify the maximum tensile stress for the outer fiber of the bending rod. CIRFT testing has also addressed a large variation in hydrogen content on the HBR rods. While the load amplitude is the dominant factor that controls the fatigue life of bending rods, the hydrogen content also has an important effect on the lifetime attained at each load range tested. In FY 15, eleven SNF rod segments from the Limerick BWR were tested using the ORNL CIRFT equipment; one test under static conditions and ten tests under dynamic loading conditions. Under static unidirectional loading, a moment of 85 N·m was obtained at a maximum curvature of 4.0 m^-1. The specimen did not show any sign of failure during three repeated loading cycles to a similar maximum curvature. Ten cyclic tests were conducted with amplitudes varying from 15.2 to 7.1 N·m. Failure was observed in nine of

BWR stability: analysis of cladding temperature for high amplitude oscillations - 146

International Nuclear Information System (INIS)

Pohl, P.; Wehle, F.

2010-01-01

Power oscillations associated with density waves in boiling water reactors (BWRs) have been studied widely. Industrial research in this area is active since the invention of the first BWR. Stability measurements have been performed in various plants during commissioning phase but especially the magnitude and divergent nature of the oscillations during the LaSalle Unit 2 nuclear power plant event on March 9, 1988, renewed concern about the state of knowledge on BWR instabilities and possible consequences to fuel rod integrity. The objective of this paper is to present a simplified stability tool, applicable for stability analysis in the non-linear regime, which extends to high amplitude oscillations where inlet reverse flow occurs. In case of high amplitude oscillations a cyclical dryout and rewetting process at the fuel rod may take place, which leads in turn to rapid changes of the heat transfer from the fuel rod to the coolant. The application of this stability tool allows for a conservative determination of the fuel rod cladding temperature in case of high amplitude oscillations during the dryout / re-wet phase. Moreover, it reveals in good agreement to experimental findings the stabilizing effect of the reverse bundle inlet flow, which might be obtained for large oscillation amplitudes. (authors)

Design of a mixed recharge with MOX assemblies of greater relation of moderation for a BWR reactor; Diseno de una recarga mixta con ensambles MOX de mayor relacion de moderacion para un reactor BWR

Energy Technology Data Exchange (ETDEWEB)

Ramirez S, J.R.; Alonso V, G.; Palacios H, J. [ININ, Carretera Mexico-Toluca Km. 36.5, 52045 Estado de Mexico (Mexico)]. e-mail: jrrs@nuclear.inin.mx

2004-07-01

The study of the fuel of mixed oxides of uranium and plutonium (MOX) it has been topic of investigation in many countries of the world and those are even discussed in many places the benefits of reprocessing the spent fuel to extract the plutonium created during the irradiation of the fuel in the nuclear power reactors. At the moment those reactors that have been loaded partially with MOX fuel, are mainly of the type PWR where a mature technology has been achieved in some countries like they are France, Belgium and England, however the experience with reactors of the type BWR is more limited and it is continued studying the best way to introduce this type of fuel in BWRs, one of the main problems to introduce MOX in reactors BWR is the neutronic design of the same one, existing different concepts to introduce the plutonium in the assemblies of fuel and one of them is the one of increasing the relationship of moderation of the assemble. In this work a MOX fuel assemble design is presented and the obtained results so far in the ININ. These results indicate that the investigated concept has some exploitable advantages in the use of the MOX fuel. (Author)

Effect of chemical composition on corrosion resistance of Zircaloy fuel cladding tube for BWR

International Nuclear Information System (INIS)

Inagaki, Masahisa; Akahori, Kimihiko; Kuniya, Jirou; Masaoka, Isao; Suwa, Masateru; Maru, Akira; Yasuda, Teturou; Maki, Hideo.

1990-01-01

Effects of Fe and Ni contents on nodular corrosion susceptibility and hydrogen pick-up of Zircaloy were investigated. Total number of 31 Zr alloys having different chemical compositions; five Zr-Sn-Fe-Cr alloys, eight Zr-Sn-Fe-Ni alloys and eighteen Zr-Sn-Fe-Ni-Cr alloys, were melted and processed to thin plates for the corrosion tests in the environments of a high temperature (510degC) steam and a high temperature (288degC) water. In addition, four 450 kg ingots of Zr-Sn-Fe-Ni-Cr alloys were industrially melted and BWR fuel cladding tubes were manufactured through a current material processing sequence to study their producibility, tensile properties and corrosion resistance. Nodular corrosion susceptibility decreased with increasing Fe and Ni contents of Zircaloys. It was seen that the improved Zircaloys having Fe and Ni contents in the range of 0.30 [Ni]+0.15[Fe]≥0.045 (w%) showed no susceptibility to nodular corrosion. An increase of Fe content resulted in a decrease of hydrogen pick-up fraction in both steam and water environments. An increase of Fe and Ni content of Zircaloys in the range of Fe≤0.25 w% and Ni≤0.1 w% did not cause the changes in tensile properties and fabricabilities of fuel cladding tube. The fuel cladding tube of improved Zircaloy, containing more amount of Fe and Ni than the upper limit of Zircaloy-2 specification showed no susceptibility to nodular corrosion even in the 530degC steam test. (author)

PWR and BWR light water reactor systems in the USA and their fuel cycle

International Nuclear Information System (INIS)

Crawford, W.D.

1977-01-01

Light water reactor operating experience in the USA can be considered to date from the choice of the PWR for use in the naval reactor programme and the subsequent construction and operation of the nuclear power plant at Shippingport in 1957. The development of the BWR in 1954 and its selection for the plant at Dresden in 1959 established this concept as the other major reactor type in the US nuclear power programme. The subsequent growth profile is presented. A significant operating record has been accumulated concerning the availability of each of these reactor types. In addition, the use and performance of BWRs and PWRs in meeting system load requirements is discussed. The growing concern regarding possible terrorist activities and other potential threats has resulted in systems and procedures designed to ensure effective safeguards at nuclear power installations; current measures are described. Environmental effects of operating plants are subject to both radiological and non-radiological monitoring. The operating results achieved and the types of modifications that have been required of operating plants by the Nuclear Regulatory Commission are reviewed. Both fuel cycles are examined in terms of: fuel burnup experience and prospects for improvement; natural uranium resources; enrichment capacity; reprocessing and recycle; and the interrelationships among the latter three factors. High-level waste management currently involving on-site storage of spent fuel is discussed in terms of available capacity and plans for expansion. The US electric utility industry viewpoint regarding an ultimate programme for waste management is outlined. Finally, the current economics and future cost trends of nuclear power plants are evaluated. (author)

Latest experiences in inspecting the inside of BWR vessel shields

Energy Technology Data Exchange (ETDEWEB)

Alberdi, R.; Gonzalez, E.

2001-07-01

In the last few years, the owners of BWR nuclear power plants have been forced to address new fuel shield inspection requirements, TECNATOM has responded to this situation by launching the TEIDE projects, which include development of an inspection machine and the corresponding Non-Destructive Tests to examine the inside of this shield. With these projects, TECNATOM has performed more than 12 fuel shield inspections in different countries. This article describes the experience gained in the last three years. (Author)

BWR type reactors

International Nuclear Information System (INIS)

Watanabe, Shoichi

1986-01-01

Purpose: To enable to remove water not by way of mechanical operation in a reactor core and improve the fuel economy in BWR type reactors. Constitution: A hollow water removing rod of a cross-like profile made of material having a smaller neutron absorption cross section than the moderator is disposed to the water gap for each of unit structures composed of four fuel assemblies, and water is charged and discharged to and from the water removing rod. Water is removed from the water removing rod to decrease the moderators in the water gap to carry out neutron spectrum shift operation from the initial to the medium stage of reactor core cycles. At the final stage of the cycle, airs in the water removing rod are extracted and the moderator is introduced. The moderator is filled and the criticality is maintained with the accumulated nuclear fission materials. The neutron spectrum shift operation can be attained by eliminating hydrothermodynamic instability and using a water removing rod of a simple structure. (Horiuchi, T.)

Solutions to commercializing metal hydride hydrogen storage products

International Nuclear Information System (INIS)

Tomlinson, J.J.; Belanger, R.

2004-01-01

'Full text:' Whilst the concept of a Hydrogen economy in the broad sense may for some analysts and Fuel Cell technology developers be an ever moving target the use of hydrogen exists and is growing in other markets today. The use of hydrogen is increasing. Who are the users? What are their unique needs? How can they better be served? As the use of hydrogen increases there are things we can do to improve the perception and handling of hydrogen as an industrial gas that will impact the future issues of hydrogen as a fuel thereby assisting the mainstream availability of hydrogen fuel a reality. Factors that will induce change in the way hydrogen is used, handled, transported and stored are the factors to concentrate development efforts on. Other factors include: cost; availability; safety; codes and standards; and regulatory authorities acceptance of new codes and standards. New methods of storage and new devices in which the hydrogen is stored will influence and bring about change and increased use. New innovative products based on Metal Hydride hydrogen storage will address some of the barriers to widely distributed hydrogen as a fuel or energy carrier to which successful fuel cell product commercialization is subject. Palcan has developed innovative products based on it's Rare Earth Metal Hydride alloy. Some of these innovations will aid the distribution of hydrogen as a fuel and offer alternatives to the existing hydrogen user and to the Fuel Cell product developer. An overview of the products and how these products will affect the distribution and use of hydrogen as an industrial gas and fuel is presented. (author)

Study on the scattering law and scattering kernel of hydrogen in zirconium hydride

International Nuclear Information System (INIS)

Jiang Xinbiao; Chen Wei; Chen Da; Yin Banghua; Xie Zhongsheng

1999-01-01

The nuclear analytical model of calculating scattering law and scattering kernel for the uranium zirconium hybrid reactor is described. In the light of the acoustic and optic model of zirconium hydride, its frequency distribution function f(ω) is given and the scattering law of hydrogen in zirconium hydride is obtained by GASKET. The scattering kernel σ l (E 0 →E) of hydrogen bound in zirconium hydride is provided by the SMP code in the standard WIMS cross section library. Along with this library, WIMS is used to calculate the thermal neutron energy spectrum of fuel cell. The results are satisfied

BWR normal water chemistry guidelines: 1986 revision

International Nuclear Information System (INIS)

1988-09-01

Boiling water reactors (BWRs) have experienced stress corrosion cracking in the reactor cooling system piping resulting in adverse impacts on plant availability and personnel radiation exposure. The BWR Owners Group and EPRI have sponsored a major research and development program to provide remedies for this stress corrosion cracking problem. This work shows that the likelihood of cracking depends on the plant's water chemistry performance (environment) as well as on material condition and stress level. Plant experience and other research demonstrate that water quality also affects fuel performance and radiation field buildup in BWRs. This report,''BWR Normal Water Chemistry Guidelines: 1986 Revision,'' presents suggested generic water chemistry specifications, justifies the proposed water chemistry limits, suggests responses to out-of-specification water chemistry, discusses available chemical analysis methods as well as data management and surveillance schemes, and details the management philosophy required to successfully implement a water chemistry control program. An appendix contains recommendations for water quality of auxiliary systems. 73 refs., 20 figs., 9 tabs

BWR radiation buildup control with ionic zinc

International Nuclear Information System (INIS)

Marble, W.J.; Wood, C.J.; Leighty, C.E.; Green, T.A.

1986-01-01

In 1983 a hypothesis was disclosed which suggested that the presence of ionic zinc in the reactor water of the BWR could reduce radiation buildup. This hypothesis was developed from correlations of plant data, and subsequently, from laboratory experiments which demonstrated clearly that ionic zinc inhibits the corrosion of stainless steel. The benefits of zinc addition have been measured at the Vallecitos Nuclear Center under and EPRI/GE project. Experimentation and analyses have been performed to evaluate the impact of intentional zinc addition on the IGSCC characteristics of primary system materials and on the performance of the nuclear fuel. It has been concluded that no negative effects are expected. The author conclude that the intentional addition of ionic zinc to the BWR reactor water at a concentration of approximately 10 ppb will provide major benefits in controlling the Co-60 buildup on primary system stainless steel surfaces. The intentional addition of zinc is now a qualified technique for use in BWRs

PREDICTIVE METHODS FOR STABILITY MARGIN IN BWR

OpenAIRE

MELARA SAN ROMÁN, JOSÉ

2016-01-01

[EN] Power and flow oscillations in a BWR are very undesirable. One of the major concerns is to ensure, during power oscillations, compliance with GDC 10 and 12. GDC 10 requires that the reactor core be designed with appropriate margin to assure that specified acceptable fuel design limits will not be exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. GDC 12 requires assurance that power oscillations which can result in conditions ...

Assessment of boiling transition analysis code against data from NUPEC BWR full-size fine-mesh bundle tests

International Nuclear Information System (INIS)

Utsuno, Hideaki; Ishida, Naoyuki; Masuhara, Yasuhiro; Kasahara, Fumio

2004-01-01

Transient BT analysis code TCAPE based on mechanistic methods coupled with subchannel analysis has been developed for the evaluation on fuel integrity under abnormal operations in BWR. TCAPE consisted mainly of the drift-flux model, the cross-flow model, the film model and the heat transfer model. Assessment of TCAPE has been performed against data from BWR full-size fine-mesh bundle tests (BFBT), which consisted of two major parts: the void distribution measurement and the critical power measurement. Code and data comparison was made for void distributions with varying number of unheated rods in simulated actual fuel assembly. Prediction of steady-state critical power was compared with the measurement on full-scale bundle under a range of BWR operational conditions. Although the cross-sectional averaged void fraction was underestimated when it became lower, the accuracy was obtained that the averaged ratio 0.910 and its standard deviation 0.076. The prediction of steady-state critical power agreed well with the data in the range of BWR operations, where the prediction accuracy was obtained that the averaged ratio 0.997 and its standard deviation 0.043. These results demonstrated that TCAPE is well capable to predict two-phase flow distribution and liquid film dryout phenomena occurring in BWR rod bundles. Part of NUPEC BFBT database will be made available for an international benchmark exercise. The code assessment shall be continued against the OECD/NRC benchmark based on BFBT database. (author)

MCTP, a code for the thermo-mechanical analysis of a fuel rod of BWR type reactors (Neutron part); MCTP, un codigo para el analisis termo-mecanico de una barra combustible de reactores tipo BWR (Parte Neutronica)

Energy Technology Data Exchange (ETDEWEB)

Hernandez L, H; Ortiz V, J [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

2003-07-01

In the National Institute of Nuclear Research of Mexico a code for the thermo-mechanical analysis of the fuel rods of the BWR type reactors of the Nucleo electric Central of Laguna Verde is developed. The code solves the diffusion equation in cylindrical coordinates with several energy groups. The code, likewise, calculates the temperature distribution and power distribution in those fuel rods. The code is denominated Multi groups With Temperatures and Power (MCTP). In the code, the energy with which the fission neutrons are emitted it is divided in six groups. They are also considered the produced perturbations by the changes in the temperatures of the materials that constitute the fuel rods, the content of fission products, the uranium consumption and in its case the gadolinium, as well as the plutonium production. In this work there are present preliminary results obtained with the code, using data of operation of the Nucleo electric Central of Laguna Verde. (Author)

Experimental and numerical investigations of BWR fuel bundle inlet flow

International Nuclear Information System (INIS)

Hoashi, E; Morooka, S; Ishitori, T; Komita, H; Endo, T; Honda, H; Yamamoto, T; Kato, T; Kawamura, S

2009-01-01

We have been studying the mechanism of the flow pattern near the fuel bundle inlet of BWR using both flow visualization test and computational fluid dynamics (CFD) simulation. In the visualization test, both single- and multi-bundle test sections were used. The former test section includes only a corner orifice facing two support beams and the latter simulates 16 bundles surrounded by four beams. An observation window is set on the side of the walls imitating the support beams upstream of the orifices in both test sections. In the CFD simulation, as well as the visualization test, the single-bundle model is composed of one bundle with a corner orifice and the multi-bundle model is a 1/4 cut of the test section that includes 4 bundles with the following four orifices: a corner orifice facing the corner of the two neighboring support beams, a center orifice at the opposite side from the corner orifice, and two side orifices. Twin-vortices were observed just upstream of the corner orifice in the multi-bundle test as well as the single-bundle test. A single-vortex and a vortex filament were observed at the side orifice inlet and no vortex was observed at the center orifice. These flow patterns were also predicted in the CFD simulation using Reynolds Stress Model as a turbulent model and the results were in good agreement with the test results mentioned above. (author)

International comparison calculations for a BWR lattice with adjacent gadolinium pins

International Nuclear Information System (INIS)

Maeder, C.; Wydler, P.

1984-09-01

The results of burnup calculations for a simplified BWR fuel element with two adjacent gadolinium rods are presented and discussed. Ten complete solutions were contributed by Denmark, France, Italy (3), Japan (3), Switzerland and the UK. Partial results obtained from Poland and the USA are included in an Appendix. (Auth.)

Predicting Hydride Donor Strength via Quantum Chemical Calculations of Hydride Transfer Activation Free Energy.

Science.gov (United States)

Alherz, Abdulaziz; Lim, Chern-Hooi; Hynes, James T; Musgrave, Charles B

2018-01-25

We propose a method to approximate the kinetic properties of hydride donor species by relating the nucleophilicity (N) of a hydride to the activation free energy ΔG ⧧ of its corresponding hydride transfer reaction. N is a kinetic parameter related to the hydride transfer rate constant that quantifies a nucleophilic hydridic species' tendency to donate. Our method estimates N using quantum chemical calculations to compute ΔG ⧧ for hydride transfers from hydride donors to CO 2 in solution. A linear correlation for each class of hydrides is then established between experimentally determined N values and the computationally predicted ΔG ⧧ ; this relationship can then be used to predict nucleophilicity for different hydride donors within each class. This approach is employed to determine N for four different classes of hydride donors: two organic (carbon-based and benzimidazole-based) and two inorganic (boron and silicon) hydride classes. We argue that silicon and boron hydrides are driven by the formation of the more stable Si-O or B-O bond. In contrast, the carbon-based hydrides considered herein are driven by the stability acquired upon rearomatization, a feature making these species of particular interest, because they both exhibit catalytic behavior and can be recycled.

VIM Monte Carlo versus CASMO comparisons for BWR advanced fuel designs

International Nuclear Information System (INIS)

Pallotta, A.S.; Blomquist, R.N.

1994-01-01

Eigenvalues and two-dimensional fission rate distributions computed with the CASMO-3G lattice physics code and the VIM Monte Carlo Code are compared. The cases assessed are two advanced commercial BWR pin bundle designs. Generally, the two codes show good agreement in K inf , fission rate distributions, and control rod worths

Comparative analysis of results between CASMO, MCNP and Serpent for a suite of Benchmark problems on BWR reactors

International Nuclear Information System (INIS)

Xolocostli M, J. V.; Vargas E, S.; Gomez T, A. M.; Reyes F, M. del C.; Del Valle G, E.

2014-10-01

In this paper a comparison is made in analyzing the suite of Benchmark problems for reactors type BWR between CASMO-4, MCNP6 and Serpent code. The Benchmark problem consists of two different geometries: a fuel cell of a pin and assembly type BWR. To facilitate the study of reactors physics in the fuel pin their nuclear characteristics are provided to detail, such as burnt dependence, the reactivity of selected nuclide, etc. With respect to the fuel assembly, the presented results are regarding to infinite multiplication factor for burning different steps and different vacuum conditions. Making the analysis of this set of Benchmark problems provides comprehensive test problems for the next fuels generation of BWR reactors with high extended burned. It is important to note that when making this comparison the purpose is to validate the methodologies used in modeling for different operating conditions, if the case is of other BWR assembly. The results will be within a range with some uncertainty, considering that does not depend on code that is used. Escuela Superior de Fisica y Matematicas of Instituto Politecnico Nacional (IPN (Mexico) has accumulated some experience in using Serpent, due to the potential of this code over other commercial codes such as CASMO and MCNP. The obtained results for the infinite multiplication factor are encouraging and motivate the studies to continue with the generation of the X S of a core to a next step a respective nuclear data library is constructed and this can be used by codes developed as part of the development project of the Mexican Analysis Platform of Nuclear Reactors AZTLAN. (Author)

Compact PEM fuel cell system combined with all-in-one hydrogen generator using chemical hydride as a hydrogen source

International Nuclear Information System (INIS)

Kim, Jincheol; Kim, Taegyu

2015-01-01

Highlights: • Compact fuel cell system was developed for a portable power generator. • Novel concept using an all-in-one reactor for hydrogen generation was proposed. • Catalytic reactor, hydrogen chamber and separator were combined in a volume. • The system can be used to drive fuel cell-powered unmanned autonomous systems. - Abstract: Compact fuel cell system was developed for a portable power generator. The power generator features a polymer electrolyte membrane fuel cell (PEMFC) using a chemical hydride as a hydrogen source. The hydrogen generator extracted hydrogen using a catalytic hydrolysis from a sodium borohydride alkaline solution. A novel concept using an all-in-one reactor was proposed in which a catalyst, hydrogen chamber and byproduct separator were combined in a volume. In addition, the reactor as well as a pump, cooling fans, valves and controller was integrated in a single module. A 100 W PEMFC stack was connected with the hydrogen generator and was evaluated at various load conditions. It was verified that the stable hydrogen supply was achieved and the developed system can be used to drive fuel cell-powered unmanned autonomous systems.

Post-Irradiation Examinations for Resolving Fuel Issues in Long Term Storage

International Nuclear Information System (INIS)

Karlsson, Joakim K.H.; Alvarez Holston, Anna-Maria

2014-01-01

In many countries extended long term dry storage is the solution for storage of spent nuclear fuel for the foreseeable future. The expected storage times have increased over the last years and today storage times of up to 300 years is anticipated. With such long storage times, requirements on transportability and retrievability of the fuel have become more important. Hitherto most investigations on fuel behaviour during dry storage have been focused on cladding creep and the impact of hydrogen and hydrides in the cladding. Creep data gives input to creep models and creep to rupture data helps to set criteria for maximum allowable internal rod pressure. Hydrides lower the ductility of the cladding and this is more pronounced with radially oriented hydrides. As the temperature decreases over time in a dry storage cask dissolved hydrogen will precipitate forming hydrides in addition to hydrides already present. Assuming there is sufficient hoop stress in the cladding, the new hydrides would be radially oriented. Together with lost ductility Delayed Hydride Cracking (DHC) could be a potential mechanism for rod failure over tens of years of dry storage as the temperature drops from about 350 deg. C to 150 deg. C. Hydride embrittlement and the DHC mechanism have been studied in the first Studsvik Cladding Integrity Project (SCIP), although the focus in this program has mainly been on higher temperatures relevant for operating conditions rather than on dry storage conditions. In addition to the mechanisms mentioned there are other failure mechanisms that could potentially threaten the cladding fuel integrity and retrievability. In case there is residual water or moisture available in the cask, or even in the fuel due to existing fuel failures, radiolysis gives free hydrogen and oxygen. In failed fuel this may cause fuel oxidation and swelling affecting fuel integrity. The hydrogen gas pressure will not threaten the cask but be available for cladding uptake. Furthermore

TVA experience in BWR reload design and licensing

International Nuclear Information System (INIS)

Robertson, J.D.

1986-01-01

TVA has developed and implemented the capability to perform BWR reload core design and licensing analyses. The advantages accruing from this capability include the tangible cost-savings from performing reload analyses in-house. Also, ''intangible'' benefits such as increased operating flexibility and the ability to accommodate multivendor fuel designs have been demonstrated. The major disadvantage with performing in-house analyses is the cost associated with development and maintenance of the analytical methods and staff expertise

Plutonium isotopic composition of high burnup spent fuel discharged from light water reactors

International Nuclear Information System (INIS)

Nakano, Yoshihiro; Okubo, Tsutomu

2011-01-01

Highlights: → Pu isotopic composition of fuel affects FBR core nuclear characteristics very much. → Spent fuel compositions of next generation LWRs with burnup of 70 GWd/t were obtained. → Pu isotopic composition and amount in the spent fuel with 70 GWd/t were evaluated. → Spectral shift rods of high burnup BWR increases the fissile Pu fraction of spent fuel. → Wide fuel rod pitch of high burnup PWR lowers the fissile Pu fraction of spent fuel. - Abstract: The isotopic composition and amount of plutonium (Pu) in spent fuel from a high burnup boiling water reactor (HB-BWR) and a high burnup pressurized water reactor (HB-PWR), each with an average discharge burnup of 70 GWd/t, were estimated, in order to evaluate fast breeder reactor (FBR) fuel composition in the transition period from LWRs to FBRs. The HB-BWR employs spectral shift rods and the neutron spectrum is shifted through the operation cycle. The weight fraction of fissile plutonium (Puf) isotopes to the total plutonium in HB-BWR spent fuel after 5 years cooling is 62%, which is larger than that of conventional BWRs with average burnup of 45 GWd/t, because of the spectral shift operation. The amount of Pu produced in the HB-BWR is also larger than that produced in a conventional BWR. The HB-PWR uses a wider pitch 17 x 17 fuel rod assembly to optimize neutron slowing down. The Puf fraction of HB-PWR spent fuel after 5 years cooling is 56%, which is smaller than that of conventional PWRs with average burnup of 49 GWd/t, mainly because of the wider pitch. The amount of Pu produced in the HB-PWR is also smaller than that in conventional PWRs.

Hydrogen transmission/storage with a metal hydride/organic slurry

Energy Technology Data Exchange (ETDEWEB)

Breault, R.W.; Rolfe, J.; McClaine, A. [Thermo Power Corp., Waltham, MA (United States)

1998-08-01

Thermo Power Corporation has developed a new approach for the production, transmission, and storage of hydrogen. In this approach, a chemical hydride slurry is used as the hydrogen carrier and storage media. The slurry protects the hydride from unanticipated contact with moisture in the air and makes the hydride pumpable. At the point of storage and use, a chemical hydride/water reaction is used to produce high-purity hydrogen. An essential feature of this approach is the recovery and recycle of the spent hydride at centralized processing plants, resulting in an overall low cost for hydrogen. This approach has two clear benefits: it greatly improves energy transmission and storage characteristics of hydrogen as a fuel, and it produces the hydrogen carrier efficiently and economically from a low cost carbon source. The preliminary economic analysis of the process indicates that hydrogen can be produced for $3.85 per million Btu based on a carbon cost of $1.42 per million Btu and a plant sized to serve a million cars per day. This compares to current costs of approximately $9.00 per million Btu to produce hydrogen from $3.00 per million Btu natural gas, and $25 per million Btu to produce hydrogen by electrolysis from $0.05 per Kwh electricity. The present standard for production of hydrogen from renewable energy is photovoltaic-electrolysis at $100 to $150 per million Btu.

Methyl Iodide Decomposition at BWR Conditions

International Nuclear Information System (INIS)

Pop, Mike; Bell, Merl

2012-09-01

Based on favourable results from short-term testing of methanol addition to an operating BWR plant, AREVA has performed numerous studies in support of necessary Engineering and Plant Safety Evaluations prior to extended injection of methanol. The current paper presents data from a study intended to provide further understanding of the decomposition of methyl iodide as it affects the assessment of methyl iodide formation with the application of methanol at BWR Plants. This paper describes the results of the decomposition testing under UV-C light at laboratory conditions and its effect on the subject methyl iodide production evaluation. The study as to the formation and decomposition of methyl iodide as it is effected by methanol addition is one phase of a larger AREVA effort to provide a generic plant Safety Evaluation prior to long-term methanol injection to an operating BWR. Other testing phases have investigated the compatibility of methanol with fuel construction materials, plant structural materials, plant consumable materials (i.e. elastomers and coatings), and ion exchange resins. Methyl iodide is known to be very unstable, typically preserved with copper metal or other stabilizing materials when produced and stored. It is even more unstable when exposed to light, heat, radiation, and water. Additionally, it is known that methyl iodide will decompose radiolytically, and that this effect may be simulated using ultra-violet radiation (UV-C) [2]. In the tests described in this paper, the use of a UV-C light source provides activation energy for the formation of methyl iodide. Thus is similar to the effect expected from Cherenkov radiation present in a reactor core after shutdown. Based on the testing described in this paper, it is concluded that injection of methanol at concentrations below 2.5 ppm in BWR applications to mitigate IGSCC of internals is inconsequential to the accident conditions postulated in the FSAR as they are related to methyl iodide formation

Impact analysis of modifying the composition of the nuclear fuel of a BWR with beryllium oxide; Analisis del impacto de modificar la composicion del combustible nuclear de un BWR con oxido de berilio

Energy Technology Data Exchange (ETDEWEB)

Gallardo V, J. M.; Morales S, J. B., E-mail: euqrop@hotmail.com [UNAM, Facultad de Ingenieria, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

2013-10-15

The beryllium oxide (Be O) presents excellent physical properties, especially its high thermal conductivity that contrasts clearly with that of the uranium dioxide (UO{sub 2}) used at the present as fuel in a great number of nuclear plants. The present work models a nuclear reactor cooled by light water in boiling with two external recirculation loops (BWR/5) using the code for the transitory analysis and postulated accidents Trac-B F1, implementing a UO{sub 2} mixture and different fractions of Be O, with the objective of improving the thermal conductivity of the fuel. The numeric results and the realized analyses indicate that when adding a fraction in volume of 10% the central temperature decreases in 30.4% in stationary state, while during the large break loss of coolant accident the peak cladding temperature diminishes in 7%. Although the real interaction of the mixture has not been determined experimentally, the obtained results are promising. (Author)

The Oxidation Products of Aluminum Hydride and Boron Aluminum Hydride Clusters

Science.gov (United States)

2016-01-04

AFRL-AFOSR-VA-TR-2016-0075 The Oxidation Products of Aluminum Hydride and Boron Aluminum Hydride Clusters KIT BOWEN JOHNS HOPKINS UNIV BALTIMORE MD...2. REPORT TYPE Final Performance 3. DATES COVERED (From - To) 30-09-2014 to 29-09-2015 4. TITLE AND SUBTITLE The Oxidation Products of Aluminum ...Hydride and Boron Aluminum Hydride Clusters 5a.  CONTRACT NUMBER 5b.  GRANT NUMBER FA9550-14-1-0324 5c.  PROGRAM ELEMENT NUMBER 61102F 6. AUTHOR(S) KIT

Benchmark problem suite for reactor physics study of LWR next generation fuels

International Nuclear Information System (INIS)

Yamamoto, Akio; Ikehara, Tadashi; Ito, Takuya; Saji, Etsuro

2002-01-01

This paper proposes a benchmark problem suite for studying the physics of next-generation fuels of light water reactors. The target discharge burnup of the next-generation fuel was set to 70 GWd/t considering the increasing trend in discharge burnup of light water reactor fuels. The UO 2 and MOX fuels are included in the benchmark specifications. The benchmark problem consists of three different geometries: fuel pin cell, PWR fuel assembly and BWR fuel assembly. In the pin cell problem, detailed nuclear characteristics such as burnup dependence of nuclide-wise reactivity were included in the required calculation results to facilitate the study of reactor physics. In the assembly benchmark problems, important parameters for in-core fuel management such as local peaking factors and reactivity coefficients were included in the required results. The benchmark problems provide comprehensive test problems for next-generation light water reactor fuels with extended high burnup. Furthermore, since the pin cell, the PWR assembly and the BWR assembly problems are independent, analyses of the entire benchmark suite is not necessary: e.g., the set of pin cell and PWR fuel assembly problems will be suitable for those in charge of PWR in-core fuel management, and the set of pin cell and BWR fuel assembly problems for those in charge of BWR in-core fuel management. (author)

Zirconium-hydride solid zero power reactor and its application research

International Nuclear Information System (INIS)

Lin Shenghuo; Luo Zhanglin; Su Zhuting

1994-10-01

The Zirconium Hydride Solid Zero Power Reactor built at China Institute of Atomic Energy is introduced. In the reactor Zirconium-hydride is used as moderator, plexiglass as reflector and U 3 O 8 with enrichment of 20% as the fuel, Since its initial criticality, the physical characteristics and safety features have been measured with the result showing that the reactor has sound stability and high sensitivity, etc. It has been successfully used for the personnel training and for the testing of reactor control instruments and experiment devices. It also presents the special advantage for the pre-research of some applications

ORCOST-2, PWR, BWR, HTGR, Fossil Fuel Power Plant Cost and Economics

International Nuclear Information System (INIS)

Fuller, L.C.; Myers, M.L.

1975-01-01

1 - Description of problem or function: ORCOST2 estimates the cost of electrical energy production from single-unit steam-electric power plants. Capital costs and operating and maintenance costs are calculated using base cost models which are included in the program for each of the following types of plants: PWR, BWR, HTGR, coal, oil, and gas. The user may select one of several input/output options for calculation of capital cost, operating and maintenance cost, levelized energy costs, fixed charge rate, annual cash flows, cumulative cash flows, and cumulative discounted cash flows. Options include the input of capital cost and/or fixed charge rate to override the normal calculations. Transmission and distribution costs are not included. Fuel costs must be input by the user. 2 - Method of solution: The code follows the guidelines of AEC Report NUS-531. A base capital-cost model and a base operating- and maintenance-cost model are selected and adjusted for desired size, location, date, etc. Costs are discounted to the year of first commercial operation and levelized to provide annual cost of electric power generation. 3 - Restrictions on the complexity of the problem: The capital cost models are of doubtful validity outside the 500 to 1500 MW(e) range

A preliminary evaluation of the ability of from-reactor casks to geometrically accommodate commercial LWR spent nuclear fuel

International Nuclear Information System (INIS)

Andress, D.; McLeod, N.B.; Rahimi, M.; Joy, D.S.; Peterson, R.W.

1991-01-01

The DOE has sponsored a number of cask design efforts to define several transportation casks to accommodate the various assemblies expected to be accepted by the Federal Waste Management System. At this time, three preliminary cask designs have been selected for the final design - the GA-4 and GA-9 truck casks and the BR-100 rail cask. The GA-4 cask is designed for PWR fuel only; the GA-9 cask is a longer cask with less shielding designed for BWR fuel only; and the BR-100 cask is designed to accommodate both PWR and BWR fuels. In total, this assessment indicates that the current Initiative I cask designs can be expected to dimensionally accommodate 100% of the PWR fuel assemblies (other than the extra-long South Texas Fuel) with control elements removed, and >90% of the assemblies having the control elements as an integral part of the fuel assembly. For BWR assemblies, >99% of the assemblies can be accommodated with fuel channels removed. Because of the button and spring interference, the basket openings in these casks will not accommodate assemblies in the BWR/2,3 and BWR/4-6 fuel classes with the fuel channels in place

Oxidation kinetics of hydride-bearing uranium metal corrosion products

International Nuclear Information System (INIS)

Totemeier, T.C.; Pahl, R.G.; Frank, S.M.

1998-01-01

The oxidation behavior of hydride-bearing uranium metal corrosion products from zero power physics reactor (ZPPR) fuel plates was studied using thermo-gravimetric analysis (TGA) in environments of Ar-4%O 2 , Ar-9%O 2 , and Ar-20%O 2 . Ignition of corrosion product samples from two moderately corroded plates was observed between 125 C and 150 C in all environments. The rate of oxidation above the ignition temperature was found to be dependent only on the net flow rate of oxygen in the reacting gas. Due to the higher net oxygen flow rate, burning rates increased with increasing oxygen concentration. Oxidation rates below the ignition temperature were much slower and decreased with increasing test time. The hydride contents of the TGA samples from the two moderately corroded plates, determined from the total weight gain achieved during burning, were 47-61 wt% and 29-39 wt%. Samples from a lightly corroded plate were not reactive; X-ray diffraction (XRD) confirmed that they contained little hydride. (orig.)

Fuel assembly for BWR type reactor

International Nuclear Information System (INIS)

Kato, Shigeru.

1993-01-01

In the fuel assembly of the present invention, a means for mounting and securing short fuel rods is improved. Not only long fuel rods but also short fuel rods are disposed in channel of the fuel assembly to improve reactor safety. The short fuel rods are supported by a screw means only at the lower end plug. The present invention prevents the support for the short fuel rod from being unreliable due to the slack of the screw by the pressure of inflowing coolants. That is, coolant abutting portions such as protrusions or concave grooves are disposed at a portion in the channel box where coolants flowing from the lower tie plate, as an uprising stream, cause collision. With such a constitution, a component caused by the pressure of the flowing coolants is formed. The component acts as a rotational moment in the direction of screwing the male threads of the short fuel rod into the end plug screw hole. Accordingly, the screw is not slackened, and the short fuel rods are mounted and secured certainly. (I.S.)

Experimental validation of CASMO-4E and CASMO-5M for radial fission rate distributions in a westinghouse SVEA-96 Optima2 BWR fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Grimm, P.; Perret, G. [Paul Scherrer Inst., CH-5232 Villigen PSI (Switzerland)

2012-07-01

Measured and calculated radial total fission rate distributions are compared for the three axial sections of a Westinghouse SVEA-96 Optima2 BWR fuel assembly, comprising 96, 92 and 84 fuel rods, respectively. The measurements were performed on a full-size fuel assembly in the PROTEUS zero-power experimental facility. The measured fission rates are compared to the results of the CASMO-4E and CASMO-5M fuel assembly codes. Detailed measured geometrical data were used in the models, and effects of the surrounding zones of the reactor were taken into account by correction factors derived from MCNPX calculations. The results of the calculations agree well with those of the experiments, with root-mean-square deviations between 1.2% and 1.5% and maximum deviations of 3-4%. The quality of the predictions by CASMO-4E and CASMO-5M is comparable. (authors)

Accommodation stresses in hydride precipitates by synchrotron x-ray diffraction

International Nuclear Information System (INIS)

Santisteban, J R; Vicente, M A; Vizcaino, P; Banchik, A D; Almer, J

2012-01-01

Hydride-forming materials (Zr, Ti, Nb, etc) are affected by a sub-critical crack growth mechanism that involves the diffusion of H to the stressed region ahead of a crack, followed by nucleation and fracture of hydrides at the crack tip [1]. The phenomenon is intermittent, with the crack propagating through the hydride and stopping when it reaches the matrix. By repeating these processes, the crack propagates through a component at a rate that is highly dependent on the temperature history of the component. Most research effort to understand this phenomenon has occurred within the nuclear industry, as it affects the safe operation of pressure tubes (Zr2.5%Nb) and the long-term storage of nuclear fuel (Zircaloy cladding). Stress-induced hydride formation is a consequence of the volume dilatation that accompanies hydride formation (of the order of 15%), which is elastoplastically accommodated by the matrix and precipitate. Compressive stresses are expected within hydride precipitates due to the constraint imposed by the matrix. Such 'accommodation' stresses are essential ingredients in all theoretical models developed to assess the crack growth rate dependence on operational variables such as temperature, applied stress intensity factor, or overall H concentration [2]. Yet little experimental information is available about the magnitude and directionality of such accommodation stresses. Synchrotron X-ray diffraction is the only technique capable of quantifying such stresses. Here we briefly describe the fundaments of the technique, when used through an area detector placed in transmission geometry. The results of the experiments have allowed us to produce a comprehensive picture about the magnitude and origin of accommodation stresses in δ zirconium hydride platelets (author)

Evaluation on transmutation performance of minor actinides with high-flux BWR

International Nuclear Information System (INIS)

Setiawan, M.B.; Kitamoto, A.; Taniguchi, A.

2001-01-01

The performance of high-flux BWR (HFBWR) for burning and/or transmutation (B/T) treatment of minor actinides (MA) and long-lived fission products (LLFP) was discussed herein for estimating an advanced waste disposal with partitioning and transmutation (P and T). The concept of high-flux B/T reactor was based on a current 33 GWt-BWR, to transmute the mass of long-lived transuranium (TRU) to short-lived fission products (SLFP). The nuclide selected for B/T treatment was MA (Np-237, Am-241, and Am-243) included in the discharged fuel of LWR. The performance of B/T treatment of MA was evaluated by a new function, i.e. [F/T ratio], defined by the ratio of the fission rate to the transmutation rate in the core, at an arbitrary burn-up, due to all MA nuclides. According to the results, HFBWR could burn and/or transmute MA nuclides with higher fission rate than BWR, but the fission rate did not increase proportionally to the flux increment, due to the higher rate of neutron adsorption. The higher B/T fraction of MA would result in the higher B/T capacity, and will reduce the units of HFBWR needed for the treatment of a constant mass of MA. In addition, HFBWR had a merit of higher mass transmutation compared to the reference BWR, under the same mass loading of MA

High-pressure hydriding of Zircaloy

International Nuclear Information System (INIS)

Kim, Y.S.

1996-01-01

The hydriding characteristics of Zircaloy-2(Zry), sponge zirconium (as a liner on Zry plate), and crystal-bar zirconium exposed to pure H 2 at 0.1 MPa or 7 MPa and 400 C were determined in a thermogravimetric apparatus. The morphology of the hydrided specimens was also examined by optical microscopy. For all specimen types, the rate of hydriding in 7 MPa H 2 was two orders of magnitude greater than in 0.1 MPa H 2 . For Zry, uniform bulk hydriding was revealed by hydride precipitates at room temperature and on one occasion, a sunburst hydride. In addition, all specimen types exhibited a hydride surface layer. In a duplex Zry/sponge-Zr specimen, Zry is more heavily hydrided than the sponge Zr layer. (orig.)

Design of a mixed recharge with MOX assemblies of greater relation of moderation for a BWR reactor

International Nuclear Information System (INIS)

Ramirez S, J.R.; Alonso V, G.; Palacios H, J.

2004-01-01

The study of the fuel of mixed oxides of uranium and plutonium (MOX) it has been topic of investigation in many countries of the world and those are even discussed in many places the benefits of reprocessing the spent fuel to extract the plutonium created during the irradiation of the fuel in the nuclear power reactors. At the moment those reactors that have been loaded partially with MOX fuel, are mainly of the type PWR where a mature technology has been achieved in some countries like they are France, Belgium and England, however the experience with reactors of the type BWR is more limited and it is continued studying the best way to introduce this type of fuel in BWRs, one of the main problems to introduce MOX in reactors BWR is the neutronic design of the same one, existing different concepts to introduce the plutonium in the assemblies of fuel and one of them is the one of increasing the relationship of moderation of the assemble. In this work a MOX fuel assemble design is presented and the obtained results so far in the ININ. These results indicate that the investigated concept has some exploitable advantages in the use of the MOX fuel. (Author)

The BWR Stability Issue

International Nuclear Information System (INIS)

D'Auria, F.

2008-01-01

The purpose of this paper is to supply general information about Boiling Water Reactor (BWR) stability. The main concerned topics are: phenomenological aspects, experimental database, modelling features and capabilities, numerical models, three-dimensional modelling, BWR system performance during stability, stability monitoring and licensing aspects.

Hydrogen generation using silicon nanoparticles and their mixtures with alkali metal hydrides

Science.gov (United States)

Patki, Gauri Dilip

Hydrogen is a promising energy carrier, for use in fuel cells, engines, and turbines for transportation or mobile applications. Hydrogen is desirable as an energy carrier, because its oxidation by air releases substantial energy (thermally or electrochemically) and produces only water as a product. In contrast, hydrocarbon energy carriers inevitably produce CO2, contributing to global warming. While CO2 capture may prove feasible in large stationary applications, implementing it in transportation and mobile applications is a daunting challenge. Thus a zero-emission energy carrier like hydrogen is especially needed in these cases. Use of H2 as an energy carrier also brings new challenges such as safe handling of compressed hydrogen and implementation of new transport, storage, and delivery processes and infrastructure. With current storage technologies, hydrogen's energy per volume is very low compared to other automobile fuels. High density storage of compressed hydrogen requires combinations of high pressure and/or low temperature that are not very practical. An alternative for storage is use of solid light weight hydrogenous material systems which have long durability, good adsorption properties and high activity. Substantial research has been conducted on carbon materials like activated carbon, carbon nanofibers, and carbon nanotubes due to their high theoretical hydrogen capacities. However, the theoretical values have not been achieved, and hydrogen uptake capacities in these materials are below 10 wt. %. In this thesis we investigated the use of silicon for hydrogen generation. Hydrogen generation via water oxidation of silicon had been ignored due to slow reaction kinetics. We hypothesized that the hydrogen generation rate could be improved by using high surface area silicon nanoparticles. Our laser-pyrolysis-produced nanoparticles showed surprisingly rapid hydrogen generation and high hydrogen yield, exceeding the theoretical maximum of two moles of H2 per

Thermohydraulic stability coupled to the neutronic in a BWR; Estabilidad termohidraulica acoplada a la neutronica en un BWR

Energy Technology Data Exchange (ETDEWEB)

Calleros M, G.; Zapata Y, M.; Gomez H, R.A.; Mendez M, A. [Comision Federal de Electricidad, Central Nucleoelectrica de Laguna Verde, Carretera Cardel-Nautla Km. 42.5, Mpio. Alto Lucero, Veracruz (Mexico); Castlllo D, R. [ININ, Carretera Mexico-Toluca Km 36.5, La Marquesa, Estado de Mexico (Mexico)]. e-mail: gcm9acpp@cfe.gob.mx

2006-07-01

In a BWR type reactor the phenomenon of the nuclear fission is presented, in which are liberated in stochastic form neutrons, originating that the population of the same ones varies in statistic form around a mean value. This variation will cause that when the neutron flow impacts on the neutron detectors, its are had as a result neutron flow signals with fluctuations around an average value. In this article it is shown that it conforms it lapses the time, this variations in the neutron flow (and therefore, in the flow signal due only to the fission), they presented oscillations inside a stable range, which won't be divergent. Considering that the BWR is characterized because boiling phenomena are presented, which affect the moderation of the neutrons, additional variations will be had in the signal coming from the neutron detectors, with relationship to the fission itself, which will be influenced by the feedback of the moderator's reactivity and of the temperature of the fuel pellet. Also, as the BWR it has coupled control systems to maintain the coolant level one and of the thermal power of the reactor, for each control action it was affected the neutron population. This means that the reactor could end up straying of a stable state condition. By it previously described, the study of the thermohydraulic stability coupled to the neutronic is complex. In this work it is shown the phenomenology, the mathematical models and the theoretical behavior associated to the stability of the BWR type reactor; the variables that affect it are identified, the models that reproduce the behavior of the thermohydraulic stability coupled to the neutronic, the way to maintain stable the reactor and the instrumentation that can settle to detect and to suppress uncertainties is described. In particular, is make reference to the evolution of the methods to maintain the stability of the reactor and the detection system and suppression of uncertainties implemented in the

Development of a BWR loading pattern design system based on modified genetic algorithms and knowledge

International Nuclear Information System (INIS)

Martin-del-Campo, Cecilia; Francois, Juan Luis; Avendano, Linda; Gonzalez, Mario

2004-01-01

An optimization system based on Genetic Algorithms (GAs), in combination with expert knowledge coded in heuristics rules, was developed for the design of optimized boiling water reactor (BWR) fuel loading patterns. The system was coded in a computer program named Loading Pattern Optimization System based on Genetic Algorithms, in which the optimization code uses GAs to select candidate solutions, and the core simulator code CM-PRESTO to evaluate them. A multi-objective function was built to maximize the cycle energy length while satisfying power and reactivity constraints used as BWR design parameters. Heuristic rules were applied to satisfy standard fuel management recommendations as the Control Cell Core and Low Leakage loading strategies, and octant symmetry. To test the system performance, an optimized cycle was designed and compared against an actual operating cycle of Laguna Verde Nuclear Power Plant, Unit I

Design and integration of a hydrogen storage on metallic hydrides

International Nuclear Information System (INIS)

Botzung, M.

2008-01-01

This work presents a hydrogen storage system using metal hydrides for a Combined Heat and Power (CHP) system. Hydride storage technology has been chosen due to project specifications: high volumetric capacity, low pressures (≤ 3.5 bar) and low temperatures (≤ 75 C: fuel cell temperature). During absorption, heat from hydride generation is dissipated by fluid circulation. An integrated plate-fin type heat exchanger has been designed to obtain good compactness and to reach high absorption/desorption rates. At first, the storage system has been tested in accordance with project specifications (absorption 3.5 bar, desorption 1.5 bar). Then, the hydrogen charge/discharge times have been decreased to reach system limits. System design has been used to simulate thermal and mass comportment of the storage tank. The model is based on the software Fluent. We take in consideration heat and mass transfers in the porous media during absorption/desorption. The hydride thermal and mass behaviour has been integrated in the software. The heat and mass transfers experimentally obtained have been compared to results calculated by the model. The influence of experimental and numerical parameters on the model behaviour has also been explored. (author) [fr

Calculation of the linear heat generation rates which violate the thermomechanical limit of plastic deformation of the fuel cladding in function of the burn up of a BWR fuel rod type

International Nuclear Information System (INIS)

Lucatero, M.A.; Hernandez L, H.

2003-01-01

The linear heat generation rates (LHGR) for a BWR type generic fuel rod, as function of the burnup that violate the thermomechanical limit of circumferential plastic deformation of the can (canning) in nominal operation in stationary state of the fuel rod are calculated. The evaluation of the LHGR in function of the burnt of the fuel, is carried out under the condition that the deformation values of the circumferential plastic deformation of the can exceeds in 0.1 the thermomechanical value operation limit of 1%. The results of the calculations are compared with the generation rates of linear operation heat in function of the burnt for this fuel rod type. The calculations are carried out with the FEMAXI-V and RODBURN codes. The results show that for exhibitions or burnt between 0 and 16,000 M Wd/tU a minimum margin of 160.8 W/cm exists among LHGR (439.6 W/cm) operation peak for the given fuel and maximum LHGR of the fuel (calculated) to reach 1.1% of circumferential plastic deformation of the can, for the peak factor of power of 1.40. For burnt of 20,000 MWd/tU and 60,000 MWd/tU exist a margin of 150.3 and 298.6 W/cm, respectively. (Author)

Updating of the costs of the nuclear fuels of the equilibrium reloading of the A BWR and EPR reactors

International Nuclear Information System (INIS)

Ortega C, R.F.

2008-01-01

In the last two and a half years, the price of the uranium in the market spot has ascended of US$20.00 dollars by lb U 3O 8 in January, 2005 to a maximum of US$137.00 dollars by Ib U 3 O 8 by the middle of 2007. At the moment this price has been stabilized in US$90.00 dollars by Ib U 3 O 8 such for the market spot, like for the long term contracts. In this work the reasons of this increment are analyzed, as well as their impact in the fuel prices of the balance recharge of the advanced reactors of boiling water (A BWR) and of the advanced water at pressure reactors (EPR). (Author)

Containment venting as a mitigation technique for BWR MARK I plant ATWS

International Nuclear Information System (INIS)

Harrington, R.M.

1987-01-01

Containment venting is studied as a mitigation strategy for preventing or delaying severe fuel damage following hypothetical BWR Anticipated Transient Without Scram (ATWS) accidents initiated by MSIV-closure, and compounded by failure of the Standby Liquid Control (SLC) system injection of sodium pentaborate solution and by the failure of manually initiated control rod insertion. The venting of primary containment after reaching 75 psia (0.52 MPa) is found to result in the release of the vented steam inside the reactor building, and to result in inadequate Net Positive Suction Head (NPSH) for any system pumping from the pressure suppression pool. CONTAIN code calculations show that personnel access to large portions of the reactor building would be lost soon after the initiation of venting and that the temperatures reached would be likely to result in independent equipment failures. It is concluded that containment venting would be more likely to cause or to hasten the onset of severe fuel damage than to prevent or to delay it. Two alternative strategies that do not require containment venting, but that could delay or prevent severe fuel damage, are analyzed. BWR-LTAS code results are presented for a successful mitigation strategy in which the reactor vessel is depressurized, and for one in which the reactor vessel remains at pressure

Spent fuel disassembly hardware and other non-fuel bearing components: characterization, disposal cost estimates, and proposed repository acceptance requirements

Energy Technology Data Exchange (ETDEWEB)

Luksic, A.T.; McKee, R.W.; Daling, P.M.; Konzek, G.J.; Ludwick, J.D.; Purcell, W.L.

1986-10-01

There are two categories of waste considered in this report. The first is the spent fuel disassembly (SFD) hardware. This consists of the hardware remaining after the fuel pins have been removed from the fuel assembly. This includes end fittings, spacer grids, water rods (BWR) or guide tubes (PWR) as appropriate, and assorted springs, fasteners, etc. The second category is other non-fuel-bearing (NFB) components the DOE has agreed to accept for disposal, such as control rods, fuel channels, etc., under Appendix E of the standard utiltiy contract (10 CFR 961). It is estimated that there will be approximately 150 kg of SFD and NFB waste per average metric ton of uranium (MTU) of spent uranium. PWR fuel accounts for approximately two-thirds of the average spent-fuel mass but only 50 kg of the SFD and NFB waste, with most of that being spent fuel disassembly hardware. BWR fuel accounts for one-third of the average spent-fuel mass and the remaining 100 kg of the waste. The relatively large contribution of waste hardware in BWR fuel, will be non-fuel-bearing components, primarily consisting of the fuel channels. Chapters are devoted to a description of spent fuel disassembly hardware and non-fuel assembly components, characterization of activated components, disposal considerations (regulatory requirements, economic analysis, and projected annual waste quantities), and proposed acceptance requirements for spent fuel disassembly hardware and other non-fuel assembly components at a geologic repository. The economic analysis indicates that there is a large incentive for volume reduction.

Development of the BWR Dry Core Initial and Boundary Conditions for the SNL XR2 Experiments; TOPICAL

International Nuclear Information System (INIS)

Ott, L.J.

1994-01-01

The objectives of the Boiling Water Reactor Experimental Analysis and Model Development for Severe Accidents (BEAMD) Program at the Oak Ridge National Laboratory (ORNL) are: (1) the development of a sound quantitative understanding of boiling water reactor (BWR) core melt progression; this includes control blade and channel box effects, metallic melt relocation and possible blockage formation under severe accident conditions, and (2) provision of BWR melt progression modeling capabilities in SCDAP/RELAP5 (consistent with the BWR experimental data base). This requires the assessment of current modeling of BWR core melt progression against the expanding BWR data base. Emphasis is placed upon data from the BWR tests in the German CORA test facility and from the ex-reactor experiments[Sandia National Laboratories (SNL)] on metallic melt relocation and blockage formation in BWRs, as well as upon in-reactor data from the Annular Core Research Reactor (ACRR) DF-4 BWR test (conducted in 1986 at SNL). The BEAMD Program is a derivative of the BWR Severe Accident Technology Programs at ORNL. The ORNL BWR programs have studied postulated severe accidents in BWRs and have developed a set of models specific to boiling water reactor response under severe accident conditions. These models, in an experiment-specific format, have been successfully applied to both pretest and posttest analyses of the DF-4 experiment, and the BWR severe fuel damage (SFD) experiments performed in the CORA facility at the Kernforschungszentrum Karlsruhe (KfK) in Germany, resulting in excellent agreement between model prediction and experiment. The ORNL BWR models have provided for more precise predictions of the conditions in the BWR experiments than were previously available. This has provided a basis for more accurate interpretation of the phenomena for which the experiments are performed. The experiment-specific models, as used in the ORNL DF-4 and CORA BWR experimental analyses, also provide a basis

Ultrapure hydrogen thermal compressor based on metal hydrides for fuel cells and hybrid vehicles

International Nuclear Information System (INIS)

Almasan, V.; Biris, A.; Coldea, I.; Lupu, D.; Misan, I.; Popeneciu, G.; Ardelean, O.

2007-01-01

Full text: In hydrogen economy, efficient compressors are indispensable elements in the storage, transport and distribution of the produced hydrogen. Energetic efficient technologies can contribute to H 2 pipelines transport to the point of use and to distribute H 2 by refuelling stations. Characteristic for metal hydrides systems is the wide area of possibilities to absorb hydrogen at low pressure from any source of hydrogen, to store and deliver it hydrogen at high pressure (compression ratio more than 30). On the basis of innovative concepts and advanced materials for H 2 storage/compression (and fast thermal transfer), a fast mass (H 2 ) and heat transfer unit will be developed suitable to be integrated in a 3 stage thermal compressor. Metal hydrides used for a three stage hydrogen compression system must have different equilibrium pressures, namely: for stage 1, low pressure H 2 absorption and resistant to poisoning with impurities of hydrogen, for stage 2, medium pressure H 2 absorption and for stage 3, high pressure hydrogen delivery (120 bar). In the case of compression device based on metallic hydrides the most important properties are the hydrogen absorption/desorption rate, a smaller process enthalpy and a great structural stability on long term hydrogen absorption/desorption cycling. These properties require metal hydrides with large differences between the hydrogen absorption and desorption pressures at equilibrium, within a rather small temperature range. The main goal of this work is to search and develop metal hydride integrated systems for hydrogen purification, storage and compression. After a careful screening three hydrogen absorbing alloys will be selected. After selection, the work up of the alloys composition on the bases of detailed solid state studies, new multi-component alloys will be developed, with suitable thermodynamic and kinetic properties for a hydrogen compressor. The results of the study are the following: new types of hydrogen

Oxidation kinetics of hydride-bearing uranium metal corrosion products

Science.gov (United States)

Totemeier, Terry C.; Pahl, Robert G.; Frank, Steven M.

The oxidation behavior of hydride-bearing uranium metal corrosion products from Zero Power Physics Reactor (ZPPR) fuel plates was studied using thermo-gravimetric analysis (TGA) in environments of Ar-4%O 2, Ar-9%O 2, and Ar-20%O 2. Ignition of corrosion product samples from two moderately corroded plates was observed between 125°C and 150°C in all environments. The rate of oxidation above the ignition temperature was found to be dependent only on the net flow rate of oxygen in the reacting gas. Due to the higher net oxygen flow rate, burning rates increased with increasing oxygen concentration. Oxidation rates below the ignition temperature were much slower and decreased with increasing test time. The hydride contents of the TGA samples from the two moderately corroded plates, determined from the total weight gain achieved during burning, were 47-61 wt% and 29-39 wt%. Samples from a lightly corroded plate were not reactive; X-ray diffraction (XRD) confirmed that they contained little hydride.

Influence of temperature and hydrogen content on stress-induced radial hydride precipitation in Zircaloy-4 cladding

Energy Technology Data Exchange (ETDEWEB)

Desquines, J., E-mail: jean.desquines@irsn.fr; Drouan, D.; Billone, M.; Puls, M.P.; March, P.; Fourgeaud, S.; Getrey, C.; Elbaz, V.; Philippe, M.

2014-10-15

Radial hydride precipitation in stress relieved Zircaloy-4 fuel claddings is studied using a new thermal–mechanical test. Two maximum temperatures for radial hydride precipitation heat treatment are studied, 350 and 450 °C with hydrogen contents ranging between 50 and 600 wppm. The new test provides two main results of interest: the minimum hoop stress required to precipitate radial hydrides and a maximum stress above which, all hydrides precipitate in the radial direction. Based on these two extreme stress conditions, a model is derived to determine the stress level required to obtain a given fraction of radial hydrides after high temperature thermal–mechanical heat treatment. The proposed model is validated using metallographic observation data on pressurized tubes cooled down under constant pressure. Most of the samples with reoriented hydrides are further subjected to a ductility test. Using finite element modeling, the test results are analyzed in terms of crack nucleation within radial hydrides at the outer diameter and crack growth through the thickness of the tubular samples. The combination of test results shows that samples with hydrogen contents of about 100 wppm had the lowest ductility.

High Burnup Fuel Behaviour under LOCA Conditions as Observed in Halden Reactor Experiments

International Nuclear Information System (INIS)

Kolstad, E.; Wiesenack, W.; Oberlander, B.; Tverberg, T.

2013-01-01

In the context of assessing the validity of safety criteria for loss of coolant accidents with high burnup fuel, the OECD Halden Reactor Project has implemented an integral in-pile LOCA test series. In this series, fuel fragmentation and relocation, axial gas communication in high burnup rods as affected by gap closure and fuel- clad bonding, and secondary cladding oxidation and hydriding are of major interest. In addition, the data are being used for code validation as well as model development and verification. So far, nine tests with irradiated fuel segments (burnup 40-92 MW.d.kg -1 ) from PWR, BWR and VVER commercial nuclear power plants have been carried out. The in-pile measurements and the PIE results show a good repeatability of the experiments. The paper describes the experimental setup as well as the principal features and main results of these tests. Fuel fragmentation and relocation have occurred to varying degrees in these tests. The paper compares the conditions leading to the presence or absence of fuel fragmentation, e.g., burnup and loss of constraint. Axial gas flow is an important driving force for clad ballooning, fuel relocation and fuel expulsion. The experiments have provided evidence that such gas flow can be impeded in high burnup fuel with a potential impact on the ballooning and fuel dispersal. Although the results of the Halden LOCA tests are, to some extent, amplified by conditions and features deliberately introduced into the test series, the fuel behaviour identified in the Halden tests has an impact on the safety assessment of high burnup fuel and should give rise to improvements of the predictive capabilities of LOCA modelling codes. (author)

BWR spent fuel storage cask performance test. Volume 2. Pre- and post-test decay heat, heat transfer, and shielding analyses

International Nuclear Information System (INIS)

Wiles, L.E.; Lombardo, N.J.; Heeb, C.M.; Jenquin, U.P.; Michener, T.E.; Wheeler, C.L.; Creer, J.M.; McCann, R.A.

1986-06-01

This report describes the decay heat, heat transfer, and shielding analyses conducted in support of performance testing of a Ridhihalgh, Eggers and Associates REA 2033 boiling water reactor (BWR) spent fuel storage cask. The cask testing program was conducted for the US Department of Energy (DOE) Commercial Spent Fuel Management Program by the Pacific Northwest Laboratory (PNL) and by General Electric at the latters' Morris Operation (GE-MO) as reported in Volume I. The analyses effort consisted of performing pretest calculations to (1) select spent fuel for the test; (2) symmetrically load the spent fuel assemblies in the cask to ensure lateral symmetry of decay heat generation rates; (3) optimally locate temperature and dose rate instrumentation in the cask and spent fuel assemblies; and (4) evaluate the ORIGEN2 (decay heat), HYDRA and COBRA-SFS (heat transfer), and QAD and DOT (shielding) computer codes. The emphasis of this second volume is on the comparison of code predictions to experimental test data in support of the code evaluation process. Code evaluations were accomplished by comparing pretest (actually pre-look, since some predictions were not completed until testing was in progress) predictions with experimental cask testing data reported in Volume I. No attempt was made in this study to compare the two heat transfer codes because results of other evaluations have not been completed, and a comparison based on one data set may lead to erroneous conclusions

BWR power oscillation evaluation methodologies in core design

International Nuclear Information System (INIS)

Hotta, Akitoshi

1995-01-01

At the initial stage of BWR development, the power oscillation due to the nuclear-thermal interaction originated in random boiling phenomena and nuclear void feedback was feared. But it was shown that under the high pressure condition in the normal operation of recent commercial BWRs, the core is in very stable state. However, power oscillation events have been observed in actual machines, and it is necessary to do the stability evaluation that sufficiently reflects the detailed operation conditions of actual plants. As the cause of power oscillation events, the instability of control system and nuclear-thermal coupling instability are important, and their mechanisms are explained. As the model for analyzing the stability of BWR core, the nuclear-thermal coupling model in frequency domain is the central existence. As the information for the design, the parameters of fuel assemblies, and the nuclear parameters and the thermohydraulic parameters of cores are enumerated. LAPUR-TSI is a nuclear-thermal coupling model. The analysis system in the software of Tokyo Electric Power Co. is outlined, and the analysis model was verified. (K.I.)

Developing and modeling of the 'Laguna Verde' BWR CRDA benchmark

International Nuclear Information System (INIS)

Solis-Rodarte, J.; Fu, H.; Ivanov, K.N.; Matsui, Y.; Hotta, A.

2002-01-01

Reactivity initiated accidents (RIA) and design basis transients are one of the most important aspects related to nuclear power reactor safety. These events are re-evaluated whenever core alterations (modifications) are made as part of the nuclear safety analysis performed to a new design. These modifications usually include, but are not limited to, power upgrades, longer cycles, new fuel assembly and control rod designs, etc. The results obtained are compared with pre-established bounding analysis values to see if the new core design fulfills the requirements of safety constraints imposed on the design. The control rod drop accident (CRDA) is the design basis transient for the reactivity events of BWR technology. The CRDA is a very localized event depending on the control rod insertion position and the fuel assemblies surrounding the control rod falling from the core. A numerical benchmark was developed based on the CRDA RIA design basis accident to further asses the performance of coupled 3D neutron kinetics/thermal-hydraulics codes. The CRDA in a BWR is a mostly neutronic driven event. This benchmark is based on a real operating nuclear power plant - unit 1 of the Laguna Verde (LV1) nuclear power plant (NPP). The definition of the benchmark is presented briefly together with the benchmark specifications. Some of the cross-sections were modified in order to make the maximum control rod worth greater than one dollar. The transient is initiated at steady-state by dropping the control rod with maximum worth at full speed. The 'Laguna Verde' (LV1) BWR CRDA transient benchmark is calculated using two coupled codes: TRAC-BF1/NEM and TRAC-BF1/ENTREE. Neutron kinetics and thermal hydraulics models were developed for both codes. Comparison of the obtained results is presented along with some discussion of the sensitivity of results to some modeling assumptions

A simplified spatial model for BWR stability

International Nuclear Information System (INIS)

Berman, Y.; Lederer, Y.; Meron, E.

2012-01-01

A spatial reduced order model for the study of BWR stability, based on the phenomenological model of March-Leuba et al., is presented. As one dimensional spatial dependence of the neutron flux, fuel temperature and void fraction is introduced, it is possible to describe both global and regional oscillations of the reactor power. Both linear stability analysis and numerical analysis were applied in order to describe the parameters which govern the model stability. The results were found qualitatively similar to past results. Doppler reactivity feedback was found essential for the explanation of the different regions of the flow-power stability map. (authors)

Development of advanced neutron radiography for inspection on irradiated fuels and materials (2). Observation of hydride and oxide film on zircaloy cladding by using neutron radiography

International Nuclear Information System (INIS)

Yasuda, Ryou; Nakata, Masahito; Mastubayashi, Masahito; Harada, Katsuya

2001-02-01

Neutron radiography has been used as available diagnosis method of integrity on irradiated fuels, and has not been employed for estimating hydride and oxide film, which are influenced on integrity of Zircaloy cladding. Preliminary tests for PIE were carried out to assess possibility of neutron radiography as evaluation tool for hydrided and oxide film on the cladding. In these experiments, Zircaloy claddings with controlled amount of hydrogen absorption (200, 500, and 1000ppm) and thickness of oxide film were radiographed in center axis and in side directions of cladding tube by neutron imaging plate method. It is noted that thickness of oxide film was formed range from 7 μ m to 70 μ m at various temperatures (973, 1173, and 1323K) under steam atmosphere on the Zircaloy claddings. CT (Computed Tomography) restructure calculation was carried out to obtain cross section image of the claddings non-destructively. The Radiographs were qualitatively investigated about structure formation area and dependence of hydrogen absorption amount on PSL (Photo Simulated Luminescence) and CT values using by image analysis processor. At the results of imaging plate test, obvious difference was not found out between hydride formation (except for 1000ppm cladding) and standard claddings in side direction image. However, on the center axis direction image, outer circumference in the cladding cross-section that corresponded with hydride segregation area became blacker. In the case of oxide film formed cladding images, although oxide film could not find out on all speciments in the radiographs taken at the center axis and side directions, cross-section of claddings heat-processed at 973K showed appreciable blackness increasing with oxide film thickness on the radiographs. On the other hand, there is no effective difference between images of oxide film formed claddings processed at 1173K and 1323K and that of standard cladding. In CT image of 1000ppm hydrogen absorbed cladding, it is

Investigation of Lithium Metal Hydride Materials for Mitigation of Deep Space Radiation

Science.gov (United States)

Rojdev, Kristina; Atwell, William

2016-01-01

Radiation exposure to crew, electronics, and non-metallic materials is one of many concerns with long-term, deep space travel. Mitigating this exposure is approached via a multi-faceted methodology focusing on multi-functional materials, vehicle configuration, and operational or mission constraints. In this set of research, we are focusing on new multi-functional materials that may have advantages over traditional shielding materials, such as polyethylene. Metal hydride materials are of particular interest for deep space radiation shielding due to their ability to store hydrogen, a low-Z material known to be an excellent radiation mitigator and a potential fuel source. We have previously investigated 41 different metal hydrides for their radiation mitigation potential. Of these metal hydrides, we found a set of lithium hydrides to be of particular interest due to their excellent shielding of galactic cosmic radiation. Given these results, we will continue our investigation of lithium hydrides by expanding our data set to include dose equivalent and to further understand why these materials outperformed polyethylene in a heavy ion environment. For this study, we used HZETRN 2010, a one-dimensional transport code developed by NASA Langley Research Center, to simulate radiation transport through the lithium hydrides. We focused on the 1977 solar minimum Galactic Cosmic Radiation environment and thicknesses of 1, 5, 10, 20, 30, 50, and 100 g/cm2 to stay consistent with our previous studies. The details of this work and the subsequent results will be discussed in this paper.

Fuel assembly for FBR type reactor and reactor core thereof

International Nuclear Information System (INIS)

Kobayashi, Kaoru.

1998-01-01

The present invention provides a fuel assembly to be loaded to a reactor core of a large sized FBR type reactor, in which a coolant density coefficient can be reduced without causing power peaking in the peripheral region of neutron moderators loaded in the reactor core. Namely, the fuel assembly for the FBR type reactor comprises a plurality of fission product-loaded fuel rods and a plurality of fertile material-loaded fuel rods and one or more rods loading neutron moderators. In this case, the plurality of fertile material-loaded fuel rods are disposed to the peripheral region of the neutron moderator-loaded rods. The plurality of fission product-loaded fuel rods are disposed surrounding the peripheral region of the plurality of fertile material-loaded fuel rods. The neutron moderator comprises zirconium hydride, yttrium hydride and calcium hydride. The fission products are mixed oxide fuels. The fertile material comprises depleted uranium or natural uranium. (I.S.)

Fuel design with low peak of local power for BWR reactors with increased nominal power

International Nuclear Information System (INIS)

Perusquia C, R.; Montes, J.L.; Hernandez, J.L.; Ortiz, J.J.; Castillo, A.

2006-01-01

The Federal Commission of Electricity recently announcement the beginning of the works related with the increase of the power to 120% of the original nominal one in the Boiling Water Reactors (BWR) of the Laguna Verde Central (CLV): In the National Institute of Nuclear Research (ININ) are carried out studies of the impact on the design of the recharge of derived fuel of this increase. One of the main effects of the power increase type that it is promoting, is the increment of the flow of generated vapor, what takes, to a bigger fraction of vacuum in the core presenting increased values of the maximum fraction to the limit, so much of the ratio of lineal heat generation (XFLPD) as of the ratio of critic power (MFLCPR). In the made studies, it is found that these fractions rise lineally with the increase of the nominal power. Considering that the reactors of the CLV at the moment operate to 105% of the original nominal power, it would imply an increment of the order of 13.35% in the XFLPD and in the MFLCPR operating to a nominal power of 120% of the original one. This would propitiate bigger problems to design appropriately the fuel cycle and the necessity, almost unavoidable, of to resort to a fuel assembly type more advanced for the recharges of the cores. As option, in the ININ the feasibility of continuing using the same type of it fuel assembles that one has come using recently in the CLV, the type GE12 is analyzed. To achieve it was outlined to diminish the peak factor of local power (LPPF) of the power cells that compose the fuel recharge in 13.35%. It was started of a fuel design previously used in the recharge of the unit 1 cycle 12 and it was re-design to use it in the recharge design of the cycle 13 of the unit 1, considering an increase to 120% of the original power and the same requirements of cycle extension. For the re-design of the fuel assembly cell it was used the PreDiCeldas computer program developed in the ININ. It was able to diminish the LPPF

Fuel assembly for BWR type reactor

International Nuclear Information System (INIS)

Ueda, Makoto

1990-01-01

Various considerations are applied to fuel rods for improving the fuel burnup degree. If a gap between the fuel rods is changed, this varies the easiness for the flow of coolants depending on places, to reduce the thermal margin. Then, it is noted for the distribution of stresses generated due to the difference of water pressure caused by the difference of water streams between the inside and the outside of a channel box, and composite value, of stresses upon occurrence of earthquakes, neutron irradiation and a channel creep phenomenon caused by the stresses of due to the water pressure difference described above, the thickness of the channel box is increased in the upstream and decreased toward the downstream. Further, fuel spacers at the position where the thickness of the channel box is changed are spaced apart from the channel box so as not to brought into contact with the channel box. This can contribute to the reduction of coolants pressure loss, improvement of critical power and improvement of reactivity, as well as remarkably moderate local stresses applied from the fuel spacers to the channel box due to horizontal vibrations upon occurrence of earthquakes to improve the integrity of fuel assembly. (N.H.)

Development of advanced BWR

International Nuclear Information System (INIS)

Toyota, Masatoshi

1982-01-01

The Japanese technology and domestic production of BWR type nuclear power plants have been established through the experiences in the construction and operation of BWRs in addition to the technical agreement with the General Electric Co. In early days, the plants experienced some trouble such as stress corrosion cracking and some inconvenience in the operation and maintenance. The government, electric power companies and BWR manufacturers have endeavored to standardize and improve the design of LWRs for the purpose of improving the safety, reliability and the rate of operation and reducing the radiation exposure dose of plant workers. The first and second stages of the standardization and improvement of LWRs have been completed. Five manufacturers of BWRs in the world have continued the conceptual design of a new version of BWR power plants. It was concluded that this is the most desirable version of BWR nuclear power stations, but the technical and economic evaluation must be made before the commercial application. Six electric power companies and three manufacturers of BWRs in Japan set up the organization to develop the technology in cooperation. The internal pump system, the new control rod drive mechanism and others are the main features. (Kako, I.)

Fuel assembly for FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki.

1995-01-01

Ordinary sodium bond-type fuel pins using nitride fuels, carbide fuels or metal fuels and pins incorporated with hydride moderators are loaded in a wrapper tube at a ratio of from 2 to 10% based on the total number of fuel pins. The hydride moderators are sealed in the hydride moderator incorporated pins at the position only for a range from the upper end to a reactor core upper position of substantially 1/4 of the height of the reactor core from the upper end of the reactor core as a center. Then, even upon occurrence of ULOF (loss of flow rate scram failure phenomenon), it gives characteristic of reducing the power only by a doppler coefficient and not causing boiling of coolant sodium but providing stable cooling to the reactor core. Therefore, a way of thinking on the assurance of passive safety is simplified to make a verification including on the reactor structure unnecessary. In an LMFBR type reactor using the fuel assembly, a critical experiment for confirming accuracy of nuclear design is sufficient for the item required for study and development, which provides a great economical effect. (N.H.)

Parametric study of the behaviour of a pre irradiated BWR fuel rod under conditions of LOCA simulated in the halden in pile test system with the FALCON code

Energy Technology Data Exchange (ETDEWEB)

Khvostov, G.; Zimmermann, M. A. [Laboratory for Reactor Physics and Systems Behaviour, Paul Scherrer Institut, Villigen (Switzerland); Ledergerber, G. [Kernkraftwerk Leibstadt AG, Leibstadt (Switzerland); Kolstad, E. [Institute for Energy Technology - OECD Halden Reactor Project, Halden (Norway); Montgomery, R. O. [Anatech Corporation, San Diego (United States)

2008-10-15

A new LOCA test at Halden was planned as the first experiment within the Halden LOCA program addressing the behaviour of commercially irradiated BWR fuel of medium burn up with burst of the cladding expected to occur at a temperature of about 1050.deg.C, which is essentially higher than in the preceding experiments. The specific measures to be adopted have been suggested based upon a parametric study using the FALCON fuel behaviour code and aimed at an optimized design of the test fuel rod for the given high target cladding temperature of 1150 .deg. C (peak local). The analysis has shown a reasonable agreement with the fundamental experimental findings, such as correlations of NUREG 0630, as well as consistency with the data from Halden LOCA testing available so far. Thus, a general conclusion is drawn about the applicability of the methodology developed at PSI to the analysis of LWR fuel rod behaviour during LOCA, in consideration of the effects of fuel burn up.

Fuel cell design using a new heuristic method

International Nuclear Information System (INIS)

Perusquia, R.; Montes T, J. L.; Ortiz S, J. J.; Castillo M, A.

2014-10-01

In this paper a new method for the pre-design of a typical fuel cell with a structural array of 10 x 10 fuel elements for a BWR is presented. The method is based on principles of maximum dispersion and minimum peaks of local power within the array of fuel elements. The pre-design of the fuel cells is made by simulation in two dimensions (2-D) through the cells physics code CASMO-4. For this purpose of pre-design the search process is guided by an objective function which is a combination of the main neutronic parameters of the fuel cell. The results show that the method is a promising tool that could be used for the design of fuel cells for use in a nuclear plant BWR. (Author)

Hydride embrittlement in zircaloy components

Energy Technology Data Exchange (ETDEWEB)

Lobo, Raquel M.; Andrade, Arnaldo H.P.; Castagnet, Mariano, E-mail: rmlobo@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

Zirconium alloys are used in nuclear reactor cores under high-temperature water environment. During service, hydrogen is generated by corrosion processes, and it is readily absorbed by these materials. When hydrogen concentration exceeds the terminal solid solubility, the excess hydrogen precipitates as zirconium hydride (ZrH{sub 2}) platelets or needles. Zirconium alloys components can fail by hydride cracking if they contain large flaws and are highly stressed. Zirconium alloys are susceptible to a mechanism for crack initiation and propagation termed delayed hydride cracking (DHC). The presence of brittle hydrides, with a K{sub Ic} fracture toughness of only a few MPa{radical}m, results in a severe loss in ductility and toughness when platelet normal is oriented parallel to the applied stress. In plate or tubing, hydrides tend to form perpendicular to the thickness direction due to the texture developed during fabrication. Hydrides in this orientation do not generally cause structural problems because applied stresses in the through-thickness direction are very low. However, the high mobility of hydrogen in a zirconium lattice enables redistribution of hydrides normal to the applied stress direction, which can result in localized embrittlement. When a platelet reaches a critical length it ruptures. If the tensile stress is sufficiently great, crack initiation starts at some of these hydrides. Crack propagation occurs by repeating the same process at the crack tip. Delayed hydride cracking can degrade the structural integrity of zirconium alloys during reactor service. The paper focuses on the fracture mechanics and fractographic aspects of hydride material. (author)

Conference 'Chemistry of hydrides' Proceedings

International Nuclear Information System (INIS)

1991-07-01

This collection of thesis of conference of Chemistry hydrides presents the results of investigations concerning of base questions of chemistry of nonorganic hydrides, including synthesis questions, studying of physical and chemical properties, thermodynamics, analytical chemistry, investigation of structure, equilibriums in the systems of metal-hydrogen, behaviour of nonorganic hydrides in non-water mediums and applying investigations in the chemistry area and technology of nonorganic hydrides

Radiation field control at the latest BWR plants -- design principle, operational experience and future subjects

Energy Technology Data Exchange (ETDEWEB)

Uchida, Shunsuke [Energy Research Lab., Ibaraki (Japan); Ohsumi, Katsumi; Takashima, Yoshie [Hitachi Works, Ibaraki (Japan)

1995-03-01

Improvements of operational procedures to control water chemistry, e.g., nickel/iron control, as well as application of hardware improvements for reducing radioactive corrosion products resulted in an extremely low occupational exposure of less than 0.5 man.Sv/yr without any serious impact on the radwaste system, for BWR plants involved in the Japanese Improvement and Standardization Program. Recently, {sup 60}C radioactively in the reactor water has been increasing due to less crud fixation on the two smooth surfaces of new type high performance fuels and to the pH drop caused by chromium oxide anions released from stainless steel structures and pipings. This increase must be limited by changes in water chemistry, e.g., applications of modified nickel/iron ratio control and weak alkali control. Controlled water chemistry to optimize three points, the plant radiation level and integrities of fuel and structural materials, is the primary future subject for BWR water chemistry.

Aluminum hydride as a hydrogen and energy storage material: Past, present and future

Energy Technology Data Exchange (ETDEWEB)

Graetz, J., E-mail: graetz@bnl.gov [Sustainable Energy Technologies Department, Brookhaven National Laboratory, Upton, NY (United States); Reilly, J.J. [Sustainable Energy Technologies Department, Brookhaven National Laboratory, Upton, NY (United States); Yartys, V.A.; Maehlen, J.P. [Institute for Energy Technology, Kjeller (Norway); Bulychev, B.M. [Department of Chemistry, Lomonosov Moscow State University, Moscow (Russian Federation); Antonov, V.E. [Institute of Solid State Physics, Russian Academy of Sciences, Chernogolovka (Russian Federation); Tarasov, B.P. [Institute of Problems of Chemical Physics, Russian Academy of Sciences, Chernogolovka (Russian Federation); Gabis, I.E. [Department of Physics, Saint-Petersburg State University, St. Petersburg (Russian Federation)

2011-09-15

Aluminum hydride (AlH{sub 3}) and its associated compounds make up a fascinating class of materials that have motivated considerable scientific and technological research over the past 50 years. Due primarily to its high energy density, AlH{sub 3} has become a promising hydrogen and energy storage material that has been used (or proposed for use) as a rocket fuel, explosive, reducing agent and as a hydrogen source for portable fuel cells. This review covers the past, present and future research on aluminum hydride and includes the latest research developments on the synthesis of {alpha}-AlH{sub 3} and the other polymorphs (e.g., microcrystallization reaction, batch and continuous methods), crystallographic structures, thermodynamics and kinetics (e.g., as a function of crystallite size, catalysts and surface coatings), high-pressure hydrogenation experiments and possible regeneration routes.

BWR fuel cycle optimization using neural networks

International Nuclear Information System (INIS)

Ortiz-Servin, Juan Jose; Castillo, Jose Alejandro; Pelta, David Alejandro

2011-01-01

Highlights: → OCONN a new system to optimize all nuclear fuel management steps in a coupled way. → OCON is based on an artificial recurrent neural network to find the best combination of partial solutions to each fuel management step. → OCONN works with a fuel lattices' stock, a fuel reloads' stock and a control rod patterns' stock, previously obtained with different heuristic techniques. → Results show OCONN is able to find good combinations according the global objective function. - Abstract: In nuclear fuel management activities for BWRs, four combinatorial optimization problems are solved: fuel lattice design, axial fuel bundle design, fuel reload design and control rod patterns design. Traditionally, these problems have been solved in separated ways due to their complexity and the required computational resources. In the specialized literature there are some attempts to solve fuel reloads and control rod patterns design or fuel lattice and axial fuel bundle design in a coupled way. In this paper, the system OCONN to solve all of these problems in a coupled way is shown. This system is based on an artificial recurrent neural network to find the best combination of partial solutions to each problem, in order to maximize a global objective function. The new system works with a fuel lattices' stock, a fuel reloads' stock and a control rod patterns' stock, previously obtained with different heuristic techniques. The system was tested to design an equilibrium cycle with a cycle length of 18 months. Results show that the new system is able to find good combinations. Cycle length is reached and safety parameters are fulfilled.

Influence of Climatic Factors on the Efficiency of Disposal Metal- Hydride Unit for the Double-Fuel Low-Speed Internal Combustion Engine of Gas Tankers

OpenAIRE

Cherednichenko, Oleksandr Costyntunovich; Tkach, Mykhaylo Romanovich

2017-01-01

Contemporary tendencies in the development of ship power engineering have been analyzed. Consideration was given to the specific features of the transportation of liquefied natural gas by gas tankers. The prospects of utilization of the secondary energy resources of marine double-fuel low-speed diesel engines were defined. The metal hydride units of a continuous action were offered for this purpose. The need for the estimation of the influence of climatic factors on the efficiency of disposal...

Comparison of thorium-based fuels with different fissile components in existing BWRs

International Nuclear Information System (INIS)

Bjoerk, Klara Insulander; Fhager, Valentin; Demaziere, Christophe

2009-01-01

Three different types of thorium based BWR fuel have been developed, in each of which thorium was combined with a different fissile component, the three components being reactor grade plutonium, uranium enriched to 20% in uranium 235 and pure uranium 233. A BWR nuclear bundle design, based on the geometrical fuel assembly design GE14, was developed for each of these fissile components. The properties and performance of the corresponding fuel assemblies were investigated via full core calculations carried out for an existing BWR and compared with the ones of an ordinary Low Enriched Uranium (LEU) fuel, which was developed for reference. The fuel assemblies and cores were designed to meet existing fuel design criteria, and were then analyzed with regards to reactivity coefficients, delayed neutron fractions, control rod worths and shutdown margins. The results show that all three alternatives seem to be feasible, although some difficulties remain with complying with the thermal limits, and with the moderator temperature and coolant void coefficients of the U-233 containing fuel being positive under some circumstances. (author)

Calculation of nuclide inventory, decay power, activity and dose rates for spent nuclear fuel

International Nuclear Information System (INIS)

Haakansson, Rune

2000-03-01

The nuclide inventory was calculated for a BWR and a PWR fuel element, with burnups of 38 and 55 MWd/kg uranium for the BWR fuel, and 42 and 60 MWd/kg uranium for the PWR fuel. The calculations were performed for decay times of up to 300,000 years. Gamma and neutron dose rates have been calculated at a distance of 1 m from a bare fuel element and outside the spent fuel canister. The calculations were performed using the CASMO-4 code

Improved point-kinetics model for the BWR control rod drop accident

International Nuclear Information System (INIS)

Neogy, P.; Wakabayashi, T.; Carew, J.F.

1985-01-01

A simple prescription to account for spatial feedback weighting effects in RDA (rod drop accident) point-kinetics analyses has been derived and tested. The point-kinetics feedback model is linear in the core peaking factor, F/sub Q/, and in the core average void fraction and fuel temperature. Comparison with detailed spatial kinetics analyses indicates that the improved point-kinetics model provides an accurate description of the BWR RDA

Fuel R and D to improve fuel reliability

International Nuclear Information System (INIS)

Yang, Rosa; Cheng, Bo; Deshon, Jeff; Edsinger, Kurt; Ozer, Odelli

2006-01-01

Light water reactor fuel is operating in an increasingly challenging environment. Fuel burnup extension and cycle length increase both can increase the local duty. Reactor water chemistry modifications for the purpose of protection the plant system materials have the potential of increasing fuel surface deposition and cladding corrosion and hydriding. The status of fuel performance in US reactors is summarized and an update of the Fuel Reliability Program' established by the utility industry to ensure reliability is provided. (author)

Descriptions of reference LWR facilities for analysis of nuclear fuel cycles. Appendixes

International Nuclear Information System (INIS)

Schneider, K.J.; Kabele, T.J.

1979-09-01

The appendixes present the calculations that were used to derive the release factors discussed for each fuel cycle facility in Volume I. Appendix A presents release factor calculations for a surface mine, underground mine, milling facility, conversion facility, diffusion enrichment facility, fuel fabrication facility, PWR, BWR, and reprocessing facility. Appendix B contains additional release factors calculated for a BWR, PWR, and a reprocessing facility. Appendix C presents release factors for a UO 2 fuel fabrication facility

GENUSA Fuel Evolution

Energy Technology Data Exchange (ETDEWEB)

Choithramani, Sylvia; Malpica, Maria [ENUSA Industrias Avanzadas, GENUSA, Josefa Valcarcel, 26 28027 Madrid (Spain); Fawcett, Russel [Global Nuclear Fuel (United States)

2009-06-15

GNF ENUSA Nuclear Fuel S.A. (GENUSA) was formed in Madrid in May 1996. GENUSA is a corporation organized and existing under the laws of Spain, jointly owned by GNF-A and ENUSA. GENUSA consolidates all European BWR fuel marketing activities of GNF-A and ENUSA, primarily providing marketing and project management. In its standard way of operating, it will obtain engineering, components and conversion from GNF-A and engineering, fabrication and fuel related services from ENUSA. GENUSA's development philosophy over the past decades has been to introduce evolutionary designs, supported by our global experience base, that deliver the performance needed by our customers to meet their operating strategies. GENUSA considers, as one of our strengths, the ever-increasing experience base that provides the foundation for such evolutionary changes. This experience is supported and complemented with an even greater GNF experience. Over the last 40 years, GNF and ENUSA have designed, fabricated, and placed in operation over 144,000 BWR fuel bundles containing over 9.7 million fuel rods. This experience base represents the widest range of operating conditions of any BWR fuel vendor, reflecting varying reactor power densities, operating strategies, and water chemistry environments. It covers operating periods of up to {approx}10 years and bundle average exposures up to 68 MWd/kgU.. It provides the confirmation of our understanding and ability to model fuel performance behavior, and has been instrumental in the identification and characterization of each encountered failure mechanism. With the knowledge gained from this extensive experience base, mitigating actions have been developed and progressively implemented by GENUSA as part of a continuous program toward improved fuel reliability and performance. GENUSA's evolutionary product introduction strategy has been extremely successful. There has been a continuous stream of new products/processes that were developed to

Power generator in BWR type reactors

International Nuclear Information System (INIS)

Yoshida, Kenji.

1984-01-01

Purpose: To enable to perform stable and dynamic conditioning operation for nuclear fuels in BWR type reactors. Constitution: The conditioning operation for the nuclear fuels is performed by varying the reactor core thermal power in a predetermined pattern by changing the predetermined power changing pattern of generator power, the rising rate of the reactor core thermal power and the upper limit for the rising power of the reactor core thermal power are calculated and the power pattern for the generator is corrected by a power conditioning device such that the upper limit for the thermal power rising rate and the upper limit for the thermal power rising rate are at the predetermined levels. Thus, when the relation between the reactor core thermal power and the generator electrical power is fluctuated, the fluctuation is detected based on the variation in the thermal power rising rate and the limit value for the thermal power rising rate, and the correction is made to the generator power changing pattern so that these values take the predetermined values to thereby perform the stable conditioning operation for the nuclear fuels. (Moriyama, K.)

Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

Science.gov (United States)

González-Robles, E.; Serrano-Purroy, D.; Sureda, R.; Casas, I.; de Pablo, J.

2015-10-01

The denominated instant release fraction (IRF) is considered in performance assessment (PA) exercises to govern the dose that could arise from the repository. A conservative definition of IRF comprises the total inventory of radionuclides located in the gap, fractures, and the grain boundaries and, if present, in the high burn-up structure (HBS). The values calculated from this theoretical approach correspond to an upper limit that likely does not correspond to what it will be expected to be instantaneously released in the real system. Trying to ascertain this IRF from an experimental point of view, static leaching experiments have been carried out with two commercial UO2 spent nuclear fuels (SNF): one from a pressurized water reactor (PWR), labelled PWR, with an average burn-up (BU) of 52 MWd/kgU and fission gas release (FGR) of 23.1%, and one from a boiling water reactor (BWR), labelled BWR, with an average BU of and 53 MWd/kgU and FGR of 3.9%. One sample of each SNF, consisting of fuel and cladding, has been leached in bicarbonate water during one year under oxidizing conditions at room temperature (25 ± 5)°C. The behaviour of the concentration measured in solution can be divided in two according to the release rate. All radionuclides presented an initial release rate that after some days levels down to a slower second one, which remains constant until the end of the experiment. Cumulative fraction of inventory in aqueous phase (FIAPc) values has been calculated. Results show faster release in the case of the PWR SNF. In both cases Np, Pu, Am, Cm, Y, Tc, La and Nd dissolve congruently with U, while dissolution of Zr, Ru and Rh is slower. Rb, Sr, Cs and Mo, dissolve faster than U. The IRF of Cs at 10 and 200 days has been calculated, being (3.10 ± 0.62) and (3.66 ± 0.73) for PWR fuel, and (0.35 ± 0.07) and (0.51 ± 0.10) for BWR fuel.

Hydrogen Outgassing from Lithium Hydride

Energy Technology Data Exchange (ETDEWEB)

Dinh, L N; Schildbach, M A; Smith, R A; Balazs1, B; McLean II, W

2006-04-20

Lithium hydride is a nuclear material with a great affinity for moisture. As a result of exposure to water vapor during machining, transportation, storage and assembly, a corrosion layer (oxide and/or hydroxide) always forms on the surface of lithium hydride resulting in the release of hydrogen gas. Thermodynamically, lithium hydride, lithium oxide and lithium hydroxide are all stable. However, lithium hydroxides formed near the lithium hydride substrate (interface hydroxide) and near the sample/vacuum interface (surface hydroxide) are much less thermally stable than their bulk counterpart. In a dry environment, the interface/surface hydroxides slowly degenerate over many years/decades at room temperature into lithium oxide, releasing water vapor and ultimately hydrogen gas through reaction of the water vapor with the lithium hydride substrate. This outgassing can potentially cause metal hydriding and/or compatibility issues elsewhere in the device. In this chapter, the morphology and the chemistry of the corrosion layer grown on lithium hydride (and in some cases, its isotopic cousin, lithium deuteride) as a result of exposure to moisture are investigated. The hydrogen outgassing processes associated with the formation and subsequent degeneration of this corrosion layer are described. Experimental techniques to measure the hydrogen outgassing kinetics from lithium hydride and methods employing the measured kinetics to predict hydrogen outgassing as a function of time and temperature are presented. Finally, practical procedures to mitigate the problem of hydrogen outgassing from lithium hydride are discussed.

Hydrogen Storage using Metal Hydrides in a Stationary Cogeneration System

International Nuclear Information System (INIS)

Botzung, Maxime; Chaudourne, Serge; Perret, Christian; Latroche, Michel; Percheron-Guegan, Annick; Marty Philippe

2006-01-01

In the frame of the development of a hydrogen production and storage unit to supply a 40 kW stationary fuel cell, a metal hydride storage tank was chosen according to its reliability and high energetic efficiency. The study of AB5 compounds led to the development of a composition adapted to the project needs. The absorption/desorption pressures of the hydride at 75 C (2 / 1.85 bar) are the most adapted to the specifications. The reversible storage capacity (0.95 %wt) has been optimized to our work conditions and chemical kinetics is fast. The design of the Combined Heat and Power CHP system requires 5 kg hydrogen storage but in a first phase, only a 0.1 kg prototype has been realised and tested. Rectangular design has been chosen to obtain good compactness with an integrated plate fin type heat exchanger designed to reach high absorption/desorption rates. In this paper, heat and mass transfer characteristics of the Metal Hydride tank (MH tank) during absorption/desorption cycles are given. (authors)

Simplification of neural network model for predicting local power distributions of BWR fuel bundle using learning algorithm with forgetting

International Nuclear Information System (INIS)

Tanabe, Akira; Yamamoto, Toru; Shinfuku, Kimihiro; Nakamae, Takuji; Nishide, Fusayo.

1995-01-01

Previously a two-layered neural network model was developed to predict the relation between fissile enrichment of each fuel rod and local power distribution in a BWR fuel bundle. This model was obtained intuitively based on 33 patterns of training signals after an intensive survey of the models. Recently, a learning algorithm with forgetting was reported to simplify neural network models. It is an interesting subject what kind of model will be obtained if this algorithm is applied to the complex three-layered model which learns the same training signals. A three-layered model which is expanded to have direct connections between the 1st and the 3rd layer elements has been constructed and the learning method of normal back propagation was applied first to this model. The forgetting algorithm was then added to this learning process. The connections concerned with the 2nd layer elements disappeared and the 2nd layer has become unnecessary. It took a longer computing time by an order to learn the same training signals than the simple back propagation, but the two-layered model was obtained autonomously from the expanded three-layered model. (author)

CERDEC Fuel Cell Team: Military Transitions for Soldier Fuel Cells

Science.gov (United States)

2008-10-27

Fuel Cell (DMFC) (PEO Soldier) Samsung: 20W DMFC (CRADA) General Atomics & Jadoo: 50W Ammonia Borane Fueled PEMFC Current Fuel Cell Team Efforts...Continued Ardica: 20W Wearable PEMFC operating on Chemical Hydrides Spectrum Brands w/ Rayovac: Hydrogen Generators and Alkaline Fuel Cells for AA...100W Ammonia Borane fueled PEMFC Ultralife: 150W sodium borohydride fueled PEMFC Protonex: 250W RMFC and Power Manager (ARO) NanoDynamics: 250W SOFC

Mechanical properties and fracture of titanium hydrides

International Nuclear Information System (INIS)

Koketsu, Hideyuki; Taniyama, Yoshihiro; Yonezu, Akio; Cho, Hideo; Ogawa, Takeshi; Takemoto, Mikio; Nakayama, Gen

2006-01-01

Titanium hydrides tend to suffer fracture when their thicknesses reach a critical thickness. Morphology and mechanical property of the hydrides are, however, not well known. The study aims to reveal the hydride morphology and fracture types of the hydrides. Chevron shaped plate hydrides were found to be produced on the surface of pure titanium (Grade 1) and Grade 7 titanium absorbing hydrogen. There were tree types of fracture of the hydrides, i.e., crack in hydride layer, exfoliation of the layer and shear-type fracture of the hydride plates, during the growth of the hydrides and deformation. We next estimated the true stress-strain curves of the hydrides on Grade 1 and 7 titanium using the dual Vickers indentation method, and the critical strain causing the Mode-I fine crack by indentation. Fracture strength and strain of the hydrides in Grade 1 titanium were estimated as 566 MPa and 4.5%, respectively. Those of the hydride in Grade 7 titanium were 498 MPa and 16%. Though the fracture strains estimated from the plastic instability of true stress-strain curves were approximately the half of those estimated by finite element method, the titanium hydrides were estimated to possess some extent of toughness or plastic deformation capability. (author)

Validation and application of the system code ATHLET-CD for BWR severe accident analyses

Energy Technology Data Exchange (ETDEWEB)

Di Marcello, Valentino, E-mail: valentino.marcello@kit.edu; Imke, Uwe; Sanchez, Victor

2016-10-15

Highlights: • We present the application of the system code ATHLET-CD code for BWR safety analyses. • Validation of core in-vessel models is performed based on KIT CORA experiments. • A SB-LOCA scenario is simulated on a generic German BWR plant up to vessel failure. • Different core reflooding possibilities are investigated to mitigate the accident consequences. • ATHLET-CD modelling features reflect the current state of the art of severe accident codes. - Abstract: This paper is aimed at the validation and application of the system code ATHLET-CD for the simulation of severe accident phenomena in Boiling Water Reactors (BWR). The corresponding models for core degradation behaviour e.g., oxidation, melting and relocation of core structural components are validated against experimental data available from the CORA-16 and -17 bundle tests. Model weaknesses are discussed along with needs for further code improvements. With the validated ATHLET-CD code, calculations are performed to assess the code capabilities for the prediction of in-vessel late phase core behaviour and reflooding of damaged fuel rods. For this purpose, a small break LOCA scenario for a generic German BWR with postulated multiple failures of the safety systems was selected. In the analysis, accident management measures represented by cold water injection into the damaged reactor core are addressed to investigate the efficacy in avoiding or delaying the failure of the reactor pressure vessel. Results show that ATHLET-CD is applicable to the description of BWR plant behaviour with reliable physical models and numerical methods adopted for the description of key in-vessel phenomena.

Investigation of water-logged spent fuel rods under dry storage conditions

International Nuclear Information System (INIS)

Kohli, R.; Pasupathi, V.

1986-09-01

Tests were conducted to determine the amount of moisture contained in breached, water-logged spent fuel rods and the rate of release. Two well-characterized BWR fuel rods with reactor-induced breaches were tested in a hot cell. These rods contained approximately 6 to 10 g of moisture, most of which was released during heating tests simulating normal cask drying operations. Additional testing with two intentionally defected fuel rods (BWR and PWR) was performed to evaluate the effect of the cladding breach on migration of moisture along the length of the fuel rod. The results showed that the moisture released from reactor-breached spent fuel rods was insufficient to cause degradation of fuel or dry storage system components

Hydrogen charging, hydrogen content analysis and metallographic examination of hydride in zirconium alloys

International Nuclear Information System (INIS)

Singh, R.N.; Kishore, R.; Mukherjee, S.; Roychowdhury, S.; Srivastava, D.; Sinha, T.K.; De, P.K.; Banerjee, S.; Gopalan, B.; Kameswaran, R.; Sheelvantra, Smita S.

2003-12-01

Gaseous and electrolytic hydrogen charging techniques for introducing controlled amount of hydrogen in zirconium alloy is described. Zr-1wt%Nb fuel tube, zircaloy-2 pressure tube and Zr-2.5Nb pressure tube samples were charged with up to 1000 ppm of hydrogen by weight using one of the aforementioned methods. These hydrogen charged Zr-alloy samples were analyzed for estimating the total hydrogen content using inert gas fusion technique. Influence of sample surface preparation on the estimated hydrogen content is also discussed. In zirconium alloys, hydrogen in excess of the terminal solid solubility precipitates out as brittle hydride phase, which acquire platelet shaped morphology due to its accommodation in the matrix and can make the host matrix brittle. The F N number, which represents susceptibility of Zr-alloy tubes to hydride embrittlement was measured from the metallographs. The volume fraction of the hydride phase, platelet size, distribution, interplatelet spacing and orientation were examined metallographically using samples sliced along the radial-axial and radial-circumferential plane of the tubes. It was observed that hydride platelet length increases with increase in hydrogen content. Considering the metallographs generated by Materials Science Division as standard, metallographs prepared by the IAEA round robin participants for different hydrogen concentration was compared. It is felt that hydride micrographs can be used to estimate not only that approximate hydrogen concentration of the sample but also its size, distribution and orientation which significantly affect the susceptibility to hydride embrittlement of these alloys. (author)

BWR-stability investigation at Forsmark 1

International Nuclear Information System (INIS)

Bergdahl, B.G.; Reisch, F.; Oguma, R.; Lorenzen, J.; Aakerhielm, F.

1988-01-01

A series of noise measurements have been conducted at Forsmark 1 during start-up operation after the revision summer '87. The main purpose was to investigate BWR-stability problems, i.e. resonant power oscillations of 0.5 Hz around 65% power and 4100 kg/s core flow, which tend to arise at high power and low core flow conditions. The analysis was performed to estimate the noise source which gives rise to the oscillation, to evaluate the measure of stability, i.e. the Decay Ratio (Dr) as well as to investigate other safety related problems. The result indicates that the oscillation is due to the dynamic coupling between the neutron kinetics and thermal hydraulics via void reactivity feedback. The Dr ranged between values of 0.7 and > 0.9, instead of expected 0.6 (Dr=1 is defined as instability). These high values imply that the core cannot suppress oscillations fast enough and a small perturbation can cause scram. Further it was found that the entire core is oscillating in phase (LPRM's) with varying strength where any connection to the consequences of different fuel (8x8, 9x9) being present simultaneously cannot be excluded. This report elucidates the importance of an on-line BWR-stability surveillance system with functions like stability condition monitoring and control system diagnosis. (orig.)

Analysis of natural circulation BWR dynamics with stochastic and deterministic methods

International Nuclear Information System (INIS)

VanderHagen, T.H.; Van Dam, H.; Hoogenboom, J.E.; Kleiss, E.B.J.; Nissen, W.H.M.; Oosterkamp, W.J.

1986-01-01

Reactor kinetic, thermal hydraulic and total plant stability of a natural convection cooled BWR was studied using noise analysis and by evaluation of process responses to control rod steps and to steamflow control valve steps. An estimate of the fuel thermal time constant and an impression of the recirculation flow response to power variations was obtained. A sophisticated noise analysis method resulted in more insight into the fluctuations of the coolant velocity

Revised uranium--plutonium cycle PWR and BWR models for the ORIGEN computer code

International Nuclear Information System (INIS)

Croff, A.G.; Bjerke, M.A.; Morrison, G.W.; Petrie, L.M.

1978-09-01

Reactor physics calculations and literature searches have been conducted, leading to the creation of revised enriched-uranium and enriched-uranium/mixed-oxide-fueled PWR and BWR reactor models for the ORIGEN computer code. These ORIGEN reactor models are based on cross sections that have been taken directly from the reactor physics codes and eliminate the need to make adjustments in uncorrected cross sections in order to obtain correct depletion results. Revised values of the ORIGEN flux parameters THERM, RES, and FAST were calculated along with new parameters related to the activation of fuel-assembly structural materials not located in the active fuel zone. Recommended fuel and structural material masses and compositions are presented. A summary of the new ORIGEN reactor models is given

Sophistication of operator training using BWR plant simulator

International Nuclear Information System (INIS)

Ohshiro, Nobuo; Endou, Hideaki; Fujita, Eimitsu; Miyakita, Kouji

1986-01-01

In Japanese nuclear power stations, owing to the improvement of fuel management, thorough maintenance and inspection, and the improvement of facilities, high capacity ratio has been attained. The thorough training of operators in nuclear power stations also contributes to it sufficiently. The BWR operator training center was established in 1971, and started the training of operators in April, 1974. As of the end of March, 1986, more than 1800 trainees completed training. At present, in the BWR operator training center, No.1 simulator of 800 MW class and No.2 simulator of 1100 MW class are operated for training. In this report, the method, by newly adopting it, good result was obtained, is described, that is, the method of introducing the feeling of being present on the spot into the place of training, and the new testing method introduced in retraining course. In the simulator training which is apt to place emphasis on a central control room, the method of stimulating trainees by playing the part of correspondence on the spot and heightening the training effect of multiple monitoring was tried, and the result was confirmed. The test of confirmation on the control board was added. (Kako, I.)

Hydrogen storage in metallic hydrides: the hydrides of magnesium-nickel alloys

International Nuclear Information System (INIS)

Silva, E.P. da.

1981-01-01

The massive and common use of hydrogen as an energy carrier requires an adequate solution to the problem of storing it. High pressure or low temperatures are not entirely satisfactory, having each a limited range of applications. Reversible metal hydrides cover a range of applications intermediate to high pressure gas and low temperature liquid hydrogen, retaining very favorable safety and energy density characteristics, both for mobile and stationary applications. This work demonstrates the technical viability of storing hydrogen in metal hydrides of magnesium-nickel alloys. Also, it shows that technology, a product of science, can be generated within an academic environment, of the goal is clear, the demand outstanding and the means available. We review briefly theoretical models relating to metal hydride properties, specially the thermodynamics properties relevant to this work. We report our experimental results on hydrides of magnesium-nickel alloys of various compositions including data on structure, hydrogen storage capacities, reaction kinetics, pressure-composition isotherms. We selected a promising alloy for mass production, built and tested a modular storage tank based on the hydrides of the alloy, with a capacity for storing 10 Nm sup(3) of hydrogen of 1 atm and 20 sup(0)C. The tank weighs 46,3 Kg and has a volume of 21 l. (author)

Radiotoxicity study of a boiling water reactor core design based on a thorium-uranium fuel concept

International Nuclear Information System (INIS)

Nunez C, A.; Espinosa P, G.

2007-01-01

Full text: The innovative design of a Boiling Water Reactor (BWR) equilibrium core using the thorium-uranium (blanket-seed) concept in the same integrated fuel assembly is presented in this paper. The lattice design uses the thorium conversion capability to 233 U in a BWR spectrum. A core design was developed to achieve an equilibrium cycle of one effective full power year in a standard BWR. A comparison of the toxicity of the spent fuel showed that toxicity is lower in the thorium cycle than other commercial fuels as UO 2 and MOX (uranium and plutonium) in case of the one-through cycle for LWR. (Author)

Feasibility assessment of burnup credit in the criticality analysis of shipping casks with boiling water reactor spent fuel

International Nuclear Information System (INIS)

Broadhead, B.L.

1991-08-01

Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This ''burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent fuel casks used for transportation and storage. Recently, analyses have demonstrated the technical feasibility and estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This report summarizes the extension of the previous PWR technical feasibility assessment to boiling water reactor (BWR) fuel. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. Two different aspects of fuel characterization were considered:l first, the generation of burn- up dependent material interaction probabilities; second, the prediction of material inventories over time (depletion). After characterizing the spent fuel at various stages of exposure and decay, three dimensional (3-D) models for an infinite array of assemblies and, in several cases, infinite arrays of assemblies in a typical shipping cask basket were analyzed. Results for assemblies without a basket provide reactivity control requirements as a function of burnup and decay, while results including the basket allow assessment of typical basket configurations to provide sufficient reactivity control for spent BWR fuel. Resulting basket worths and reactivity trends over time are then evaluated to determine whether burnup credit is needed and feasible in BWR applications

Hydriding and dehydriding characteristics of small-scale DU and ZrCo beds

Energy Technology Data Exchange (ETDEWEB)

Chung, Dongyou; Lee, Jungmin; Koo, Daeseo [Korea Atomic Energy Research Institute, 989-111 Daedeokdaero, Yuseong, Daejeon 305-353 (Korea, Republic of); Chung, Hongsuk, E-mail: hschung1@kaeri.kr [Korea Atomic Energy Research Institute, 989-111 Daedeokdaero, Yuseong, Daejeon 305-353 (Korea, Republic of); Kim, Ki Hwan [Korea Atomic Energy Research Institute, 989-111 Daedeokdaero, Yuseong, Daejeon 305-353 (Korea, Republic of); Kang, Hyun-Goo; Chang, Min Ho [National Fusion Research Institute, 113 Gwahakro, Yuseong, Daejeon 305-333 (Korea, Republic of); Camp, Patrick [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Jung, Ki Jung; Cho, Seungyon; Yun, Sei-Hun; Kim, Chang Shuk [National Fusion Research Institute, 113 Gwahakro, Yuseong, Daejeon 305-333 (Korea, Republic of); Yoshida, Hiroshi [Fusion Science Consultant, 3288-10 Sakado-cho, Mito-shi 310-0841, Ibakaki-ken (Japan); Paek, Seungwoo; Lee, Hansoo [Korea Atomic Energy Research Institute, 989-111 Daedeokdaero, Yuseong, Daejeon 305-353 (Korea, Republic of)

2013-10-15

Highlights: • We have designed and fabricated a twosome small-scale getter bed for a comparison of ZrCo with DU on the hydriding/dehydriding properties. • We provide preliminary experimental results of our ZrCo and DU beds. -- Abstract: With the development of fusion technology, it will be necessary to store large amounts of tritium during the nuclear fusion fuel cycle. Stable metal tritides are viewed as potential candidates for the high-density storage of tritium. Metal tritide formers offer a safe and convenient method for tritium storage. For the storage, supply, and recovery of hydrogen isotopes, zirconium cobalt (ZrCo) and depleted uranium (DU) have been extensively proposed. Thus, we have designed and fabricated two identical small-scale getter beds for a comparison of ZrCo with DU on the hydriding/dehydriding properties. After the powderization of the metals, the hydriding/dehydriding performance at different stoichiometries of ZrCo and DU was measured. We provide preliminary experimental results of our ZrCo and DU beds.

BWR Full Integral Simulation Test (FIST) Phase II test results and TRAC-BWR model qualification

International Nuclear Information System (INIS)

Sutherland, W.A.; Alamgir, M.; Findlay, J.A.; Hwang, W.S.

1985-10-01

Eight matrix tests were conducted in the FIST Phase I. These tests investigated the large break, small break and steamline break LOCA's, as well as natural circulation and power transients. There are nine tests in Phase II of the FIST program. They include the following LOCA tests: BWR/6 LPCI line break, BWR/6 intermediate size recirculation break, and a BWR/4 large break. Steady state natural circulation tests with feedwater makeup performed at high and low pressure, and at high pressure with HPCS makeup, are included. Simulation of a transient without rod insertion, and with controlled depressurization, was performed. Also included is a simulation of the Peach Bottom turbine trip test. The final two tests simulated a failure to maintain water level during a postulated accident. A FIST program objective is to assess the TRAC code by comparisons with test data. Two post-test predictions made with TRACB04 are compared with Phase II test data in this report. These are for the BWR/6 LPCI line break LOCA, and the Peach Bottom turbine trip test simulation

Report on the basic design of a hydrogen transportation system utilizing metal hydrides and the evaluation thereon; Kinzoku suisokabutsu wo riyoshita suiso yuso system no kihon sekkei to sono hyoka ni kansuru hokokusho

Energy Technology Data Exchange (ETDEWEB)

NONE

1981-03-26

This paper describes a hydrogen transportation system utilizing metal hydrides. For a storage method for moving, metal hydrides having high hydrogen containing performance like Mg-based hydrides would have high portability, less weight disadvantage, and high economic performance. In the fixed location storage, metal hydrides are superior in safety and maintenance cost to the conventional high-pressure gas holder and liquefied hydrogen storage. Because of their high dependence on equilibrium pressure and temperature, the significance of development thereof is large as the source of high-pressure hydrogen generation and motive force. More effective utilization of low-level heat, and separation and refining of hydrogen may also be expected. With regard to fuel supply for hydrogen fueled automobiles, metal hydrides are better in safety and total energy cost than liquefied hydrogen, but have a number of disadvantageous points in weight demerit. Eliminating the weight demerit would be the central issue of the development. Accompanying the development of hydrogen fueled automobiles, there are a number of technological elements to be developed on fuel supply system, such as storage, moving and transportation in hydrogen manufacturing sites, and filling and storage at using sites. Arranging the related infrastructures would be the issue. (NEDO)

Prototypical spent nuclear fuel rod consolidation equipment: Phase 2, Final design report: Volume 4, Appendices: Part 3

International Nuclear Information System (INIS)

Ciez, A.P.

1987-01-01

The purpose of this manual is to provide assembly, installation, operation, maintenance, and off-normal recovery procedures for the Consolidation Equipment. The Consolidation System is a horizontal, dry system capable of processing one Pressurized Water Reactor (PWR) fuel assembly or one Boiling Water Reactor (BWR) fuel assembly at a time. The system will process all spent PWR and BWR fuels from the commercial US nuclear power reactor industry. Component changeouts for various fuel types have been minimized to reduce costs, required in-cell module storage space, and to increase efficiency by decreasing set-up time between fuel consolidation campaigns. The most important feature of the Westinghouse system is the ability to control the fuel rods at all times during the consolidation process from rod extraction, through canister loading. This features assures that the rods from two PWR fuel assemblies or four BWR fuel assemblies (minimum) can be loaded into one consolidated rods canister

Best-estimate analysis development for BWR systems

International Nuclear Information System (INIS)

Sutherland, W.A.; Alamgir, M.; Kalra, S.P.; Beckner, W.D.

1986-01-01

The Full Integral Simulation Test (FIST) Program is a three pronged approach to the development of best-estimate analysis capability for BWR systems. An experimental program in the FIST BWR system simulator facility extends the LOCA data base and adds operational transients data. An analytical method development program with the BWR-TRAC computer program extends the modeling of BWR specific components and major interfacing systems, and improves numerical techniques to reduce computer running time. A method qualification program tests TRAC-B against experiments run in the FIST facility and extends the results to reactor system applications. With the completion and integration of these three activities, the objective of a best-estimate analysis capability has been achieved. (author)

The influence of stress state on the reorientation of hydrides in a zirconium alloy

International Nuclear Information System (INIS)

Cinbiz, Mahmut N.; Koss, Donald A.; Motta, Arthur T.

2016-01-01

Hydride reorientation can occur in spent nuclear fuel cladding when subjected to a tensile hoop stress above a threshold value during cooling. Because in these circumstances the cladding is under a multiaxial stress state, the effect of stress biaxiality on the threshold stress for hydride reorientation is investigated using hydrided CWSR Zircaloy-4 sheet specimens containing ∼180 wt ppm of hydrogen and subjected to a two-cycle thermo-mechanical treatment. The study is based on especially designed specimens within which the stress biaxiality ratios range from uniaxial (σ_2/σ_1 = 0) to “near-equibiaxial” tension (σ_2/σ_1 = 0.8). The threshold stress is determined by mapping finite element calculations of the principal stresses and of the stress biaxiality ratio onto the hydride microstructure obtained after the thermo-mechanical treatment. The results show that the threshold stress (maximum principal stress) decreases from 155 to 75 MPa as the stress biaxiality increases from uniaxial to “near-equibiaxial” tension.

Panorama of the BWR reactors - Evolution of the concept

Energy Technology Data Exchange (ETDEWEB)

Novotny, C.; Uhrig, E. [AREVA NP GmbH, Safety Engineering Department - PEPS-G (Germany)

2012-01-15

Nowadays, a fleet of more than 50 boiling water reactors (BWR) are in operation in the world. This article gives a short overview on the developments of nuclear power plants of the BWR type, with a focus on the European builds. It describes the technical bases from the early designs in the fifties, sketches the innovations of the sixties and seventies in the types BWR 69 and 72 (Baulinie 69 and 72) and gives an outlook of a possible next generation BWR. A promising approach in recent BWR developments is the the combination of passive safety systems with established design basis

BWR water chemistry guidelines and PWR primary water chemistry guidelines in Japan – Purpose and technical background

Energy Technology Data Exchange (ETDEWEB)

Kawamura, Hirotaka, E-mail: kawamuh@criepi.denken.or.jp [Central Research Institute of Electric Power Industry (Japan); Hirano, Hideo [Central Research Institute of Electric Power Industry (Japan); Katsumura, Yousuke [University of Tokyo (Japan); Uchida, Shunsuke [Tohoku University (Japan); Mizuno, Takayuki [Mie University (Japan); Kitajima, Hideaki; Tsuzuki, Yasuo [Japan Nuclear Safety Institute (Japan); Terachi, Takumi [Institute of Nuclear Safety System, Inc. (Japan); Nagase, Makoto; Usui, Naoshi [Hitachi-GE Nuclear Energy, Ltd. (Japan); Takagi, Junichi; Urata, Hidehiro [Toshiba Corporation (Japan); Shoda, Yasuhiko; Nishimura, Takao [Mitsubishi Heavy Industry, Ltd. (Japan)

2016-12-01

Highlights: • Framework of BWR/PWR water chemistry Guidelines in Japan are presented. • Guideline necessity, definitions, philosophy and technical background are mentioned. • Some guideline settings for control parameters and recommendations are explaines. • Chemistry strategy is also mentioned. - Abstract: After 40 years of light water reactor (LWR) operations in Japan, the sustainable development of water chemistry technologies has aimed to ensure the highest coolant system component integrity and fuel reliability performance for maintaining LWRs in the world; additionally, it aimed to achieve an excellent dose rate reduction. Although reasonable control and diagnostic parameters are utilized by each boiling water reactor (BWR) and pressurized water reactor (PWR) owner, it is recognized that specific values are not shared among everyone involved. To ensure the reliability of BWR and PWR operation and maintenance, relevant members of the Atomic Energy Society of Japan (AESJ) decided to establish guidelines for water chemistry. The Japanese BWR and PWR water chemistry guidelines provide strategies to improve material and fuel reliability performance as well as to reduce dosing rates. The guidelines also provide reasonable “control values”, “diagnostic values” and “action levels” for multiple parameters, and they stipulate responses when these levels are exceeded. Specifically, “conditioning parameters” are adopted in the Japanese PWR primary water chemistry guidelines. Good practices for operational conditions are also discussed with reference to long-term experience. This paper presents the purpose, technical background and framework of the preliminary water chemistry guidelines for Japanese BWRs and PWRs. It is expected that the guidelines will be helpful as an introduction to achieve safety and reliability during operations.

Genusa Bepu methodologies for the safety analysis of BWRs; Metodologias Bepu de Genusa para el analisis de seguridad de reactores BWR

Energy Technology Data Exchange (ETDEWEB)

Trueba, M.; Garcia, J.; Goodson, C.; Ibarra, L.

2016-08-01

This article describes the BEPU methodologies developed by General Electric-Hitachi (GEH) for the evaluation of the BWR reactor safety analysis based on the TRACG best-estimate code. These methodologies are applicable to a wide range of events, operational transients (AOO), anticipated transients without scram (ATWS), loss of coolant accidents (LOCA) and instability events; to different BWR types operating commercially. General Electric (GE( designs and other vendors, including Generation III+ESBWR; to the new operation strategies, and to all types of BWR fuel. Their application achieves, among other benefits, a better understanding of the overall plant response and an improvement in margins to the operating limits; thus, the increase of flexibility in reactor operation and reduction in generation costs. (Author)

Fuel assembly and reactor core

International Nuclear Information System (INIS)

Aoyama, Motoo; Koyama, Jun-ichi; Uchikawa, Sadao; Bessho, Yasunori; Nakajima, Akiyoshi; Maruyama, Hiromi; Ozawa, Michihiro; Nakamura, Mitsuya.

1990-01-01

The present invention concerns fuel assemblies charged in a BWR type reactor and the reactor core. The fuel assembly comprises fuel rods containing burnable poisons and fuel rods not containing burnable poisons. Both of the highest and the lowest gadolinia concentrations of the fuel rods containing gadolinia as burnable poisons are present in the lower region of the fuel assembly. This can increase the spectral shift effect without increasing the maximum linear power density. (I.N.)

Influence of hydrides orientation on strain, damage and failure of hydrided zircaloy-4; Influence de l'orientation des hydrures sur les modes de deformation, d'endommagement et de rupture du zircaloy-4 hydrure

Energy Technology Data Exchange (ETDEWEB)

Racine, A

2005-09-15

In pressurized water reactors of nuclear power plants, fuel pellets are contained in cladding tubes, made of Zirconium alloy, for instance Zircaloy-4. During their life in the primary water of the reactor (155 bars, 300 C), cladding tubes are oxidized and consequently hydrided. A part of the hydrogen given off precipitates as Zirconium hydrides in the bulk material and embrittles the material. This embrittlement depends on many parameters, among which hydrogen content and orientation of hydrides with respect to the applied stress. This investigation is devoted to the influence of the orientation of hydrides with respect to the applied stress on strain, damage and failure mechanisms. Macroscopic and SEM in-situ ring tensile tests are performed on cladding tube material (unirradiated cold worked stress-relieved Zircaloy-4) hydrided with about 200 and 500 wppm hydrogen, and with different main hydrides orientation: either parallel or perpendicular to the circumferential tensile direction. We get the mechanical response of the material as a function of hydride orientation and hydrogen content and we investigate the deformation, damage and failure mechanisms. In both cases, digital image correlation techniques are used to estimate local and global strain distributions. Neither the tensile stress-strain response nor the global and local strain modes are significantly affected by hydrogen content or hydride orientation, but the failure modes are strongly modified. Indeed, only 200 wppm radial hydrides embrittle Zy-4: sample fail in the elastic domain at about 350 MPa before strain bands could develop; whereas in other cases sample reach at least 750 MPa before necking and final failure, in ductile or brittle mode. To model this particular heterogeneous material behavior, a non-coupled damage approach which takes into account the anisotropic distribution of the hydrides is proposed. Its parameters are identified from the macroscopic strain field measurements and a