WorldWideScience

Sample records for scaled pwr experiments

  1. Scaling studies - PWR

    International Nuclear Information System (INIS)

    Sonneck, G.

    1983-05-01

    A RELAP 4/MOD 6 study was made based on the blowdown phase of the intermediate break experiment LOFT L5-1. The method was to set up a base model and to vary parametrically some areas where it is known or suspected that LOFT differs from a commercial PWR. The aim was not to simulate LOFT or a PWR exactly but to understand the influence of the following parameters on the thermohydraulic behaviour of the system and the clad temperature: stored heat in the downcomer (LOFT has rather large filler blocks in this part of the pressure vessel); bypass between downcomer and upper plenum; and core length. The results show that LOFT is prototypical for all calculated blowdowns. As the clad temperatures decrease with decreasing stored energy in the downcomer, increased bypass and increased core length, LOFT results seem to be realistic as long as realistic bypass sizes are considered; they are conservative in the two other areas. (author)

  2. UPTF experiment: Effect of full-scale geometry on countercurrent flow behaviour in PWR downcomer

    International Nuclear Information System (INIS)

    Liebert, J.; Weiss, P.

    1989-01-01

    Four separate effects tests (13 runs) have been performed at UPTF - a 1:1 scale test facility - to investigate the thermal-hydraulic phenomena in the full-scale downcomer of a PWR during end-of-blowdown, refill and reflood phases. Special attention has been paid to the effects of geometry - cold leg arrangement - and ECC-water subcooling on downcomer countercurrent flow and ECC bypass behaviour. A synopsis of the most significant events and a comparison of countercurrent flow limitation (CCFL) data from UPTF and 1/5 scale test facility of Creare are given. The CCFL results of UPTF are compared to data predicted by an empirical correlation developed at Creare, based on the modified dimensionless Wallis parameter J * . A significant effect of cold leg arrangement on CCFL was observed leading to strongly heterogeneous flow condition in the downcomer. CCFL in front of cold leg 1 adjacent to the broken loop exists even for very low steam flow rates. Therefore the benefit of strong water subcooling is not as much as expected. The existing flooding correlation of Creare predicts the full-scale downcomer CCFL insufficiently. New flooding correlations are required to describe the CCFL process adequately. (orig.)

  3. PWR standardization: The French experience

    International Nuclear Information System (INIS)

    Bacher, P.E.

    1987-01-01

    After a short historical review of the French PWR programme with 45000 MWe in operation and 15000 MWe under construction, the paper first develops the objectives and limits of the standardizatoin policy. Implementation of standardization is described through successive reactor series and feedback of experience, together with its impact on safety and on codes and standards. Present benefits of standardization range from low engineering costs to low backfitting costs, via higher quality, reduction in construction times and start-up schedules and improved training of operators. The future of the French programme into the 1990's is again with an advanced standardized series, the N4-1400 MW plant. There is no doubt that the very positive experience with standardization is relevant to any country trying to achieve self-reliance in the nuclear power field. (author)

  4. Vibration behavior of PWR reactor internals Model experiments and analysis

    International Nuclear Information System (INIS)

    Assedo, R.; Dubourg, M.; Livolant, M.; Epstein, A.

    1975-01-01

    In the late 1971, the CEA and FRAMATOME decided to undertake a comprehensive joint program of studying the vibration behavior of PWR internals of the 900 MWe, 50 cycle, 3 loop reactor series being built by FRAMATOME in France. The PWR reactor internals are submitted to several sources of excitation during normal operation. Two main sources of excitation may effect the internals behavior: the large flow turbulences which could generate various instabilities such as: vortex shedding: the pump pressure fluctuations which could generate acoustic noise in the circuit at frequencies corresponding to shaft speed frequencies or blade passing frequencies, and their respective harmonics. The flow induced vibrations are of complex nature and the approach selected, for this comprehensive program, is semi-empirical and based on both theoretical analysis and experiments on a reduced scale model and full scale internals. The experimental support of this program consists of: the SAFRAN test loop which consists of an hydroelastic similitude of a 1/8 scale model of a PWR; harmonic vibration tests in air performed on full scale reactor internals in the manufacturing shop; the GENNEVILLIERS facilities which is a full flow test facility of primary pump; the measurements carried out during start up on the Tihange reactor. This program will be completed in April 1975. The results of this program, the originality of which consists of studying separately the effects of random excitations and acoustic noises, on the internals behavior, and by establishing a comparison between experiments and analysis, will bring a major contribution for explaining the complex vibration phenomena occurring in a PWR

  5. The latest full-scale PWR simulator in Japan

    International Nuclear Information System (INIS)

    Nishimuru, Y.; Tagi, H.; Nakabayashi, T.

    2004-01-01

    The latest MHI Full-scale Simulator has an excellent system configuration, in both flexibility and extendability, and has highly sophisticated performance in PWR simulation by the adoption of CANAC-II and PRETTY codes. It also has an instructive character to display the plant's internal status, such as RCS condition, through animation. Further, the simulation has been verified to meet a functional examination at model plant, and with a scale model test result in a two-phase flow event, after evaluation for its accuracy. Thus, the Simulator can be devoted to a sophisticated and broad training course on PWR operation. (author)

  6. Experiments on natural circulation during PWR severe accidents and their analysis

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Stewart, W.A.; Sha, W.T.

    1988-01-01

    Buoyancy-induced natural circulation flows will occur during the early-part of PWR high pressure accident scenarios. These flows affect several key parameters; in particular, the course of such accidents will most probably change due to local failures occurring in the primary coolant system (CS) before substantial core degradation. Natural circulation flow patterns were measured in a one-seventh scale PWR PCS facility at Westinghouse RandD laboratories. The measured flow and temperature distributions are report in this paper. The experiments were analyzed with the COMMIX code and good agreement was obtained between data and calculations. 10 refs., 8 figs., 2 tabs

  7. Reduced scale PWR passive safety system designing by genetic algorithms

    International Nuclear Information System (INIS)

    Cunha, Joao J. da; Alvim, Antonio Carlos M.; Lapa, Celso Marcelo Franklin

    2007-01-01

    This paper presents the concept of 'Design by Genetic Algorithms (DbyGA)', applied to a new reduced scale system problem. The design problem of a passive thermal-hydraulic safety system, considering dimensional and operational constraints, has been solved. Taking into account the passive safety characteristics of the last nuclear reactor generation, a PWR core under natural circulation is used in order to demonstrate the methodology applicability. The results revealed that some solutions (reduced scale system DbyGA) are capable of reproducing, both accurately and simultaneously, much of the physical phenomena that occur in real scale and operating conditions. However, some aspects, revealed by studies of cases, pointed important possibilities to DbyGA methodological performance improvement

  8. Fire experiences: principal lessons learned, application in PWR power plants

    International Nuclear Information System (INIS)

    Schoemacker, M.

    1984-01-01

    The article reviews the principal design rules to be borne in mind for PWR nuclear units installation. These rule takes into account: the specific character of materials involved (safety aspect for nuclear construction), experience acquired as a result of fires in EDF production units, and the results obtained from tests carried out by the EDF at Fort de Chelles between 1980 and 1982, especially in the field of PVC cables [fr

  9. Gadolinia experience and design for PWR fuel cycles

    International Nuclear Information System (INIS)

    Stephenson, L. C.

    2000-01-01

    The purpose of this paper is to describe Siemens Power Corporation's (SPC) current experience with the burnable absorber gadolinia in PWR fuel assemblies, including optimized features of SPC's PWR gadolinia designs, and comparisons with other burnable absorbers. Siemens is the world leader in PWR gadolinia experience. More than 5,900 Siemens PWR gadolinia-bearing fuel assemblies have been irradiated. The use of gadolinia-bearing fuel provides significant flexibility in fuel cycle designs, allows for low radial leakage fuel management and extended operating cycles, and reduces BOC (beginning-of-cycle) soluble boron concentrations. The optimized use of an integral burnable neutron absorber is a design feature which provides improved economic performance for PWR fuel assemblies. This paper includes a comparison between three different types of integral burnable absorbers: gadolinia, Zirconium diboride and erbia. Fuel cycle design studies performed by Siemens have shown that the enrichment requirements for 18-24 month fuel cycles utilizing gadolinia or zirconium diboride integral fuel burnable absorbers can be approximately the same. Although a typical gadolinia residual penalty for a cycle design of this length is as low as 0.02-0.03 wt% U-235, the design flexibility of gadolinia allows for very aggressive low-leakage core loading plans which reduces the enrichment requirements for gadolinia-bearing fuel. SPC has optimized its use of gadolinia in PWR fuel cycles. Typically, low (2-4) weight percent Gd 2 O 3 is used for beginning to middle of cycle reactivity hold down as well as soluble boron concentration holddown at BOC. Higher concentrations of Gd 2 O 3 , such as 6 and 8 wt%, are used to control power peaking in assemblies later in the cycle. SPC has developed core strategies that maximize the use of lower gadolinia concentrations which significantly reduces the gadolinia residual reactivity penalty. This optimization includes minimizing the number of rods with

  10. A reduced scale two loop PWR core designed with particle swarm optimization technique

    International Nuclear Information System (INIS)

    Lima Junior, Carlos A. Souza; Pereira, Claudio M.N.A; Lapa, Celso M.F.; Cunha, Joao J.; Alvim, Antonio C.M.

    2007-01-01

    Reduced scale experiments are often employed in engineering projects because they are much cheaper than real scale testing. Unfortunately, designing reduced scale thermal-hydraulic circuit or equipment, with the capability of reproducing, both accurately and simultaneously, all physical phenomena that occur in real scale and at operating conditions, is a difficult task. To solve this problem, advanced optimization techniques, such as Genetic Algorithms, have been applied. Following this research line, we have performed investigations, using the Particle Swarm Optimization (PSO) Technique, to design a reduced scale two loop Pressurized Water Reactor (PWR) core, considering 100% of nominal power and non accidental operating conditions. Obtained results show that the proposed methodology is a promising approach for forced flow reduced scale experiments. (author)

  11. FLUOLE-2: An Experiment for PWR Pressure Vessel Surveillance

    Directory of Open Access Journals (Sweden)

    Thiollay Nicolas

    2016-01-01

    Full Text Available FLUOLE-2 is a benchmark-type experiment dedicated to 900 and 1450 MWe PWR vessels surveillance dosimetry. This two-year program started in 2014 and will end in 2015. It will provide precise experimental data for the validation of the neutron spectrum propagation calculation from core to vessel. It is composed of a square core surrounded by a stainless steel baffe and internals: PWR barrel is simulated by steel structures leading to different steel-water slides; two steel components stand for a surveillance capsule holder and for a part of the pressure vessel. Measurement locations are available on the whole experimental structure. The experimental knowledge of core sources will be obtained by integral gamma scanning measurements directly on fuel pins. Reaction rates measured by calibrated fission chambers and a large set of dosimeters will give information on the neutron energy and spatial distributions. Due to the low level neutron flux of EOLE ZPR a special, high efficiency, calibrated gamma spectrometry device will be used for some dosimeters, allowing to measure an activity as low as 7. 10−2 Bq per sample. 103mRh activities will be measured on an absolute calibrated X spectrometry device. FLUOLE-2 experiment goal is to usefully complete the current experimental benchmarks database used for the validation of neutron calculation codes. This two-year program completes the initial FLUOLE program held in 2006–2007 in a geometry representative of 1300 MWe PWR.

  12. The N4 plant: culmination of French PWR experience

    International Nuclear Information System (INIS)

    Bellet, J.; Houyez, A.; Journet, J.; Pierrard, J.H.

    1985-01-01

    The model N4 series of 1400MWe class PWR plants has been developed in France from a unique base of technical and operating experience. It meets the French government's requirement for a reactor free of constraints due to licensing agreements with overseas companies, with enhanced safety features and incorporating the lessons of Three Mile Island. In particular, improvements have been made to the reactor vessel, the steam generators, the primary pumps and control systems. The units are capable of daily load following and extended operation between refuelling. The N4 plant includes a new design of turbine-generator. (author)

  13. MOX and UOX PWR fuel performances EDF operating experience

    International Nuclear Information System (INIS)

    Provost, Jean-Luc; Debes, Michel

    2005-01-01

    Based on a large program of experimentations implemented during the 90s, the industrial achievement of new FAs designs with increased performances opens up new prospects. The currently UOX fuels used on the 58 EDF PWR units are now authorized up to a maximum FA burn-up of 52 GWd/t with a large experience from 45 to 50 GWd/t. Today, the new products, along with the progress made in the field of calculation methods, still enable to increase further the fuel performances with respect to the safety margins. Thus, the conditions are met to implement in the next years new fuel managements on each NPPs series of the EDF fleet with increased enrichment (up to 4.5%) and irradiation limits (up to 62 GWd/t). The recycling of plutonium is part of EDF's reprocessing/recycling strategy. Up to now, 20 PWR 900 MW reactors are managed in MOX hybrid management. The feedback experience of 18 years of PWR operation with MOX is satisfactory, without any specific problem regarding manoeuvrability or plant availability. EDF is now looking to introduce MOX fuels with a higher plutonium content (up to 8.6%) equivalent to natural uranium enriched to 3.7%. It is the goal of the MOX Parity core management which achieve balance of MOX and UOX fuel performance with a significant increase of the MOX average discharge burn-up (BU max: 52 GWd/t for MOX and UOX). The industrial maturity of new FAs designs, with increased performances, allows the implementation in the next years of new fuel managements on each NPPs series of the EDF fleet. The scheduling of the implementation of the new fuel managements on the PWRs fleet is a great challenge for EDF, with important stakes: the nuclear KWh cost decrease with the improvement of the plant operation performance. (author)

  14. A simple analytical scaling method for a scaled-down test facility simulating SB-LOCAs in a passive PWR

    International Nuclear Information System (INIS)

    Lee, Sang Il

    1992-02-01

    A Simple analytical scaling method is developed for a scaled-down test facility simulating SB-LOCAs in a passive PWR. The whole scenario of a SB-LOCA is divided into two phases on the basis of the pressure trend ; depressurization phase and pot-boiling phase. The pressure and the core mixture level are selected as the most critical parameters to be preserved between the prototype and the scaled-down model. In each phase the high important phenomena having the influence on the critical parameters are identified and the scaling parameters governing the high important phenomena are generated by the present method. To validate the model used, Marviken CFT and 336 rod bundle experiment are simulated. The models overpredict both the pressure and two phase mixture level, but it shows agreement at least qualitatively with experimental results. In order to validate whether the scaled-down model well represents the important phenomena, we simulate the nondimensional pressure response of a cold-leg 4-inch break transient for AP-600 and the scaled-down model. The results of the present method are in excellent agreement with those of AP-600. It can be concluded that the present method is suitable for scaling the test facility simulating SB-LOCAs in a passive PWR

  15. Steam Generator Chemical Cleaning Application: Korean Experience in PWR NPP

    International Nuclear Information System (INIS)

    Hwang, In-Ho; Varrin-Jr, Robert-D.; Little, Michael-J.; Oh, Yeon-Ok; Choo, Seong-Jib; Park, Jin-Hyeok

    2012-09-01

    Korea Hydro and Nuclear Power (KHNP) performed an EPRI/SGOG chemical cleaning of the secondary side of the steam generators at Ulchin Unit 3 (UCN3) in March 2011 and at Ulchin Unit 4 (UCN4) in September 2011. The steam generator chemical cleaning (SGCC) was performed with venting at the top-of-tube sheet (TTS) and at tube support plates (TSPs) 4, 5, 6, 7, 8, 9, and 10. A primary objective of this SGCC was to address outer diameter stress corrosion cracking (ODSCC), which has been observed at the TTS and TSPs in the UCN3 SGs. The EPRI/SGOG process has been shown to effectively reduce prevailing ODSCC rates at the TTS and TSPs, particularly when applied with periodic venting in this application. This was the first full-length SGCC campaign with venting performed in Korea. Ulchin Unit 3 commenced commercial operation in August 1998 and Ulchin Unit 4 commenced commercial operation in December 1999. UCN3 and UCN4 are a two-loop pressurized water reactor (PWR) of the Korea Standard Nuclear Plant (KSNP) design. The SGs contain high-temperature mill annealed (HTMA) Alloy 600 tubing and are similar in design to the Combustion Engineering CE-80. The KSNP SGs have been susceptible to outer diameter stress corrosion cracking (ODSCC), which is consistent with operating experience for other SGs containing Alloy 600HTMA tubing material. The UCN3/4 SGs have recently begun to experience ODSCC. Hankook Jungsoo Industries Co., Ltd (HaJI) was selected as the cleaning vendor by KHNP. To date, HaJI has completed five Advanced Scale Conditioning Agent (ASCA) cleaning applications and two EPRI/SGOG Steam Generator Chemical Cleaning (SGCC) campaigns for KHNP. The goal of total deposit removal of the applications were successfully achieved and the amounts are 3,579 kg at UCN3 and 3,786 kg at UCN4 which values were estimated before each cleaning by analysing ECT signal and liquid samples from the SGs. The deposits from the SGs were primarily composed of magnetite. There were no chemical

  16. 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)

  17. Numerical modeling of in-vessel melt water interaction in large scale PWR`s

    Energy Technology Data Exchange (ETDEWEB)

    Kolev, N.I. [Siemens AG, KWU NA-M, Erlangen (Germany)

    1998-01-01

    This paper presents a comparison between IVA4 simulations and FARO L14, L20 experiments. Both experiments were performed with the same geometry but under different initial pressures, 51 and 20 bar respectively. A pretest prediction for test L21 which is intended to be performed under an initial pressure of 5 bar is also presented. The strong effect of the volume expansion of the evaporating water at low pressure is demonstrated. An in-vessel simulation for a 1500 MW el. PWR is presented. The insight gained from this study is: that at no time are conditions for the feared large scale melt-water intermixing at low pressure in force, with this due to the limiting effect of the expansion process which accelerates the melt and the water into all available flow paths. (author)

  18. Natural circulation in a scaled PWR integral test facility

    International Nuclear Information System (INIS)

    Kiang, R.L.; Jeuck, P.R. III

    1987-01-01

    Natural circulation is an important mechanism for cooling a nuclear power plant under abnormal operating conditions. To study natural circulation, we modeled a type of pressurized water reactor (PWR) that incorporates once-through steam generators. We conducted tests of single-phase natural circulations, two-phase natural circulations, and a boiler condenser mode. Because of complex geometry, the natural circulations observed in this facility exhibit some phenomena not commonly seen in a simple thermosyphon loop

  19. Reduced scaling of thermal-hydraulic circuits for studies of PWR reactors natural circulation

    International Nuclear Information System (INIS)

    Botelho, D.A.

    1993-01-01

    The Ishii et al. hydrodynamic similarity criteria for natural circulation were used for scaling reduced models of prototype passive residual heat removal system of a 600 M We PWR. The physical scales of the thermohydraulic parameters obtained presented a reasonable agreement when compared with simplified analytic models of the systems. (author)

  20. Short-term calculations to supplement the RS 16 B PWR experiments with internals (PWR1 to PWR5), using the LECK 4 computer code

    International Nuclear Information System (INIS)

    Hughes, G.; Mueller, R.

    1980-03-01

    Within the framework of research project RS 16 B sponsored by the German BMFT a series of a blowdown experiments, DWR1 to DWR5, were performed using a vessel with dummy internals under conditions similar to those in a PWR. The prime objective of these experiments was the investigation of the highly transient blowdown phenomena in the discharge nozzle and the determination of the induced loads on the internals. As a partner in the project, KWU carried out both pre-test predictions and post-test analyses of these experiments using, among others, the computer code LECK 4. For the most severe blowdown test DWR5, the influence of the most important model parameters on the blowdown analysis was investigated in detail. These investigations suggest that, similar to the long-term analyses, calculations using the homogeneous critical flow model would improve agreement between calculation and experiment. (orig./RW) [de

  1. Operating experience with PWR in the FRG (Federal Republic of Germany)

    International Nuclear Information System (INIS)

    Cramer, H.

    1980-01-01

    Operating experiences with PWR's in the Federal Republic of Germany has been exclusively with KWU turnkey power plants with U tube steam generators. Such experience started with the 345 MW Obrigheim plant in 1968 and includes the 670 MW Stade plant, the 1200/1300 MW Biblis plants, The 900 MW Neckarwestheim and the 1300 MW Unterweser plants. (E.G.) [pt

  2. Experiment of the downcomer effective water head during a reflood phase of PWR LOCA

    International Nuclear Information System (INIS)

    Sudo, Yukio; Murao, Yoshio

    1978-12-01

    The results and analysis are described of a downcomer effective water head experiment. Downcomer effective water head is the driving force to feed an emergency coolant to the core during a reflood phase of PWR LOCA. The test rig has dimensions of the full-scale height and gap. Experimental conditions are: downcomer wall temperature = 250 0 -- 300 0 C, back pressure = 1 atm, coolant temperature = 98 0 -- 100 0 C, extraction water velocity = 0 -- 2 cm/s, and gap size = 200 mm. The effective water head histories obtained by experiment were compared with those predicted from the heat release from the downcomer walls. The heat release was calculated from the temperature histories indicated by thermocouples instrumented in and on the walls during experiment. The following were revealed: (1) The relation of heat flux and superheat (q vs ΔT sub(s)) obtained in the experiment is much different from that in pool boiling. (2) The predicted effective water head is in good agreement with the experimental one after 120 sec from the initiation of coolant injection. (3) The effect of extraction water velocity is negligible. (4) The effect of initial wall temperatures is evident. (author)

  3. Generation of SCALE 6 Input Data File for Cross Section Library of PWR Spent Fuel

    International Nuclear Information System (INIS)

    Jeong, Chang Joon; Cho, Dong Keun

    2010-11-01

    In order to obtain the cross section libraries of the Korean Pressurized water reactor (PWR) spent fuel (SF), SCALE 6 code input files have been generated. The PWR fuel data were obtained from the nuclear design report (NDR) of the current operating PWRs. The input file were prepared for 16 fuel types such as 4 types of Westinghouse 14x14, 3 types of OPR-1000 16x16, 4 types of Westinghouse 16x16, and 6 types of Westinghouse 17x17. For each fuel type, 5 kinds of fuel enrichments have been considered such as 1.5, 2.0 ,3.0, 4.0 and 5.0 wt%. In the SCALE 6 calculation, a ENDF-V 44 group was used. The 25 burnup step until 72000 MWD/T was used. A 1/4 symmetry model was used for 16x16 and 17x17 fuel assembly, and 1/2 symmetry model was used for 14x14 fuel assembly The generated cross section libraries will be used for the source-term analysis of the PWR SF

  4. Experiments on the risk of sump plugging on French PWR

    International Nuclear Information System (INIS)

    Armand, Y.; Mattei, J.M.; Vicena, I.; Gubco, V.; Batalik, J.; Murani, J.; Davydov, M.; Melikhov, O.I.; Blinkov, V.N.

    2003-01-01

    The 'Institut de Radioprotection et de Surete Nucleaire' has decided to perform an experimental program of studies on the risk of sump plugging in a 900 MWe pressurized water reactor (PWR). A general overview of the literature has been conducted between October 1999 and November 2000, ending by the definition of an approach methodology and the written of technical specifications. The problem identified gave rise to an European call for tenders leading to four contracts signed in november 2001. Three contracts are managed by VUEZ (Slovakia), the last by EREC (Russia). The methodology and the facilities devoted to these studies are presented in this paper, the preliminary results observed are also presented. The objectives for 2003 to finalize the program and the safety assessment are presented. (authors)

  5. 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

  6. SCALE 5.1 Predictions of PWR Spent Nuclear Fuel Isotopic Compositions

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-03-01

    The purpose of this calculation report is to document the comparison to measurement of the isotopic concentrations for pressurized water reactor (PWR) spent nuclear fuel determined with the Standardized Computer Analysis for Licensing Evaluation (SCALE) 5.1 (Ref. ) epletion calculation method. Specifically, the depletion computer code and the cross-section library being evaluated are the twodimensional (2-D) transport and depletion module, TRITON/NEWT,2, 3 and the 44GROUPNDF5 (Ref. 4) cross-section library, respectively, in the SCALE .1 code system.

  7. Estimation of PWR spent fuel composition using SCALE and SWAT code systems

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Hee Sung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Kenya, Suyama; Hiroshi, Okuno [Japan Atomic Energy Research Institute, Tokyo (Japan)

    2001-05-01

    The isotopic composition calculations were performed for 26 spent fuel samples from Obrigheim PWR reactor and 55 spent fuel samples from 7 PWR reactors using SCALE4.4 SAS2H with 27, 44 and 238 group cross-section libraries and SWAT with 107 group cross-section library. For convenience, the ratio of the measured to calculated value was used as a parameter. The four kinds of the calculation results were compared with the measured data. For many important nuclides for burnup credit criticality safety evaluation, the four methods applied in this study showed good coincidence with measurements in general. More precise observations showed the following results. Less unity ratios were found for Pu-239 and -241 for selected 16 samples out of the 26 samples from Obrigheim reactor. Larger than unity ratios were found for Am-241 for both the 16 and 55 samples. Larger than unity ratios were found for Sm-149 for the 55 samples. In the case of 26 sample SWAT was generally accompanied by larger ratios than those of SAS2H with some exceptions. Based on the measured-to-calculated ratios for 71 samples of a combined set in which 16 selected samples and 55 samples were included, the correction factors that should be multiplied to the calculated isotopic compositions were generated for a conservative estimate of the neutron multiplication factor of a system containing PWR spent fuel, taking burnup credit into account.

  8. Medium scale test study of chemical cleaning technique for secondary side of SG in PWR

    International Nuclear Information System (INIS)

    Zhang Mengqin; Zhang Shufeng; Yu Jinghua; Hou Shufeng

    1997-08-01

    The medium scale test study of chemical cleaning technique for removing corrosion product (Fe 3 O 4 ) in secondary side of SG in PWR has been completed. The test has been carried out in a medium scale test loop. The medium scale test evaluated the effect of the chemical cleaning technique (temperature, flow rate, cleaning time, cleaning process), the state of corrosion product deposition on magnetite (Fe 3 O 4 ) solubility and safety of materials of SG in cleaning process. The inhibitor component of chemical cleaning agent has been improved by electrochemical linear polarization method, the effect of inhibitor on corrosion resistance of materials have been examined in the medium scale test loop, the most components of chemical cleaning agent have been obtained, the EDTA is main component in cleaning agent. The electrochemical method for monitor corrosion of materials during cleaning process has been completed in the laboratory. The study of the medium scale test of chemical cleaning technique have had the optimum chemical cleaning technique for remove corrosion product in SG secondary side of PWR. (9 refs., 4 figs., 11 tabs.)

  9. A feasibility assessment for incorporating of passive RHRS into large scale active PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S O; Sub, S Y; Kim, Y S; Chang, M H; Park, J K [Korea Atomic Energy Research Inst., Taejon (Korea, Republic of)

    1996-12-01

    A feasibility study was carried out for the possible incorporation of passive RHRS (Residual Heat Removal System) into a large-scale of active PWR plant. Four kinds of system configurations were considered. For each case its performance and impacts on plant safety, cost, licensing, operation and maintenance were evaluated. The evaluation came up with a finding of PRHRS with a gravity feed tank as most probable design concept. However, considering rearrangement of structure and pipe routing inside and outside containment, it is concluded that implementation of the PRHRS concept into well developed active plants is not desirable at present. (author). 6 refs, 7 figs, 1 tab.

  10. A feasibility assessment for incorporating of passive RHRS into large scale active PWR

    International Nuclear Information System (INIS)

    Kim, S.O.; Sub, S.Y.; Kim, Y.S.; Chang, M.H.; Park, J.K.

    1996-01-01

    A feasibility study was carried out for the possible incorporation of passive RHRS (Residual Heat Removal System) into a large-scale of active PWR plant. Four kinds of system configurations were considered. For each case its performance and impacts on plant safety, cost, licensing, operation and maintenance were evaluated. The evaluation came up with a finding of PRHRS with a gravity feed tank as most probable design concept. However, considering rearrangement of structure and pipe routing inside and outside containment, it is concluded that implementation of the PRHRS concept into well developed active plants is not desirable at present. (author). 6 refs, 7 figs, 1 tab

  11. Zircaloy oxidation and cladding deformation in PWR-specific CORA experiments

    International Nuclear Information System (INIS)

    Minato, K.; Hering, W.; Hagen, S.

    1991-07-01

    Out-of-pile bundle experiments (zircaloy 4) are performed in the CORA facility to investigate the behavior of PWR fuel elements during severe fuel damage (SFD) accidents. Within the international cooperation the most significant phenomena such as cladding deformation, oxidation (especially the zirconium/steam reaction), melt formation, melt release, and relocation which were found in all tests have been analyzed. (orig./MM) [de

  12. Experiments for simulating a great leak in the primary coolant circuit of a PWR type reactor

    International Nuclear Information System (INIS)

    Liebig, E.

    1977-01-01

    A loss of coolant accident is to be simulated on a high pressure test rig. The accident is initiated by an externally induced rupture of a pair of rupture-disks installed in a coolant ejection device. Several problems of simulating leaks in the primary coolant circuit of PWR type reactors are dealt with. The selection of appropriate rupture-disks for such experiments is described

  13. Thermal stability of morpholine, AMP and sarcosine in PWR secondary systems. Laboratory and loop experiments

    International Nuclear Information System (INIS)

    Feron, D.; Lambert, I.

    1991-01-01

    Laboratory and loop tests have been carried out in order to investigate the thermal stability of three amines (morpholine, AMP and sarcosine) in PWR secondary conditions. Laboratory experiments have been performed in a titanium autoclave at 300 deg C. The results pointed out high thermal decomposition rates of AMP and sarcosine. A decomposition mechanism is proposed for the 3 amines. Loop tests have been performed in order to compare steam cycle conditioning with ammonia, morpholine and AMP. The amine concentrations and the decomposition products such as acetate and formate have been followed around the secondary circuit of the ORION loop which reproduces the main physico-chemical characteristics of a PWR secondary circuit. These concentrations are reported together with the evolution of cationic conductivities. The influence of oxygen concentration on amine thermal stability has been observed. Results are expressed also in terms of decomposition rates and of relative volatility

  14. Estimation of fire frequency from PWR operating experience

    International Nuclear Information System (INIS)

    Bertrand, R.; Bonneval, F.; Barrachin, G.; Bonino, F.

    1998-01-01

    In the framework of a fire probabilistic safety assessment (Fire PSA), the French Institute for Nuclear Safety and Protection (IPSN) has developed a method for estimating the frequency of fire in a nuclear power plant room. This method is based on the analysis of French Pressurized Water Reactors operating experience. The method adopted consists is carrying out an in-depth analysis of fire-related incidents. A database has been created including 202 fire events reported in 900 MWe and 1300 MWe reactors from the start of their commercial operation up to the first of March 1994, which represents a cumulated service life of 508 reactor-years. For each reported fire, several data were recorded among which: The operating state of the reactor in the stage preceding the fire, the building in which the fire broke out, the piece of equipment or the human intervention which caused the fire. Operating experience shows that most fires are initiated by electrical problems (short-circuits, arcing, faulty contacts, etc.) and that human intervention also plays an important role (grinding, cutting, welding, cleaning, etc.). A list of equipment and of human interventions which proved to be possible fire sources was therefore drawn up. the items of this list were distributed in 19 reference groups defined by taking into account the nature of the potential ignition source (transformers, electrical cabinets, pumps, fans, etc.). The fire frequency assigned to each reference group was figured out using the operating experience information of the database. The fire frequency in a room is considered to be made out of two contributions: one due to equipment which is proportional to the number of pieces of equipment from each reference group contained in the room, and a second one which is due to human interventions and assumed to be uniform throughout the reactor. Formulas to assess the fire frequencies in a room, the reactor being in a shutdown state or at power, are then proposed

  15. Experience of Areva in fuel services for PWR and BWR

    International Nuclear Information System (INIS)

    Morales, I.

    2015-01-01

    AREVA being an integrated supplier of fuel assemblies has included in its strategy to develop services and solutions to customers who desire to improve the performance and safety of their fuel. These services go beyond the simple 'after sale' services that can be expected from a fuel supplier: The portfolio of AREVA includes a wide variety of services, from scientific calculations to fuel handling services in a nuclear power plant. AREVA is committed to collaborate and to propose best-in-class solutions that really make the difference for the customer, based on 40 years of Fuel design and manufacturing experience. (Author)

  16. SCALE6 Hybrid Deterministic-Stochastic Shielding Methodology for PWR Containment Calculations

    International Nuclear Information System (INIS)

    Matijevic, Mario; Pevec, Dubravko; Trontl, Kresimir

    2014-01-01

    The capabilities and limitations of SCALE6/MAVRIC hybrid deterministic-stochastic shielding methodology (CADIS and FW-CADIS) are demonstrated when applied to a realistic deep penetration Monte Carlo (MC) shielding problem of full-scale PWR containment model. The ultimate goal of such automatic variance reduction (VR) techniques is to achieve acceptable precision for the MC simulation in reasonable time by preparation of phase-space VR parameters via deterministic transport theory methods (discrete ordinates SN) by generating space-energy mesh-based adjoint function distribution. The hybrid methodology generates VR parameters that work in tandem (biased source distribution and importance map) in automated fashion which is paramount step for MC simulation of complex models with fairly uniform mesh tally uncertainties. The aim in this paper was determination of neutron-gamma dose rate distribution (radiation field) over large portions of PWR containment phase-space with uniform MC uncertainties. The sources of ionizing radiation included fission neutrons and gammas (reactor core) and gammas from activated two-loop coolant. Special attention was given to focused adjoint source definition which gave improved MC statistics in selected materials and/or regions of complex model. We investigated benefits and differences of FW-CADIS over CADIS and manual (i.e. analog) MC simulation of particle transport. Computer memory consumption by deterministic part of hybrid methodology represents main obstacle when using meshes with millions of cells together with high SN/PN parameters, so optimization of control and numerical parameters of deterministic module plays important role for computer memory management. We investigated the possibility of using deterministic module (memory intense) with broad group library v7 2 7n19g opposed to fine group library v7 2 00n47g used with MC module to fully take effect of low energy particle transport and secondary gamma emission. Compared with

  17. Basic investigation of particle swarm optimization performance in a reduced scale PWR passive safety system design

    International Nuclear Information System (INIS)

    Cunha, Joao J. da; Lapa, Celso Marcelo F.; Alvim, Antonio Carlos M.; Lima, Carlos A. Souza; Pereira, Claudio Marcio do N.A.

    2010-01-01

    This work presents a methodology to investigate the viability of using particle swarm optimization technique to obtain the best combination of physical and operational parameters that lead to the best adjusted dimensionless groups, calculated by similarity laws, that are able to simulate the most relevant physical phenomena in single-phase flow under natural circulation and to offer an appropriate alternative reduced scale design for reactor primary loops with this flow characteristics. A PWR reactor core, under natural circulation, based on LOFT test facility, was used as the case study. The particle swarm optimization technique was applied to a problem with these thermo-hydraulics conditions and results demonstrated the viability and adequacy of the method to design similar systems with these characteristics.

  18. Basic investigation of particle swarm optimization performance in a reduced scale PWR passive safety system design

    Energy Technology Data Exchange (ETDEWEB)

    Cunha, Joao J. da [Eletronuclear Eletrobras Termonuclear, Gerencia de Analise de Seguranca Nuclear, Rua da Candelaria, 65, 7o andar. Centro, Rio de Janeiro 20091-906 (Brazil); Lapa, Celso Marcelo F., E-mail: lapa@ien.gov.b [Instituto de Engenharia Nuclear, Divisao de Reatores/PPGIEN, P.O. Box 68550, Rua Helio de Almeida 75 Cidade Universitaria, Ilha do Fundao, Rio de Janeiro 21941-972 (Brazil); Instituto Nacional de Ciencia e Tecnologia de Reatores Nucleares Inovadores (Brazil); Alvim, Antonio Carlos M. [Universidade Federal do Rio de Janeiro, COPPE/Nuclear, P.O. Box 68509, Cidade Universitaria, Ilha do Fundao s/n, Rio de Janeiro 21945-970 (Brazil); Instituto Nacional de Ciencia e Tecnologia de Reatores Nucleares Inovadores (Brazil); Lima, Carlos A. Souza [Instituto de Engenharia Nuclear, Divisao de Reatores/PPGIEN, P.O. Box 68550, Rua Helio de Almeida 75 Cidade Universitaria, Ilha do Fundao, Rio de Janeiro 21941-972 (Brazil); Instituto Politecnico, Universidade do Estado do Rio de Janeiro, Pos-Graduacao em Modelagem Computacional, Rua Alberto Rangel, s/n, Vila Nova, Nova Friburgo 28630-050 (Brazil); Pereira, Claudio Marcio do N.A. [Instituto de Engenharia Nuclear, Divisao de Reatores/PPGIEN, P.O. Box 68550, Rua Helio de Almeida 75 Cidade Universitaria, Ilha do Fundao, Rio de Janeiro 21941-972 (Brazil); Instituto Nacional de Ciencia e Tecnologia de Reatores Nucleares Inovadores (Brazil)

    2010-03-15

    This work presents a methodology to investigate the viability of using particle swarm optimization technique to obtain the best combination of physical and operational parameters that lead to the best adjusted dimensionless groups, calculated by similarity laws, that are able to simulate the most relevant physical phenomena in single-phase flow under natural circulation and to offer an appropriate alternative reduced scale design for reactor primary loops with this flow characteristics. A PWR reactor core, under natural circulation, based on LOFT test facility, was used as the case study. The particle swarm optimization technique was applied to a problem with these thermo-hydraulics conditions and results demonstrated the viability and adequacy of the method to design similar systems with these characteristics.

  19. Methods of studying oxide scales grown on zirconium alloys in autoclaves and in a PWR

    International Nuclear Information System (INIS)

    Blank, H.; Bart, G.; Thiele, H.

    1992-01-01

    The analysis of water-side corrosion of zirconium alloys has been a field of research for more than 25 years, but the details of the mechanisms involved still cannot be put into a coherent picture. Improved methods are required to establish the details of the microstructure of the oxide scales. A new approach has been made for a general analysis of oxide specimens from scales grown on the zirconium-based cladding alloys of PWR rods in order to analyse the morphology of these scales, the topography of the oxide/metal interface and the crystal structures close to this interface: a) Instead of using the conventional pickling solutions, the Zr-alloys are dissolved using a 'softer' solution (Br 2 in an organic solvent) in order to avoid damage to the oxide at the oxide/metal interface to be analysed by SEM (scanning electron microscopy). A second advantage of this method is easy etching of the grain structure of Zr-alloys for SEM analysis; b) By using the particular properties of the oxide scales, the corrosion-rate-determining innermost part of the oxide layer at the oxide/metal interface can be separated from the rest of the oxide scale and then analysed by SEM, STEM (scanning transmission electron microscopy), TEM (transmission electron microscopy) and electron diffraction after dissolution of the alloy. Examples are given from oxides grown on Zr-alloys in a pressurized water reactor and in autoclaves. (author) 8 figs., 3 tabs., 9 refs

  20. PWR water chemistry controls: a perspective on industry initiatives and trends relative to operating experience and the EPRI PWR water chemistry guidelines

    International Nuclear Information System (INIS)

    Fruzzetti, K.; Choi, S.; Haas, C.; Pender, M.; Perkins, D.

    2010-01-01

    An effective PWR water chemistry control program must address the following goals: Minimize materials degradation (e.g., PWSCC, corrosion of fuel, corrosion damage of steam generator (SG) tubes); Maintain fuel integrity and good performance; Minimize corrosion product transport (e.g., transport and deposition on the fuel, transport into the SGs where it can foul tube surfaces and create crevice environments for the concentration of corrosive impurities); Minimize dose rates. Water chemistry control must be optimized to provide overall improvement considering the sometimes variant constraints of the goals listed above. New technologies are developed for continued mitigation of materials degradation, continued fuel integrity and good performance, continued reduction of corrosion product transport, and continued minimization of plant dose rates. The EPRI chemistry program, in coordination with other EPRI programs, strives to improve these areas through application of chemistry initiatives, focusing on these goals. This paper highlights the major initiatives and issues with respect to PWR primary and secondary system chemistry and outlines the recent, on-going, and proposed work to effectively address them. These initiatives are presented in light of recent operating experience, as derived from EPRI's PWR chemistry monitoring and assessment program, and EPRI's water chemistry guidelines. (author)

  1. PWR Secondary Water Chemistry Control Status: A Summary of Industry Initiatives, Experience and Trends Relative to the EPRI PWR Secondary Water Chemistry Guidelines

    International Nuclear Information System (INIS)

    Fruzzetti, Keith; Choi, Samuel

    2012-09-01

    The latest revision of the EPRI Pressurized Water Reactor (PWR) Secondary Water Chemistry Guidelines was issued in February 2009. The Guidelines continue to focus on minimizing stress corrosion cracking (SCC) of steam generator tubes, as well as minimizing degradation of other major components / subsystems of the secondary system. The Guidelines provide a technically-based framework for a plant-specific and effective PWR secondary water chemistry program. With the issuance of Revision 7 of the Guidelines in 2009, many plants have implemented changes that allow greater flexibility on startup. For example, the previous Guidelines (Revision 6) contained a possible low power hold at 5% power and a possible mid power hold at approximately 30% power based on chemistry constraints. Revision 7 has established a range over which a plant-specific value can be chosen for the possible low power hold (between 5% and 15%) and mid power hold (between 30% and 50%). This has provided plants the ability to establish significant plant evolutions prior to reaching the possible power hold; such as establishing seal steam to the condenser, placing feed pumps in service, or initiating forward flow of heater drains. The application of this flexibility in the industry will be explored. This paper also highlights the major initiatives and industry trends with respect to PWR secondary chemistry; and outlines the recent work to effectively address them. These will be presented in light of recent operating experience, as derived from EPRI's PWR Chemistry Monitoring and Assessment (CMA) program (which contains more than 400 cycles of operating chemistry data). (authors)

  2. PWR core and spent fuel pool analysis using scale and nestle

    International Nuclear Information System (INIS)

    Murphy, J. E.; Maldonado, G. I.; St Clair, R.; Orr, D.

    2012-01-01

    The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

  3. PWR core and spent fuel pool analysis using scale and nestle

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, J. E.; Maldonado, G. I. [Dept. of Nuclear Engineering, Univ. of Tennessee, Knoxville, TN 37996-2300 (United States); St Clair, R.; Orr, D. [Duke Energy, 526 S. Church St, Charlotte, NC 28202 (United States)

    2012-07-01

    The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

  4. PKL/K9, Refill and Reflood Experiment in a Simulated PWR Primary System (PKL)

    International Nuclear Information System (INIS)

    1981-01-01

    1 - Description of test facility: PKL-facility simulates the essential primary system components of a typical West German 1300 PWR with regard to their thermohydraulic behaviour. The facility essentially consists of the pressure vessel with the heated bundle, the downcomer simulator, the primary loops with the components steam generator and pump simulator, the injection devices, the break geometry simulator, as well as the separators connected thereto, and the test containment to maintain a back-pressure at the location of break which is expected to be typical for emergency conditions. The number of heater rods and the cross-sections of the testing plant are on a reduced scale 1:134 in comparison with a typical German PWR. The elevations and locations are essentially full scale. Pressure vessel: The space between the pressure vessel and the inner core casing is sealed from the core region and the upper and lower plenum and connected with the upper plenum only by a pressure equalization line. The rod bundle surrounded by the inner core casing consists of 340 rods, 337 of which are indirect electrically heated. The test bundle cross-section as well as a heater element with the measuring elevations, the original-KWU-spacers and the axial power profile (7 power steps) are described. Downcomer: The downcomer is simulated by the downcomer nozzle region and the downcomer U-tube. The cold leg injection takes place both directly in the downcomer nozzle region and in the lines of t he intact single and double loop near to the downcomer nozzle region. A cylindrical insertion and repulsing metal sheets are installed in the downcomer nozzle region in order to avoid the emergency injection points into the broken loop. 2 - Description of test: Test K 9 out of a series PKL-IB was conducted on May 30, 1979 by Kraftwerk Union (KWU) at Erlangen (Germany). The objective of the integral cold leg injection test K 9 (double-ended 200%-break) was to investigate after a LOCA the refill and

  5. Validation of a methodology to develop a test facility in reduced scale related to boron dispersion in a pressurizer of an iPWR

    Energy Technology Data Exchange (ETDEWEB)

    Nascimento, Samira R.V.; Lira, Carlos A.B.O.; Lapa, Celso M.F.; Lima, Fernando R.A.; Bezerra, Jair L.; Silva, Mário A.B., E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Centro de Tecnologia e Geociências. Departamento de Energia Nuclear; Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Centro Regional de Ciências Nucleares do Nordeste (CRCN/CNEN-PE), Recife, PE (Brazil)

    2017-11-01

    The conception and the project of a 1:200 reduced scale test facility have been developed in earlier researches. Such a facility aims to investigate boron homogenization process inside the pressurizer of an iPWR (integral PWR) by considering water mixing from this component with that coming from the reactor core. For this kind of reactor, the pressurizer is located at the top of the pressure vessel demanding the need of identifying the proper mechanisms in order to warrant an adequate homogenization for the water mixture. Once the installation of the experimental setup was concluded, its behavior has been analyzed by considering the concentration of a tracer diluted in the circulation water, whose measurements were obtained at the pressurizer outlet orifices. Two experiments representing boration(boron concentration increase)/deboration(boron concentration decrease) scenarios have been accomplished. Sample acquisition was carried out for every ten minutes during a total time equal to 180 minutes. Results showed that the combination of Fractional Scaling Analysis with local Froude number consisted of an appropriate methodology to provide the reduced scale test facility parameters, inasmuch the measured concentrations from the experiments reproduced the theoretical behavior with sufficient accuracy. (author)

  6. Validation of a methodology to develop a test facility in reduced scale related to boron dispersion in a pressurizer of an iPWR

    International Nuclear Information System (INIS)

    Nascimento, Samira R.V.; Lira, Carlos A.B.O.; Lapa, Celso M.F.; Lima, Fernando R.A.; Bezerra, Jair L.; Silva, Mário A.B.

    2017-01-01

    The conception and the project of a 1:200 reduced scale test facility have been developed in earlier researches. Such a facility aims to investigate boron homogenization process inside the pressurizer of an iPWR (integral PWR) by considering water mixing from this component with that coming from the reactor core. For this kind of reactor, the pressurizer is located at the top of the pressure vessel demanding the need of identifying the proper mechanisms in order to warrant an adequate homogenization for the water mixture. Once the installation of the experimental setup was concluded, its behavior has been analyzed by considering the concentration of a tracer diluted in the circulation water, whose measurements were obtained at the pressurizer outlet orifices. Two experiments representing boration(boron concentration increase)/deboration(boron concentration decrease) scenarios have been accomplished. Sample acquisition was carried out for every ten minutes during a total time equal to 180 minutes. Results showed that the combination of Fractional Scaling Analysis with local Froude number consisted of an appropriate methodology to provide the reduced scale test facility parameters, inasmuch the measured concentrations from the experiments reproduced the theoretical behavior with sufficient accuracy. (author)

  7. State of the art report of exponential experiments with PWR spent nuclear fuel

    International Nuclear Information System (INIS)

    Ro, Seung Gy; Park, Sung Won; Park, Kwang Joon; Kim, Jong Hoon; Hong, Kwon Pyo; Shin, Hee Sung

    2000-09-01

    Exponential experiment method is discussed for verifying the computer code system of the nuclear criticality analysis which makes it possible to apply for the burnup credit in storage, transportation, and handling of spent nuclear fuel. In this report, it is described that the neutron flux density distribution in the exponential experiment system which consists of a PWR spent fuel in a water pool is measured by using 252 Cf neutron source and a mini-fission chamber, and therefrom the exponential decay coefficient is determined. Besides, described is a method for determining the absolute thermal neutron flux density by means of the Cd cut-off technique in association with a gold foil. Also a method is described for analyzing the energy distribution of γ-ray from the gold foil activation detector in detail

  8. Maturity of the PWR

    International Nuclear Information System (INIS)

    Bacher, P.; Rapin, M.; Aboudarham, L.; Bitsch, D.

    1983-03-01

    Figures illustrating the predominant position of the PWR system are presented. The question is whether on the basis of these figures the PWR can be considered to have reached maturity. The following analysis, based on the French program experience, is an attempt to pinpoint those areas in which industrial maturity of the PWR has been attained, and in which areas a certain evolution can still be expected to take place

  9. PWR-related integral safety experiments in the PKL 111 test facility SBLOCA under beyond-design-basis accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Weber, P.; Umminger, K.J.; Schoen, B. [Siemens AG Power Generation Group (KWU), Erlangen (France)

    1995-09-01

    The thermal hydraulic behavior of a PWR during beyond-design-basis accident scenarios is of vital interest for the verification and optimization of accident management procedures. Within the scope of the German reactor safety research program experiments were performed in the volumetrically scaled PKL 111 test facility by Siemens/KWU. This highly instrumented test rig simulates a KWU-design PWR (1300 MWe). In particular, the latest tests performed related to a SBLOCA with additional system failures, e.g. nitrogen entering the primary system. In the case of a SBLOCA, it is the goal of the operator to put the plant in a condition where the decay heat can be removed first using the low pressure emergency core cooling system and then the residual heat removal system. The experimental investigation presented assumed the following beyond-design-basis accident conditions: 0.5% break in a cold leg, 2 of 4 steam generators (SGs) isolated on the secondary side (feedwater- and steam line-valves closed), filled with steam on the primary side, cooldown of the primary system using the remaining two steam generators, high pressure injection system only in the two loops with intact steam generators, if possible no operator actions to reach the conditions for residual heat removal system activation. Furthermore, it was postulated that 2 of the 4 hot leg accumulators had a reduced initial water inventory (increased nitrogen inventory), allowing nitrogen to enter the primary system at a pressure of 15 bar and nearly preventing the heat transfer in the SGs ({open_quotes}passivating{close_quotes} U-tubes). Due to this the heat transfer regime in the intact steam generators changed remarkably. The primary system showed self-regulating system effects and heat transfer improved again (reflux-condenser mode in the U-tube inlet region).

  10. Utilization of ''CONTACT'' experiments to improve the fission gas release knowledge in PWR fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Charles, M; Abassin, J J; Bruet, M; Baron, D; Melin, P

    1983-03-01

    The CONTACT experiments, which were carried out by the French CEA, within the framework of a CEA-FRAMATOME collaboration agreement, bear on the behaviour of in-pile irradiated PWR fuel rods. We will focus here upon their results dealing with fission gas release. The experimental device is briefly described, then the following results are given: the kinetics of stable fission gas release for various linear ratings; the instantaneous fractional release rates of radioactive gases versus their decay constant in the range 1.5 10/sup -6/-3.6 10/sup -3/s/sup -1/, for various burnups, as also the influence of fuel temperature. Moreover, the influence of the nature and the pressure of the filling gas upon the release is presented for various linear ratings. The experimental results are discussed and analysed with the purpose to model various physical phenomena involved in the release (low-temperature mechanisms, diffusion).

  11. A survey of blockage measurement methods used in PWR multi-rod experiments

    Energy Technology Data Exchange (ETDEWEB)

    Hindle, E.D.; Jones, C.; Whitty, S. (AEA Reactor Services, Springfield (UK))

    1986-05-01

    The deformation characteristics of Zircaloy multi-rod arrays are being investigated in laboratory and in-reactor tests, and heat transfer experiments are being carried out on pre-deformed arrays. The primary objective is to demonstrate that cladding distension occurring under hypothetical loss-of-coolant accident (LOCA) conditions will not impede the PWR emergency coolant flow during the reflood stage to the extent that unacceptably high cladding temperatures are reached, i.e. that a coolable geometry is maintained. This Report critically reviews the current methods for measuring blockage in multi-rod arrays and discusses their application. A new definition which overcomes the deficiencies of the previous methods is proposed even though it still has drawbacks in the case of overall blockage measurement. A method for automatically measuring the individual rod strain, general cluster blockage sub-channel blockage and sub-channel perimeter changes is described and the results from a deformed array presented. (author).

  12. A survey of blockage measurement methods used in PWR multi-rod experiments

    International Nuclear Information System (INIS)

    Hindle, E.D.; Jones, C.; Whitty, S.

    1986-05-01

    The deformation characteristics of Zircaloy multi-rod arrays are being investigated in laboratory and in-reactor tests, and heat transfer experiments are being carried out on pre-deformed arrays. The primary objective is to demonstrate that cladding distension occurring under hypothetical loss-of-coolant accident (LOCA) conditions will not impede the PWR emergency coolant flow during the reflood stage to the extent that unacceptably high cladding temperatures are reached, i.e. that a coolable geometry is maintained. This Report critically reviews the current methods for measuring blockage in multi-rod arrays and discusses their application. A new definition which overcomes the deficiencies of the previous methods is proposed even though it still has drawbacks in the case of overall blockage measurement. A method for automatically measuring the individual rod strain, general cluster blockage sub-channel blockage and sub-channel perimeter changes is described and the results from a deformed array presented. (author)

  13. French analytic experiment on the high specific burnup of PWR fuels in normal conditions

    International Nuclear Information System (INIS)

    Bruet, M.; Atabek, R.; Houdaille, B.; Baron, D.

    1982-04-01

    Hydrostatic density determinations made on UO 2 pellets of different kinds irradiated in conditions representative of PWR conditions enable the internal swelling rate of the UO 2 to be ascertained. A mean value of 0.8% per 10 4 MWdt -1 (u) up to a specific burnup of 45000 MWdt -1 (u) may be deduced from this experimental basis. These results agree well with those obtained in the TANGO experiments in which UO 2 balls were irradiated in quasi isothermal conditions and without stress. Further, the open porosity of oxide closes progressively and the change in the total porosity is thus very limited (under 1% at 45000 MWdt -1 (u)). With respect to the swelling of the pellets the rise in the specific burnup would not appear therefore to be a problem. The behaviour of recrystallized zircaloy 4 claddings remains satisfactory with respect to creep and growth during irradiation [fr

  14. Utilization of ''CONTACT'' experiments to improve the fission gas release knowledge in PWR fuel rods

    International Nuclear Information System (INIS)

    Charles, M.; Abassin, J.J.; Bruet, M.

    1983-01-01

    The CONTACT experiments, which were carried out by the French CEA, within the framework of a CEA-FRAMATOME collaboration agreement, bear on the behaviour of in-pile irradiated PWR fuel rods. We will focus here upon their results dealing with fission gas release. The experimental device is briefly described, then the following results are given: the kinetics of stable fission gas release for various linear ratings; the instantaneous fractional release rates of radioactive gases versus their decay constant in the range 1.5 10 -6 -3.6 10 -3 s -1 , for various burnups, as also the influence of fuel temperature. Moreover, the influence of the nature and the pressure of the filling gas upon the release is presented for various linear ratings. The experimental results are discussed and analysed with the purpose to model various physical phenomena involved in the release (low-temperature mechanisms, diffusion)

  15. Numerical experiment designs. Study of vibratory behaviour of PWR'S control rod clusters

    International Nuclear Information System (INIS)

    Bosselut, D.; Soulier, B.; Regnier, G.

    1997-01-01

    The application of Experiment Design method to Finite Element Model (FEM) calculations is an original way of performing parametric studies. It has been used at EDF to simulate on a large parametric domain the vibrations of PWR's control rod cluster and to analyse the rod wear process. In the first part the FEM and the location of excitation sources are described. The calculated values are: rod displacement in the guiding cards, shock forces on the guiding cards and the wear power produced. In the second part, the computed Experiment Domain is described. This method approaches the response surface by a second degree polynomial. The retained model is composed for every parameters of all linear, quadratic and interaction terms (26 coefficients). In all, 34 polynomials have been built to approach the effective shock forces and the mean wear power at each of the 17 guiding points. In the third part the building of the computer Experiment Design is detailed: by Doehlert design adaptation to take into account a qualitative parameter, design optimization by adding four well chosen experiments and finally, design extension by passing from 4 to 6 parameters. In the last part, all the information deduced from application of this method are presented. The influence of parameters on calculated effective shock forces has been determined along rods and response surface have been easily approximated. The systematism and closeness of Experiment Design technique is underlined. Easy simulation of all the response domain by polynomial approach allows comparison with experimental results. (authors)

  16. Design of the control room of the N4-type PWR: main features and feedback operating experience

    International Nuclear Information System (INIS)

    Peyrouton, J.M.; Guillas, J.; Nougaret, Ch.

    2004-01-01

    This article presents the design, specificities and innovating features of the control room of the N4-type PWR. A brief description of control rooms of previous 900 MW and 1300 MW -type PWR allows us to assess the change. The design of the first control room dates back to 1972, at that time 2 considerations were taken into account: first the design has to be similar to that of control rooms for thermal plants because plant operators were satisfied with it and secondly the normal operating situation has to be privileged to the prejudice of accidental situations just as it was in a thermal plant. The turning point was the TMI accident that showed the weight of human factor in accidental situations in terms of pilot team, training, procedures and the ergonomics of the work station. The impact of TMI can be seen in the design of 1300 MW-type PWR. In the beginning of the eighties EDF decided to launch a study for a complete overhaul of the control room concept, the aim was to continue reducing the human factor risk and to provide a better quality of piloting the plant in any situation. The result is the control room of the N4-type PWR. Today the cumulated feedback experience of N4 control rooms represents more than 20 years over a wide range of situations from normal to incidental, a survey shows that the N4 design has fulfilled its aims. (A.C.)

  17. Pressure vessel fracture studies pertaining to a PWR LOCA-ECC thermal shock: experiments TSE-1 and TSE-2

    International Nuclear Information System (INIS)

    Cheverton, R.D.

    1976-09-01

    The LOCA-ECC Thermal Shock Program was established to investigate the potential for flaw propagation in pressurized-water reactor (PWR) vessels during injection of emergency core coolant following a loss-of-coolant accident. Studies thus far have included fracture mechanics analyses of typical PWRs, the design and construction of a thermal shock test facility, determination of material properties for test specimens, and two thermal shock experiments with 0.53-m-OD (21-in.) by 0.15-m-wall (6-in.) cylindrical test specimens. The PWR calculations indicated that under some circumstances crack propagation could be expected and that experiments should be conducted for cracks that would have the potential for propagation at least halfway through the wall

  18. Experiment and analyses on intentional secondary-side depressurization during PWR small break LOCA. Effects of depressurization rate and break area on core liquid level behavior

    International Nuclear Information System (INIS)

    Asaka, Hideaki; Ohtsu, Iwao; Anoda, Yoshinari; Kukita, Yutaka

    1997-01-01

    The effects of the secondary-side depressurization rate and break area on the core liquid level behavior during a PWR small-break LOCA were studied using experimental data from the Large Scale Test Facility (LSTF) and by using analysis results obtained with a JAERI modified version of RELAP5/MOD3 code. The LSTF is a 1/ 48 volumetrically scaled full-height integral model of a Westinghouse-type PWR. The code reproduced the thermal-hydraulic responses, observed in the experiment, for important parameters such as the primary and secondary side pressures and core liquid level behavior. The sensitivity of the core minimum liquid level to the depressurization rate and break area was studied by using the code assessed above. It was found that the core liquid level took a local minimum value for a given break area as a function of secondary side depressurization rate. Further efforts are, however, needed to quantitatively define the maximum core temperature as a function of break area and depressurization rate. (author)

  19. International experience with a multidisciplinary table top exercise for response to a PWR accident

    International Nuclear Information System (INIS)

    Lakey, J.R.A.

    1996-01-01

    Table Top Exercises are used for the training of emergency response personnel from a wide range of disciplines whose duties range from strategic to tactical, from managerial to operational. The exercise reported in this paper simulates the first two or three hours of an imaginary accident on a generic PWR site (named Seaside or Lakeside depending on its location). It is designed to exercise the early response of staff of the utility, government, local authority and the media and some players represent the public. The relatively few scenarios used for this exercise are based on actual events scaled to give off-site consequences which demand early assessment and therefore stress the communication procedures. The exercise is applicable in different cultures and has been used in over 20 short courses held in the USA, UK, Sweden, Prague, and Hong Kong. There are two styles of support for players: a linear program which ensures that all players follow the desired path through the event and an open program which is triggered by umpires (who play the reactor crew from a script) and by requests from other players. In both cases the exercise ends with a Press Conference. Players have an initial briefing and are assigned to roles; those who must speak at interviews and at the Press Conference arc given separate briefing by an expert in Public Affairs. The exercise runs with up to six groups and the communication rate reaches about 30 to 40 messages per hour for each group. The exercise can be applied to test management and communication systems and to study human response to emergencies because the merits of individual players are highlighted in the relatively stressful conditions of the initial stage of an accident. For some players the exercise is the first time that they have been required to carry out their task in front of other people

  20. Operating experience with an on-line vibration control system for PWR main coolant pumps

    International Nuclear Information System (INIS)

    Runkel, J.; Stegemann, D.; Vortriede, A.

    1996-01-01

    The main circulation pumps are key components of nuclear power plants with pressurized water reactors, because the availability of the main circulation pumps has a direct influence on the availability and electrical output of the entire plant. The on-line automatic vibration control system ASMAS was developed for early failure detection during the normal operation of the main circulation pumps in order to avoid unexpected outages and to establish the possibility of preventive maintenance of the pumps. This system is permanently and successfully operating in three German 1300 MW el NPP's with PWR and has been successfully tested in a 350 MW el NPP with a PWR. (orig.)

  1. Operating experience with an on-line vibration control system for PWR main coolant pumps

    International Nuclear Information System (INIS)

    Runkel, J.; Stegemann, D.; Vortriede, A.

    1998-01-01

    The main circulation pumps are key components of nuclear power plants with pressurized water reactors (PWRs), because the availability of the main circulation pumps has a direct influence on the availability and electrical output of the entire plant. The on-line automatic vibration control system ASMAS was developed for early failure detection during the normal operation of the main circulation pumps in order to avoid unexpected outages and to establish the possibility of preventive maintenance of the pumps. This system is permanently and successfully operating in three German 1300 MW e1 NPP's with PWR and has been successfully tested in a 350 MW e1 NPP with a PWR. (orig.)

  2. Analysis of a bending test on a full-scale PWR hot leg elbow containing a surface crack

    Energy Technology Data Exchange (ETDEWEB)

    Delliou, P. le [Electricite de France, EDF, 77 - Moret-sur-Loing (France). Dept. MTC; Julisch, P.; Hippelein, K. [Stuttgart Univ. (Germany). Staatliche Materialpruefungsanstalt; Bezdikian, G. [Electricite de France, EDF, 92 - Paris la Defense (France). Direction Production Transport

    1998-11-01

    EDF, in co-operation with Framatome, has conducted a large research programme on the mechanical behaviour of thermally aged cast duplex stainless steel elbows, which are part of the main primary circuit of French PWR. One important task of this programme consisted of testing a full-scale PWR hot leg elbow. The elbow contained a semi-elliptical circumferential notch machined on the outer surface of the intrados as well as casting defects located on the flanks. To simulate the end-of-life condition of the component regarding material toughness, it had undergone a 2400 hours ageing heat treatment at 400 C. The test preparation and execution, as well as the material characterization programme, were committed to MPA. The test was conducted under constant internal pressure and in-plane bending (opening mode) at 200 C. For safety reasons, it took place on an open air-site: the Meppen military test ground. At the maximum applied moment (6000 kN.m), the notch did not initiate. This paper presents the experimental results and the fracture mechanics analysis of the test, based on finite element calculations. (orig.)

  3. PWR boron dilution transients. Thermal-hydraulic analyses of PKL-E experiments

    International Nuclear Information System (INIS)

    Pietro Alessandro Di Maio; Antonino Tomasello; Giuseppe Vella

    2005-01-01

    Full text of publication follows: In the field of PWR reactor safety, one of the topics that is currently of major interest worldwide is that of inadvertent boron dilution events. The safety issue involved in such scenarios is that inadvertent transport into the reactor core of un-borated water - or water having only a low boron concentration - can lead to local recriticality and possibly to power excursions. Studies on various accidental sequences that could initiate boron dilution events revealed that some SBLOCAs, occurring in the primary system, lead to reflux condenser conditions and subsequent re-establishment of natural circulation are of particular significance. In this work, the first field of analysis is related to the investigation of the thermal - hydraulic conditions that could lead to boron dilution events, such as the stop of natural circulation within primary system and the subsequent start of reflux condenser functioning mode. The investigation of the primary thermal - hydraulic conditions has been performed using the experimental results obtained in the PKL test integral facility in which some SBLOCA sequences have been carried out. Particular useful were the PKL III E experiments data whose results have been numerically reproduced using the code Relap5/MOD3.3/Beta code, contributing to understand the complex thermalhydraulic phenomena related to a PWR boron dilution event. The second field of analysis is related to the effects that possible displacements of un-borated water slugs towards the Reactor Pressure Vessel (RPV) could have on the core reactivity. A numerical approach using the Relap5 reactor kinetics model has been adopted to integrate the experimental thermal - hydraulic data obtained in the PKL III E tests. A careful analysis has been performed in order to establish which core conditions at incident start could produce the largest reactivity increase as a consequence of restarting of natural circulation during the primary system

  4. PWR boron dilution transients. Thermal-hydraulic analyses of PKL-E experiments

    Energy Technology Data Exchange (ETDEWEB)

    Pietro Alessandro Di Maio; Antonino Tomasello; Giuseppe Vella [Dipartimento di Ingegneria Nucleare, Viale delle Scienze, 90128 Palermo (Italy)

    2005-07-01

    Full text of publication follows: In the field of PWR reactor safety, one of the topics that is currently of major interest worldwide is that of inadvertent boron dilution events. The safety issue involved in such scenarios is that inadvertent transport into the reactor core of un-borated water - or water having only a low boron concentration - can lead to local recriticality and possibly to power excursions. Studies on various accidental sequences that could initiate boron dilution events revealed that some SBLOCAs, occurring in the primary system, lead to reflux condenser conditions and subsequent re-establishment of natural circulation are of particular significance. In this work, the first field of analysis is related to the investigation of the thermal - hydraulic conditions that could lead to boron dilution events, such as the stop of natural circulation within primary system and the subsequent start of reflux condenser functioning mode. The investigation of the primary thermal - hydraulic conditions has been performed using the experimental results obtained in the PKL test integral facility in which some SBLOCA sequences have been carried out. Particular useful were the PKL III E experiments data whose results have been numerically reproduced using the code Relap5/MOD3.3/Beta code, contributing to understand the complex thermalhydraulic phenomena related to a PWR boron dilution event. The second field of analysis is related to the effects that possible displacements of un-borated water slugs towards the Reactor Pressure Vessel (RPV) could have on the core reactivity. A numerical approach using the Relap5 reactor kinetics model has been adopted to integrate the experimental thermal - hydraulic data obtained in the PKL III E tests. A careful analysis has been performed in order to establish which core conditions at incident start could produce the largest reactivity increase as a consequence of restarting of natural circulation during the primary system

  5. Application of the Particle Swarm Optimization (PSO) technique to the thermal-hydraulics project of a PWR reactor core in reduced scale

    International Nuclear Information System (INIS)

    Lima Junior, Carlos Alberto de Souza

    2008-09-01

    The reduced scale models design have been employed by engineers from several different industries fields such as offshore, spatial, oil extraction, nuclear industries and others. Reduced scale models are used in experiments because they are economically attractive than its own prototype (real scale) because in many cases they are cheaper than a real scale one and most of time they are also easier to build providing a way to lead the real scale design allowing indirect investigations and analysis to the real scale system (prototype). A reduced scale model (or experiment) must be able to represent all physical phenomena that occurs and further will do in the real scale one under operational conditions, e.g., in this case the reduced scale model is called similar. There are some different methods to design a reduced scale model and from those two are basic: the empiric method based on the expert's skill to determine which physical measures are relevant to the desired model; and the differential equation method that is based on a mathematical description of the prototype (real scale system) to model. Applying a mathematical technique to the differential equation that describes the prototype then highlighting the relevant physical measures so the reduced scale model design problem may be treated as an optimization problem. Many optimization techniques as Genetic Algorithm (GA), for example, have been developed to solve this class of problems and have also been applied to the reduced scale model design problem as well. In this work, Particle Swarm Optimization (PSO) technique is investigated as an alternative optimization tool for such problem. In this investigation a computational approach, based on particle swarm optimization technique (PSO), is used to perform a reduced scale two loop Pressurized Water Reactor (PWR) core, considering 100% of nominal power operation on a forced flow cooling circulation and non-accidental operating conditions. A performance comparison

  6. Field experience on Zn injection on PWR plants with a view to dose rate reduction

    International Nuclear Information System (INIS)

    Roumiguiere, F.

    2005-01-01

    Operating experience acquired at PWR plants shows that zinc injection in the primary coolant at low concentration (∼5 ppb) is a very effective tool to achieve a reduction of the dose rate build-up. The beneficial effect of zinc consists on improving the protective layer characteristics of the reactor coolant system surfaces, which results in a lower pickup of activated products (Co-60, Co-58), and consequently a reduction of the associated dose rates. Zinc injection was introduced at the Unit B of the Biblis Power Station in September 1996 and at the Obrigheim Nuclear Power Station in February 1998, as a measure for reduction of radiation fields. The effectiveness of the method and its compatibility with the overall plant was examined in a rather comprehensive surveillance program at these plants. The already published data show that zinc injection did not lead to any operating restrictions or other negative effects on plants systems and components. Zinc injection is still being implemented today at these plants. Zinc injection is considered today as a mature technique and is now being successfully applied at a number of PWRs in Germany, Brazil, USA and Japan, with the support of Framatome-ANP. Several PWRs in Europe and Asia are preparing for zinc chemistry in the near future. The method is inexpensive and easy to apply. Its implementation is highly advisable in terms of the cost/benefit criterion following the ALARA principle. This paper gives an overview of the experience gathered with the method. The main subject addressed by the paper is the evolution of dose rates at the primary system and work-related doses since introduction of the method. In German PWRs with Incoloy 800 steam generator tubing material (Ni-content ∼32%), the observed reductions correspond to a decrease in dose rates of around 10 to 15% per year following, as predicted, the half-life time of 60 Co. Overall reductions in high radiation areas are now in the range of 50% after 5 years of

  7. Study on scaling law of PWR natural circulation with motion condition

    International Nuclear Information System (INIS)

    Lu Donghua; Xiao Zejun; Chen Bingde

    2009-01-01

    For some nuclear reactors installed on automobiles, boats or deep sea vehicles, it is an important way to investigate their system safety by performing natural circulation experiments under motion condition. This paper studied the natural circulation on moving plants based on work of static natural circulation scaling method. With rigid motion theory, acceleration at each point was obtained on primary system and introduced to momentum equation. Thus a set of motion similar criteria were obtained. Furthermore, equal and unequal height simulation were analyzed. As to the unequal one, non isochronous simulation was needed for displacement and angular acceleration. (authors)

  8. DOMPAC dosimetry experiment. Neutronic simulation of the thickness of a PWR pressure vessel. Irradiation damages

    International Nuclear Information System (INIS)

    Alberman, A.; Faure, M.; Thierry, M.; Hoclet, O.; Le Dieu de Ville, A.; Nimal, J.C.; Soulat, P.

    1979-01-01

    For suitable extrapolation of irradiated PWR ferritic steel results, proper irradiation of the pressure vessel has been 'simulated' in test reactor. For this purpose, a huge steel block (20 cm in depth) was loaded with Saclay's graphite (GAMIN) and tungsten damage detectors. Core-block water gap was optimized through spectrum indexes method, by ANISN and SABINE codes so that spectrum in 1/4 thickness matches with ANISN computations for PWR Fessenheim 1. A good experimental agreement is found with calculated dpa damage gradient. 3D Monte Carlo computation (TRIPOLI), was performed on the DOMPAC device, and spectrum indexes evolution was found consistent with experimental results. Surveillance rigs behind a 'thermal shield' were also simulated, including damage and activation monitors. Dosimetry results give an order of magnitude of accuracies involved in projecting steel sample embrittlement to the pressure vessel [fr

  9. Twenty-five years of transient counting experience in French PWR units

    Energy Technology Data Exchange (ETDEWEB)

    Barthelet, B. [Electricite de France (EDF DPN), 93 - Saint-Denis (France); Savoldelli, D.; Fritz, R. [Electricite de France (EDF DPN), 93 - Noisy le Grand (France)

    2001-07-01

    For nearly twenty five years, EDF has been checking that the actual operating transients are neither more severe nor more numerous than the design basis transients. This activity of transient cycle counting and bookkeeping has enabled EDF to own a database of more than 800 reactor.years for the PWR units. The current method of transient cycle counting is presented. In the paper, we will point out the main results of transient cycle counting and lessons learned. In general, the frequencies of transients are lower than the design frequencies. In few cases, they are higher, such as the transient frequencies of the RCS lines connected to auxiliary systems often due to operating procedures or particular periodic testing. Few periodic tests were not taken into account in the design basis transient file ; they have been detected thanks to the transient cycle counting. In the last 1980's, we achieved the first updating of the design basis transient file for the PWR 900 MWe series. In the early 1990's, we updated the design basis transient file of the PWR 1300 MWe series. In fact, since design and start-up, the operating conditions have been modified (fuel cycle with stretch-out, modification of the hot leg and cold leg temperatures for the PWR 1300 MWe,...). This was the cause of many unclassified transients. In the new design basis transient file, we have created new transients and increased the frequencies of some of them. This has enabled to consider the updated design basis transient file more representative of actual operating transients. For some years, we have increasingly associated the operators with the transient cycle counting concern. We noticed progress (decreased frequencies of most transients). (authors)

  10. Twenty-five years of transient counting experience in French PWR units

    International Nuclear Information System (INIS)

    Barthelet, B.; Savoldelli, D.; Fritz, R.

    2001-01-01

    For nearly twenty five years, EDF has been checking that the actual operating transients are neither more severe nor more numerous than the design basis transients. This activity of transient cycle counting and bookkeeping has enabled EDF to own a database of more than 800 reactor.years for the PWR units. The current method of transient cycle counting is presented. In the paper, we will point out the main results of transient cycle counting and lessons learned. In general, the frequencies of transients are lower than the design frequencies. In few cases, they are higher, such as the transient frequencies of the RCS lines connected to auxiliary systems often due to operating procedures or particular periodic testing. Few periodic tests were not taken into account in the design basis transient file ; they have been detected thanks to the transient cycle counting. In the last 1980's, we achieved the first updating of the design basis transient file for the PWR 900 MWe series. In the early 1990's, we updated the design basis transient file of the PWR 1300 MWe series. In fact, since design and start-up, the operating conditions have been modified (fuel cycle with stretch-out, modification of the hot leg and cold leg temperatures for the PWR 1300 MWe,...). This was the cause of many unclassified transients. In the new design basis transient file, we have created new transients and increased the frequencies of some of them. This has enabled to consider the updated design basis transient file more representative of actual operating transients. For some years, we have increasingly associated the operators with the transient cycle counting concern. We noticed progress (decreased frequencies of most transients). (authors)

  11. Dosimetry and fluence calculations on french PWR vessels comparisons between experiments and calculations

    International Nuclear Information System (INIS)

    Nimal, J.C.; Bourdet, L.; Guilleret, J.C.; Hedin, F.

    1988-01-01

    Fluence and damage calculations on PWR pressure vessels and irradiation test specimens are presented for two types of reactor: the franco-belgian (reactor CHOOZ) and the french reactors (CPY program). Comparisons with measurements are given for activation foils and fission detectors; most of them are about irradiation test specimen locations; comparisons are made for the Chooz plant on vessel stainless steel samplings and in the reactor pit

  12. Comparison of thermal behavior of different PWR fuel rod simulators for LOCA experiments

    International Nuclear Information System (INIS)

    Casal, V.; Malang, S.; Rust, K.

    1982-10-01

    For experimental investigations of a loss-of-coolant accident (LOCA) of a PWR electrical heater rods are applied as thermal fuel rod simulators. To substitute heater rods from the SEMISCALE program by INTERATOM-KfK heater rods in a current experimental program at the Instituut for Energiteknikk-(OECD-Halden), the thermodynamic behavior of different heater rods during a LOCA were compared. The results show, that SEMISCALE-heater rods can be replaced by those fabricated by INTERATOM. (orig.) [de

  13. Experience and optimisation of ethanolamine treatment for a PWR secondary system

    International Nuclear Information System (INIS)

    Eeden, Nestor-van; Matthee, Felix; Montshiwagae, Maleke

    2012-09-01

    The secondary systems of both pressurised water reactor (PWR) units at Koeberg Nuclear Power Station (KNPS) on the West Coast of South Africa have been on ethanolamine (ETA) treatment since 1999 / 2000. Prior to conversion to ETA Chemistry the secondary regime was ammonia / hydrazine all volatile treatment (AVT) which aimed to achieve a pH 25C of 9.65 in the feedwater. ETA was selected as the preferred corrosion mitigation agent as opposed to the alternative secondary treatment technology available at the time to combat flow accelerated corrosion (FAC). The main advantage of ETA over ammonia is that ETA has a lower relative volatility and thus it favours the water phase while ammonia favours the steam phase. Plant chemistry, design and financial aspects were considered in the decision making for KNPS. Implementation of ETA at KNPS initially targeted 2 mg ETA/kg in the feedwater and 4 mg ETA/kg in the heater drains tank. It was found that under these conditions it was necessary to simultaneously inject ammonia to reach the pH specification in the feedwater. While using only pH as a limiting parameter, with ETA on manual control and ammonia on automated control, it was possible for the ETA concentration to drop below the target levels. To prevent this from occurring, during 2002 higher target concentrations for ETA were established of between 3.0 and 3.5 mg ETA/kg in the feedwater and with an upper limit value of 4 mg ETA/kg. In 2003, this upper limiting value and the target concentrations were discarded when it was decided to attain the pH specification without injecting ammonia. When dosing only ETA and hydrazine (in the absence of ammonia dosing) the pH specification was only achievable at greater than 4 mg ETA/kg. This specification was again revised in 2007 and changed to 4 mg ETA/kg as a minimum limit; with no upper limit value. Dosing was still manually controlled but now using feedwater conductivity measurement based on a correlation established with ETA

  14. Interpretation of out of line control rod experiments for 1300 MWE PWR

    Energy Technology Data Exchange (ETDEWEB)

    Leroy, J.L.; Garcia-Fernandez, L.

    1988-01-01

    The present note summarizes the studies we performed recently in order to search a 2D reconstruction procedure for the 1300 MWE PWR power shape, starting from data coming out from thermocouples placed on several fuel assemblies. In classical PWR design, only a few assemblies are equipped with measurement devices, so that it is necessary to interpolate among measure points in order to obtain a complete coverage of the core. A mathematical approach based on the splitting of the power into a reference steady state nominal shape and some ''influence'' and harmonic functions was chosen. The reference steady state power shape, which corresponds to the full power operating mode, is obtained via direct mobile chamber measurements. The perturbations due to the control rod movements are accounted for by specific ''influence'' functions: moreover, harmonics are used to reconstruct the minor effects due to xenon tilts, rod out of line positions and all actual mechanical and thermohydraulic inhomogeneities. The weighting coefficients of the functions are evaluated by a least square method, starting from the distribution of the deviations among the measurements and the reference values.

  15. International standard problem ISP36. Cora-W2 experiment on severe fuel damage for a Russian type PWR

    International Nuclear Information System (INIS)

    1996-01-01

    An OECD/NEA-CSNI International Standard Problem (ISP) has been performed on the experimental comparison basis of the severe fuel damage experiment CORA-W2. The out-of-pile experiment CORA-W2 was executed in February 1993 at he Forschungszentrum Karlsruhe. The objective of this experiment was the investigation of the behavior of a Russian type PWR fuel element (VVER-1000) during early core degradation. The main difference between a Western type and a Russian type PWR bundle is the B 4 C absorber rod instead of AgInCd. Measured quantities ar boundary conditions, bundle temperature, hydrogen generation and the final bundle configurations after cooldown. The ISP was conducted as a blind exercise. Boundary conditions were estimated using ATHLET-CD. Six different severe accident codes were used. The comparisons between experimental and analytical results were grouped by codes and examined separately. The thermal behavior up to significant oxidation has been predicted quite well. Larger deviations have been observed for the oxidation-induced temperature escalation, both time of onset and maximum temperature as well. The bundle behavior is greatly influenced by chemical interactions involving B 4 C absorber rod material, which failed relatively early at low temperature due to eutectic interaction between B 4 C and SS cladding as well as the SS guide tube. Regarding the complex material interaction larger differences can be recognized between calculated and measured results because of inappropriate models for material relocation and solidification processes and the lack of models describing the interactions of absorber rod materials with the fuel rods. For the total amount of H 2 generated, acceptable agreement could be achieved, if the total of oxidized zirconium was calculated correctly. The oxidation of stainless steel components and B 4 C were not treated. In general the confidence in code predictions decreases with processing core damage. (N.T.)

  16. International standard problem ISP36. Cora-W2 experiment on severe fuel damage for a Russian type PWR

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-31

    An OECD/NEA-CSNI International Standard Problem (ISP) has been performed on the experimental comparison basis of the severe fuel damage experiment CORA-W2. The out-of-pile experiment CORA-W2 was executed in February 1993 at he Forschungszentrum Karlsruhe. The objective of this experiment was the investigation of the behavior of a Russian type PWR fuel element (VVER-1000) during early core degradation. The main difference between a Western type and a Russian type PWR bundle is the B{sub 4}C absorber rod instead of AgInCd. Measured quantities ar boundary conditions, bundle temperature, hydrogen generation and the final bundle configurations after cooldown. The ISP was conducted as a blind exercise. Boundary conditions were estimated using ATHLET-CD. Six different severe accident codes were used. The comparisons between experimental and analytical results were grouped by codes and examined separately. The thermal behavior up to significant oxidation has been predicted quite well. Larger deviations have been observed for the oxidation-induced temperature escalation, both time of onset and maximum temperature as well. The bundle behavior is greatly influenced by chemical interactions involving B{sub 4}C absorber rod material, which failed relatively early at low temperature due to eutectic interaction between B{sub 4}C and SS cladding as well as the SS guide tube. Regarding the complex material interaction larger differences can be recognized between calculated and measured results because of inappropriate models for material relocation and solidification processes and the lack of models describing the interactions of absorber rod materials with the fuel rods. For the total amount of H{sub 2} generated, acceptable agreement could be achieved, if the total of oxidized zirconium was calculated correctly. The oxidation of stainless steel components and B{sub 4}C were not treated. In general the confidence in code predictions decreases with processing core damage. 36 refs.

  17. The PWR integrated leak rate test, a review of experiences and results

    International Nuclear Information System (INIS)

    Keogh, P.

    1985-01-01

    The paper reviews the Integrated Leak Rate Test (ILRT) as it has been carried out in the USA and as reported in papers in European countries. The test procedures are critically appraised and recommendations are given for modifications to them. The values used in a PWR are identified as a main source of leaks and possibilities for improvement are discussed. The use of a part pressure test and its limitations are considered. A part pressure test cannot give the same assurance as a full pressure test but may be useful for the identification of gross leaks. Secondary effects such as weather and the use of Van de Waals equations are considered and are found to be not important for concrete containments. (orig.)

  18. Safety considerations of PWR's

    International Nuclear Information System (INIS)

    Arnold, W.H. Jr.

    1977-01-01

    The safety of the central station pressurized water reactor is well established and substantiated by its excellent operating record. Operating data from 55 reactors of this type have established a record of safe operating history unparalleled by any modern large scale industry. The 186 plants under construction require a continuing commitment to maintain this outstanding record. The safety of the PWR has been further verified by the recently completed Reactor Safety Study (''Rasmussen'' Report). Not only has this study confirmed the exceptionally low risk associated with PWR operation, it has also introduced a valuable new tool in the decision making process. PWR designs, utilizing the philosophy of defense in depth, provide the bases for evaluating margins of safety. The design of the reactor coolant system, the containment system, emergency core cooling system and other related systems and components provide substantial margins of safety under both normal and postulated accident conditions even considering simultaneous effects of earthquakes and other environmental phenomena. Margins of safety in the assessment of various postulated accident conditions, with emphasis on the postulated loss of reactor coolant accident (LOCA), have been evaluated in depth as exemplified by the comprehensive ECCS rulemaking hearings followed by imposition of very conservative Nuclear Regulatory Commission requirements. When evaluated on an engineering best estimate approach, the significant margins to safety for a LOCA become more apparent. Extensive test programs have also substantiated margins to safety limits. These programs have included both separate effects and systems tests. Component testing has also been performed to substantiate performance levels under adverse combinations of environmental stress. The importance of utilizing past experience and of optimizing the deployment of incremental resources is self evident. Recent safety concerns have included specific areas such

  19. Determination of vibration amplitudes and neutron-mechanical scale factors in the PWR nuclear power plant

    International Nuclear Information System (INIS)

    Kostic, Lj.; Heidemann, P.; Runkel, J.

    1997-01-01

    Displacements of vibrating reactor components which can not be measured by other means during normal reactor operation can be determined through the scale factors from the neutron spectra of signals measured by the standard in-core neutron instrumentation. Neutron-mechanical scale factors are determined for the vibrations of fuel assemblies and reactor pressure vessel/core barrel system using the signals of in-core neutron detectors and accelerometers. (author)

  20. Laser anemometry measurements of natural circulation flow in a scale model PWR system

    International Nuclear Information System (INIS)

    Kadambi, J.R.; Schneider, S.J.

    1990-01-01

    This paper reports on experimental studies conducted to investigate the natural circulation of a single-phase fluid in a scale model pressurized water reactor system during a postulated degraded core accident. A half-section of a 1/7 scale model with a plexiglass adiabatic window was used. Water and sulfurhexafluoride (SF 6 ) were used as the fluid. Laser-Doppler anemometry (LDA) was used in marking the velocity measurements along the center plane of the model at five elevations

  1. Corrosion in PWR steam generator tubes made of alloy 600TT: overview of operating experience, NDE and safety issues

    International Nuclear Information System (INIS)

    Curieres, I. de; Sollier, T.; Delaval, C.

    2015-01-01

    About 60 PWR plants worldwide are operating with steam generator tubes made of alloy 600TT, among which 27 are located in France. This alloy is susceptible to corrosion, both on the primary and secondary side in every fleet, though with different kinetics or extent. It is noteworthy that many of the primary side corrosion issues can be clearly explained by design or operating conditions. However, studies show that all the secondary side issues are much hardly explained by simple considerations. This paper will give an overview of the international operating experience of this alloy and indicate the associated controllability and safety-related issues. An emphasis will be put on the manufacturing, chemistry and specificities of the different fleets. The French situation will be reviewed in this frame. (authors)

  2. Concept of scaled test facility for simulating the PWR thermalhydraulic behaviour

    International Nuclear Information System (INIS)

    Silva Filho, E.

    1990-01-01

    This work deals with the design of a scaled test facility of a typical pressurized water reactor plant, to simulation of small break Loss-of-Coolant Accident. The computer code RELAP 5/ MOD1 has been utilized to simulate the accident and to compare the test facility behaviour with the reactor plant one. The results demonstrate similar thermal-hydraulic behaviours of the two sistema. (author)

  3. PWR internals segmentation and packaging experience in the U.S

    International Nuclear Information System (INIS)

    Kreitman, P.J.

    2008-01-01

    Seven commercial nuclear power plants of the Pressurized Water Reactor (PWR) design have been permanently shut down in the US to date. Six of these plants have been decommissioned using dismantling methods. The remaining plant, Indian Point 1, located in Buchanan, NY, has been placed in Safe Storage. Of the six dismantled plants, five underwent extensive segmentation, separation, and packaging of their internal components to allow removal and disposal of the reactor vessel assembly. Only the Trojan Plant in Portland, Oregon, was able to ship the reactor vessel assembly with its internals intact to the final disposal site near Richland, Washington. An important part of the planning for the upcoming decommissioning of similar plants in Europe such as Chooz A in France and Zorita in Spain will be to study the lessons learned from the US efforts and apply the best practices to future projects. This paper will chronicle the evolution of the reactor internals segmentation and packaging process to date, including the planning, methodology, equipment, waste management, and packaging strategy. (author)

  4. Experience in the use of low concentration gadolinia as a PWR fuel burnable absorber

    International Nuclear Information System (INIS)

    Mildrum, C.M.; Segovia, M.A.

    2001-01-01

    A description is provided of the low concentration gad design being used in the Spanish 3-loop 17 x 17 fueled PWR's. This design uses a relatively small number of high concentration gadolinia fuel rods (6 and 8 w/o Gd2O3) with a large number of low concentration gad rods (2 w/o Gd2O3). The 2 w/o gad rods substitute, in part, the high concentration gad rods, thereby helping reduce the end of cycle reactivity penalty from the residual absorption in the gadolinium. The low concentration gad design is advantageous for long cycles (more than 18 months) and plant up-rating scenarios in that the soluble boron concentration increases that would otherwise result for these situations are avoided. These boron concentration increases could have potentially adverse effects on the plant, since the moderator temperature coefficient (MTC) is made less negative, the effectiveness of the boron shutdown safety systems is reduced, and the safety margins are eroded for some accidents, such as for boron dilution events. This paper also reviews the APA nuclear design code system performance for the low concentration gad design. (author)

  5. Operating experience with steam generator water chemistry in Japanese PWR plants

    International Nuclear Information System (INIS)

    Onimura, K.; Hattori, T.

    1991-01-01

    Since the first PWR plant in Japan started its commercial operation in 1970, seventeen plants are operating as of the end of 1990. First three units initially applied phosphate treatment as secondary water chemistry control and then changed to all volatile treatment (AVT) due to phosphate induced wastage of steam generator tubing. The other fourteen units operate exclusively under AVT. In Japan, several corrosion phenomena of steam generator tubing, resulted from secondary water chemistry, have been experienced, but occurrence of those phenomena has decreased by means of improvement on impurity management, boric acid treatment and high hydrazine operation. Recently secondary water chemistry in Japanese plants are well maintained in every stage of operation. This paper introduces brief summary of the present status of steam generators and secondary water chemistry in Japan and ongoing activities of investigation for future improvement of reliability of steam generator. History and present status of secondary water chemistry in Japanese PWRs were introduced. In order to get improved water chemistry, the integrity of secondary system equipments is essential and the improvement in water chemistry has been achieved with the improvement in equipments and their usage. As a result of those efforts, present status of secondary water is excellent. However, further development for crevice chemistry monitoring technique and an advanced water chemistry data management system is desired for the purpose of future improvement of reliability of steam generator

  6. Performance of core exit thermocouple for PWR accident management action in vessel top break LOCA simulation experiment at OECD/NEA ROSA project

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Nakamura, Hideo

    2009-01-01

    Presented are experiment results of the Large Scale Test Facility (LSTF) conducted at the Japan Atomic Energy Agency (JAEA) with a focus on core exit thermocouple (CET) performance to detect core overheat during a vessel top break loss-of-coolant accident (LOCA) simulation experiment. The CET temperatures are used to start accident management (AM) action to quickly depressurize steam generator (SG) secondary side in case of core temperature excursion. Test 6-1 is first test of the OECD/NEA ROSA Project started in 2005, simulating withdraw of a control rod drive mechanism penetration nozzle at the vessel top head. The break size is equivalent to 1.9% cold leg break. The AM action was initiated when CET temperature rose up to 623K. There was no reflux water fallback onto the CETs during the core heat-up period. The core overheat, however, was detected with a time delay of about 230s. In addition, a large temperature discrepancy was observed between the CETs and the hottest core region. This paper clarifies the reason of time delay and temperature discrepancy between the CETs and heated core during boil-off including three-dimensional steam flows in the core and core exit. The paper discusses applicability of the LSTF CET performance to pressurized water reactor (PWR) conditions and a possibility of alternative indicators for earlier AM action than in Test 6-1 is studied by using symptom-based plant parameters such as a reactor vessel water level detection. (author)

  7. Boron dilution transients during natural circulation flow in PWR-Experiments and CFD simulations

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, Thomas [Forschungszentrum Dresden-Rossendorf (FZD)-Institute of Safety Research, P.O. Box 510119, D-01314 Dresden (Germany)], E-mail: T.Hoehne@fzd.de; Kliem, Soeren; Rohde, Ulrich; Weiss, Frank-Peter [Forschungszentrum Dresden-Rossendorf (FZD)-Institute of Safety Research, P.O. Box 510119, D-01314 Dresden (Germany)

    2008-08-15

    Coolant mixing in the cold leg, downcomer and the lower plenum of pressurized water reactors is an important phenomenon mitigating the reactivity insertion into the core. Therefore, mixing of the de-borated slugs with the ambient coolant in the reactor pressure vessel was investigated at the four loops 1:5 scaled Rossendorf coolant mixing model (ROCOM) mixing test facility. In particular thermal hydraulics analyses have shown, that weakly borated condensate can accumulate in the pump loop seal of those loops, which do not receive a safety injection. After refilling of the primary circuit, natural circulation in the stagnant loops can re-establish simultaneously and the de-borated slugs are shifted towards the reactor pressure vessel (RPV). In the ROCOM experiments, the length of the flow ramp and the initial density difference between the slugs and the ambient coolant was varied. From the test matrix experiments with 0 resp. 2% density difference between the de-borated slugs and the ambient coolant were used to validate the CFD software ANSYS CFX. To model the effects of turbulence on the mean flow a higher order Reynolds stress turbulence model was employed and a mesh consisting of 6.4 million hybrid elements was utilized. Only the experiments and CFD calculations with modeled density differences show stratification in the downcomer. Depending on the degree of density differences the less dense slugs flow around the core barrel at the top of the downcomer. At the opposite side, the lower borated coolant is entrained by the colder safety injection water and transported to the core. The validation proves that ANSYS CFX is able to simulate appropriately the flow field and mixing effects of coolant with different densities.

  8. Analyses of PWR spent fuel composition using SCALE and SWAT code systems to find correction factors for criticality safety applications adopting burnup credit

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Hee Sung; Suyama, Kenya; Mochizuki, Hiroki; Okuno, Hiroshi; Nomura, Yasushi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-01-01

    The isotopic composition calculations were performed for 26 spent fuel samples from the Obrigheim PWR reactor and 55 spent fuel samples from 7 PWR reactors using the SAS2H module of the SCALE4.4 code system with 27, 44 and 238 group cross-section libraries and the SWAT code system with the 107 group cross-section library. For the analyses of samples from the Obrigheim PWR reactor, geometrical models were constructed for each of SCALE4.4/SAS2H and SWAT. For the analyses of samples from 7 PWR reactors, the geometrical model already adopted in the SCALE/SAS2H was directly converted to the model of SWAT. The four kinds of calculation results were compared with the measured data. For convenience, the ratio of the measured to calculated values was used as a parameter. When the ratio is less than unity, the calculation overestimates the measurement, and the ratio becomes closer to unity, they have a better agreement. For many important nuclides for burnup credit criticality safety evaluation, the four methods applied in this study showed good coincidence with measurements in general. More precise observations showed, however: (1) Less unity ratios were found for Pu-239 and -241 for selected 16 samples out of the 26 samples from the Obrigheim reactor (10 samples were deselected because their burnups were measured with Cs-137 non-destructive method, less reliable than Nd-148 method the rest 16 samples were measured with); (2) Larger than unity ratios were found for Am-241 and Cm-242 for both the 16 and 55 samples; (3) Larger than unity ratios were found for Sm-149 for the 55 samples; (4) SWAT was generally accompanied by larger ratios than those of SAS2H with some exceptions. Based on the measured-to-calculated ratios for 71 samples of a combined set in which 16 selected samples and 55 samples were included, the correction factors that should be multiplied to the calculated isotopic compositions were generated for a conservative estimate of the neutron multiplication factor

  9. Current experience and a new modeling on water hammer due to steam condensation in PWR secondary system

    International Nuclear Information System (INIS)

    Kawanishi, K.; Kasahara, J.; Ueno, T.; Suzuta, T.

    1998-01-01

    There have been possibilities to occur water hammer in pipelines of turbine system for nuclear or fossil fuel power plants. According to the NUREG report, approximately 150 events have been reported since 1969, we also have an experience recently. Water hammer occurs due to sudden steam condensation with pressure pulse. This kind of pressure pulses has been made by alternative producing and condensing of steam slug in the pipe and its frequency relates subcooling and pipe structures. This paper presents our current experience on water hammer with some experimental studies. The present experiment has been performed to obtain the data base for evaluating the pressure pulses. The test pipe was horizontal tubes with dead end connected to vertical tube which simulating drain line in PWR secondary system. The main results are shown as follows; Magnitude of pressure pulse depends drain velocity and initial subcooling. Pipe structure effects on the frequency and continual time of water hammer phenomenon. A new modeling for quantitative explanation of the phenomena is also presented

  10. A new approach to designing reduced scale thermal-hydraulic experiments

    International Nuclear Information System (INIS)

    Lapa, Celso M.F.; Sampaio, Paulo A.B. de; Pereira, Claudio M.N.A.

    2004-01-01

    Reduced scale experiments are often employed in engineering because they are much cheaper than real scale testing. Unfortunately, though, it is difficult to design a thermal-hydraulic circuit or equipment in reduced scale capable of reproducing, both accurately and simultaneously, all the physical phenomena that occur in real scale and operating conditions. This paper presents a methodology to designing thermal-hydraulic experiments in reduced scale based on setting up a constrained optimization problem that is solved using genetic algorithms (GAs). In order to demonstrate the application of the methodology proposed, we performed some investigations in the design of a heater aimed to simulate the transport of heat and momentum in the core of a pressurized water reactor (PWR) at 100% of nominal power and non-accident operating conditions. The results obtained show that the proposed methodology is a promising approach for designing reduced scale experiments

  11. Experience in the use of low concentration gadolinia as a PWR fuel burnable absorber

    International Nuclear Information System (INIS)

    Mildrum, C.M.; Segovia, M.A.

    2001-01-01

    A description is provided of the low concentration gad design being used in the Spanish 3-loop 17 x 17 fueled PWR's. This design uses a relatively small number of high concentration gadolinia fuel rods (6 and 8 w/o Gd 2 O 3 ) with a large number of low concentration gad rods (2 w/o Gd 2 O 3 ). The 2 w/o gad rods substitute, in part, the high concentration gad rods, thereby helping reduce the end of cycle reactivity penalty from the residual absorption in the gadolinium. The low concentration gad design is advantageous for long cycles (18+ months) and plant up-rating scenarios in that the soluble boron concentration increases that would otherwise result for these situations are avoided. These boron concentration increases could have potentially adverse effects on the plant, since the moderator temperature coefficient (MTC) is made less negative, the effectiveness of the boron shutdown safety systems is reduced, and the safety margins are eroded for some accidents, such as for boron dilution events. These increases in the boron concentration would also require the plant to operate at higher lithium (Li) concentrations in the coolant in order to maintain the pH level at the desired value. Operation at the higher Li concentrations is undesirable because of the concerns over the potential impact on the fuel assembly material performance (e.g., crud and corrosion). This paper also reviews the APA (Alpha/Phoenix-P/ANC) nuclear design code system performance for the low concentration gad design. The design system performance for the reload cores that have or are employing this design has been completely satisfactory. The performance and accuracy of the nuclear design methodology is found to be as good for this design as for the reload cores that use exclusively high gad concentrations, or those that use WABA's - the discrete burnable absorber (BA) used prior to its substitution for gadolinium. (authors)

  12. PWR secondary water chemistry guidelines

    International Nuclear Information System (INIS)

    Bell, M.J.; Blomgren, J.C.; Fackelmann, J.M.

    1982-10-01

    Steam generators in pressurized water reactor (PWR) nuclear power plants have experienced tubing degradation by a variety of corrosion-related mechanisms which depend directly on secondary water chemistry. As a result of this experience, the Steam Generator Owners Group and EPRI have sponsored a major program to provide solutions to PWR steam generator problems. This report, PWR Secondary Water Chemistry Guidelines, in addition to presenting justification for water chemistry control parameters, discusses available analytical methods, data management and surveillance, and the management philosophy required to successfully implement the guidelines

  13. Validation of PWR core seismic models with shaking table tests on interacting scale 1 fuel assemblies

    International Nuclear Information System (INIS)

    Viallet, E.; Bolsee, G.; Ladouceur, B.; Goubin, T.; Rigaudeau, J.

    2003-01-01

    The fuel assembly mechanical strength must be justified with respect to the lateral loads under accident conditions, in particular seismic loads. This justification is performed by means of time-history analyses with dynamic models of an assembly row in the core, allowing for assembly deformations, impacts at grid locations and reactor coolant effects. Due to necessary simplifications, the models include 'equivalent' parameters adjusted with respect to dynamic characterisation tests of the fuel assemblies. Complementing such tests on isolated assemblies by an overall model validation with shaking table tests on interacting assemblies is obviously desirable. Seismic tests have been performed by French CEA (Commissariat a l'Energie Atomique) on a row of six full scale fuel assemblies, including two types of 17 x 17 12ft design. The row models are built according to the usual procedure, with preliminary characterisation tests performed on a single assembly. The test-calculation comparisons are made for two test configurations : in air and in water. The relatively large number of accelerograms (15, used for each configuration) is also favourable to significant comparisons. The results are presented for the impact forces at row ends, displacements at mid assembly, and also 'statistical' parameters. Despite a non-negligible scattering in the results obtained with different accelerograms, the calculations prove realistic, and the modelling process is validated with a good confidence level. This satisfactory validation allows to evaluate precisely the margins in the seismic design methodology of the fuel assemblies, and thus to confirm the safety of the plants in case of seismic event. (author)

  14. Comparison of problems and experience of core operation with distorted fuel element assemblies in VVER-1000 and PWR reactors

    International Nuclear Information System (INIS)

    Afanas'ev, A.

    1999-01-01

    The main reactors leading to distortion of fuel element assemblies during reactor operation were studied. A series of actions which compensate this effect was proposed. Criteria of operation limitation in VVER-1000 and PWR reactors are described

  15. Numerical simulations of counter-current two-phase flow experiments in a PWR hot leg model using an interfacial area density model

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, Thomas, E-mail: t.hoehne@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf (HZDR), Institute of Safety Research, P.O. Box 510 119, D-01314 Dresden (Germany); Deendarlianto,; Lucas, Dirk [Helmholtz-Zentrum Dresden-Rossendorf (HZDR), Institute of Safety Research, P.O. Box 510 119, D-01314 Dresden (Germany)

    2011-10-15

    In order to improve the understanding of counter-current two-phase flows and to validate new physical models, CFD simulations of 1/3rd scale model of the hot leg of a German Konvoi PWR with rectangular cross section was performed. Selected counter-current flow limitation (CCFL) experiments at the Helmholtz-Zentrum Dresden-Rossendorf (HZDR) were calculated with ANSYS CFX 12.1 using the multi-fluid Euler-Euler modeling approach. The transient calculations were carried out using a gas/liquid inhomogeneous multiphase flow model coupled with a k-{omega} turbulence model for each phase. In the simulation, the surface drag was approached by a new correlation inside the Algebraic Interfacial Area Density (AIAD) model. The AIAD model allows the detection of the morphological form of the two phase flow and the corresponding switching via a blending function of each correlation from one object pair to another. As a result this model can distinguish between bubbles, droplets and the free surface using the local liquid phase volume fraction value. A comparison with the high-speed video observations shows a good qualitative agreement. The results indicated that quantitative agreement of the CCFL characteristics between calculation and experimental data was obtained. The goal is to provide an easy usable AIAD framework for all Code users, with the possibility of the implementation of their own correlations.

  16. Numerical simulations of counter-current two-phase flow experiments in a PWR hot leg model using an interfacial area density model

    Energy Technology Data Exchange (ETDEWEB)

    Hohne, T.; Deendarlianto; Vallee, C.; Lucas, D.; Beyer, M., E-mail: t.hoehne@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf (HZDR), Inst. of Safety Research, Dresden (Germany)

    2011-07-01

    In order to improve the understanding of counter-current two-phase flows and to validate new physical models, CFD simulations of 1/3rd scale model of the hot leg of a German Konvoi PWR with rectangular cross section was performed. Selected counter-current flow limitation (CCFL) experiments at the Helmholtz-Zentrum Dresden- Rossendorf (HZDR) were calculated with ANSYS CFX 12.1 using the multi-fluid Euler-Euler modeling approach. The transient calculations were carried out using a gas/liquid inhomogeneous multiphase flow model coupled with a SST turbulence model for each phase. In the simulation, the surface drag was approached by a new correlation inside the Algebraic Interfacial Area Density (AIAD) model. The AIAD model allows the detection of the morphological form of the two phase flow and the corresponding switching via a blending function of each correlation from one object pair to another. As a result this model can distinguish between bubbles, droplets and the free surface using the local liquid phase volume fraction value. A comparison with the high-speed video observations shows a good qualitative agreement. The results indicated that quantitative agreement of the CCFL characteristics between calculation and experimental data was obtained. The goal is to provide an easy usable AIAD framework for all ANSYS CFX users, with the possibility of the implementation of their own correlations. (author)

  17. Within AREVA, FRAMATOME ANP and its worldwide experience with PWR and BWR fuels

    International Nuclear Information System (INIS)

    Watteau, Michel; Esteve, Bernard; Giese, Ulrich; Matheson, John

    2002-01-01

    Faced with obvious energy procurement security needs and the increasing concern about global warming, many countries are making a lucid analysis of their energy situation and reconsidering the multiple assets of nuclear energy. After the European Commission's Green Paper evaluation which was endorsed by the European Parliament, the United States gave a strong signal to the whole world by deciding to extent the operating life time of its existing NPPs and by envisioning the construction of new ones. In Asia, here in Korea, and in Japan, the People's Republic of China, Taiwan, large-scale nuclear power plant programs are being pursued. It was in this context, with the aim of ever-greater competitiveness, that the AREVA group was conceived. The aim is for all our skills to have a higher profile on the international markets, so that we are in a stronger position to develop a leadership in our two main high tech sectors of interconnect - electronics and nuclear. In the nuclear sector, the pooling of the Cogema and Framatome ANP forces is enabling AREVA to offer a comprehensive service package ranging from uranium mining to decommissioning, encompassing the design and construction of plants and their fuel; AREVA's experience is grounded in unequalled know-how. Further, with the CEA, a multidisciplinary research organization in charge of anticipating the emerging technologies, as a close partner, AREVA has a unique strategic vision. With this set-up, AREVA has the financial resources it needs to forge the alliances necessary for its development, so that it can best confront international competition and meet the requirements of its customers world-wide

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

  19. Scaling criteria for rock dynamic experiments

    Energy Technology Data Exchange (ETDEWEB)

    Crowley, Barbara K [Lawrence Radiation Laboratory, University of California, Livermore, CA (United States)

    1970-05-01

    A set of necessary conditions for performing scaled rock dynamics experiments is derived from the conservation equations of continuum mechanics. Performing scaled experiments in two different materials is virtually impossible because of the scaling restrictions imposed by two equations of state. However, performing dynamically scaled experiments in the same material is possible if time and distance use the same scaling factor and if the effects of gravity are insignificant. When gravity becomes significant, dynamic scaling is no longer possible. To illustrate these results, example calculations of megaton and kiloton experiments are considered. (author00.

  20. Application of the Particle Swarm Optimization (PSO) technique to the thermal-hydraulics project of a PWR reactor core in reduced scale; Aplicacao da tecnica de otimizacao por enxame de particulas no projeto termo-hidraulico em escala reduzida do nucleo de um reator PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lima Junior, Carlos Alberto de Souza

    2008-09-15

    The reduced scale models design have been employed by engineers from several different industries fields such as offshore, spatial, oil extraction, nuclear industries and others. Reduced scale models are used in experiments because they are economically attractive than its own prototype (real scale) because in many cases they are cheaper than a real scale one and most of time they are also easier to build providing a way to lead the real scale design allowing indirect investigations and analysis to the real scale system (prototype). A reduced scale model (or experiment) must be able to represent all physical phenomena that occurs and further will do in the real scale one under operational conditions, e.g., in this case the reduced scale model is called similar. There are some different methods to design a reduced scale model and from those two are basic: the empiric method based on the expert's skill to determine which physical measures are relevant to the desired model; and the differential equation method that is based on a mathematical description of the prototype (real scale system) to model. Applying a mathematical technique to the differential equation that describes the prototype then highlighting the relevant physical measures so the reduced scale model design problem may be treated as an optimization problem. Many optimization techniques as Genetic Algorithm (GA), for example, have been developed to solve this class of problems and have also been applied to the reduced scale model design problem as well. In this work, Particle Swarm Optimization (PSO) technique is investigated as an alternative optimization tool for such problem. In this investigation a computational approach, based on particle swarm optimization technique (PSO), is used to perform a reduced scale two loop Pressurized Water Reactor (PWR) core, considering 100% of nominal power operation on a forced flow cooling circulation and non-accidental operating conditions. A performance

  1. Noise analysis and mimic experiments for loose part accident in the primary coolant loop of PWR

    International Nuclear Information System (INIS)

    Yang Xiuzhou; Cheng Tingxiang; Zhang Bin

    1994-01-01

    The basic principle of loose part monitoring is to detect and measure the structure transfer sound generated by impacting of metal loose part with accelerators and to identify and diagnose by the micro-processor. This paper introduces the theoretical base of loose part monitoring, the location and mass estimation of loose part, and three mimic experiment applying noise analysis techniques. It provides some useful preparations for the development of loose part monitoring system

  2. French experience in transient data collection and fatigue monitoring of PWR's nuclear steam supply system

    International Nuclear Information System (INIS)

    Sabaton, M.; Morilhat, P.; Savoldelli, D.; Genette, P.

    1995-10-01

    Electricite de France (EDF), the french national electricity company, is operating 54 standardized pressurizer water reactors. This about 500 reactor-years experience in nuclear stations operation and maintenance area has allowed EDF to develop its own strategy for monitoring of age-related degradations of NPP systems and components relevant for plant safety and reliability. After more than fifteen years of experience in regulatory transient data collection and seven years of successful fatigue monitoring prototypes experimentation, EDF decided to design a new system called SYSFAC (acronym for SYsteme de Surveillance en FAtigue de la Chaudiere) devoted to transient logging and thermal fatigue monitoring of the reactor coolant pressure boundary. The system is fully automatic and directly connected to the on-site data acquisition network without any complementary instrumentation. A functional transient detection module and a mechanical transient detection module are in charge of the general transient data collection. A fatigue monitoring module is aimed towards a precise surveillance of five specific zones particularly sensible to thermal fatigue. After the first step of preliminary studies, the industrial phase of the SYSFAC project is currently going on, with hardware and software tests and implementation. The first SYSFAC system will be delivered to the pilot power plant by the beginning of 1996. The extension to all EDF's nuclear 900 MW is planned after one more year of feedback experience. (authors). 12 refs., 3 figs

  3. Recent advances in PWR fuel design and performance experience at ABB-CENF

    International Nuclear Information System (INIS)

    Corsetti, Lawrence V.

    2004-01-01

    Utilities in the United States continue to move towards longer cycles and higher burnups to improve fuel cycle economics. This has placed increased demands for improved burnable absorber concepts. Zircaloy-4 corrosion behavior remains a high burnup performance concern. Over the past several years there has also been increasing utility interest in fuel reliability improvements. The development and application of erbia as a burnable absorber mixed directly with urania fuel will be discussed. Debris resistant fuel assembly designs and operating experience are reviewed. Oxide thickness measurements showing the improved corrosion resistance of optimized, low-tin Zircaloy-4 are presented. (author)

  4. ROX PWR

    International Nuclear Information System (INIS)

    Akie, H.; Yamashita, T.; Shirasu, N.; Takano, H.; Anoda, Y.; Kimura, H.

    1999-01-01

    For an efficient burnup of excess plutonium from nuclear reactors spent fuels and dismantled warheads, plutonium rock-like oxide(ROX) fuel has been investigated. The ROX fuel is expected to provide high Pu transmutation capability, irradiation stability and chemical and geological stability. While, a zirconia-based ROX(Zr-ROX)-fueled PWR core has some problems of Doppler reactivity coefficient and power peaking factor. For the improvement of these characteristics, two approaches were considered: the additives such as UO 2 , ThO 2 and Er 2 O 3 , and a heterogeneous core with Zr-ROX and UO 2 assemblies. As a result, the additives UO 2 + Er 2 O 3 are found to sufficiently improve the reactivity coefficients and accident behavior, and to flatten power distribution. On the other hand, in the 1/3Zr-ROX + 2/3UO 2 heterogeneous core, further reduction of power peaking seems necessary. (author)

  5. Corrosion in PWR stainless steel components: a TSO perspective based on operating experience and expertises

    International Nuclear Information System (INIS)

    Curieres, I. de

    2015-01-01

    Stainless steels are used commonly in many circuits of a nuclear power plant. Particularly, they are the prime materials for the inside surface of the primary circuit. Their operating experience has been good, though a number of cases of degradations due to corrosion have been reported the last ten years. This number of events is increasing and many studies of damaged parts become available. Based on the operating experience and these studies, IRSN will provide its perspective on the safety-related issues associated with the corrosion of stainless steel components. It appears that today's knowledge is not sufficient to define relevant criteria or to determine the exact set of parameters which leads to SCC (Stress Corrosion Cracking) of stainless steels. As a consequence, the best strategy remains an inspection and repair/replacement one. Moreover many cases show the influence of pollutants in the SCC events. This emphasizes the fact that chemistry parameters are strongly connected to safety issues, with respect to the stainless steels integrity

  6. French PWR safety philosophy

    International Nuclear Information System (INIS)

    Conte, M.

    1986-05-01

    Increasing knowledge and lessons learned from starting and operating experience of French nuclear power plants, completed by the experience learned from the operation of foreign reactors, has contributed to the improvement of French PWR design and safety philosophy. Based on a deterministic approach, the French safety analysis was progressively completed by a probabilistic approach, each of them having possibilities and limits. As a consequence of the global risk objective set in 1977 for nuclear reactors, safety analysis was extended to the evaluation of events more complex than the conventional ones, and later to the evaluation of the feasibility of the offsite emergency plans in case of severe accidents

  7. Advanced Scale Conditioning Agent (ASCA) Applications: 2012 Experience Update

    International Nuclear Information System (INIS)

    Little, Michael-J.; Varrin, Robert-D.; Pellman, Aaron-T.; Kreider, Marc A.

    2012-09-01

    ASCAs are a group of dilute chemical treatments for removing deposited corrosion products from the secondary side of PWR steam generators (SGs). Each ASCA formulation is customized to achieve plant-specific goals that can include: - Partial dissolution and structural modification of the tube scale present on free span surfaces through full bundle treatment, lowering the deposit loading and enhancing SG thermal performance levels through creation of a scale structure marked by increased boiling efficiency, - Softening and partial removal of deposits present in the broached flow holes in the tube support plates, reducing the risks of level oscillations and flow-induced vibration (FIV), - Chemical removal of copper from tube scale and tube sheet deposits, reducing the risk of rapid tube corrosion caused by the oxidized conditions promoted by some copper species, and - Dissolution of hardness species from consolidated top-of-tube sheet (TTS) collars to enhance collar removal through water-jetting and other mechanical cleaning techniques. Regardless of the cleaning objectives for a particular plant, all ASCA processes are designed to minimize corrosion, waste disposal costs, and impact on the outage schedule. To date, about 40 ASCA applications have been carried out in four (4) countries. This paper provides an update of the industry experience gained during these applications, including results demonstrating the ability of ASCA processes to meet the goals outlined above. Experience at multiple units, including several repeat ASCA applications, has demonstrated significant heat-transfer benefits (i.e., steam pressure increases of up to 1-2 bar (15-30 psi)). ASCA applications also regularly achieve significant reductions in TSP blockage (i.e., up to 30% absolute increases in available flow area in broached flow holes) and have been successful in eliminating level oscillations caused by excessive broached-hole blockage. (authors)

  8. Design of the control room of the N4-type PWR: main features and feedback operating experience; La salle de commande du palier N4: principales caracteristiques et retour d'experience d'exploitation

    Energy Technology Data Exchange (ETDEWEB)

    Peyrouton, J.M.; Guillas, J.; Nougaret, Ch. [Electricite de France (EDF/DPN/CAPE), 93 - Saint-Denis (France)

    2004-07-01

    This article presents the design, specificities and innovating features of the control room of the N4-type PWR. A brief description of control rooms of previous 900 MW and 1300 MW -type PWR allows us to assess the change. The design of the first control room dates back to 1972, at that time 2 considerations were taken into account: first the design has to be similar to that of control rooms for thermal plants because plant operators were satisfied with it and secondly the normal operating situation has to be privileged to the prejudice of accidental situations just as it was in a thermal plant. The turning point was the TMI accident that showed the weight of human factor in accidental situations in terms of pilot team, training, procedures and the ergonomics of the work station. The impact of TMI can be seen in the design of 1300 MW-type PWR. In the beginning of the eighties EDF decided to launch a study for a complete overhaul of the control room concept, the aim was to continue reducing the human factor risk and to provide a better quality of piloting the plant in any situation. The result is the control room of the N4-type PWR. Today the cumulated feedback experience of N4 control rooms represents more than 20 years over a wide range of situations from normal to incidental, a survey shows that the N4 design has fulfilled its aims. (A.C.)

  9. ROX PWR

    Energy Technology Data Exchange (ETDEWEB)

    Akie, H.; Yamashita, T.; Shirasu, N.; Takano, H.; Anoda, Y.; Kimura, H. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-12-01

    For an efficient burnup of excess plutonium from nuclear reactors spent fuels and dismantled warheads, plutonium rock-like oxide(ROX) fuel has been investigated. The ROX fuel is expected to provide high Pu transmutation capability, irradiation stability and chemical and geological stability. While, a zirconia-based ROX(Zr-ROX)-fueled PWR core has some problems of Doppler reactivity coefficient and power peaking factor. For the improvement of these characteristics, two approaches were considered: the additives such as UO{sub 2}, ThO{sub 2} and Er{sub 2}O{sub 3}, and a heterogeneous core with Zr-ROX and UO{sub 2} assemblies. As a result, the additives UO{sub 2}+ Er{sub 2}O{sub 3} are found to sufficiently improve the reactivity coefficients and accident behavior, and to flatten power distribution. On the other hand, in the 1/3Zr-ROX + 2/3UO{sub 2} heterogeneous core, further reduction of power peaking seems necessary. (author)

  10. Experience of Areva in fuel services for PWR and BWR; Experiencia de Areva en servicios de combustible para PWR y BWR

    Energy Technology Data Exchange (ETDEWEB)

    Morales, I.

    2015-07-01

    AREVA being an integrated supplier of fuel assemblies has included in its strategy to develop services and solutions to customers who desire to improve the performance and safety of their fuel. These services go beyond the simple 'after sale' services that can be expected from a fuel supplier: The portfolio of AREVA includes a wide variety of services, from scientific calculations to fuel handling services in a nuclear power plant. AREVA is committed to collaborate and to propose best-in-class solutions that really make the difference for the customer, based on 40 years of Fuel design and manufacturing experience. (Author)

  11. CHF experiments of tight pitch lattice rod bundles under PWR pressure condition for development of reduced moderation water reactor

    International Nuclear Information System (INIS)

    Araya, Fumimasa; Nakatsuka, Toru; Yoritsune, Tsutomu

    2002-10-01

    In order to improve plutonium utilization, design studies of reduced moderation water reactors which have hard neutron energy spectrum have been carried out at Division of Energy System Research of Japan Atomic Energy Research Institute (JAERI). At present, triangle, tight pitch lattice cores with about 1 mm gap width between fuel rods have been focused in the neutronic core design. Since a degradation of the heat removal from the fuel rods is worried, an evaluation of heat removal capability i.e. critical heat flux becomes one of important evaluation items in the feasibility study. However, any of published data base, which can be applicable to the evaluation on such narrow gap width cores, does not exist. Therefore, in the present study, in order to accumulate applicable data and to confirm applicability of an evaluation methodology of critical heat flux, basic experiments on the critical heat flux were performed using the test sections consisted of 7 heater rods bundles with the gap widths of 1.5, 1.0 and 0.6 mm under the PWR pressure conditions. The present report describes the experimental apparatus, experimental conditions and accumulated data. Analysis results of the data and the applicability of the evaluation methodology used for the design work are also discussed in this report. As the results of the experiment, it was found that the critical heat flux increased as the mass flux and the inlet subcooling increased. In the region of the mass flux less than about 2,000 kg/m 2 /s, the critical heat flux decreased as the gap width decreased. In the larger mass flux region, obvious trend of effects of the gap width on critical heat flux were not observed due to data scatterings. The flow-area-averaged thermal-equilibrium quality at the CHF position was in the higher ranges from 0.3 to 0.8 in the cases of gap widths of 1.0 and 0.6 mm, and 0.1 to 0.3 in the 1.5 mm case. Based on the experimental results such that the CHFs occurred in the higher quality range and

  12. Pressure vessel fracture studies pertaining to a PWR LOCA-ECC thermal shock: experiments TSE-3 and TSE-4 and update of TSE-1 and TSE-2 analysis

    International Nuclear Information System (INIS)

    Cheverton, R.D.; Bolt, S.E.

    1977-01-01

    The LOCA-ECC Thermal Shock Program was established to investigate the potential for flaw propagation in pressurized-water reactor (PWR) vessels during injection of emergency core coolant following a loss-of-coolant accident. Studies thus far have included fracture mechanics analyses of typical PWRs, the design and construction of a thermal shock test facility, determination of material properties for test specimens, and four thermal shock experiments with 0.53-m-OD (21-in.) by 0.15-m-wall (6-in.) cylindrical test specimens. In the first experiment, initiation was not expected and did not occur, although there was a small amount of subcritical crack growth. In the second experiment, initiation of a semicircular flaw took place as expected; the final length along the surface was about four times the initial length, but there was no radial growth. The third and fourth experiments were similar, and the long axial flaw initiated in good agreement with predictions

  13. Method and Result of Experiment for Support of Technical Solutions in the Field of Perfection of a Nuclear Fuel Cycle for Future PWR Reactors

    International Nuclear Information System (INIS)

    Ostrovskiy, V.; Kudryavtsev, E.; Tutnov, I.

    2011-01-01

    The paper presents the basics of approach of planning and carrying out of experiments to validate safety PWR reactors of the future when accepting technical solutions concerning using of improved fuel rods in fuel assembly. Basic principles and criteria used for the validation of technical solutions and developments in improving of nuclear fuel cycle of PWR reactors of the future are presented from the point of safety of future operation of modified fuel rods. We explore the questions of safety operation of PWR reactors with fuel assemblies, containing fuel rods with different length of fuel. The paper discusses the ways of solving of important tasks of critical facility experiments conducting for verification of new technical solutions in the sphere of PWR nuclear fuel cycle improvement on the base of international standards ISO 2000:9000 and functional safety recommendations of IEC (International Electromechanical Commission). New Federal laws of Russian Federation define the main principle for demands to NPP and any supplier of nuclear techniques. The principle is 'quantity indicators of risk should not exceed comprehensible social size of the established indicators of safety for any moment of operation of NPP'. On the other hand the second principle should be applied to extraction of the greatest benefit from operation of the equipment, systems or the NPP as whole: 'The long operation and full commercial use of resource and service properties of the equipment, systems and the NPP as a whole'. Realization of this principle assumes development and introduction of new technical solutions for a validation of guarantees of safety of the future operation of NPP or it separate components. Solving the practical problems of a validation of safety use of fuel rods with the increased length of a fuel column in fuel assembly in nuclear reactors of the future, we should choose new strategies and programs of verification experiments on the base of the analysis of guarantees

  14. The scaling of experiments on volcanic systems

    Directory of Open Access Journals (Sweden)

    Olivier eMERLE

    2015-06-01

    Full Text Available In this article, the basic principles of the scaling procedure are first reviewed by a presentation of scale factors. Then, taking an idealized example of a brittle volcanic cone intruded by a viscous magma, the way to choose appropriate analogue materials for both the brittle and ductile parts of the cone is explained by the use of model ratios. Lines of similarity are described to show that an experiment simulates a range of physical processes instead of a unique natural case. The pi theorem is presented as an alternative scaling procedure and discussed through the same idealized example to make the comparison with the model ratio procedure. The appropriateness of the use of gelatin as analogue material for simulating dyke formation is investigated. Finally, the scaling of some particular experiments such as pyroclastic flows or volcanic explosions is briefly presented to show the diversity of scaling procedures in volcanology.

  15. College Experience Scale (EExU

    Directory of Open Access Journals (Sweden)

    Angélica Juárez

    2017-07-01

    Full Text Available The experience of being a university student (University Experiences had been poorly studied so far. However, research in this field can provide valuable information about the quality of academic life, wellbeing or stress in this population. There is a lack of psychological tests that explore this theoretical construct. The aim of this study was to develop and validate a scale for measuring Univeristy Experiences, for that reason 314 college students were invited to participate for the validation. This students coursed different careers and reported 20 years old as average age. The University Experiences Scale (EExU has adequate psychometric properties. It has a structure of four factors: experience satisfaction, support perception, experience perception and life style adjustment. This factors explain 43.1% of the variance. The grouping of the factors of the College Experience Scale concurs with data reported in the literature about such concept, however this is the first questionnaire designed for measuring it. We anticipate that future studies will seek to verify the performance of the scale in different populations of students and analyze its psychometric properties and its possible association with other psychological variables that affect college students and their health.

  16. Evaluation of PWR steam generator water hammer. Final technical report, June 1, 1976--December 31, 1976

    International Nuclear Information System (INIS)

    Block, J.A.; Crowley, C.J.; Rothe, P.H.; Wallis, G.B.; Young, L.R.

    1977-05-01

    An investigation of waterhammer in the main feedwater piping of PWR steam generators due to water slugs formed in the steam generator feedring is reported. The relevant evidence from PWR operation and testing is compiled and summarized. The state-of-the-art of analysis of related phenomena is reviewed. Original exploratory modeling experiments at 1 / 10 and 1 / 4 scale are reported. Bounding analyses of the behavior are performed and several key phenomena have been identified for the first time. Recommendations to the Nuclear Regulatory Commission are made

  17. Direct Drive Wave Energy Buoy – 33rd scale experiment

    Energy Technology Data Exchange (ETDEWEB)

    Rhinefrank, Kenneth E. [Columbia Power Technologies, Inc.; Lenee-Bluhm, Pukha [Columbia Power Technologies, Inc.; Prudell, Joseph H. [Columbia Power Technologies, Inc.; Schacher, Alphonse A.; Hammagren, Erik J.; Zhang, Zhe [Columbia Power Technologies, Inc.

    2013-07-29

    Columbia Power Technologies (ColPwr) and Oregon State University (OSU) jointly conducted a series of tests in the Tsunami Wave Basin (TWB) at the O.H. Hinsdale Wave Research Laboratory (HWRL). These tests were run between November 2010 and February 2011. Models at 33rd scale representing Columbia Power’s Manta series Wave Energy Converter (WEC) were moored in configurations of one, three and five WEC arrays, with both regular waves and irregular seas generated. The primary research interest of ColPwr is the characterization of WEC response. The WEC response will be investigated with respect to power performance, range of motion and generator torque/speed statistics. The experimental results will be used to validate a numerical model. The primary research interests of OSU include an investigation into the effects of the WEC arrays on the near- and far-field wave propagation. This report focuses on the characterization of the response of a single WEC in isolation. To facilitate understanding of the commercial scale WEC, results will be presented as full scale equivalents.

  18. Corrosion of PWR steam generators

    International Nuclear Information System (INIS)

    Garnsey, R.

    1979-01-01

    Some designs of pressurized water reactor (PWR) steam generators have experienced a variety of corrosion problems which include stress corrosion cracking, tube thinning, pitting, fatigue, erosion-corrosion and support plate corrosion resulting in 'denting'. Large international research programmes have been mounted to investigate the phenomena. The operational experience is reviewed and mechanisms which have been proposed to explain the corrosion damage are presented. The implications for design development and for boiler and feedwater control are discussed. (author)

  19. Engineering-scale dust control experiments

    International Nuclear Information System (INIS)

    Winberg, M.R.; Pawelko, R.J.; Jacobs, N.C.; Thompson, D.N.

    1990-12-01

    This report presents the results of engineering scale dust-control experiments relating to contamination control during handling of transuranic waste. These experiments focused on controlling dust during retrieval operations of buried waste where waste and soil are intimately mixed. Sources of dust generation during retrieval operations include digging, dumping, and vehicle traffic. Because contaminants are expected to attach to soil particles and move with the generated dust, control of the dust spread may be the key to contamination control. Dust control techniques examined in these experiments include the use of misting systems, soil fixatives, and dust suppression agents. The Dryfog Ultrasonic Misting Head, manufactured by Sonics, Incorporated, and ENTAC, an organic resin derived from tree sap manufactured by ENTAC Corporation, were tested. The results of the experiments include product performance and recommended application methods. 19 figs., 7 refs., 6 tabs

  20. Experience feedback examination in PWR type reactors operating for the 1997-1999 period; Examen du retour d'experience en exploitation des reacteurs a eau sous pression pour la periode 1997-1999

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    The present report is relative to the examination that the permanent group has made on the experience feedback in operation for PWR type reactors for the period 1997-1999 that was on eleven themes chosen by the Nuclear Safety and Radiation Protection Authority. It used analysis reports made by I.R.S.N. in support of four meetings of the permanent group devoted to this examination from April 2001 to June 2002. The different themes were operating uncertainties, machining to vibrations, analysis of incidents and gaseous releases, circuits, human factors, behaviour of electric batteries, risk of cold source loss. (N.C.)

  1. An induction linac injector for scaled experiments

    International Nuclear Information System (INIS)

    Rutkowski, H.L.; Faltens, A.; Pike, C.; Brodzik, D.; Johnson, R.M.; Vanecek, D.; Hewett, D.W.

    1991-04-01

    An injector is being developed at LBL that would serve as the front end of a scaled induction linac accelerator technology experiment for heavy ion fusion. The ion mass being used is in the range 10--18. It is a multi-beam device intended to accelerate up to 2 MeV with 500 mA in each beam. The first half of the accelerating column has been built and experiments with one carbon beam are underway at the 1 MeV level. 5 refs., 1 fig

  2. A scaling study of the natural circulation flow of the ex-vessel core catcher cooling system of a 1400MW PWR for designing a scale-down test facility

    International Nuclear Information System (INIS)

    Rhee, Bo. W.; Ha, K. S.; Park, R. J.; Song, J. H.

    2012-01-01

    A scaling study on the steady state natural circulation flow along the flow path of the ex-vessel core catcher cooling system of 1400MWe PWR is described. The scaling criteria for reproducing the same thermalhydraulic characteristics of the natural circulation flow as the prototype core catcher cooling system in the scale-down test facility is derived and the resulting natural circulation flow characteristics of the prototype and scale-down facility analyzed and compared. The purpose of this study is to apply the similarity law to the prototype EU-APR1400 core catcher cooling system and the model test facility of this prototype system and derive a relationship between the heating channel characteristics and the down-comer piping characteristics so as to determine the down-comer pipe size and the orifice size of the model test facility. As the geometry and the heating wall heat flux of the heating channel of the model test facility will be the same as those of the prototype core catcher cooling system except the width of the heating channel is reduced, the axial distribution of the coolant quality (or void fraction) is expected to resemble each other between the prototype and model facility. Thus using this fact, the down-comer piping design characteristics of the model facility can be determined from the relationship derived from the similarity law

  3. ILSE-ESQ injector scaled experiment

    International Nuclear Information System (INIS)

    Henestroza, E.; Eylon, S.; Yu, S.; Grote, D.

    1993-01-01

    A 2 MeV, 800 mA, K + injector for the Heavy Ion Fusion Induction Linac Systems Experiments (ISLE) is under development at LBL. It consists of a 500keV-1MeV diode pre-injector followed by an electrostatic quadrupole accelerator (ESQ). One of the key issues for the ESQ centers around the control of beam aberrations due to the open-quotes energy effectclose quotes: in a strong electrostatic quadrupole field, ions at beam edge will have energies very different from those on the axis. The resulting kinematic distortions lead to S-shaped phase spaces, which, if uncorrected, will lead eventually to emittance growth. These beam aberrations can be minimized by increasing the injection energy and/or strengthening the beam focusing. It may also be possible to compensate for the open-quotes energy effectclose quotes by proper shaping of the quadrupoles electrodes. In order to check the physics of the open-quotes energy effectclose quotes of the ESQ design a scaled experiment has been designed that will accommodate the parameters of the source, as well as the voltage limitations, of the Single Beam Transport Experiment (SBTE). Since the 500 KeV pre-injector delivers a 4 cm converging beam, a quarter-scale experiment will fit the 1 cm converging beam of the SBTE source. Also, a 10 mA beam in SBTE, and the requirement of equal perveance in both systems, forces all the voltages to scale down by a factor 0.054. Results from this experiment and corresponding 3D PIC simulations will be presented

  4. ILSE-ESQ injector scaled experiment

    International Nuclear Information System (INIS)

    Henestroza, E.; Eylon, S.; Yu, S.; Grote, D.

    1993-05-01

    A 2 MeV, 800 mA, K + injector for the Heavy Ion Fusion Induction Linac Systems Experiments (ISLE) is under development at LBL. It consists of a 500 keV-1MeV diode preinjector followed by an electrostatic quadrupole accelerator (ESQ). One of the key issues for the ESQ centers around the control of beam aberrations due to the ''energy effect'': in a strong electrostatic quadrupole field, ions at beam edge will have energies very different from those on the axis. The resulting kinematic distortions lead to S-shaped phase spaces, which, if uncorrected, will lead eventually to emittance growth. These beam aberrations can be minimized by increasing the injection energy and/or strengthening the beam focusing. It may also be possible to compensate for the ''energy effect'' by proper shaping of the quadrupoles electrodes. In order to check the physics of the ''energy effect'' of the ESQ design a scaled experiment has been designed that will accommodate the parameters of the source, as well as the voltage limitations, of the Single Beam Transport Experiment (SBTE). Since the 500 keV pre-injector delivers a 4 cm converging beam, a quarter-scale experiment will fit the 1 cm converging beam of the SBTE source. Also, a 10 mA beam in SBTE, and the requirement of equal perveance in both systems, forces all the voltages to scale down by a factor 0.054. Results from this experiment and corresponding 3D PIC simulations will be presented

  5. French PWR Safety Philosophy

    International Nuclear Information System (INIS)

    Conte, M. M.

    1986-01-01

    The first 900 MWe units, built under the American Westinghouse licence and with reference to the U. S. regulation, were followed by 28 standardized units, C P1 and C P2 series. Increasing knowledge and lessons learned from starting and operating experience of French nuclear power plants, completed by the experience learned from the operation of foreign reactors, has contributed to the improvement of French PWR design and safety philosophy. As early as 1976, this experience was taken into account by French Safety organisms to discuss, with Electricite de France, the safety options for the planned 1300 MWe units, P4 and P4 series. In 1983, the new reactor scheduled, Ni4 series 1400 MWe, is a totally French design which satisfies the French regulations and other French standards and codes. Based on a deterministic approach, the French safety analysis was progressively completed by a probabilistic approach each of them having possibilities and limits. Increasing knowledge and lessons learned from operating experience have contributed to the French safety philosophy improvement. The methodology now applied to safety evaluation develops a new facet of the in depth defense concept by taking highly unlikely events into consideration, by developing the search of safety consistency of the design, and by completing the deterministic approach by the probabilistic one

  6. PWR plant construction in Japan

    International Nuclear Information System (INIS)

    Tamura, Toshifumi

    2002-01-01

    The construction methods based on the experiences on the Nuclear Island, which is a critical path in the total construction schedule, have been studied and reconsidered in order to construct by more reliable and economical method. So various improved construction method are being applied and the duration of construction is being reduced continuously. So various improved construction method are being applied and the duration of construction is being reduced continuously. In this paper, the history of construction of twenty-three (23) PWR Plant, the actual construction methods and schedule of Ohi-3/4, to which the many improved methods were applied during their construction, are introduced mainly with the improved points for previously constructed plants. And also the situation of construction method for the next PWR Plant is simply explained

  7. Full Scale Experiment with Interactive Urban Lighting

    DEFF Research Database (Denmark)

    Poulsen, Esben Skouboe; Andersen, Hans Jørgen; Jensen, Ole B.

    2012-01-01

    and region of occupancy of persons in the town square were monitored in real time by computer vision analyses of thermal images from 3 cameras monitoring the twin square. The results of the computer vision analyses were used to control the illumination from 16 3.5 meter high RGB LED Lamps that were......This paper presents and discusses the results of a full-scale interactive urban illumination experiment. The experiment investigates how human motion intensities can be used as input for controlling the illumination of a town square in the city of Aalborg in Denmark. The trajectory, velocity...... distributed across the square in an irregular grid. The lamps were DMX controlled. Using architectural models as sketching tools, 4 different illumination designs were developed and tested for a week in January. The result shows that in general people immersed in the square did not notice that the light...

  8. The PWR cores management

    International Nuclear Information System (INIS)

    Barral, J.C.; Rippert, D.; Johner, J.

    2000-01-01

    During the meeting of the 25 january 2000, organized by the SFEN, scientists and plant operators in the domain of the PWR debated on the PWR cores management. The five first papers propose general and economic information on the PWR and also the fast neutron reactors chains in the electric power market: statistics on the electric power industry, nuclear plant unit management, the ITER project and the future of the thermonuclear fusion, the treasurer's and chairman's reports. A second part offers more technical papers concerning the PWR cores management: performance and optimization, in service load planning, the cores management in the other countries, impacts on the research and development programs. (A.L.B.)

  9. Highlights of the French program on PWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Pages, J P [CEA Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires

    1997-12-01

    The presentation reviews the French programme on PWR fuel including the overall results of the year 1996 for nuclear operation; fuel management and economy; French nuclear electricity generation sites; production of nuclear generated electricity; energy availability of the 900 and 1,300 Mw PWR units; average radioactive liquid releases excluding tritium per unit; plutonium recycling experience.

  10. Stress intensity factor at the tip of cladding incipient crack in RIA-simulating experiments for high-burnup PWR fuels

    International Nuclear Information System (INIS)

    Udagawa, Yutaka; Suzuki, Motoe; Sugiyama, Tomoyuki; Fuketa, Toyoshi

    2009-01-01

    RIA-simulating experiments for high-burnup PWR fuels have been performed in the NSRR, and the stress intensity factor K 1 at the tip of cladding incipient crack has been evaluated in order to investigate its validity as a PCMI failure threshold under RIA conditions. An incipient crack depth was determined by observation of metallographs. The maximum hydride-rim thickness in the cladding of the test fuel rod was regarded as the incipient crack depth in each test case. Hoop stress in the cladding periphery during the pulse power transient was calculated by the RANNS code. K 1 was calculated based on crack depth and hoop stress. According to the RANNS calculation, PCMI failure cases can be divided into two groups: failure in the elastic phase and failure in the plastic phase. In the former case, elastic deformation was predominant around the incipient crack at failure time. K 1 is available only in this case. In the latter, plastic deformation was predominant around the incipient crack at failure time. Failure in the elastic phase never occurred when K 1 was less than 17 MPa m 1/2 . For failure in the plastic phase, the plastic hoop strain of the cladding periphery at failure time clearly showed a tendency to decrease with incipient crack depth. The combination of K 1 , for failure in the elastic phase, and plastic hoop strain at failure, for failure in the plastic phase, can be an effective index of PCMI failure under RIA conditions. (author)

  11. Pre-test prediction and post-test analysis of PWR fuel rod ballooning in the MT-3 in-pile LOCA simulation experiment in the NRU reactor

    International Nuclear Information System (INIS)

    Donaldson, A.T.; Horwood, R.A.; Healey, T.

    1983-01-01

    The USNRC and the UKAEA have jointly funded a series of in-pile LOCA simulation experiments in the Canadian NRU reactor in order to secure further information on the thermal hydraulic and clad deformation response of PWR fuel rod bundles. Test MT-3 in the series was performed using reflood rate and rod internal pressure conditions specified by the UK nuclear industry. The parameters were selected to ensure the development of a near-isothermal clad temperature history during which zircaloy was required to balloon and rupture near the alpha-alpha/beta phase transition. Specification of the reflood rate conditions was assisted by the performance of a precursor test on an unpressurised rod bundle and by complementary application of appropriate thermal hydraulic analyses. Identification of the rod internal pressure needed to cause ballooning and rupture was achieved using a creep deformation model, BALLOON, in conjunction with the clad thermal history defined by the prior thermal hydraulic test. This paper presents the basis of the BALLOON analysis and describes its application in calculating the fill gas pressure for rods MT-3, their axial ballooning profile and the clad temperature at peak radial strain elevations. (author)

  12. Large Scale Experiments on Spacecraft Fire Safety

    Science.gov (United States)

    Urban, David; Ruff, Gary A.; Minster, Olivier; Fernandez-Pello, A. Carlos; Tien, James S.; Torero, Jose L.; Legros, Guillaume; Eigenbrod, Christian; Smirnov, Nickolay; Fujita, Osamu; hide

    2012-01-01

    Full scale fire testing complemented by computer modelling has provided significant knowhow about the risk, prevention and suppression of fire in terrestrial systems (cars, ships, planes, buildings, mines, and tunnels). In comparison, no such testing has been carried out for manned spacecraft due to the complexity, cost and risk associated with operating a long duration fire safety experiment of a relevant size in microgravity. Therefore, there is currently a gap in knowledge of fire behaviour in spacecraft. The entire body of low-gravity fire research has either been conducted in short duration ground-based microgravity facilities or has been limited to very small fuel samples. Still, the work conducted to date has shown that fire behaviour in low-gravity is very different from that in normal gravity, with differences observed for flammability limits, ignition delay, flame spread behaviour, flame colour and flame structure. As a result, the prediction of the behaviour of fires in reduced gravity is at present not validated. To address this gap in knowledge, a collaborative international project, Spacecraft Fire Safety, has been established with its cornerstone being the development of an experiment (Fire Safety 1) to be conducted on an ISS resupply vehicle, such as the Automated Transfer Vehicle (ATV) or Orbital Cygnus after it leaves the ISS and before it enters the atmosphere. A computer modelling effort will complement the experimental effort. Although the experiment will need to meet rigorous safety requirements to ensure the carrier vehicle does not sustain damage, the absence of a crew removes the need for strict containment of combustion products. This will facilitate the possibility of examining fire behaviour on a scale that is relevant to spacecraft fire safety and will provide unique data for fire model validation. This unprecedented opportunity will expand the understanding of the fundamentals of fire behaviour in spacecraft. The experiment is being

  13. Scaled Eagle Nebula Experiments on NIF

    Energy Technology Data Exchange (ETDEWEB)

    Pound, Marc W. [Univ. of Maryland, College Park, MD (United States)

    2017-03-28

    We performed scaled laboratory experiments at the National Ignition Facility laser to assess models for the creation of pillar structures in star-forming clouds of molecular hydrogen, in particular the famous Pillars of the Eagle Nebula. Because pillars typically point towards nearby bright ultraviolet stars, sustained directional illumination appears to be critical to pillar formation. The experiments mock up illumination from a cluster of ultraviolet-emitting stars, using a novel long duration (30--60 ns), directional, laser-driven x-ray source consisting of multiple radiation cavities illuminated in series. Our pillar models are assessed using the morphology of the Eagle Pillars observed with the Hubble Space Telescope, and measurements of column density and velocity in Eagle Pillar II obtained at the BIMA and CARMA millimeter wave facilities. In the first experiments we assess a shielding model for pillar formation. The experimental data suggest that a shielding pillar can match the observed morphology of Eagle Pillar II, and the observed Pillar II column density and velocity, if augmented by late time cometary growth.

  14. Small-Scale Experiments.10-gallon drum experiment summary

    Energy Technology Data Exchange (ETDEWEB)

    Rosenberg, David M.

    2015-02-05

    A series of sub-scale (10-gallon) drum experiments were conducted to characterize the reactivity, heat generation, and gas generation of mixtures of chemicals believed to be present in the drum (68660) known to have breached in association with the radiation release event at the Waste Isolation Pilot Plant (WIPP) on February 14, 2014, at a scale expected to be large enough to replicate the environment in that drum but small enough to be practical, safe, and cost effective. These tests were not intended to replicate all the properties of drum 68660 or the event that led to its breach, or to validate a particular hypothesis of the release event. They were intended to observe, in a controlled environment and with suitable diagnostics, the behavior of simple mixtures of chemicals in order to determine if they could support reactivity that could result in ignition or if some other ingredient or event would be necessary. There is a significant amount of uncertainty into the exact composition of the barrel; a limited sub-set of known components was identified, reviewed with Technical Assessment Team (TAT) members, and used in these tests. This set of experiments was intended to provide a framework to postulate realistic, data-supported hypotheses for processes that occur in a “68660-like” configuration, not definitively prove what actually occurred in 68660.

  15. Large Scale Experiments on Spacecraft Fire Safety

    DEFF Research Database (Denmark)

    Urban, David L.; Ruff, Gary A.; Minster, Olivier

    2012-01-01

    -based microgravity facilities or has been limited to very small fuel samples. Still, the work conducted to date has shown that fire behaviour in low-gravity is very different from that in normal-gravity, with differences observed for flammability limits, ignition delay, flame spread behaviour, flame colour and flame......Full scale fire testing complemented by computer modelling has provided significant knowhow about the risk, prevention and suppression of fire in terrestrial systems (cars, ships, planes, buildings, mines, and tunnels). In comparison, no such testing has been carried out for manned spacecraft due...... to the complexity, cost and risk associ-ated with operating a long duration fire safety experiment of a relevant size in microgravity. Therefore, there is currently a gap in knowledge of fire behaviour in spacecraft. The entire body of low-gravity fire research has either been conducted in short duration ground...

  16. Psychometric properties of the Dissociative Experiences Scale.

    Science.gov (United States)

    Dubester, K A; Braun, B G

    1995-04-01

    The test-retest reliability of the Dissociative Experiences Scale (DES; Bernstein EM, Putnam FW [1986] Development, reliability, and validity of a dissociation scale. The Journal of Nervous and Mental Disease 174:727-735) in a clinical sample was found to be .93 for the total DES score and .95, .89, and .82 for the three subscale scores of amnesia, depersonalization-derealization, and absorption (dissociative identity disorder [DID], DSM-IV), respectively. Test-retest reliabilities within diagnostic groups of multiple personality disorder, dissociative disorder not otherwise specified, and a general other category of psychiatric diagnoses were obtained for total and subscale scores on the DES. These ranged from .78 to .96. Tests of mean scores across the two test sessions showed the total and subscale scores to be temporally stable. The DES was also found to be highly internally consistent: Cronbach's alphas of .96 and .97 were observed for the total DES scores taken at times 1 and 2, respectively. Construct validity of the DES was demonstrated by differentiation among the subscale scores in a repeated-measures analysis of variance (F[2,154] = 32.03, p < or = .001). Normality and general distribution issues were also addressed and provided a rationale for using the DES with parametric statistics. Reasons why the DES (as it was originally designed) is not appropriate as a dependent measure in outcome research are discussed, along with needed future research. Implications of the findings for the clinical usefulness of the DES as a diagnostic instrument are noted.

  17. Scaling laws for spherical pinch experiments

    International Nuclear Information System (INIS)

    Singh, D.P.; Palleschi, V.; Vaselli, M.

    1991-01-01

    In spherical pinch (SP) experiments, the plasma heated at the center of a cell to reach ignition temperature is confined by imploding shock waves for a time long enough to satisfy the Lawson criterion for plasma fusion. In earlier theoretical studies, the expansion of the central plasma either is neglected or is assumed to be radially uniform. The energy is considered to be deposited instantaneously at the center of the cell and the nonlinear heat conduction equation is solved to study the temporal evolution of the central plasma. Incorporating the ignition condition for the average temperature of the expanding fireball, and its confinement by imploding convergent shock waves, which may be fired from the periphery of the cell with some time delay, the scaling laws for satisfying the Lawson criterion are investigated in detail. The relevant calculations indicate that the cumulative effects of the convergent shock waves in the vicinity of the center of the cell play an important role in these scaling laws. (author)

  18. Application of the “best representativity” method to a PWR fuel calculation using the critical experiments at the Toshiba NCA facility

    International Nuclear Information System (INIS)

    Umano, Takuya; Yoshioka, Kenichi; Obara, Toru

    2015-01-01

    reactor (PWR) fuel assembly. Under combination of NCA critical experiments, the representativity factor became closer to unity. Simultaneously, a correction of the infinite multiplication factor was realized. Results were firstly compared among different combinations of experiments. Comparisons with the results of other calculation methods were also conducted. Whole results were explained with physical considerations

  19. Image-processing of time-averaged interface distributions representing CCFL characteristics in a large scale model of a PWR hot-leg pipe geometry

    International Nuclear Information System (INIS)

    Al Issa, Suleiman; Macián-Juan, Rafael

    2017-01-01

    Highlights: • CCFL characteristics are investigated in PWR large-scale hot-leg pipe geometry. • Image processing of air-water interface produced time-averaged interface distributions. • Time-averages provide a comparative method of CCFL characteristics among different studies. • CCFL correlations depend upon the range of investigated water delivery for Dh ≫ 50 mm. • 1D codes are incapable of investigating CCFL because of lack of interface distribution. - Abstract: Countercurrent Flow Limitation (CCFL) was experimentally investigated in a 1/3.9 downscaled COLLIDER facility with a 190 mm pipe’s diameter using air/water at 1 atmospheric pressure. Previous investigations provided knowledge over the onset of CCFL mechanisms. In current article, CCFL characteristics at the COLLIDER facility are measured and discussed along with time-averaged distributions of the air/water interface for a selected matrix of liquid/gas velocities. The article demonstrates the time-averaged interface as a useful method to identify CCFL characteristics at quasi-stationary flow conditions eliminating variations that appears in single images, and showing essential comparative flow features such as: the degree of restriction at the bend, the extension and the intensity of the two-phase mixing zones, and the average water level within the horizontal part and the steam generator. Consequently, making it possible to compare interface distributions obtained at different investigations. The distributions are also beneficial for CFD validations of CCFL as the instant chaotic gas/liquid interface is impossible to reproduce in CFD simulations. The current study shows that final CCFL characteristics curve (and the corresponding CCFL correlation) depends upon the covered measuring range of water delivery. It also shows that a hydraulic diameter should be sufficiently larger than 50 mm in order to obtain CCFL characteristics comparable to the 1:1 scale data (namely the UPTF data). Finally

  20. Measuring experience of hospitality : scale development and validation

    NARCIS (Netherlands)

    Pijls-Hoekstra, Ruth; Groen, Brenda H.; Galetzka, Mirjam; Pruyn, Adriaan T.H.

    This paper describes the development of the Experience of Hospitality Scale (EH-Scale) for assessing hospitality in service environments from a guest point of view. In contrast to other scales, which focus specifically on staff behaviour, the present scale focuses on the experience of hospitality

  1. Measuring the experience of hospitality : Scale development and validation

    NARCIS (Netherlands)

    Pijls-Hoekstra, Ruth; Groen, Brenda H.; Galetzka, Mirjam; Pruyn, Adriaan T.H.

    2017-01-01

    This paper identifies what customers experience as hospitality and subsequently presents a novel and compact assessment scale for measuring customers’ experience of hospitality at any kind of service organization. The Experience of Hospitality Scale (EH-Scale) takes a broader perspective compared to

  2. F.E.M. of PWR's control rod cluster. Parametrical study of vibrating behavior by an Experiment Design method

    International Nuclear Information System (INIS)

    Bosselut, D.; Soulier, B.

    1997-03-01

    Some finite element models have been performed at EDF to simulate the vibrations of rod cluster and to analyse the wear phenomenon of rods using parametrical studies. In the first part, one of the finite element models is presented. The location of excitation sources is described. The calculated values are: rod displacement in the guiding cards, shock forces on the guiding cards and wear power produced. In the second part, a parametrical study is presented for a given computer experiment domain with an Experimental Design method. The building of the computer experiment design is described. The used polynomial model has all linear, quadratic and interactive terms for each of the 6 parameters (26 coefficients), 34 polynomials have been built to approach the effective shock forces and the mean wear power at each of the 17 guiding points. In the last part, the influence of parameters on calculated mean wear power is shown along rods and some responses surfaces are visualized. Systematism and closeness of experiment design technique is underlined. Easy simulation of all the response domain by polynomial approach, allows comparison with experiment feedback. (author)

  3. ABB advanced BWR and PWR fuel

    International Nuclear Information System (INIS)

    Junkrans, S.; Helmersson, S.; Andersson, S.

    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 BWR and PWR. 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 10x10 BWR fuel, where ABB is the only vendor to date with multi batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of BWR and PWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its customers. (orig.)

  4. Engineering development of a digital replacement protection system at an operating US PWR nuclear power plant: Installation and operational experiences

    Energy Technology Data Exchange (ETDEWEB)

    Miller, M.H. [Duke Power Co., Seneca, SC (United States)

    1995-04-01

    The existing Reactor Protection Systems (RPSs) at most US PWRs are systems which reflect 25 to 30 year-old designs, components and manufacturing techniques. Technological improvements, especially in relation to modern digital systems, offer improvements in functionality, performance, and reliability, as well as reductions in maintenance and operational burden. The Nuclear power industry and the US nuclear regulators are poised to move forward with the issues that have slowed the transition to modern digital replacements for nuclear power plant safety systems. The electric utility industry is now more than ever being driven by cost versus benefit decisions. Properly designed, engineered, and installed digital systems can provide adequate cost-benefit and allow continued nuclear generated electricity. This paper describes various issues and areas related to an ongoing RPS replacement demonstration project which are pertinant for a typical US nuclear plant to consider cost-effective replacement of an aging analog RPS with a modern digital RPS. The following subject areas relative to the Oconee Nuclear Station ISAT{trademark} Demonstrator project are discussed: Operator Interface Development; Equipment Qualification; Validation and Verification of Software; Factory Testing; Field Changes and Verification Testing; Utility Operational, Engineering and Maintenance; Experiences with Demonstration System; and Ability to operate in parallel with the existing Analog RPS.

  5. Engineering development of a digital replacement protection system at an operating US PWR nuclear power plant: Installation and operational experiences

    International Nuclear Information System (INIS)

    Miller, M.H.

    1995-01-01

    The existing Reactor Protection Systems (RPSs) at most US PWRs are systems which reflect 25 to 30 year-old designs, components and manufacturing techniques. Technological improvements, especially in relation to modern digital systems, offer improvements in functionality, performance, and reliability, as well as reductions in maintenance and operational burden. The Nuclear power industry and the US nuclear regulators are poised to move forward with the issues that have slowed the transition to modern digital replacements for nuclear power plant safety systems. The electric utility industry is now more than ever being driven by cost versus benefit decisions. Properly designed, engineered, and installed digital systems can provide adequate cost-benefit and allow continued nuclear generated electricity. This paper describes various issues and areas related to an ongoing RPS replacement demonstration project which are pertinant for a typical US nuclear plant to consider cost-effective replacement of an aging analog RPS with a modern digital RPS. The following subject areas relative to the Oconee Nuclear Station ISAT trademark Demonstrator project are discussed: Operator Interface Development; Equipment Qualification; Validation and Verification of Software; Factory Testing; Field Changes and Verification Testing; Utility Operational, Engineering and Maintenance; Experiences with Demonstration System; and Ability to operate in parallel with the existing Analog RPS

  6. Improvement of two calculation methods of the detectors activation and on the PWR 900 MWe vessel. Comparison with the experiment

    International Nuclear Information System (INIS)

    Kitsos, S.

    1992-11-01

    The conditions of the vessel ageing (damages through irradiation) that mostly determine the life time of a nuclear reactor depend on the dose received. For the determination of this dose we use two calculation methods: one exact method using the TRIPOLI code that solves the Boltzmann equation with the Monte-Carlo method and one simplified method based on the point-kernel method. The advantages of the second method which is fast (easy reproduction of the results) and deterministic (the effects of difference are possible) compared to the first one which is long and statistical make its development necessary. The qualifications of these methods are done by comparison with the experiment that we reach as following: for the first method, we check the programming of TRIPOLI code (representativity of the collision) and its alignment with SN codes, and we modify the second method in order to use variable linear attenuation coefficients inside each medium to represent better the effects of the spectrum and of the reflection. As part of the checking of the basic physical data and their mode of representation, we present a study of the influence of the energy group averaging of the cross sections and of the number of the groups, as well as study of the influence of the cross sections origin

  7. PWR core design calculations

    International Nuclear Information System (INIS)

    Trkov, A.; Ravnik, M.; Zeleznik, N.

    1992-01-01

    Functional description of the programme package Cord-2 for PWR core design calculations is presented. Programme package is briefly described. Use of the package and calculational procedures for typical core design problems are treated. Comparison of main results with experimental values is presented as part of the verification process. (author) [sl

  8. Next generation PWR

    International Nuclear Information System (INIS)

    Tanaka, Toshihiko; Fukuda, Toshihiko; Usui, Shuji

    2001-01-01

    Development of LWR for power generation in Japan has been intended to upgrade its reliability, safety, operability, maintenance and economy as well as to increase its capacity in order, since nuclear power generation for commercial use was begun on 1970, to steadily increase its generation power. And, in Japan, ABWR (advanced BWR) of the most promising LWR in the world, was already used actually and APWR (advanced PWR) with the largest output in the world is also at a step of its actual use. And, development of the APWR in Japan was begun on 1980s, and is at a step of plan on construction of its first machine at early of this century. However, by large change of social affairs, economy of nuclear power generation is extremely required, to be positioned at an APWR improved development reactor promoted by collaboration of five PWR generation companies and the Mitsubishi Electric Co., Ltd. Therefore, on its development, investigation on effect of change in social affairs on nuclear power stations was at first carried out, to establish a design requirement for the next generation PWR. Here were described on outline, reactor core design, safety concept, and safety evaluation of APWR+ and development of an innovative PWR. (G.K.)

  9. SBLOCA AND LOFW EXPERIMENTS IN A SCALED-DOWN IET FACILITY OF REX-10 REACTOR

    Directory of Open Access Journals (Sweden)

    YEON-GUN LEE

    2013-06-01

    Full Text Available This paper presents an experimental investigation of the small-break loss-of-coolant accident (SBLOCA and the loss-of-feedwater accident (LOFW in a scaled integral test facility of REX-10. REX-10 is a small integral-type PWR in which the coolant flow is driven by natural circulation, and the RCS is pressurized by the steam-gas pressurizer. The postulated accidents of REX-10 include the system depressurization initiated by the break of a nitrogen injection line connected to the steam-gas pressurizer and the complete loss of normal feedwater flow by the malfunction of control systems. The integral effect tests on SBLOCA and LOFW are conducted at the REX-10 Test Facility (RTF, a full-height full-pressure facility with reduced power by 1/50. The SBLOCA experiment is initiated by opening a flow passage out of the pressurizer vessel, and the LOFW experiment begins with the termination of the feedwater supply into the helical-coil steam generator. The experimental results reveal that the RTF can assure sufficient cooldown capability with the simulated PRHRS flow during these DBAs. In particular, the RTF exhibits faster pressurization during the LOFW test when employing the steam-gas pressurizer than the steam pressurizer. This experimental study can provide unique data to validate the thermal-hydraulic analysis code for REX-10.

  10. Minor actinide transmutation on PWR burnable poison rods

    International Nuclear Information System (INIS)

    Hu, Wenchao; Liu, Bin; Ouyang, Xiaoping; Tu, Jing; Liu, Fang; Huang, Liming; Fu, Juan; Meng, Haiyan

    2015-01-01

    Highlights: • Key issues associated with MA transmutation are the appropriate loading pattern. • Commercial PWRs are the only choice to transmute MAs in large scale currently. • Considerable amount of MA can be loaded to PWR without disturbing k eff markedly. • Loading MA to PWR burnable poison rods for transmutation is an optimal loading pattern. - Abstract: Minor actinides are the primary contributors to long term radiotoxicity in spent fuel. The majority of commercial reactors in operation in the world are PWRs, so to study the minor actinide transmutation characteristics in the PWRs and ultimately realize the successful minor actinide transmutation in PWRs are crucial problem in the area of the nuclear waste disposal. The key issues associated with the minor actinide transmutation are the appropriate loading patterns when introducing minor actinides to the PWR core. We study two different minor actinide transmutation materials loading patterns on the PWR burnable poison rods, one is to coat a thin layer of minor actinide in the water gap between the zircaloy cladding and the stainless steel which is filled with water, another one is that minor actinides substitute for burnable poison directly within burnable poison rods. Simulation calculation indicates that the two loading patterns can load approximately equivalent to 5–6 PWR annual minor actinide yields without disturbing the PWR k eff markedly. The PWR k eff can return criticality again by slightly reducing the boric acid concentration in the coolant of PWR or removing some burnable poison rods without coating the minor actinide transmutation materials from PWR core. In other words, loading minor actinide transmutation material to PWR does not consume extra neutron, minor actinide just consumes the neutrons which absorbed by the removed control poisons. Both minor actinide loading patterns are technically feasible; most importantly do not need to modify the configuration of the PWR core and

  11. Scaling the drop size in coflow experiments

    International Nuclear Information System (INIS)

    Castro-Hernandez, E; Gordillo, J M; Gundabala, V; Fernandez-Nieves, A

    2009-01-01

    We perform extensive experiments with coflowing liquids in microfluidic devices and provide a closed expression for the drop size as a function of measurable parameters in the jetting regime that accounts for the experimental observations; this expression works irrespective of how the jets are produced, providing a powerful design tool for this type of experiments.

  12. Scaling the drop size in coflow experiments

    Energy Technology Data Exchange (ETDEWEB)

    Castro-Hernandez, E; Gordillo, J M [Area de Mecanica de Fluidos, Universidad de Sevilla, Avenida de los Descubrimientos s/n, 41092 Sevilla (Spain); Gundabala, V; Fernandez-Nieves, A [School of Physics, Georgia Institute of Technology, Atlanta, GA 30332 (United States)], E-mail: jgordill@us.es

    2009-07-15

    We perform extensive experiments with coflowing liquids in microfluidic devices and provide a closed expression for the drop size as a function of measurable parameters in the jetting regime that accounts for the experimental observations; this expression works irrespective of how the jets are produced, providing a powerful design tool for this type of experiments.

  13. PWR fuel behavior: lessons learned from LOFT

    International Nuclear Information System (INIS)

    Russell, M.L.

    1981-01-01

    A summary of the experience with the Loss-of-Fluid Test (LOFT) fuel during loss-of-coolant experiments (LOCEs), operational and overpower transient tests and steady-state operation is presented. LOFT provides unique capabilities for obtaining pressurized water reactor (PWR) fuel behavior information because it features the representative thermal-hydraulic conditions which control fuel behavior during transient conditions and an elaborate measurement system to record the history of the fuel behavior

  14. Status of developing advanced PWR in Japan

    International Nuclear Information System (INIS)

    Iida, Yotaro

    1982-01-01

    During past eleven years since the first PWR power plant, Mihama Unit 1 of Kansai Electric Power Co., started the commercial operation in 1970, Mitsubishi Heavy Industries has endeavored to improve PWR technologies on the basis of the advice from electric power companies and the technical information to overcome difficulties in PWR power plants. Now, the main objective is to improve the overall plant performance, and the rate of operation of Japanese PWR power plants has significantly risen. The improvement of the reliability, the shortening of regular inspection period and the reduction of radioactive waste handling were attempted. In view of the satisfactory operational experience of Westinghouse type PWRs, the basic reactor concept has not been changed so far. Mitsubishi and Westinghouse reached basic agreement in August, 1981, to develop a spectral shift type large capacity reactor as the advanced PWRs for Japan. This type of PWRs hab higher degree of freedom for extended fuel cycle operation and enhances the advantage of entire fuel cycle economy, particularly the significant reduction of uranium use. The improved neutron economy is attainable by reducing neutron loss, and the core design with low power density and the economical use of plutonium are advantageous for the fuel cycle economy. (Kako, I.)

  15. Plutonium recycling in PWR

    International Nuclear Information System (INIS)

    Youinou, G.; Girieud, R.; Guigon, B.

    2000-01-01

    Two concepts of 100% MOX PWR cores are presented. They are designed such as to minimize the consequences of the introduction of Pu on the core control. The first one has a high moderation ratio and the second one utilizes an enriched uranium support. The important design parameters as well as their capabilities to multi recycle Pu are discussed. We conclude with the potential interest of the two concepts. (author)

  16. The integrated PWR

    International Nuclear Information System (INIS)

    Gautier, G.M.

    2002-01-01

    This document presents the integrated reactors concepts by a presentation of four reactors: PIUS, SIR, IRIS and CAREM. The core conception, the operating, the safety, the economical aspects and the possible users are detailed. From the performance of the classical integrated PWR, the necessity of new innovative fuels utilization, the research of a simplified design to make easier the safety and the KWh cost decrease, a new integrated reactor is presented: SCAR 600. (A.L.B.)

  17. Biodegradation of creosote compounds: Comparison of experiments at different scales

    DEFF Research Database (Denmark)

    Broholm, K.; Arvin, Erik

    2001-01-01

    of the pyrroles on the biodegradation of benzene, and the biodegradation of benzothiophene occurs only in the presence of a primary substrate. The experiments show that some biodegradation processes of organic compounds may be common to different microorganisms.......This paper compares the results of biodegradation experiments with creosote compounds performed at different scales. The experiments include field observations, field experiments, large-scale intact laboratory column experiments, model fracture experiments, and batch experiments. Most...... of the experiments were conducted with till or ground water from the field site at Ringe on the island of Funen. Although the experiments were conducted on different scales, they revealed that some phenomena-e.g., an extensive biodegradation potential of several of the creosote compounds, the inhibitory influence...

  18. Reactor control system. PWR

    International Nuclear Information System (INIS)

    2009-01-01

    At present, 23 units of PWR type reactors have been operated in Japan since the start of Mihama Unit 1 operation in 1970 and various improvements have been made to upgrade operability of power stations as well as reliability and safety of power plants. As the share of nuclear power increases, further improvements of operating performance such as load following capability will be requested for power stations with more reliable and safer operation. This article outlined the reactor control system of PWR type reactors and described the control performance of power plants realized with those systems. The PWR control system is characterized that the turbine power is automatic or manually controlled with request of the electric power system and then the nuclear power is followingly controlled with the change of core reactivity. The system mainly consists of reactor automatic control system (control rod control system), pressurizer pressure control system, pressurizer water level control system, steam generator water level control system and turbine bypass control system. (T. Tanaka)

  19. AGR v PWR

    International Nuclear Information System (INIS)

    Green, D.

    1986-01-01

    When the Central Electricity Generating Board (CEGB) invited tenders and placed a contract for the Advanced Gas Cooled Reactor (AGR) at Dungeness B in 1965 -preferring it to the Pressurised Water Reactor (PWR) -the AGR was lamentably ill developed. The effects of the decision were widely felt, for it took the British nuclear industry off the light water reactor highway of world reactor business and up and idiosyncratic private highway of its own, excluding it altogether from any material export business in the two decades which followed. Yet although the UK may have made wrong decisions in rejecting the PWR in 1965, that does not mean that it can necessarily now either correct them, or redeem their consequence, by reversing the choice in 1985. In the 20 years since 1965 the whole world economic and energy picture has been transformed and the national picture with it. Picking up the PWR now could prove as big a disaster as rejecting it may have been in 1965. (author)

  20. Water chemistry in PWR

    International Nuclear Information System (INIS)

    Abe, Kenji

    1987-01-01

    This article outlines major features and basic concept of the secondary system of PWR's and water properties control measures adopted in recent PWR plants. The secondary system of a PWR consists of a condenser cooling pipe (aluminum-brass, titanium, or stainless steel), low-pressure make-up water heating pipe (aluminum-brass or stainless steel), high-ressure make-up water heating pipe (cupro-nickel or stainless steel), steam generator heat-transfer pipe (Inconel 600 or 690), and bleed/drain pipe (carbon steel, low alloy steel or stainless steel). Other major pipes and equipment are made of carbon steel or stainless steel. Major troubles likely to be caused by water in the secondary system include reduction in wall thickness of the heat-transfer pipe, stress corrosion cracking in the heat-transfer pipe, and denting. All of these are caused by local corrosion due to concentration of purities contained in water. For controlling the water properties in the secondary system, it is necessary to prevent impurities from entering the system, to remove impurities and corrosion products from the system, and to prevent corrosion of apparatus making up the system. Measures widely adopted for controlling the formation of IGA include the addition of boric acid for decreasing the concentration of free alkali and high hydrazine operation for providing a highly reducing atmospere. (Nogami, K.)

  1. 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.

  2. Experience with the INES scale in Egypt

    International Nuclear Information System (INIS)

    Rashad, S.M.

    1997-01-01

    Thirty years experience with Egypt first Research Reactor (ET-RR-1) operation, was introduced focusing on the famous events that were initiated and the procedures that were taken for their recovery or mitigation is given. Four out of seven events can be attributed to human errors, the events if classified using the INES

  3. PWR: 10 years after and perspectives

    International Nuclear Information System (INIS)

    1990-01-01

    These proceedings of the SFEN days on PWR (Ten years after and perspectives) comprise 13 conferences bearing on: - From the occurential approach to the state approach - Evolution of calculating tools - Human factors and safety - Reactor safety in the PWR 2000 - The PWR and the electrical power grid load follow - Fuel aspect of PWR management - PWR chemistry evolution - Balance of radiation protection - PWR modifications balance and influence on reactor operation - Design and maintenance of reactor components: 4 conferences [fr

  4. Application of the RELAP5 to the analysis of large scale integral experiments

    International Nuclear Information System (INIS)

    D'Auria, F.; Galassi, G. M.

    2000-01-01

    The present paper discusses the application of the code-modalisation to the analysis of experiments performed in the UPTF and the PANDA facility available in Germany and in Switzerland, respectively. The UPTF simulates all the internals of the reactor pressure vessel of a Pressurized Water Reactor with 1/1 scale of the geometric dimensions (diameters and lengths). The PANDA simulates the containment system of a Simplified Biling Water Reactor and the vessel of the reactor. The considered experimental dana base includes the occurrence of three-dimensional phenomena that are relevant to the refill phase of a large break Loss of Coolant Accident in a PWR and to the coupling between primary system and containment performance in a SBWR. The application of the code also required to adapt and to extend the methodology for modalisation development. The results from the qualitatively point of view are satisfactory as far as the performance of code-modalisation is concerned and do not show any important code limitation. They also confirm the suitability of the code for applications in the nuclear technology. However, this conclusion should be considered as preliminary because of the lack of independent proofs. (author)

  5. Some thermalhydraulics of closure head adapters in a 3 loops PWR

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, F.; Daubert, O.; Hecker, M. [EDF/DER/National Hydraulics Laboratory, Chatou (France)] [and others

    1995-09-01

    In 1993 a R&D action, based on numerical simulations and experiments on PWR`s upper head was initiated. This paper presents the test facility TRAVERSIN (a scale model of a 900 MW PWR adapter) and the calculations performed on the geometry of different upper head sections with the Thermalhydraulic Finite Element Code N3S used for 2D and 3D computations. The paper presents the method followed to bring the adapter and upper head study to a successful conclusion. Two complementary approaches are performed to obtain global results on complete fluid flow in the upper head and local results on the flow around the adapters of closure head. A validation test case of these experimental and numerical tools is also presented.

  6. The physics of musical scales: Theory and experiment

    Science.gov (United States)

    Durfee, Dallin S.; Colton, John S.

    2015-10-01

    The theory of musical scales involves mathematical ratios, harmonic resonators, beats, and human perception and provides an interesting application of the physics of waves and sound. We first review the history and physics of musical scales, with an emphasis on four historically important scales: twelve-tone equal temperament, Pythagorean, quarter-comma meantone, and Ptolemaic just intonation. We then present an easy way for students and teachers to directly experience the qualities of different scales using MIDI synthesis.

  7. The PWR spectral code GELS. Pt. 1

    International Nuclear Information System (INIS)

    Penndorf, K.; Schult, F.; Schulz, G.

    1976-01-01

    The code procedures group constant libraries for the static PWR design of whatever fuel cycle - Uranium, Thorium, or Plutonium. The whole reach of temperatures is covered and the treatment of strong lumped absorbers as control or burnable poison pins is included. The main features are: 1) Good accuracy in spite of not fitting the material data to critical experiments; 2) speed and relatively low computer equipment; 3) restriction to PWR's only. In case of demands for higher accuracy there is a further restriction concerning the library data of the epithermal resonance absorbers: They are strictly valid only for several special lattice geometrics. Three samples are given each representing a typical application of the code. Two of them likewise are demonstrations of recalculated experiments. (orig.) [de

  8. Full-scale and time-scale heating experiments at Stripa: preliminary results

    International Nuclear Information System (INIS)

    Cook, N.G.W.; Hood, Michael; California Univ., Berkeley

    1978-01-01

    Two full-scale heating experiments and a time-scale heating experiment have recently been started in granite 340 meters below surface. The purpose of the full-scale heating experiments is to assess the near-field effects of thermal loading for the design of an underground repository of nuclear wastes. That of the time-scale heating experiments is to obtain field data of the interaction between heaters and its effect on the rock mass during a period of about two years, which corresponds to about twenty years of full-scale operation. Geological features of the rock around each experiment have been mapped carefully, and temperatures, stresses and displacements induced in the rock by heating have been calculated in advance of the experiments. Some 800 different measurements are recorded at frequent intervals by a computer system situated underground. These data can be compared at any time with predictions made earlier on video display units underground

  9. Weighing Photons Using Bathroom Scales: A Thought Experiment

    Science.gov (United States)

    Huggins, Elisha

    2010-01-01

    Jay Orear, in his introductory physics text, defined the weight of a person as the reading one gets when standing on a (properly calibrated) bathroom scale. Here we will use Jay's definition of weight in a thought experiment to measure the weight of a photon. The thought experiment uses the results of the Pound-Rebka-Snider experiments, Compton…

  10. Dynamically Scaled Model Experiment of a Mooring Cable

    Directory of Open Access Journals (Sweden)

    Lars Bergdahl

    2016-01-01

    Full Text Available The dynamic response of mooring cables for marine structures is scale-dependent, and perfect dynamic similitude between full-scale prototypes and small-scale physical model tests is difficult to achieve. The best possible scaling is here sought by means of a specific set of dimensionless parameters, and the model accuracy is also evaluated by two alternative sets of dimensionless parameters. A special feature of the presented experiment is that a chain was scaled to have correct propagation celerity for longitudinal elastic waves, thus providing perfect geometrical and dynamic scaling in vacuum, which is unique. The scaling error due to incorrect Reynolds number seemed to be of minor importance. The 33 m experimental chain could then be considered a scaled 76 mm stud chain with the length 1240 m, i.e., at the length scale of 1:37.6. Due to the correct elastic scale, the physical model was able to reproduce the effect of snatch loads giving rise to tensional shock waves propagating along the cable. The results from the experiment were used to validate the newly developed cable-dynamics code, MooDy, which utilises a discontinuous Galerkin FEM formulation. The validation of MooDy proved to be successful for the presented experiments. The experimental data is made available here for validation of other numerical codes by publishing digitised time series of two of the experiments.

  11. Scaling and extrapolation of hydrogen distribution experiments

    International Nuclear Information System (INIS)

    Karwat, H.

    1986-01-01

    The containment plays an important role in predicting the residual risk to the environment under severe accident conditions. Risk analyses show that massive fission product release from the reactor fuel can occur only if during a loss of coolant the core is severely damaged and a containment failure is anticipated. Large amounts of hydrogen inevitably, are formed during the core degradation and will be released into the containment. More combustible gases are produced later when the coremelt will contact the containment concrete. Thus a potential for an early containment failure exists if a massive hydrogen deflagration cannot be excluded. A more remote cause for early containment failure may be an energetic steam explosion which requires a number of independent conditions when the molten core material contacts residual coolant water. The prediction of the containment loads caused by a hydrogen combustion is dependent on the prediction of the combustion mode. In the paper an attempt is made to identify on basis of a dimensional analysis such areas for which particular care must be exercised when scale experimental evidence is interpreted and extrapolated with the aid of a computer code or a system of computer codes. The study is restricted to fluid dynamic phenomena of the gas distribution process within the containment atmosphere. The gas sources and the mechanical response of containment structures are considered as given boundary conditions under which the containment is to be analyzed

  12. Developing the Cyber Victimization Experiences and Cyberbullying Behaviors Scales.

    Science.gov (United States)

    Betts, Lucy R; Spenser, Karin A

    2017-01-01

    The reported prevalence rates of cyber victimization experiences and cyberbullying behaviors vary. Part of this variation is likely due to the diverse definitions and operationalizations of the constructs adopted in previous research and the lack of psychometrically robust measures. Through 2 studies, the authors developed (Study 1) and evaluated (Study 2) the cyber victimization experiences and cyberbullying behaviors scales. Participants in Study 1 were 393 (122 boys, 171 girls) and in Study 2 were 345 (153 boys, 192 girls) 11-15-year-olds who completed measures of cyber victimization experiences, cyberbullying behaviors, face-to-face victimization experiences, face-to-face bullying behaviors, and social desirability. The 3-factor cyber victimization experiences scale comprised threat, shared images, and personal attack. The 3-factor cyberbullying behaviors scale comprised sharing images, gossip, and personal attack. Both scales demonstrated acceptable internal consistency and convergent validity.

  13. The Conceptual Design of Innovative Safe PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Han-Gon [Centural Research Institute, Daejeon (Korea, Republic of); Heo, Sun [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2016-10-15

    Most of countries operating NPPs have been performed post-Fukushima improvements as short-term countermeasure to enhance the safety of operating NPPs. Separately, vendors have made efforts on developing passive safety systems as long-term and ultimate countermeasures. AP1000 designed by Westinghouse Electric Company has passive safety systems including the passive emergency core cooling system (PECCS), the passive residual heat removal system (PRHRS), and the passive containment cooling system (PCCS). ESBWR designed by GE-Hitachi also has passive safety systems consisting of the isolation condenser system, the gravity driven cooling system and the PCCS. Other countries including China and Russia have made efforts on developing passive safety systems for enhancing the safety of their plants. In this paper, we summarize the design goals and main design feature of innovative safe PWR, iPOWER which is standing for Innovative Passive Optimized World-wide Economical Reactor, and show the developing status and results of research projects. To mitigate an accident without electric power and enhance the safety level of PWR, the conceptual designs of passive safety system and innovative safe PWR have been performed. It includes the PECCS for core cooling and the PCCS for containment cooling. Now we are performing the small scale and separate effect tests for the PECCS and the PCCS and preparing the integral effect test for the PECCS and real scale test for the PCCS.

  14. Experimental modelling of core debris dispersion from the vault under a PWR pressure vessel: Part 1

    International Nuclear Information System (INIS)

    Macbeth, R.V.; Trenberth, R.

    1987-12-01

    Modelling experiments have been done on a 1/25 scale model in Perspex of the vault under a PWR pressure vessel. Various liquids have been used to simulate molten core debris assumed to have fallen on to the vault floor from a breach at the bottom of the pressure vessel. High pressure air and helium have been used to simulate the discharge of steam and gas from the breach. The dispersion of liquid via the vault access shafts has been measured. Photographs have been taken of fluid flow patterns and velocity profiles have been obtained. The requirements for further experiments are indicated. (author)

  15. PWR decontamination feasibility study

    Energy Technology Data Exchange (ETDEWEB)

    Silliman, P.L.

    1978-12-18

    The decontamination work which has been accomplished is reviewed and it is concluded that it is worthwhile to investigate further four methods for decontamination for future demonstration. These are: dilute chemical; single stage strong chemical; redox processes; and redox/chemical in combination. Laboratory work is recommended to define the agents and processes for demonstration and to determine the effect of the solvents on PWR materials. The feasibility of Indian Point 1 for decontamination demonstrations is discussed, and it is shown that the system components of Indian Point 1 are well suited for use in demonstrations.

  16. PWR decontamination feasibility study

    International Nuclear Information System (INIS)

    Silliman, P.L.

    1978-01-01

    The decontamination work which has been accomplished is reviewed and it is concluded that it is worthwhile to investigate further four methods for decontamination for future demonstration. These are: dilute chemical; single stage strong chemical; redox processes; and redox/chemical in combination. Laboratory work is recommended to define the agents and processes for demonstration and to determine the effect of the solvents on PWR materials. The feasibility of Indian Point 1 for decontamination demonstrations is discussed, and it is shown that the system components of Indian Point 1 are well suited for use in demonstrations

  17. PWR core design calculations

    Energy Technology Data Exchange (ETDEWEB)

    Trkov, A; Ravnik, M; Zeleznik, N [Inst. Jozef Stefan, Ljubljana (Slovenia)

    1992-07-01

    Functional description of the programme package Cord-2 for PWR core design calculations is presented. Programme package is briefly described. Use of the package and calculational procedures for typical core design problems are treated. Comparison of main results with experimental values is presented as part of the verification process. (author) [Slovenian] Opisali smo programski paket CORD-2, ki se uporablja pri projektnih izracunih sredice pri upravljanju tlacnovodnega reaktorja. Prikazana je uporaba paketa in racunskih postopkov za tipicne probleme, ki nastopajo pri projektiranju sredice. Primerjava glavnih rezultatov z eksperimentalnimi vrednostmi je predstavljena kot del preveritvenega procesa. (author)

  18. QuickEval: a web application for psychometric scaling experiments

    Science.gov (United States)

    Van Ngo, Khai; Storvik, Jehans J.; Dokkeberg, Christopher A.; Farup, Ivar; Pedersen, Marius

    2015-01-01

    QuickEval is a web application for carrying out psychometric scaling experiments. It offers the possibility of running controlled experiments in a laboratory, or large scale experiment over the web for people all over the world. It is a unique one of a kind web application, and it is a software needed in the image quality field. It is also, to the best of knowledge, the first software that supports the three most common scaling methods; paired comparison, rank order, and category judgement. It is also the first software to support rank order. Hopefully, a side effect of this newly created software is that it will lower the threshold to perform psychometric experiments, improve the quality of the experiments being carried out, make it easier to reproduce experiments, and increase research on image quality both in academia and industry. The web application is available at www.colourlab.no/quickeval.

  19. Thermal-hydraulic study of integrated steam generator in PWR

    International Nuclear Information System (INIS)

    Osakabe, Masahiro

    1989-01-01

    One of the safety aspects of innovative reactor concepts is the integration of steam generators (SGs) into the reactor vessel in the case of the pressurized water reactor (PWR). All of the reactor system components including the pressurizer are within the reactor vessel in the SG integrated PWR. The simple heat transfer code was developed for the parametric study of the integrated SG. The code was compared to the once-through 19-tube SG experiment and the good agreement between the experimental results and the code predictions was obtained. The assessed code was used for the parametric study of the integrated once-through 16 m-straight-tube SG installed in the annular downcomer. The proposed integrated SG as a first attempt has approximately the same tube size and pitch as the present PWR and the SG primary and secondary sides in the present PWR is inverted in the integrated PWR. Based on the study, the reactor vessel size of the SG integrated PWR was calculated. (author)

  20. EPRI PWR primary water chemistry guidelines revision

    International Nuclear Information System (INIS)

    McElrath, Joel; Fruzzetti, Keith

    2014-01-01

    EPRI periodically updates the PWR Primary Water Chemistry Guidelines as new information becomes available and as required by NEI 97-06 (Steam Generator Program Guidelines) and NEI 03-08 (Guideline for the Management of Materials Issues). The last revision of the PWR water chemistry guidelines identified an optimum primary water chemistry program based on then-current understanding of research and field information. This new revision provides further details with regard to primary water stress corrosion cracking (PWSCC), fuel integrity, and shutdown dose rates. A committee of industry experts, including utility specialists, nuclear steam supply system (NSSS) and fuel vendor representatives, Institute of Nuclear Power Operations (INPO) representatives, consultants, and EPRI staff collaborated in reviewing the available data on primary water chemistry, reactor water coolant system materials issues, fuel integrity and performance issues, and radiation dose rate issues. From the data, the committee updated the water chemistry guidelines that all PWR nuclear plants should adopt. The committee revised guidance with regard to optimization to reflect industry experience gained since the publication of Revision 6. Among the changes, the technical information regarding the impact of zinc injection on PWSCC initiation and dose rate reduction has been updated to reflect the current level of knowledge within the industry. Similarly, industry experience with elevated lithium concentrations with regard to fuel performance and radiation dose rates has been updated to reflect data collected to date. Recognizing that each nuclear plant owner has a unique set of design, operating, and corporate concerns, the guidelines committee has retained a method for plant-specific optimization. Revision 7 of the Pressurized Water Reactor Primary Water Chemistry Guidelines provides guidance for PWR primary systems of all manufacture and design. The guidelines continue to emphasize plant

  1. PWR burnable absorber evaluation

    International Nuclear Information System (INIS)

    Cacciapouti, R.J.; Weader, R.J.; Malone, J.P.

    1995-01-01

    The purpose of the study was to evaluate the relative neurotic efficiency and fuel cycle cost benefits of PWR burnable absorbers. Establishment of reference low-leakage equilibrium in-core fuel management plans for 12-, 18- and 24-month cycles. Review of the fuel management impact of the integral fuel burnable absorber (IFBA), erbium and gadolinium. Calculation of the U 3 O 8 , UF 6 , SWU, fuel fabrication, and burnable absorber requirements for the defined fuel management plans. Estimation of fuel cycle costs of each fuel management plan at spot market and long-term market fuel prices. Estimation of the comparative savings of the different burnable absorbers in dollar equivalent per kgU of fabricated fuel. (author)

  2. PWR systems transient analysis

    International Nuclear Information System (INIS)

    Kennedy, M.F.; Peeler, G.B.; Abramson, P.B.

    1985-01-01

    Analysis of transients in pressurized water reactor (PWR) systems involves the assessment of the response of the total plant, including primary and secondary coolant systems, steam piping and turbine (possibly including the complete feedwater train), and various control and safety systems. Transient analysis is performed as part of the plant safety analysis to insure the adequacy of the reactor design and operating procedures and to verify the applicable plant emergency guidelines. Event sequences which must be examined are developed by considering possible failures or maloperations of plant components. These vary in severity (and calculational difficulty) from a series of normal operational transients, such as minor load changes, reactor trips, valve and pump malfunctions, up to the double-ended guillotine rupture of a primary reactor coolant system pipe known as a Large Break Loss of Coolant Accident (LBLOCA). The focus of this paper is the analysis of all those transients and accidents except loss of coolant accidents

  3. PWR degraded core analysis

    International Nuclear Information System (INIS)

    Gittus, J.H.

    1982-04-01

    A review is presented of the various phenomena involved in degraded core accidents and the ensuing transport of fission products from the fuel to the primary circuit and the containment. The dominant accident sequences found in the PWR risk studies published to date are briefly described. Then chapters deal with the following topics: the condition and behaviour of water reactor fuel during normal operation and at the commencement of degraded core accidents; the generation of hydrogen from the Zircaloy-steam and the steel-steam reactions; the way in which the core deforms and finally melts following loss of coolant; debris relocation analysis; containment integrity; fission product behaviour during a degraded core accident. (U.K.)

  4. Steam generators in PWR's

    International Nuclear Information System (INIS)

    Michel, R.

    1974-01-01

    The steam generator of the PWR operates according to the principle of natural circulation. It consists of a U-shaped tube bundle whose free ends are welded to a bottom plate. The tube bundle is surrounded by a cylinder jacket which has slots closely above the bottom or tube plate. The feed water mixed with boiling water enters the tube bundle through these slots. Because of its buoyancy, the steam-water mixture flows upwards. Below the tube plate there are chambers for distributing and collecting pressurized water separated by means of a partition wall. By omitting some tubes, a free alloy is created so that the tubes in the center get sufficient water, too. By asymmetrical arrangement of the partition wall it is further possible to limit the tube alloy only to the inlet side for pressurized water. The flow over the tube plate is thus improved on the inlet side. (DG) [de

  5. Materials performance in operating PWR steam generators

    International Nuclear Information System (INIS)

    Weeks, J.R.

    1975-01-01

    The Inconel-600 tubing in operating PWR steam generators has developed leaks due to intergranular stress corrosion cracking or a general wastage attack, originating from the secondary side of the tubing. Corrosion has been limited to those areas of the steam generators where limited coolant circulation and high heat flux have caused impurities to concentrate. Wastage or pitting attack has always been associated with local concentration of sodium hydrogen phosphates, whereas stress corrosion has been associated with local concentration of sodium or potassium hydroxides. The only instance of stress corrosion originating from the primary side occurred on cold-worked tubing when hydrogen was not added to getter oxygen, and LiOH was not added to raise the pH of the primary coolant. All PWR manufacturers are now recommending that the phosphate treatment of the secondary coolant be abandoned in favor of an all-volatile treatment. Experience in operating plants has shown, however, that removal of phosphate-rich sludge deposits is difficult, and that further wastage and/or intergranular stress corrosion may develop; the residual sodium phosphates gradually convert by reaction with corrosion product hydroxides to sodium hydroxide, which remains concentrated in the limited flow areas. Improvements in circulation patterns have been achieved by inserting flow baffles in some PWR steam generators. Inservice monitoring by eddy current techniques is useful for detecting corrosion-induced defects in the tubing, but irreproducibility in field examinations can lead to uncertainties interpreting the results. (U.S.)

  6. Safety aspects of the using Gd as burnable poison in PWR's

    International Nuclear Information System (INIS)

    Vandenberg, C.; Bonet, H.; Charlier, A.

    1978-01-01

    The experience of BELGONUCLEAIRE in using Gd in LWR's has indicated the safety related advantages of this burnable poison. The successfully operation of the BR3 PWR power plant with 5% of Gd rods is presented and extrapolated to large PWR's. (authro)

  7. Reconnection Scaling Experiment (RSX): Magnetic Reconnection in Linear Geometry

    Science.gov (United States)

    Intrator, T.; Sovinec, C.; Begay, D.; Wurden, G.; Furno, I.; Werley, C.; Fisher, M.; Vermare, L.; Fienup, W.

    2001-10-01

    The linear Reconnection Scaling Experiment (RSX) at LANL is a new experiment that can create MHD relevant plasmas to look at the physics of magnetic reconnection. This experiment can scale many relevant parameters because the guns that generate the plasma and current channels do not depend on equilibrium or force balance for startup. We describe the experiment and initial electrostatic and magnetic probe data. Two parallel current channels sweep down a long plasma column and probe data accumulated over many shots gives 3D movies of magnetic reconnection. Our first data tries to define an operating regime free from kink instabilities that might otherwise confuse the data and shot repeatability. We compare this with MHD 2 fluid NIMROD simulations of the single current channel kink stability boundary for a variety of experimental conditions.

  8. Development of Cost Estimation Methodology of Decommissioning for PWR

    International Nuclear Information System (INIS)

    Lee, Sang Il; Yoo, Yeon Jae; Lim, Yong Kyu; Chang, Hyeon Sik; Song, Geun Ho

    2013-01-01

    The permanent closure of nuclear power plant should be conducted with the strict laws and the profound planning including the cost and schedule estimation because the plant is very contaminated with the radioactivity. In Korea, there are two types of the nuclear power plant. One is the pressurized light water reactor (PWR) and the other is the pressurized heavy water reactor (PHWR) called as CANDU reactor. Also, the 50% of the operating nuclear power plant in Korea is the PWRs which were originally designed by CE (Combustion Engineering). There have been experiences about the decommissioning of Westinghouse type PWR, but are few experiences on that of CE type PWR. Therefore, the purpose of this paper is to develop the cost estimation methodology and evaluate technical level of decommissioning for the application to CE type PWR based on the system engineering technology. The aim of present study is to develop the cost estimation methodology of decommissioning for application to PWR. Through the study, the following conclusions are obtained: · Based on the system engineering, the decommissioning work can be classified as Set, Subset, Task, Subtask and Work cost units. · The Set and Task structure are grouped as 29 Sets and 15 Task s, respectively. · The final result shows the cost and project schedule for the project control and risk management. · The present results are preliminary and should be refined and improved based on the modeling and cost data reflecting available technology and current costs like labor and waste data

  9. Developing the cyber victimisation experiences and cyber bullying behaviours scales

    OpenAIRE

    Betts, LR; Spenser, KA

    2017-01-01

    The reported prevalence rates of cyber victimisation experiences and cyber bullying behaviours vary. Part of this variation is likely due to the diverse definitions and operationalisations of the constructs adopted in previous research and the lack of psychometrically robust measures. Through two studies, the current research developed (Study 1) and evaluated (Study 2) the cyber victimisation experiences and cyber bullying behaviours scales. In Study 1 393 (122 male, 171 female), and in Study...

  10. PWR [pressurized water reactor] pressurizer transient response: Final report

    International Nuclear Information System (INIS)

    Murphy, S.I.

    1987-08-01

    To predict PWR pressurizer transients, Ahl proposed a three region model with a universal coefficient to represent condensation on the water surface. Specifically, this work checks the need for three regions and the modeling of the interfacial condensation coefficient. A computer model has been formulated using the basic mass and energy conservation laws. A two region vapor and liquid model was first used to predict transients run on a one-eleventh scale Freon pressurizer. These predictions verified the need for a second liquid region. As a result, a three region model was developed and used to predict full-scale pressurizer transients at TMI-2, Shippingport, and Stade. Full-scale pressurizer predictions verified the three region model and pointed out the shortcomings of Ahl's universal condensation coefficient. In addition, experiments were run using water at low pressure to study interface condensation. These experiments showed interface condensation to be significant only when spray flow is turned on; this result was incorporated in the final three region model

  11. A Validation Study of the Adolescent Dissociative Experiences Scale

    Science.gov (United States)

    Keck Seeley, Susan. M.; Perosa, Sandra, L.; Perosa, Linda, M.

    2004-01-01

    Objective: The purpose of this study was to further the validation process of the Adolescent Dissociative Experiences Scale (A-DES). In this study, a 6-item Likert response format with descriptors was used when responding to the A-DES rather than the 11-item response format used in the original A-DES. Method: The internal reliability and construct…

  12. Large-scale silviculture experiments of western Oregon and Washington.

    Science.gov (United States)

    Nathan J. Poage; Paul D. Anderson

    2007-01-01

    We review 12 large-scale silviculture experiments (LSSEs) in western Washington and Oregon with which the Pacific Northwest Research Station of the USDA Forest Service is substantially involved. We compiled and arrayed information about the LSSEs as a series of matrices in a relational database, which is included on the compact disc published with this report and...

  13. Reproducible and controllable induction voltage adder for scaled beam experiments

    Energy Technology Data Exchange (ETDEWEB)

    Sakai, Yasuo; Nakajima, Mitsuo; Horioka, Kazuhiko [Department of Energy Sciences, Tokyo Institute of Technology, 4259 Nagatsuta, Midori-ku, Yokohama 226-8502 (Japan)

    2016-08-15

    A reproducible and controllable induction adder was developed using solid-state switching devices and Finemet cores for scaled beam compression experiments. A gate controlled MOSFET circuit was developed for the controllable voltage driver. The MOSFET circuit drove the induction adder at low magnetization levels of the cores which enabled us to form reproducible modulation voltages with jitter less than 0.3 ns. Preliminary beam compression experiments indicated that the induction adder can improve the reproducibility of modulation voltages and advance the beam physics experiments.

  14. Valve testing for UK PWR safety applications

    International Nuclear Information System (INIS)

    George, P.T.; Bryant, S.

    1989-01-01

    Extensive testing and development has been done by the Central Electricity Generating Board (CEGB) to support the design, construction and operation of Sizewell B, the UK's first PWR. A Blowdown Rig for the Assessment of Valve Operability - (BRAVO) has been constructed at the CEGB Marchwood Engineering Laboratory to reproduce PWR Pressurizer fluid conditions for the full scale testing of Pressurizer Relief System (PRS) valves. A full size tandem pair of Pilot Operated Safety Relief Valves (POSRVs) is being tested under the full range of pressurizer fluid conditions. Tests to date have produced important data on the performance of the valve in its Cold Overpressure protection mode of operation and on methods for the in-service testing of the valve. Also, a full size pressurizer safety valve has been tested under full PRS fluid conditions to develop a methodology for the pre-service testing of the Sizewell valves. Further work will be carried out to develop procedures for the in-service testing of the valve. In the Main Steam Safety Valve test program carried out at the Siemens-KWU Test Facilities, a single MSSV from three potential suppliers was tested under full secondary system conditions. The test results have been analyzed and are reflected in the CEGB's arrangements for the pre-service and in-service testing of the Sizewell MSSVs. Valves required to interrupt pipebreak flow must be qualified for this duty by testing or a combination of testing and analysis. To obtain guidance on the performance of such tests gate and globe valves have been subjected to simulated pipebreaks under PWR primary circuit conditions. In the light of problems encountered with gate valve closure under these conditions, further tests are currently being carried out on the BRAVO facility on a gate valve, in preparation for the full scale flow interruption qualification testing of the Sizewell main steam isolation valve

  15. TRANSPORT CHARACTERISTICS OF SELECTED PWR LOCA GENERATED DEBRIS

    International Nuclear Information System (INIS)

    MAJI, A. K.; MARSHALL, B.

    2000-01-01

    In the unlikely event of a Loss of Coolant Accident (LOCA) in a pressurized water reactor (PWR), break jet impingement would dislodge thermal insulation FR-om nearby piping, as well as other materials within the containment, such as paint chips, concrete dust, and fire barrier materials. Steam/water flows induced by the break and by the containment sprays would transport debris to the containment floor. Subsequently, debris would likely transport to and accumulate on the suction sump screens of the emergency core cooling system (ECCS) pumps, thereby potentially degrading ECCS performance and possibly even failing the ECCS. In 1998, the U. S. Nuclear Regulatory Commission (NRC) initiated a generic study (Generic Safety Issue-191) to evaluate the potential for the accumulation of LOCA related debris on the PWR sump screen and the consequent loss of ECCS pump net positive suction head (NPSH). Los Alamos National Laboratory (LANL), supporting the resolution of GSI-191, was tasked with developing a method for estimating debris transport in PWR containments to estimate the quantity of debris that would accumulate on the sump screen for use in plant specific evaluations. The analytical method proposed by LANL, to predict debris transport within the water that would accumulate on the containment floor, is to use computational fluid dynamics (CFD) combined with experimental debris transport data to predict debris transport and accumulation on the screen. CFD simulations of actual plant containment designs would provide flow data for a postulated accident in that plant, e.g., three-dimensional patterns of flow velocities and flow turbulence. Small-scale experiments would determine parameters defining the debris transport characteristics for each type of debris. The containment floor transport methodology will merge debris transport characteristics with CFD results to provide a reasonable and conservative estimate of debris transport within the containment floor pool and

  16. The Phoenix series large scale LNG pool fire experiments.

    Energy Technology Data Exchange (ETDEWEB)

    Simpson, Richard B.; Jensen, Richard Pearson; Demosthenous, Byron; Luketa, Anay Josephine; Ricks, Allen Joseph; Hightower, Marion Michael; Blanchat, Thomas K.; Helmick, Paul H.; Tieszen, Sheldon Robert; Deola, Regina Anne; Mercier, Jeffrey Alan; Suo-Anttila, Jill Marie; Miller, Timothy J.

    2010-12-01

    The increasing demand for natural gas could increase the number and frequency of Liquefied Natural Gas (LNG) tanker deliveries to ports across the United States. Because of the increasing number of shipments and the number of possible new facilities, concerns about the potential safety of the public and property from an accidental, and even more importantly intentional spills, have increased. While improvements have been made over the past decade in assessing hazards from LNG spills, the existing experimental data is much smaller in size and scale than many postulated large accidental and intentional spills. Since the physics and hazards from a fire change with fire size, there are concerns about the adequacy of current hazard prediction techniques for large LNG spills and fires. To address these concerns, Congress funded the Department of Energy (DOE) in 2008 to conduct a series of laboratory and large-scale LNG pool fire experiments at Sandia National Laboratories (Sandia) in Albuquerque, New Mexico. This report presents the test data and results of both sets of fire experiments. A series of five reduced-scale (gas burner) tests (yielding 27 sets of data) were conducted in 2007 and 2008 at Sandia's Thermal Test Complex (TTC) to assess flame height to fire diameter ratios as a function of nondimensional heat release rates for extrapolation to large-scale LNG fires. The large-scale LNG pool fire experiments were conducted in a 120 m diameter pond specially designed and constructed in Sandia's Area III large-scale test complex. Two fire tests of LNG spills of 21 and 81 m in diameter were conducted in 2009 to improve the understanding of flame height, smoke production, and burn rate and therefore the physics and hazards of large LNG spills and fires.

  17. Design of an experiment to measure the decay heat of an irradiated PWR fuel: MERCI experiment; Conception d'une experience de mesure de la puissance residuelle d'un combustible irradie: l'experience MERCI

    Energy Technology Data Exchange (ETDEWEB)

    Bourganel, St

    2002-11-01

    After a reactor shutdown, a significant quantity of energy known as 'decay heat' continues to be generated from the irradiated fuel. This heat source is due to the disintegration energy of fission products and actinides. Decay heat determination of an irradiated fuel is of the utmost importance for safety analysis as the design cooling systems, spent fuel transport, or handling. Furthermore, the uncertainty on decay heat has a straight economic impact. The unloading fuel spent time is an example. The purpose of MERCI experiment (irradiated fuel decay heat measurement) consists in qualifying computer codes, particularly the DARWIN code system developed by the CEA in relation to industrial organizations, as EDF, FRAMATOME and COGEMA. To achieve this goal, a UOX fuel is irradiated in the vicinity of the OSIRIS research reactor, and then the decay heat is measured by using a calorimeter. The objective is to reduce the decay heat uncertainties from 8% to 3 or 4% at short cooling times. A full simulation on computer of the MERCI experiment has been achieved: fuel irradiation analysis is performed using transport code TRIPOLI4 and evolution code DARWIN/PEPIN2, and heat transfer with CASTEM2000 code. The results obtained are used for the design of this experiment. Moreover, we propose a calibration procedure decreasing the influence of uncertainty measurements and an interpretation method of the experimental results and evaluation of associated uncertainties. (author)

  18. Design of an experiment to measure the decay heat of an irradiated PWR fuel: MERCI experiment; Conception d'une experience de mesure de la puissance residuelle d'un combustible irradie: l'experience MERCI

    Energy Technology Data Exchange (ETDEWEB)

    Bourganel, St

    2002-11-01

    After a reactor shutdown, a significant quantity of energy known as 'decay heat' continues to be generated from the irradiated fuel. This heat source is due to the disintegration energy of fission products and actinides. Decay heat determination of an irradiated fuel is of the utmost importance for safety analysis as the design cooling systems, spent fuel transport, or handling. Furthermore, the uncertainty on decay heat has a straight economic impact. The unloading fuel spent time is an example. The purpose of MERCI experiment (irradiated fuel decay heat measurement) consists in qualifying computer codes, particularly the DARWIN code system developed by the CEA in relation to industrial organizations, as EDF, FRAMATOME and COGEMA. To achieve this goal, a UOX fuel is irradiated in the vicinity of the OSIRIS research reactor, and then the decay heat is measured by using a calorimeter. The objective is to reduce the decay heat uncertainties from 8% to 3 or 4% at short cooling times. A full simulation on computer of the MERCI experiment has been achieved: fuel irradiation analysis is performed using transport code TRIPOLI4 and evolution code DARWIN/PEPIN2, and heat transfer with CASTEM2000 code. The results obtained are used for the design of this experiment. Moreover, we propose a calibration procedure decreasing the influence of uncertainty measurements and an interpretation method of the experimental results and evaluation of associated uncertainties. (author)

  19. Steam Generator Owners Group PWR secondary water chemistry guidelines

    International Nuclear Information System (INIS)

    Welty, C.S. Jr.; Green, S.J.

    1985-01-01

    In 1981 the Steam Generator Owners Group (SGOG), a group of domestic and foreign pressurized water reactor (PWR) owners, developed and issued the PWR secondary water chemistry guidelines. The guidelines were prepared in response to the growing recognition that a majority of the problems causing reduced steam generator reliability (e.g., denting, wasteage, pitting, etc.) were related to secondary (steam) side water purity. The guidelines were subsequently issued as an Electric Power Research Institute (EPRI) report. In 1984 they were revised to reflect industry experience in adopting the original issuance and to incorporate new information on causes of corrosion damage. The guidelines have been endorsed and their adoption recommended by the SGOG

  20. Advanced ion exchange resins for PWR condensate polishing

    International Nuclear Information System (INIS)

    Hoffman, B.; Tsuzuki, S.

    2002-01-01

    The severe chemical and mechanical requirements of a pressurized water reactor (PWR) condensate polishing plant (CPP) present a major challenge to the design of ion exchange resins. This paper describes the development and initial operating experience of improved cation and anion exchange resins that were specifically designed to meet PWR CPP needs. Although this paper focuses specifically on the ion exchange resins and their role in plant performance, it is also recognized and acknowledged that excellent mechanical design and operation of the CPP system are equally essential to obtaining good results. (authors)

  1. Development of the bullying and health experiences scale.

    Science.gov (United States)

    Beran, Tanya; Stanton, Lauren; Hetherington, Ross; Mishna, Faye; Shariff, Shaheen

    2012-11-09

    Until recently, researchers have studied forms of bullying separately. For 40 years, research has looked at the traditional forms of bullying, including physical (eg, hitting), verbal (eg, threats), and social (eg, exclusion). Attention focused on cyberbullying in the early 2000s. Although accumulating research suggests that bullying has multiple negative effects for children who are targeted, these effects excluded cyberbullying from the definition of bullying. This paper responds to the need for a multidimensional measure of the impact of various forms of bullying. We used a comprehensive definition of bullying, which includes all of its forms, to identify children who had been targeted or who had participated in bullying. We then examined various ways in which they were impacted. We used an online method to administer 37 impact items to 377 (277 female, 100 male) children and youth, to develop and test the Bullying and Health Experience Scale. A principal components analysis of the bullying impact items with varimax rotation resulted in 8 factors with eigenvalues greater than one, explaining 68.0% of the variance. These scales include risk, relationships, anger, physical injury, drug use, anxiety, self-esteem, and eating problems, which represent many of the cognitive, psychological, and behavioral consequences of bullying. The Cronbach alpha coefficients for the 8 scales range from .73 to .90, indicating good inter-item consistency. Comparisons between the groups showed that children involved in bullying had significantly higher negative outcomes on all scales than children not involved in bullying. The high Cronbach alpha values indicate that the 8 impact scales provide reliable scores. In addition, comparisons between the groups indicate that the 8 scales provide accurate scores, with more negative outcomes reported by children involved in bullying compared to those who are not involved in bullying. This evidence of reliability and validity indicates that

  2. Pressurizer and steam-generator behavior under PWR transient conditions

    International Nuclear Information System (INIS)

    Wahba, A.B.; Berta, V.T.; Pointner, W.

    1983-01-01

    Experiments have been conducted in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR), at the Idaho National Engineering Laboratory, in which transient phenomena arising from accident events with and without reactor scram were studied. The main purpose of the LOFT facility is to provide data for the development of computer codes for PWR transient analyses. Significant thermal-hydraulic differences have been observed between the measured and calculated results for those transients in which the pressurizer and steam generator strongly influence the dominant transient phenomena. Pressurizer and steam generator phenomena that occurred during four specific PWR transients in the LOFT facility are discussed. Two transients were accompanied by pressurizer inflow and a reduction of the heat transfer in the steam generator to a very small value. The other two transients were accompanied by pressurizer outflow while the steam generator behavior was controlled

  3. Development of Large-Scale Spacecraft Fire Safety Experiments

    DEFF Research Database (Denmark)

    Ruff, Gary A.; Urban, David L.; Fernandez-Pello, A. Carlos

    2013-01-01

    exploration missions outside of low-earth orbit and accordingly, more complex in terms of operations, logistics, and safety. This will increase the challenge of ensuring a fire-safe environment for the crew throughout the mission. Based on our fundamental uncertainty of the behavior of fires in low...... of the spacecraft fire safety risk. The activity of this project is supported by an international topical team of fire experts from other space agencies who conduct research that is integrated into the overall experiment design. The large-scale space flight experiment will be conducted in an Orbital Sciences...

  4. Make use of EDF orientations in PWR fuel management

    International Nuclear Information System (INIS)

    Gloaguen, A.

    1989-01-01

    The EDF experience acquired permits to allow the PWR fuel performances and to make use of better management. In this domain low progress can be given considerable financial profits. The industrial and commercial structures, the time constant of the fuel cycle, has for consequence that the electric utilities can take advantage only progressively of the expected profits [fr

  5. Non-dimensional scaling of impact fast ignition experiments

    International Nuclear Information System (INIS)

    Farley, D R; Shigemori, K; Murakami, M; Azechi, H

    2008-01-01

    Recent experiments at the Osaka University Institute for Laser Engineering (ILE) showed that 'Impact Fast Ignition' (IFI) could increase the neutron yield of inertial fusion targets by two orders of magnitude [1]. IFI utilizes the thermal and kinetic energy of a laser-accelerated disk to impact an imploded fusion target. ILE researchers estimate a disk velocity of 10 8 cm/sec is needed to ignite the fusion target [2]. To be able to study the IFI concept using lasers different from that at ILE, appropriate non-dimensionalization of the flow should be done. Analysis of the rocket equation gives parameters needed for producing similar IFI results with different lasers. This analysis shows that a variety of laboratory-scale commercial lasers could produce results useful to full-scale ILE experiments

  6. Engineering development for a small-scale recirculator experiment

    International Nuclear Information System (INIS)

    Newton, M.A.; Deadrick, F.J.; Hanks, R.L.; Hawkins, S.A.; Holm, K.A.; Kirbie, H.C.; Karpenko, V.P.; Nattrass, L.A.; Longinotti, D.B.

    1995-01-01

    Lawrence Livermore National Laboratory (LLNL) is evaluating the physics and technology of recirculating induction accelerators for heavy-ion inertial-fusion drivers. As part of this evaluation, the authors are building a small-scale recirculator to demonstrate the concept and to use as a test bed for the development of recirculator technologies. System designs have been completed and components are presently being designed and developed for the small-scale recirculator. This paper discusses results of the design and development activities that are presently being conducted to implement the small-scale recirculator experiments. An, overview of the system design is presented along with a discussion of the implications of this design on the mechanical and electrical hardware. The paper focuses primarily on discussions of the development and design of the half-lattice period hardware and the advanced solid-state modulator

  7. Sizewell 'B' PWR reference design

    International Nuclear Information System (INIS)

    1982-04-01

    The reference design for a PWR power station to be constructed as Sizewell 'B' is presented in 3 volumes containing 14 chapters and in a volume of drawings. The report describes the proposed design and provides the basis upon which the safety case and the Pre-Construction Safety Report have been prepared. The station is based on a 3425MWt Westinghouse PWR providing steam to two turbine generators each of 600 MW. The layout and many of the systems are based on the SNUPPS design for Callaway which has been chosen as the US reference plant for the project. (U.K.)

  8. Blind-blind prediction by RELAP5/MOD1 for a 0.1% very small cold-leg break experiment at ROSA-IV large-scale test facility

    International Nuclear Information System (INIS)

    Koizumi, Y.; Kumamaru, H.; Kukita, Y.; Kawaji, M.; Osakabe, M.; Schultz, R.R.; Tanaka, M.; Tasaka, K.

    1986-01-01

    The large-scale test facility (LSTF) of the Rig of Safety Assessment No. 4 (ROSA-IV) program is a volumetrically scaled (1/48) pressurized water reactor (PWR) system with an electrically heated core used for integral simulation of small break loss-of-coolant accidents (LOCAs) and operational transients. The 0.1% very small cold-leg break experiment was conducted as the first integral experiment at the LSTF. The test provided a good opportunity to truly assess the state-of-the-art predictability of the safety analysis code RELAP5/MODI CY18 through a blind-blind prediction of the experiment since there was no prior experience in analyzing the experimental data with the code; furthermore, detailed operational characteristics of LSTF were not yet known. The LOCA transient was mitigated by high-pressure charging pump injection to the primary system and bleed and feed operation of the secondary system. The simulated reactor system was safely placed in hot standby condition by engineered safety features similar to those on a PWR. Natural circulation flow was established to effectively remove the decay heat generated in the core. No cladding surface temperature excursion was observed. The RELAP5 code showed good capability to predict thermal-hydraulic phenomena during the very small break LOCA transient. Although all the information needed for the analysis by the RELAP5 code was obtained solely from the engineering drawings for fabrication and the operational specifications, the code predicted key phenomena satisfactorily

  9. Micro-scale hydrological field experiments in Romania

    Directory of Open Access Journals (Sweden)

    Minea Gabriel

    2016-02-01

    Full Text Available The paper (communication presents an overview of hydrologic field experiments at micro-scale in Romania. In order to experimentally investigate micro (plot-scale hydrological impact of soil erosion, the National Institute of Hydrology and Water Management founded Voineşti Experimental Basin (VES in 1964 and the Aldeni Experimental Basins (AEB in 1984. AEB and VES are located in the Curvature Subcarpathians. Experimental plots are organized in a double systems and have an area of 80 m2 (runoff plots at AEB and 300 m2 (water balance plots at VES. Land use of plot: first plot ”grass-land” is covered with perennial grass and second plot (control consists in ”bare soil”. Over the latter one, the soil is hoeing, which results in a greater development of infiltration than in the first plot. Experimental investigations at micro-scale are aimed towards determining the parameters of the water balance equation, during natural and artificial rainfalls, researching of flows and soil erosion processes on experimental plots, extrapolating relations involving runoff coefficients from a small scale to medium scale. Nowadays, the latest evolutions in data acquisition and transmission equipment are represented by sensors (such as: sensors to determinate the soil moisture content. Exploitation and dissemination of hydrologic data is accomplished by research themes/projects, year-books of basic data and papers.

  10. Field-scale colloid migration experiments in a granite fracture

    International Nuclear Information System (INIS)

    Vilks, P.; Frost, L.H.; Bachinski, D.B.

    1997-01-01

    An understanding of particle migration in fractured rock, required to assess the potential for colloid-facilitated transport of radionuclides, can best be evaluated when the results of laboratory experiments are demonstrated in the field. Field-scale migration experiments with silica colloids were carried out at AECL's Underground Research Laboratory (URL), located in southern Manitoba, to develop the methodology for large-scale migration experiments and to determine whether colloid transport is possible over distances up to 17 m. In addition, these experiments were designed to evaluate the effects of flow rate and flow path geometry, and to determine whether colloid tracers could be used to provide additional information on subsurface transport to that provided by conservative tracers alone. The colloid migration studies were carried out as part of AECL's Transport Properties in Highly Fractured Rock Experiment, the objective of which was to develop and demonstrate methods for evaluating the solute transport characteristics of zones of highly fractured rock. The experiments were carried out within fracture zone 2 as two-well recirculating, two-well non-recirculating, and convergent flow tests, using injection rates of 5 and 101 min -1 . Silica colloids with a 20 nm size were used because they are potentially mobile due to their stability, small size and negative surface charge. The shapes of elution profiles for colloids and conservative tracers were similar, demonstrating that colloids can migrate over distances of 17 m. The local region of drawdown towards the URL shaft affected colloid migration and, to a lesser extent, conservative tracer migration within the flow field established by the two-well tracer tests. These results indicate that stable colloids, with sizes as small as 20 nm, have different migration properties from dissolved conservative tracers. (author)

  11. PWR AXIAL BURNUP PROFILE ANALYSIS

    International Nuclear Information System (INIS)

    J.M. Acaglione

    2003-01-01

    The purpose of this activity is to develop a representative ''limiting'' axial burnup profile for pressurized water reactors (PWRs), which would encompass the isotopic axial variations caused by different assembly irradiation histories, and produce conservative isotopics with respect to criticality. The effect that the low burnup regions near the ends of spent fuel have on system reactivity is termed the ''end-effect''. This calculation will quantify the end-effects associated with Pressurized Water Reactor (PWR) fuel assemblies emplaced in a hypothetical 21 PWR waste package. The scope of this calculation covers an initial enrichment range of 3.0 through 5.0 wt% U-235 and a burnup range of 10 through 50 GWd/MTU. This activity supports the validation of the process for ensuring conservative generation of spent fuel isotopics with respect to criticality safety applications, and the use of burnup credit for commercial spent nuclear fuel. The intended use of these results will be in the development of PWR waste package loading curves, and applications involving burnup credit. Limitations of this evaluation are that the limiting profiles are only confirmed for use with the B andW 15 x 15 fuel assembly design. However, this assembly design is considered bounding of all other typical commercial PWR fuel assembly designs. This calculation is subject to the Quality Assurance Requirements and Description (QARD) because this activity supports investigations of items or barriers on the Q-list (YMP 2001)

  12. Scaling studies and conceptual experiment designs for NGNP CFD assessment

    Energy Technology Data Exchange (ETDEWEB)

    D. M. McEligot; G. E. McCreery

    2004-11-01

    The objective of this report is to document scaling studies and conceptual designs for flow and heat transfer experiments intended to assess CFD codes and their turbulence models proposed for application to prismatic NGNP concepts. The general approach of the project is to develop new benchmark experiments for assessment in parallel with CFD and coupled CFD/systems code calculations for the same geometry. Two aspects of the complex flow in an NGNP are being addressed: (1) flow and thermal mixing in the lower plenum ("hot streaking" issue) and (2) turbulence and resulting temperature distributions in reactor cooling channels ("hot channel" issue). Current prismatic NGNP concepts are being examined to identify their proposed flow conditions and geometries over the range from normal operation to decay heat removal in a pressurized cooldown. Approximate analyses have been applied to determine key non-dimensional parameters and their magnitudes over this operating range. For normal operation, the flow in the coolant channels can be considered to be dominant turbulent forced convection with slight transverse property variation. In a pressurized cooldown (LOFA) simulation, the flow quickly becomes laminar with some possible buoyancy influences. The flow in the lower plenum can locally be considered to be a situation of multiple hot jets into a confined crossflow -- with obstructions. Flow is expected to be turbulent with momentumdominated turbulent jets entering; buoyancy influences are estimated to be negligible in normal full power operation. Experiments are needed for the combined features of the lower plenum flows. Missing from the typical jet experiments available are interactions with nearby circular posts and with vertical posts in the vicinity of vertical walls - with near stagnant surroundings at one extreme and significant crossflow at the other. Two types of heat transfer experiments are being considered. One addresses the "hot channel" problem, if necessary

  13. Monitoring of the energy scale in the KATRIN neutrino experiment

    CERN Document Server

    AUTHOR|(CDS)2083282

    The question of the absolute mass scale of neutrinos is of particular interest for particle physics, astrophysics, and cosmology. The KATRIN experiment (KArlsruhe TRItium Neutrino experiment) aims to address the effective electron antineutrino mass from the shape of the tritium $\\beta$-spectrum with an unprecedented sensitivity of 0.2 eV/c$^2$. One of the major systematic effects concerns the experimental energy scale, which has to be stable at the level of only a few parts in a million. For its calibration and monitoring the monoenergetic electrons emitted in the internal conversion of $\\gamma$-transition of the metastable isotope $^{83\\mathrm{m}}$Kr will be extensively applied. The aim of this thesis is to address the problem of KATRIN energy scale distortions and its monitoring in detail. The source of electrons based on $^{83\\mathrm{m}}$Kr embedded in a solid as well as the source based on gaseous $^{83\\mathrm{m}}$Kr are studied. Based on the experimental results an approach for the continuous stability m...

  14. Microfluidic Experiments Studying Pore Scale Interactions of Microbes and Geochemistry

    Science.gov (United States)

    Chen, M.; Kocar, B. D.

    2016-12-01

    Understanding how physical phenomena, chemical reactions, and microbial behavior interact at the pore-scale is crucial to understanding larger scale trends in groundwater chemistry. Recent studies illustrate the utility of microfluidic devices for illuminating pore-scale physical-biogeochemical processes and their control(s) on the cycling of iron, uranium, and other important elements 1-3. These experimental systems are ideal for examining geochemical reactions mediated by microbes, which include processes governed by complex biological phenomenon (e.g. biofilm formation, etc.)4. We present results of microfluidic experiments using a model metal reducing bacteria and varying pore geometries, exploring the limitations of the microorganisms' ability to access tight pore spaces, and examining coupled biogeochemical-physical controls on the cycling of redox sensitive metals. Experimental results will provide an enhanced understanding of coupled physical-biogeochemical processes transpiring at the pore-scale, and will constrain and compliment continuum models used to predict and describe the subsurface cycling of redox-sensitive elements5. 1. Vrionis, H. A. et al. Microbiological and geochemical heterogeneity in an in situ uranium bioremediation field site. Appl. Environ. Microbiol. 71, 6308-6318 (2005). 2. Pearce, C. I. et al. Pore-scale characterization of biogeochemical controls on iron and uranium speciation under flow conditions. Environ. Sci. Technol. 46, 7992-8000 (2012). 3. Zhang, C., Liu, C. & Shi, Z. Micromodel investigation of transport effect on the kinetics of reductive dissolution of hematite. Environ. Sci. Technol. 47, 4131-4139 (2013). 4. Ginn, T. R. et al. Processes in microbial transport in the natural subsurface. Adv. Water Resour. 25, 1017-1042 (2002). 5. Scheibe, T. D. et al. Coupling a genome-scale metabolic model with a reactive transport model to describe in situ uranium bioremediation. Microb. Biotechnol. 2, 274-286 (2009).

  15. Multi-scale modelling for HEDP experiments on Orion

    Science.gov (United States)

    Sircombe, N. J.; Ramsay, M. G.; Hughes, S. J.; Hoarty, D. J.

    2016-05-01

    The Orion laser at AWE couples high energy long-pulse lasers with high intensity short-pulses, allowing material to be compressed beyond solid density and heated isochorically. This experimental capability has been demonstrated as a platform for conducting High Energy Density Physics material properties experiments. A clear understanding of the physics in experiments at this scale, combined with a robust, flexible and predictive modelling capability, is an important step towards more complex experimental platforms and ICF schemes which rely on high power lasers to achieve ignition. These experiments present a significant modelling challenge, the system is characterised by hydrodynamic effects over nanoseconds, driven by long-pulse lasers or the pre-pulse of the petawatt beams, and fast electron generation, transport, and heating effects over picoseconds, driven by short-pulse high intensity lasers. We describe the approach taken at AWE; to integrate a number of codes which capture the detailed physics for each spatial and temporal scale. Simulations of the heating of buried aluminium microdot targets are discussed and we consider the role such tools can play in understanding the impact of changes to the laser parameters, such as frequency and pre-pulse, as well as understanding effects which are difficult to observe experimentally.

  16. Reduction experiment of iron scale by adding waste plastics.

    Science.gov (United States)

    Zhang, Chongmin; Chen, Shuwen; Miao, Xincheng; Yuan, Hao

    2009-01-01

    The special features of waste plastics in China are huge in total amount, various in type and dispersive in deposition. Therefore, it is necessary to try some new ways that are fit to Chinese situation for disposing waste plastics as metallurgical raw materials more effectively and flexibly. Owing to its high ferrous content and less impurity, the iron scale became ideal raw material to produce pure iron powder. One of the methods to produce pure iron powder is Hoganas Method, by which, after one or multistage of reduction steps, the iron scale can be reduced pure iron powder. However, combining utilization of waste plastics and iron powder production, a series of reduction experiments were arranged and investigated, which is hoped to take use of both thermal and chemical energy contained in waste plastics as well as to improve the reducing condition of iron scale, and hence to develop a new metallurgical way of disposing waste plastics. The results show that under these experimental conditions, the thermal-decomposition of water plastics can conduce to an increase of porosity in the reduction systems. Moreover, better thermodynamics and kinetics conditions for the reduction of scale can be reached. As a result, the reduction rate is increased.

  17. Scaling effects concerning the analysis of small break experiments

    International Nuclear Information System (INIS)

    Austregesilo Filho, H.

    1985-01-01

    Some scaling effects related to the experimental facilities as well as to the analytical models used for the design and safety analysis of nuclear power plants are discussed or the basis of phenomena expected to occur during small-break loss - of - coolant accidents. The results of isolated small-break experiments should not be directly extrapolated to the safety analysis of commercial reactors, due to the scaling distortions inherent to the test facilities. With respect to the analytical models used to simulate thermohydraulic processes in experimental facilities, their eventual dependence relative to the system dimension should be examined in order to assess their applicability to the safety analysis of commercial power plants. (Author) [pt

  18. Experimental study on secondary depressurization action for PWR vessel bottom small break LOCA with HPI failure and gas inflow (ROSA-V/LSTF test SB-PV-03)

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

    2005-06-01

    A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility (LSTF) of ROSA-V program to study effects of accident management (AM) measures on core cooling, which is important in case of high pressure injection (HPI) system failure during an SBLOCA at a pressurized water reactor (PWR). The LSTF is a full-height and 1/48 volume-scaled facility simulating 4-loop Westinghouse-type PWR (3423 MWt). The experiment, SB-PV-03, simulated a PWR vessel bottom SBLOCA with a rupture of ten instrument-tubes which is equivalent to 0.2% cold leg break. Total HPI failure, non-condensable gas inflow from accumulator injection system (AIS) and operator AM actions on steam generator (SG) secondary depressurization at a rate of -55 K/h and auxiliary feedwater (AFW) supply for 30 minutes were assumed as experiment conditions. It is clarified that the AM actions are effective on primary system depressurization until the end of AIS injection at 1.6 MPa, but thereafter become less effective due to inflow of the non-condensable gas, resulting in delay of low pressure injection (LPI) actuation and whole core heatup under continuous water discharge through the bottom break. The report describes these thermohydraulic phenomena related with transient primary coolant mass and AM actions in addition to estimation of non-condensable gas behavior which affected primary-to-secondary heat transfer. (author)

  19. An experimental study on effective depressurization actions for PWR vessel bottom small break LOCA with HPI failure and gas inflow (ROSA-V test SB-PV-04)

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

    2006-03-01

    A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility (LSTF) of ROSA-V program to study effects of accident management (AM) measures on core cooling, which are important in case of total failure of high pressure injection (HPI) system during an SBLOCA at a pressurized water reactor (PWR). The LSTF is a full-height and 1/48 volume-scaled facility simulating a 4-loop Westing-house-type PWR (3423 MWt). The experiment, SB-PV-04, simulated a PWR vessel bottom SBLOCA with a rupture of ten instrument-tubes which is equivalent to 0.2% cold leg break. It is clarified that AM actions with steam generator (SG) rapid depressurization by fully opening relief valves and auxiliary feedwater supply are effective to avoid core uncovery by actuating the low pressure injection (LPI) system though the primary depressurization is degraded by non-condensable gas inflow to the primary loops from the accumulator injection system. The effective core cooling was established by the rapid depressurization which contributed to preserve larger primary coolant mass than in the previous experiment (SB-PV-03) which was conducted with smaller primary cooling rate of -55 K/h as AM actions. (author)

  20. Dimensional analysis of small-scale steam explosion experiments

    International Nuclear Information System (INIS)

    Huh, K.; Corradini, M.L.

    1986-01-01

    Dimensional analysis applied to Nelson's small-scale steam explosion experiments to determine the qualitative effect of each relevant parameter for triggering a steam explosion. According to experimental results, the liquid entrapment model seems to be a consistent explanation for the steam explosion triggering mechanism. The three-dimensional oscillatory wave motion of the vapor/liquid interface is analyzed to determine the necessary conditions for local condensation and production of a coolant microjet to be entrapped in fuel. It is proposed that different contact modes between fuel and coolant may involve different initiation mechanisms of steam explosions

  1. Scaling up debris-flow experiments on a centrifuge

    Science.gov (United States)

    Hung, C.; Capart, H.; Crone, T. J.; Grinspum, E.; Hsu, L.; Kaufman, D.; Li, L.; Ling, H.; Reitz, M. D.; Smith, B.; Stark, C. P.

    2013-12-01

    Boundary forces generated by debris flows can be powerful enough to erode bedrock and cause considerable damage to infrastructure during runout. Formulation of an erosion-rate law for debris flows is therefore a high priority, and it makes sense to build such a law around laboratory experiments. However, running experiments big enough to generate realistic boundary forces is a logistical challenge to say the least [1]. One alternative is to run table-top simulations with unnaturally weak but fast-eroding pseudo-bedrock, another is to extrapolate from micro-erosion of natural substrates driven by unnaturally weak impacts; hybrid-scale experiments have also been conducted [2]. Here we take a different approach in which we scale up granular impact forces by running our experiments under enhanced gravity in a geotechnical centrifuge [3]. Using a 40cm-diameter rotating drum [2] spun at up to 100g, we generate debris flows with an effective depth of over several meters. By varying effective gravity from 1g to 100g we explore the scaling of granular flow forces and the consequent bed and wall erosion rates. The velocity and density structure of these granular flows is monitored using laser sheets, high-speed video, and particle tracking [4], and the progressive erosion of the boundary surfaces is measured by laser scanning. The force structures and their fluctuations within the granular mass and at the boundaries are explored with contact dynamics numerical simulations that mimic the lab experimental conditions [5]. In this presentation we summarize these results and discuss how they can contribute to the formulation of debris-flow erosion law. [1] Major, J. J. (1997), Journal of Geology 105: 345-366, doi:10.1086/515930 [2] Hsu, L. (2010), Ph.D. thesis, University of California, Berkeley [3] Brucks, A., et al (2007), Physical Review E 75, 032301, doi:10.1103/PhysRevE.75.032301 [4] Spinewine, B., et al (2011), Experiments in Fluids 50: 1507-1525, doi: 10.1007/s00348

  2. Explicating Experience: Development of a Valid Scale of Past Hazard Experience for Tornadoes.

    Science.gov (United States)

    Demuth, Julie L

    2018-03-23

    People's past experiences with a hazard theoretically influence how they approach future risks. Yet, past hazard experience has been conceptualized and measured in wide-ranging, often simplistic, ways, resulting in mixed findings about its relationship with risk perception. This study develops a scale of past hazard experiences, in the context of tornadoes, that is content and construct valid. A conceptual definition was developed, a set of items were created to measure one's most memorable and multiple tornado experiences, and the measures were evaluated through two surveys of the public who reside in tornado-prone areas. Four dimensions emerged of people's most memorable experience, reflecting their awareness of the tornado risk that day, their personalization of the risk, the intrusive impacts on them personally, and impacts experienced vicariously through others. Two dimensions emerged of people's multiple experiences, reflecting common types of communication received and negative emotional responses. These six dimensions are novel in that they capture people's experience across the timeline of a hazard as well as intangible experiences that are both direct and indirect. The six tornado experience dimensions were correlated with tornado risk perceptions measured as cognitive-affective and as perceived probability of consequences. The varied experience-risk perception results suggest that it is important to understand the nuances of these concepts and their relationships. This study provides a foundation for future work to continue explicating past hazard experience, across different risk contexts, and for understanding its effect on risk assessment and responses. © 2018 Society for Risk Analysis.

  3. Crack growth rate of PWR piping

    International Nuclear Information System (INIS)

    Bethmont, M.; Doyen, J.J.; Lebey, J.

    1979-01-01

    The Aquitaine 1 program, carried out jointly by FRAMATOME and the CEA is intended to improve knowledge about cracking mechanisms in AISI 316 L austenitic stainless steel under conditions similar to those of the PWR environment (irradiation excluded). Experiments of fatigue crack growth are performed on piping elements, scale 1/4 of primary pipings, by means of internal hydraulic cyclic pressure. Interpretation of results requires a knowledge of the stress intensity factor Ksub(I) at the front of the crack. Results of a series of calculations of Ksub(I) obtained by different methods for defects of finite and infinite length (three dimensional calculations) are given in the paper. The following have been used: calculations by finite elements, calculations by weight function. Notches are machined on the test pipes, which are subjected to internal hydraulic pressure cycles, under cold conditions, to initiate a crack at the tip of the notch. They are then cycled at a frequency of 4 cycles/hour on on water demineralised loop at a temperature of 280 0 C, the pressure varying at each cycle between approximately 160 bars and 3 bars. After each test, a specimen containing the defect is taken from the pipe for micrographic analysis. For the first test the length of the longitudinal external defect is assumed infinite. The number of cycles carried out is 5880 cycles. Two defects are machined in the tube for the second test. The number of cycles carried out is N = 440. The tests are performed under hot conditions (T = 280 0 C). For the third test two defects are analysed under cold and hot conditions. The number of cycles carried out for the external defect is 7000 when hot and 90000 when cold. The number of cycles for the internal defect is 1650 when hot and 68000 when cold. In order to interpret the results, the data da/dN are plotted on a diagram versus ΔK. Comparisons are made between these results and the curves from laboratory tests

  4. Parameterized representation of macroscopic cross section for PWR reactor

    International Nuclear Information System (INIS)

    Fiel, João Cláudio Batista; Carvalho da Silva, Fernando; Senra Martinez, Aquilino; Leal, Luiz C.

    2015-01-01

    Highlights: • This work describes a parameterized representation of the homogenized macroscopic cross section for PWR reactor. • Parameterization enables a quick determination of problem-dependent cross-sections to be used in few group calculations. • This work allows generating group cross-section data to perform PWR core calculations without computer code calculations. - Abstract: The purpose of this work is to describe, by means of Chebyshev polynomials, a parameterized representation of the homogenized macroscopic cross section for PWR fuel element as a function of soluble boron concentration, moderator temperature, fuel temperature, moderator density and 235 92 U enrichment. The cross-section data analyzed are fission, scattering, total, transport, absorption and capture. The parameterization enables a quick and easy determination of problem-dependent cross-sections to be used in few group calculations. The methodology presented in this paper will allow generation of group cross-section data from stored polynomials to perform PWR core calculations without the need to generate them based on computer code calculations using standard steps. The results obtained by the proposed methodology when compared with results from the SCALE code calculations show very good agreement

  5. Large scale FCI experiments in subassembly geometry. Test facility and model experiments

    International Nuclear Information System (INIS)

    Beutel, H.; Gast, K.

    A program is outlined for the study of fuel/coolant interaction under SNR conditions. The program consists of a) under water explosion experiments with full size models of the SNR-core, in which the fuel/coolant system is simulated by a pyrotechnic mixture. b) large scale fuel/coolant interaction experiments with up to 5kg of molten UO 2 interacting with liquid sodium at 300 deg C to 600 deg C in a highly instrumented test facility simulating an SNR subassembly. The experimental results will be compared to theoretical models under development at Karlsruhe. Commencement of the experiments is expected for the beginning of 1975

  6. PWR-to-PWR fuel cycle model using dry process

    International Nuclear Information System (INIS)

    Iqbal, M.; Jeong, Chang Joon; Rho, Gyu Hong

    2002-03-01

    PWR-to-PWR fuel cycle model has been developed to recycle the spent fuel using the dry fabrication process. Two types of fuels were considered; first fuel was based on low initial enrichment with low discharge burnup and second one was based on more initial enrichment with high discharge burnup in PWR. For recycling calculations, the HELIOS code was used, in which all of the available fission products were considered. The decay of 10 years was applied for reuse of the spent fuel. Sensitivity analysis for the fresh feed material enrichment has also been carried out. If enrichment of the mixing material is increased the saving of uranium reserves would be decreased. The uranium saving of low burned fuel increased from 4.2% to 7.4% in fifth recycling step for 5 wt% to 19.00wt% mixing material enrichment. While for high burned fuel, there was no uranium saving, which implies that higher uranium enrichment required than 5 wt%. For mixing of 15 wt% enriched fuel, the required mixing is about 21.0% and 37.0% of total fuel volume for low and high burned fuel, respectively. With multiple recycling, reductions in waste for low and high burned fuel became 80% and 60%, for first recycling, respectively. In this way, waste can be reduced more and the cost of the waste disposal reduction can provide the economic balance

  7. Gas explosion characterization, wave propagation (small-scale experiments)

    International Nuclear Information System (INIS)

    Larsen, G.C.

    1985-01-01

    A number of experiments have been performed with blast waves arising from the ignition of homogeneous and well defined mixtures of methane, oxygen and nitrogen, contained within spherical balloons with controlled initial dimensions. In the initial small scale experiments pressure characteristics, ground reflection phenomena and pressure distribution on box like obstacles were studied. Both configurations with one box and two closely spaced boxes have been considered, and a wave-wave interaction phenomenom was observed in the case of closely spaced obstacles. Main emphasis has been placed on the half scale field experiments. In these, the maximum flame speed has been of the order of 100 m/s, resulting in positive peak pressures of 50-100.10 2 Pa in 5 - 10 m distance from the source. The explosion process was found to be reasonable symmetric. The attenuation of the blast wave due to vegetation and the influence of obstacles as banks, walls and houses on the pressure field have been investigated. The presence of the bank and the house was felt in a zone with a length corresponding to a typical dimension of the obstacles, whereas the overall pressure field is shown to be unaffected by the type of obstacles and vegetation investigated. For the wall and house, reflection factors have been established, and some variation over the surface has been measured. The scatter of the pressure measurements is estimated for stable, neutral and unstable atmospheric conditions, and an attempt to determine the ground reflection factor has been performed. Finally the accelerations of a house exposed to the blast wave have been examined

  8. Full-scale fire experiments on vertical horizontal cable trays

    International Nuclear Information System (INIS)

    Mangs, J.; Keski-Rahkonen, O.

    1997-10-01

    Two full-scale fire experiments on PVC cables used in nuclear power plants were carried out, one with cables in vertical position and one with cables in horizontal position. The vertical cable bundle, 3 m high, 300 mm wide and 30 mm thick, was attached to a steel cable ladder. The vertical bundle experiment was carried out in nearly free space with three walls near the cable ladder guiding air flow in order to stabilise flames. The horizontal cable experiment was carried out in a small room with five cable bundles attached to steel cable ladders. Three of the 2 m long cable bundles were located in an array, equally spaced above each other near one long side of the room and two correspondingly near the opposite long side. The vertical cable bundle was ignited with a small propane gas burner beneath the lower edge of the bundle. The horizontal cable bundles were ignited with a small propane burner beneath the lowest bundle in an array of three bundles. Rate of heat release by means of oxygen consumption calorimetry, mass change, CO 2 , CO and smoke production rate and gas, wall and cable surface temperatures were measured as a function of time, as well as time to sprinkler operation and failure of test voltage in cables. Additionally, the minimum rate of heat release needed to ignite the bundle was determined. This paper concentrates on describing and recording the experimental set-up and the data obtained. (orig.)

  9. Scaled electron experiments at the University of Maryland

    International Nuclear Information System (INIS)

    Haber, I.; Bai, G.; Bernal, S.; Beaudoin, B.; Feldman, D.; Fiorito, R.B.; Godlove, T.F.; Kishek, R.A.; O'Shea, P.G.; Quinn, B.; Papadopoulos, C.; Reiser, M.; Rodgers, J.; Stratakis, D.; Sutter, D.; Thangaraj, J.C.T.; Tian, K.; Walter, M.; Wu, C.

    2007-01-01

    The University of Maryland Electron Ring (UMER) and the Long Solenoid Experiment (LSE) are two electron machines that were designed explicitly to study the physics of space-charge-dominated beams. The operating parameters of these machines can be varied by choice of apertures and gun operating conditions to access a wide range of parameters that reproduce, on a scaled basis, the full nonlinear time-dependent physics that is expected in much costlier ion systems. Early operation of these machines has demonstrated the importance of the details of beam initial conditions in determining the downstream evolution. These machines have also been a convenient tested for benchmarking simulation codes such as WARP, and for development of several novel diagnostic techniques. We present our recent experience with multi-turn operation as well as recent longitudinal and transverse physics experiments and comparisons to simulation results. Development of novel diagnostic techniques such as time-dependent imaging using optical transition radiation and tomographic beam reconstruction are also described

  10. PACTEL and PWR PACTEL Test Facilities for Versatile LWR Applications

    Directory of Open Access Journals (Sweden)

    Virpi Kouhia

    2012-01-01

    Full Text Available This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVER-440-type reactors. The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP-33. PACTEL has been upgraded and modified on a case-by-case basis. The latest facility configuration, the PWR PACTEL facility, was constructed for research activities associated with the EPR-type reactor. A significant design basis is to utilize certain parts of PACTEL, and at the same time, to focus on a proper construction of two new loops and vertical steam generators with an extensive instrumentation. The PWR PACTEL benchmark exercise was launched in 2010 with a small break loss-of-coolant accident test as the chosen transient. Both facilities, PACTEL and PWR PACTEL, are maintained fully operational side by side.

  11. PACTEL and PWR PACTEL Test Facilities for Versatile LWR Applications

    International Nuclear Information System (INIS)

    Virpi Kouhia, V.; Purhonen, H.; Riikonen, V.; Puustinen, M.; Kyrki-Rajamaki, R.; Vihavainen, J.

    2012-01-01

    This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVER-440-type reactors. The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP-33. PACTEL has been upgraded and modified on a case-by-case basis. The latest facility configuration, the PWR PACTEL facility, was constructed for research activities associated with the EPR-type reactor. A significant design basis is to utilize certain parts of PACTEL, and at the same time, to focus on a proper construction of two new loops and vertical steam generators with an extensive instrumentation. The PWR PACTEL benchmark exercise was launched in 2010 with a small break loss-of-coolant accident test as the chosen transient. Both facilities, PACTEL and PWR PACTEL, are maintained fully operational side by side.

  12. Diffusion Experiments in Opalinus Clay: Laboratory, Large-Scale Diffusion Experiments and Microscale Analysis by RBS.

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Gutierrez, M.; Alonso de los Rios, U.; Missana, T.; Cormenzana, J.L.; Mingarro, M.; Morejon, J.; Gil, P.

    2008-08-06

    The Opalinus Clay (OPA) formation in the Zurcher Weiland (Switzerland) is a potential host rock for a repository for high-level radioactive waste. Samples collected in the Mont Terri Underground Rock Laboratory (URL), where the OPA formation is located at a depth between -200 and -300 m below the surface, were used to study the radionuclide diffusion in clay materials. Classical laboratory essays and a novel experimental set-up for large-scale diffusion experiments were performed together to a novel application of the nuclear ion beam technique Rutherford Backscattering Spectrometry (RBS), to understand the transport properties of the OPA and to enhance the methodologies used for in situ diffusion experiments. Through-Diffusion and In-Diffusion conventional laboratory diffusion experiments were carried out with HTO, 36{sup C}l-, I-, 22{sup N}a, 75{sup S}e, 85{sup S}r, 233{sup U}, 137{sup C}s, 60{sup C}o and 152{sup E}u. Large-scale diffusion experiments were performed with HTO, 36{sup C}l, and 85{sup S}r, and new experiments with 60{sup C}o, 137{sup C}s and 152{sup E}u are ongoing. Diffusion experiments with RBS technique were done with Sr, Re, U and Eu. (Author) 38 refs.

  13. Progress in heavy ion driven inertial fusion energy: From scaled experiments to the integrated research experiment

    International Nuclear Information System (INIS)

    Barnard, J.J.; Ahle, L.E.; Baca, D.; Bangerter, R.O.; Bieniosek, F.M.; Celata, C.M.; Chacon-Golcher, E.; Davidson, R.C.; Faltens, A.; Friedman, A.; Franks, R.M.; Grote, D.P.; Haber, I.; Henestroza, E.; Hoon, M.J.L. de; Kaganovich, I.; Karpenko, V.P.; Kishek, R.A.; Kwan, J.W.; Lee, E.P.; Logan, B.G.; Lund, S.M.; Meier, W.R.; Molvik, A.W.; Olson, C.; Prost, L.R.; Qin, H.; Rose, D.; Sabbi, G.-L.; Sangster, T.C.; Seidl, P.A.; Sharp, W.M.; Shuman, D.; Vay, J.-L.; Waldron, W.L.; Welch, D.; Yu, S.S.

    2001-01-01

    The promise of inertial fusion energy driven by heavy ion beams requires the development of accelerators that produce ion currents (∼100's Amperes/beam) and ion energies (∼1-10 GeV) that have not been achieved simultaneously in any existing accelerator. The high currents imply high generalized perveances, large tune depressions, and high space charge potentials of the beam center relative to the beam pipe. Many of the scientific issues associated with ion beams of high perveance and large tune depression have been addressed over the last two decades on scaled experiments at Lawrence Berkeley and Lawrence Livermore National Laboratories, the University of Maryland, and elsewhere. The additional requirement of high space charge potential (or equivalently high line charge density) gives rise to effects (particularly the role of electrons in beam transport) which must be understood before proceeding to a large scale accelerator. The first phase of a new series of experiments in Heavy Ion Fusion Virtual National Laboratory (HIF VNL), the High Current Experiments (HCX), is now being constructed at LBNL. The mission of the HCX will be to transport beams with driver line charge density so as to investigate the physics of this regime, including constraints on the maximum radial filling factor of the beam through the pipe. This factor is important for determining both cost and reliability of a driver scale accelerator. The HCX will provide data for design of the next steps in the sequence of experiments leading to an inertial fusion energy power plant. The focus of the program after the HCX will be on integration of all of the manipulations required for a driver. In the near term following HCX, an Integrated Beam Experiment (IBX) of the same general scale as the HCX is envisioned. The step which bridges the gap between the IBX and an engineering test facility for fusion has been designated the Integrated Research Experiment (IRE). The IRE (like the IBX) will provide an

  14. PWR system reliability improvement activities

    International Nuclear Information System (INIS)

    Yoshikawa, Yuichiro

    1985-01-01

    In Japan lacking in energy resources, it is our basic energy policy to accelerate the development program of nuclear power, thereby reducing our dependence. As referred to in the foregoing, every effort has been exerted on our part to improve the PWR system reliability by dint of the so-called 'HOMEMADE' TQC activities, which is our brain-child as a result of applying to the energy industry the quality control philosophy developed in the field of manufacturing industry

  15. Seismic proving test of PWR reactor containment vessel

    International Nuclear Information System (INIS)

    Akiyama, H.; Yoshikawa, T.; Tokumaru, Y.

    1987-01-01

    The seismic reliability proving tests of nuclear power plant facilities are carried out by Nuclear Power Engineering Test Center (NUPEC), using the large-scale, high-performance vibration of Tadotsu Engineering Laboratory, and sponsored by the Ministry of International Trade and Industry (MITI). In 1982, the seismic reliability proving test of PWR containment vessel started using the test component of reduced scale 1/3.7 and the test component proved to have structural soundness against earthquakes. Subsequently, the detailed analysis and evaluation of these test results were carried out, and the analysis methods for evaluating strength against earthquakes were established. Whereupon, the seismic analysis and evaluation on the actual containment vessel were performed by these analysis methods, and the safety and reliability of the PWR reactor containment vessel were confirmed

  16. Report on the PWR-radiation protection/ALARA Committee

    Energy Technology Data Exchange (ETDEWEB)

    Malone, D.J. [Consumers Power Co., Covert, MI (United States)

    1995-03-01

    In 1992, representatives from several utilities with operational Pressurized Water Reactors (PWR) formed the PWR-Radiation Protection/ALARA Committee. The mission of the Committee is to facilitate open communications between member utilities relative to radiation protection and ALARA issues such that cost effective dose reduction and radiation protection measures may be instituted. While industry deregulation appears inevitable and inter-utility competition is on the rise, Committee members are fully committed to sharing both positive and negative experiences for the benefit of the health and safety of the radiation worker. Committee meetings provide current operational experiences through members providing Plant status reports, and information relative to programmatic improvements through member presentations and topic specific workshops. The most recent Committee workshop was facilitated to provide members with defined experiences that provide cost effective ALARA performance.

  17. Simple biogas desulfurization by microaeration - Full scale experience.

    Science.gov (United States)

    Jeníček, P; Horejš, J; Pokorná-Krayzelová, L; Bindzar, J; Bartáček, J

    2017-08-01

    Hydrogen sulfide in biogas is common problem during anaerobic treatment of wastewater with high sulfate concentration (breweries, distilleries, etc.) and needs to be removed before biogas utilization. Physico-chemical desulfurization methods are energetically demanding and expensive compare to biochemical methods. Microaeration, i.e. dosing of small amount of air, is suitable and cost effective biochemical method of sulfide oxidation to elemental sulfur. It has been widely used in biogas plants, but its application in anaerobic reactors for wastewater treatment has been rarely studied or tested. The lack of full-scale experience with microaeration in wastewater treatment plants has been overcome by evaluating the results of seven microaerobic digesters in central Europe. The desulfurization efficiency has been more than 90% in most of the cases. Moreover, microaeration improved the degradability of COD and volatile suspended solids. Copyright © 2017 Elsevier Ltd. All rights reserved.

  18. PWR secondary water chemistry guidelines: Revision 3

    International Nuclear Information System (INIS)

    Lurie, S.; Bucci, G.; Johnson, L.; King, M.; Lamanna, L.; Morgan, E.; Bates, J.; Burns, R.; Eaker, R.; Ward, G.; Linnenbom, V.; Millet, P.; Paine, J.P.; Wood, C.J.; Gatten, T.; Meatheany, D.; Seager, J.; Thompson, R.; Brobst, G.; Connor, W.; Lewis, G.; Shirmer, R.; Gillen, J.; Kerns, M.; Jones, V.; Lappegaard, S.; Sawochka, S.; Smith, F.; Spires, D.; Pagan, S.; Gardner, J.; Polidoroff, T.; Lambert, S.; Dahl, B.; Hundley, F.; Miller, B.; Andersson, P.; Briden, D.; Fellers, B.; Harvey, S.; Polchow, J.; Rootham, M.; Fredrichs, T.; Flint, W.

    1993-05-01

    An effective, state-of-the art secondary water chemistry control program is essential to maximize the availability and operating life of major PWR components. Furthermore, the costs related to maintaining secondary water chemistry will likely be less than the repair or replacement of steam generators or large turbine rotors, with resulting outages taken into account. The revised PWR secondary water chemistry guidelines in this report represent the latest field and laboratory data on steam generator corrosion phenomena. This document supersedes Interim PWR Secondary Water Chemistry Recommendations for IGA/SCC Control (EPRI report TR-101230) as well as PWR Secondary Water Chemistry Guidelines--Revision 2 (NP-6239)

  19. On the consistency of scale among experiments, theory, and simulation

    Science.gov (United States)

    McClure, James E.; Dye, Amanda L.; Miller, Cass T.; Gray, William G.

    2017-02-01

    As a tool for addressing problems of scale, we consider an evolving approach known as the thermodynamically constrained averaging theory (TCAT), which has broad applicability to hydrology. We consider the case of modeling of two-fluid-phase flow in porous media, and we focus on issues of scale as they relate to various measures of pressure, capillary pressure, and state equations needed to produce solvable models. We apply TCAT to perform physics-based data assimilation to understand how the internal behavior influences the macroscale state of two-fluid porous medium systems. A microfluidic experimental method and a lattice Boltzmann simulation method are used to examine a key deficiency associated with standard approaches. In a hydrologic process such as evaporation, the water content will ultimately be reduced below the irreducible wetting-phase saturation determined from experiments. This is problematic since the derived closure relationships cannot predict the associated capillary pressures for these states. We demonstrate that the irreducible wetting-phase saturation is an artifact of the experimental design, caused by the fact that the boundary pressure difference does not approximate the true capillary pressure. Using averaging methods, we compute the true capillary pressure for fluid configurations at and below the irreducible wetting-phase saturation. Results of our analysis include a state function for the capillary pressure expressed as a function of fluid saturation and interfacial area.

  20. Experimental modelling of core debris dispersion from the vault under a PWR pressure vessel. Pt. 2

    International Nuclear Information System (INIS)

    Rose, P.W.

    1987-12-01

    In previous experiments, done on a 1/25 scale model in Perspex of the vault under a PWR pressure vessel, the instrument tubes support structure built into the vault was not included. It consists of a number of grids made up of fairly massive steel girders. These have now been added to the model and experiments performed using water to simulate molten core debris assumed to have fallen on to the vault floor and high-pressure air to simulate the discharge of steam or gas from the assumed breach at the bottom of the pressure vessel. The results show that the tubes support structure considerably reduces the carry-over of liquid via the vault access shafts. (author)

  1. Full-scale and time-scale heating experiments at Stripa: preliminary results. Technical project report No. 11

    International Nuclear Information System (INIS)

    Cook, N.G.W.; Hood, M.

    1978-12-01

    Two full-scale heating experiments and a time-scale heating experiment have recently been started in granite 340 meters below surface. The purpose of the full-scale heating experiments is to assess the near-field effects of thermal loading for the design of an underground repository of nuclear wastes. That of the time-scale heating experiments is to obtain field data of the interaction between heaters and its effect on the rock mass during a period of about two years, which corresponds to about twenty years of full-scale operation. Geological features of the rock around each experiment have been mapped carefully, and temperatures, stresses and displacements induced in the rock by heating have been calculated in advance of the experiments. Some 800 different measurements are recorded at frequent intervals by a computer system situated underground. These data can be compared at any time with predictions made earlier on video display units underground

  2. Virtual neutron scattering experiments - Training and preparing students for large-scale facility experiments

    Directory of Open Access Journals (Sweden)

    Julie Hougaard Overgaard

    2016-11-01

    Full Text Available Dansk Vi beskriver, hvordan virtuelle eksperimenter kan udnyttes i et læringsdesign ved at forberede de studerende til hands-on-eksperimenter ved storskalafaciliteter. Vi illustrerer designet ved at vise, hvordan virtuelle eksperimenter bruges på Niels Bohr Institutets kandidatkursus om neutronspredning. I den sidste uge af kurset, rejser studerende til et storskala neutronspredningsfacilitet for at udføre neutronspredningseksperimenter. Vi bruger studerendes udsagn om deres oplevelser til at argumentere for, at arbejdet med virtuelle experimenter forbereder de studerende til at engagere sig mere frugtbart med eksperimenter ved at lade dem fokusere på fysikken og relevante data i stedet for instrumenternes funktion. Vi hævder, at det er, fordi de kan overføre deres erfaringer med virtuelle eksperimenter til rigtige eksperimenter. Vi finder dog, at læring stadig er situeret i den forstand, at kun kendskab til bestemte eksperimenter overføres. Vi afslutter med at diskutere de muligheder, som virtuelle eksperimenter giver. English We describe how virtual experiments can be utilized in a learning design that prepares students for hands-on experiments at large-scale facilities. We illustrate the design by showing how virtual experiments are used at the Niels Bohr Institute in a master level course on neutron scattering. In the last week of the course, students travel to a large-scale neutron scattering facility to perform real neutron scattering experiments. Through student interviews and survey answers, we argue, that the virtual training prepares the students to engage more fruitfully with experiments by letting them focus on physics and data rather than the overwhelming instrumentation. We argue that this is because they can transfer their virtual experimental experience to the real-life situation. However, we also find that learning is still situated in the sense that only knowledge of particular experiments is transferred. We proceed to

  3. Stochastic optimization of loading pattern for PWR

    International Nuclear Information System (INIS)

    Smuc, T.; Pevec, D.

    1994-01-01

    The application of stochastic optimization methods in solving in-core fuel management problems is restrained by the need for a large number of proposed solutions loading patterns, if a high quality final solution is wanted. Proposed loading patterns have to be evaluated by core neutronics simulator, which can impose unrealistic computer time requirements. A new loading pattern optimization code Monte Carlo Loading Pattern Search has been developed by coupling the simulated annealing optimization algorithm with a fast one-and-a-half dimensional core depletion simulator. The structure of the optimization method provides more efficient performance and allows the user to empty precious experience in the search process, thus reducing the search space size. Hereinafter, we discuss the characteristics of the method and illustrate them on the results obtained by solving the PWR reload problem. (authors). 7 refs., 1 tab., 1 fig

  4. Evaluation of tight-pitch PWR cores

    International Nuclear Information System (INIS)

    Correa, F.; Driscoll, M.J.; Lanning, D.D.

    1979-08-01

    The impact of tight pinch cores on the consumption of natural uranium ore has been evaluated for two systems of coupled PWR's namely one particular type of thorium system - 235 U/UO 2 : Pu/ThO 2 : 233 U/ThO 2 - and the conventional recycle-mode uranium system - 235 U/UO 2 : Pu/UO 2 . The basic parameter varied was the fuel-to-moderator volume ratio (F/M) of the (uniform) lattice for the last core in each sequence. Although methods and data verification in the range of present interest, 0.5 (current lattices) 1.0, the EPRI-LEOPARD and LASER programs used for the thorium and uranium calculations, respectively, were successfully benchmarked against several of the more pertinent experiments

  5. Integrated laboratory scale demonstration experiment of the hybrid sulphur cycle and preliminary scale-up

    International Nuclear Information System (INIS)

    Leybros, J.; Rivalier, P.; Saturnin, A.; Charton, S.

    2010-01-01

    The hybrid sulphur cycle is today one of the most promising processes to produce hydrogen on a massive scale within the scope of high temperature nuclear reactors development. Thus, the Fuel Cycle Technology Department at CEA Marcoule is involved in studying the hybrid sulphur process from a technical and economical performance standpoint. Based on mass and energy balance calculations, a ProsimPlus TM flow sheet and a commercial plant design were prepared. This work includes a study on sizing of the main equipment. The capital cost has been estimated using the major characteristics of main equipment based upon formulae and charts published in literature. A specific approach has been developed for electrolysers. Operational costs are also proposed for a plant producing 1000 mol/s H 2 . Bench scale and pilot experiments must focus on the electrochemical step due to limited experimental data. Thus, a pilot plant with a hydrogen capacity of 100 NL/h was built with the aim of acquiring technical and technological data for electrolysis. This pilot plant was designed to cover a wide range of operating conditions: sulphuric acid concentrations up to 60 wt.%, temperatures up to 100 deg. C and pressures up to 10 bar. New materials and structures recently developed for fuel cells, which are expected to yield significant performance improvements when applied to classical electrochemical processes, will be tested. All experiments will be coupled with phenomenological simulation tools developed jointly with the experimental programme. (authors)

  6. Development of nuclear thermal hydraulic verification test and evaluation technology - Development of fundamental technique for experiment of natural circulation phenomena in PWR systems

    Energy Technology Data Exchange (ETDEWEB)

    Park, Goon Cherl; Lee, Tae Ho; Kim, Moon Oh; Kim, Hak Joon [Seoul National University, Seoul (Korea)

    2000-04-01

    The dimensional analysis applied two-fluid model of CFX-4,2 were performed. For verification of analysis results, experimental measurement data of two-phase flow parameters in subcooled boiling flow were produced for vertical(0 deg) and inclination (60 deg). And through comparison analysis and experiments the application possibility of various two -phase flow models and the analysis ability of code were evaluated. Measurement technique of bubble velocity in two-phase flow using backscattering standard LDV was investigated from slug to bubbly flow regime. The range of velocity measured is from 0.2 to 1.5 m/s and that of bubble size is from 2 to 20 mm. For local temperature of boiling flow measurement, microthermocouple were manufactured and local liquid and vapor temperatures were measured in pool boiling and boiling flow. 66 refs., 74 figs., 4 tabs. (Author)

  7. Advanced accumulator for PWR

    International Nuclear Information System (INIS)

    Ichimura, Taiki; Chikahata, Hideyuki

    1997-01-01

    Advanced accumulators have been incorporated into the APWR design in order to simplify the safety system configuration and to improve reliability. The advanced accumulators refill the reactor vessel with a large discharge flow rate in a large LOCA, then switch to a small flow rate to continue safety injection for core reflooding. The functions of the conventional accumulator and the low head safety injection pump are integrated into this advanced accumulator. Injection performance tests simulating LOCA conditions and visualization tests for new designs have been carried out. This paper describes the APWR ECCS configuration, the advanced accumulator design and some of the injection performance and visualization test results. It was verified that the flow resistance of the advanced accumulator is independent of the model scale. The similarity law and performance data of the advanced accumulator for applying APWR was established. (author)

  8. Failure of PWR-RHRS under cold shutdown conditions: Experimental results from the PKL test facility

    International Nuclear Information System (INIS)

    Mandl, R.M.; Umminger, K.J.; Logt, J.V.D.

    1991-01-01

    The Residual Heat Removal System (RHRS) of a PWR is designed to transfer thermal energy from the core after plant shutdown and maintain the plant in cold shutdown or refuelling conditions for extended periods of time. Initial reactor cooling after shutdown is achieved by dissipating heat through the steam generators (SGs) and discharging steam to the condenser by means of the Turbine Bypass System (TBS). When the reactor coolant temperature has dropped to about 160C and pressure has been reduced to 30 bar the RHRS is placed into operation. it reduces the coolant temperature to 50C within 20 hours after shutdown. The time margin for establishing alternate methods of heat removal following a failure of the RHRS depends on the Reactor Coolant System (RCS) temperature, the decay heat rate and the amount of RCS inventory. During some shutdown operations the RCS may be partially drained (e. g. to perform SG inspections). Decreased primary system inventory can significantly reduce the time available to recover the RHRS's function prior to bulk boiling and possible core uncovery. In the PKL test facility, which simulates a 1,300 MWe 4-loop PWR on a scale 1:145, a failure of RHRS under cold shutdown conditions was performed. This presentation gives a brief description of the test facility followed by the test objectives and results of this experiment

  9. Gas entrainment by one single French PWR spray, SARNET-2 spray benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Malet, J., E-mail: jeanne.malet@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire, Saclay (France); Mimouni, S., E-mail: stephane.mimouni@edf.fr [Electricité de France, EDF MF2E, Chatou (France); Manzini, G., E-mail: giovanni.manzini@rse-web.it [RSE, Milano (Italy); Xiao, J., E-mail: jianjun.xiao@kit.edu [IKET, KIT, Karlsruhe (Germany); Vyskocil, L., E-mail: vyl@ujv.cz [UJV Rez (Czech Republic); Siccama, N.B., E-mail: siccama@nrg.eu [NRG, Safety and Power (Netherlands); Huhtanen, R., E-mail: risto.huhtanen@vtt.fi [VTT, PO Box 1000, FI-02044 VTT (Finland)

    2015-02-15

    Highlights: • This paper presents a benchmark performed in the frame of the SARNET-2 EU project. • It concerns momentum transfer between a PWR spray and the surrounding gas. • The entrained gas velocities can vary up to 100% from one code to another. • Simplified boundary conditions for sprays are generally used by the code users. • It is shown how these simplified conditions impact the gas entrainment. - Abstract: This paper presents a benchmark performed in the frame of the SARNET-2 EU project, dealing with momentum transfer between a real-scale PWR spray and the surrounding gas. It presents a description of the IRSN tests on the CALIST facility, the participating codes (8 contributions), code-experiment and code-to-code comparisons. It is found that droplet velocities are almost well calculated one meter below the spray nozzle, even if the spread of the spray is not recovered and the values of the entrained gas velocity vary up to 100% from one code to another. Concerning sensitivity analysis, several ‘simplifications’ have been made by the contributors, especially based on the boundary conditions applied at the location where droplets are injected. It is shown here that such simplifications influence droplet and entrained gas characteristics. The next step will be to translate these conclusions in terms of variables representative of interesting parameters for nuclear safety.

  10. Gas entrainment by one single French PWR spray, SARNET-2 spray benchmark

    International Nuclear Information System (INIS)

    Malet, J.; Mimouni, S.; Manzini, G.; Xiao, J.; Vyskocil, L.; Siccama, N.B.; Huhtanen, R.

    2015-01-01

    Highlights: • This paper presents a benchmark performed in the frame of the SARNET-2 EU project. • It concerns momentum transfer between a PWR spray and the surrounding gas. • The entrained gas velocities can vary up to 100% from one code to another. • Simplified boundary conditions for sprays are generally used by the code users. • It is shown how these simplified conditions impact the gas entrainment. - Abstract: This paper presents a benchmark performed in the frame of the SARNET-2 EU project, dealing with momentum transfer between a real-scale PWR spray and the surrounding gas. It presents a description of the IRSN tests on the CALIST facility, the participating codes (8 contributions), code-experiment and code-to-code comparisons. It is found that droplet velocities are almost well calculated one meter below the spray nozzle, even if the spread of the spray is not recovered and the values of the entrained gas velocity vary up to 100% from one code to another. Concerning sensitivity analysis, several ‘simplifications’ have been made by the contributors, especially based on the boundary conditions applied at the location where droplets are injected. It is shown here that such simplifications influence droplet and entrained gas characteristics. The next step will be to translate these conclusions in terms of variables representative of interesting parameters for nuclear safety

  11. Air scaling and modeling studies for the 1/5-scale mark I boiling water reactor pressure suppression experiment

    Energy Technology Data Exchange (ETDEWEB)

    Lai, W.; McCauley, E.W.

    1978-01-04

    Results of table-top model experiments performed to investigate pool dynamics effects due to a postulated loss-of-coolant accident (LOCA) for the Peach Bottom Mark I boiling water reactor containment system guided subsequent conduct of the 1/5-scale torus experiment and provided new insight into the vertical load function (VLF). Pool dynamics results were qualitatively correct. Experiments with a 1/64-scale fully modeled drywell and torus showed that a 90/sup 0/ torus sector was adequate to reveal three-dimensional effects; the 1/5-scale torus experiment confirmed this.

  12. Air scaling and modeling studies for the 1/5-scale mark I boiling water reactor pressure suppression experiment

    International Nuclear Information System (INIS)

    Lai, W.; McCauley, E.W.

    1978-01-01

    Results of table-top model experiments performed to investigate pool dynamics effects due to a postulated loss-of-coolant accident (LOCA) for the Peach Bottom Mark I boiling water reactor containment system guided subsequent conduct of the 1/5-scale torus experiment and provided new insight into the vertical load function (VLF). Pool dynamics results were qualitatively correct. Experiments with a 1/64-scale fully modeled drywell and torus showed that a 90 0 torus sector was adequate to reveal three-dimensional effects; the 1/5-scale torus experiment confirmed this

  13. Analysis of core physics and thermal-hydraulics results of control rod withdrawal experiments in the LOFT facility

    International Nuclear Information System (INIS)

    Varacalle, D.J. Jr.; Chen, T.H.; Harvego, E.A.; Ollikkala, H.

    1983-01-01

    Two anticipated transient experiments simulating an uncontrolled control rod withdrawal event in a pressurized water reactor (PWR) were conducted in the Loss-of-Fluid Test (LOFT) Facility at the Idaho National Engineering Laboratory. The scaled LOFT 50-MW(t) PWR includes most of the principal features of larger commercial PWRs. The experiments tested the ability of reactor analysis codes to accurately calculate core reactor physics and thermal-hydraulic phenomena in an integral reactor system. The initial conditions and scaled operating parameters for the experiments were representative of those expected in a commercial PWR. In both experiments, all four LOFT control rod assemblies were withdrawn at a reactor power of 37.5 MW and a system pressure of 14.8 MPa

  14. Validating Westinghouse atom 16 x 16 and 18 x 18 PWR fuel performance

    International Nuclear Information System (INIS)

    Andersson, S.; Gustafson, J.; Jourdain, P.; Lindstroem, L.; Hallstadius, L.; Hofling, C.G.

    2001-01-01

    Westinghouse Atom designs and fabricates PWR fuel for all major European fuel types: 17 x 17 standard (12 ft) and 17 x 17 XL (14 ft) for Westinghouse type PWRs, and 16 x 16 and 18 x 18 fuel for Siemens type PWRs. The W Atom PWR fuel designs are based on the extensive Westinghouse CE PWR fuel experience from combustion engineering type PWRs. The W atom designs utilise basic design features from the W CE fuel tradition, such as all-Zircaloy mid grids and the proven ( 6 rod years) Guardian TM debris catcher, which is integrated in the bottom Inconel grid. Several new features have been developed to meet with stringent European requirements originating from requirements on very high burnup, in combination with low-leakage core operating strategies and high coolant temperatures. The overall reliability of the Westinghouse Atom PWR fuel is very high; no fuel failure has been detected since 1997. (orig.)

  15. Large scale laboratory diffusion experiments in clay rocks

    International Nuclear Information System (INIS)

    Garcia-Gutierrez, M.; Missana, T.; Mingarro, M.; Martin, P.L.; Cormenzana, J.L.

    2005-01-01

    Full text of publication follows: Clay formations are potential host rocks for high-level radioactive waste repositories. In clay materials the radionuclide diffusion is the main transport mechanism. Thus, the understanding of the diffusion processes and the determination of diffusion parameters in conditions as similar as possible to the real ones, are critical for the performance assessment of deep geological repository. Diffusion coefficients are mainly measured in the laboratory using small samples, after a preparation to fit into the diffusion cell. In addition, a few field tests are usually performed for confirming laboratory results, and analyse scale effects. In field or 'in situ' tests the experimental set-up usually includes the injection of a tracer diluted in reconstituted formation water into a packed off section of a borehole. Both experimental systems may produce artefacts in the determination of diffusion coefficients. In laboratory the preparation of the sample can generate structural change mainly if the consolidated clay have a layered fabric, and in field test the introduction of water could modify the properties of the saturated clay in the first few centimeters, just where radionuclide diffusion is expected to take place. In this work, a large scale laboratory diffusion experiment is proposed, using a large cylindrical sample of consolidated clay that can overcome the above mentioned problems. The tracers used were mixed with clay obtained by drilling a central hole, re-compacted into the hole at approximately the same density as the consolidated block and finally sealed. Neither additional treatment of the sample nor external monitoring are needed. After the experimental time needed for diffusion to take place (estimated by scoping calculations) the block was sampled to obtain a 3D distribution of the tracer concentration and the results were modelled. An additional advantage of the proposed configuration is that it could be used in 'in situ

  16. Maintenance service for major component of PWR plant. Replacement of pressurizer safe end weld

    International Nuclear Information System (INIS)

    Miyoshi, Yoshiyuki; Kobayashi, Yuki; Yamamoto, Kazuhide; Ueda, Takeshi; Suda, Naoki; Shintani, Takashi

    2017-01-01

    In October 2016, MHI completed the replacement of safe end weld of pressurizer (Pz) of Ringhals unit 3, which was the first maintenance work for main component of pressurized water reactor (PWR) plant in Europe. For higher reliability and longer lifetime of PWR plant, MHI has conducted many kinds of maintenance works of main components of PWR plants in Japan against stress corrosion cracking due to aging degradation. Technical process for replacement of Pz safe end weld were established by MHI. MHI has experienced the work for 21 PWR units in Japan. That of Ringhals unit 3 was planned and conducted based on the experiences. In this work, Alloy 600 used for welds of nozzles of Pz was replaced with Alloy 690. Alloy 690 is more corrosive-resistant than Alloy 600. Specially designed equipment and technical process were developed and established by MHI to replace safe end weld of Pz and applied for the Ringhals unit 3 as a first application in Europe. The application had been performed in success and achieved the planned replacement work duration and total radiation dose by using sophisticated machining and welding equipment designed to meet the requirements to be small, lightweight and remote-controlled and operating by well skilled MHI personnel experienced in maintenance activities for major components of PWR plant in Japan. The success shows that the experience, activities and technology developed in Japan for main components of PWR plant shall be applicable to contribute reliable operations of nuclear power plants in Europe and other countries. (author)

  17. Conceptual design of simplified PWR

    International Nuclear Information System (INIS)

    Tabata, Hiroaki

    1996-01-01

    The limited availability for location of nuclear power plant in Japan makes plants with higher power ratings more desirable. Having no intention of constructing medium-sized plants as a next generation standard plant, Japanese utilities are interested in applying passive technologies to large ones. So, Japanese utilities have studied large passive plants based on AP600 and SBWR as alternative future LWRs. In a joint effort to develop a new generation nuclear power plant which is more friendly to operator and maintenance personnel and is economically competitive with alternative sources of power generation, JAPC and Japanese Utilities started the study to modify AP600 and SBWR, in order to accommodate the Japanese requirements. During a six year program up to 1994, basic concepts for 1000 MWe class Simplified PWR (SPWR) and Simplified BWR (SBWR) were developed, though there still remain several areas to be improved. These studies have now stepped into the phase of reducing construction cost and searching for maximum power rating that can be attained by reasonably practical technology. These results also suggest that it is hopeful to develop a large 3-loop passive plant (∼1200 MWe). Since Korea mainly deals with PWR, this paper summarizes SPWR study. The SPWR is jointly studied by JAPC, Japanese PWR Utilities, EdF, WH and Mitsubishi Heavy Industry. Using the AP-600 reference design as a basis, we enlarged the plant size to 3-loops and added engineering features to conform with Japanese practice and Utilities' preference. The SPWR program definitively confirmed the feasibility of a passive plant with an NSSS rating about 1000 MWe and 3 loops. (J.P.N.)

  18. Polynomial parameterized representation of macroscopic cross section for PWR reactor

    International Nuclear Information System (INIS)

    Fiel, Joao Claudio B.

    2015-01-01

    The purpose of this work is to describe, by means of Tchebychev polynomial, a parameterized representation of the homogenized macroscopic cross section for PWR fuel element as a function of soluble boron concentration, moderator temperature, fuel temperature, moderator density and 235 U 92 enrichment. Analyzed cross sections are: fission, scattering, total, transport, absorption and capture. This parameterization enables a quick and easy determination of the problem-dependent cross-sections to be used in few groups calculations. The methodology presented here will enable to provide cross-sections values to perform PWR core calculations without the need to generate them based on computer code calculations using standard steps. The results obtained by parameterized cross-sections functions, when compared with the cross-section generated by SCALE code calculations, or when compared with K inf , generated by MCNPX code calculations, show a difference of less than 0.7 percent. (author)

  19. PWR accident management realated tests: some Bethsy results

    International Nuclear Information System (INIS)

    Clement, P.; Chataing, T.; Deruaz, R.

    1993-01-01

    The BETHSY integral test facility which is a scaled down model of a 3 loop FRAMATOME PWR and is currently operated at the Nuclear Center of Grenoble, forms an important part of the French strategy for PWR Accident Management. In this paper the features of both the facility and the experimental program are presented. Two accident transients: a total loss of feedwater and a 2'' cold leg break in case of High Pressure Safety Injection System failure, involving either Event Oriented - or State Oriented-Emergency Operating Procedures (EO-EOP or SO-EOP) are described and the system response analyzed. CATHARE calculation results are also presented which illustrate the ability of this code to adequately predict the key phenomena of these transients. (authors). 13 figs., 11 refs., 2 tabs

  20. Resonant Wave Energy Converters: Small-scale field experiments and first full-scale prototype

    International Nuclear Information System (INIS)

    Arena, Felice; Fiamma, Vincenzo; Iannolo, Roberto; Laface, Valentina; Malara, Giovanni; Romolo, Alessandra; Strati Federica Maria

    2015-01-01

    The Resonant Wave Energy Converter 3 (REWEC3) is a device belonging to the family of Oscillating Water Columns (OWCs), that can convert the energy of incident waves into electrical energy via turbines. In contrast to classical OWCs, it incorporates a small vertical U-shaped duct to connect the water column to the open wave field. This article shows the results of a small-scale field experiment involving a REWEC3 designed for working with a 2 kW turbine. Then, the next experimental activity on a REWEC3 installed in the NOEL laboratory with the collaboration of ENEA, is presented. Finally, the first prototype of ReWEC3 under construction in Civitavecchia (Rome, Italy) is shown. The crucial features of the construction stage are discussed and some initial performances are provided. [it

  1. Graph processing platforms at scale: practices and experiences

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Seung-Hwan [ORNL; Lee, Sangkeun (Matt) [ORNL; Brown, Tyler C [ORNL; Sukumar, Sreenivas R [ORNL; Ganesh, Gautam [ORNL

    2015-01-01

    Graph analysis unveils hidden associations of data in many phenomena and artifacts, such as road network, social networks, genomic information, and scientific collaboration. Unfortunately, a wide diversity in the characteristics of graphs and graph operations make it challenging to find a right combination of tools and implementation of algorithms to discover desired knowledge from the target data set. This study presents an extensive empirical study of three representative graph processing platforms: Pegasus, GraphX, and Urika. Each system represents a combination of options in data model, processing paradigm, and infrastructure. We benchmarked each platform using three popular graph operations, degree distribution, connected components, and PageRank over a variety of real-world graphs. Our experiments show that each graph processing platform shows different strength, depending the type of graph operations. While Urika performs the best in non-iterative operations like degree distribution, GraphX outputforms iterative operations like connected components and PageRank. In addition, we discuss challenges to optimize the performance of each platform over large scale real world graphs.

  2. Frictional sliding in layered rock: laboratory-scale experiments

    International Nuclear Information System (INIS)

    Buescher, B.J.; Perry, K.E. Jr.; Epstein, J.S.

    1996-09-01

    The work is part of the rock mechanics effort for the Yucca Mountain Site Characterization Program. The laboratory-scale experiments are intended to provide high quality data on the mechanical behavior of jointed structures that can be used to validate complex numerical models for rock-mass behavior. Frictional sliding between simulated rock joints was studied using phase shifting moire interferometry. A model, constructed from stacks of machined and sandblasted granite plates, contained a central hole bore normal to the place so that frictional slip would be induced between the plates near the hole under compressive loading. Results show a clear evolution of slip with increasing load. Since the rock was not cycled through loading- unloading, the quantitative differences between the three data sets are probably due to a ''wearing-in'' effect. The highly variable spatial frequency of the data is probably due to the large grain size of the granite and the stochastic frictional processes. An unusual feature of the evolution of slip with increasing load is that as the load gets larger, some plates seem to return to a null position. Figs, 6 refs

  3. A study on effective system depressurization during a PWR vessel bottom break LOCA with HPI failure and gas inflow prevention. ROSA-V/LSTF test SB-PV-05

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

    2006-11-01

    A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility (LSTF) of ROSA-V program to study effects of accident management (AM) measures on core cooling, which are important in case of total failure of high pressure injection (HPI) system during an SBLOCA at a pressurized water reactor (PWR). The LSTF is a full-height and 1/48 volume-scaled facility simulating a 4-loop Westinghouse-type PWR (3423 MWt). The experiment, SB-PV-05, simulated a PWR vessel bottom SBLOCA with a rupture of nine instrument tubes, which is equivalent to 0.18% cold leg break. It is clarified that AM actions with steam generator (SG) depressurization to achieve a primary loop cooling rate at -55 K/h and auxiliary feedwater supply for 30 minutes are effective to avoid core uncovery by actuating the low pressure injection (LPI) system. It is also shown through the comparison with the previous experiment of SB-PV-03 that prevention of non-condensable gas inflow from the accumulator injection system (AIS) is very important to actuate the LPI to achieve adequate core cooling. This report presents experiment results of SB-PV-05 in detail and shows the effects of gas inflow prevention on core cooling through the estimation of primary coolant mass and energy balance in the primary system. (author)

  4. Surveillance of vibrations in PWR

    International Nuclear Information System (INIS)

    Espefaelt, R.; Lorenzen, J.; Aakerhielm, F.

    1980-07-01

    The core of a PWR - including fuel elements, internal structure, control rods and core support structure inside the pressure vessel - is subjected to forces which can cause vibrations. One sensitive means to detect and analyse such vibrations is by means of the noise from incore and excore neutron detector signals. In this project noise recordings have been made on two occasions in the Ringhals 2 plant and the obtained data been analysed using the Studsvik Noise Analysis Program System (SNAPS). The results have been intepreted and a detailed description of the vibrational status of the core and pressure vessel internals has been produced. On the basis of the obtained results it is proposed that neutron signal noise analysis should be performed at each PWR plant in the beginning, middle and end of each fuel cycle and an analysis be made using the methods developed in the project. It would also provide a contribution to a higher degree of preparedness for diagnostic tasks in case of unexpected and abnormal events. (author)

  5. Some thermohydraulics of closure head adapters in a 3 loops PWR

    International Nuclear Information System (INIS)

    Hofmann, F.; Daubert, O.; Bertrand, C.; Hecker, M.; Arnoux-Guisse, F.; Bonnin, O.

    1995-12-01

    In 1993 a R and D action, based on numerical simulations and experiments on PWR's upper head was initiated. This paper presents the test facility TRAVERSIN (a scale model of a 900 MW adapter) and calculations performed on the geometry of different upper head sections with the Thermalhydraulic Finite Element Code N3S used for 2D and 3D computations. The paper presents the method followed to bring the adapter and upper head study to a successful conclusion. Two complementary approaches are performed to obtain global results on complete fluid flow in the upper head and local results on the flow around the adapters of closure head. A validation test case of these experimental and numerical tools is also presented. (authors). 7 refs., 9 figs., 1 tab

  6. The development of flow test technology for PWR fuel assembly

    International Nuclear Information System (INIS)

    Chung, Moon Ki; Cha, Chong Hee; Chung, Chang Hwan; Chun, Se Young; Song, Chul Hwa; Chung, Heung Joon; Won, Soon Yeun; Cho, Yeong Rho; Kim, Bok Deuk

    1988-05-01

    KAERI has an extensive program to develope PWR fuel assembly. In relation to the program, development of flow test technology is needed to evaluate the thermal hydraulic compactibility and mechanical integrity of domestically fabricated nuclear fuels. A high-pressure and high-temperature flow test facility was designed to test domestically fabricated fuel assembly. The test section of the facility has capacity of a 6x6 full length PWR fuel assembly. A flow test rig was designed and installed at Cold Test Loop to carry out model experiments with 5x5 rod assembly under atmosphere pressure to get information about the characteristics of pressure loss of spacer grids and velocity distribution in the subchannels. LDV measuring technology was established using TSI's Laser Dopper Velocimeter 9100-3 System

  7. Condensate polishing guidelines for PWR and BWR plants

    International Nuclear Information System (INIS)

    Robbins, P.; Crinigan, P.; Graham, B.; Kohlmann, R.; Crosby, C.; Seager, J.; Bosold, R.; Gillen, J.; Kristensen, J.; McKeen, A.; Jones, V.; Sawochka, S.; Siegwarth, D.; Keeling, D.; Polidoroff, T.; Morgan, D.; Rickertsen, D.; Dyson, A.; Mills, W.; Coleman, L.

    1993-03-01

    Under EPRI sponsorship, an industry committee, similar in form and operation to other guideline committees, was created to develop Condensate Polishing Guidelines for both PWR and BWR systems. The committee reviewed the available utility and water treatment industry experience on system design and performance and incorporated operational and state-of-the-art information into document. These guidelines help utilities to optimize present condensate polisher designs as well as be a resource for retrofits or new construction. These guidelines present information that has not previously been presented in any consensus industry document. The committee generated guidelines that cover both deep bed and powdered resin systems as an integral part of the chemistry of PWR and BWR plants. The guidelines are separated into sections that deal with the basis for condensate polishing, system design, resin design and application, data management and performance and management responsibilities

  8. Conceptual study of advanced PWR core design

    International Nuclear Information System (INIS)

    Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong.

    1997-09-01

    The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs

  9. Conceptual study of advanced PWR core design

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong

    1997-09-01

    The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs.

  10. Computational fluid dynamics (CFD) round robin benchmark for a pressurized water reactor (PWR) rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Shin K., E-mail: paengki1@tamu.edu; Hassan, Yassin A.

    2016-05-15

    Highlights: • The capabilities of steady RANS models were directly assessed for full axial scale experiment. • The importance of mesh and conjugate heat transfer was reaffirmed. • The rod inner-surface temperature was directly compared. • The steady RANS calculations showed a limitation in the prediction of circumferential distribution of the rod surface temperature. - Abstract: This study examined the capabilities and limitations of steady Reynolds-Averaged Navier–Stokes (RANS) approach for pressurized water reactor (PWR) rod bundle problems, based on the round robin benchmark of computational fluid dynamics (CFD) codes against the NESTOR experiment for a 5 × 5 rod bundle with typical split-type mixing vane grids (MVGs). The round robin exercise against the high-fidelity, broad-range (covering multi-spans and entire lateral domain) NESTOR experimental data for both the flow field and the rod temperatures enabled us to obtain important insights into CFD prediction and validation for the split-type MVG PWR rod bundle problem. It was found that the steady RANS turbulence models with wall function could reasonably predict two key variables for a rod bundle problem – grid span pressure loss and the rod surface temperature – once mesh (type, resolution, and configuration) was suitable and conjugate heat transfer was properly considered. However, they over-predicted the magnitude of the circumferential variation of the rod surface temperature and could not capture its peak azimuthal locations for a central rod in the wake of the MVG. These discrepancies in the rod surface temperature were probably because the steady RANS approach could not capture unsteady, large-scale cross-flow fluctuations and qualitative cross-flow pattern change due to the laterally confined test section. Based on this benchmarking study, lessons and recommendations about experimental methods as well as CFD methods were also provided for the future research.

  11. Conceptual study on advanced PWR system

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Yoon Young; Chang, M H; Yu, K J; Lee, D J; Cho, B H; Kim, H Y; Yoon, J H; Lee, Y J; Kim, J P; Park, C T; Seo, J K; Kang, H S; Kim, J I; Kim, Y W; Kim, Y H

    1997-07-01

    In this study, the adoptable essential technologies and reference design concept of the advanced reactor were developed and related basic experiments were performed. (1) Once-through Helical Steam Generator: a performance analysis computer code for heli-coiled steam generator was developed for thermal sizing of steam generator and determination of thermal-hydraulic parameters. (2) Self-pressurizing pressurizer : a performance analysis computer code for cold pressurizer was developed. (3) Control rod drive mechanism for fine control : type and function were surveyed. (4) CHF in passive PWR condition : development of the prediction model bundle CHF by introducing the correction factor from the data base. (5) Passive cooling concepts for concrete containment systems: development of the PCCS heat transfer coefficient. (6) Steam injector concepts: analysis and experiment were conducted. (7) Fluidic diode concepts : analysis and experiment were conducted. (8) Wet thermal insulator : tests for thin steel layers and assessment of materials. (9) Passive residual heat removal system : a performance analysis computer code for PRHRS was developed and the conformance to EPRI requirement was checked. (author). 18 refs., 55 tabs., 137 figs.

  12. Conceptual study on advanced PWR system

    International Nuclear Information System (INIS)

    Bae, Yoon Young; Chang, M. H.; Yu, K. J.; Lee, D. J.; Cho, B. H.; Kim, H. Y.; Yoon, J. H.; Lee, Y. J.; Kim, J. P.; Park, C. T.; Seo, J. K.; Kang, H. S.; Kim, J. I.; Kim, Y. W.; Kim, Y. H.

    1997-07-01

    In this study, the adoptable essential technologies and reference design concept of the advanced reactor were developed and related basic experiments were performed. 1) Once-through Helical Steam Generator: a performance analysis computer code for heli-coiled steam generator was developed for thermal sizing of steam generator and determination of thermal-hydraulic parameters. 2) Self-pressurizing pressurizer : a performance analysis computer code for cold pressurizer was developed. 3) Control rod drive mechanism for fine control : type and function were surveyed. 4) CHF in passive PWR condition : development of the prediction model bundle CHF by introducing the correction factor from the data base. 5) Passive cooling concepts for concrete containment systems: development of the PCCS heat transfer coefficient. 6) Steam injector concepts: analysis and experiment were conducted. 7) Fluidic diode concepts : analysis and experiment were conducted. 8) Wet thermal insulator : tests for thin steel layers and assessment of materials. 9) Passive residual heat removal system : a performance analysis computer code for PRHRS was developed and the conformance to EPRI requirement was checked. (author). 18 refs., 55 tabs., 137 figs

  13. 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

    Energy Technology Data Exchange (ETDEWEB)

    González-Robles, E., E-mail: ernesto.gonzalez-robles@kit.edu [CTM Centre Tecnològic, Plaça de la Ciència 2, 08243 Manresa (Spain); Serrano-Purroy, D. [European Commission - EC, Joint Research Centre (JRC), Institute for Transuranium Elements - ITU, Postfach 2340, D-76125 Karlsruhe (Germany); Sureda, R. [CTM Centre Tecnològic, Plaça de la Ciència 2, 08243 Manresa (Spain); Casas, I. [Chemical Engineering Department, Universitat Politècnica de Catalunya, Av. Diagonal 647, 08028 Barcelona (Spain); Pablo, J. de [CTM Centre Tecnològic, Plaça de la Ciència 2, 08243 Manresa (Spain); Chemical Engineering Department, Universitat Politècnica de Catalunya, Av. Diagonal 647, 08028 Barcelona (Spain)

    2015-10-15

    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 UO{sub 2} 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 (FIAP{sub c}) 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.

  14. PWR thermocouple mechanical sealing structure

    International Nuclear Information System (INIS)

    Shen Qiuping; He Youguang

    1991-08-01

    The PWR in-core temperature detection device, which is one of measures to insure reactor safety operation, is to monitor and diagnose reactor thermal power output and in-core power distribution. The temperature detection device system uses thermocouples as measuring elements with stainless steel protecting sleeves. The thermocouple has a limited service time and should be replaced after its service time has reached. A new sealing device for the thermocouples of reactor in-core temperature detection system has been developed to facilitate replacement. The structure is complete tight under high temperature and pressure without any leakage and seepage, and easy to be assembled or disassembled in radioactive environment. The device is designed to make it possible to replace the thermocouple one by one if necessary. This is a new, simple and practical structure

  15. Dither Gyro Scale Factor Calibration: GOES-16 Flight Experience

    Science.gov (United States)

    Reth, Alan D.; Freesland, Douglas C.; Krimchansky, Alexander

    2018-01-01

    This poster is a sequel to a paper presented at the 34th Annual AAS Guidance and Control Conference in 2011, which first introduced dither-based calibration of gyro scale factors. The dither approach uses very small excitations, avoiding the need to take instruments offline during gyro scale factor calibration. In 2017, the dither calibration technique was successfully used to estimate gyro scale factors on the GOES-16 satellite. On-orbit dither calibration results were compared to more traditional methods using large angle spacecraft slews about each gyro axis, requiring interruption of science. The results demonstrate that the dither technique can estimate gyro scale factors to better than 2000 ppm during normal science observations.

  16. Results of intermediate-scale hot isostatic press can experiments

    International Nuclear Information System (INIS)

    Nelson, L.O.; Vinjamuri, K.

    1995-05-01

    Radioactive high-level waste (HLW) has been managed at the Idaho Chemical Processing Plant (ICPP) for a number of years. Since 1963, liquid HLW has been solidified into a granular solid (calcine). Presently, over 3,800 m 3 of calcine is stored in partially-underground stainless steel bins. Four intermediate- scale HLW can tests (two 6-in OD x 12-in tall and two 4-in OD x 7-in tall) are described and compared to small-scale HIP can tests (1- to 3-in OD x 1- to 4.5-in tall). The intermediate-scale HIP cans were loaded with a 70/30 calcine/frit blend and HIPped at an off-site facility at 1050 degrees C; and 20 ksi. The dimensions of two cans (4-in OD x 7-in tall) were monitored during the HIP cycle with eddy-current sensors. The sensor measurements indicated that can deformation occurs rapidly at 700 degrees C; after which, there is little additional can shrinkage. HIP cans were subjected to a number of analyses including calculation of the overall packing efficiency (56 to 59%), measurement of glass-ceramic (3.0 to 3.2 g/cc), 14-day MCC-1 leach testing (total mass loss rates 2 day), and scanning electron microscopy (SEM). Based on these analyses, the glass-ceramic material produced in intermediate-scale cans is similar to material produced in small-scale cans. No major scale-up problems were indicated. Based on the packing efficiency observed in intermediate- and small-scale tests, the overall packing efficiency of production-scale (24-in OD x 36- to 190-in tall) cans would be approximately 64% for a pre-HIP right-circular cylinder geometry. An efficiency of 64% would represent a volume reduction factor of 2.5 over a candidate glass waste prepared at 33 wt% waste loading

  17. Energy confinement scaling in tokamaks: some implications of recent experiments with ohmic and strong auxiliary heating

    International Nuclear Information System (INIS)

    Goldston, R.J.

    1984-02-01

    Recent results from confinement scaling experiments on tokamaks with ohmic and strong auxiliary heating are reviewed. An attempt is made to draw these results together into a low-density ohmic confinement scaling law, and a scaling law for confinement with auxiliary heating. The auxiliary heating confinement law may also serve to explain the saturation in tau/sub E/ vs anti n/sub e/ observed in some ohmic heating density scaling experiments

  18. The Design of a Fire Source in Scale-Model Experiments with Smoke Ventilation

    DEFF Research Database (Denmark)

    Nielsen, Peter Vilhelm; Brohus, Henrik; la Cour-Harbo, H.

    2004-01-01

    The paper describes the design of a fire and a smoke source for scale-model experiments with smoke ventilation. It is only possible to work with scale-model experiments where the Reynolds number is reduced compared to full scale, and it is demonstrated that special attention to the fire source...... (heat and smoke source) may improve the possibility of obtaining Reynolds number independent solutions with a fully developed flow. The paper shows scale-model experiments for the Ofenegg tunnel case. Design of a fire source for experiments with smoke ventilation in a large room and smoke movement...

  19. Babcock and Wilcox advanced PWR development

    International Nuclear Information System (INIS)

    Kulynych, G.E.; Lemon, J.E.

    1986-01-01

    The Babcock and Wilcox 600 MWe PWR design is discussed. Main features of the new B-600 design are improvements in reactor system configuration, glandless coolant pumps, safety features, core design and steam generators

  20. Hydrogen production in a PWR during LOCA

    International Nuclear Information System (INIS)

    Cassette, P.

    1983-12-01

    The purpose of this paper is to provide information on hydrogen generation during LOCA in French 900 MW PWR power plants. The design basis accident is taken into account as well as more severe accidents assuming failure of emergency systems

  1. Large scale experiments as a tool for numerical model development

    DEFF Research Database (Denmark)

    Kirkegaard, Jens; Hansen, Erik Asp; Fuchs, Jesper

    2003-01-01

    Experimental modelling is an important tool for study of hydrodynamic phenomena. The applicability of experiments can be expanded by the use of numerical models and experiments are important for documentation of the validity of numerical tools. In other cases numerical tools can be applied...

  2. HUNGARIAN EXPERIENCES WITH THE BELIEFS ABOUT ATTRACTIVENESS SCALE.

    Science.gov (United States)

    Czeglédi, Edit; Szabo, Kornélia

    2016-03-30

    Sociocultural influences regarding bodily appearance and their psychological consequences play a considerable role in the development and maintenance of body image disturbance and eating disorders. The purpose of the study was to explore the psychometric properties of the Beliefs About Attractiveness Scale-Revised and its correlates among young adults in Hungary. In our cross-sectional online study, participants were 18-35 years old (N = 820, 40% male). self-reported anthropometric data, Beliefs About Attractiveness Scale-Revised, Eating Disorder Inventory, SCOFF questionnaire, Sociocultural Attitudes Towards Appearance Questionnaire-3, and Rosenberg Self-esteem Scale. The exploratory factor analysis showed that the fit indices of the three-factor solution are acceptable (χ²(₁₇₁)) = 5124.8, p scales were confirmed. Among those who were at risk of developing an eating disorder, all of the measured beliefs were significantly greater than among those who were not at risk (thin: Z = 6.501, p Scale-Revised is a reliable, valid measure and we suggest its introduction into Hungarian research. Relationships between beliefs about attractiveness and self- esteem, body image and eating disorders suggest intervention opportunities in with regards to prevention and treatment of eating disorders.

  3. Structure-dynamic model verification calculation of PWR 5 tests

    International Nuclear Information System (INIS)

    Engel, R.

    1980-02-01

    Within reactor safety research project RS 16 B of the German Federal Ministry of Research and Technology (BMFT), blowdown experiments are conducted at Battelle Institut e.V. Frankfurt/Main using a model reactor pressure vessel with a height of 11,2 m and internals corresponding to those in a PWR. In the present report the dynamic loading on the pressure vessel internals (upper perforated plate and barrel suspension) during the DWR 5 experiment are calculated by means of a vertical and horizontal dynamic model using the CESHOCK code. The equations of motion are resolved by direct integration. (orig./RW) [de

  4. Sulfur-Iodine Integrated Lab Scale Experiment Development

    Energy Technology Data Exchange (ETDEWEB)

    Russ, Ben

    2011-05-27

    The sulfur-iodine (SI) cycle was deermined to be the best cycle for coupling to a high temperature reactor (HTR) because of its high efficiency and potential for further improvement. The Japanese Atomic Energy Agency (JAEA) has also selected the SI process for further development and has successfully completed bench-scale demonstrations of the SI process at atmospheric pressure. JEA also plans to proceed with pilot-scale demonstrations of the SI process and eventually plans to couple an SI demonstration plant to its High Temperature Test Reactor (HHTR). As part of an international NERI project, GA, SNL, and the Frech Commissariat L'Energie Atomique performed laboratory-scale demonstrations of the SI process at prototypical temperatures and pressures. This demonstration was performed at GA in San Diego, CA and concluded in April 2009.

  5. Sulfur-Iodine Integrated Lab Scale Experiment Development

    International Nuclear Information System (INIS)

    Russ, Ben

    2011-01-01

    The sulfur-iodine (SI) cycle was determined to be the best cycle for coupling to a high temperature reactor (HTR) because of its high efficiency and potential for further improvement. The Japanese Atomic Energy Agency (JAEA) has also selected the SI process for further development and has successfully completed bench-scale demonstrations of the SI process at atmospheric pressure. JEA also plans to proceed with pilot-scale demonstrations of the SI process and eventually plans to couple an SI demonstration plant to its High Temperature Test Reactor (HHTR). As part of an international NERI project, GA, SNL, and the Frech Commissariat L'Energie Atomique performed laboratory-scale demonstrations of the SI process at prototypical temperatures and pressures. This demonstration was performed at GA in San Diego, CA and concluded in April 2009.

  6. Alloy development for high burnup cladding (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, R. [Kraftwerk Union AG, Mulheim (Germany); Jeong, Y.H.; Baek, K.H.; Kim, S.J.; Choi, B.K.; Kim, J.M. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-04-01

    An overview on current alloy development for high burnup PWR fuel cladding is given. It is mainly based on literature data. First, the reasons for an increase of the current mean discharge burnup from 35 MWd / kg(U) to 70 MWd / kg(U) are outlined. From the material data, it is shown that a batch average burnup of 60-70 MWd / kg(U), as aimed by many fuel vendors, can not be achieved with stand (=ASTM-) Zry-4 cladding tubes without violating accepted design criteria. Specifically criteria which limit maximum oxide scale thickness and maximum hydrogen content, and to a less degree, maximum creep and growth rate, can not be achieved. The development potential of standard Zry-4 is shown. Even when taking advantage of this potential, it is shown that an 'improved' Zry-4 is reaching its limits when it achieves the target burnup. The behavior of some Zr alloys outside the ASTM range is shown, and the advantages and disadvantages of the 3 alloy groups (ZrSn+transition metals, ZrNb, ZrSnNb+transition metals) which are currently considered to have the development potential for high burnup cladding materials are depicted. Finally, conclusions are drawn. (author). 14 refs., 11 tabs., 82 figs.

  7. Critical experiments for large scale enriched uranium solution handling

    International Nuclear Information System (INIS)

    Tanner, J.E.; Forehand, H.M.

    1985-01-01

    The authors have performed 17 critical experiments with a concentrated aqueous uranyl nitrate solution contained in an annular cylindrical tank, with annular cylindrical absorbers of stainless steel and/or polyethylene inside. k/sub eff/ calculated by KENO IV, employing 16-group Hansen-Roach cross sections, average 0.977. There is a variation of the calculational bias among the separate experiments, but it is too small to allow assigning it to specific components of the equipment. They are now performing critical experiments with a more concentrated uranyl nitrate solution in pairs of very squat cylindrical tanks with disc shaped absorbers and reflectors of carbon steel, stainless steel, nitronic-50, plain and borated polyethylene. These experiments are in support of upgrading fuel reprocessing at the Idaho Chemical Processing Plant

  8. Comparison of PWR-IMF and FR fuel cycles

    International Nuclear Information System (INIS)

    Darilek, Petr; Zajac, Radoslav; Breza, Juraj; Necas, Vladimir

    2007-01-01

    The paper gives a comparison of PWR (Russia origin VVER-440) cycle with improved micro-heterogeneous inert matrix fuel assemblies and FR cycle. Micro-heterogeneous combined assembly contains transmutation pins with Pu and MAs from burned uranium reprocessing and standard uranium pins. Cycle analyses were performed by HELIOS spectral code and SCALE code system. Comparison is based on fuel cycle indicators, used in the project RED-IMPACT - part of EU FP6. Advantages of both closed cycles are pointed out. (authors)

  9. Medium-scale melt-sodium fragmentation experiments

    International Nuclear Information System (INIS)

    Chu, T.Y.; Beattie, A.G.; Drotning, W.D.; Powers, D.A.

    1979-01-01

    The results of a series of fragmentation experiments involving up to 20 Kg of thermitically produced high temperature melts and 23 Kg of sodium are presented. Except for one experiment where some centimeter size particles are observed, the fragment distributions seem to be in the range of previous data. Spatial distribution of the fragments in the debris bed appears to be stratified. Scanning electron micrographs of fragments indicate fragmentation to be occurring in the molten state for the more intense interactions observed. Interaction data obtained show quiescent periods of 0.5 to 1.5 second between pressure pulses. The force impulse values per unit mass of melt seems to be in the same range as previous experiments

  10. A large scale double beta and dark matter experiment: GENIUS

    International Nuclear Information System (INIS)

    Hellmig, J.; Klapdor-Kleingrothaus, H.V.

    1997-01-01

    The recent results from the HEIDELBERG-MOSCOW experiment have demonstrated the large potential of double beta decay to search for new physics beyond the Standard Model. To increase by a major step the present sensitivity for double beta decay and dark matter search much bigger source strengths and much lower backgrounds are needed than used in experiments under operation at present or under construction. We present here a study of a project proposed recently, which would operate one ton of 'naked' enriched germanium-detectors in liquid nitrogen as shielding in an underground setup (GENIUS). It improves the sensitivity to neutrino masses to 0.01 eV. A ten ton version would probe neutrino masses even down to 10 -3 eV. The first version would allow to test the atmospheric neutrino problem, the second at least part of the solar neutrino problem. Both versions would allow in addition significant contributions to testing several classes of GUT models. These are especially tests of R-parity breaking supersymmetry models, leptoquark masses and mechanism and right-handed W-boson masses comparable to LHC. The second issue of the experiment is the search for dark matter in the universe. The entire MSSM parameter space for prediction of neutralinos as dark matter particles could be covered already in a first step of the full experiment with the same purity requirements, but using only 100 kg of 76 Ge or even of natural Ge making the experiment competitive to LHC in the search for supersymmetry.The layout of the proposed experiment is discussed and the shielding and purity requirements are studied using GEANT Monte Carlo simulations. As a demonstration of the feasibility of theexperiment first results of operating a 'naked' Ge detector in liquid nitrogen are presented. (orig.)

  11. A large scale field experiment in the Amazon basin (LAMBADA/BATERISTA)

    NARCIS (Netherlands)

    Dolman, A.J.; Kabat, P.; Gash, J.H.C.; Noilhan, J.; Jochum, A.M.; Nobre, C.

    1995-01-01

    A description is given of a large-scale field experiment planned in the Amazon basin, aimed at assessing the large-scale balances of energy, water and carbon dioxide. The embedding of this experiment in global change programmes is described, viz. the Biospheric Aspects of the Hydrological Cycle

  12. Electrochemical removal of salts from masonry - Experiences from pilot scale

    DEFF Research Database (Denmark)

    Ottosen, Lisbeth M.; Rörig-Dalgaard, Inge; Villumsen, Arne

    2008-01-01

    A pilot experiment with newly developed electrodes was tested for removal of contaminating salts from brick masonry where plaster peeling was a problem. A high concentration of sulfate was found at the height where the paint peeling was most pronounced. The concentrations of chloride and nitrate ...

  13. Measuring Instructional Differentiation in a Large-Scale Experiment

    Science.gov (United States)

    Williams, Ryan T.; Swanlund, Andrew; Miller, Shazia; Konstantopoulos, Spyros; Eno, Jared; van der Ploeg, Arie; Meyers, Coby

    2014-01-01

    This study operationalizes four measures of instructional differentiation: one for Grade 2 English language arts (ELA), one for Grade 2 mathematics, one for Grade 5 ELA, and one for Grade 5 mathematics. Our study evaluates their measurement properties of each measure in a large field experiment: the Indiana Diagnostic Assessment Tools Study, which…

  14. Influence of structure improvement of guide tubes and bundles in pressurized water reactor (PWR) on drop of control rods

    International Nuclear Information System (INIS)

    Shen Xiuzhong; Yu Pingan; Yang Guanyue

    1996-01-01

    In order to alleviate the cross hydraulic load on control rod guide tubes and bundles, some protective sleeves are added to those near the upper plenum outlet nozzles (4 symmetric bundles: 02-26, 03-25, 11-29, 12-28). In a 1/4 scale transparent model of the PWR upper plenum of Qinshan Nuclear Power Station, water was chosen as the fluid and hydraulic experiments with improved control rod guide tubes and bundles were carried out. The results were carefully compared with those of the experiments with unimproved control rod guide tubes and bundles. It is concluded that adding protective sleeves to the control rod guide tubes and bundles near the outlet nozzles will help to lighten the hydraulic load on them and make certain of the free movement and rapid dropping of control rods in the tubes and bundles in emergency by order

  15. Kinetics experiments and bench-scale system: Background, design, and preliminary experiments

    International Nuclear Information System (INIS)

    Rofer, C.K.

    1987-10-01

    The project, Supercritical Water Oxidation of Hazardous Chemical Waste, is a Hazardous Waste Remedial Actions Program (HAZWRAP) Research and Development task being carried out by the Los Alamos National Laboratory. Its objective is to obtain information for use in understanding the basic technology and for scaling up and applying oxidation in supercritical water as a viable process for treating a variety of DOE-DP waste streams. This report gives the background and rationale for kinetics experiments on oxidation in supercritical water being carried out as a part of this HAZWRAP Research and Development task. It discusses supercritical fluid properties and their relevance to applying this process to the destruction of hazardous wastes. An overview is given of the small emerging industry based on applications of supercritical water oxidation. Factors that could lead to additional applications are listed. Modeling studies are described as a basis for the experimental design. The report describes plug flow reactor and batch reactor systems, and presents preliminary results. 28 refs., 4 figs., 5 tabs

  16. Recent development for improving of PWR flexibility to load follow and frequency control operation

    International Nuclear Information System (INIS)

    Dubourg, M.

    1983-08-01

    In order to adjust the PWR electricity generation to the consumption network, new operating conditions were established. Those new conditions generate additional mechanical and thermal sollicitations due to the frequent motion of control rod banks, consisting of mechanical fatigue cycling and wear at the level of control rode drive mechanisms, control rods and guide tubes, wear and thermal fatigue cycling at the level of fuel assemblies. This paper presents the various aspects of this program including identification of the most critical areas of components, basic research in laboratories for resolving wear problems in PWR environment, improvement of local hydraulics for reducing loads, and endurance testing of full scale components on testing facilities

  17. PWR vessel flaw distribution development

    International Nuclear Information System (INIS)

    Rosinski, S.T.; Kennedy, E.L.; Foulds, J.R.; Kinsman, K.M.

    1990-01-01

    This paper reports on PWR pressure vessels which operate under NRC rules and regulatory guides intended to prevent failure of the vessels. Plants failing to meet the operating criteria specified under these rules and regulations are required to analytically demonstrate fitness for service in order to continue operation. The initial flaw size or distribution of initial vessel flaws is a key input to the required vessel integrity analyses. However, the flaw distribution assumed in the development of the NRC Regulations and recommended for the plant specific analyses is potentially over-conservative. This is because the distribution is based on the limited amount of vessel inspection data available at the time the criteria were being developed and does not take full advantage of the more recent and reliable domestic vessel inspection results. The U.S. Department of Energy is funding an effort through Sandia National Laboratories to investigate the possibility of developing a new flaw distribution based on the increased amount and improved reliability of domestic vessel inspection data. Results of Phase I of the program indicate that state-of-the-art NDE systems' capabilities are sufficient for development of a new flaw distribution that could ultimately provide life extension benefits over the presently required operating practice

  18. Upgrading of PWR plant simulators

    International Nuclear Information System (INIS)

    Wada, Tomonori; Sasaki, Kazunori; Nakaishi, Hirokazu.

    1989-01-01

    For the education and training of operators in electric power plants, simulators have been employed, and it is well known that their effect is great. There are operation training simulators which simulate the dynamic characteristics of plants and all the machinery and equipment that operators handle, and train the procedure of restoration at the time of abnormality in plants, education simulators which can analyze the dynamic characteristics of plants efficiently in a short time, and offer information by visualizing phenomena with three-dimensional display and others so as to be easily understandable, and forecast simulators which do the analysis forecasting plant behavior at the time of abnormality in plants, and investigate the necessity of the guide for operation procedure and the countermeasures at the time of emergency. In this explanation, the upgrading of operation training simulators which have been put already in training is discussed. The constitution of simulator system and the instructor function, the outline of PWR plant simulation models comprising thermal flow model, pump model, leak model and so on, the techniques of increasing simulator speed, and the example of analysis using the NUPAC code are reported. (K.I.)

  19. PWR secondary water chemistry study

    International Nuclear Information System (INIS)

    Pearl, W.L.; Sawochka, S.G.

    1977-02-01

    Several types of corrosion damage are currently chronic problems in PWR recirculating steam generators. One probable cause of damage is a local high concentration of an aggressive chemical even though only trace levels are present in feedwater. A wide variety of trace chemicals can find their way into feedwater, depending on the sources of condenser cooling water and the specific feedwater treatment. In February 1975, Nuclear Water and Waste Technology Corporation (NWT), was contracted to characterize secondary system water chemistry at five operating PWRs. Plants were selected to allow effects of cooling water chemistry and operating history on steam generator corrosion to be evaluated. Calvert Cliffs 1, Prairie Island 1 and 2, Surry 2, and Turkey Point 4 were monitored during the program. Results to date in the following areas are summarized: (1) plant chemistry variations during normal operation, transients, and shutdowns; (2) effects of condenser leakage on steam generator chemistry; (3) corrosion product transport during all phases of operation; (4) analytical prediction of chemistry in local areas from bulk water chemistry measurements; and (5) correlation of corrosion damage to chemistry variation

  20. Integral type small PWR with stand-alone safety

    International Nuclear Information System (INIS)

    Makihara, Yoshiaki

    2001-01-01

    A feasibility study is achieved on an integral type small PWR with stand-alone safety. It is designed to have the following features. (1) The coolant does not leak out at any accidental condition. (2) The fuel failure does never occur while it is supposed on the large scale PWR at the design base accident. (3) At any accidental condition the safety is secured without any support from the outside (stand-alone safety secure). (4) It has self-regulating characteristics and easy controllability. The above features can be satisfied by integrate the steam generator and CRDM in the reactor vessel while the pipe line break has to be considered on the conventional PWR. Several counter measures are planned to satisfy the above features. The economy feature is also attained by several simplifications such as (1) elimination of main coolant piping and pressurizer by the integration of primary cooling system and self-pressurizing, (2) elimination of RCP by application of natural circulating system, (3) elimination of ECCS and accumulator by application of static safety system, (4) large scale volume reduction of the container vessel by application of integrated primary cooling system, (5) elimination of boric acid treatment by deletion of chemical shim. The long operation period such as 10 years can be attained by the application of Gd fuel in one batch refueling. The construction period can be shortened by the standardizing the design and the introduction of modular component system. Furthermore the applicability of the reduced modulation core is also considered. (K. Tsuchihashi)

  1. Experiments in a real scale maglev vehicle prototype

    Energy Technology Data Exchange (ETDEWEB)

    Sotelo, G G; Stephan, R M [Department of Electrical Engineering - Federal University of Rio de Janeiro. Address: CT, Bl. I-2000 Sala 148, Cidade Universitaria, Rio de Janeiro. PO-BOX: 68553, CEP.: 21941 - 972 (Brazil); Costa, G C [Universidade Estadual da Zona Oeste (Brazil); Dias, D H N; Machado, O J; David, E D; Andrade, R de Jr, E-mail: sotelo@coe.ufrj.b

    2010-06-01

    A Brazilian real scale magnetically levitated transport system prototype is under development at the Federal University of Rio de Janeiro. To test this system a 180 m long line has been projected and it will be concluded by the end of 2010. A superconducting linear bearing (SLB) is used to replace the wheels of a conventional train. High temperature superconductor bulks placed inside cryostats attached to the vehicle and a magnetic rail composes the SLB. To choose the magnetic rail for the test line three different rails, selected in a previous simulation work, were built and tested. They are composed by Nd-Fe-B and steel, arranged in a flux concentrator topology. The magnetic flux density for those magnetic rails was mapped. Also, the levitation force between those rails and the superconductor cryostat, for several cooling gaps, were measured to select the best rail geometry to be used in the real scale line. The SLB allows building a light vehicle with distributed load, silent and high energy efficient. The proposed vehicle is composed of four modules with just 1.5 m of length each one and it can transport up to 24 passengers. The test line having two curves with 45 m radius and a 15% acclivity ramp is also presented.

  2. Scaling NS-3 DCE Experiments on Multi-Core Servers

    Science.gov (United States)

    2016-06-15

    MPTCP) using the same software in DCE. In the experiment, only two wireless links ( LTE and Wi-Fi) are setup to examine MPTCP, resulting in limited...performance drop on the blade server. Our investigation then turned to other straightforward measures including the following: • We reduced the amount...simulation with varying numbers of cores and measured the run time. To pin the simulation to a specific set of cores, we switched from using 0:00   2:00

  3. Large-scale experience with biological treatment of contaminated soil

    International Nuclear Information System (INIS)

    Schulz-Berendt, V.; Poetzsch, E.

    1995-01-01

    The efficiency of biological methods for the cleanup of soil contaminated with total petroleum hydrocarbons (TPH) and polycyclic aromatic hydrocarbons (PAH) was demonstrated by a large-scale example in which 38,000 tons of TPH- and PAH-polluted soil was treated onsite with the TERRAFERM reg-sign degradation system to reach the target values of 300 mg/kg TPH and 5 mg/kg PAH. Detection of the ecotoxicological potential (Microtox reg-sign assay) showed a significant decrease during the remediation. Low concentrations of PAH in the ground were treated by an in situ technology. The in situ treatment was combined with mechanical measures (slurry wall) to prevent the contamination from dispersing from the site

  4. Rasch scaling paranormal belief and experience: structure and semantics of Thalbourne's Australian Sheep-Goat Scale.

    Science.gov (United States)

    Lange, Rense; Thalbourne, Michael A

    2002-12-01

    Research on the relation between demographic variables and paranormal belief remains controversial given the possible semantic distortions introduced by item and test level biases. We illustrate how Rasch scaling can be used to detect such biases and to quantify their effects, using the Australian Sheep-Goal Scale as a substantive example. Based on data from 1.822 respondents, this test was Rasch scalable, reliable, and unbiased at the test level. Consistent with other research in which unbiased measures of paranormal belief were used, extremely weak age and sex effects were found (partial eta2 = .005 and .012, respectively).

  5. Improvement in PWR flexibility the french program 1975-1995

    International Nuclear Information System (INIS)

    Gautier, A.; Miossec, C.

    1985-12-01

    Between 1975 and 1985, a substantial effort was launched in France to greatly improve PWR's flexibility, resulting in the current situation where both frequency control and load follow are now routinely performed on most plants in operation. Based on rapidly accumulating operational experience and on all expertise acquired in the past decade, a second-generation core control strategy is now being finalized for application on all future 1400 MW plants (with commercial operation scheduled in 1992 for first unit). This 20-year program is discussed

  6. An evaluation of the Positive Emotional Experiences Scale: A preliminary analysis

    Directory of Open Access Journals (Sweden)

    Rene van Wyk

    2016-11-01

    Full Text Available Orientation: The positive organisational behaviour movement emphasises the advantages of psychological strengths in business. The psychological virtues of positive emotional experiences can potentially promote human strengths to the advantages of business functioning and the management of work conditions. This is supported by Fredrickson’s broaden-and-build theory that emphasises the broadening of reactive thought patterns through experiences of positive emotions. Research purpose: A preliminary psychometric evaluation of a positive measurement of dimensions of emotional experiences in the workplace, by rephrasing the Kiefer and Barclay Toxic Emotional Experiences Scale. Motivation for the study: This quantitative Exploratory Factor Analysis investigates the factorial structure and reliability of the Positive Emotional Experiences Scale, a positive rephrased version of the Toxic Emotional Experiences Scale. Research approach, design and method: This Exploratory Factor Analysis indicates an acceptable three-factor model for the Positive Emotional Experiences Scale. These three factors are: (1 psychological recurrent positive state, (2 social connectedness and (3 physical refreshed energy, with strong Cronbach’s alphas of 0.91, 0.91 and 0.94, respectively. Main findings: The three-factor model of the Positive Emotional Experiences Scale provides a valid measure in support of Fredrickson’s theory of social, physical and psychological endured personal resources that build positive emotions. Practical/Managerial implications: Knowledge gained on positive versus negative emotional experiences could be applied by management to promote endured personal resources that strengthen positive emotional experiences. Contribution/value-add: The contribution of this rephrased Positive Emotional Experiences Scale provides a reliable measure of assessment of the social, physical and endured psychological and personal resources identified in Fredrickson

  7. Preliminary safety analysis of the PWR with accident-tolerant fuels during severe accident conditions

    International Nuclear Information System (INIS)

    Wu, Xiaoli; Li, Wei; Wang, Yang; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng; Liu, Tong; Deng, Yongjun; Huang, Heng

    2015-01-01

    Highlights: • Analysis of severe accident scenarios for a PWR fueled with ATF system is performed. • A large-break LOCA without ECCS is analyzed for the PWR fueled with ATF system. • Extended SBO cases are discussed for the PWR fueled with ATF system. • The accident-tolerance of ATF system for application in PWR is illustrated. - Abstract: Experience gained in decades of nuclear safety research and previous nuclear accidents direct to the investigation of passive safety system design and accident-tolerant fuel (ATF) system which is now becoming a hot research point in the nuclear energy field. The ATF system is aimed at upgrading safety characteristics of the nuclear fuel and cladding in a reactor core where active cooling has been lost, and is preferable or comparable to the current UO 2 –Zr system when the reactor is in normal operation. By virtue of advanced materials with improved properties, the ATF system will obviously slow down the progression of accidents, allowing wider margin of time for the mitigation measures to work. Specifically, the simulation and analysis of a large break loss of coolant accident (LBLOCA) without ECCS and extended station blackout (SBO) severe accident are performed for a pressurized water reactor (PWR) loaded with ATF candidates, to reflect the accident-tolerance of ATF

  8. Experimental investigation of stratified two-phase flows in the hot leg of a PWR for CFD validation

    Energy Technology Data Exchange (ETDEWEB)

    Vallee, Christophe; Lucas, Dirk [Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany). Inst. of Fluid Dynamics; Tomiyama, Akio [Kobe Univ (Japan). Graduate School of Engineering; Murase, Michio [Institute of Nuclear Safety System, Inc. (INSS), Fukui (Japan)

    2012-12-15

    Stratified 2-phase flows were investigated in 2 different models of the hot leg of a pressurised water reactor (PWR) in order to provide experimental data for the development and validation of computational fluid dynamics (CFD) codes. Therefore, the local flow structure was visualised with a high-speed video camera. Moreover, one test section was designed with a rectangular cross-section to achieve optimal observation conditions. The phenomenon of counter-current flow limitation (CCFL) was investigated, which may affect the reflux condenser cooling mode in some accident scenarios. The experiments were conducted with air and water at room temperature and maximum pressures of 3 bar as well as with steam and saturated water at boundary conditions of up to 50 bar and 264 C. The measured CCFL characteristics were compared with similar experimental data and correlations available in the literature. This shows that the channel height is the characteristic length to be used in the Wallis parameter for channels with rectangular cross-sections. Furthermore, the experimental results confirm that the Wallis similarity is appropriate to scale CCFL in the hot leg of a PWR over a wide range of pressure and temperature conditions. Finally, an image processing algorithm was developed to recognise the stratified interface in the camera frames. Subsequently, the interfacial structure along the hot leg was visualised by the representation of the probability distribution of the water level. (orig.)

  9. Experience Scaling Up Manufacturing of Emerging Photovoltaic Technologies

    Energy Technology Data Exchange (ETDEWEB)

    Braun, G. W.; Skinner, D. E.

    2007-01-01

    This report examines two important generic photovoltaic technologies at particularly revealing stages of development, i.e., the stages between R&D and stable commercial production and profitable sales. Based on two historical cases, it attempts to shed light on the difference between: (1) costs and schedules validated by actual manufacturing and market experience, and (2) estimated costs and schedules that rely on technology forecasts and engineering estimates. The amorphous Silicon case also identifies some of the costs that are incurred in meeting specific market requirements, while the Cadmium Telluride case identifies many of the operational challenges involved in transferring R&D results to production. The transition between R&D and commercial success takes a great deal of time and money for emerging energy conversion technologies in general. The experience reported here can be instructive to those managing comparable efforts, and to their investors. It can also be instructive to R&D managers responsible for positioning such new technologies for commercial success.

  10. Experience gained in bench scale and pilot scale fluidised bed processing

    CSIR Research Space (South Africa)

    Hadley, TD

    2005-01-01

    Full Text Available of titanium dioxide. Expertise in the design and commissioning of industrial-scale plants has led to the supply (through licensees) of a biomass sludge incinerator/boiler generating 26t/h steam, a 20 MW high-sulphur pitch incinerator and a 12 MW fluidised bed...

  11. Development and examination of the psychometric properties of the Learning Experience Scale in nursing.

    Science.gov (United States)

    Takase, Miyuki; Imai, Takiko; Uemura, Chizuru

    2016-06-01

    This paper examines the psychometric properties of the Learning Experience Scale. A survey method was used to collect data from a total of 502 nurses. Data were analyzed by factor analysis and the known-groups technique to examine the construct validity of the scale. In addition, internal consistency was evaluated by Cronbach's alpha, and stability was examined by test-retest correlation. Factor analysis showed that the Learning Experience Scale consisted of five factors: learning from practice, others, training, feedback, and reflection. The scale also had the power to discriminate between nurses with high and low levels of nursing competence. The internal consistency and the stability of the scale were also acceptable. The Learning Experience Scale is a valid and reliable instrument, and helps organizations to effectively design learning interventions for nurses. © 2015 Wiley Publishing Asia Pty Ltd.

  12. Mechanistically-Based Field-Scale Models of Uranium Biogeochemistry from Upscaling Pore-Scale Experiments and Models

    International Nuclear Information System (INIS)

    Tim Scheibe; Alexandre Tartakovsky; Brian Wood; Joe Seymour

    2007-01-01

    Effective environmental management of DOE sites requires reliable prediction of reactive transport phenomena. A central issue in prediction of subsurface reactive transport is the impact of multiscale physical, chemical, and biological heterogeneity. Heterogeneity manifests itself through incomplete mixing of reactants at scales below those at which concentrations are explicitly defined (i.e., the numerical grid scale). This results in a mismatch between simulated reaction processes (formulated in terms of average concentrations) and actual processes (controlled by local concentrations). At the field scale, this results in apparent scale-dependence of model parameters and inability to utilize laboratory parameters in field models. Accordingly, most field modeling efforts are restricted to empirical estimation of model parameters by fitting to field observations, which renders extrapolation of model predictions beyond fitted conditions unreliable. The objective of this project is to develop a theoretical and computational framework for (1) connecting models of coupled reactive transport from pore-scale processes to field-scale bioremediation through a hierarchy of models that maintain crucial information from the smaller scales at the larger scales; and (2) quantifying the uncertainty that is introduced by both the upscaling process and uncertainty in physical parameters. One of the challenges of addressing scale-dependent effects of coupled processes in heterogeneous porous media is the problem-specificity of solutions. Much effort has been aimed at developing generalized scaling laws or theories, but these require restrictive assumptions that render them ineffective in many real problems. We propose instead an approach that applies physical and numerical experiments at small scales (specifically the pore scale) to a selected model system in order to identify the scaling approach appropriate to that type of problem. Although the results of such studies will

  13. Mechanistically-Based Field-Scale Models of Uranium Biogeochemistry from Upscaling Pore-Scale Experiments and Models

    Energy Technology Data Exchange (ETDEWEB)

    Tim Scheibe; Alexandre Tartakovsky; Brian Wood; Joe Seymour

    2007-04-19

    Effective environmental management of DOE sites requires reliable prediction of reactive transport phenomena. A central issue in prediction of subsurface reactive transport is the impact of multiscale physical, chemical, and biological heterogeneity. Heterogeneity manifests itself through incomplete mixing of reactants at scales below those at which concentrations are explicitly defined (i.e., the numerical grid scale). This results in a mismatch between simulated reaction processes (formulated in terms of average concentrations) and actual processes (controlled by local concentrations). At the field scale, this results in apparent scale-dependence of model parameters and inability to utilize laboratory parameters in field models. Accordingly, most field modeling efforts are restricted to empirical estimation of model parameters by fitting to field observations, which renders extrapolation of model predictions beyond fitted conditions unreliable. The objective of this project is to develop a theoretical and computational framework for (1) connecting models of coupled reactive transport from pore-scale processes to field-scale bioremediation through a hierarchy of models that maintain crucial information from the smaller scales at the larger scales; and (2) quantifying the uncertainty that is introduced by both the upscaling process and uncertainty in physical parameters. One of the challenges of addressing scale-dependent effects of coupled processes in heterogeneous porous media is the problem-specificity of solutions. Much effort has been aimed at developing generalized scaling laws or theories, but these require restrictive assumptions that render them ineffective in many real problems. We propose instead an approach that applies physical and numerical experiments at small scales (specifically the pore scale) to a selected model system in order to identify the scaling approach appropriate to that type of problem. Although the results of such studies will

  14. Kinetic Interpretation of Nitrogen Removal in Pilot Scale Experiments

    DEFF Research Database (Denmark)

    Harremoës, Poul; Sinkjær, Ole

    1995-01-01

    with biological and chemical phosphorus removal. Nitrification and denitrification rates have been measured in batch tests on activated sludge extracted from the pilot plants and by measuring transient concentrations during the alternating mode of operation in the aerobic and anoxic tanks. The data were......Pilot plant experiments have been performed over a period of four years in order to establish an experimental basis for the upgrading of the treatment plants of The City of Copenhagen to nutrient removal. The choice of design is the alternating mode of operating biological nitrogen removal...... normalized to standard conditions by correcting them according to the kinetic theory. The average normalized nitrification rate was measured to be between 54 and 60 mg NH~-N/(g VSSn~t' h) by different test methods at 7°C. The denitrification rate was measured to vary between 0.85 and 0.95mg NO~--N/(g VSS. h...

  15. Scaled beam merging experiment for heavy ion inertial fusion

    Directory of Open Access Journals (Sweden)

    P. A. Seidl

    2003-09-01

    Full Text Available Transverse beam combining is a cost-saving option employed in many designs for heavy ion fusion drivers. However, the resultant transverse phase space dilution must be minimized so as not to sacrifice focusability at the target. A prototype combining experiment has been completed employing four 3-mA Cs^{+} beams injected at 160 keV. The focusing elements upstream of the merge consist of four quadrupoles and a final combined-function element (quadrupole and dipole. Following the merge, the resultant single beam is transported in a single alternating gradient channel where the subsequent evolution of the distribution function is diagnosed. The results are in fair agreement with particle-in-cell simulations. They indicate that for some heavy ion fusion driver designs, the phase space dilution from merging is acceptable.

  16. Gas explosion characterization, wave propagation (small scale experiments)

    International Nuclear Information System (INIS)

    Larsen, G.C.

    1985-08-01

    A number of experiments have been performed with blast waves arising from the ignition of homogeneous and well defined mixtures of methane, oxygen and nitrogen, contained within spherical balloons with controlled initial dimensions. The pressure characteristics has been studied for blast waves with and without influence from reflected waves. The influence of obstacles in the flow field has also been treated. Both configuration with one box and two closely spaced boxes have been considered, and a wave-wave interaction phenomenon was observed in the case of closely spaced obstacles. Moreover reflection coefficients have been established and some pressure variations over the surfaces have been observed. An acoustic appriximation has been used to model the blast wave originating from an expanding sphere. It has been demonstrated, that the generated pressure pulse is very sensitive to the expansion rate. Calculated and measured data have been compared, and a reasonable agreement has been found. (author)

  17. Pilot scale experiments on radiation vulcanization of NR latex

    Science.gov (United States)

    Ridwan, M.

    The potential of irradiated latex as raw material of commercial use is under testing on pilot plant scale in Indonesia which has 225 kCi Co-60 irradiation facility and can irradiate 1000 tonnes of centrifuged latex per annum. The facility was jointly designed by BATAN of Indonesia and JAERI of Japan and was jointly financed by UNDP/IAEA, Government of Japan and Government of Indonesia under UNDP/IAEA Regional Cooperative Agreement Project on Industrial Application of Isotopes and Radiation Technology. The facility is a water pool type and can accomodate 400 kCi Co-60. The Co-60 rack has two shapes, plate and cylindrical shapes. The plate shape source is used for natural rubber latex irradiation and the cylindrical one is used for other irradiation services. The vulcanization system consists of three major components : emulsification unit ( height : 650 mm, diameter 500 mm ), mixing unit ( height : 1900mm, diameter 1200 mm ) and vulcanization reactor ( height : 1800 mm, diameter 1300 mm ). The first two components are located outside shielded room while the third one-in irradiation room. The radiation vulcanization process is a much simpler energy saving process comparedto the conventional thermal process which has two vulcanization steps before and after dipping. The physical and mechanical properties of irradiated NR Latex are comparable to those of sulfur vulcanized, and depend on many factors such as irradiation dose, sensitizer content, dry rubber content and storage time.

  18. Pilot scale experiments on radiation vulcanization of NR latex

    International Nuclear Information System (INIS)

    Ridwan, M.

    1985-01-01

    The potential of irradiated latex as raw material of commercial use is under testing on pilot plant scale in Indonesia which has 225 kCi Co-60 irradiation facility and can irradiate 1000 tonnes of centrifuged latex per annum. The facility was jointly designed by BATAN of Indonesia and JAERI of Japan and was jointly financed by UNDP/IAEA, Government of Japan and Government of Indonesia under UNDP/IAEA Regional Cooperative Agreement Project on Industrial Application of Isotopes and Radiation Technology. The facility is a water pool type and can accommodate 400 kCi Co-60. The Co-60 rack has two shapes, plate and cylindrical shapes. The plate shape source is used for natural rubber latex irradiation and the cylindrical one is used for other irradiation services. The vulcanization system consists of three major components: emulsification unit, mixing unit and vulcanization reactor. The first two components are located outside shielded room while the third one in irradiation room. The radiation vulcanization process is a much simpler energy saving process compared to the conventional thermal process which has two vulcanization steps before and after dipping. The physical and mechanical properties of irradiated NR latex are comparable to those of sulfur vulcanized. (author)

  19. Comprehensive exergetic and economic comparison of PWR and hybrid fossil fuel-PWR power plants

    International Nuclear Information System (INIS)

    Sayyaadi, Hoseyn; Sabzaligol, Tooraj

    2010-01-01

    A typical 1000 MW Pressurized Water Reactor (PWR) nuclear power plant and two similar hybrid 1000 MW PWR plants operate with natural gas and coal fired fossil fuel superheater-economizers (Hybrid PWR-Fossil fuel plants) are compared exergetically and economically. Comparison is performed based on energetic and economic features of three systems. In order to compare system at their optimum operating point, three workable base case systems including the conventional PWR, and gas and coal fired hybrid PWR-Fossil fuel power plants considered and optimized in exergetic and exergoeconomic optimization scenarios, separately. The thermodynamic modeling of three systems is performed based on energy and exergy analyses, while an economic model is developed according to the exergoeconomic analysis and Total Revenue Requirement (TRR) method. The objective functions based on exergetic and exergoeconomic analyses are developed. The exergetic and exergoeconomic optimizations are performed using the Genetic Algorithm (GA). Energetic and economic features of exergetic and exergoeconomic optimized conventional PWR and gas and coal fired Hybrid PWR-Fossil fuel power plants are compared and discussed comprehensively.

  20. Analytical technical of lightning surges induced on grounding mesh of PWR nuclear power plant

    International Nuclear Information System (INIS)

    Ikeda, I.; Tani, M.; Yonezawa, T.

    1990-01-01

    An analytical lightning surge technique is needed to make a qualitative and predictive evaluation of transient voltages induced on local grounding meshes and instrumentation cables by a lightning strike on a lightning rod in a PWR plant. This paper discusses an experiment with lightning surge impulses in a PWR plant which was setup to observe lightning caused transient voltages. Experimental data when compared with EMTP simulation results improved the simulation method. The improved method provides a good estimation of induced voltages on grounding meshes and instrumentation cables

  1. Analysis of the alternatives for the chemical treatment of the secondary circuit of PWR power plants

    International Nuclear Information System (INIS)

    Lopes, J.P.G.; Silva Neto, A.J. da; Braganca Junior, A.; Dominguez, D.

    1990-01-01

    The operational experiences within PWR power plants shows that the major problems which affect the plant availability occurs in the secondary side, mainly in the steam generators and condenser. The aim of this report is to perform an evaluation of the main chemical treatment processes, which are applied to the secondary side of PWR power plants in order to reduce the corrosion problems to which are exposed the system equipment, minimizing in this way the shut down and maintenance cost for repairs and replacement of damaged components. (author)

  2. French experience in transient data collection and fatigue monitoring of PWR`s nuclear steam supply system; Experience francaise sur la comptabilisation des transitoires et la surveillance en fatigue des chaudieres REP

    Energy Technology Data Exchange (ETDEWEB)

    Sabaton, M.; Morilhat, P.; Savoldelli, D.; Genette, P.

    1995-10-01

    Electricite de France (EDF), the french national electricity company, is operating 54 standardized pressurizer water reactors. This about 500 reactor-years experience in nuclear stations operation and maintenance area has allowed EDF to develop its own strategy for monitoring of age-related degradations of NPP systems and components relevant for plant safety and reliability. After more than fifteen years of experience in regulatory transient data collection and seven years of successful fatigue monitoring prototypes experimentation, EDF decided to design a new system called SYSFAC (acronym for SYsteme de Surveillance en FAtigue de la Chaudiere) devoted to transient logging and thermal fatigue monitoring of the reactor coolant pressure boundary. The system is fully automatic and directly connected to the on-site data acquisition network without any complementary instrumentation. A functional transient detection module and a mechanical transient detection module are in charge of the general transient data collection. A fatigue monitoring module is aimed towards a precise surveillance of five specific zones particularly sensible to thermal fatigue. After the first step of preliminary studies, the industrial phase of the SYSFAC project is currently going on, with hardware and software tests and implementation. The first SYSFAC system will be delivered to the pilot power plant by the beginning of 1996. The extension to all EDF`s nuclear 900 MW is planned after one more year of feedback experience. (authors). 12 refs., 3 figs.

  3. Application of burnup credit for PWR spent fuel storage pool

    International Nuclear Information System (INIS)

    Shin, Hee Sung; Ro, Seung-Gy; Bae, Kang Mok; Kim, Ik Soo; Shin, Young Joon

    1999-01-01

    A study on the application of burnup credit for a PWR spent fuel storage pool has been investigated using a computer code system such as CSAS6 module of SCALE 4.3 in association with 44-group SCALE cross-section library. The calculation bias of the code system at a 95% probability with a 95% confidence level seems to be 0.00951 by benchmarking the system for forty six experimental data. With the aid of this computer code system, criticality analysis has been performed for the PWR spent fuel storage pool. Uncertainties due to postulated abnormal and accidental conditions, and manufacturing tolerance such as stainless steel thickness of storage rack, fuel enrichment, fuel density and box size have statistically been combined and resulted in 0.00674. Also, isotopic correction factor which was based on the calculated and measured concentration of 43 isotopes for both selected actinides and fission products important in burnup credit application has been taken into account in the criticality analysis. It is revealed that the minimum burnup with the corrected isotopic concentrations as required for the safe storage is 5,730 MWd/tU in enriched fuel of 5.0 wt%. (author)

  4. Parallel GPU implementation of PWR reactor burnup

    International Nuclear Information System (INIS)

    Heimlich, A.; Silva, F.C.; Martinez, A.S.

    2016-01-01

    Highlights: • Three GPU algorithms used to evaluate the burn-up in a PWR reactor. • Exhibit speed improvement exceeding 200 times over the sequential. • The C++ container is expansible to accept new nuclides chains. - Abstract: This paper surveys three methods, implemented for multi-core CPU and graphic processor unit (GPU), to evaluate the fuel burn-up in a pressurized light water nuclear reactor (PWR) using the solutions of a large system of coupled ordinary differential equations. The reactor physics simulation of a PWR reactor spends a long execution time with burnup calculations, so performance improvement using GPU can imply in better core design and thus extended fuel life cycle. The results of this study exhibit speed improvement exceeding 200 times over the sequential solver, within 1% accuracy.

  5. Large Scale Experiments of Multihop Networks in Mobile Scenarios

    Directory of Open Access Journals (Sweden)

    Yacine Benchaïb

    2016-03-01

    Full Text Available This paper presents the latest advances in our research work focused on VIRMANEL and SILUMOD, a couple of tools developed for research in wireless mobile multihop networks. SILUMOD is a domain specific language dedicated to the definition of mobility models. This language contains key- words and special operators that make it easy to define a mobility model and calculate the positions of a trajectory. These positions are sent to VIRMANEL, a tool that man- ages virtual machines corresponding to mobile nodes, emu- lates their movements and the resulting connections and dis- connections, and displays the network evolution to the user, thanks to its graphical user interface. The virtualization ap- proach we take here allows to run real code and to test real protocol implementations without deploying an important experimental platform. For the experimentation of a large number of virtual mobile nodes, we defined and implemented a new algorithm for the nearest neighbor search to find the nodes that are within communication range. We then car- ried out a considerable measurement campaign in order to evaluate the performance of this algorithm. The results show that even with an experiment using a large number of mobile nodes, our algorithm make it possible to evaluate the state of connectivity between mobile nodes within a reasonable time and number of operations.

  6. Emergency response preparedness: the French experience of large scale exercises

    International Nuclear Information System (INIS)

    Chanson, D.; Desnoyers, B.; Chabane, J.M.

    2004-01-01

    In compliance with the IAEA regulations for the transport of radioactive material in the event of accidents during transport of radioactive material, emergency provisions to protect persons, property and environment have to be established and developed by the relevant national organisations. In France, the prefect of the department where the accident occurs is responsible for decisions and measures required to ensure the protection of both population and property at risk owing to the accident. During an accident, the ministers concerned provide the prefect with recommendations and information, in order to help him take the requisite decisions. On their side, the nuclear industry and transport companies also have to be prepared to intervene and to support the authorities at their request, depending on their capacities and their specialities. To prepare the emergency teams properly and acquire effective emergency plans, training exercises have to be conducted regularly with every ministerial department involved, the nuclear industry and transport companies, members of the public and the media. Then, the feedback from such exercises shall be taken into account to improve the emergency procedures. This paper will introduce: - emergency response preparedness: what is required by the relevant regulations? - emergency response preparedness: how is France organised? - the French experience of conducting large training exercises simulating accidents involving the transport of radioactive material; - the main difficulties and lessons learned; - the perspectives

  7. Transferability of results of small scale experiments to real structures

    International Nuclear Information System (INIS)

    Schmitt, W.; Siegele, D.; Kordisch, H.; Baudendistel, E.

    1983-01-01

    The good agreement of the experimental J-values and the numerical J confirms the experimental procedure to evaluate J from the work done on the specimen. This is important for the application of single specimen techniques. The next logical steps in the chain of transferability will now be - after the verification of the three-dimensional crack growth calculations - the experimental and numerical analysis of configuration closer to real structures, e.g. part-through surface flaws in plates and pipes. Starting from a semi-elliptical fatigue flaw the dark regime of stable tearing does no longer follow the original elliptical shape. The explanation of this behavior can only be expected if all possible three-dimensional effects are taken into account. Those three-dimensional effects are also apparent even in compact specimens if no sidegrooves are used. In a first test of the three-dimensional crack growth capabilities of the IWM version of ADINA this experiment has been simulated and loaded up to a displacement of about half the final displacement in the test. (orig./RW)

  8. Preliminary study on direct recycling of spent PWR fuel in PWR system

    International Nuclear Information System (INIS)

    Waris, Abdul; Nuha; Novitriana; Kurniadi, Rizal; Su'ud, Zaki

    2012-01-01

    Preliminary study on direct recycling of PWR spent fuel to support SUPEL (Straight Utilization of sPEnt LWR fuel in LWR system) scenario has been conducted. Several spent PWR fuel compositions in loaded PWR fuel has been evaluated to obtain the criticality of reactor. The reactor can achieve it criticality for U-235 enrichment in the loaded fresh fuel is at least 4.0 a% with the minimum fraction of the spent fuel in the core is 15.0 %. The neutron spectra become harder with the escalating of U-235 enrichment in the loaded fresh fuel as well as the amount of the spent fuel in the core.

  9. Fuel assemblies for PWR type reactors: fuel rods, fuel plates. CEA work presentation

    International Nuclear Information System (INIS)

    Delafosse, Jacques.

    1976-01-01

    French work on PWR type reactors is reported: basic knowledge on Zr and its alloys and on uranium oxide; experience gained on other programs (fast neutron and heavy water reactors); zircaloy-2 or zircaloy-4 clad UO 2 fuel rods; fuel plates consisting of zircaloy-2 clad UO 2 squares of thickness varying between 2 and 4mm [fr

  10. A simulated test of physical starting and reactor physics on zero power facility of PWR

    International Nuclear Information System (INIS)

    Yao Zewu; Ji Huaxiang; Chen Zhicheng; Yao Zhiquan; Chen Chen; Li Yuwen

    1995-01-01

    The core neutron economics has been verified through experiments conducted at a zero power reactor with baffles of various thickness. A simulated test of physical starting of Qinshan PWR has been introduced. The feasibility and safety of the programme are verified. The research provides a valuable foundation for developing physical starting programme

  11. Atmea launches Atmea1 the mid-sized generation 3+ PWR you can rely on

    International Nuclear Information System (INIS)

    2008-01-01

    ATMEA, a daughter company of AREVA NP and Mitsubishi Heavy Industries, is developing and will supply ATMEA1, the most advanced 1100 MWe PWR plant with the combination of the unique set of competence and experience of its parent companies. This folder presents the ATMEA1 reactor main features. (J.S.)

  12. Fully predictive simulation of real-scale cable tray fire based on small-scale laboratory experiments

    Energy Technology Data Exchange (ETDEWEB)

    Beji, Tarek; Merci, Bart [Ghent Univ. (Belgium). Dept. of Flow, Heat and Combustion Mechanics; Bonte, Frederick [Bel V, Brussels (Belgium)

    2015-12-15

    This paper presents a computational fluid dynamics (CFD)-based modelling strategy for real-scale cable tray fires. The challenge was to perform fully predictive simulations (that could be called 'blind' simulations) using solely information from laboratory-scale experiments, in addition to the geometrical arrangement of the cables. The results of the latter experiments were used (1) to construct the fuel molecule and the chemical reaction for combustion, and (2) to estimate the overall pyrolysis and burning behaviour. More particularly, the strategy regarding the second point consists of adopting a surface-based pyrolysis model. Since the burning behaviour of each cable could not be tracked individually (due to computational constraints), 'groups' of cables were modelled with an overall cable surface area equal to the actual value. The results obtained for one large-scale test (a stack of five horizontal trays) are quite encouraging, especially for the peak Heat Release Rate (HRR) that was predicted with a relative deviation of 3 %. The time to reach the peak is however overestimated by 4.7 min (i.e. 94 %). Also, the fire duration is overestimated by 5 min (i.e. 24 %). These discrepancies are mainly attributed to differences in the HRRPUA (heat release rate per unit area) profiles between the small-scale and large-scale. The latter was calculated by estimating the burning area of cables using video fire analysis (VFA).

  13. SCALES: SEVIRI and GERB CaL/VaL area for large-scale field experiments

    Science.gov (United States)

    Lopez-Baeza, Ernesto; Belda, Fernando; Bodas, Alejandro; Crommelynck, Dominique; Dewitte, Steven; Domenech, Carlos; Gimeno, Jaume F.; Harries, John E.; Jorge Sanchez, Joan; Pineda, Nicolau; Pino, David; Rius, Antonio; Saleh, Kauzar; Tarruella, Ramon; Velazquez, Almudena

    2004-02-01

    The main objective of the SCALES Project is to exploit the unique opportunity offered by the recent launch of the first European METEOSAT Second Generation geostationary satellite (MSG-1) to generate and validate new radiation budget and cloud products provided by the GERB (Geostationary Earth Radiation Budget) instrument. SCALES" specific objectives are: (i) definition and characterization of a large reasonably homogeneous area compatible to GERB pixel size (around 50 x 50 km2), (ii) validation of GERB TOA radiances and fluxes derived by means of angular distribution models, (iii) development of algorithms to estimate surface net radiation from GERB TOA measurements, and (iv) development of accurate methodologies to measure radiation flux divergence and analyze its influence on the thermal regime and dynamics of the atmosphere, also using GERB data. SCALES is highly innovative: it focuses on a new and unique space instrument and develops a new specific validation methodology for low resolution sensors that is based on the use of a robust reference meteorological station (Valencia Anchor Station) around which 3D high resolution meteorological fields are obtained from the MM5 Meteorological Model. During the 1st GERB Ground Validation Campaign (18th-24th June, 2003), CERES instruments on Aqua and Terra provided additional radiance measurements to support validation efforts. CERES instruments operated in the PAPS mode (Programmable Azimuth Plane Scanning) focusing the station. Ground measurements were taken by lidar, sun photometer, GPS precipitable water content, radiosounding ascents, Anchor Station operational meteorological measurements at 2m and 15m., 4 radiation components at 2m, and mobile stations to characterize a large area. In addition, measurements during LANDSAT overpasses on June 14th and 30th were also performed. These activities were carried out within the GIST (GERB International Science Team) framework, during GERB Commissioning Period.

  14. A large scale field experiment in the Amazon Basin (Lambada/Bateristca)

    Energy Technology Data Exchange (ETDEWEB)

    Dolman, A.J.; Kabat, P.; Gash, J.H.C.; Noilhan, J.; Jochum, A.M.; Nobre, C. [Winand Staring Centre, Wageningen (Netherlands)

    1994-12-31

    A description is given of a large scale field experiment planned in the Amazon Basin, aiming to assess the large scale balances of energy, water and CO{sub 2}. The background for this experiment, the embedding in global change programmes of IGBP/BAHC and WCRP/GEWEX is described. A proposal by four European groups aimed at designing the experiment with the help of mesoscale models is described and a possible European input to this experiment is suggested. 24 refs., 1 app.

  15. Using a Virtual Experiment to Analyze Infiltration Process from Point to Grid-cell Size Scale

    Science.gov (United States)

    Barrios, M. I.

    2013-12-01

    The hydrological science requires the emergence of a consistent theoretical corpus driving the relationships between dominant physical processes at different spatial and temporal scales. However, the strong spatial heterogeneities and non-linearities of these processes make difficult the development of multiscale conceptualizations. Therefore, scaling understanding is a key issue to advance this science. This work is focused on the use of virtual experiments to address the scaling of vertical infiltration from a physically based model at point scale to a simplified physically meaningful modeling approach at grid-cell scale. Numerical simulations have the advantage of deal with a wide range of boundary and initial conditions against field experimentation. The aim of the work was to show the utility of numerical simulations to discover relationships between the hydrological parameters at both scales, and to use this synthetic experience as a media to teach the complex nature of this hydrological process. The Green-Ampt model was used to represent vertical infiltration at point scale; and a conceptual storage model was employed to simulate the infiltration process at the grid-cell scale. Lognormal and beta probability distribution functions were assumed to represent the heterogeneity of soil hydraulic parameters at point scale. The linkages between point scale parameters and the grid-cell scale parameters were established by inverse simulations based on the mass balance equation and the averaging of the flow at the point scale. Results have shown numerical stability issues for particular conditions and have revealed the complex nature of the non-linear relationships between models' parameters at both scales and indicate that the parameterization of point scale processes at the coarser scale is governed by the amplification of non-linear effects. The findings of these simulations have been used by the students to identify potential research questions on scale issues

  16. Load-following operation of PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Jong Hwa; Oh, Soo Yul; Koo, Yang Hyun; Lee, Jae Han [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-12-01

    The load-following operation of nuclear power plants will become inevitable due to the increased nuclear share in the total electricity generation. As a groundwork for the load-following capability of the Korean next generation PWRs, the state-of-the-art has been reviewed. The core control principles and methods are the main subject in this review as well as the impact of load-following operations on the fuel performance and on the mechanical integrity of components. To begin with, it was described what the load-following operation is and in what view point the technology should be reviewed. Afterwards the load-following method, performance and problems in domestic 900 MWe class PWRs were discussed, and domestic R and D works were summarized. Foreign technologies were also reviewed. They include Mode G and Mode X of Foratom, D and L bank method of KWU, the method using PSCEA of ABB-CE, and MSHIM of Westinghouse. The load-following related special features of Foratom`s N4 plant, KWU`s plants, ABB-CE`s Systems 80+, and Westinghouse`s AP600 were described in each technology review. The review concluded that the capability of N4 plant with Mode X is the best and the methods in System, 80+ and AP600 would require verifications for the continued and usual load-following operation. It was recommended that the load-following operation experiences in domestic PWRs under operation be required to settle down the capability for the future. In addition, a more enhanced technology is required for the Korean next generation PWR regardless what the reference plant concept is. 30 figs., 19 tabs., 75 refs. (Author).

  17. Transient analysis of blowdown thrust force under PWR LOCA

    International Nuclear Information System (INIS)

    Yano, Toshikazu; Miyazaki, Noriyuki; Isozaki, Toshikuni

    1982-10-01

    The analytical results of blowdown characteristics and thrust forces were compared with the experiments, which were performed as pipe whip and jet discharge tests under the PWR LOCA conditions. The blowdown thrust forces obtained by Navier-Stokes momentum equation about a single-phase, homogeneous and separated two-phase flow, assuming critical pressure at the exit if a critical flow condition was satisfied. The following results are obtained. (1) The node-junction method is useful for both the analyses of the blowdown thrust force and of the water hammer phenomena. (2) The Henry-Fauske model for subcooled critical flow is effective for the analysis of the maximum thrust force under the PWR LOCA conditions. The jet thrust parameter of the analysis and experiment is equal to 1.08. (3) The thrust parameter of saturated blowdown has the same one with the value under pressurized condition when the stagnant pressure is chosen as the saturated one. (4) The dominant terms of the blowdown thrust force in the momentum equation are the pressure and momentum terms except that the acceleration term has large contribution only just after the break. (5) The blowdown thrust force in the analysis greatly depends on the selection of the exit pressure. (author)

  18. The BG18, a B(U)F type package used for the transport of irradiated fuel rods - return of experience

    Energy Technology Data Exchange (ETDEWEB)

    Juergen, S.; Herman, S. [Transnubel, Dessel (Belgium)

    2004-07-01

    The purpose of this presentation is to share the return of experience of Transnubel after a period of nearly 3 years operation of the BG18 package in several nuclear power plants and hot cell facilities. This package has been used mainly for the shipment of full scale as well as samples of irradiated fuel rods - UOX or MOX, PWR or BWR.

  19. The BG18, a B(U)F type package used for the transport of irradiated fuel rods - return of experience

    International Nuclear Information System (INIS)

    Juergen, S.; Herman, S.

    2004-01-01

    The purpose of this presentation is to share the return of experience of Transnubel after a period of nearly 3 years operation of the BG18 package in several nuclear power plants and hot cell facilities. This package has been used mainly for the shipment of full scale as well as samples of irradiated fuel rods - UOX or MOX, PWR or BWR

  20. PWR cold-leg small break loca with faulty HPI

    International Nuclear Information System (INIS)

    Kumamaru, H.; Kukita, Y.

    1991-01-01

    The ROSA-IV Large Scale Test Facility (LSTF) is a 1/48 volumetrically-scaled model of a pressurized water reactor (PWR). At the LSTF are performed cold-leg small-break loss-of-coolant accident (LOCA) tests with faulty high pressure injection (HPI) system for break areas from 0.5% to 10% and an intentional primary system depressurization test following a small-break LOCA test. A simple prediction model is proposed for prediction of times of major events. Test data and calculations show that intentional primary system depressurization with use of the pressurizer power-operated relief valves (PORVs) is effective for break areas of approximately 0.5% or less, is unnecessary for breaks of 5% or more, and is insufficient for intermediate break areas to maintain adequate core cooling. (author)

  1. PWR reactor vessel in-service-inspection according to RSEM

    International Nuclear Information System (INIS)

    Algarotti, Marc; Dubois, Philippe; Hernandez, Luc; Landez, Jean Paul

    2006-01-01

    Nuclear services experience Framatome ANP (an AREVA and Siemens company) has designed and constructed 86 Pressurized Water Reactors (PWR) around the world including the three units lately commissioned at Ling Ao in the People's Republic of China and ANGRA 2 in Brazil; the company provided general and specialized outage services supporting numerous outages. Along with the American and German subsidiaries, Framatome ANP Inc. and Framatome ANP GmbH, Framatome ANP is among the world leading nuclear services providers, having experience of over 500 PWR outages on 4 continents, with current involvement in more than 50 PWR outages per year. Framatome ANP's experience in the examinations of reactor components began in the 1970's. Since then, each unit (American, French and German companies) developed automated NDT inspection systems and carried out pre-service and ISI (In-Service Inspections) using a large range of NDT techniques to comply with each utility expectations. These techniques have been validated by the utilities and the safety authorities of the countries where they were implemented. Notably Framatome ANP is fully qualified to provide full scope ISI services to satisfy ASME Section XI requirements, through automated NDE tasks including nozzle inspections, reactor vessel head inspections, steam generator inspections, pressurizer inspections and RPV (Reactor Pressure Vessel) inspections. Intercontrole (Framatome ANP subsidiary dedicated in supporting ISI) is one of the leading NDT companies in the world. Its main activity is devoted to the inspection of the reactor primary circuit in French and foreign PWR Nuclear Power Plants: the reactor vessel, the steam generators, the pressurizer, the reactor internals and reactor coolant system piping. NDT methods mastered by Intercontrole range from ultrasonic testing to eddy current and gamma ray examinations, as well as dye penetrant testing, acoustic monitoring and leak testing. To comply with the high requirements of

  2. PWR reactors for BBR nuclear power plants

    International Nuclear Information System (INIS)

    Structure and functioning of the nuclear steam generator system developed by BBR and its components are described. Auxiliary systems, control and load following behaviour and fuel management are discussed and the main data of PWR given. The brochure closes with a perspective of the future of the Muelheim-Kaerlich nuclear power plant. (GL) [de

  3. Thermohydraulic calculations of PWR primary circuits

    International Nuclear Information System (INIS)

    Botelho, D.A.

    1984-01-01

    Some mathematical and numerical models from Retran computer codes aiming to simulate reactor transients, are presented. The equations used for calculating one-dimensional flow are integrated using mathematical methods from Flash code, with steam code to correlate the variables from thermodynamic state. The algorithm obtained was used for calculating a PWR reactor. (E.G.) [pt

  4. Reliability of PWR type nuclear power plants

    International Nuclear Information System (INIS)

    Ribeiro, A.A.T.; Muniz, A.A.

    1978-12-01

    Results of the analysis of factors influencing the reliability of international nuclear power plants of the PWR type are presented. The reliability factor is estimated and the probability of its having lower values than a certain specified value is discussed. (Author) [pt

  5. Coolant monitoring systems for PWR reactors

    International Nuclear Information System (INIS)

    Luzhnov, A.M.; Morozov, V.V.; Tsypin, S.G.

    1987-01-01

    The ways of improving information capacity of existing monitoring systems and the necessity of designing new ones for coolant monitoring are reviewed. A wide research program on development of coolant monitoring systems in PWR reactors is analyzed. The possible applications of in-core and out-of-core detectors for coolant monitoring are demonstrated

  6. Secondary systems of PWR and BWR

    International Nuclear Information System (INIS)

    Schindler, N.

    1981-01-01

    The secondary systems of a nuclear power plant comprises the steam, condensate and feedwater cycle, the steam plant auxiliary or ancillary systems and the cooling water systems. The presentation gives a general review about the main systems which show a high similarity of PWR and BWR plants. (orig./RW)

  7. Simulation model of a PWR power plant

    International Nuclear Information System (INIS)

    Larsen, N.

    1987-03-01

    A simulation model of a hypothetical PWR power plant is described. A large number of disturbances and failures in plant function can be simulated. The model is written as seven modules to the modular simulation system for continuous processes DYSIM and serves also as a user example of this system. The model runs in Fortran 77 on the IBM-PC-AT. (author)

  8. Utilization of thorium in PWR type reactors

    International Nuclear Information System (INIS)

    Correa, F.

    1977-01-01

    Uranium 235 consumption is comparatively evaluated with thorium cycle for a PWR type reactor. Modifications are only made in fuels components. U-235 consumption is pratically unchanged in both cycles. Some good results are promised to the mixed U-238/Th-232 fuel cycle in 1/1 proportion [pt

  9. Improvement of PWR reliability by corrosion prevention

    International Nuclear Information System (INIS)

    Takamatsu, Hiroshi

    1999-01-01

    Since first PWR in Japan started commercial operation in 1970, we have encountered the various modes of corrosion on primary and secondary side components. We have paid much efforts for resolving these corrosion problems, that is, investigating the causes of corrosion and establishing the countermeasures for these corrosion. We summarize these efforts in this article. (author)

  10. An evaluation of tight - pitch PWR cores

    International Nuclear Information System (INIS)

    Correa, F.

    1980-01-01

    The subtask of a project carried out at MIT (Massachusetts Institute of Technology) for DOE (Department of Energy) as part of their NASAP/INFCE - related effects involving the optimization of PWR lattices in the recycle model is summarized. (E.G.) [pt

  11. Most experiments done so far with limited plants. Large-scale testing ...

    Indian Academy of Sciences (India)

    First page Back Continue Last page Graphics. Most experiments done so far with limited plants. Large-scale testing needs to be done with objectives such as: Apart from primary transformants, their progenies must be tested. Experiments on segregation, production of homozygous lines, analysis of expression levels in ...

  12. Disturbed Experience of Self : Psychometric Analysis of the Self-Experience Lifetime Frequency Scale (SELF)

    NARCIS (Netherlands)

    Heering, Henriette Dorothee; Goedhart, Saskia; Bruggeman, Richard; Cahn, Wiepke; de Haan, Lieuwe; Kahn, Rene S.; Meijer, Carin J.; Myin-Germeys, Inez; van Os, Jim; Wiersma, Durk

    2016-01-01

    Background: Schizophrenia is characterized by positive and negative symptoms, but recently anomalous self-experiences, e.g. exaggerated self-consciousness (hyperreflectivity), receive more attention as an important symptom domain in schizophrenia patients. The semi-structured interview, the

  13. Disturbed Experience of Self : Psychometric Analysis of the Self-Experience Lifetime Frequency Scale (SELF)

    NARCIS (Netherlands)

    Heering, Henriëtte Dorothée; Goedhart, Saskia; Bruggeman, Richard; Cahn, Wiepke; de Haan, Lieuwe; Kahn, René S; Meijer, Carin J; Myin-Germeys, Inez; van Os, Jim; Wiersma, Durk

    2016-01-01

    BACKGROUND: Schizophrenia is characterized by positive and negative symptoms, but recently anomalous self-experiences, e.g. exaggerated self-consciousness (hyperreflectivity), receive more attention as an important symptom domain in schizophrenia patients. The semi-structured interview, the

  14. Integrating the philosophy and psychology of aesthetic experience: development of the aesthetic experience scale.

    Science.gov (United States)

    Stamatopoulou, Despina

    2004-10-01

    This study assessed the dynamic relationship between person and object in aesthetic experience. Patterns of the structure of aesthetic experience were derived from a conceptual model based on philosophical and psychological ideas. These patterns were further informed by interviewing individuals with extensive involvement in aesthetic activities and 25 secondary students. Accordingly, patterns were tested by developing a large pool of items attempting to identify measurable structural components of aesthetic experience. Refined first in a pilot study, the 36-item questionnaire was administered to 652 Greek students, aged from 13 to 15 years. Correlation matrices and exploratory factor analyses on principal components were used to examine internal structural relationships. The obliquely rotated five-factor solution of the refined instrument accounted for the 44.1% of the total variance and was combatible with the conceptual model of aesthetic experience, indicating the plausibility of both. The internal consistency of the items was adequate and external correlational analysis offered preliminary support for subsequent development of a self-report measure that serves to operationalize the major constructs of aesthetic experience in the general adolescent population. The results also raise theoretical issues for those interested in empirical aesthetics, suggesting that in experiential functioning, expressive perception and affect may play a more constructive role in cognitive processes than is generally acknowledged.

  15. LOFT/LP-FW-1, Loss of Fluid Test, PWR Response to Loss-of-Feedwater Transient

    International Nuclear Information System (INIS)

    1989-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The first OECD LOFT experiment was conducted on February 20, 1983. It was designed to evaluate the generic PWR system response during a complete loss-of-feedwater transient. The objective of the experiment was to investigate the performance of primary 'feed and bleed' using a 'bleed' from the PORV and 'feed' from the HPIS to provide decay heat removal and system pressure reduction while maintaining the primary coolant inventory. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

  16. Beta and gamma dose calculations for PWR and BWR containments

    International Nuclear Information System (INIS)

    King, D.B.

    1989-07-01

    Analyses of gamma and beta dose in selected regions in PWR and BWR containment buildings have been performed for a range of fission product releases from selected severe accidents. The objective of this study was to determine the radiation dose that safety-related equipment could experience during the selected severe accident sequences. The resulting dose calculations demonstrate the extent to which design basis accident qualified equipment could also be qualified for the severe accident environments. Surry was chosen as the representative PWR plant while Peach Bottom was selected to represent BWRs. Battelle Columbus Laboratory performed the source term release analyses. The AB epsilon scenario (an intermediate to large LOCA with failure to recover onsite or offsite electrical power) was selected as the base case Surry accident, and the AE scenario (a large break LOCA with one initiating event and a combination of failures in two emergency cooling systems) was selected as the base case Peach Bottom accident. Radionuclide release was bounded for both scenarios by including spray operation and arrested sequences as variations of the base scenarios. Sandia National Laboratories used the source terms to calculate dose to selected containment regions. Scenarios with sprays operational resulted in a total dose comparable to that (2.20 x 10 8 rads) used in current equipment qualification testing. The base case scenarios resulted in some calculated doses roughly an order of magnitude above the current 2.20 x 10 8 rad equipment qualification test region. 8 refs., 23 figs., 12 tabs

  17. A new approach to PWR power control using intelligent techniques

    International Nuclear Information System (INIS)

    Boroushaki, M.; Ghofrani, M.B.; Lucas, C.; Yazdanpanah, M.J.; Sadati, N.

    2004-01-01

    Improved load following capability is one of the main technical performances of advanced PWR(APWR). Controlling the nuclear reactor core during load following operation encounters some difficulties. These difficulties mainly arise from nuclear reactor core limitations in local power peaking, while the core is subject to large and sharp variation of local power density during transients. Axial offset (A.O) is the parameter usually used to represent of core power peaking, in form of a practical parameter. This paper, proposes a new intelligent approach to A.o control of PWR nuclear reactors core during load following operation. This method uses a neural network model of the core to predict the dynamic behavior of the core and a fuzzy critic based on the operator knowledge and experience for the purpose of decision-making during load following operations. Simulation results show that this method can use optimum control rod groups maneuver with variable overlapping and may improve the reactor load following capability

  18. Interface tracking simulations of bubbly flows in PWR relevant geometries

    Energy Technology Data Exchange (ETDEWEB)

    Fang, Jun, E-mail: jfang3@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Rasquin, Michel, E-mail: michel.rasquin@colorado.edu [Aerospace Engineering Department, University of Colorado, Boulder, CO 80309 (United States); Bolotnov, Igor A., E-mail: igor_bolotnov@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States)

    2017-02-15

    Highlights: • Simulations were performed for turbulent bubbly flows in PWR subchannel geometry. • Liquid turbulence is fully resolved by direct numerical simulation approach. • Bubble behavior is captured using level-set interface tracking method. • Time-averaged single- and two-phase turbulent flow statistical quantities are obtained. - Abstract: The advances in high performance computing (HPC) have allowed direct numerical simulation (DNS) approach coupled with interface tracking methods (ITM) to perform high fidelity simulations of turbulent bubbly flows in various complex geometries. In this work, we have chosen the geometry of the pressurized water reactor (PWR) core subchannel to perform a set of interface tracking simulations (ITS) with fully resolved liquid turbulence. The presented research utilizes a massively parallel finite-element based code, PHASTA, for the subchannel geometry simulations of bubbly flow turbulence. The main objective for this research is to demonstrate the ITS capabilities in gaining new insight into bubble/turbulence interactions and assisting the development of improved closure laws for multiphase computational fluid dynamics (M-CFD). Both single- and two-phase turbulent flows were studied within a single PWR subchannel. The analysis of numerical results includes the mean gas and liquid velocity profiles, void fraction distribution and turbulent kinetic energy profiles. Two sets of flow rates and bubble sizes were used in the simulations. The chosen flow rates corresponded to the Reynolds numbers of 29,079 and 80,775 based on channel hydraulic diameter (D{sub h}) and mean velocity. The finite element unstructured grids utilized for these simulations include 53.8 million and 1.11 billion elements, respectively. This has allowed to fully resolve all the turbulence scales and the deformable interfaces of individual bubbles. For the two-phase flow simulations, a 1% bubble volume fraction was used which resulted in 17 bubbles in

  19. The verification of PWR-fuel code for PWR in-core fuel management

    International Nuclear Information System (INIS)

    Surian Pinem; Tagor M Sembiring; Tukiran

    2015-01-01

    In-core fuel management for PWR is not easy because of the number of fuel assemblies in the core as much as 192 assemblies so many possibilities for placement of the fuel in the core. Configuration of fuel assemblies in the core must be precise and accurate so that the reactor operates safely and economically. It is necessary for verification of PWR-FUEL code that will be used in-core fuel management for PWR. PWR-FUEL code based on neutron transport theory and solved with the approach of multi-dimensional nodal diffusion method many groups and diffusion finite difference method (FDM). The goal is to check whether the program works fine, especially for the design and in-core fuel management for PWR. Verification is done with equilibrium core search model at three conditions that boron free, 1000 ppm boron concentration and critical boron concentration. The result of the average burn up fuel assemblies distribution and power distribution at BOC and EOC showed a consistent trend where the fuel with high power at BOC will produce a high burn up in the EOC. On the core without boron is obtained a high multiplication factor because absence of boron in the core and the effect of fission products on the core around 3.8 %. Reactivity effect at 1000 ppm boron solution of BOC and EOC is 6.44 % and 1.703 % respectively. Distribution neutron flux and power density using NODAL and FDM methods have the same result. The results show that the verification PWR-FUEL code work properly, especially for core design and in-core fuel management for PWR. (author)

  20. Contribution to the experimental qualification of PWR fuel storage calculations

    International Nuclear Information System (INIS)

    Marsault, Philippe.

    1980-12-01

    Experiments were carried out on assemblies representative of those used in PWR reactors in a configuration made critical with a driver zone. In this way, certain parameters were able to be measured using current classical techniques. As the multiplication factor for a group of assemblies cannot be determined directly, substitutions were made with an equivalent homogeneous lattice in which Laplacian measurements could be made. The k(infinite) factor was obtained by introducing a migration area which can only be obtained from calculations. Experimental storage studies realized during the CRISTO 1 campaign utilize: 1) a lattice with 4 14x14 pin assemblies immersed in ordinary water; 2) a lattice with 4 14x14 pin assemblies and 3) a regular lattice. The CRISTO experiment enabled criticality calculations to be qualified with these lattices for storage under accidental conditions [fr

  1. Pore-scale and Continuum Simulations of Solute Transport Micromodel Benchmark Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Oostrom, Martinus; Mehmani, Yashar; Romero Gomez, Pedro DJ; Tang, Y.; Liu, H.; Yoon, Hongkyu; Kang, Qinjun; Joekar Niasar, Vahid; Balhoff, Matthew; Dewers, T.; Tartakovsky, Guzel D.; Leist, Emily AE; Hess, Nancy J.; Perkins, William A.; Rakowski, Cynthia L.; Richmond, Marshall C.; Serkowski, John A.; Werth, Charles J.; Valocchi, Albert J.; Wietsma, Thomas W.; Zhang, Changyong

    2016-08-01

    Four sets of micromodel nonreactive solute transport experiments were conducted with flow velocity, grain diameter, pore-aspect ratio, and flow focusing heterogeneity as the variables. The data sets were offered to pore-scale modeling groups to test their simulators. Each set consisted of two learning experiments, for which all results was made available, and a challenge experiment, for which only the experimental description and base input parameters were provided. The experimental results showed a nonlinear dependence of the dispersion coefficient on the Peclet number, a negligible effect of the pore-aspect ratio on transverse mixing, and considerably enhanced mixing due to flow focusing. Five pore-scale models and one continuum-scale model were used to simulate the experiments. Of the pore-scale models, two used a pore-network (PN) method, two others are based on a lattice-Boltzmann (LB) approach, and one employed a computational fluid dynamics (CFD) technique. The learning experiments were used by the PN models to modify the standard perfect mixing approach in pore bodies into approaches to simulate the observed incomplete mixing. The LB and CFD models used these experiments to appropriately discretize the grid representations. The continuum model use published non-linear relations between transverse dispersion coefficients and Peclet numbers to compute the required dispersivity input values. Comparisons between experimental and numerical results for the four challenge experiments show that all pore-scale models were all able to satisfactorily simulate the experiments. The continuum model underestimated the required dispersivity values and, resulting in less dispersion. The PN models were able to complete the simulations in a few minutes, whereas the direct models needed up to several days on supercomputers to resolve the more complex problems.

  2. Probing the frontiers of particle physics with tabletop-scale experiments.

    Science.gov (United States)

    DeMille, David; Doyle, John M; Sushkov, Alexander O

    2017-09-08

    The field of particle physics is in a peculiar state. The standard model of particle theory successfully describes every fundamental particle and force observed in laboratories, yet fails to explain properties of the universe such as the existence of dark matter, the amount of dark energy, and the preponderance of matter over antimatter. Huge experiments, of increasing scale and cost, continue to search for new particles and forces that might explain these phenomena. However, these frontiers also are explored in certain smaller, laboratory-scale "tabletop" experiments. This approach uses precision measurement techniques and devices from atomic, quantum, and condensed-matter physics to detect tiny signals due to new particles or forces. Discoveries in fundamental physics may well come first from small-scale experiments of this type. Copyright © 2017 The Authors, some rights reserved; exclusive licensee American Association for the Advancement of Science. No claim to original U.S. Government Works.

  3. Study of influence of an experiment scale on cylinder test results

    Directory of Open Access Journals (Sweden)

    Waldemar A. Trzciński

    2014-03-01

    Full Text Available In the work, influence of a scale of experiment on the results of cylindrical test used todetermine the acceleration capabilities of explosives was analyzed. Explosives used in ammunition(TNT, hexogen and explosives for civil purpose (ammonals were selected for testing. Copper tubeswith different diameters and wall thickness were used. Conclusions are drawn regarding the advisabilityof increasing or decreasing the scale of the cylinder test.[b]Keywords[/b]: explosives, acceleration ability, cylinder test

  4. Scaling options for integral experiments for molten salt fluid mechanics and heat transfer

    International Nuclear Information System (INIS)

    Philippe Bardet; Per F Peterson

    2005-01-01

    Full text of publication follows: Molten fluoride salts have potentially large benefits for use in high-temperature heat transport in fission and fusion energy systems, due to their very very low vapor pressures at high temperatures. Molten salts have high volumetric heat capacity compared to high-pressure helium and liquid metals, and have desirable safety characteristics due to their chemical inertness and low pressure. Therefore molten salts have been studied extensively for use in fusion blankets, as an intermediate heat transfer fluid for thermochemical hydrogen production in the Next Generation Nuclear Plant, as a primary coolant for the Advanced High Temperature Reactor, and as a solvent for fuel in the Molten Salt Reactor. This paper presents recent progress in the design and analysis of scaled thermal hydraulics experiments for molten salt systems. We have identified a category of light mineral oils that can be used for scaled experiments. By adjusting the length, velocity, average temperature, and temperature difference scales of the experiment, we show that it is possible to simultaneously match the Reynolds (Re), Froude (Fr), Prandtl (Pr) and Rayleigh (Ra) numbers in the scaled experiments. For example, the light mineral oil Penreco Drakesol 260 AT can be used to simulate the molten salt flibe (Li 2 BeF 4 ). At 110 deg. C, the oil Pr matches 600 deg. C flibe, and at 165 deg. C, the oil Pr matches 900 deg. C flibe. Re, Fr, and Ra can then be matched at a length scale of Ls/Lp = 0.40, velocity scale of U s /U p = 0.63, and temperature difference scale of ΔT s /ΔT p = 0.29. The Weber number is then matched within a factor of two, We s /We p = 0.7. Mechanical pumping power scales as Qp s /Qp p = 0.016, while heat inputs scale as Qh s /Qh p = 0.010, showing that power inputs to scaled experiments are very small compared to the prototype system. The scaled system has accelerated time, t s /t p = 0.64. When Re, Fr, Pr and Ra are matched, geometrically scaled

  5. Methodology for the LABIHS PWR simulator modernization

    Energy Technology Data Exchange (ETDEWEB)

    Jaime, Guilherme D.G.; Oliveira, Mauro V., E-mail: gdjaime@ien.gov.b, E-mail: mvitor@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    The Human-System Interface Laboratory (LABIHS) simulator is composed by a set of advanced hardware and software components whose goal is to simulate the main characteristics of a Pressured Water Reactor (PWR). This simulator serves for a set of purposes, such as: control room modernization projects; designing of operator aiding systems; providing technological expertise for graphical user interfaces (GUIs) designing; control rooms and interfaces evaluations considering both ergonomics and human factors aspects; interaction analysis between operators and the various systems operated by them; and human reliability analysis in scenarios considering simulated accidents and normal operation. The simulator runs in a PA-RISC architecture server (HPC3700), developed nearby 2000's, using the HP-UX operating system. All mathematical modeling components were written using the HP Fortran-77 programming language with a shared memory to exchange data from/to all simulator modules. Although this hardware/software framework has been discontinued in 2008, with costumer support ceasing in 2013, it is still used to run and operate the simulator. Due to the fact that the simulator is based on an obsolete and proprietary appliance, the laboratory is subject to efficiency and availability issues, such as: downtime caused by hardware failures; inability to run experiments on modern and well known architectures; and lack of choice of running multiple simulation instances simultaneously. This way, there is a need for a proposal and implementation of solutions so that: the simulator can be ported to the Linux operating system, running on the x86 instruction set architecture (i.e. personal computers); we can simultaneously run multiple instances of the simulator; and the operator terminals run remotely. This paper deals with the design stage of the simulator modernization, in which it is performed a thorough inspection of the hardware and software currently in operation. Our goal is to

  6. Methodology for the LABIHS PWR simulator modernization

    Energy Technology Data Exchange (ETDEWEB)

    Jaime, Guilherme D.G.; Oliveira, Mauro V., E-mail: gdjaime@ien.gov.b, E-mail: mvitor@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    The Human-System Interface Laboratory (LABIHS) simulator is composed by a set of advanced hardware and software components whose goal is to simulate the main characteristics of a Pressured Water Reactor (PWR). This simulator serves for a set of purposes, such as: control room modernization projects; designing of operator aiding systems; providing technological expertise for graphical user interfaces (GUIs) designing; control rooms and interfaces evaluations considering both ergonomics and human factors aspects; interaction analysis between operators and the various systems operated by them; and human reliability analysis in scenarios considering simulated accidents and normal operation. The simulator runs in a PA-RISC architecture server (HPC3700), developed nearby 2000's, using the HP-UX operating system. All mathematical modeling components were written using the HP Fortran-77 programming language with a shared memory to exchange data from/to all simulator modules. Although this hardware/software framework has been discontinued in 2008, with costumer support ceasing in 2013, it is still used to run and operate the simulator. Due to the fact that the simulator is based on an obsolete and proprietary appliance, the laboratory is subject to efficiency and availability issues, such as: downtime caused by hardware failures; inability to run experiments on modern and well known architectures; and lack of choice of running multiple simulation instances simultaneously. This way, there is a need for a proposal and implementation of solutions so that: the simulator can be ported to the Linux operating system, running on the x86 instruction set architecture (i.e. personal computers); we can simultaneously run multiple instances of the simulator; and the operator terminals run remotely. This paper deals with the design stage of the simulator modernization, in which it is performed a thorough inspection of the hardware and software currently in operation. Our goal is to

  7. Methodology for the LABIHS PWR simulator modernization

    International Nuclear Information System (INIS)

    Jaime, Guilherme D.G.; Oliveira, Mauro V.

    2011-01-01

    The Human-System Interface Laboratory (LABIHS) simulator is composed by a set of advanced hardware and software components whose goal is to simulate the main characteristics of a Pressured Water Reactor (PWR). This simulator serves for a set of purposes, such as: control room modernization projects; designing of operator aiding systems; providing technological expertise for graphical user interfaces (GUIs) designing; control rooms and interfaces evaluations considering both ergonomics and human factors aspects; interaction analysis between operators and the various systems operated by them; and human reliability analysis in scenarios considering simulated accidents and normal operation. The simulator runs in a PA-RISC architecture server (HPC3700), developed nearby 2000's, using the HP-UX operating system. All mathematical modeling components were written using the HP Fortran-77 programming language with a shared memory to exchange data from/to all simulator modules. Although this hardware/software framework has been discontinued in 2008, with costumer support ceasing in 2013, it is still used to run and operate the simulator. Due to the fact that the simulator is based on an obsolete and proprietary appliance, the laboratory is subject to efficiency and availability issues, such as: downtime caused by hardware failures; inability to run experiments on modern and well known architectures; and lack of choice of running multiple simulation instances simultaneously. This way, there is a need for a proposal and implementation of solutions so that: the simulator can be ported to the Linux operating system, running on the x86 instruction set architecture (i.e. personal computers); we can simultaneously run multiple instances of the simulator; and the operator terminals run remotely. This paper deals with the design stage of the simulator modernization, in which it is performed a thorough inspection of the hardware and software currently in operation. Our goal is to

  8. Experiments to investigate direct containment heating phenomena with scaled models of the Surry Nuclear Power Plant

    International Nuclear Information System (INIS)

    Blanchat, T.K.; Allen, M.D.; Pilch, M.M.

    1994-01-01

    The Containment Technology Test Facility (CTTF) and the Surtsey Test Facility at Sandia National Laboratories (SNL) are used to perform scaled experiments for the Nuclear Regulatory Commission (NRC) that simulate High Pressure Melt Ejection (HPME) accidents in a nuclear power plant (NPP). These experiments are designed to investigate the effects of direct containment heating (DCH) phenomena on the containment load. High-temperature, chemically reactive melt is ejected by high-pressure steam into a scale model of a reactor cavity. Debris is entrained by the steam blowdown into a containment model where specific phenomena, such as the effect of subcompartment structures, prototypic atmospheres, and hydrogen generation and combustion, can be studied

  9. The MIRAGE project: large scale radionuclide transport investigations and integral migration experiments

    International Nuclear Information System (INIS)

    Come, B.; Bidoglio, G.; Chapman, N.

    1986-01-01

    Predictions of radionuclide migration through the geosphere must be supported by large-scale, long-term investigations. Several research areas of the MIRAGE Project are devoted to acquiring reliable data for developing and validating models. Apart from man-made migration experiments in boreholes and/or underground galleries, attention is paid to natural geological migration systems which have been active for very long time spans. The potential role of microbial activity, either resident or introduced into the host media, is also considered. In order to clarify basic mechanisms, smaller scale ''integral'' migration experiments under fully controlled laboratory conditions are also carried out using real waste forms and representative geological media. (author)

  10. Confirmation tests of PWR surveillance capsule shipping container

    International Nuclear Information System (INIS)

    Tomita, N.; Ue, K.; Ohashi, M.; Asada, K.; Yoneda, Y.

    1980-01-01

    Mitsubishi Heavy Industries, Ltd. carried out the confirmation tests to confirm the reliability of the PWR surveillance capsule shipping container and to collect cask design data using a 10-ton weight full scale model at Kobe Shipyard and Engine Works. This report presents the outline of these tests. The B Type container was a cylinder 3289 mm long, 1080 mm in diameter and designed in accordance with the new modified Japanese regulations similar to IAEA regulation. These tests consist of four 9 m drop tests, two 1 m puncture tests, a fire test and an immersion test. In conclusion, safetyness of this container has been proved and various technical data for cask design were also collected through these tests. (author)

  11. Fuel management optimization for a PWR

    International Nuclear Information System (INIS)

    Dumas, M.; Robeau, D.

    1981-04-01

    This study is aimed to optimize the refueling pattern of a PWR. Two methods are developed, they are based on a linearized form of the optimization problem. The first method determines a feasible solution in two steps; in the first one the original problem is replaced by a relaxed one which is solved by the Method of Approximation Programming. The second step is based on the Branch and Bound method to find the feasible solution closest to the solution obtained in the first step. The second method starts from a given refueling pattern and tries to improve this pattern by the calculation of the effects of 2 by 2, 3 by 3 and 4 by 4 permutations on the objective function. Numerical results are given for a typical PWR refueling using the two methods

  12. RSK-guidelines for PWR reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The RSK guidelines for PWA reactors of April 24, 1974, have been revised and amended in this edition. The RSK presents a summary of safety requirements to be observed in the design, construction, and operation of PWR reactors in the form of guidelines. From January 1979 onwards these guidelines will be the basis of siting and safety considerations for new PWR reactors, and newly built nuclear power plants will have to form these guidelines. They are not binding for existing nuclear power plants under construction or in operation. It will be a matter of individual discussion whether or not the guidelines will be applied in these plants. The main purpose of the guidelines is to facilitate discussion among RSK members and to give early information on necessary safety requirements. If the guidelines are observed by producers and operators, the RSK will make statements on individual projects at short notice. (orig./HP) [de

  13. Optimization of reload core design for PWR

    International Nuclear Information System (INIS)

    Shen Wei; Xie Zhongsheng; Yin Banghua

    1995-01-01

    A direct efficient optimization technique has been effected for automatically optimizing the reload of PWR. The objective functions include: maximization of end-of-cycle (EOC) reactivity and maximization of average discharge burnup. The fuel loading optimization and burnable poison (BP) optimization are separated into two stages by using Haling principle. In the first stage, the optimum fuel reloading pattern without BP is determined by the linear programming method using enrichments as control variable, while in the second stage the optimum BP allocation is determined by the flexible tolerance method using the number of BP rods as control variable. A practical and efficient PWR reloading optimization program based on above theory has been encoded and successfully applied to Qinshan Nuclear Power Plant (QNP) cycle 2 reloading design

  14. Assessment of TRAC-PF1/MOD1 code for large break LOCA in PWR

    International Nuclear Information System (INIS)

    Akimoto, Hajime; Ohnuki, Akira; Murao, Yoshio; Abe, Yutaka.

    1993-03-01

    As the first step of the REFLA/TRAC code development, the TRAC/PF1/MOD1 code has been assessed for various experiments that simulate postulated large-break loss-of-coolant accident (LBLOCA) in PWR to understand the predictive capability and to identify the problem areas of the code. The assessment calculations were performed for separate effect tests for critical flow, counter current flow, condensation at cold leg and reflood as well as integral tests to understand predictability for individual phenomena. This report summarizes results from the assessment calculations of the TRAC-PF1/MOD1 code for LBLOCA in PWR. The assessment calculations made clear the predictive capability and problem areas of the TRAC-PF1/MOD1 code for LBLOCA in PWR. The areas, listed below, should be improved for more realistic and effective simulation of LBLOCA in PWR: (1) core heat transfer model during blowdown, (2) ECC bypass model at downcomer during refill, (3) condensation model during accumulator injection, and (4) core thermal hydraulic model during reflood. (author) 57 refs

  15. Development and application of integrated digital I and C system in Japanese PWR plants

    International Nuclear Information System (INIS)

    Tominaga, M.

    1995-01-01

    The Integrated Digital Instrumentation and Control (I and C) System has been developed and applied to non-safety grade I and C systems in the latest 5 Japanese PWR plants in 1990's. Based on the experience in these plants, the Integrated Digital I and C System will be planned to apply also to safety grade I and C systems in Advanced PWR (APWR) as the overall application of digital technology. The basic design task has been just started for APWR which is to be in commercial operation in early 2000's and under the development about various issues of safety grade digital I and C systems. On the other hand, in conventional Japanese PWR plants, digital I and C systems have been applied step by step since 1980's. For example, digital I and C systems for radio-active waste processing system have been adopted to 13 units, and dedicated digital I and C systems for Local loop control system to 8 units. The trend and status of development and application of the digital I and C systems, especially the Integrated Digital I and C System in Japanese PWR plants are presented. (5 refs., 4 figs.)

  16. Notes on a Dramaturgical Analysis of Unequal Small-Scale Corruption Experiences

    Directory of Open Access Journals (Sweden)

    Edgar Daniel Manchinelly Mota

    2017-10-01

    Full Text Available In the last two decades, corruption has emerged as a relevant subject on a worldwide scale, because of its negative effects on the economy and State institutions, among other things. Research has focused on the macro aspects of corruption, emphasizing its causes and consequences. However, small-scale corruption has not been studied in such detail. This document proposes a theoretical-methodological framework for a dramaturgical analysis of small-scale corruption, with the aim of demonstrating that it is a stratified interaction. In this sense, corruption is an unequal experience for citizens, which depends on individuals’ social position.

  17. Global Energy and Water Cycle Experiment (GEWEX) and the Continental-scale International Project (GCIP)

    Science.gov (United States)

    Vane, Deborah

    1993-01-01

    A discussion of the objectives of the Global Energy and Water Cycle Experiment (GEWEX) and the Continental-scale International Project (GCIP) is presented in vugraph form. The objectives of GEWEX are as follows: determine the hydrological cycle by global measurements; model the global hydrological cycle; improve observations and data assimilation; and predict response to environmental change. The objectives of GCIP are as follows: determine the time/space variability of the hydrological cycle over a continental-scale region; develop macro-scale hydrologic models that are coupled to atmospheric models; develop information retrieval schemes; and support regional climate change impact assessment.

  18. Fostering sustainable small-scale investments: lessons from experience and ideas for intervention and innovation

    International Nuclear Information System (INIS)

    Lucia, R.J. de

    2000-01-01

    This article presents lessons from experience pertinent to implementing small-scale natural resource and related investment projects in developing countries. It outlines ideas for intervention and innovation to foster such investments. Particular emphasis is placed on private-sector participation in these investments. Following a brief presentation of the economic development and other arguments that support intervention and innovation in support of such small-scale natural resource investments, the article discusses many of the lessons learned from experience. These lessons reinforce those touched upon in the investment-specific discussions in the previous articles. The experience and associated lessons that are examined include experiences in both developing and developed countries and encompass nor just lessons from natural resource investments, but also small-scale investments in other sectors. Financing and other innovations which facilitate meeting the challenges are drawn from relevant experience where barriers to investment have been surmounted. Options for programme and project interventions to increase market penetration of small-scale investments and achieve the associated development linkages and synergies are suggested. These suggestions are aimed especially at governments and bilateral and multilateral development finance and development assistance entities. It is those players who might support such interventions in collaboration with local financial institutions and other market players. (author)

  19. EDF/CIDEN - ONECTRA: PWR decontamination

    International Nuclear Information System (INIS)

    Fayolle, P.; Orcel, H.; Wertz, L.

    2010-01-01

    In the context of PWR circuit renewal (expected in 2011) and their decontamination, an analysis of data coming from cartography and on site decontamination measurements as well as from premise modelling by means of the PANTHERE radioprotection code, is presented. Several French PWRs have been studied. After a presentation of code principles and operation, the authors discuss the radiological context of a workstation, and give an assessment of the annual dose associated with maintenance operations with or without decontamination

  20. Optimum fuel use in PWR reactors

    International Nuclear Information System (INIS)

    Neubauer, W.

    1979-07-01

    An optimization program was developed to calculate minimum-cost refuelling schedules for PWR reactors. Optimization was made over several cycles, without any constraints (equilibrium cycle). In developing the optimization program, special consideration was given to an individual treatment of every fuel element and to a sufficiently accurate calculation of all the data required for safe reactor operation. The results of the optimization program were compared with experimental values obtained at Obrigheim nuclear power plant. (orig.) [de

  1. Chemical and radiochemical specifications - PWR power plants

    International Nuclear Information System (INIS)

    Stutzmann, A.

    1997-01-01

    Published by EDF this document gives the chemical specifications of the PWR (Pressurized Water Reactor) nuclear power plants. Among the chemical parameters, some have to be respected for the safety. These parameters are listed in the STE (Technical Specifications of Exploitation). The values to respect, the analysis frequencies and the time states of possible drops are noticed in this document with the motion STE under the concerned parameter. (A.L.B.)

  2. GAIA: AREVAs New PWR fuel assembly design

    Energy Technology Data Exchange (ETDEWEB)

    Vollmert, N.; Gentet, G.; Louf, P.H.; Mindt, M.; O' Brian, J.; Peucker, J.

    2015-07-01

    GAIA is the label of a new PWR Fuel Assembly design developed by AREVA with the objective to provide its customers an advanced fuel assembly design regarding both robustness and performance. Since 2012 GAIA lead fuel assemblies are under irradiation in a Swedish reactor and since 2015 in a U.S. reactor. Visual inspections and examinations carried out so far during the outages confirmed the intended reliability, robustness and the performance enhancement of the design. (Author)

  3. Shielding design for PWR in France

    International Nuclear Information System (INIS)

    Champion, G.; Charransol; Le Dieu de Ville, A.; Nimal, J.C.; Vergnaud, T.

    1983-05-01

    Shielding calculation scheme used in France for PWR is presented here for 900 MWe and 1300 MWe plants built by EDF the French utility giving electricity. Neutron dose rate at areas accessible by personnel during the reactor operation is calculated and compared with the measurements which were carried out in 900 MWe units up to now. Measurements on the first French 1300 MWe reactor are foreseen at the end of 1983

  4. Organization patterns of PWR power plants

    International Nuclear Information System (INIS)

    Leicman, J.

    1980-01-01

    Organization patterns are shown for the St. Lucia 1, North Anna, Sequoyah, and Beaver Valley nuclear power plants, for a typical PWR power plant in the USA, for the Biblis/RWE-KWU nuclear power plants and for a four-unit nuclear power plant operated by Electricite de France as well as for the Loviisa power plant. Organization patterns are also shown for relatively independent and non-independent nuclear power plants according to IAEA recommendations. (J.P.)

  5. Sensitivity analysis of a PWR pressurizer

    International Nuclear Information System (INIS)

    Bruel, Renata Nunes

    1997-01-01

    A sensitivity analysis relative to the parameters and modelling of the physical process in a PWR pressurizer has been performed. The sensitivity analysis was developed by implementing the key parameters and theoretical model lings which generated a comprehensive matrix of influences of each changes analysed. The major influences that have been observed were the flashing phenomenon and the steam condensation on the spray drops. The present analysis is also applicable to the several theoretical and experimental areas. (author)

  6. T Plant removal of PWR Chiller Subsystem

    International Nuclear Information System (INIS)

    Dana, C.M.

    1994-01-01

    The PWR Pool Chiller System is not longer required for support of the Shippingport Blanket Fuel Assemblies Storage. The Engineering Work Plan will provide the overall coordination of the documentation and physical changes to deactivate the unneeded subsystem. The physical removal of all energy sources for the Chiller equipment will be covered under a one time work plan. The documentation changes will be covered using approved Engineering Change Notices and Procedure Change Authorizations as needed

  7. Hydroelastic model of PWR reactor internals SAFRAN 1 - Validation of a vibration calculation method

    International Nuclear Information System (INIS)

    Epstein, A.; Gibert, R.J.; Jeanpierre, F.; Livolant, M.

    1978-01-01

    The SAFRAN 1 test loop consists of an hydroelastic similitude of a 1/8 scale model of a 3 loop P.W.R. Vibrations of the main internals (thermal shield and core barrel) and pressure fluctuations in water thin sections between vessel and internals, and in inlet and outlet pipes, have been measured. The calculation method consists of: an evaluation of the main vibration and acoustic sources owing to the flow (unsteady jet impingement on the core barrel, turbulent flow in a water thin section). A calculation of the internal modal parameters taking into account the inertial effects of fluid (the computer codes AQUAMODE and TRISTANA have been used). A calculation of the acoustic response of the circuit (the computer code VIBRAPHONE has been used). The good agreement between the calculation and the experimental results allows using this method with better security for the prediction of the vibration levels of full scale P.W.R. internals

  8. Model-based fault detection and isolation of a PWR nuclear power plant using neural networks

    International Nuclear Information System (INIS)

    Far, R.R.; Davilu, H.; Lucas, C.

    2008-01-01

    The proper and timely fault detection and isolation of industrial plant is of premier importance to guarantee the safe and reliable operation of industrial plants. The paper presents application of a neural networks-based scheme for fault detection and isolation, for the pressurizer of a PWR nuclear power plant. The scheme is constituted by 2 components: residual generation and fault isolation. The first component generates residuals via the discrepancy between measurements coming from the plant and a nominal model. The neutral network estimator is trained with healthy data collected from a full-scale simulator. For the second component detection thresholds are used to encode the residuals as bipolar vectors which represent fault patterns. These patterns are stored in an associative memory based on a recurrent neutral network. The proposed fault diagnosis tool is evaluated on-line via a full-scale simulator detected and isolate the main faults appearing in the pressurizer of a PWR. (orig.)

  9. Nondestructive examination requirements for PWR vessel internals

    International Nuclear Information System (INIS)

    Spanner, J.

    2015-01-01

    This paper describes the requirements for the nondestructive examination of pressurized water reactor (PWR) vessel internals in accordance with the requirements of the EPRI Material Reliability Program (MRP) inspection standard for PWR internals (MRP-228) and the American Society of Mechanical Engineers Section XI In-service Inspection. The MRP vessel internals examinations have been performed at nuclear plants in the USA since 2009. The objective of the inspection standard is to provide the requirements for the nondestructive examination (NDE) methods implemented to support the inspection and evaluation of the internals. The inspection standard contains requirements specific to the inspection methodologies involved as well as requirements for qualification of the NDE procedures, equipment and personnel used to perform the vessel internals inspections. The qualification requirements for the NDE systems will be summarized. Six PWR plants in the USA have completed inspections of their internals using the Inspection and Evaluation Guideline (MRP-227) and the Inspection Standard (MRP-228). Examination results show few instances of service-induced degradation flaws, as expected. The few instances of degradation have mostly occurred in bolting

  10. Modelling activity transport behavior in PWR plant

    International Nuclear Information System (INIS)

    Henshaw, Jim; McGurk, John; Dickinson, Shirley; Burrows, Robert; Hinds, Kelvin; Hussey, Dennis; Deshon, Jeff; Barrios Figueras, Joan Pau; Maldonado Sanchez, Santiago; Fernandez Lillo, Enrique; Garbett, Keith

    2012-09-01

    The activation and transport of corrosion products around a PWR circuit is a major concern to PWR plant operators as these may give rise to high personnel doses. The understanding of what controls dose rates on ex-core surfaces and shutdown releases has improved over the years but still several questions remain unanswered. For example the relative importance of particle and soluble deposition in the core to activity levels in the plant is not clear. Wide plant to plant and cycle to cycle variations are noted with no apparent explanations why such variations are observed. Over the past few years this group have been developing models to simulate corrosion product transport around a PWR circuit. These models form the basis for the latest version of the BOA code and simulate the movement of Fe and Ni around the primary circuit. Part of this development is to include the activation and subsequent transport of radioactive species around the circuit and this paper describes some initial modelling work in this area. A simple model of activation, release and deposition is described and then applied to explain the plant behaviour at Sizewell B and Vandellos II. This model accounts for activation in the core, soluble and particulate activity movement around the circuit and for activity capture ex-core on both the inner and outer oxides. The model gives a reasonable comparison with plant observations and highlights what controls activity transport in these plants and importantly what factors can be ignored. (authors)

  11. Diffusion Experiments with Opalinus and Callovo-Oxfordian Clays: Laboratory, Large-Scale Experiments and Microscale Analysis by RBS

    International Nuclear Information System (INIS)

    Garcia-Gutierrez, M.; Alonso, U.; Missana, T.; Cormenzana, J.L.; Mingarro, M.; Morejon, J.; Gil, P.

    2009-01-01

    Consolidated clays are potential host rocks for deep geological repositories for high-level radioactive waste. Diffusion is the main transport process for radionuclides (RN) in these clays. Radionuclide (RN) diffusion coefficients are the most important parameters for Performance Assessment (PA) calculations of clay barriers. Different diffusion methodologies were applied at a laboratory scale to analyse the diffusion behaviour of a wide range of RN. Main aims were to understand the transport properties of different RNs in two different clays and to contribute with feasible methodologies to improve in-situ diffusion experiments, using samples of larger scale. Classical laboratory essays and a novel experimental set-up for large-scale diffusion experiments were performed, together to a novel application of the nuclear ion beam technique Rutherford Backscattering Spectrometry (RBS), for diffusion analyses at the micrometer scale. The main experimental and theoretical characteristics of the different methodologies, and their advantages and limitations are here discussed. Experiments were performed with the Opalinus and the Callovo-Oxfordian clays. Both clays are studied as potential host rock for a repository. Effective diffusion coefficients ranged between 1.10 - 10 to 1.10 - 12 m 2 /s for neutral, low sorbing cations (as Na and Sr) and anions. Apparent diffusion coefficients for strongly sorbing elements, as Cs and Co, are in the order of 1.10-13 m 2 /s; europium present the lowest diffusion coefficient (5.10 - 15 m 2 /s). The results obtained by the different approaches gave a comprehensive database of diffusion coefficients for RN with different transport behaviour within both clays. (Author) 42 refs

  12. Diffusion Experiments with Opalinus and Callovo-Oxfordian Clays: Laboratory, Large-Scale Experiments and Microscale Analysis by RBS

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Gutierrez, M.; Alonso, U.; Missana, T.; Cormenzana, J.L.; Mingarro, M.; Morejon, J.; Gil, P.

    2009-09-25

    Consolidated clays are potential host rocks for deep geological repositories for high-level radioactive waste. Diffusion is the main transport process for radionuclides (RN) in these clays. Radionuclide (RN) diffusion coefficients are the most important parameters for Performance Assessment (PA) calculations of clay barriers. Different diffusion methodologies were applied at a laboratory scale to analyse the diffusion behaviour of a wide range of RN. Main aims were to understand the transport properties of different RNs in two different clays and to contribute with feasible methodologies to improve in-situ diffusion experiments, using samples of larger scale. Classical laboratory essays and a novel experimental set-up for large-scale diffusion experiments were performed, together to a novel application of the nuclear ion beam technique Rutherford Backscattering Spectrometry (RBS), for diffusion analyses at the micrometer scale. The main experimental and theoretical characteristics of the different methodologies, and their advantages and limitations are here discussed. Experiments were performed with the Opalinus and the Callovo-Oxfordian clays. Both clays are studied as potential host rock for a repository. Effective diffusion coefficients ranged between 1.10{sup -}10 to 1.10{sup -}12 m{sup 2}/s for neutral, low sorbing cations (as Na and Sr) and anions. Apparent diffusion coefficients for strongly sorbing elements, as Cs and Co, are in the order of 1.10-13 m{sup 2}/s; europium present the lowest diffusion coefficient (5.10{sup -}15 m{sup 2}/s). The results obtained by the different approaches gave a comprehensive database of diffusion coefficients for RN with different transport behaviour within both clays. (Author) 42 refs.

  13. Zircaloy-4 corrosion in PWR's

    International Nuclear Information System (INIS)

    Fyfitch, S.; Smalley, W.R.; Roberts, E.

    1985-01-01

    Zircaloy-4 waterside corrosion has been studied extensively in the nuclear industry for a number of years. Following the early crud-related corrosion failures in the Saxton test reactor, Westinghouse undertook numerous programs to minimize crud deposition on fuel rods in power reactors through primary coolant chemistry control. Modern plants today are operating with improved coolant chemistry guidelines, and crud deposition levels are very low in proportion to earlier experience. Zircaloy-4 corrosion under a variety of coolant chemistry, heat flux and exposure conditions has been studied extensively. Experience to date, even in relatively high coolant temperature plants, has indicated that -for both fuel cladding and structural components- Zircaloy-4 waterside corrosion performance has been excellent. Recognizing future industry trends, however, which will result in Zircaloy-4 being subjected to ever increasing corrosion duties, Westinghouse will continue accumulating Zircaloy-4 corrosion experience in large power plants. 13 refs.

  14. Implementation in free software of the PWR type university nucleo electric simulator (SU-PWR)

    International Nuclear Information System (INIS)

    Valle H, J.; Hidago H, F.; Morales S, J.B.

    2007-01-01

    Presently work is shown like was carried out the implementation of the University Simulator of Nucleo-electric type PWR (SU-PWR). The implementation of the simulator was carried out in a free software simulation platform, as it is Scilab, what offers big advantages that go from the free use and without cost of the product, until the codes modification so much of the system like of the program with the purpose of to improve it or to adapt it to future routines and/or more advanced graphic interfaces. The SU-PWR shows the general behavior of a PWR nuclear plant (Pressurized Water Reactor) describing the dynamics of the plant from the generation process of thermal energy in the nuclear fuel, going by the process of energy transport toward the coolant of the primary circuit the one which in turn transfers this energy to the vapor generators of the secondary circuit where the vapor is expanded by means of turbines that in turn move the electric generator producing in this way the electricity. The pressurizer that is indispensable for the process is also modeled. Each one of these stages were implemented in scicos that is the Scilab tool specialized in the simulation. The simulation was carried out by means of modules that contain the differential equation that mathematically models each stage or equipment of the PWR plant. The result is a series of modules that based on certain entrances and characteristic of the system they generate exits that in turn are the entrance to other module. Because the SU-PWR is an experimental project in early phase, it is even work and modifications to carry out, for what the models that are presented in this work can vary a little the being integrated to the whole system to simulate, but however they already show clearly the operation and the conformation of the plant. (Author)

  15. Activity transport models for PWR primary circuits; PWR-ydinvoimalaitoksen primaeaeripiirin aktiivisuuskulkeutumismallit

    Energy Technology Data Exchange (ETDEWEB)

    Tanner, V; Rosenberg, R [VTT Chemical Technology, Otaniemi (Finland)

    1995-03-01

    The corrosion products activated in the primary circuit form a major source of occupational radiation dose in the PWR reactors. Transport of corrosion activity is a complex process including chemistry, reactor physics, thermodynamics and hydrodynamics. All the mechanisms involved are not known and there is no comprehensive theory for the process, so experimental test loops and plant data are very important in research efforts. Several activity transport modelling attempts have been made to improve the water chemistry control and to minimise corrosion in PWR`s. In this research report some of these models are reviewed with special emphasis on models designed for Soviet VVER type reactors. (51 refs., 16 figs., 4 tabs.).

  16. Program of monitoring PWR fuel in Spain; Programa de Vigilancia de Combustible pwr en Espana

    Energy Technology Data Exchange (ETDEWEB)

    Martinez Murillo, J. C.; Quecedo, M.; Munoz-Roja, C.

    2015-07-01

    In the year 2000 the PWR utilities: Centrales Nucleares Almaraz-Trillo (CNAT) and Asociacion Nuclear Asco-Vandellos (ANAV), and ENUSA Industrias Avanzadas developed and executed a coordinated strategy named PIC (standing for Coordinated Research Program), for achieving the highest level of fuel reliability. The paper will present the scope and results of this program along the years and will summarize the way the changes are managed to ensure fuel integrity. The excellent performance of the ENUSA manufactured fuel in the PWR Spanish NPPs is the best indicator that the expectations on this program are being met. (Author)

  17. The Large-Scale Biosphere-Atmosphere Experiment in Amazonia: Analyzing Regional Land Use Change Effects.

    Science.gov (United States)

    Michael Keller; Maria Assunção Silva-Dias; Daniel C. Nepstad; Meinrat O. Andreae

    2004-01-01

    The Large-Scale Biosphere-Atmosphere Experiment in Amazonia (LBA) is a multi-disciplinary, multinational scientific project led by Brazil. LBA researchers seek to understand Amazonia in its global context especially with regard to regional and global climate. Current development activities in Amazonia including deforestation, logging, cattle ranching, and agriculture...

  18. ECOLOGICAL RESEARCH IN THE LARGE-SCALE BIOSPHERE–ATMOSPHERE EXPERIMENT IN AMAZONIA: EARLY RESULTS.

    Science.gov (United States)

    M. Keller; A. Alencar; G. P. Asner; B. Braswell; M. Bustamente; E. Davidson; T. Feldpausch; E. Fern ndes; M. Goulden; P. Kabat; B. Kruijt; F. Luizao; S. Miller; D. Markewitz; A. D. Nobre; C. A. Nobre; N. Priante Filho; H. Rocha; P. Silva Dias; C von Randow; G. L. Vourlitis

    2004-01-01

    The Large-scale Biosphere–Atmosphere Experiment in Amazonia (LBA) is a multinational, interdisciplinary research program led by Brazil. Ecological studies in LBA focus on how tropical forest conversion, regrowth, and selective logging influence carbon storage, nutrient dynamics, trace gas fluxes, and the prospect for sustainable land use in the Amazon region. Early...

  19. Deepening, Broadening and Scaling up: a Framework for Steering Transition Experiments. : Essay 02

    NARCIS (Netherlands)

    S.J.M. van den Bosch (Suzanne); J. Rotmans (Jan)

    2008-01-01

    markdownabstractThis essay presents a conceptual framework for analyzing and influencing the contribution of small-scale experiments to transitions towards a more sustainable society. This framework is aimed at providing academics and practitioners with a theoretical and practice oriented

  20. Ecological research in the large-scale biosphere-atmosphere experiment in Amazonia: early results

    NARCIS (Netherlands)

    Keller, M.; Alencar, A.; Asner, G.P.; Braswell, B.; Bustamante, M.; Davidson, E.; Feldpausch, T.; Fernandes, E.; Goulden, M.; Kabat, P.; Kruijt, B.; Luizão, F.; Miller, S.; Markewitz, D.; Nobre, A.D.; Nobre, C.A.; Priante Filho, N.; Rocha, da H.; Silva Dias, P.; Randow, von C.; Vourlitis, G.L.

    2004-01-01

    The Large-scale Biosphere-Atmosphere Experiment in Amazonia (LBA) is a multinational, interdisciplinary research program led by Brazil. Ecological studies in LBA focus on how tropical forest conversion, regrowth, and selective logging influence carbon storage,. nutrient dynamics, trace gas fluxes,

  1. Flux ropes and 3D dynamics in the relaxation scaling experiment

    International Nuclear Information System (INIS)

    Intrator, T P; Feng, Y; Weber, T E; Swan, H O; Sun, X; Dorf, L; Sears, J A

    2013-01-01

    Flux ropes form basic building blocks for magnetic dynamics in many plasmas, are macroscopic analogues of magnetic field lines, and are irreducibly three dimensional (3D). We have used the relaxation scaling experiment (RSX) to study flux ropes, and have found many new features involving 3D dynamics, kink instability driven reconnection, nonlinearly stable but kinking flux ropes, and large flows. (paper)

  2. Database of full-scale laboratory experiments on wave-driven sand transport processes

    NARCIS (Netherlands)

    van der Werf, Jebbe J.; Schretlen, Johanna Lidwina Maria; Ribberink, Jan S.; O'Donoghue, Tom

    2009-01-01

    A new database of laboratory experiments involving sand transport processes over horizontal, mobile sand beds under full-scale non-breaking wave and non-breaking wave-plus-current conditions is described. The database contains details of the flow and bed conditions, information on which quantities

  3. An Examination and Validation of an Adapted Youth Experience Scale for University Sport

    Science.gov (United States)

    Rathwell, Scott; Young, Bradley W.

    2016-01-01

    Limited tools assess positive development through university sport. Such a tool was validated in this investigation using two independent samples of Canadian university athletes. In Study 1, 605 athletes completed 99 survey items drawn from the Youth Experience Scale (YES 2.0), and separate a priori measurement models were evaluated (i.e., 99…

  4. Data scaling and temperature calibration in time-resolved photocrystallographic experiments

    DEFF Research Database (Denmark)

    Schmøkel, Mette Stokkebro; Kaminski, Radoslaw; Benedict, Jason B.

    2010-01-01

    -steady-state experiments conducted at conventional sources, but not negligible in synchrotron studies in which very short laser exposures may be adequate. The relative scaling of the light-ON and light-OFF data and the correction for temperature differences between the two sets are discussed....

  5. Low concentration NP preoxidation condition for PWR decontamination

    International Nuclear Information System (INIS)

    Huang Fuduan; Yu Degui; Lu Jingju; Ding Dejun; Zhao Yukun

    1991-02-01

    To use preoxidation condition with low concentration NP (nitric acid permanganate) instead of conventional high concentration AP (alkline permanganate ) for PWR oxidation decontamination (POD) was summarized. Experiments including three parts have been performed. The defilming performance and decontamination factor of preoxidation with low concentration NP, which is 100, 10 times lower than that of AP are better than that with high concentration AP. The reason has been studied with the aid of prefilmed specimens of corrosion potential measuring in NP solution and chromium release in NP and AP solutions. The behaviour of alloy 13 prefilmed specimen in NP preoxidation solution is different from 18-8 ss and Incoloy 800. In the low acidity, the corrosion potential moves toward positive direction as the acidity becomes high

  6. Experimental study on reflooding in advanced tight lattice PWR

    International Nuclear Information System (INIS)

    Hori, K.; Kodama, J.; Teramae, T.

    2000-01-01

    This paper is related to the experimental study on the feasibility of core cooling by re-flooding in a large break loss of coolant accident (LOCA) for the advanced tight lattice pressurized water reactor (PWR). The tight lattice core design should be adopted to improve the conversion ratio. Major one of the key questions of such tight lattice core is the cooling capability under the re-flood condition in a large break LOCA. Forced feed bottom re-flooding experiments have been performed by use of a 4x4 triangular array rod bundle. The rod gap is 0.5 mm, 1.0 mm, or 1.5 mm. The measured peak temperature is below around 1273 K even in case of 1.0/0.5 mm rod gap. And, the evaluation based on the experimental results of rod temperatures and core pressure drop also shows that the core cooling under re-flooding condition is feasible. (author)

  7. Toward an early detection of PWR control rod anomalous dropping

    International Nuclear Information System (INIS)

    Blazquez, J.; Vallejo, I.

    1998-01-01

    Some anomalous PWR control rods dropping occurred in the past. It is assumed to be caused by a geometrical deformation of its guide tube, which might be related with neutron fluence and its sharp changes. Now at days, this problem is an open field of research, oriented to the understanding and prevention of the event. Work here is focused toward early detection. A differential equation modelling control rod free fall movement is found. There result three acceleration terms: gravity; friction with fluid; and friction with its guide tube. From recorded Plant measurements, both friction coefficients are estimated. The one from guide tube experiences a large variation in case of anomalous dropping; so relationship with neutron fluence is proposed for the prevention purpose. (Author)

  8. Optimization of the decontamination in EDF PWR power plants

    International Nuclear Information System (INIS)

    Gosset, P.; Dupin, M.; Buisine, D.; Buet, J.F.; Brunel, V.

    2002-01-01

    The optimisation of decontamination in EDF PWR power plants is the result of a permanent collaborative work between the plant operators, the subcontractors, central services of nuclear power division of EDF. This collaborative work enables the saving of all the feedback experience. The main operations carried out on nuclear sites like mechanical decontamination of valves, use of the ''EMMAC'' process on big components (replacement of steam generator, hydraulic parts of the reactor coolant pumps), use of foam on pools walls and divers in highly contaminated pools have been discussed. This paper shows that the choice of decontamination processes is very dependant on the components, on the dose rate reduction to be aimed and on the possibility to treat the waste on site. (authors)

  9. Neutronal aspects of PWR control for transient load following

    International Nuclear Information System (INIS)

    Cossic, A.

    1985-01-01

    The purpose of this thesis is to qualify the CRONOS diffusion code on a load transient in grey mode control. First of all, we have established a general axial calculational model and studied the important physical phenomena: xenon oscillation, grey rods absorption, radial leaks modelling, effect of the initial conditions in Iodine and Xenon. In a second stage, a three dimensional calculation has been performed, the results of which have been compared to a PWR 900 TRICASTIN 3 experiment and have been in good agreement. In the last part, we show that the results of the axial model using one-dimensional CRONOS calculations are quite consistent with the three-dimensional calculation [fr

  10. PWR-blowdown heat transfer separate effects program

    International Nuclear Information System (INIS)

    Thomas, D.G.

    1976-01-01

    The ORNL Pressurized-Water Reactor Blowdown Heat Transfer (PWR-BDHT) Program is an experimental separate-effects study of the relations among the principal variables that can alter the rate of blowdown, the presence of flow reversal and rereversal, time delay to critical heat flux, the rate at which dryout progresses, and similar time-related functions that are important to LOCA analysis. Primary test results are obtained from the Thermal-Hydraulic Test Facility (THTF). Supporting experiments are carried out in several additional test loops - the Forced Convection Test Facility (FCTF), an air-water loop, a transient steam-water loop, and a low-temperature water mockup of the THTF heater rod bundle. The studies to date are described

  11. Behaviour of organic iodides under pwr accident conditions

    International Nuclear Information System (INIS)

    Alm, M.

    1982-01-01

    Laboratory experiments were performed to study the behaviour of radioactive methyl iodide under PWR loss-of-coolant conditions. The pressure relief equipment consisted of an autoclave for simulating the primary circuit and of an expansion vessel for simulating the conditions after a rupture in the reactor coolant system. After pressure relief, the composition of the CH 3 sup(127/131)I-containing steam-air mixture within the expansion vessel was analysed at 80 0 C over a period of 42 days. On the basis of the values measured and of data taken from the literature, both qualitative and quantitative assessments have been made as to the behaviour of radioactive methyl iodide in the event of loss-of-coolant accidents. (author)

  12. Design of large steam turbines for PWR power stations

    International Nuclear Information System (INIS)

    Hobson, G.

    1984-01-01

    The authors review the thermodynamic cycle requirements for use with pressurized-water reactors, outline the way thermal efficiency is maximized, and discuss the special nature of the wet-steam cycle associated with turbines for this type of reactor. Machine and cycle parameters are optimized to achieve high thermal efficiency, particular attention being given to arrangements for water separation and steam reheating and to provisions for feedwater heating. Principles and details of mechanical design are considered for a range both of full-speed turbines running at 3000 rev/min on 50 Hz systems and of half-speed turbines running at 1800 rev/min on 60 Hz systems. The importance of service experience with nuclear wet-stream turbines, and its relevance to the design of modern turbines for PWR applications, is discussed. (author)

  13. PWR station blackout transient simulation in the INER integral system test facility

    International Nuclear Information System (INIS)

    Liu, T.J.; Lee, C.H.; Hong, W.T.; Chang, Y.H.

    2004-01-01

    Station blackout transient (or TMLB' scenario) in a pressurized water reactor (PWR) was simulated using the INER Integral System Test Facility (IIST) which is a 1/400 volumetrically-scaled reduce-height and reduce-pressure (RHRP) simulator of a Westinghouse three-loop PWR. Long-term thermal-hydraulic responses including the secondary boil-off and the subsequent primary saturation, pressurization and core uncovery were simulated based on the assumptions of no offsite and onsite power, feedwater and operator actions. The results indicate that two-phase discharge is the major depletion mode since it covers 81.3% of the total amount of the coolant inventory loss. The primary coolant inventory has experienced significant re-distribution during a station blackout transient. The decided parameter to avoid the core overheating is not the total amount of the coolant inventory remained in the primary core cooling system but only the part of coolant left in the pressure vessel. The sequence of significant events during transient for the IIST were also compared with those of the ROSA-IV large-scale test facility (LSTF), which is a 1/48 volumetrically-scaled full-height and full-pressure (FHFP) simulator of a PWR. The comparison indicates that the sequence and timing of these events during TMLB' transient studied in the RHRP IIST facility are generally consistent with those of the FHFP LSTF. (author)

  14. Strengthening scaling up through learning from implementation: comparing experiences from Afghanistan, Bangladesh and Uganda.

    Science.gov (United States)

    Bennett, Sara; Mahmood, Shehrin Shaila; Edward, Anbrasi; Tetui, Moses; Ekirapa-Kiracho, Elizabeth

    2017-12-28

    Many effective innovations and interventions are never effectively scaled up. Implementation research (IR) has the promise of supporting scale-up through enabling rapid learning about the intervention and its fit with the context in which it is implemented. We integrate conceptual frameworks addressing different dimensions of scaling up (specifically, the attributes of the service or innovation being scaled, the actors involved, the context, and the scale-up strategy) and questions commonly addressed by IR (concerning acceptability, appropriateness, adoption, feasibility, fidelity to original design, implementation costs, coverage and sustainability) to explore how IR can support scale-up. We draw upon three IR studies conducted by Future Health Systems (FHS) in Afghanistan, Bangladesh and Uganda. We reviewed project documents from the period 2011-2016 to identify information related to the dimensions of scaling up. Further, for each country, we developed rich descriptions of how the research teams approached scaling up, and how IR contributed to scale-up. The rich descriptions were checked by FHS research teams. We identified common patterns and differences across the three cases. The three cases planned quite different innovations/interventions and had very different types of scale-up strategies. In all three cases, the research teams had extensive prior experience within the study communities, and little explicit attention was paid to contextual factors. All three cases involved complex interactions between the research teams and other stakeholders, among stakeholders, and between stakeholders and the intervention. The IR planned by the research teams focussed primarily on feasibility and effectiveness, but in practice, the research teams also had critical insights into other factors such as sustainability, acceptability, cost-effectiveness and appropriateness. Stakeholder analyses and other project management tools further complemented IR. IR can provide

  15. Assessment of void swelling in austenitic stainless steel PWR core internals

    International Nuclear Information System (INIS)

    Chung, H.M.

    2006-01-01

    As many pressurized water reactors (PWRs) age and life extension of the aged plants is considered, void swelling behavior of austenitic stainless steel (SS) core internals has become the subject of increasing attention. In this report, the available database on void swelling and density change of austenitic SSs was critically reviewed. Irradiation conditions, test procedures, and microstructural characteristics were carefully examined, and key factors that are important to determine the relevance of the database to PWR conditions were evaluated. Most swelling data were obtained from steels irradiated in fast breeder reactors at temperatures >385 C and at dose rates that are orders of magnitude higher than PWR dose rates. Even for a given irradiation temperature and given steel, the integral effects of dose and dose rate on void swelling should not be separated. It is incorrect to extrapolate swelling data on the basis of 'progressive compounded multiplication' of separate effects of factors such as dose, dose rate, temperature, steel composition, and fabrication procedure. Therefore, the fast reactor data should not be extrapolated to determine credible void swelling behavior for PWR end-of-life (EOL) or life-extension conditions. Although the void swelling data extracted from fast reactor studies is extensive and conclusive, only limited amounts of swelling data and information have been obtained on microstructural characteristics from discharged PWR internals or steels irradiated at temperatures and at dose rates comparable to those of a PWR. Based on this relatively small amount of information, swelling in thin-walled tubes and baffle bolts in a PWR is not considered a concern. As additional data and relevant research becomes available, the newer results should be integrated with existing data, and the worthiness of this conclusion should continue to be scrutinized. PWR baffle reentrant corners are the most likely location to experience high swelling rates, and

  16. Test requirements for the integral effect test to simulate Korean PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Song, Chul Hwa; Park, C. K.; Lee, S. J.; Kwon, T. S.; Yun, B. J.; Chung, M. K

    2001-02-01

    In this report, the test requirements are described for the design of the integral effect test facility to simulate Korean PWR plants. Since the integral effect test facility should be designed so as to simulate various thermal hydraulic phenomena, as closely as possible, to be occurred in real plants during operation or anticipated transients, the design and operational characteristics of the reference plants (Korean Standard Nuclear Plant and Korean Next Generation Reactor)were analyzed in order to draw major components, systems, and functions to be satisfied or simulated in the test facility. The test matrix is set up by considering major safety concerns of interest and the test objectives to confirm and enhance the safety of the plants. And the analysis and prioritization of the test matrix leads to the general design requirements of the test facility. Based on the general design requirements, the design criteria is set up for the basic and detailed design of the test facility. And finally it is drawn the design requirements specific to the fluid system and measurement system of the test facility. The test requirements in this report will be used as a guideline to the scaling analysis and basic design of the test facility. The test matrix specified in this report can be modified in the stage of main testing by considering the needs of experiments and circumstances at that time.

  17. Test requirements for the integral effect test to simulate Korean PWR plants

    International Nuclear Information System (INIS)

    Song, Chul Hwa; Park, C. K.; Lee, S. J.; Kwon, T. S.; Yun, B. J.; Chung, M. K.

    2001-02-01

    In this report, the test requirements are described for the design of the integral effect test facility to simulate Korean PWR plants. Since the integral effect test facility should be designed so as to simulate various thermal hydraulic phenomena, as closely as possible, to be occurred in real plants during operation or anticipated transients, the design and operational characteristics of the reference plants (Korean Standard Nuclear Plant and Korean Next Generation Reactor)were analyzed in order to draw major components, systems, and functions to be satisfied or simulated in the test facility. The test matrix is set up by considering major safety concerns of interest and the test objectives to confirm and enhance the safety of the plants. And the analysis and prioritization of the test matrix leads to the general design requirements of the test facility. Based on the general design requirements, the design criteria is set up for the basic and detailed design of the test facility. And finally it is drawn the design requirements specific to the fluid system and measurement system of the test facility. The test requirements in this report will be used as a guideline to the scaling analysis and basic design of the test facility. The test matrix specified in this report can be modified in the stage of main testing by considering the needs of experiments and circumstances at that time

  18. Experimental study of effect of initial clad temperature on reflood phenomena during PWR-LOCA

    International Nuclear Information System (INIS)

    Sugimoto, Jun; Murao, Yoshio

    1983-01-01

    Integral system tests with the Cylindrical Core Test Facility (CCTF) were performed to investigate the effect of the initial clad temperature on the reflood phenomena in a PWR-LOCA. The initial peak clad temperatures in these three tests were 871, 968 and 1,047K, respectively. The feedback of the system on the core inlet mass flow rate was estimated to be little influenced by the variation of the initial clad temperature except for the first 20s in the transient. The observed temperature rise from the reflood initiation was lower with the higher initial clad temperature. This qualitatively agreed with the results of the small scale forced feed reflood experiments. However, the magnitude of the temperature rise in CCTF was significantly low due to the high initial core inlet mass flow rate. Also observed were the multi-dimensional thermal behaviors for the three cases in the CCTF wide core. The analysis codes REFLA and TRAC reasonably predicted the effect of the initial clad temperature on the core thermo-hydraulics under the simulated core inlet flow conditions. However, the calculated temperature rise of the maximum powered rod based on the one-dimensional core analysis was higher than that of the average powered rod, which contradicts the tendency observed in CCTF tests. (author)

  19. Boron mixing transients in a 900 MW PWR vessel for a reactor start-up operation

    International Nuclear Information System (INIS)

    Alvarez, D.; Martin, A.; Schneider, J.P.

    1995-01-01

    In 1991 a R and D action, based on numerical simulations and experiments on PWRs'S primary coolant temperature or boron mixing capabilities, was initiated. This paper presents the test facility BORA-BORA (a 1/5th scaled mock-up of a 900 MW PWR vessel) and the Thermalhydraulic Finite Element Code N3S used for 3D calculations performed on the accurate geometry of the plant. As a validation test case of these experimental and numerical tools, we present the results obtained on the primary coolant mixing capabilities in the vessel with the three loops balanced in mass flow rate. The second part of this report deals with the mixing of a clear water plug in the vessel when a primary coolant pump start-up. The results are obtained in the mock-up in terms of boron concentration at the core inlet for several clear water plug volumes. The numerical results give the complete fluid flow and boron concentration patterns but comparisons were made at the core inlet. (author). 15 refs., 9 figs., 1 tab

  20. A large-scale forest fragmentation experiment: the Stability of Altered Forest Ecosystems Project

    Science.gov (United States)

    Ewers, Robert M.; Didham, Raphael K.; Fahrig, Lenore; Ferraz, Gonçalo; Hector, Andy; Holt, Robert D.; Kapos, Valerie; Reynolds, Glen; Sinun, Waidi; Snaddon, Jake L.; Turner, Edgar C.

    2011-01-01

    Opportunities to conduct large-scale field experiments are rare, but provide a unique opportunity to reveal the complex processes that operate within natural ecosystems. Here, we review the design of existing, large-scale forest fragmentation experiments. Based on this review, we develop a design for the Stability of Altered Forest Ecosystems (SAFE) Project, a new forest fragmentation experiment to be located in the lowland tropical forests of Borneo (Sabah, Malaysia). The SAFE Project represents an advance on existing experiments in that it: (i) allows discrimination of the effects of landscape-level forest cover from patch-level processes; (ii) is designed to facilitate the unification of a wide range of data types on ecological patterns and processes that operate over a wide range of spatial scales; (iii) has greater replication than existing experiments; (iv) incorporates an experimental manipulation of riparian corridors; and (v) embeds the experimentally fragmented landscape within a wider gradient of land-use intensity than do existing projects. The SAFE Project represents an opportunity for ecologists across disciplines to participate in a large initiative designed to generate a broad understanding of the ecological impacts of tropical forest modification. PMID:22006969

  1. A large-scale forest fragmentation experiment: the Stability of Altered Forest Ecosystems Project.

    Science.gov (United States)

    Ewers, Robert M; Didham, Raphael K; Fahrig, Lenore; Ferraz, Gonçalo; Hector, Andy; Holt, Robert D; Kapos, Valerie; Reynolds, Glen; Sinun, Waidi; Snaddon, Jake L; Turner, Edgar C

    2011-11-27

    Opportunities to conduct large-scale field experiments are rare, but provide a unique opportunity to reveal the complex processes that operate within natural ecosystems. Here, we review the design of existing, large-scale forest fragmentation experiments. Based on this review, we develop a design for the Stability of Altered Forest Ecosystems (SAFE) Project, a new forest fragmentation experiment to be located in the lowland tropical forests of Borneo (Sabah, Malaysia). The SAFE Project represents an advance on existing experiments in that it: (i) allows discrimination of the effects of landscape-level forest cover from patch-level processes; (ii) is designed to facilitate the unification of a wide range of data types on ecological patterns and processes that operate over a wide range of spatial scales; (iii) has greater replication than existing experiments; (iv) incorporates an experimental manipulation of riparian corridors; and (v) embeds the experimentally fragmented landscape within a wider gradient of land-use intensity than do existing projects. The SAFE Project represents an opportunity for ecologists across disciplines to participate in a large initiative designed to generate a broad understanding of the ecological impacts of tropical forest modification.

  2. Behavior of a nine-rod PWR bundle under power-cooling-mismatch conditions

    International Nuclear Information System (INIS)

    Gunnerson, F.S.; Sparks, D.T.

    1979-01-01

    An experiment to characterize the behavior of a nine-rod pressurized water reactor (PWR) fuel bundle operating during power-cooling-mismatch (PCM) conditions has been conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory (INEL). The experiment, designated Test PCM-5, is part of a series of PCM experiments designed to evaluate light water reactor (LWR) fuel rod response under postulated accident conditions. Test PCM-5 was the first nine-rod bundle experiment in the PCM test series. The primary objectives and the results of the experiment are described

  3. SCALE system cross-section validation for criticality safety analysis

    International Nuclear Information System (INIS)

    Hathout, A.M.; Westfall, R.M.; Dodds, H.L. Jr.

    1980-01-01

    The purpose of this study is to test selected data from three cross-section libraries for use in the criticality safety analysis of UO 2 fuel rod lattices. The libraries, which are distributed with the SCALE system, are used to analyze potential criticality problems which could arise in the industrial fuel cycle for PWR and BWR reactors. Fuel lattice criticality problems could occur in pool storage, dry storage with accidental moderation, shearing and dissolution of irradiated elements, and in fuel transport and storage due to inadequate packing and shipping cask design. The data were tested by using the SCALE system to analyze 25 recently performed critical experiments

  4. Experiment using laboratory scale extruder. Fluid behavior in twin-screw extruder

    International Nuclear Information System (INIS)

    Suzuki, Hiroshi; Miura, Akihiko

    1999-09-01

    All evidences and chemical data suggest non-chemical heating mechanism raised the filling temperature of the bituminized product. But they indicate the filling temperature was higher than before at the incident. We estimated the physical heat mechanism in the extruder. It is well known that the viscous-heating occurs in mixing process in extruders. In order to confirm the behavior of the torque and temperature, some experiment using laboratory scale extruder were performed. The result of the experiment using laboratory scale extruder showed that the phenomena of salt enrichment and salt accumulation were observed and they raised mixture temperature at the decreased feed rate. These phenomena depend on the feed rate. It is considered that they have large contribution to heat transportation and operational torque due to the friction between screw and mixture. In this report, all experiment result are explained. (author)

  5. Plasma performance and scaling laws in the RFX-mod reversed-field pinch experiment

    International Nuclear Information System (INIS)

    Innocente, P.; Alfier, A.; Canton, A.; Pasqualotto, R.

    2009-01-01

    The large range of plasma currents (I p = 0.2-1.6 MA) and feedback-controlled magnetic boundary conditions of the RFX-mod experiment make it well suited to performing scaling studies. The assessment of such scaling, in particular those on temperature and energy confinement, is crucial both for improving the operating reversed-field pinch (RFP) devices and for validating the RFP configuration as a candidate for the future fusion reactors. For such a purpose scaling laws for magnetic fluctuations, temperature and energy confinement have been evaluated in stationary operation. RFX-mod scaling laws have been compared with those obtained from other RFP devices and numerical simulations. The role of the magnetic boundary has been analysed, comparing discharges performed with different active control schemes of the edge radial magnetic field.

  6. Development of the Italian Version of the Near-Death Experience Scale

    Directory of Open Access Journals (Sweden)

    Francesca Pistoia

    2018-02-01

    Full Text Available Near-death experiences (NDEs have been defined as any conscious perceptual experience occurring in individuals pronounced clinically dead or who came very close to physical death. They are frequently reported by patients surviving a critical injury and, intriguingly, they show common features across different populations. The tool traditionally used to assess NDEs is the NDE Scale, which is available in the original English version. The aim of this study was to develop the Italian version of the NDE Scale and to assess its reliability in a specific clinical setting. A process of translation of the original scale was performed in different stages in order to obtain a fully comprehensible and accurate Italian translation. Later, the scale was administered to a convenience sample of patients who had experienced a condition of coma and were, at the time of assessment, fully conscious and able to provide information as requested by the scale. Inter-rater and test–retest reliability, assessed by the weighted Cohen’s kappa (Kw, were estimated. A convenience sample of 20 subjects [mean age ± standard deviation (SD 51.6 ± 17.1, median time from injury 3.5 months, interquartile range (IQR 2–10] was included in the study. Inter-rater [Kw 0.77 (95% CI 0.67–0.87] and test–retest reliability [Kw 0.96 (95% CI 0.91–1.00] showed good to excellent values for the total scores of the Italian NDE Scale and for subanalyses of each single cluster of the scale. An Italian Version of the NDE Scale is now available to investigate the frequency of NDE, the causes for NDE heterogeneity across different life-threatening conditions, and the possible neural mechanisms underlying NDE phenomenology.

  7. Study on thermo-hydraulic behavior during reflood phase of a PWR-LOCA

    International Nuclear Information System (INIS)

    Sugimoto, Jun

    1989-01-01

    This paper describes thermo-hydraulic behavior during the reflood phase in a postulated large-break loss-of-coolant accident (LOCA) of a PWR. In order to better predict the reflood transient in a nuclear safety analysis specific analytical models have been developed for, saturated film boiling heat transfer in inverted slung flow, the effect of grid spacers on core thermo-hydraulics, overall system thermo-hydraulic behavior, and the thermal response similarity between nuclear fuel rods and simulated rods. A heat transfer correlation has been newly developed for saturated film boiling based on a 4 x 4-rod experiment conducted at JAERI. The correlation provides a good agreement with existing experiments except in the vicinity of grid spacer locations. An analytical model has then been developed addressing the effect of grid spacers. The thermo-hydraulic behavior near the grid spacers was found to be predicted well with this model by considering the breakup of droplets in dispersed flow and water accumulation above the grid spacers in inverted slung flow. A system analysis code has been developed which couples the one-dimensional core and multi-loop primary system component models. It provides fairly good agreement with system behavior obtained in a large-scale integral reflood experiment with active primary system components. An analytical model for the radial temperature distribution in a rod has been developed and verified with data from existing experiments. It was found that a nuclear fuel rod has a lower cladding temperature and an earlier quench time than an electrically heated rod in a typical reflood condition. (author)

  8. Origins and Scaling of Hot-Electron Preheat in Ignition-Scale Direct-Drive Inertial Confinement Fusion Experiments.

    Science.gov (United States)

    Rosenberg, M J; Solodov, A A; Myatt, J F; Seka, W; Michel, P; Hohenberger, M; Short, R W; Epstein, R; Regan, S P; Campbell, E M; Chapman, T; Goyon, C; Ralph, J E; Barrios, M A; Moody, J D; Bates, J W

    2018-02-02

    Planar laser-plasma interaction (LPI) experiments at the National Ignition Facility (NIF) have allowed access for the first time to regimes of electron density scale length (∼500 to 700  μm), electron temperature (∼3 to 5 keV), and laser intensity (6 to 16×10^{14}  W/cm^{2}) that are relevant to direct-drive inertial confinement fusion ignition. Unlike in shorter-scale-length plasmas on OMEGA, scattered-light data on the NIF show that the near-quarter-critical LPI physics is dominated by stimulated Raman scattering (SRS) rather than by two-plasmon decay (TPD). This difference in regime is explained based on absolute SRS and TPD threshold considerations. SRS sidescatter tangential to density contours and other SRS mechanisms are observed. The fraction of laser energy converted to hot electrons is ∼0.7% to 2.9%, consistent with observed levels of SRS. The intensity threshold for hot-electron production is assessed, and the use of a Si ablator slightly increases this threshold from ∼4×10^{14} to ∼6×10^{14}  W/cm^{2}. These results have significant implications for mitigation of LPI hot-electron preheat in direct-drive ignition designs.

  9. Origins and Scaling of Hot-Electron Preheat in Ignition-Scale Direct-Drive Inertial Confinement Fusion Experiments

    Science.gov (United States)

    Rosenberg, M. J.; Solodov, A. A.; Myatt, J. F.; Seka, W.; Michel, P.; Hohenberger, M.; Short, R. W.; Epstein, R.; Regan, S. P.; Campbell, E. M.; Chapman, T.; Goyon, C.; Ralph, J. E.; Barrios, M. A.; Moody, J. D.; Bates, J. W.

    2018-01-01

    Planar laser-plasma interaction (LPI) experiments at the National Ignition Facility (NIF) have allowed access for the first time to regimes of electron density scale length (˜500 to 700 μ m ), electron temperature (˜3 to 5 keV), and laser intensity (6 to 16 ×1014 W /cm2 ) that are relevant to direct-drive inertial confinement fusion ignition. Unlike in shorter-scale-length plasmas on OMEGA, scattered-light data on the NIF show that the near-quarter-critical LPI physics is dominated by stimulated Raman scattering (SRS) rather than by two-plasmon decay (TPD). This difference in regime is explained based on absolute SRS and TPD threshold considerations. SRS sidescatter tangential to density contours and other SRS mechanisms are observed. The fraction of laser energy converted to hot electrons is ˜0.7 % to 2.9%, consistent with observed levels of SRS. The intensity threshold for hot-electron production is assessed, and the use of a Si ablator slightly increases this threshold from ˜4×10 14 to ˜6 ×1014 W /cm2 . These results have significant implications for mitigation of LPI hot-electron preheat in direct-drive ignition designs.

  10. Development of the vacuum drying process for the PWR spent nuclear fuel dry storage

    Energy Technology Data Exchange (ETDEWEB)

    Baeg, Chagn Yeal; Cho, Chun Hyung [Korea Radioactive Waste Agency, Daejeon (Korea, Republic of)

    2016-12-15

    This paper describes the development of a dry operation process for PWR spent nuclear fuel, which is currently stored in the domestic NPP's storage pool, using a dual purpose metal cask. Domestic NNPs have had experience with wet type transportation of PWR spent nuclear fuel between neighboring NPPs since the early 1990s, but no experience with dry type operation. For this reason, we developed a specific operation process and also confirmed the safety of the major cask components and its spent nuclear fuel during the dual purpose metal cask operation process. We also describe the short term operation process that was established to be completed within 21 hours and propose the allowable working time for each step (15 hours for wet process, 3 hours for drain process and 3 hours for vacuum drying process)

  11. Computer codes for the study of the loss of coolant accident of PWR reactors

    International Nuclear Information System (INIS)

    Gomolinski, M.; Menessier, D.; Tellier, N.

    1975-01-01

    The CEA has undertaken a large programme to study the consequence on the core of the LOCA of a PWR. In the programme, simultaneously carried out experiments and the development of the calculations means are described. Several experiments such as OMEGA, ERSEC and PHEBUS tests, which provide data to check the computer codes are outlined briefly in the paper. For analysis of the LOCA of a PWR, a series of computer codes, which are at present in use or under development, are linked with each other. The codes are DANAIDES for blowdown, CERES for refill and reflood, THETA-1B and FLIRA for heat up calculation during the blow-down and the reflooding period respectively. FLIRA-PASTEL, a combination of FLIRA and PASTEL which calculate the stress and deformations of material using the finite element method, will be used in place of FLIRA. The basic models and flowcharts of the above codes are described in the paper

  12. Steam turbines for PWR stations

    International Nuclear Information System (INIS)

    Muscroft, J.

    1989-01-01

    The thermodynamic cycle requirements and mechanical design features applying to modern GEC 3000 rev/min steam turbines for pressurised water reactor power stations are reviewed. The most recent developments include machines of 630 MW and 985 MW output which are currently under construction. The importance of service experience with nuclear wet steam turbines associated with a variety of types of water cooled reactor and its relevance to the design of modern 3000 rev/min turbines for pressurised water reactor applications is emphasised. (author)

  13. PWR vessel inspection performance improvements

    International Nuclear Information System (INIS)

    Blair Fairbrother, D.; Bodson, Francis

    1998-01-01

    A compact robot for ultrasonic inspection of reactor vessels has been developed that reduces setup logistics and schedule time for mandatory code inspections. Rather than installing a large structure to access the entire weld inspection area from its flange attachment, the compact robot examines welds in overlapping patches from a suction cup anchor to the shell wall. The compact robot size allows two robots to be operated in the vessel simultaneously. This significantly reduces the time required to complete the inspection. Experience to date indicates that time for vessel examinations can be reduced to fewer than four days. (author)

  14. Zinc injection in German PWR plants

    International Nuclear Information System (INIS)

    Streit, K.

    2004-01-01

    Operating experience acquired at PWR NNPs shows that zinc injection at low concentrations of 5 ppb is a very effective source term reduction measure. This method does not lead to any operating restrictions or other negative effects on plant systems and components. The nuclear industry has been very successful in reducing radiation exposures within the past two decades. Annual exposures could be significantly decreased and are now at a level of around 1 man-Sv per plant and year. This great success can mainly be attributed to the general commitment of plant operators to maintaining radiation exposures of workers in the controlled access area as low as reasonably achievable (ALARA principle). The ALARA principle, of course, also implies evaluation of the economic benefit of radiation protection measures. Radiation source term reduction has drawn increasing attention of plant operators in recent years. For the new PWRs cobalt-based alloys in the primary system have successively been eliminated already at the design and construction phase within the last decade. Use of wear-resistant cobalt-free substitute materials in combination with the general use of advanced alloys for the steam generator tubing of PWRs resulted in low values for the two most common sources of plant radiation fields, namely 58 Co and 60 Co. Investigations showed that the beneficial effect of zinc can be related to its high affinity for mixed spinel oxide phases, resulting in the following two basic effects: -Zinc is incorporated preferentially into the oxide layer on primary system surfaces and thus reduces pickup of 58 Co and 60 Co and - Zinc can displace cobalt isotopes from existing oxide layers. In German PWRs with Incoloy 800 steam generator tubing material (Ni-content -32%), the observed reductions correspond to a decrease in dose rates of around 10 to 15% per year and thus follow, as predicted, the half-life time of 60 Co. Overall reductions in high radiation areas are now in the range of

  15. The Work Experience Measurement Scale (WEMS): a useful tool in workplace health promotion.

    Science.gov (United States)

    Nilsson, Petra; Andersson, H Ingemar; Ejlertsson, Göran

    2013-01-01

    To present validity data for the Work Experience Measurement Scale (WEMS), an instrument measuring multifaceted work experience from a salutogenic health resource perspective as a contrast to the more common pathogenic risk perspective, by exploring WEMS relationship to established measurements that are positively related to health and work. A salutogenic perspective focuses on finding conditions and resources in life, for example at work, that can enhance the individual's health and strength, instead of those causing illness and weakness. This study was carried out in 2009 at a Swedish hospital with a web-based survey (WEMS) to 770 employees. Different occupational groups at the hospital participated. Additional questionnaires used at the same time were the Utrecht Work Engagement Scale (UWES-9), the Salutogenic Health Indicator Scale (SHIS), the General Self-Efficacy scale (GSE), and three questions about self-rated health, general well-being, and quality of life. Cronbach's Alpha of WEMS sub-indices were in the interval of 0.85-0.96. Convergent validity and discriminant validity of WEMS and its sub-indices were shown to be satisfying by correlations. In addition, WEMS demonstrated the ability to discriminate between groups. WEMS sub-indices discriminated even better between groups than the total index. The WEMS proved to be a workplace health promotion questionnaire that was able to measure experiences of work from a salutogenic perspective. The WEMS has a potential of being a useful tool in workplace health promotion to enhance positive human capabilities and resources to improve work performance.

  16. Scale for positive aspects of caregiving experience: development, reliability, and factor structure.

    Science.gov (United States)

    Kate, N; Grover, S; Kulhara, P; Nehra, R

    2012-06-01

    OBJECTIVE. To develop an instrument (Scale for Positive Aspects of Caregiving Experience [SPACE]) that evaluates positive caregiving experience and assess its psychometric properties. METHODS. Available scales which assess some aspects of positive caregiving experience were reviewed and a 50-item questionnaire with a 5-point rating was constructed. In all, 203 primary caregivers of patients with severe mental disorders were asked to complete the questionnaire. Internal consistency, test-retest reliability, cross-language reliability, split-half reliability, and face validity were evaluated. Principal component factor analysis was run to assess the factorial validity of the scale. RESULTS. The scale developed as part of the study was found to have good internal consistency, test-retest reliability, cross-language reliability, split-half reliability, and face validity. Principal component factor analysis yielded a 4-factor structure, which also had good test-retest reliability and cross-language reliability. There was a strong correlation between the 4 factors obtained. CONCLUSION. The SPACE developed as part of this study has good psychometric properties.

  17. Particle generation methods applied in large-scale experiments on aerosol behaviour and source term studies

    International Nuclear Information System (INIS)

    Swiderska-Kowalczyk, M.; Gomez, F.J.; Martin, M.

    1997-01-01

    In aerosol research aerosols of known size, shape, and density are highly desirable because most aerosols properties depend strongly on particle size. However, such constant and reproducible generation of those aerosol particles whose size and concentration can be easily controlled, can be achieved only in laboratory-scale tests. In large scale experiments, different generation methods for various elements and compounds have been applied. This work presents, in a brief from, a review of applications of these methods used in large scale experiments on aerosol behaviour and source term. Description of generation method and generated aerosol transport conditions is followed by properties of obtained aerosol, aerosol instrumentation used, and the scheme of aerosol generation system-wherever it was available. An information concerning aerosol generation particular purposes and reference number(s) is given at the end of a particular case. These methods reviewed are: evaporation-condensation, using a furnace heating and using a plasma torch; atomization of liquid, using compressed air nebulizers, ultrasonic nebulizers and atomization of liquid suspension; and dispersion of powders. Among the projects included in this worked are: ACE, LACE, GE Experiments, EPRI Experiments, LACE-Spain. UKAEA Experiments, BNWL Experiments, ORNL Experiments, MARVIKEN, SPARTA and DEMONA. The aim chemical compounds studied are: Ba, Cs, CsOH, CsI, Ni, Cr, NaI, TeO 2 , UO 2 Al 2 O 3 , Al 2 SiO 5 , B 2 O 3 , Cd, CdO, Fe 2 O 3 , MnO, SiO 2 , AgO, SnO 2 , Te, U 3 O 8 , BaO, CsCl, CsNO 3 , Urania, RuO 2 , TiO 2 , Al(OH) 3 , BaSO 4 , Eu 2 O 3 and Sn. (Author)

  18. Modelling solute dispersion in periodic heterogeneous porous media: Model benchmarking against intermediate scale experiments

    Science.gov (United States)

    Majdalani, Samer; Guinot, Vincent; Delenne, Carole; Gebran, Hicham

    2018-06-01

    This paper is devoted to theoretical and experimental investigations of solute dispersion in heterogeneous porous media. Dispersion in heterogenous porous media has been reported to be scale-dependent, a likely indication that the proposed dispersion models are incompletely formulated. A high quality experimental data set of breakthrough curves in periodic model heterogeneous porous media is presented. In contrast with most previously published experiments, the present experiments involve numerous replicates. This allows the statistical variability of experimental data to be accounted for. Several models are benchmarked against the data set: the Fickian-based advection-dispersion, mobile-immobile, multirate, multiple region advection dispersion models, and a newly proposed transport model based on pure advection. A salient property of the latter model is that its solutions exhibit a ballistic behaviour for small times, while tending to the Fickian behaviour for large time scales. Model performance is assessed using a novel objective function accounting for the statistical variability of the experimental data set, while putting equal emphasis on both small and large time scale behaviours. Besides being as accurate as the other models, the new purely advective model has the advantages that (i) it does not exhibit the undesirable effects associated with the usual Fickian operator (namely the infinite solute front propagation speed), and (ii) it allows dispersive transport to be simulated on every heterogeneity scale using scale-independent parameters.

  19. Investigation of small break loss-of-coolant phenomena in a small scale nonnuclear test facility

    International Nuclear Information System (INIS)

    Cozzuol, J.M.; Fauble, T.J.; Harvego, E.A.

    1980-01-01

    A small-scale nonnuclear integral test facility designed to simulate a pressurized water reactor (PWR) system was used to evaluate the effects of a small break loss-of-coolant accident (LOCA) on the system thermal-hydraulic response. The experiment approximated a 2.5% (11-cm diameter) communicative break in the cold leg of a PWR, and included initial conditions which were similar to conditions in a PWR operating at full power. The 2.5% break size ensured that the nominal break flow rate was greater than the high pressure injection system (HPIS) flow rate, thus providing the potential for a continuous system depressurization. The sequence of events was similar to that used in evaluation model analysis of small break loss-of-coolant accidents, and included simulated reactor scram and loss of offsite power. Comparisions of experimental data with computer code calculations are used to demonstrate the capabilities and limitations of integral system calculations used to predict phenomena which can be important in the assessment of a small break LOCA in a PWR

  20. Thermal anchoring of wires in large scale superconducting coil test experiment

    International Nuclear Information System (INIS)

    Patel, Dipak; Sharma, A.N.; Prasad, Upendra; Khristi, Yohan; Varmora, Pankaj; Doshi, Kalpesh; Pradhan, S.

    2013-01-01

    Highlights: • We addressed how thermal anchoring in large scale coil test is different compare to small cryogenic apparatus? • We did precise estimation of thermal anchoring length at 77 K and 4.2 K heat sink in large scale superconducting coil test experiment. • We addressed, the quality of anchoring without covering entire wires using Kapton/Teflon tape. • We obtained excellent results in temperature measurement without using GE Varnish by doubling estimated anchoring length. -- Abstract: Effective and precise thermal anchoring of wires in cryogenic experiment is mandatory to measure temperature in milikelvin accuracy and to avoid unnecessary cooling power due to additional heat conduction from room temperature (RT) to operating temperature (OT) through potential, field, displacement and stress measurement instrumentation wires. Instrumentation wires used in large scale superconducting coil test experiments are different compare to cryogenic apparatus in terms of unique construction and overall diameter/area due to errorless measurement in large time-varying magnetic field compare to small cryogenic apparatus, often shielded wires are used. Hence, along with other variables, anchoring techniques and required thermal anchoring length are entirely different in this experiment compare to cryogenic apparatus. In present paper, estimation of thermal anchoring length of five different types of instrumentation wires used in coils test campaign at Institute for Plasma Research (IPR), India has been discussed and some temperature measurement results of coils test campaign have been presented

  1. A large-scale soil-structure interaction experiment: Part I design and construction

    International Nuclear Information System (INIS)

    Tang, H.T.; Tang, Y.K.; Wall, I.B.; Lin, E.

    1987-01-01

    In the simulated earthquake experiments (SIMQUAKE) sponsored by EPRI, the detonation of vertical arrays of explosives propagated wave motions through the ground to the model structures. Although such a simulation can provide information about dynamic soil-structure interaction (SSI) characteristics in a strong motion environment, it lacks seismic wave scattering characteristics for studying seismic input to the soil-structure system and the effect of different kinds of wave composition to the soil-structure response. To supplement the inadequacy of the simulated earthquake SSI experiment, the Electric Power Research Institute (EPRI) and the Taiwan Power Company (Taipower) jointly sponsored a large scale SSI experiment in the field. The objectives of the experiment are: (1) to obtain actual strong motion earthquakes induced database in a soft-soil environment which will substantiate predictive and design SSI models;and (2) to assess nuclear power plant reactor containment internal components dynamic response and margins relating to actual earthquake-induced excitation. These objectives are accomplished by recording and analyzing data from two instrumented, scaled down, (1/4- and 1/12-scale) reinforced concrete containments sited in a high seismic region in Taiwan where a strong-motion seismic array network is located

  2. Reactive flow modeling of small scale detonation failure experiments for a baseline non-ideal explosive

    Energy Technology Data Exchange (ETDEWEB)

    Kittell, David E.; Cummock, Nick R.; Son, Steven F. [School of Mechanical Engineering, Purdue University, West Lafayette, Indiana 47907 (United States)

    2016-08-14

    Small scale characterization experiments using only 1–5 g of a baseline ammonium nitrate plus fuel oil (ANFO) explosive are discussed and simulated using an ignition and growth reactive flow model. There exists a strong need for the small scale characterization of non-ideal explosives in order to adequately survey the wide parameter space in sample composition, density, and microstructure of these materials. However, it is largely unknown in the scientific community whether any useful or meaningful result may be obtained from detonation failure, and whether a minimum sample size or level of confinement exists for the experiments. In this work, it is shown that the parameters of an ignition and growth rate law may be calibrated using the small scale data, which is obtained from a 35 GHz microwave interferometer. Calibration is feasible when the samples are heavily confined and overdriven; this conclusion is supported with detailed simulation output, including pressure and reaction contours inside the ANFO samples. The resulting shock wave velocity is most likely a combined chemical-mechanical response, and simulations of these experiments require an accurate unreacted equation of state (EOS) in addition to the calibrated reaction rate. Other experiments are proposed to gain further insight into the detonation failure data, as well as to help discriminate between the role of the EOS and reaction rate in predicting the measured outcome.

  3. Scaling up community mobilisation through women's groups for maternal and neonatal health: experiences from rural Bangladesh

    Directory of Open Access Journals (Sweden)

    Nahar Tasmin

    2012-01-01

    Full Text Available Abstract Background Program coverage is likely to be an important determinant of the effectiveness of community interventions to reduce neonatal mortality. Rigorous examination and documentation of methods to scale-up interventions and measure coverage are scarce, however. To address this knowledge gap, this paper describes the process and measurement of scaling-up coverage of a community mobilisation intervention for maternal, child and neonatal health in rural Bangladesh and critiques this real-life experience in relation to available literature on scaling-up. Methods Scale-up activities took place in nine unions in rural Bangladesh. Recruitment and training of those who deliver the intervention, communication and engagement with the community and other stakeholders and active dissemination of intervention activities are described. Process evaluation and population survey data are presented and used to measure coverage and the success of scale-up. Results The intervention was scaled-up from 162 women's groups to 810, representing a five-fold increase in population coverage. The proportion of women of reproductive age and pregnant women who were engaged in the intervention increased from 9% and 3%, respectively, to 23% and 29%. Conclusions Examination and documentation of how scaling-up was successfully initiated, led, managed and monitored in rural Bangladesh provide a deeper knowledge base and valuable lessons. Strong operational capabilities and institutional knowledge of the implementing organisation were critical to the success of scale-up. It was possible to increase community engagement with the intervention without financial incentives and without an increase in managerial staff. Monitoring and feedback systems that allow for periodic programme corrections and continued innovation are central to successful scale-up and require programmatic and operational flexibility.

  4. Plot-scale field experiment of surface hydrologic processes with EOS implications

    Science.gov (United States)

    Laymon, Charles A.; Macari, Emir J.; Costes, Nicholas C.

    1992-01-01

    Plot-scale hydrologic field studies were initiated at NASA Marshall Space Flight Center to a) investigate the spatial and temporal variability of surface and subsurface hydrologic processes, particularly as affected by vegetation, and b) develop experimental techniques and associated instrumentation methodology to study hydrologic processes at increasingly large spatial scales. About 150 instruments, most of which are remotely operated, have been installed at the field site to monitor ground atmospheric conditions, precipitation, interception, soil-water status, and energy flux. This paper describes the nature of the field experiment, instrumentation and sampling rationale, and presents preliminary findings.

  5. The main security problems encountered in the definition of the PWR 1300 MWe stage

    International Nuclear Information System (INIS)

    Guimbail, H.; Auvergnon, F.

    1980-01-01

    The main problems of security encountered in the definition of the PWR 1300 stage originated from the necessity to ensure the continuity of a stage with the preceding one, the overture to technical progress, the acceptance of responsibility for the gaines experience and the control of cost prices. A few examples show how desirable it is for a project to be managed with the design and construction rules stabilized as early as possible [fr

  6. Preventive detection of incipient failure and improvement of availability of French PWR using acoustic emission

    International Nuclear Information System (INIS)

    Audenard, B.; Marini, J.

    1982-08-01

    Laboratory tests, on site experience gained on PWR during start up test as well as during nominal functioning have given FRAMATOME very great confidence in A.E. techniques for preventive detection of incidents. Loose part and leakage monitoring are already being used on an industrial basis. Crack growth detection and monitoring are still in the investigation phase and various. Research and Development programs are presently being carried out

  7. AP1000. The PWR revisited

    International Nuclear Information System (INIS)

    Gaio, P.

    2006-01-01

    The distinguishing features of Westinghouse's AP1000 advanced passive pressurized water reactor are highlighted. In particular, the AP1000's passive safety features are described as well as their implications for simplifying the design, construction, and operation of this design compared to currently operating plants, and significantly increasing safety margins over current plants as well. The AP1000 design specifically incorporates the knowledge acquired from the substantial accumulation of power reactor operating experience and benefits from the application of the Probabilistic Risk Assessment in the design process itself. The AP1000 design has been certified by the US Nuclear Regulatory Commission under its new rules for licensing new nuclear plants, 10 CFR Part 52, and is the subject of six combined Construction and Operating License applications now being developed. Currently the AP1000 design is being assessed against the EUR Rev C requirements for new nuclear power plants in Europe. (author)

  8. Study of anticipated transient without scram for PWR

    International Nuclear Information System (INIS)

    Pu Jilong.

    1985-01-01

    Anticipated Transient Without Scram (ATWS) of PWR, the one of the 'Unresolved Safety Issue' with NRC, has been investigated for many years. The latest analysis done by the author considers the PWR's inherent stability and long-term performence under the condition of ATWS combined with SBLOCA and studies the sensitivity of several assumptions, which shows positive results

  9. Pushing back the boundaries of PWR fuel performance

    International Nuclear Information System (INIS)

    Sofer, G.A.; Skogen, F.B.; Brown, C.A.; Fresk, Y.U.

    1985-01-01

    In today's fiercely competitive PWR reload market utilities are benefiting from a variety of design innovations which are helping to cut fuel cycle costs and to improve fuel performance. An advanced PWR fuel design from Exxon, for example, currently under evaluation at the Ginna plant in the United States, offers higher burn-up and greater power cycling. (author)

  10. An economic analysis code used for PWR fuel cycle

    International Nuclear Information System (INIS)

    Liu Dingqin

    1989-01-01

    An economic analysis code used for PWR fuel cycle is developed. This economic code includes 12 subroutines representing vavious processes for entire PWR fuel cycle, and indicates the influence of the fuel cost on the cost of the electricity generation and the influence of individual process on the sensitivity of the fuel cycle cost

  11. Sizewell: proposed site for Britain's first PWR power station

    International Nuclear Information System (INIS)

    1980-10-01

    The pamphlet covers the following points, very briefly: nuclear power - a success story; the Government's nuclear programme; why Sizewell; the PWR (with diagram); the PWR at Sizewell (with aerial view) (location; size; cooling water; road access; fuel transport; construction; employment; environment; screening; the next steps (licensing procedures, etc.); safety; further information). (U.K.)

  12. Chemical decontamination solutions: Effects on PWR equipment

    International Nuclear Information System (INIS)

    Pezze, C.M.; Colvin, E.R.; Aspden, R.G.

    1992-01-01

    A critical objective for the nuclear industry is the reduction of personnel exposure to radiation. Reductions have been achieved through industry's radiation management programs including training and radiation awareness concepts. Increased plant maintenance and higher radiation fields at many sites continue to raise concerns. To alleviate the radiation exposure problem, the sources of radiation which contribute to personnel exposure must be removed from the plant. A feasible was of significantly reducing these sources from a Pressurized Water Reactor (PWR) is to chemically decontaminate the entire reactor coolant system (RCS). A program was conducted to determine the technical acceptability of using certain dilute chemical solvent processes for full RCS chemical decontamination. The two processes evaluated were CAN-DEREM and LOMI. The purpose of the program was to define and complete a systematic evaluation of the major issues that need to be addressed for the successful decontamination of the entire RCS and affected portions of the auxiliary systems of a four-loop PWR system. A test program was designed to evaluate the corrosion effects of the two decontamination processes under expected plant conditions. Materials and sample configurations dictated by generic PWR components were evaluated. The testing also included many standard corrosion coupons. The test data were then used to assess the impact of chemical decontamination on the physical condition and operability of the components, equipment and mechanical systems that make up the RCS. An overview of the test program, sample configurations, data and engineering evaluations is presented. The data demonstrate that through detailed engineering evaluations of corrosion data and equipment function, the impact of full RCS chemical decontamination on plant equipment is established

  13. Contribution for the improvement of pressurized thermal shock assessment methodologies in PWR pressure vessels

    International Nuclear Information System (INIS)

    Gomes, Paulo de Tarso Vida

    2005-01-01

    The structural integrity assessment of nuclear reactor pressure vessel, concerned to Pressurized Thermal Shock (PTS) accidents, became a necessity and has been investigated since the eighty's. The recognition of the importance of PTS assessment has led the international nuclear technology community to devote a considerable research effort directed to the complete integrity assessment process of the Reactor Pressure Vessels (VPR). Researchers in Europe, Japan and U.S.A. have concentrated efforts in the VPR structural and fracture analysis, conducting experiments to best understand how specific factors act on the behavior of discontinuities, under PTS loading conditions. The main goal of this work is to study de structural behavior of an 'in scale' PWR nuclear reactor pressure vessel model, containing actual discontinuities, under loading conditions generated by a pressurized thermal shock. To construct the pressure vessel model utilized in this research, the approach developed by Barroso (1995) and based on likelihood studies, related to thermal-hydraulic behavior during the PTS was employed. To achieve the objective of this research, a new methodology to generate cracks, with known geometry and localization in the vessel model wall was developed. Additionally, an hydraulic circuit, able to flood the vessel model, heated to 300 deg C, with 10 m 3 of water at 8 deg C, in 170 seconds, was built. Thermo-hydraulic calculations using RELAP5/M0D 3.2.2γ computational code were done, to estimate the temperature profiles during the cooling time. The resulting data subsidized the thermo-structural calculations that were accomplished using ANSYS 7.01 computational code, for both 2D and 3D models. So, the stress profiles obtained with these calculations were associated with fracture mechanics concepts, to assess the crack growth behavior in the VPR model wall. After the PTS test, the VPR model was submitted to destructive and non-destructive inspections. The results

  14. Transient study of a PWR pressurizer

    International Nuclear Information System (INIS)

    Sotoma, H.

    1973-01-01

    An appropriate method for the calculation and transient performance of the pressurizer of a pressurized water reactor is presented. The study shows a digital program of simulation of pressurizer dynamics based on the First Law of Thermodynamic and Laws of Heat and Mass Transfer. The importance of the digital program that was written for a pressurizer of PWR, lies in the fact that, this can be of practical use in the safety analysis of a reactor of Angra dos Reis type with a power of about 500 M We. (author)

  15. Technical specifications for PWR secondary water chemistry

    International Nuclear Information System (INIS)

    Weeks, J.R.; van Rooyen, D.

    1977-08-01

    The bases for establishing Technical Specifications for PWR secondary water chemistry are reviewed. Whereas extremely stringent control of secondary water needs to be maintained to prevent denting in some units, sound bases for establishing limits that will prevent stress corrosion, wastage, and denting do not exist at the present time. This area is being examined very thoroughly by industry-sponsored research programs. Based on the evidence available to date, short term control limits are suggested; establishment of these or other limits as Technical Specifications is not recommended until the results of the research programs have been obtained and evaluated

  16. Technical basis for PWR emergency plans forming

    International Nuclear Information System (INIS)

    L'Homme, A.; Manesse, D.; Gauvain, J.; Crabol, B.

    1989-10-01

    Our speech first summarizes the french approach concerning the management of severe accidents which could occur on PWR stations. Then it defines the source-term which is being used as a general support for elaborating the emergency plans devoted to the protection of the population. It describes next the consequences of this source-term on the site and in the environment, which constitute the technical bases for defining actions of utilities and concerned authorities. It gives lastly information on the present status of the different emergency plans and the complementary work undertaken to improve them [fr

  17. Coolant degassing device for PWR type reactors

    International Nuclear Information System (INIS)

    Kita, Kaoru; Takezawa, Kazuaki; Minemoto, Masaki.

    1982-01-01

    Purpose: To efficiently decrease the rare gas concentration in primary coolants, as well as shorten the degassing time required for the periodical inspection in the waste gas processing system of a PWR type reactor. Constitution: Usual degassing method by supplying hydrogen or nitrogen to a volume control tank is replaced with a method of utilizing a degassing tower (method of flowing down processing liquid into the filled tower from above while uprising streams from the bottom of the tower thereby degassing the gases dissolved in the liquid into the steams). The degassing tower is combined with a hydrogen separator or hydrogen recombiner to constitute a waste gas processing system. (Ikeda, J.)

  18. Industrywide survey of PWR organics. Final report

    International Nuclear Information System (INIS)

    Richards, J.E.; Byers, W.A.

    1986-07-01

    Thirteen Pressurized Water reactor (PWR) secondary cycles were sampled for organic acids, total organic carbon, and inorganic anions. The distribution and removal of organics in a makeup water treatment system were investigted at an additional plant. TOC analyses were used for the analysis of makeup water systems; anion ion chromatography and ion exclusion chromatography were used for the analysis of secondary water systems. Additional information on plant operation and water chemistry was collected in a survey. The analytical and survey data were compared and correlations made

  19. Microcomputer simulation of PWR power plant pressurizer

    International Nuclear Information System (INIS)

    Araujo, L.R.A. de; Calixto Neto, J.; Martinez, A.S.; Schirru, R.

    1990-01-01

    It is presented a method for the simulation of the pressurizer behavior of a PWR power plant. The method was implanted in a microcomputer, and it considers all the devices for the pressure control (spray and relief valves, heaters, controller, etc.). The physical phenomena and the PID (Proportional + Integral + Derivative) controller were mathematically represented by linear relations, uncoupled, discretized in the time. There are three different algorithms which take into account the non-linear effects introduced by the variation of the physical properties due to the temperature and pressure, and also the mutual effects between the physical phenomena and the PID controller. (author)

  20. Minimization of PWR reactor control rods wear

    International Nuclear Information System (INIS)

    Ponzoni Filho, Pedro; Moura Angelkorte, Gunther de

    1995-01-01

    The Rod Cluster Control Assemblies (RCCA's) of Pressurized Water Reactors (PWR's) have experienced a continuously wall cladding wear when Reactor Coolant Pumps (RCP's) are running. Fretting wear is a result of vibrational contact between RCCA rodlets and the guide cards which provide lateral support for the rodlets when RCCA's are withdrawn from the core. A procedure is developed to minimize the rodlets wear, by the shuffling and axial reposition of RCCA's every operating cycle. These shuffling and repositions are based on measurement of the rodlet cladding thickness of all RCCA's. (author). 3 refs, 2 figs, 2 tabs

  1. Burst protected nuclear reactor plant with PWR

    International Nuclear Information System (INIS)

    Harand, E.; Michel, E.

    1978-01-01

    In the PWR, several integrated components from the steam raising unit and the main coolant pump are grouped around the reactor pressure vessel in a multiloop circuit and in a vertical arrangement. For safety reasons all primary circuit components and pipelines are situated in burst protection covers. To reduce the area of the plant straight tube steam raising units with forced circulation are used as steam raising units. The boiler pumps are connected to the vertical tubes and to the pressure vessel via double pipelines made as twin chamber pipes. (DG) [de

  2. PWR life time: the EDF project

    International Nuclear Information System (INIS)

    Noel, R.; Reynes, L.; Mercier, J.P.

    1987-01-01

    Operating a very large number of standardized PWR units which supply today 70% of French power generation, Electricite de France is highly interested in getting the best estimate of the safe and economical life of these plants. An extensive program of work has been undertaken in this respect. The studies have first to go through all available data on aging process, survey and maintenance of a limited number of major components. This review will lead to recommendation of complementary work in these fields. The first conclusions are that these units are able to perform a long service time, under provision of careful survey and maintenance [fr

  3. 14C Behaviour in PWR coolant

    International Nuclear Information System (INIS)

    Sims, Howard; Dickinson Shirley; Garbett, Keith

    2012-09-01

    Although 14 C is produced in relatively small amounts in PWR coolant, it is important to know its fate, for example whether it is released by gaseous discharge, removed by absorption on ion exchange (IX) resins or deposited on the fuel pin surfaces. 14 C can exist in a range of possible chemical forms: inorganic carbon compounds (probably mainly CO 2 ), elemental carbon, and organic compounds such as hydrocarbons. This paper presents results from a preliminary survey of the possible reactions of 14 C in PWR coolant. The main conclusions of the study are: - A combination of thermal and radiolytic reactions controls the chemistry of 14 C in reactor coolant. A simple chemical kinetic model predicts that CH 3 OH would be the initial product from radiolytic reactions of 14 C following its formation from 17 O. CH 3 OH is predicted to arise as a result of reactions of OH . with CH 4 and CH 3 , and it persists because there is no known radiation chemical reduction mechanism. - Thermodynamic considerations show that CH 3 OH can be thermally reduced to CH 4 in PWR conditions, although formation of CO 2 from small organics is the most thermodynamically favourable outcome. Such reactions could be catalysed on active nickel surfaces in the primary circuit. - Limited plant data would suggest that CH 4 is the dominant form in PWR and CO 2 in BWR. This implies that radiation chemistry may be important in determining the speciation. - Addition of acetate does not affect the amount of 14 C formed, but the addition of large amounts of stable carbon would lead to a large range of additional products, some of which would be expected to deposit on fuel pin surfaces as high molecular weight hydrocarbons. However, the subsequent thermal decomposition reactions of these products are not known. - Acetate addition may represent a small input of 12 C compared with organic material released from CVCS resins, although the importance of this may depend on whether that is predominantly soluble

  4. Environmental surveillance of PWR power stations

    International Nuclear Information System (INIS)

    Conti, M.

    1980-01-01

    The action of Electricite de France with respect to the environment of PWR nuclear power stations is essentially centred on prevention. Controls are carried out at two levels: - before the power station goes on stream (radioecological study), - when the power station is operational. The purpose of the controls effected on the radioactive effluents and the environment is to check that the maximum discharge rate stipulated in the corresponding orders is complied with and to ensure that there are no anomalies in the environment [fr

  5. Advancing PWR fuel to meet customer needs

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, F W

    1987-03-01

    Since the introduction of the Optimized Fuel Assembly (OFA) for PWRs in the late 1970s, Westinghouse has continued to work with the utility customers to identify the greatest needs for further advance in fuel performance and reliability. The major customer requirements include longer fuel cycle at lower costs, increased fuel discharge burn-up, enhanced operating flexibility, all accompanied by even greater reliability. In response to these needs, Westinghouse developed Vantage 5 PWR fuel. To optimize reactor operations, Vantage 5 fuel features distinct advantages: integral fuel burnable absorbers, axial and radial blankets, intermediate flow mixers, a removable top nozzle, and assembly modifications to accommodate increased discharge burn-up.

  6. Recent development in PWR zinc injection

    International Nuclear Information System (INIS)

    Ocken, H.; Fruzzetti, K.; Frattini, P.; Wood, C.J.

    2002-01-01

    Zinc injection to the reactor coolant system (RCS) of PWRs holds the promise to alleviate two key challenges facing PWR plant operators: (1) reducing degradation of coolant system materials, including nickel-base alloy tubing and lower alloy penetrations due to stress corrosion cracking, and (2) lowering shutdown dose rates. Primary water stress corrosion cracking (PWSCC) is a dominant tube failure mode at many plants. This paper summarizes recent observations from U. S. and international PWRs that have implemented zinc injection, focusing primarily on coolant chemistry and dose rate issues. It also provides a look at the future direction of EPRI-sponsored projects on this topic. (authors)

  7. Modeling and experiments of biomass combustion in a large-scale grate boiler

    DEFF Research Database (Denmark)

    Yin, Chungen; Rosendahl, Lasse; Kær, Søren Knudsen

    2007-01-01

    is inherently more difficult due to the complexity of the solid biomass fuel bed on the grate, the turbulent reacting flow in the combustion chamber and the intensive interaction between them. This paper presents the CFD validation efforts for a modern large-scale biomass-fired grate boiler. Modeling...... and experiments are both done for the grate boiler. The comparison between them shows an overall acceptable agreement in tendency. However at some measuring ports, big discrepancies between the modeling and the experiments are observed, mainly because the modeling-based boundary conditions (BCs) could differ...

  8. Theoretical explanation of present mirror experiments and linear stability of larger scaled machines

    International Nuclear Information System (INIS)

    Berk, H.L.; Baldwin, D.E.; Cutler, T.A.; Lodestro, L.L.; Maron, N.; Pearlstein, L.D.; Rognlien, T.D.; Stewart, J.J.; Watson, D.C.

    1976-01-01

    A quasilinear model for the evolution of the 2XIIB mirror experiment is presented and shown to reproduce the time evolution of the experiment. From quasilinear theory it follows that the energy lifetime is the Spitzer electron drag time for T/sub e/ approximately less than 0.1T/sub i/. By computing the stability boundary of the DCLC mode, with warm plasma stabilization, the electron temperature is predicted as a function of radial scale length. In addition, the effect of finite length corrections to the Alfven cyclotron mode is assessed

  9. Validation of the containment code Sirius: interpretation of an explosion experiment on a scale model

    International Nuclear Information System (INIS)

    Blanchet, Y.; Obry, P.; Louvet, J.; Deshayes, M.; Phalip, C.

    1979-01-01

    The explicit 2-D axisymmetric Langrangian code SIRIUS, developed at the CEA/DRNR, Cadarache, deals with transient compressive flows in deformable primary tanks with more or less complex internal component geometries. This code has been subjected to a two-year intensive validation program on scale model experiments and a number of improvements have been incorporated. This paper presents a recent calculation of one of these experiments using the SIRIUS code, and the comparison with experimental results shows the encouraging possibilities of this Lagrangian code

  10. Simulating Small-Scale Experiments of In-Tunnel Airblast Using STUN and ALE3D

    Energy Technology Data Exchange (ETDEWEB)

    Neuscamman, Stephanie [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Glenn, Lewis [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Schebler, Gregory [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); McMichael, Larry [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Glascoe, Lee [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2011-09-12

    This report details continuing validation efforts for the Sphere and Tunnel (STUN) and ALE3D codes. STUN has been validated previously for blast propagation through tunnels using several sets of experimental data with varying charge sizes and tunnel configurations, including the MARVEL nuclear driven shock tube experiment (Glenn, 2001). The DHS-funded STUNTool version is compared to experimental data and the LLNL ALE3D hydrocode. In this particular study, we compare the performance of the STUN and ALE3D codes in modeling an in-tunnel airblast to experimental results obtained by Lunderman and Ohrt in a series of small-scale high explosive experiments (1997).

  11. FEBEX: Full-Scale engineered barriers experiment in crystalline host-rock: preoperational phase. Synthesized report

    International Nuclear Information System (INIS)

    1997-01-01

    The FEBEX project is being cofinanced by the EC under contract F 14WCT950006. In addition to the EC, seven partners from three countries of the EU. (France, Germany, and Spain) as well as one from EFTA (Switzerland) are participating in the project. ENRESA is the coordinating partner with NAGRA assisting in coordinating some aspects. The project consists of two large-scale tests and a series of complimentary laboratory tests. The work is being executed by the following organizations: CIEMAT, AITEMIN, UP-DIT (CIMNE), ULC, CSIC-Zaidin, and UPM (SPAIN) ANDRA and G.3S (FRANCE) GRS (GERMANY). This report includes a synthesized description of the experiment from its conception through the installation of the two large-scale tests (from the middle of 1994 to the beginning of 1997, preoperation stage). The experiment is described in detail in a series of specific reports. (Author)

  12. Validating Bayesian truth serum in large-scale online human experiments.

    Science.gov (United States)

    Frank, Morgan R; Cebrian, Manuel; Pickard, Galen; Rahwan, Iyad

    2017-01-01

    Bayesian truth serum (BTS) is an exciting new method for improving honesty and information quality in multiple-choice survey, but, despite the method's mathematical reliance on large sample sizes, existing literature about BTS only focuses on small experiments. Combined with the prevalence of online survey platforms, such as Amazon's Mechanical Turk, which facilitate surveys with hundreds or thousands of participants, BTS must be effective in large-scale experiments for BTS to become a readily accepted tool in real-world applications. We demonstrate that BTS quantifiably improves honesty in large-scale online surveys where the "honest" distribution of answers is known in expectation on aggregate. Furthermore, we explore a marketing application where "honest" answers cannot be known, but find that BTS treatment impacts the resulting distributions of answers.

  13. Cosmogenic activation of germanium used for tonne-scale rare event search experiments

    Science.gov (United States)

    Wei, W.-Z.; Mei, D.-M.; Zhang, C.

    2017-11-01

    We report a comprehensive study of cosmogenic activation of germanium used for tonne-scale rare event search experiments. The germanium exposure to cosmic rays on the Earth's surface are simulated with and without a shielding container using Geant4 for a given cosmic muon, neutron, and proton energy spectrum. The production rates of various radioactive isotopes are obtained for different sources separately. We find that fast neutron induced interactions dominate the production rate of cosmogenic activation. Geant4-based simulation results are compared with the calculation of ACTIVIA and the available experimental data. A reasonable agreement between Geant4 simulations and several experimental data sets is presented. We predict that cosmogenic activation of germanium can set limits to the sensitivity of the next generation of tonne-scale experiments.

  14. Scales

    Science.gov (United States)

    Scales are a visible peeling or flaking of outer skin layers. These layers are called the stratum ... Scales may be caused by dry skin, certain inflammatory skin conditions, or infections. Examples of disorders that ...

  15. Recent validation experience with multigroup cross-section libraries and scale

    International Nuclear Information System (INIS)

    Bowman, S.M.; Wright, R.Q.; DeHart, M.D.; Parks, C.V.; Petrie, L.M.

    1995-01-01

    This paper will discuss the results obtained and lessons learned from an extensive validation of new ENDF/B-V and ENDF/B-VI multigroup cross-section libraries using analyses of critical experiments. The KENO V. a Monte Carlo code in version 4.3 of the SCALE computer code system was used to perform the critical benchmark calculations via the automated SCALE sequence CSAS25. The cross-section data were processed by the SCALE automated problem-dependent resonance-processing procedure included in this sequence. Prior to calling KENO V.a, CSAS25 accesses BONAMI to perform resonance self-shielding for nuclides with Bondarenko factors and NITAWL-II to process nuclides with resonance parameter data via the Nordheim Integral Treatment

  16. Feasibility analysis of large length-scale thermocapillary flow experiment for the International Space Station

    Science.gov (United States)

    Alberts, Samantha J.

    The investigation of microgravity fluid dynamics emerged out of necessity with the advent of space exploration. In particular, capillary research took a leap forward in the 1960s with regards to liquid settling and interfacial dynamics. Due to inherent temperature variations in large spacecraft liquid systems, such as fuel tanks, forces develop on gas-liquid interfaces which induce thermocapillary flows. To date, thermocapillary flows have been studied in small, idealized research geometries usually under terrestrial conditions. The 1 to 3m lengths in current and future large tanks and hardware are designed based on hardware rather than research, which leaves spaceflight systems designers without the technological tools to effectively create safe and efficient designs. This thesis focused on the design and feasibility of a large length-scale thermocapillary flow experiment, which utilizes temperature variations to drive a flow. The design of a helical channel geometry ranging from 1 to 2.5m in length permits a large length-scale thermocapillary flow experiment to fit in a seemingly small International Space Station (ISS) facility such as the Fluids Integrated Rack (FIR). An initial investigation determined the proposed experiment produced measurable data while adhering to the FIR facility limitations. The computational portion of this thesis focused on the investigation of functional geometries of fuel tanks and depots using Surface Evolver. This work outlines the design of a large length-scale thermocapillary flow experiment for the ISS FIR. The results from this work improve the understanding thermocapillary flows and thus improve technological tools for predicting heat and mass transfer in large length-scale thermocapillary flows. Without the tools to understand the thermocapillary flows in these systems, engineers are forced to design larger, heavier vehicles to assure safety and mission success.

  17. Mark I 1/5-scale boiling water reactor pressure suppresion experiment quick-look report

    International Nuclear Information System (INIS)

    Lai, W.; Collins, E.K.

    1977-01-01

    This report is intended as a ''quick-look'' report summarizing the experimental results obtained from pressure suppression experiment numbers 2.1, 2.2, and 2.3 that were performed on the Lawrence Livermore Laboratory's 1/5-scale boiling water reactor (BWR) Mark I pressure suppression experimental facility on April 26, 1977. A brief description of the general nature of the tests and a summary of the actual tests that were performed are given

  18. IKONOS imagery for the Large Scale Biosphere–Atmosphere Experiment in Amazonia (LBA).

    Science.gov (United States)

    George Hurtt; Xiangming Xiao; Michael Keller; Michael Palace; Gregory P. Asner; Rob Braswell; Brond& #305; Eduardo S. zio; Manoel Cardoso; Claudio J.R. Carvalho; Matthew G. Fearon; Liane Guild; Steve Hagen; Scott Hetrick; Berrien Moore III; Carlos Nobre; Jane M. Read; S& aacute; Tatiana NO-VALUE; Annette Schloss; George Vourlitis; Albertus J. Wickel

    2003-01-01

    The LBA-ECO program is one of several international research components under the Brazilian-led Large Scale Biosphere–Atmosphere Experiment in Amazonia (LBA). The field-oriented research activities of this study are organized along transects and include a set of primary field sites, where the major objective is to study land-use change and ecosystem dynamics, and a...

  19. Wafer-scale fabrication of glass-FEP-glass microfluidic devices for lipid bilayer experiments.

    Science.gov (United States)

    Bomer, Johan G; Prokofyev, Alexander V; van den Berg, Albert; Le Gac, Séverine

    2014-12-07

    We report a wafer-scale fabrication process for the production of glass-FEP-glass microdevices using UV-curable adhesive (NOA81) as gluing material, which is applied using a novel "spin & roll" approach. Devices are characterized for the uniformity of the gluing layer, presence of glue in the microchannels, and alignment precision. Experiments on lipid bilayers with electrophysiological recordings using a model pore-forming polypeptide are demonstrated.

  20. Large scale experiments with a 5 MW sodium/air heat exchanger for decay heat removal

    International Nuclear Information System (INIS)

    Stehle, H.; Damm, G.; Jansing, W.

    1994-01-01

    Sodium experiments in the large scale test facility ILONA were performed to demonstrate proper operation of a passive decay heat removal system for LMFBRs based on pure natural convection flow. Temperature and flow distributions on the sodium and the air side of a 5 MW sodium/air heat exchanger in a natural draught stack were measured during steady state and transient operation in good agreement with calculations using a two dimensional computer code ATTICA/DIANA. (orig.)

  1. The construction and evaluation of a scale of consumer shopping experience

    Directory of Open Access Journals (Sweden)

    B. A. Gillham

    2003-10-01

    Full Text Available Generally speaking consumer psychology have focused primarily on the problem solving and decision making processes underlying buying behaviour. As a result, the experience of consumers engaging in the activity of shopping has been neglected. The purpose of this study was to construct and evaluate a scale of consumer shopping experience. The instrument indicated that shopping experience is not merely a means to an end, but in many cases an end in itself. The Shopping Experience Survey consisting of 54 items was developed and administered to 426 respondents consisting of students at a large tertiary institution. The study indicated that the Shopping Experience Survey is highly reliable (rxx=0,965 and sensitive to group differences. Opsomming Benaderings tot verbruikersielkunde fokus in die algemeen hoofsaaklik op die probleemoplossings- en besluitnemingsprosesse onderliggend aan aankoopgedrag. Gevolglik het die ervarings wat verbruikers tydens die aankoopproses beleef, nie veel aandag geniet nie. Die doel van die onderhawige studie was om ’n skaal van verbruikeraankoopbelewenisse te konstrueer en te evalueer. Die instrument het aangedui dat die aankoopbelewenis nie ’n middel tot ’n doel is nie, maar ‘n doel insigself is. Die Shopping Experience Survey, bestaande uit 54 items, is ontwikkel en geadministreer aan 426 respondente van ’n groot residensiële tersiêre instelling. Met die studie is bevind dat die Shopping Experience Survey hoogs betroubaar is (rxx=0,965 en sensitief is vir groepverskille.

  2. A large-scale soil-structure interaction experiment: Design and construction

    International Nuclear Information System (INIS)

    Tang, H.T.; Tang, Y.K.; Stepp, J.C.; Wall, I.B.; Lin, E.; Cheng, S.C.; Lee, S.K.

    1989-01-01

    This paper describes the design and construction phase of the Large-Scale Soil-Structure Interaction Experiment project jointly sponsored by EPRI and Taipower. The project has two objectives: 1. to obtain an earthquake database which can be used to substantiate soil-structure interaction (SSI) models and analysis methods; and 2. to quantify nuclear power plant reactor containment and internal components seismic margin based on earthquake experience data. These objectives were accomplished by recording and analyzing data from two instrumented, scaled down, reinforced concrete containment structures during seismic events. The two model structures are sited in a high seismic region in Taiwan (SMART-1). A strong-motion seismic array network is located at the site. The containment models (1/4- and 1/12-scale) were constructed and instrumented specially for this experiment. Construction was completed and data recording began in September 1985. By November 1986, 18 strong motion earthquakes ranging from Richter magnitude 4.5 to 7.0 were recorded. (orig./HP)

  3. Development and Validation of a Scale Assessing Mental Health Clinicians' Experiences of Associative Stigma.

    Science.gov (United States)

    Yanos, Philip T; Vayshenker, Beth; DeLuca, Joseph S; O'Connor, Lauren K

    2017-10-01

    Mental health professionals who work with people with serious mental illnesses are believed to experience associative stigma. Evidence suggests that associative stigma could play an important role in the erosion of empathy among professionals; however, no validated measure of the construct currently exists. This study examined the convergent and discriminant validity and factor structure of a new scale assessing the associative stigma experiences of clinicians working with people with serious mental illnesses. A total of 473 clinicians were recruited from professional associations in the United States and participated in an online study. Participants completed the Clinician Associative Stigma Scale (CASS) and measures of burnout, quality of care, expectations about recovery, and self-efficacy. Associative stigma experiences were commonly endorsed; eight items on the 18-item scale were endorsed as being experienced "sometimes" or "often" by over 50% of the sample. The new measure demonstrated a logical four-factor structure: "negative stereotypes about professional effectiveness," "discomfort with disclosure," "negative stereotypes about people with mental illness," and "stereotypes about professionals' mental health." The measure had good internal consistency. It was significantly related to measures of burnout and quality of care, but it was not related to measures of self-efficacy or expectations about recovery. Findings suggest that the CASS is internally consistent and shows evidence of convergent validity and that associative stigma is commonly experienced by mental health professionals who work with people with serious mental illnesses.

  4. Evaluation of Surface Runoff Generation Processes Using a Rainfall Simulator: A Small Scale Laboratory Experiment

    Science.gov (United States)

    Danáčová, Michaela; Valent, Peter; Výleta, Roman

    2017-12-01

    of 5 mm/min was used to irrigate a corrupted soil sample. The experiment was undertaken for several different slopes, under the condition of no vegetation cover. The results of the rainfall simulation experiment complied with the expectations of a strong relationship between the slope gradient, and the amount of surface runoff generated. The experiments with higher slope gradients were characterised by larger volumes of surface runoff generated, and by shorter times after which it occurred. The experiments with rainfall simulators in both laboratory and field conditions play an important role in better understanding of runoff generation processes. The results of such small scale experiments could be used to estimate some of the parameters of complex hydrological models, which are used to model rainfall-runoff and erosion processes at catchment scale.

  5. Data assimilation and PWR primary measurement

    International Nuclear Information System (INIS)

    Mercier, Thibaud

    2015-01-01

    A Pressurized Water Reactor (PWR) Reactor Coolant System (RCS) is a highly complex physical process: heterogeneous power, flow and temperature distributions are difficult to be accurately measured, since instrumentations are limited in number, thus leading to the relevant safety and protection margins. EDF R and D is seeking to assess the potential benefits of applying Data Assimilation to a PWR's RCS (Reactor Coolant System) measurements, in order to improve the estimators for parameters of a reactor's operating setpoint, i.e. improving accuracy and reducing uncertainties and biases of measured RCS parameters. In this thesis, we define a 0D semi-empirical model for RCS, satisfying the description level usually chosen by plant operators, and construct a Monte-Carlo Method (inspired from Ensemble Methods) in order to use this model with Data Assimilation tools. We apply this method on simulated data in order to assess the reduction of uncertainties on key parameters: results are beyond expectations, however strong hypothesis are required, implying a careful preprocessing of input data. (author)

  6. Advanced high conversion PWR: preliminary analysis

    International Nuclear Information System (INIS)

    Golfier, H.; Bellanger, V.; Bergeron, A.; Dolci, F.; Gastaldi, B.; Koberl, O.; Mignot, G.; Thevenot, C.

    2007-01-01

    In this paper, physical aspects of a HCPWR (High Conversion Light Water Reactor), which is an innovative PWR fuelled with mixed oxide and having a higher conversion ratio due to a lower moderation ratio. Moderation ratios lower than unity are considered which has led to low moderation PWR fuel assembly designs. The objectives of this parametric study are to define a feasibility area with regard to the following neutronic aspects: moderation ratio, Pu loading, reactor spectrum, irradiation time, and neutronic coefficients. Important thermohydraulic parameters are the pressure drop, the critical heat flux, the maximum temperature in the fuel rod and the pumping power. The thermohydraulic analysis shows that a range of moderation ratios from 0.8 to 1.2 is technically possible. A compromise between improved fuel utilization and research and development effort has been found for the moderation ration of about 1. The parametric study shows that there are 2 ranges of interest for the moderation ratio: -) moderation ratio between 0.8 and 1.2 with reduced fissile heights (> 3 m), hexagonal arrangement fuel assembly and square arrangement fuel assembly are possible; and -) moderation between 0.6 and 0.7 with a modification of the reactor operating conditions (reduction of the primary flow and of the thermal power), the fuel rods could be arranged inside a hexagonal fuel rod assembly. (A.C.)

  7. CECP, Decommissioning Costs for PWR and BWR

    International Nuclear Information System (INIS)

    Bierschbach, M.C.

    1997-01-01

    1 - Description of program or function: The Cost Estimating Computer Program CECP, designed for use on an IBM personal computer or equivalent, was developed for estimating the cost of decommissioning boiling water reactor (BWR) and light-water reactor (PWR) power stations to the point of license termination. 2 - Method of solution: Cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial volume and costs; and manpower staffing costs. Using equipment and consumables costs and inventory data supplied by the user, CECP calculates unit cost factors and then combines these factors with transportation and burial cost algorithms to produce a complete report of decommissioning costs. In addition to costs, CECP also calculates person-hours, crew-hours, and exposure person-hours associated with decommissioning. 3 - Restrictions on the complexity of the problem: The program is designed for a specific waste charge structure. The waste cost data structure cannot handle intermediate waste handlers or changes in the charge rate structures. The decommissioning of a reactor can be divided into 5 periods. 200 different items for special equipment costs are possible. The maximum amount for each special equipment item is 99,999,999$. You can support data for 10 buildings, 100 components each; ESTS1071/01: There are 65 components for 28 systems available to specify the contaminated systems costs (BWR). ESTS1071/02: There are 75 components for 25 systems available to specify the contaminated systems costs (PWR)

  8. Modeling of PWR fuel at extended burnup

    International Nuclear Information System (INIS)

    Dias, Raphael Mejias

    2016-01-01

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

  9. Workers doses in central European PWR NPPs

    International Nuclear Information System (INIS)

    Janzekovic, H.; Krizman, M.

    2003-01-01

    As is stated, the ISOE database which was established in 1992 forms an excellent basis for studies and comparisons of occupational exposure data between nuclear power plants. In the year 2001, 69% of all participating reactors were pressurised water reactors. The ISOE database presents workers' exposure from 213 participating pressurised reactors (PWR) from 27 countries in that year. Among these 32 PWRs belong to six Central European Countries. The analysis of the exposure of workers based on radiation protection performance indicators (collective dose, average dose etc.) in these PWRs could be related to some nuclear safety performance indicators for recent years using ISOE database. The comparison is made to ISOE world - wide data. In the six Central European Countries altogether 32 PWR operated in the year 2001.The international databases of performance indicators related to radiation protection as for example the ISOE or the UNSCEAR database can be use as an efficient tool in the management of radiation protection of workers in a nuclear facilities and regulatory bodies. The databases enable the study of performance trends and the improvement of radiation protection. (authors)

  10. Scaling of cratering experiments: an analytical and heuristic approach to the phenomenology

    International Nuclear Information System (INIS)

    Killian, B.G.; Germain, L.S.

    1977-01-01

    The phenomenology of cratering can be thought of as consisting of two phases. The first phase, where the effects of gravity are negligible, consists of the energy source dynamically imparting its energy to the surroundings, rock and air. As illustrated in this paper, the first phase can be scaled if: radiation effects are negligible, experiments are conducted in the same rock material, time and distance use the same scaling factor, and distances scale as the cube root of the energy. The second phase of cratering consists of the rock, with its already developed velocity field, being thrown out. It is governed by the ballistics equation, and gravity is of primary importance. This second phase of cratering is examined heuristically by examples of the ballistics equation which illustrate the basic phenomena in crater formation. When gravity becomes significant, in addition to the conditions for scaling imposed in the first phase, distances must scale inversely as the ratio of gravities. A qualitative relationship for crater radius is derived and compared with calculations and experimental data over a wide range of energy sources and gravities

  11. Measuring mobbing experiences of academic nurses: development of a mobbing scale.

    Science.gov (United States)

    Ozturk, Havva; Sokmen, Serap; Yilmaz, Fatma; Cilingir, Dilek

    2008-09-01

    The aims of this study were to develop a mobbing scale for academic nurses and to determine their mobbing experiences. Data were collected between January and June 2006 with a 60-item mobbing scale and a questionnaire composed of 6 questions concerning demographics and 10 questions regarding nurses' opinions about mobbing. For the Mobbing Scale for Academic Nurses, the content validity index was 88%, item-to-total correlations ranged from .41 to .73, Cronbach alpha was .97, and Kaiser-Meyer-Olkin measure of sampling adequacy was .72. Barlett's test yielded quite significant results (chi2= 7905.47, p = .000). The scale was composed of eight subscales. One fifth of the academic nurses experienced mobbing, and there was evidence of mobbing at university nursing schools. The mobbing scale for academic nurses can be used to collect reliable and accurate data about mobbing experienced by academic nurses. If there is mobbing in nursing faculties and schools, appropriate precautions should be taken to protect people against mobbing, and a safe and comfortable atmosphere must be created in nursing faculties and schools.

  12. Results of laboratory tests on a robust filtration system for PWR containments in the case of a serious accident

    International Nuclear Information System (INIS)

    L'Homme, A.; Berlin, M.; Beraud, G.

    1986-01-01

    A study is currently in progress in France on a simple filtration process using sand as a filtration medium which, in the event of a serious accident leading to core meltdown in a pressurized water reactor, will permit controlled and filtered releases from the containment. Laboratory tests on sand filters for aerosols have been conducted. The tests involved the use of columns of sand, 80 cm high and 20 cm in diameter, under conditions which were similar to those inside the containment of a PWR in which a serious accident has occurred. The sand granulometry, the aerosol particle size and the flow rate and steam content of the fluid to be filtered were variable parameters. The results obtained from the experiment showed that as a filtration medium for this simple filter system for reactors a sand obtainable from the Cattenom quarry was most suitable. For this sand the filtration coefficient for aerosols is greater than 10 and the pressure drop is less than 10 4 pascals. Experience has also shown that there is no risk, under the operating conditions envisaged, that the filter will become clogged by aerosols or steam from condensed water or that there will be any major escape of aerosols retained during long-term operation of the filter or caused by the vaporisation of the condensed water. A larger scale experiment is already being carried out. (author)

  13. Power ramp testing method for PWR fuel rod at research reactor

    International Nuclear Information System (INIS)

    Zhou Yidong; Zhang Peisheng; Zhang Aimin; Gao Yongguang; Wang Huarong

    2003-01-01

    A tentative power ramp test for short PWR fuel rod has been conducted at the Heavy Water Research Reactor (HWRR) in China Institute of Atomic Energy (CIAE). The test fuel rod was cooled by the circulating water in the test loop. The power ramp was realized by moving solid neutron-absorbing screen around the fuel rod. The linear power of the fuel rod increased from 220 W/cm to 340 W/cm with a power ramp rate of 20 W/cm/min. The power of the fuel rod was monitored by both in-core thermal and nuclear measurement sensors in the test rig. This test provides experiences for further developing the power ramp test methods for PWR fuel rods at research reactor. (author)

  14. Study on quality control measures of static casting main pipe in PWR nuclear power plant

    International Nuclear Information System (INIS)

    Jiang Zhenbiao; Li Guanying; Liu Zhicheng

    2013-01-01

    This study analyzes the main reasons which impact the quality of primary pipe static casting elbows in PWR-M310 nuclear power plant. The quality control measures are developed from the election and inspection of material, improving sand production and casting process, improving lean management of personnel. The static casting defects of primary pipe elbows for Fuqing Unit 1 and 2 were down to less than 50% of the former project. The quality of static casting for the primary pipe elbows was significantly improved. Moreover, the implementation saves human resources and financing to repair casting defects, and also helps to win the delivery schedule. The quality control measures are good reference for improving primary pipe casting process. This study provides valuable experience for further study of improving the quality of static casting for the primary pipe of PWR nuclear power plant. (authors)

  15. The new lattice code Paragon and its qualification for PWR core applications

    International Nuclear Information System (INIS)

    Ouisloumen, M.; Huria, H.C.; Mayhue, L.T.; Smith, R.M.; Kichty, M.J.; Matsumoto, H.; Tahara, Y.

    2003-01-01

    Paragon is a new two-dimensional transport code based on collision probability with interface current method and written entirely in Fortran 90/95. The qualification of Paragon has been completed and the results are very good. This qualification included a number of critical experiments. Comparisons to the Monte Carlo code MCNP for a wide variety of PWR assembly lattice types were also performed. In addition, Paragon-based core simulator models have been compared against PWR plant startup and operational data for a large number of plants. Some results of these calculations and also comparisons against models developed with a licensed Westinghouse lattice code, Phoenix-P, are presented. The qualification described in this paper provided the basis for the qualification of Paragon both as a validated transport code and as the nuclear data source for core simulator codes

  16. The empirical intensity of PWR primary coolant pumps failure and repair

    International Nuclear Information System (INIS)

    Milivojevicj, S.; Riznicj, J.

    1988-01-01

    The wealth of operating experience concerning PWR type and nuclear reactors that has been regularly monitored and systematically processes since 1971, enabled an analysis of the PWR primary coolant pumps operation. Failure intensity α and repair intensity μ of the pump during its working life were calculated, as these values are necessary in order to determine the reliability and availability of the pump as the basis for analyzing its effect on the safety and efficiency of the nuclear power plant. The trend of failure intensity α follows the theoretically expected changes in α over time, and this is around 10 -5 in the majority of life-time. Repair intensity μ indicates a slow rise during life-time, i.e. its faster return to operation. (author).7 refs.; 5 figs

  17. Small scale thermal-hydraulic experiment for stable operation of a pius-type reactor

    International Nuclear Information System (INIS)

    Tasaka, K.; Tamaki, M.; Imai, S.; Irianto, I.D.; Tsuji, Y.; Kukita, Y.

    1994-01-01

    Thermal-hydraulic experiments using a small-scale atmospheric pressure test loop have been performed for the Process Inherent Ultimate Safety (PIUS)-type reactor to develop the new pump speed feedback control system. Three feedback control systems based on the measurement of flow rate, differential pressure, and fluid temperature distribution in the lower density lock have been proposed and confirmed by a series of experiments. Each of the feedback control systems had been verified in the simulation experiment such as a start-up simulation test. The automatic pump speed control based on the fluid temperature at the lower density lock was quite effective to maintain the stratified interface between primary water and borated pool water for stable operation of the reactor. (author)

  18. Small scale thermal violence experiments for combined insensitive high explosive and booster materials

    Energy Technology Data Exchange (ETDEWEB)

    Rae, Philip J [Los Alamos National Laboratory; Bauer, Clare L [AWE, UK; Stennett, C [DCMT SHRIVENHAM, UK; Flower, H M [AWE, UK

    2010-01-01

    A small scale cook-off experiment has been designed to provide a violence metric for both booster and IHE materials, singly and in combination. The experiment has a simple, axisymmetric geometry provided by a 10 mm internal diameter cylindrical steel confinement up to 80 mm in length. Heating is applied from one end of the sample length creating pseudo 1-D heating profile and a thermal gradient across the sample(s). At the opposite end of the confinement to the heating block, a machined groove provides a point of rupture that generates a cylindrical fragment. The displacement of the external face of the fragment is detected by Heterodyne Velocimetry. Proof of concept experiments are reported focusing on HMX and TATB formulations, and are described in relation to confinement, ullage and heating profile. The development of a violence metric, based upon fragment velocity records is discussed.

  19. A small scale accelerator driven subcritical assembly development and demonstration experiment at LAMPF

    International Nuclear Information System (INIS)

    Wender, S.A.; Venneri, F.; Bowman, C.D.; Arthur, E.D.; Heighway, E.A.; Beard, C.A.; Bracht, R.R.; Buksa, J.J.; Chavez, W.; DeVolder, B.G.

    1994-01-01

    A small scale experiment is described that will demonstrate many of the aspects of accelerator-driven transmutation technology. This experiment uses the high-power proton beam from the Los Alamos Meson Physics Facility accelerator and will be located in the Area-A experimental hall. Beam currents of up to 1 mA will be used to produce neutrons with a molten lead target. The target is surrounded by a molten salt and graphite moderator blanket. Fissionable material can be added to the molten salt to demonstrate plutonium burning or transmutation of commercial spent fuel or energy production from thorium. The experiment will be operated at power levels up to 5 MW t

  20. Survival and leaching of Tetracycline resistant bacteria and fecal indicators from manure in field scale experiments

    DEFF Research Database (Denmark)

    Bech, Tina; Amin, Mostofa; Lægdsmand, Mette

    The spreading of manure on agricultural land is an economic and practical solution for improving soil quality; however, animal manure frequently contains zoonotic pathogenic bacteria, such as certain Eschericia coli, Salmonella spp. and Campylobacter spp. The present experiment was conducted...... as a large multidisciplinary project. Pig manure with a natural content of Tetracycline resistant bacteria and fecal indicator organisms was followed in soil columns and a field scale experiment. In the field experiment pig manure was injected into agricultural soil. The distribution and survival of natural...... occurring indicator bacteria around a manure slurry slit in the soil was followed. During a period of two months, sections of soils with different distance to the manure string were assayed to obtain information on survival and spread of bacteriophage, faecal indicators (Enterococci, Bacterioides, E. coli...