Full-Scale Mark II CRT program data report no. 10 (TEST 1203)

International Nuclear Information System (INIS)

Kukita, Yutaka; Takeshita, Isao; Yamamoto, Nobuo; Namatame, Ken; Shiba, Masayoshi

1981-03-01

Recorded data for TEST 1203 conducted on the Full-Scale Mark II CRT (Containment Response Test) Facility are presented. The test 1203 is the third test run of a series of steam discharge pool swell test. It is one of the tests where break diameter was varied parametrically, i.e., TEST 1201 (200 mm), TEST 1202 (240 mm) and TEST 1203 (220 mm). The test was successfully conducted. A drywell initial pressurization rate of 188 kPa/s was obtained, which is approximately equal to what is postulated for a hypothetical DBA (Design Basis Accident) LOCA in the BWR Mark II containment. (author)

Key Parameters for Operator Diagnosis of BWR Plant Condition during a Severe Accident

Energy Technology Data Exchange (ETDEWEB)

Clayton, Dwight A [ORNL; Poore III, Willis P [ORNL

2015-01-01

The objective of this research is to examine the key information needed from nuclear power plant instrumentation to guide severe accident management and mitigation for boiling water reactor (BWR) designs (specifically, a BWR/4-Mark I), estimate environmental conditions that the instrumentation will experience during a severe accident, and identify potential gaps in existing instrumentation that may require further research and development. This report notes the key parameters that instrumentation needs to measure to help operators respond to severe accidents. A follow-up report will assess severe accident environmental conditions as estimated by severe accident simulation model analysis for a specific US BWR/4-Mark I plant for those instrumentation systems considered most important for accident management purposes.

Leibstadt: a 950-MW(e) BWR/6 Mark-III in commercial operation

International Nuclear Information System (INIS)

Fischer, P.U.

1985-01-01

It may be somewhat premature to report on a plant that started up in 1984 as the first of General Electric's (GE's) BWR/6 Mark-III plants in the Western Hemisphere and commenced commercial operation on December 15, 1984. The theme of the session certainly applies to the overall Swiss nuclear program and the search for excellence has been our ambition out of economic and energy supply necessities. Leibstadt came on line just in time to cover the needs of the Swiss consumers during the winter of 84/85. It has provided reliable service from the outset and operated during the extreme European cold wave in January 1985 without interruption. In 1985 the plant is expected to cover approx.15% of the electricity needs of Switzerland. The encouraging start of commercial operation gives hope that with time Leibstadt will be able to approach the capacity factors of the other four Swiss nuclear power stations, which in 1984 were between 88.4 and 90.3%

A direct comparison of MELCOR 1.8.3 and MAAP4 results for several PWR ampersand BWR accident sequences

International Nuclear Information System (INIS)

Leonard, M.T.; Ashbaugh, S.G.; Cole, R.K.; Bergeron, K.D.; Nagashima, K.

1996-01-01

This paper presents a comparison of calculations of severe accident progression for several postulated accident sequences for representative Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) nuclear power plants performed with the MELCOR 1.8.3 and the MAAP4 computer codes. The PWR system examined in this study is a 1100 MWe system similar in design to a Westinghouse 3-loop plant with a large dry containment; the BWR is a 1100 MWe system similar in design to General Electric BWR/4 with a Mark I containment. A total of nine accident sequences were studied with both codes. Results of these calculations are compared to identify major differences in the timing of key events in the calculated accident progression or other important aspects of severe accident behavior, and to identify specific sources of the observed differences

BWR containments license renewal industry report; revision 1. Final report

International Nuclear Information System (INIS)

Smith, S.; Gregor, F.

1994-07-01

The U.S. nuclear power industry, through coordination by the Nuclear Management and Resources Council (NUMARC), and sponsorship by the U.S. Department of Energy (DOE) and the Electric Power Research Institute (EPRI), has evaluated age-related degradation effects for a number of major plant systems, structures, and components, in the license renewal technical Industry Reports (IR's). License renewal applicants may choose to reference these IR's in support of their plant-specific license renewal applications as an equivalent to the integrated plant assessment provisions of the license renewal rule (IOCFR54). The scope of the IR provides the technical basis for license renewal for U.S. Boiling Water Reactor (BWR) containments. The scope of the report includes containments constructed of reinforced or prestressed concrete with steel liners and freestanding stell containments. Those domestic BWR containments designated as Mark I, Mark II or Mark III are covered, but no containments are addressed before these designs. The report includes those items within the jurisdictional boundaries for metal and concrete containments defined by Section III of the ASME Boiler and Pressure Vessel Code, Division 1, Subsection NE (Class MC) and Division 2 (Class CC) and their supports, but excluding snubbers

Power oscillations in BWR reactors

International Nuclear Information System (INIS)

Espinosa P, G.

2002-01-01

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

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

International Nuclear Information System (INIS)

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

1996-01-01

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

An overview of the BWR ECCS strainer blockage issues

International Nuclear Information System (INIS)

Serkiz, A.W.; Marshall, M.L. Jr.; Elliott, R.

1996-01-01

This Paper provides a brief overview of actions taken in the mid 1980s to resolve Unresolved Safety Issue (USI) A-43, open-quotes Containment Emergency Sump Performance,close quotes and their relationship to the BWR strainer blockage issue; the importance of insights gained from the Barseback-2 (a Swedish BWR) incident in 1992 and from ECCS strainer testing and inspections at the Perry nuclear power plant in 1992 and 1993; an analysis of an US BWR/4 with a Mark I containment; an international community sharing of knowledge relevant to ECCS strainer blockage, additional experimental programs; and identification of actions needed to resolve the strainer blockage issue and the status of such efforts

Full-scale mark II CRT program data report no. 7

International Nuclear Information System (INIS)

Kukita, Yutaka; Namatame, Ken; Yamamoto, Nobuo; Takeshita, Isao; Shiba, Masayoshi

1980-03-01

The Full-Scale Mark II CRT (Containment Response Test) Program was initiated in 1977 to provide a data base for evaluation of the LOCA hydrodynamic loads in the BWR Mark II pressure suppression system. The test facility is 1/18 in volume and has a wetwell which is a full-scale replica of one 20 0 -sector of that of a reference Mark II. This report documents test data obtained from TEST 3102, which is a large (200 mm) water break test performed on June 29, 1979. The test was performed for a supernominal break area (approx. 160% of the scaled break area for a postulated double-ended break in the recirculation line) to obtain vent steam mass fluxes ranging up to the maximum design value of the actual plants. Before the initiation of the test more than 90% of the drywell air was replaced by steam and transferred into wetwell airspace to reduce the air content in the vent flow during the test. Because of this pre-test treatment (so-called prepurging) the test data obtained for high- and intermediate-steam-flux condensation oscillations are believed to be to a high degree on conservative side. (author)

Analysis of BWR/Mark III drywell failure during degraded core accidents

International Nuclear Information System (INIS)

Yang, J.W.

1983-01-01

The potential for a hydrogen detonation due to the accumulation of a large amount of hydrogen in the drywell region of a BWR Mark III containment is analyzed. Loss of integrity of the drywell wall causes a complete bypass of the suppression pool and leads to pressurization of the containment building. However, the predicted peak containment pressure does not exceed the estimates of containment failure pressure

Standard Technical Specifications, General Electric plants, BWR/4

International Nuclear Information System (INIS)

1992-09-01

This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/4, and documents the positions of the Nuclear Regulatory Commission based on the BWR Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power specifications. This report contains three volumes. This document, Volume 2, contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS

Simulation of decreasing reactor power level with BWR simulator

International Nuclear Information System (INIS)

Suwoto; Zuhair; Rivai, Abu Khalid

2002-01-01

Study on characteristic of BWR using Desktop PC Based Simulator Program was analysed. This simulator is more efficient and cheaper for analyzing of characteristic and dynamic respond than full scope simulator for decreasing power level of BW. Dynamic responses of BWR reactor was investigated during the power level reduction from 100% FP (Full Power) which is 3926 MWth to 0% FP with 25% steps and 1 % FP/sec rate. The overall results for core flow rate, reactor steam flow, feed-water flow and turbine-generator power show tendency proportional to reduction of reactor power. This results show that reactor power control in BWR could be done by control of re-circulation flow that alter the density of water used as coolant and moderator. Decreasing the re-circulation flow rate will decrease void density which has negative reactivity and also affect the position of control rods

BWR Mark III containment analyses using a GOTHIC 8.0 3D model

International Nuclear Information System (INIS)

Jimenez, Gonzalo; Serrano, César; Lopez-Alonso, Emma; Molina, M del Carmen; Calvo, Daniel; García, Javier; Queral, César; Zuriaga, J. Vicente; González, Montserrat

2015-01-01

Highlights: • The development of a 3D GOTHIC code model of BWR Mark-III containment is described. • Suppression pool modelling based on the POOLEX STB-20 and STB-16 experimental tests. • LOCA and SBO transient simulated to verify the behaviour of the 3D GOTHIC model. • Comparison between the 3D GOTHIC model and MAAP4.07 model is conducted. • Accurate reproduction of pre severe accident conditions with the 3D GOTHIC model. - Abstract: The purpose of this study is to establish a detailed three-dimensional model of Cofrentes NPP BWR/6 Mark III containment building using the containment code GOTHIC 8.0. This paper presents the model construction, the phenomenology tests conducted and the selected transient for the model evaluation. In order to study the proper settings for the model in the suppression pool, two experiments conducted with the experimental installation POOLEX have been simulated, allowing to obtain a proper behaviour of the model under different suppression pool phenomenology. In the transient analyses, a Loss of Coolant Accident (LOCA) and a Station Blackout (SBO) transient have been performed. The main results of the simulations of those transients were qualitative compared with the results obtained from simulations with MAAP 4.07 Cofrentes NPP model, used by the plant for simulating severe accidents. From this comparison, a verification of the model in terms of pressurization, asymmetric discharges and high pressure release were obtained. The completeness of this model has proved to adequately simulate the thermal hydraulic phenomena which occur in the containment during accidental sequences

Boiling water system of nuclear power plants (BWR)

International Nuclear Information System (INIS)

Martias Nurdin

1975-01-01

About 85% of the world electric generators are light water reactors. It shows that LWR is technologically and economically competitive with other generators. The Boiling Water Reactor (BWR) is one of the two systems in the LWR group. The techniques of BWR operation in several countries, especially low and moderate power BWR, are presented. The discussion is made in relation with the interconnection problems of electric installation in developing countries, including Indonesia, where the total electric energy installation is low. The high reliability and great flexibility of the operation of a boiling water reactor for a sufficiently long period are also presented. Component standardization for BWR system is discussed to get a better technological and economical performance for further development. (author)

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

International Nuclear Information System (INIS)

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

1999-01-01

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

Uncertainties in source term estimates for a station blackout accident in a BWR with Mark I containment

International Nuclear Information System (INIS)

Lee, M.; Cazzoli, E.; Liu, Y.; Davis, R.; Nourbakhsh, H.; Schmidt, E.; Unwin, S.; Khatib-Rahbar, M.

1988-01-01

In this paper, attention is limited to a single accident progression sequence, namly a station blackout accident in a BWR with a Mark I containment building. Identified as an important accident in the draft version of NUREG-1150 a station blackout involves loss of both off-site power and dc power resulting in failure of the diesels to start and in the unavailability of the high pressure injection and core isolation cooling systems. This paper illustrates the calculated uncertainties (Probability Density Functions) associated with the radiological releases into the environment for the nine fission product groups at 10 hours following the initiation of core-concrete interactions. Also shown are the results ofthe STCP base case simulation. 5 refs., 1 fig., 1 tab

BWR Mark III pressure suppression containment response to hydrogen deflagration

International Nuclear Information System (INIS)

Fuls, G.M.; Gunter, A.D.

1982-01-01

The CLASIX-3 computer program has been used to evaluate the temperature and pressure response of the BWR Mark III Suppression Containment System to hydrogen deflagration resulting from a degraded core condition. The CLASIX-3 computer program is an extension of the CLASIX program which was originally developed to analyze ice condenser containments. A brief description is given of the modifications made to CLASIX to increase its flexibility and versatility to include the capability of analyzing the Mark III Containment. Analytical results are presented for the two base case transients. The two base cases are the stuck open steam relief valve and the small break LOCA, both of which are assumed to lead to a degraded core condition and the release of hydrogen to the containment. Results include pressure and temperature response, gas concentrations and suppression pool response

BWR 90: The ABB advanced BWR design

International Nuclear Information System (INIS)

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

1999-01-01

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

Analysis of a BWR direct cycle forced circulation power plants operation

International Nuclear Information System (INIS)

Andrade, G.G. de.

1973-01-01

First, it is established a general view over the operational problems of the BWR direct cycle forced circulation power plants, and then it is analysed the possibility of the utilization of the energy purged from the turbine as an additional energy for the electrical generation. To simulate the BWR power plant and to obtain the solution of the mathematical model it was developed a computer code named ATOR which shows the feasibility of the proposed method. In this way it is shown the possibility to get a better maneuvering allowance for the BWR power plant whenever it is permitted a convenient use of the vapor extracted from the turbine for the feedwater pre-heaters of the reactor. (author)

FIX-II/3025, BWR FIX-II Pump Trip Experiment 3025, Immediate Split Size Break

International Nuclear Information System (INIS)

NILSSON, Lars; GUSTAFSSON, Per-Ake; GUSTAFSON, Lennart; JANCZAK, Rajmund; OESTERLUNDH, Ingrid

1992-01-01

1 - Description of test facility: The FIX-II facility is a volume scaled 1:777 representation of a Swedish BWR with external pumps. The pressure vessel contains a 36 rod full length bundle and a spray condenser at the top to allow steady state operation. The downcomer, bypass channels and guide tube volumes are represented by external piping. The intact loop represents three of the four external reactor loops. The broken loop is constructed such that both guillotine breaks and split breaks may be simulated. The facility is equipped with ADS-simulation, but no ECCS injection are included. The FIX-II loop is also suited to investigate response of pump trips and MSIV closures in internal pump reactors. 2 - Description of test: Test 3025 simulates an intermediate size split break in one of the four main recirculation lines. The break area was 31 per cent of the scaled down pipe area of the reactor. The initial power of the 36-rod bundle was 3.38 MW, corresponding to the hot channel power of the reactor

CFD Simulation of rigid venting of the containment of a BWR-5 Mark-II reactor; Simulacion CFD de los venteos rigidos de la contencion de un reactor BWR-5 Mark-II

Energy Technology Data Exchange (ETDEWEB)

Galindo G, I. F.; Vazquez B, A. K.; Velazquez E, L. [Instituto Nacional de Electricidad y Energias Limpias, Reforma 113, Col. Palmira, 62490 Cuernavaca, Morelos (Mexico); Tijerina S, F.; Tapia M, R., E-mail: francisco.tijerina@cfe.gob.mx [CFE, Central Nucleoelectrica Laguna Verde, Carretera Federal Cardel-Nautla Km 42.5, 91476 Municipio Alto Lucero, Veracruz (Mexico)

2016-09-15

In conditions of prolonged loss of external energy or a severe accident, venting to the atmosphere is an alternative to prevent overpressure and release of fission products from the primary containment of a nuclear reactor. Due to the importance of flow determination through rigid vents, a computational fluid dynamics (CFD) model is proposed to verify the capacity of rigid vents in the primary containment of a boiling water reactor (BWR) under different operating conditions (pressure, temperature and compositions of the fluids). The model predicts and provides detailed information on variables such as mass flow and velocity of the venting gases. In the proposed model the primary containment gas is vented to the atmosphere via rigid vents (pipes) from the dry and wet pit. Is assumed that the container is pressurized because is in a defined scenario, and at one point the venting is open and the gas released into the atmosphere. The objective is to characterize the flow and validate the CFD model for the overpressure conditions that occur in an accident such as a LOCA, Sbo, etc. The model is implemented with Ansys-Fluent general-purpose CFD software based on the geometry of the venting ducts of the containment of a BWR. The model is developed three-dimensional and resolves at steady state for compressible flow and includes the effects of the turbulence represented by the Reynolds stress model. The CFD results are compared with the values of a one-dimensional and isentropic model for compressible flow. The relative similarity of results leads to the conclusion that the proposed CFD model can help to predict the rigid venting capacity of the containment of a BWR, however more information is required for full validation of the proposed model. (Author)

Standard Technical Specifications, General Electric plants, BWR/4

International Nuclear Information System (INIS)

1992-09-01

This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the BWR Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. This document Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

The BWR Hybrid 4 control rod

International Nuclear Information System (INIS)

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

1988-01-01

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

Study of the Utilization BWR Type Nuclear Power Reactor for Desalination Process

International Nuclear Information System (INIS)

Itjeu Karliana; Sumijanto; Dhandhang Purwadi, M.

2008-01-01

The needs of fresh water increased by rapid population growth and industrials expansion, but these demands can not be prepared naturally. Following this case, seawater desalination becomes the primer option which can fulfill the need through the nuclear desalination technology. The coupled nuclear power reactor enables to supply thermal energy for auxiliary equipment and pumps operation. The utilization study of power reactor type BWR coupled with desalination process has been performed. The goal of study is to obtain characteristic data of desalted water specification which desalination system coupling with nuclear power plant produced energy for desalination process. The study is carried out by browsing data and information, and comprehensive review of thermal energy correlation between NPP with desalination process installation. According to reviewing are found that the thermal energy and electric power utilization from the nuclear power reactor are enable to remove the seawater to produce desalted water and also to operate auxiliary equipments. The assessment results is VK-300 reactor prototype, BWR type 250 MW(e) power are cogeneration unit can supplied hot steam temperature 285 °C to the extraction turbine to empower 150 MW electric power, and a part of hot steam 130 °C is use to operate desalination process and remind heat is distribute to the municipal and offices at that region. The coupled of VK-300 reactor power type BWR with desalination installation of MED type enable to produce desalted water with high quality distillate. Based on the economic calculation that the VK-300 reactor power of BWR type produced water distillate capacity is 300.000 m 3 /hour with cost US$ 0.58/m 3 . The coupling VK-300 reactor power type BWR with MED desalination plant is competitive economically. (author)

Valuation of power oscillations in a BWR after control rod banks withdrawal events

International Nuclear Information System (INIS)

Costa, A. L.; Pereira, C.; Da Silva, C. A. M.; Veloso, M. A. F.

2009-01-01

The out-of-phase mode of oscillation is a very challenging type of instability occurring in BWR (Boiling Water Reactor) and its study is relevant because of the safety implications related to the capability to promptly detect any such inadvertent occurrence by in-core neutron detectors, thus triggering the necessary countermeasures in terms of selected rod insertion or even reactor shutdown. In this work, control rod banks (CRB) withdrawal transient was considered to study the power instability occurring in a BWR. To simulate this transient, the control rod banks were continuously removed from the BWR core in different cases. The simulation resulted in a very large increase of power. To perform the instability simulations, the RELAP5/MOD3.3 thermal hydraulic system code was coupled with the PARCS/2.4 3D neutron kinetic code. Data from a real BWR, the Peach Bottom, have been used as reference conditions and reactor parameters. The trend of the mass flow rate, pressure, coolant temperature and the void fraction to four thermal hydraulic channels symmetrically located in the core with respect to the core centre, were taken. It appears that the velocity of the rod bank withdrawal is a very important aspect for reactor stability. The slowest CRB withdrawal (180 s) did not cause power perturbation while the fast removal (20 s) triggered a slow power oscillation that little by little amplified to reach levels of more 100% of the initial power after about 210 s. The investigation of the related thermo hydraulic parameters showed that the mass flow rate, the void fraction and also the coolant temperature began to oscillate at approximately the same time interval

Analysis of a Mark II containment structure for hydrodynamic loads in suppression pool

International Nuclear Information System (INIS)

Bedrosian, B.

1978-01-01

During pressure-relief modes of BWR plant operation forcing signals are introduced into the suppression pool at discrete locations: exit nozzles of SRV discharge pipes (quenchers or ramsheads). These forcing signals are transmitted through the water of the suppression pool and, after reaching the pool boundaries, act as loadings on the containment structure wetted perimeter. The response of the containment structure is influenced by the presence of water as it interacts with the structure during application of the load. An adequate analysis must account for fluid-structure interaction (FSI) effects. This paper presents an exact formulation for solving the problem. FSI effects may become significant for a given geometry if the time history of loading and the dynamic properties of the coupled fluid-structure system satisfy a defined (system related) relationship. Results of analyses and parametric/sensitivity studies performed for the steel containment structure of an 1100 Mwe BWR nuclear plant of Mark II configuration are presented. (Author)

Evaluation of JRC source term methodology using MAAP5 as a fast-running crisis tool for a BWR4 Mark I reactor

International Nuclear Information System (INIS)

Vela-García, M.; Simola, K.

2016-01-01

JRC participated in the OECD/NEA FASTRUN benchmark reviewing fast-running software tools to model fission product releases during accidents at nuclear power plants. The main goal of fast-running software tools is to foresee the accident progression, so that mitigating actions can be taken and the population can be adequately protected. Within the FASTRUN, JRC used the MAAP 4.0.8 code and developed a methodology to obtain the source term (as activity released per radioisotope) of PWR and BWR station black-out accident scenarios. The modifications made in the MAAP models were limited to a minimum number of important parameters. This aims at reproducing a crisis situation with a limited time to adapt a generic input deck. This paper presents further studies, where JRC analysed the FASTRUN BWR scenario using MAAP 5.0.2 that has the capability of calculating doses. A sensitivity study was performed with the MAAP 5.0.2 DOSE package deactivated, using the same methodology as in the case of MAAP 4.0.8 for source term calculation. The results were close to the reference LTSBO SOARCA case, independently of the methodology used. One of the benefits of using the MAAP code is the short runtime of the simulations.

BWR power oscillation evaluation methodologies in core design

International Nuclear Information System (INIS)

Hotta, Akitoshi

1995-01-01

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

Reduction of radiation exposure in Japanese BWR Nuclear Power Plants

International Nuclear Information System (INIS)

Morikawa, Yoshitake

1995-01-01

The reduction of occupational exposure to radiation during the annual inspection and maintenance outages of Japanese boiling water reactors (BWR) is one of the most important objectives for stable and reliable operation. It was shown that this radiation exposure is caused by radionuclides, such as Co-60, Co-58 and Mn-54 which are produced from the metal elements Co, Ni, and Fe present in the corrosion products of structural materials that had been irradiated by neutrons. Therefore, to reduce radiation sources and exposures in Japanese BWRs, attempts have been reinforced to remove corrosion products and activated corrosion products from the primary coolant system. This paper describes the progress of the application of these measures to Japanese BWRs. Most Japanese BWR-4 and BWR-5 type nuclear power plants started their commercial operations during the 1970s. With the elapse of time during operations, a problem came to the forefront, namely that occupational radiation exposure during plant outages gradually increased, which obstructed the smooth running of inspections and maintenance work. To overcome this problem, extensive studies to derive effective countermeasures for radiation exposure reduction were undertaken, based on the evaluation of the plants operation data

Reduction of radiation exposure in Japanese BWR Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Morikawa, Yoshitake [ISOGO Nuclear Engineering Center, Yokohama (Japan)

1995-03-01

The reduction of occupational exposure to radiation during the annual inspection and maintenance outages of Japanese boiling water reactors (BWR) is one of the most important objectives for stable and reliable operation. It was shown that this radiation exposure is caused by radionuclides, such as Co-60, Co-58 and Mn-54 which are produced from the metal elements Co, Ni, and Fe present in the corrosion products of structural materials that had been irradiated by neutrons. Therefore, to reduce radiation sources and exposures in Japanese BWRs, attempts have been reinforced to remove corrosion products and activated corrosion products from the primary coolant system. This paper describes the progress of the application of these measures to Japanese BWRs. Most Japanese BWR-4 and BWR-5 type nuclear power plants started their commercial operations during the 1970s. With the elapse of time during operations, a problem came to the forefront, namely that occupational radiation exposure during plant outages gradually increased, which obstructed the smooth running of inspections and maintenance work. To overcome this problem, extensive studies to derive effective countermeasures for radiation exposure reduction were undertaken, based on the evaluation of the plants operation data.

Application of TRAC-BD1/MOD1 to a BWR/4 feedwater control failure ATWS

International Nuclear Information System (INIS)

Rouhani, S.Z.; Giles, M.M.; Mohr, C.M. Jr.; Weaver, W.L. III.

1984-01-01

This paper begins with a short description of the Transient Reactor Analysis Code for Boiling Water Reactors (TRAC-BWR), briefly mentioning some of its main features such as specific BWR models and input structure. Next, an input model of a BWR/4 is described, and, the assumptions used in performing an analysis of the loss of a feedwater controller without scram are listed. The important features of the calculated trends in flows, pressure, reactivity, and power are shown graphically and commented in the text. A comparison of some of the main predicted trends with the calculated results from a similar study by General Electric is also presented

Experience and development of on-line BWR surveillance system at Onagawa nuclear power station unit-1

International Nuclear Information System (INIS)

Kishi, A.; Chiba, K.; Kato, K.; Ebata, S.; Ando, Y.; Sakamoto, H.

1986-01-01

ONAGAWA nuclear power station Unit-1 (Tohoku Electric Power Co.) is a BWR-4 nuclear power station of 524 MW electric power which started commercial operation in June 1984. To attain high reliability and applicability for ONAGAWA-1, Tohoku Electric Power Co. and Toshiba started a Research and Development project on plant surveillance and diagnosis from April 1982. Main purposes of this project are to: (1) Develop an on-line surveillance system and acquire its operating experience at a commercial BWR, (2) Assist in plant operation and maintenance by data acquisition and analysis, (3) Develop a new technique for plant surveillance and diagnosis. An outline of the project, operating experience gained from the on-line surveillance system and an introduction to new diagnosis techniques are reported in this paper. (author)

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

International Nuclear Information System (INIS)

Morimoto, Yuichi; Okumura, Keisuke

1992-05-01

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

CFD Simulation of rigid venting of the containment of a BWR-5 Mark-II reactor

International Nuclear Information System (INIS)

Galindo G, I. F.; Vazquez B, A. K.; Velazquez E, L.; Tijerina S, F.; Tapia M, R.

2016-09-01

In conditions of prolonged loss of external energy or a severe accident, venting to the atmosphere is an alternative to prevent overpressure and release of fission products from the primary containment of a nuclear reactor. Due to the importance of flow determination through rigid vents, a computational fluid dynamics (CFD) model is proposed to verify the capacity of rigid vents in the primary containment of a boiling water reactor (BWR) under different operating conditions (pressure, temperature and compositions of the fluids). The model predicts and provides detailed information on variables such as mass flow and velocity of the venting gases. In the proposed model the primary containment gas is vented to the atmosphere via rigid vents (pipes) from the dry and wet pit. Is assumed that the container is pressurized because is in a defined scenario, and at one point the venting is open and the gas released into the atmosphere. The objective is to characterize the flow and validate the CFD model for the overpressure conditions that occur in an accident such as a LOCA, Sbo, etc. The model is implemented with Ansys-Fluent general-purpose CFD software based on the geometry of the venting ducts of the containment of a BWR. The model is developed three-dimensional and resolves at steady state for compressible flow and includes the effects of the turbulence represented by the Reynolds stress model. The CFD results are compared with the values of a one-dimensional and isentropic model for compressible flow. The relative similarity of results leads to the conclusion that the proposed CFD model can help to predict the rigid venting capacity of the containment of a BWR, however more information is required for full validation of the proposed model. (Author)

GPE-BWR and the containment venting and filtering issue

International Nuclear Information System (INIS)

Palomo, J.; Santiago, J. de

1988-01-01

The Spanish Boiling Water Reactor Owner's Group (GPE-BWR) is formed by three utilities, owning four units: Santa Maria de Garona (46 MWe, BWR3, Mark I containment), Cofrentes (975 MWe, BWR6, Mark III containment) and Valdecaballeros (2x975 MWe, BWR6, Mark III containment) - all of the reactors having been supplied by General Electric. One of the GPE-BWR's several committees is the Safety and Licensing Committee, which follows up the evolution of severe accident topics and particularly the containment venting and filtering issue. In September 1987, the Consejo de Seguridad Nuclear (CSN), the Spanish Regulatory Body, asked the GPE-BWR to define its position on the installation of a containment venting system. The GPE-BWR created a Working Group which presented a Report on Containment Venting to the CSN in January 1987 gathered from: the US Nuclear Regulatory Commission (NRC); some US utilities; and several European countries, especially France, Germany and Sweden. CSN's review of the containment venting Report and the Action Plan proposed by the GPE-BWR finished in April 1988. The conclusion of the Report and the proposed Action Plan take into account the US NRC's identified open items on severe accidents and the R and D programs scheduled to close these items

LAPUR5 BWR stability analysis in Kuosheng nuclear power plant

International Nuclear Information System (INIS)

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

2005-01-01

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

New design procedure development of future reactor critical power estimation. (1) Practical design-by-analysis method for BWR critical power design correlation

International Nuclear Information System (INIS)

Yamamoto, Yasushi; Mitsutake, Toru

2007-01-01

For present BWR fuels, the full mock-up thermal-hydraulic test, such as the critical power measurement test, pressure drop measurement test and so on, has been needed. However, the full mock-up test required the high costs and large-scale test facility. At present, there are only a few test facilities to perform the full mock-up thermal-hydraulic test in the world. Moreover, for future BWR, the bundle size tends to be larger, because of reducing the plant construction costs and minimizing the routine check period. For instance, AB1600, improved ABWR, was proposed from Toshiba, whose bundle size was 1.2 times larger than the conventional BWR fuel size. It is too expensive and far from realistic to perform the full mock-up thermal-hydraulic test for such a large size fuel bundle. The new design procedure is required to realize the large scale bundle design development, especially for the future reactor. Therefore, the new design procedure, Practical Design-by-Analysis (PDBA) method, has been developed. This new procedure consists of the partial mock-up test and numerical analysis. At present, the subchannel analysis method based on three-fluid two-phase flow model only is a realistic choice. Firstly, the partial mock-up test is performed, for instance, the 1/4 partial mock-up bundle. Then, the first-step critical power correlation coefficients are evaluated with the measured data. The input data, such as the spacer effect model coefficient, on the subchannel analysis are also estimated with the data. Next, the radial power effect on the critical power of the full-bundle size was estimated with the subchannel analysis. Finally, the critical power correlation is modified by the subchannel analysis results. In the present study, the critical power correlation of the conventional 8x8 BWR fuel was developed with the PDBA method by 4x4 partial mock-up tests and the subchannel analysis code. The accuracy of the estimated critical power was 3.8%. The several themes remain to

Power ramp tests of BWR-MOX fuels

International Nuclear Information System (INIS)

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

1996-01-01

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

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

International Nuclear Information System (INIS)

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

1990-01-01

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

Development of a computerized operator support system for BWR power plant

International Nuclear Information System (INIS)

Monta, K.; Sekimizu, K.; Sato, N.; Araki, T.; Mori, N.

1985-01-01

A computerized operator support system for BWR power plant has been developed since 1980 supported by the Japanese government. The main functions of the systems are post trip operational guidance, disturbance analysis, standby system management, operational margin monitoring and control rod operational guidance. The former two functions aim at protection against incidents during operation of nuclear power plants and the latter three functions aim at their prevention. As the final stage of the development, these functions are combined with the plant supervision function and are organized as an advanced man-machine interface for BWR power plant. During the above process, operator task analyses are performed to enable synthesis of these support functions for right fit to operator tasks and to realize a hierarchical structure for CRT displays for right fit to operators cognitive needs. (author)

Power generator in BWR type reactors

International Nuclear Information System (INIS)

Yoshida, Kenji.

1984-01-01

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

Impact of advanced BWR core physics method on BWR core monitoring

Energy Technology Data Exchange (ETDEWEB)

Moon, H; Wells, A [Siemens Power Corporation, Richland (United States)

2000-07-01

Siemens Power Corporation recently initiated development of POWERPLEX{sup TM}-III for delivery to the Grand Gulf Nuclear Power Station. The main change introduced in POWERPLEX{sup TM}-III as compared to its predecessor POWERPLEX{sup TM}-II is the incorporation of the advances BWR core simulator MICROBURN-B2. A number of issues were identified and evaluated relating to the implementation of MICROBURN-B2 and its impact on core monitoring. MICROBURN-B2 demands about three to five times more memory and two to three times more computing time than its predecessor MICROBURN-B in POWERPLEX {sup TM}-II. POWERPLEX{sup TM}-III will improve thermal margin prediction accuracy and provide more accurate plant operating conditions to operators than POWERPLEX{sup TM}-II due to its improved accuracy in predicted TIP values and critical k-effective. The most significant advantage of POWERPLEX{sup TM}-III is its capability to monitor a relaxed rod sequence exchange operation. (authors)

Coretran/Vipre assembly critical power assessment against Nupec BWR experiments

International Nuclear Information System (INIS)

Aounallah, Y.

2001-01-01

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

Quantification of uncertainties in source term estimates for a BWR with Mark I containment

International Nuclear Information System (INIS)

Khatib-Rahbar, M.; Cazzoli, E.; Davis, R.; Ishigami, T.; Lee, M.; Nourbakhsh, H.; Schmidt, E.; Unwin, S.

1988-01-01

A methodology for quantification and uncertainty analysis of source terms for severe accident in light water reactors (QUASAR) has been developed. The objectives of the QUASAR program are (1) to develop a framework for performing an uncertainty evaluation of the input parameters of the phenomenological models used in the Source Term Code Package (STCP), and (2) to quantify the uncertainties in certain phenomenological aspects of source terms (that are not modeled by STCP) using state-of-the-art methods. The QUASAR methodology consists of (1) screening sensitivity analysis, where the most sensitive input variables are selected for detailed uncertainty analysis, (2) uncertainty analysis, where probability density functions (PDFs) are established for the parameters identified by the screening stage and propagated through the codes to obtain PDFs for the outputs (i.e., release fractions to the environment), and (3) distribution sensitivity analysis, which is performed to determine the sensitivity of the output PDFs to the input PDFs. In this paper attention is limited to a single accident progression sequence, namely; a station blackout accident in a BWR with a Mark I containment buildings. Identified as an important accident in the draft NUREG-1150 a station blackout involves loss of both off-site power and DC power resulting in failure of the diesels to start and in the unavailability of the high pressure injection and core isolation coding systems

Comparison between MARCH-3 and MAAP-3 thermal-hydraulic results for a severe accident in a BWR system with MARK-III containment

International Nuclear Information System (INIS)

Barbucci, P.; Guidi, L.; Mariotti, G.

1988-01-01

A comparison between results provided by the Source Term Code Package and by the MAAP-3 code for a PWR with full pressure containment was presented. Thereafter the same two methodologies were used to analyse a severe accident sequence in a typical BWR power plant equipped with a General Electric BWR 6 reactor, rated at 2894 MWt, and a MARK-III type containment. As a reference sequence the TQUV was chosen. This sequence is characterized by a transient (T) with loss of feedwater (Q) and loss of all Emergency Core Cooling Systems, both at high pressure (U) and, after the intervention of the Automatic Depressurization System (ADS), at low pressure (V). After the vessel, failure two basic scenarios for the containment response were analysed: in the first one the pedestal is always dry, in the second one it is fully flooded. Typical limestone/common sand and basaltic concrete compositions were considered. In the following sections the obtained results will be shown with the main purpose to point out the different phenomenological models of the two codes rather than to look for the true plant response to such a severe accident. After the presentation of the most important physical models and of the main assumptions for the analyses (sects. 2 and 4), the comparison will be performed for the in-vessel phase, in section 3, and for the ex-vessel phase, in section 5

Electrical equipment performance under severe accident conditions (BWR/Mark 1 plant analysis): Summary report

International Nuclear Information System (INIS)

Bennett, P.R.; Kolaczkowski, A.M.; Medford, G.T.

1986-09-01

The purpose of the Performance Evaluation of Electrical Equipment during Severe Accident States Program is to determine the performance of electrical equipment, important to safety, under severe accident conditions. In FY85, a method was devised to identify important electrical equipment and the severe accident environments in which the equipment was likely to fail. This method was used to evaluate the equipment and severe accident environments for Browns Ferry Unit 1, a BWR/Mark I. Following this work, a test plan was written in FY86 to experimentally determine the performance of one selected component to two severe accident environments

Reactor power control device in BWR power plant

International Nuclear Information System (INIS)

Kurosawa, Tsuneo.

1997-01-01

The present invention provides a device for controlling reactor power based on a start-up/shut down program in a BWR type reactor, as well as for detecting deviation, if occurs, of the power from the start-up/shut down program, to control a recycling flow rate control system or control rod drive mechanisms. Namely, a power instruction section successively executes the start-up/shut down program and controls the coolant recycling system and the control rod driving mechanisms to control the power. A current state monitoring and calculation section receives a process amount, calculates parameters showing the plant state, compares/monitors them with predetermined values, detecting the deviation, if occurs, of the plant state from the start-up/shut down program, and prevents output of a power increase control signal which leads to power increase. A forecasting and monitoring/calculation section forecasts and calculates the plant state when not yet executed steps of the start-up/shut down program are performed, stops the execution of the start-up/shut down program in the next step in a case of forecasting that the results of the calculation will deviate from the start-up/shut down program. (I.S.)

Over Twenty Years Of Experience In ITU TRIGA MARK-II Reactor

International Nuclear Information System (INIS)

Yavuz, Hasbi

2008-01-01

I.T.U. TRIGA MARK-II Training and Research Reactor, rated at 250 kW steady-state and 1200 MW pulsing power is the only research and training reactor owned and operated by a university in Turkey. Reactor has been operating since March 11, 1979; therefore the reactor has been operating successfully for more than twenty years. Over the twenty years of operation: - The tangential beam tube was equipped with a neutron radiography facility, which consists of a divergent collimator and exposure room; - A computerized data acquisition system was designed and installed such that all parameters of the reactor, which are observed from the console, could be monitored both in normal and pulse operations; - An electrical power calibration system was built for the thermal power calibration of the reactor; - Publications related with I.T.U. TRIGA MARK-II Training and Research Reactor are listed in Appendix; - Two majors undesired shutdown occurred; - The I.T.U. TRIGA MARK-II Training and Research Reactor is still in operation at the moment. (authors)

Identification and assessment of containment and release management strategies for a BWR Mark III containment

International Nuclear Information System (INIS)

Lin, C.C.; Lehner, J.R.; Vandenkieboom, J.J.

1992-02-01

This report identifies and assesses accident management strategies which could be important for preventing containment failure and/or mitigating the release of fission products during a severe accident in a BWR plant with a Mark III type of containment. Based on information available from probabilistic risk assessments and other existing severe accident research, and using simplified containment and release event trees, the report identifies the challenges a Mark III containment could face during the course of a severe accident, the mechanisms behind these challenges, and the strategies that could be used to mitigate the challenges. The strategies are linked to the general safety objectives which apply for containment and release management by means of a safety objective tree. The strategies were assessed by applying them to certain severe accident sequence categories deemed important for a Mark III containment because of one or more of the following characteristics: high probability of core damage, high consequences, lead to a number of challenges, and involve the failure of multiple systems

Power control system in BWR type reactors

International Nuclear Information System (INIS)

Nishizawa, Yasuo.

1980-01-01

Purpose: To control the reactor power so that the power distribution can satisfy the limiting conditions, by regulating the reactor core flow rate while monitoring the power distribution in the reactor core of a BWR type reactor. Constitution: A power distribution monitor determines the power distribution for the entire reactor core based on the data for neutron flux, reactor core thermal power, reactor core flow rate and control rod pattern from the reactor and calculates the linear power density distribution. A power up ratio computing device computes the current linear power density increase ratio. An aimed power up ratio is determined by converting the electrical power up ratio transferred from a load demand input device into the reactor core thermal power up ratio. The present reactor core thermal power up ratio is subtracted from the limiting power up ratio and the difference is sent to an operation amount indicator and the reactor core flow rate is changed in a reactor core flow rate regulator, by which the reactor power is controlled. (Moriyama, K.)

Coretran/Vipre assembly critical power assessment against Nupec BWR experiments

Energy Technology Data Exchange (ETDEWEB)

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

2001-07-01

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

Accident sequence analysis for a BWR [Boiling Water Reactor] during low power and shutdown operations

International Nuclear Information System (INIS)

Whitehead, D.W.; Hake, T.M.

1990-01-01

Most previous Probabilistic Risk Assessments have excluded consideration of accidents initiated in low power and shutdown modes of operation. A study of the risk associated with operation in low power and shutdown is being performed at Sandia National Laboratories for a US Boiling Water Reactor (BWR). This paper describes the proposed methodology for the analysis of the risk associated with the operation of a BWR during low power and shutdown modes and presents preliminary information resulting from the application of the methodology. 2 refs., 2 tabs

Adaptive control method for core power control in TRIGA Mark II reactor

Science.gov (United States)

Sabri Minhat, Mohd; Selamat, Hazlina; Subha, Nurul Adilla Mohd

2018-01-01

The 1MWth Reactor TRIGA PUSPATI (RTP) Mark II type has undergone more than 35 years of operation. The existing core power control uses feedback control algorithm (FCA). It is challenging to keep the core power stable at the desired value within acceptable error bands to meet the safety demand of RTP due to the sensitivity of nuclear research reactor operation. Currently, the system is not satisfied with power tracking performance and can be improved. Therefore, a new design core power control is very important to improve the current performance in tracking and regulate reactor power by control the movement of control rods. In this paper, the adaptive controller and focus on Model Reference Adaptive Control (MRAC) and Self-Tuning Control (STC) were applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, adaptive controller model, and control rods selection programming. The mathematical models of the reactor core were based on point kinetics model, thermal hydraulic models, and reactivity models. The adaptive control model was presented using Lyapunov method to ensure stable close loop system and STC Generalised Minimum Variance (GMV) Controller was not necessary to know the exact plant transfer function in designing the core power control. The performance between proposed adaptive control and FCA will be compared via computer simulation and analysed the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

Compact modular BWR (CM-BWR)

International Nuclear Information System (INIS)

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

2003-01-01

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

Trend of field data on pipe wall thinning for BWR power plants

International Nuclear Information System (INIS)

Hakii, Junichi; Hiranuma, Naoki; Hidaka, Akitaka

2009-01-01

Strongly motivated by every stakeholder not to repeat Mihama Nuclear Power Station pipe rupture accident in August 2004, JSME Main Committee on Codes and Standards on Power Generation Facilities immediately launched a special task force to develop Rules on Pipe Wall Thinning Management for BWR, PWR and fossil Power Plants respectively. The authors describes the process of the development of Rules for BWR Power Plans from the view point of collections and analysis of fields data of pipe wall thinning. Through its activities, the authors confirmed the existing findings, like the effect of Oxygen injection, turbulence and dependence on coolant temperature, derived from series of laboratory-scaled experiments in FAC and coolant velocities effects in LDI. Further based upon the said proven findings with field data, they explain the adequacy of major concept of the rule such as separate treatment of FAC (Flow Accelerated Corrosion) and LDI (Liquid Droplet Impingement). (author)

The optimal control of ITU TRIGA Mark II Reactor

International Nuclear Information System (INIS)

Can, Burhanettin

2008-01-01

In this study, optimal control of ITU TRIGA Mark-II Reactor is discussed. A new controller has been designed for ITU TRIGA Mark-II Reactor. The controller consists of main and auxiliary controllers. The form is based on Pontragyn's Maximum Principle and the latter is based on PID approach. For the desired power program, a cubic function is chosen. Integral Performance Index includes the mean square of error function and the effect of selected period on the power variation. YAVCAN2 Neutronic - Thermal -Hydraulic code is used to solve the equations, namely 11 equations, dealing with neutronic - thermal - hydraulic behavior of the reactor. For the controller design, a new code, KONTCAN, is written. In the application of the code, it is seen that the controller controls the reactor power to follow the desired power program. The overshoot value alters between 100 W and 500 W depending on the selected period. There is no undershoot. The controller rapidly increases reactivity, then decreases, after that increases it until the effect of temperature feedback is compensated. Error function varies between 0-1 kW. (author)

Development of a coordinated control system for BWR nuclear power plant and HVDC transmission system

International Nuclear Information System (INIS)

Ishikawa, M.; Hara, T.; Hirayama, K.; Sekiya, K.

1986-01-01

The combined use of dc and ac transmissions or so-called hybrid transmission was under study, employing both dc and ac systems to enable stable transmission of 10,000 MW of electric power generated by the BWR nuclear plant, scheduled to be built about 800 km away from the center of the load. It was thus necessary to develop a hybrid power transmission control system, the hybrid power transmission system consisting of a high voltage dc transmission system (HVDC) and an ultrahigh ac transmission system (UHVAC). It was also necessary to develop a control system for HVDC transmission which protects the BWR nuclear power plant from being influenced by any change in transmission mode that occurs as a result of faults on the UHVAC side when the entire power of the BWR plant is being sent by the HVDC transmission. This paper clarifies the requirements for the HVDC system control during hybrid transmission and also during dc transmission. The control method that satisfies these requirements was studied to develop a control algorithm

Operation status display and monitoring system for BWR nuclear power plant

International Nuclear Information System (INIS)

Wakabayashi, Yasuo; Hayakawa, Hiroyasu; Kawamura, Atsuo; Kaneda, Mitsunori.

1982-01-01

Lately, the development of the system has been made for BWR plants, which monitors the operating status not only in normal operation but also in abnormal state and also for plant safety. Recently, the improvement of man-machine interface has been tried through the practical use of technique which displays data collectively on a CRT screen relating them mutually. As one of those results, the practical use of an electronic computer and color CRT display for No. 1 unit in the Fukushima No. 2 Nuclear Power Station (2F-1), Tokyo Electric Power Co., is described. Also, new centralized control panels containing such systems were used for the 1100 MWe BWR nuclear power plants now under construction, No. 3 unit of the Fukushima No. 2 Power Station and No. 1 unit of Kashiwazaki-Kariwa Nuclear Power Station (2F-3 and K-1, respectively). The display and monitoring system in 2F-1 plant is the first one in which a computer and color CRTs were practically employed for a BWR plant in Japan, and already in commercial operation. The advanced operating status monitoring system, to which the result of evaluation of the above system was added, was incorporated in the new centralized control panels presently under production for 2F-3 and K-1 plants. The outline of the system, the functions of an electronic computer, plant operating status monitor, surveillance test guide, the automation of plant operation and auxiliary operation guide are reported for these advanced monitoring system. It was confirmed that these systems are useful means to improve the man-machine communication for plant operation minitoring. (Wakatsuki, Y.)

Development of advanced BWR

International Nuclear Information System (INIS)

Toyota, Masatoshi

1982-01-01

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

Completion of high-efficiency BWR turbine plant 'Hamaoka unit No. 4'

International Nuclear Information System (INIS)

Tsuji, Kunio; Hamaura, Norikazu; Shibashita, Naoaki; Kazama, Seiichi

1995-01-01

Accompanying the increase of capacity of nuclear power plants in Japan, the plants having heightened economical efficiency, which are supported by the improvement of thermal efficiency and the reduction of dose, are demanded. Hitachi Ltd. has completed No. 4 turbine unit of 1137 MW output in Hamaoka Nuclear Power Station, Chubu Electric Power Co., Inc., which is the largest capacity machine in Japanese BWR plants. In this unit, the moisture separator heater, the steam turbine with high efficiency, and the hollow thread film condensate filter which treats the total flow rate of condensate are used as the reheating type BWR plant for the first time in Japan, and the plan of heightened economy and operation was adopted. It was confirmed by the trial for about 10 months that the planned performance was sufficiently satisfied, and the commercial operation was started in September, 1993. The features of the 1137 MW turbine unit are explained. The turbine is of tandem six-flow exhaust condensation type. Diffuser type low pressure turbine exhaust chambers, butterfly type combination intermediate valve are adopted. The stages with the blades having moisture-separating grooves were corrected. The reliability of the shaft system was improved. The adoption of the moisture separator heater and the application of the hollow thread film type condensate filter are explained. (K.I.)

Prediction of BWR performance under the influence of Isolation Condenser-using RAMONA-4 code

International Nuclear Information System (INIS)

Khan, H.J.; Cheng, H.S.; Rohatgi, U.S.

1992-01-01

The purpose of the Boiling Water Reactor (BWR) Isolation Condenser (IC) is to passively control the reactor pressure by removing heat from the system. This type of control is expected to reduce the frequency of opening and closing of the Safety Relief Valves (SRV). A comparative analysis is done for a BWR operating with and without the influence of an IC under Main Steam Isolation Valve (MSIV) closure. A regular BWR, with forced flow and high thermal power, has been considered for analysis. In addition, the effect of ICs on the BWR performance is studied for natural convection flow at lower power and modified riser geometry. The IC is coupled to the steam dome for the steam inlet flow and the Reactor Pressure Vessel (RPV) near the feed water entrance for the condensate return flow. Transient calculations are performed using prescribed pressure set points for the SRVs and given time settings for MSIV closure. The effect of the IC on the forced flow is to reduce the rate of pressure rise and thereby decrease the cycling frequency ofthe SRVS. This is the primary objective of any operating IC in a BWR (e.g. Oyster Creek). The response of the reactor thermal and fission power, steam flow rate, collapsed liquid level, and core average void fraction are found to agree with the trend of pressure. The variations in the case of an active IC can be closely related to the creation of a time lag and changes in the cycling frequency of the SRVS. An analysis for natural convection flow in a BWR indicates that the effect of an IC on its transient performance is similar to that for the forced convection system. In this case, the MSIV closure, has resulted in a lower peak pressure due to the magnitude of reduced power. However, the effect of reduced cycling frequency of the SRV due to the IC, and the time lag between the events, are comparable to that for forced convection

Flex concept for US-A BWR extended loss of AC power events

International Nuclear Information System (INIS)

Powers, J.; Aoyagi, Y.; Kataoka, K.; Thomas, S.; Mookhoek, B.

2015-09-01

The US-Advanced Boiling Water Reactor (US-A BWR), certified by the US NRC, is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (Stp 3 and 4) Combined License Application (Cola) and incorporates numerous design and technology enhancements for improved safety performance. Nuclear Innovation North America (NINA) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The Stp 3 and 4 project has finished the US NRC technical review of the Cola, and the final safety evaluation report (FSER) is scheduled to be issued by the US NRC in 2015. Following the accident at the Fukushima Dai-ichi plant, the US-A BWR was reviewed for Beyond Design Basis Event (BDBE) safety using industry and regulatory guidance for US NRC Order EA-12-049 Order Modifying Licenses with Regard to Requirements for Mitigation of Beyond Design Basis External Events (BDBEE). By virtue of the design approach, the US-A BWR is capable of providing an indefinite coping period for a station blackout. The use of installed systems with extended coping times is a significant advantage of the US-A BWR compared to most of the plants currently operating in the U.S. In addition, the Stp 3 and 4 design incorporates enhancements consistent with the current US industry Diverse and Flexible Coping Strategies (Flex) initiative. The final technical topic requiring review by the US NRC Advisory Committee on Reactor Safeguards was the Flex Integrated Plan submitted by NINA, and this review was successfully completed. This paper summarizes the progress of the US-A BWR in licensing the Flex Integrated Plan for the project, and describes the technology and features of the US-A BWR design that contribute to safety post-Fukushima. It also provides an informational comparison of the design capabilities of the US-A BWR for extreme external events, and relates these capabilities to re

Flex concept for US-A BWR extended loss of AC power events

Energy Technology Data Exchange (ETDEWEB)

Powers, J. [Toshiba America Nuclear Energy, Charlotte, North Carolina (United States); Aoyagi, Y.; Kataoka, K. [Toshiba Corporation, Kawasaki, Kanagawa (Japan); Thomas, S.; Mookhoek, B., E-mail: jim.powers@toshiba.com [Nuclear Innovation North America, Lake Jackson, Texas (United States)

2015-09-15

The US-Advanced Boiling Water Reactor (US-A BWR), certified by the US NRC, is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (Stp 3 and 4) Combined License Application (Cola) and incorporates numerous design and technology enhancements for improved safety performance. Nuclear Innovation North America (NINA) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The Stp 3 and 4 project has finished the US NRC technical review of the Cola, and the final safety evaluation report (FSER) is scheduled to be issued by the US NRC in 2015. Following the accident at the Fukushima Dai-ichi plant, the US-A BWR was reviewed for Beyond Design Basis Event (BDBE) safety using industry and regulatory guidance for US NRC Order EA-12-049 Order Modifying Licenses with Regard to Requirements for Mitigation of Beyond Design Basis External Events (BDBEE). By virtue of the design approach, the US-A BWR is capable of providing an indefinite coping period for a station blackout. The use of installed systems with extended coping times is a significant advantage of the US-A BWR compared to most of the plants currently operating in the U.S. In addition, the Stp 3 and 4 design incorporates enhancements consistent with the current US industry Diverse and Flexible Coping Strategies (Flex) initiative. The final technical topic requiring review by the US NRC Advisory Committee on Reactor Safeguards was the Flex Integrated Plan submitted by NINA, and this review was successfully completed. This paper summarizes the progress of the US-A BWR in licensing the Flex Integrated Plan for the project, and describes the technology and features of the US-A BWR design that contribute to safety post-Fukushima. It also provides an informational comparison of the design capabilities of the US-A BWR for extreme external events, and relates these capabilities to re

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

Energy Technology Data Exchange (ETDEWEB)

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

1995-09-01

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

Method of operating BWR type power plants

International Nuclear Information System (INIS)

Koyama, Kazuaki.

1981-01-01

Purpose: To improve the operation efficiency of BWR type reactors by reducing the time from the start-up of the reactor to the start-up of the turbine and electrical generator, as well as decrease the pressure difference in each of the sections of the pressure vessel to thereby extend its life span. Method: The operation comprises switching the nuclear reactor from the shutdown mode to the start-up mode, increasing the reactor power to a predetermined level lower than a rated power while maintaining the reactor pressure to a predetermined level lower than a rated pressure, starting up a turbine and an electrical generator in the state of the predetermined reactor pressure and the reactor power to connect the electrical generator to the power transmission system and, thereafter, increasing the reactor pressure and the reactor power to the predetermined rated pressure and rated power respectively. This can shorten the time from the start-up of the reactor to the start of the power transmission system, whereby the operation efficiency of the power plant can be improved. (Moriyama, K.)

Two types of a passive safety containment for a near future BWR with active and passive safety systems

International Nuclear Information System (INIS)

Sato, Takashi; Akinaga, Makoto; Kojima, Yoshihiro

2009-01-01

The paper presents two types of a passive safety containment for a near future BWR. They are named Mark S and Mark X containment. One of their common merits is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. The PCV pressure can be moderated within the design pressure. Another merit is the capability to submerge the PCV and the RPV above the core level. The third merit is robustness against external events such as a large commercial airplane crash. Both the containments have a passive cooling core catcher that has radial cooling channels. The Mark S containment is made of reinforced concrete and applicable to a large power BWR up to 1830 MWe. The Mark X containment has the steel secondary containment and can be cooled by natural circulation of outside air. It can accommodate a medium power BWR up to 1380 MWe. In both cases the plants have active and passive safety systems constituting in-depth hybrid safety (IDHS). The IDHS provides not only hardware diversity between active and passive safety systems but also more importantly diversity of the ultimate heat sinks between the atmosphere and the sea water. Although the plant concept discussed in the paper uses well-established technology, plant performance including economy is innovatively and evolutionally improved. Nothing is new in the hardware but everything is new in the performance.

Two types of a passive safety containment for a near future BWR with active and passive safety systems

Energy Technology Data Exchange (ETDEWEB)

Sato, Takashi [Toshiba Corporation, IEC, Gen-SS, 8, Shinsugita-ho, Isogo-ku, Yokohama (Japan)], E-mail: takashi44.sato@glb.toshiba.co.jp; Akinaga, Makoto; Kojima, Yoshihiro [Toshiba Corporation, IEC, Gen-SS, 8, Shinsugita-ho, Isogo-ku, Yokohama (Japan)

2009-09-15

The paper presents two types of a passive safety containment for a near future BWR. They are named Mark S and Mark X containment. One of their common merits is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. The PCV pressure can be moderated within the design pressure. Another merit is the capability to submerge the PCV and the RPV above the core level. The third merit is robustness against external events such as a large commercial airplane crash. Both the containments have a passive cooling core catcher that has radial cooling channels. The Mark S containment is made of reinforced concrete and applicable to a large power BWR up to 1830 MWe. The Mark X containment has the steel secondary containment and can be cooled by natural circulation of outside air. It can accommodate a medium power BWR up to 1380 MWe. In both cases the plants have active and passive safety systems constituting in-depth hybrid safety (IDHS). The IDHS provides not only hardware diversity between active and passive safety systems but also more importantly diversity of the ultimate heat sinks between the atmosphere and the sea water. Although the plant concept discussed in the paper uses well-established technology, plant performance including economy is innovatively and evolutionally improved. Nothing is new in the hardware but everything is new in the performance.

Analysis of the thermal response of a BWR Mark-I containment shell to direct contact by molten core materials

International Nuclear Information System (INIS)

Kress, T.S.; Cleveland, J.C.

1988-01-01

This study was undertaken to evaluate the thermal response of a BWR Mark-I containment shell in the event of an accident severe enough for molten core materials to fall into the cavity beneath the rector vessel and eventually come into direct contact with the shell. An existing ORNL three-dimensional transient heat transport computer code, HEATING-6, was used for a specific 2-D case (and variations) for which representative melt/shell boundary conditions required as input were available from other studies. In addition to the use of HEATING-6, a simplified analytical steady-state correlation was developed and given the name BWR Liner Analysis Program (BWRLAP). BWRLAP was ''benchmarked'' by comparison with HEATING-6 and was then used to make a number of parametric calculations to investigate the sensitivities of the results to the inputs. 5 refs., 11 figs., 2 tabs

FIX-II/2032, BWR Pump Trip Experiment 2032, Simulation Mass Flow and Power Transients

International Nuclear Information System (INIS)

1988-01-01

1 - Description of test facility: In the FIX-II pump trip experiments, mass flow and power transients were simulated subsequent to a total loss of power to the recirculation pumps in an internal pump boiling water reactor. The aim was to determine the initial power limit to give dryout in the fuel bundle for the specified transient. In addition, the peak cladding temperature was measured and the rewetting was studied. 2 - Description of test: Pump trip experiment 2032 was a part of test group 2, i.e. the mass flow transient was to simulate the pump coast down with a pump inertia of 11.3 kg.m -2 . The initial power in the 36-rod bundle was 4.44 MW which gave dryout after 1.4 s from the start of the flow transient. A maximum rod cladding temperature of 457 degrees C was measured. Rewetting was obtained after 7.6 s. 3 - Experimental limitations or shortcomings: No ECCS injection systems

Peak radiated power measurement of the DOE Mark II container tag with integrated ST-676 sensor radio frequency identification device.

Energy Technology Data Exchange (ETDEWEB)

Jursich, Mark

2010-04-01

The total peak radiated power of the Department of Energy Mark II container tag was measured in the electromagnetic reverberation chamber facility at Sandia National Laboratories. The tag's radio frequency content was also evaluated for possible emissions outside the intentional transmit frequency band. No spurious emissions of any significance were found, and the radiated power conformed to the manufacturer's specifications.

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

International Nuclear Information System (INIS)

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

2006-01-01

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

Development of power change maneuvering method for BWR

International Nuclear Information System (INIS)

Fukuzaki, Takaharu; Yamada, Naoyuki; Kiguchi, Takashi; Sakurai, Mikio.

1985-01-01

A power change maneuvering method for BWR has been proposed to generate an optimal power control maneuver, which realizes the power change operation closest to a power change demand pattern under operating constraints. The method searches for the maneuver as an optimization problem, where the variables are thermal power levels sampled from the demand pattern, the performance index is defined to express the power mismatch between demand and feasible patterns, and the constraints are limit lines on the thermal power-core flow rate map and limits on keeping fuel integrity. The usable feasible direction method is utilized as the optimization algorithm, with newly developed techniques for initial value generation and step length determination, which apply one-dimensional search and inverse-interpolation methods, respectively, to realize the effective search of the optimal solution. Simulation results show that a typical computing time is about 5 min by a general purpose computer and the method has been verified to be practical even for on-line use. (author)

Identification and assessment of containment and release management strategies for a BWR Mark I containment

International Nuclear Information System (INIS)

Lin, C.C.; Lehner, J.R.

1991-09-01

This report identifies and assesses accident management strategies which could be important for preventing containment failure and/or mitigating the release of fission products during a severe accident in a BWR plant with a Mark 1 type of containment. Based on information available from probabilistic risk assessments and other existing severe accident research, and using simplified containment and release event trees, the report identifies the challenges a Mark 1 containment could face during the course of a severe accident, the mechanisms behind these challenges, and the strategies that could be used to mitigate the challenges. A safety objective tree is developed which provides the connection between the safety objectives, the safety functions, the challenges, and the strategies. The strategies were assessed by applying them to certain severe accident sequence categories which have one or more of the following characteristics: have high probability of core damage or high consequences, lead to a number of challenges, and involve the failure of multiple systems. 59 refs., 55 figs., 27 tabs

Utilization of Slovenian TRIGA Mark II reactor

International Nuclear Information System (INIS)

Snoj, L.; Smodis, B.

2010-01-01

TRIGA Mark II research reactor at the Jozef Stefan Institute [JSI] is extensively used for various applications, such as: irradiation of various samples, training and education, verification and validation of nuclear data and computer codes, testing and development of experimental equipment used for core physics tests at a nuclear power plant. The paper briefly describes the aforementioned activities and shows that even such small reactors are still indispensable in nuclear science and technology. (author)

Efficient method for simulation of BWR severe accident sequence events before core uncovery

International Nuclear Information System (INIS)

Harrington, R.M.

1984-01-01

BWR-LACP has been a versatile tool for the ORNL SASA program. The development effort was minimal, and the code is fast running and economical. Operator actions are easily simulated and the complete scope of both reactor vessel and primary containment are modeled. Valuable insights have been gained into accident sequences. A Fortran version is under development and it will be modified for application to Mark II plants

An analysis of molten-corium-induced failure of drain pipes in BWR Mark 2 containments

International Nuclear Information System (INIS)

Taleyarkhan, R.P.; Podowski, M.Z.

1991-01-01

This study has focused on mechanistic simulation and analysis of potential failure modes for inpedestal drywell drain pipes in the Limerick boiling water reactor (BWR) Mark 2 containment. Physical phenomena related to surface tension breakdown, heatup, melting, ablation, crust formation and failure, and core material relocation into drain pipes with simultaneous melting of pipe walls were modeled and analyzed. The results of analysis have been used to assess the possibility of drain pipe failure and the resultant loss of pressure-suppression capability. Estimates have been made for the timing and amount of molten corium released to the wetwell. The study has revealed that significantly different melt progression sequences can result depending upon the failure characteristics of the frozen metallic crust which forms over the drain cover during the initial stages of debris pour. Another important result is that it can take several days for the molten fuel to ablate the frozen metallic debris layer -- if the frozen layer has cooled below 1100 K before fuel attack. 10 refs., 3 figs., 4 tabs

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

International Nuclear Information System (INIS)

Pruitt, D.W.

1992-01-01

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

Standard Technical Specifications, General Electric plants, BWR/4

International Nuclear Information System (INIS)

1992-09-01

This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. This document Volume 3, contains the Bases for Sections 3.4--3.10 of the improved STS

Analysis of Fukushima Daiichi Nuclear Power Station severe accident using MAAP4.05

International Nuclear Information System (INIS)

Yoo, Jae S.; Suh, Kune Y.; Kim, Dong M.

2011-01-01

Rather extensive reactor core meltdown and partial melt-through that took place at the Fukushima Daiichi nuclear power plants (NPPs) on March 11, 2011 had been caused by a massive earthquake followed by tsunami rarely seen in history. The happening had turned into an unprecedented serious accident since the Chernobyl Unit 4 in 1986 that extended over multiple reactors simultaneously. A previous documentary survey, NUREG-1150 report provides significant insights into how a severe accident might develop at a boiling water reactor (BWR) and the range of consequences. NUREG-1150 did identify the importance of loss of power accidents for a BWR. This paper describes recent analyses of the Fukushima Daiichi NPPs severe accident. Calculations were performed with the MAAP4.05 code by modifying the parameter file for the Peach Bottom Unit 2, a BWR 4 type in the Mark-I containment. Generally this is the same type of reactor as Fukushima Daiichi Units 2 and 3. This resulted in good understanding of the response of this type of early BWRs to prolonged loss of diesel generators and batteries. There is clearly vulnerability with this early type of BWRs which would be less onerous than later existing plant and new build designs. This analysis, however, was based on rather limited amount of information obtained at the time of preparing this report and adopted various estimates and assumptions for conditions necessary to run the analysis. Hence there persists considerable uncertainty in the results. MAAP4.05 was run pursuant to the observed data and chronology of Fukushima Daiichi Units 1 through 3 reported by the Tokyo Electric Power Company (TEPCO) as well as the Japanese Government. Severe accident scenarios have not only gone far beyond the design basis, but also exceeded the extent of multiple breakdowns assumed in the preparation for such accident management measures as the malfunction or loss of all the emergency core cooling system (ECCS) combined with the extended loss of

Summary of the OECD/NRC Boiling Water Reactor Turbine Trip Benchmark - Fourth Workshop (BWR-TT4)

International Nuclear Information System (INIS)

2002-01-01

The reference problem chosen for simulation in a BWR is a Turbine Trip transient, which begins with a sudden Turbine Stop Valve (TSV) closure. The pressure oscillation generated in the main steam piping propagates with relatively little attenuation into the reactor core. The induced core pressure oscillation results in dramatic changes of the core void distribution and fluid flow. The magnitude of the neutron flux transient taking place in the BWR core is strongly affected by the initial rate of pressure rise caused by pressure oscillation and has a strong spatial variation. The correct simulation of the power response to the pressure pulse and subsequent void collapse requires a 3-D core modeling supplemented by 1-D simulation of the remainder of the reactor coolant system. A BWR TT benchmark exercise, based on a well-defined problem with complete set of input specifications and reference experimental data, has been proposed for qualification of the coupled 3-D neutron kinetics/thermal-hydraulic system transient codes. Since this kind of transient is a dynamically complex event with reactor variables changing very rapidly, it constitutes a good benchmark problem to test the coupled codes on both levels: neutronics/thermal-hydraulic coupling and core/plant system coupling. Subsequently, the objectives of the proposed benchmark are: comprehensive feedback testing and examination of the capability of coupled codes to analyze complex transients with coupled core/plant interactions by comparison with actual experimental data. The benchmark consists of three separate exercises: Exercise 1 - Power vs. Time Plant System Simulation with Fixed Axial Power Profile Table (Obtained from Experimental Data). Exercise 2 - Coupled 3-D Kinetics/Core Thermal-Hydraulic BC Model and/or 1-D Kinetics Plant System Simulation. Exercise 3 - Best-Estimate Coupled 3-D Core/Thermal-Hydraulic System Modeling. The purpose of this fourth workshop was to present and discuss final results of

Synergistic failure of BWR internals

International Nuclear Information System (INIS)

Ware, A. G.; Chang, T.Y.

1999-01-01

Boiling Water Reactor (BWR) core shrouds and other reactor internals important to safety are experiencing intergranular stress corrosion cracking (IGSCC). The United States Nuclear Regulatory Commission has followed the problem, and as part of its investigations, contracted with the Idaho National Engineering and Environmental Laboratory to conduct a risk assessment. The overall project objective is to assess the potential consequences and risks associated with the failure of IGSCC-susceptible BWR vessel internals, with specific consideration given to potential cascading and common mode effects. An initial phase has been completed in which background material was gathered and evaluated, and potential accident sequences were identified. A second phase is underway to perform a simplified, quantitative probabilistic risk assessment on a representative high-power BWR/4. Results of the initial study conducted on the jet pumps show that any cascading failures would not result in a significant increase in the core damage frequency. The methodology is currently being extended to other major reactor internals components

BWR type nuclear power plant

International Nuclear Information System (INIS)

Matsumoto, Kosuke.

1991-01-01

In a BWR type nuclear power plant in which reactor water in a reactor pressure vessel can be drained to a waste processing system by way of reactor recycling pipeways and remaining heat removal system pipeways, a pressurized air supply device is disposed for supplying air for pressurizing reactor water to the inside of the reactor pressure vessel by way of an upper head. With such a constitution, since the pressurized air sent from the pressurized air supply device above the reactor pressure vessel for the reactor water discharging pressure upon draining, the water draining pressure is increased compared with a conventional case and, accordingly, the amount of drained water is not reduced even in the latter half of draining. Accordingly, the draining efficiency can be improved and only a relatively short period of time is required till the completion of the draining, which can improve safety and save labors. (T.M.)

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-12-15

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

Calculation device for fuel power history in BWR type reactors

International Nuclear Information System (INIS)

Sakagami, Masaharu.

1980-01-01

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

Fundamentals of boiling water reactor (BWR)

International Nuclear Information System (INIS)

Bozzola, S.

1982-01-01

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

Simulation of a scenario of total loss of external and internal power (Sbo) for different vent pressures of the containment of a BWR-5

International Nuclear Information System (INIS)

Cardenas V, J.; Mugica R, C. A.; Godinez S, V.

2014-10-01

The simulation of a Station Black Out (Sbo) was realized with intervention of the vent containment by means of a rigid vent coming from the dry-well and that discharges directly to the atmosphere, with the MELCOR code version 2.1. This scenario was carried out for a BWR-5 and containment type Mark II, with a thermal power of 2317 MWt similar to the reactor of nuclear power plant of Laguna Verde. For this scenario was considered as only available system for coolant injection to the reactor to the Reactor Core Isolation Cooling (Rcic), which remained operating 4 hours with batteries bank. The Security and Relief Valves (SR V) were considered functional (by simplicity) and that they mechanically do not exceed their capacity to liberate pressure due to the performances in their safety way. The operator maneuver to perform the SR V and to de pressurize the vessel until the pressure (13 kg/cm 2 ) to operate the low pressure systems was modeled. The results cover approximately 48 hours (172000 seconds), time in which was observed the behavior of the level and pressure in the vessel. Also the scenario evolution was analyzed to different vent pressures of the primary containment (2.0, 3.0, 4.5, 6.0, and 10.0 kg/cm 2 ), the temperature profiles of the dry-well, the hydrogen accumulation in the containment, the radio-nuclides liberation through rigid vent to the atmosphere and the inventory of these. In this work an analysis of the pressure behavior in the primary containment is presented, with the purpose of minimizing liberated fission products to the environment. (Author)

The development of emergency core cooling systems in the PWR, BWR, and HWR Candu type of nuclear power plants

International Nuclear Information System (INIS)

Mursid Djokolelono.

1976-01-01

Emergency core cooling systems in the PWR, BWR, and HWR-Candu type of nuclear power plant are reviewed. In PWR and BWR the emergency cooling can be catagorized as active high pressure, active low pressure, and a passive one. The PWR uses components of the shutdown cooling system: whereas the BWR uses components of pressure suppression contaiment. HWR Candu also uses the shutdown cooling system similar to the PWR except some details coming out from moderator coolant separation and expensive cost of heavy water. (author)

Technical description and evaluation of BWR hybrid power shape monitoring system. Final report

International Nuclear Information System (INIS)

Frogner, B.; Ipaktchi, A.; Yang, C.; Grow, R.; Ho, C.; Kiguchi, T.

1982-03-01

This report discusses the method for monitoring BWR cores that has been implemented in the Power Shape Monitoring System (PSMS). The approach has been benchmarked to TIP data from three plants and five fuel cycles and the accuracy of the calculations has been evaluated by using gamma scan data from two plants. A coupled neutronics/thermal-hydraulic nodal code (NODE-B/THERM-B) is used in the PSMS. It has been demonstrated that adaptation of this code to partially fit the TIP readings followed by a statistical characterization of the remaining errors results in better accuracy and improved sensitivity for anomaly detection compared to an approach that is entirely dependent upon the detector readings. The computed power distribution has a one-sigma uncertainty of 6% for the nodal power and 4% for the bundle power. This is significantly better than the plant process computers that actually were used for monitoring those two plants where comparisons were made

Development of next BWR plant

International Nuclear Information System (INIS)

Moriya, Kumiaki; Tanikawa, Naoshi; Kinoshita, Shoichiro; Utena, Shunsuke

1995-01-01

It is expected that BWR power generation will be main nuclear power generation for long period hereafter, and in the ABWRs being constructed at present, the safety, reliability, operation performance, economical efficiency and so on are further heightend as compared with conventional BWRs. On the other hand, in order to cope with future social change, the move to develop the next reactor type following ABWRs was begun already by the cooperation of electirc power companies and plant manufacturers. Hitachi Ltd. has advanced eagerly the development of new light water reactors. Also the objective of BWR power generation hereafter is to heighten the safety, reliability, operation performance and economical efficiency, and the development has been advanced, aiming at bearing the main roles of nuclear power generation. At present, ABWRs are under construction as No. 6 and 7 plants in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc. In order to let ABWRs take root, the further improvement of economy by the standardization, the rationalization by revising the specification and the improvement of machinery and equipment is necessary. As the needs of the development of next generation BWRs, the increase of power output, the heightening of safety and economical efficiency are discussed. The concept of the next generation BWR plant aiming at the start of operation around 2010 is shown. (K.I.)

Development of next BWR plant

Energy Technology Data Exchange (ETDEWEB)

Moriya, Kumiaki; Tanikawa, Naoshi; Kinoshita, Shoichiro; Utena, Shunsuke [Hitachi Ltd., Ibaraki (Japan). Hitachi Works

1995-04-01

It is expected that BWR power generation will be main nuclear power generation for long period hereafter, and in the ABWRs being constructed at present, the safety, reliability, operation performance, economical efficiency and so on are further heightend as compared with conventional BWRs. On the other hand, in order to cope with future social change, the move to develop the next reactor type following ABWRs was begun already by the cooperation of electirc power companies and plant manufacturers. Hitachi Ltd. has advanced eagerly the development of new light water reactors. Also the objective of BWR power generation hereafter is to heighten the safety, reliability, operation performance and economical efficiency, and the development has been advanced, aiming at bearing the main roles of nuclear power generation. At present, ABWRs are under construction as No. 6 and 7 plants in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc. In order to let ABWRs take root, the further improvement of economy by the standardization, the rationalization by revising the specification and the improvement of machinery and equipment is necessary. As the needs of the development of next generation BWRs, the increase of power output, the heightening of safety and economical efficiency are discussed. The concept of the next generation BWR plant aiming at the start of operation around 2010 is shown. (K.I.).

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

International Nuclear Information System (INIS)

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

1983-09-01

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

Mark II magnetic detector for SPEAR

International Nuclear Information System (INIS)

Larsen, R.R.

1975-01-01

The Mark II Detector, presently in the design stage, is a SLAC/LBL detector project to replace the Mark I now in operation at SPEAR. While similar in concept to the Mark I it will have improved momentum resolution, shower detection, solid angle coverage for both triggering and tracking and a magnet design providing easier access to those particles transmitted through the aluminum coil

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

International Nuclear Information System (INIS)

Tautges, T.J.

1993-10-01

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

Economic analysis of hydride fueled BWR

International Nuclear Information System (INIS)

Ganda, F.; Shuffler, C.; Greenspan, E.; Todreas, N.

2009-01-01

The economic implications of designing BWR cores with hydride fuels instead of conventional oxide fuels are analyzed. The economic analysis methodology adopted is based on the lifetime levelized cost of electricity (COE). Bracketing values (1970 and 3010 $/kWe) are used for the overnight construction costs and for the power scaling factors (0.4 and 0.8) that correlate between a change in the capital cost to a change in the power level. It is concluded that a newly constructed BWR reactor could substantially benefit from the use of 10 x 10 hydride fuel bundles instead of 10 x 10 oxide fuel bundles design presently in use. The cost saving would depend on the core pressure drop constraint that can be implemented in newly constructed BWRs - it is between 2% and 3% for a core pressure drop constraint as of the reference BWR, between 9% and 15% for a 50% higher core pressure drop, and between 12% and 21% higher for close to 100% core pressure. The attainable cost reduction was found insensitive to the specific construction cost but strongly dependent on the power scaling factor. The cost advantage of hydride fuelled cores as compared to that of the oxide reference core depends only weakly on the uranium and SWU prices, on the 'per volume base' fabrication cost of hydride fuels, and on the discount rate used. To be economically competitive, the uranium enrichment required for the hydride fuelled core needs to be around 10%.

Safety Evaluation Report related to the final design approval of the GESSAR II BWR/6 Nuclear Island Design (Docket No. 50-447). Supplement No. 3

International Nuclear Information System (INIS)

1985-01-01

Supplement 3 to the Safety Evaluation Report (SER) for the application filed by General Electric Company for the final design approval for the GE BWR/6 nuclear island design has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. This report supplements the GESSAR II SER (NUREG-0979), issued in April 1983, summarizing the results of the staff's safety review of the GESSAR II BWR/6 nuclear island design. Subject to favorable resolution of the items discussed in this supplement, the staff concludes that the GESSAR II design satisfactorily addresses the severe-accident concerns described in draft NUREG-1070

Clinical evaluation of the Non-Contact Tonometer Mark II.

Science.gov (United States)

Chauhan, B C; Henson, D B

1988-09-01

The purpose of this investigation was to test the reliability of the American Optical Non-Contact Tonometer Mark II (NCT II) using the Goldmann Applanation Tonometer (GAT) as the validating instrument. The sample contained 102 consecutive patients from our University Eye Clinic, of whom one-half had 4 NCT II measurements first, followed by 4 GAT measurements; the other one-half had 4 GAT measurements first, followed by 4 NCT II measurements. No significant change in intraocular pressure (IOP) was noted over the measurement sequence with either instrument. There was no significant difference between paired NCT II and GAT readings when the NCT II was used first; however, a highly significant difference between paired readings was obtained when the GAT was used first, indicating that the GAT measurement produced a delayed reduction in the IOP. This effect did not occur with the NCT II. Although the NCT II is shown to have a good overall reliability when compared to the GAT in both protocols, the agreement between any two tonometers may be influenced greatly by the very process of taking a measurement and by the dynamic nature of the IOP.

BWR control blade replacement strategies

Energy Technology Data Exchange (ETDEWEB)

Kennard, M W [Stoller Nuclear Fuel, NAC International, Pleasantville, NY (United States); Harbottle, J E [Stoller Nuclear Fuel, NAC International, Thornbury, Bristol (United Kingdom)

2000-02-01

The reactivity control elements in a BWR, the control blades, perform three significant functions: provide shutdown margin during normal and accident operating conditions; provide overall core reactivity control; and provide axial power shaping control. As such, the blades are exposed to the core's neutron flux, resulting in irradiation of blade structural and absorber materials. Since the absorber depletes with time (if B{sub 4}C is used, it also swells) and the structural components undergo various degradation mechanisms (e.g., embrittlement, corrosion), the blades have limits on their operational lifetimes. Consequently, BWR utilities have implemented strategies that aim to maximize blade lifetimes while balancing operational costs, such as extending a refuelling outage to shuffle high exposure blades. This paper examines the blade replacement strategies used by BWR utilities operating in US, Europe and Asia by assembling information related to: the utility's specific blade replacement strategy; the impact the newer blade designs and changes in core operating mode were having on those strategies; the mechanical and nuclear limits that determined those strategies; the methods employed to ensure that lifetime limits were not exceeded during operation; and blade designs used (current and replacement blades). (author)

BWR control blade replacement strategies

International Nuclear Information System (INIS)

Kennard, M.W.; Harbottle, J.E.

2000-01-01

The reactivity control elements in a BWR, the control blades, perform three significant functions: provide shutdown margin during normal and accident operating conditions; provide overall core reactivity control; and provide axial power shaping control. As such, the blades are exposed to the core's neutron flux, resulting in irradiation of blade structural and absorber materials. Since the absorber depletes with time (if B 4 C is used, it also swells) and the structural components undergo various degradation mechanisms (e.g., embrittlement, corrosion), the blades have limits on their operational lifetimes. Consequently, BWR utilities have implemented strategies that aim to maximize blade lifetimes while balancing operational costs, such as extending a refuelling outage to shuffle high exposure blades. This paper examines the blade replacement strategies used by BWR utilities operating in US, Europe and Asia by assembling information related to: the utility's specific blade replacement strategy; the impact the newer blade designs and changes in core operating mode were having on those strategies; the mechanical and nuclear limits that determined those strategies; the methods employed to ensure that lifetime limits were not exceeded during operation; and blade designs used (current and replacement blades). (author)

The Mark II Vertex Drift Chamber

International Nuclear Information System (INIS)

Alexander, J.P.; Baggs, R.; Fujino, D.

1989-03-01

We have completed constructing and begun operating the Mark II Drift Chamber Vertex Detector. The chamber, based on a modified jet cell design, achieves 30 μm spatial resolution and 2 gas mixtures. Special emphasis has been placed on controlling systematic errors including the use of novel construction techniques which permit accurate wire placement. Chamber performance has been studied with cosmic ray tracks collected with the chamber located both inside and outside the Mark II. Results on spatial resolution, average pulse shape, and some properties of CO 2 mixtures are presented. 10 refs., 12 figs., 1 tab

Comparison of the CORA-12, 13, 17 experiments and B4 effect on the flooding behavior of BWR bundles

International Nuclear Information System (INIS)

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

1995-01-01

The CORA quench experiments 12, 13 (PWR) and 17 (BWR) are in agreement with LOFT 2 and TMI: Flooding of hot Zircaloy clad fuel rods does not result in an immediate cooldown of the bundle, but produces remarkable temporary temperature increase, connected to a strong peak in hydrogen production. The PWR tests CORA 12 and CORA 13 are of the same geometrical arrangement and test conduct, with the exception of the shorter time between power shutdown and quench initiation for CORA 13. A higher temperature of the bundle at start of quenching was the consequence. BWR test CORA 17 - with B 4 C absorber and additional Zircaloy channel box walls - was in respect to the delay-time between power shutdown and start of quenching similar to test CORA 12. All tests showed during the quench phase the temporary temperature increase, correlated to a hydrogen peak. The CORA 17 test resulted immediately after quenching in a modest increase for 20 s and changed then in a steep increase, resulting in the highest temperature and hydrogen peaks of the three tests. CORA 17 also showed a temperature increase in the lower part of the bundle, in contrast to CORA 12 and CORA 13 with temperature increase only in the upper half of the bundle. We interpret this earlier starting and stronger reaction due to the influence of the boron carbide, the absorber material of the BWR test. B 4 C has an exothermic reaction rate 4 to 9 times larger than Zry and produces 5 to 6,6 times more hydrogen. Probably the hot remained columns of B 4 C (seen in the non-quench test CORA 16) react early in the quench process with the increased upcoming steam. The bundle temperature raised by this reaction increases the reaction rate (exponential dependency) of the remaining metallic Zry. Due to the larger amount of Zry in the BWR bundle (channel box walls) and the smaller steam input during the heatup phase (2 g/s instead of 6 g/s) more metallic Zry can have survived oxidation during the heatup phase. (orig./HP)

Panorama of the BWR reactors - Evolution of the concept

Energy Technology Data Exchange (ETDEWEB)

Novotny, C.; Uhrig, E. [AREVA NP GmbH, Safety Engineering Department - PEPS-G (Germany)

2012-01-15

Nowadays, a fleet of more than 50 boiling water reactors (BWR) are in operation in the world. This article gives a short overview on the developments of nuclear power plants of the BWR type, with a focus on the European builds. It describes the technical bases from the early designs in the fifties, sketches the innovations of the sixties and seventies in the types BWR 69 and 72 (Baulinie 69 and 72) and gives an outlook of a possible next generation BWR. A promising approach in recent BWR developments is the the combination of passive safety systems with established design basis

BWR radiation exposure--experience and projection

International Nuclear Information System (INIS)

Falk, C.F.; Wilkinson, C.D.; Hollander, W.R.

1979-01-01

The BWR/6 Mark III radiation exposures are projected to be about half of those of current average operating experience of 725 man-rem. These projections are said to be realistic and based on current achievements and not on promises of future development. The several BWRs operating with low primary system radiation levels are positive evidence that radiation sources can be reduced. Improvements have been made in reducing the maintenance times for the BWR/6, and further improvements can be made by further attention to cost-effective plant arrangement and layout during detail design to improve accessibility and maintainability of each system and component

The Studsvik power transient programs Demo-Ramp II and Trans-Ramp I

International Nuclear Information System (INIS)

Bergenlid, U.; Lysell, G.; Mogard, H.; Roennberg, G.

1984-01-01

The Studsvik Demo-Ramp II och Trans-Ramp I are internationally sponsored research programs. The main objectives are similar in both programs: to study the effects on the PCI/SCC failure process of short time power transients, above the failure threshold where cladding failure (FP leakage) is expected to occur after a sufficient hold time. Demo-Ramp II is completed, whereas, at present, Trans-Ramp I is in progress. Test fuel rods of standard BWR design are used. The fuel rods have been base-irradiated in a power reactor (burn-up in the range 18 to 29 MWd/kg U) and subsequently ramp tested in the R2 reactor. Extensive examinations of the rods have been performed. In the Demo-Ramp II program a large number of incipient cladding cracks were observed to be formed more rapidly than expected, based on previous knowledge. It was possible to operate one rod for a very short time above the failure threshold without SCC crack formation. One objective of the Trans-Ramp I program is to define more closely the power-time region above the failure threshold where the rods remain intact after power transients. (author)

Study of a new automatic reactor power control for the TRIGA Mark II reactor at University of Pavia

Energy Technology Data Exchange (ETDEWEB)

Borio Di Tigliole, A.; Magrotti, G. [Laboratorio Energia Nucleare Applicata (L.E.N.A.), University of Pavia, Via Aselli 41, 27100 (Italy); Cammi, A.; Memoli, V. [Politecnico di Milano, Department of Energy, Nuclear Engineering Division (CeSNEF), Via Ponzio 34/3, 20133 Milano (Italy); Gadan, M. A. [Instrumentation and Control Department, National Atomic Energy Comission of Argentina, University of Pavia (Italy)

2009-07-01

The installation of a new Instrumentation and Control (IC) system for the TRIGA Mark-II reactor at University of Pavia has recently been completed in order to assure a safe and continuous reactor operation for the future. The intervention involved nearly the whole IC system and required a channel-by-channel component substitution. One of the most sensitive part of the intervention concerned the Automatic Reactor Power Controller (ARPC) which permits to keep the reactor at an operator-selected power level acting on the control rod devoted to the fine regulation of system reactivity. This controller installed can be set up using different control logics: currently the system is working in relay mode. The main goal of the work presented in this paper is to set up a Proportional-Integral-Derivative (PID) configuration of the new controller installed on the TRIGA reactor of Pavia so as to optimize the response to system perturbations. The analysis have shown that a continuous PID offers generally better results than the relay mode which causes power oscillations with an amplitude of 3% of the nominal power

Evaluation of TRIGA Mark II reactor in Turkey

International Nuclear Information System (INIS)

Bilge, Ali Nezihi

1990-01-01

There are two research reactors in Turkey and one of them is the university Triga Mark II reactor which was in service since 1979 both for education and industrial application purposes. The main aim of this paper is to evaluate the spectrum of the services carried by Turkish Triga Mark II reactor. In this work, statistical distribution of the graduate works and applications, by using Triga Mark II reactor is examined and evaluated. In addition to this, technical and scientific uses of this above mentioned reactor are also investigated. It was already showed that the uses and benefits of this reactor can not be limited. If the sufficient work and service is given, NDT and industrial applications can also be carried economically. (orig.)

Treatment of core components from nuclear power plants with PWR and BWR reactors - 16043

International Nuclear Information System (INIS)

Viermann, Joerg; Friske, Andreas; Radzuweit, Joerg

2009-01-01

During operation of a Nuclear Power Plant components inside the RPV get irradiated. Irradiation has an effect on physical properties of these components. Some components have to be replaced after certain neutron doses or respectively after a certain operating time of the plant. Such components are for instance water channels and control rods from Boiling Water Reactors (BWR) or control elements, poisoning elements and flow restrictors from Pressurized Water Reactors (PWR). Most of these components are stored in the fuel pool for a certain time after replacement. Then they have to be packaged for further treatment or for disposal. More than 25 years ago GNS developed a system for disposal of irradiated core components which was based on a waste container suitable for transport, storage and disposal of Intermediate Level Waste (ILW), the so-called MOSAIK R cask. The MOSAIK R family of casks is subject of a separate presentation at the ICEM 09 conference. Besides the MOSAIK R cask the treatment system developed by GNS comprised underwater shears to cut the components to size as well as different types of equipment to handle the components, the shears and the MOSAIK R casks in the fuel pool. Over a decade of experience it showed that this system although effective needed improvement for BWR plants where many water channels and control rods had to be replaced after a certain operating time. Because of the large numbers of components the time period needed to cut the components in the pool had a too big influence on other operational work like rearranging of fuel assemblies in the pool. The system was therefore further developed and again a suitable cask was the heart of the solution. GNS developed the type MOSAIK R 80 T, a cask that is capable to ship the unsegmented components with a length of approx. 4.5 m from the Power plants to an external treatment centre. This treatment centre consisting of a hot cell installation with a scrap shear, super-compactor and a heavy

Core damage frequency prespectives for BWR 3/4 and Westinghouse 4-loop plants based on IPE results

International Nuclear Information System (INIS)

Dingman, S.; Camp, S.; LaChance, J.; Mary Drouin

1995-01-01

This paper discusses the core damage frequency (CDF) insights gained by analyzing the results of the Individual Plant Examinations (IPES) for two groups of plants: boiling water reactor (BWR) 3/4 plants with Reactor Core Isolation Cooling systems, and Westinghouse 4-loop plants. Wide variability was observed for the plant CDFs and for the CDFs of the contributing accident classes. On average, transients-with loss of injection, station blackout sequences, and transients with loss of decay heat removal are important contributors for the BWR 3/4 plants, while transients, station blackout sequences, and loss-of-coolant accidents are important for the Westinghouse 4-loop plants. The key factors that contribute to the variability in the results are discussed. The results are often driven by plant-specific design and operational characteristics, but differences in modeling approaches are also important for some accident classes

Interpretation of incore noise measurements in BWR's

International Nuclear Information System (INIS)

Dam, H. van

1982-01-01

A survey is given of the main incentives for power reactor noise research and the differences and similarities of noise in power and zero power systems are touched on. The basic characteristics of the adjoint method in reactor noise theory are treated. The detector adjoint functions describe the transfer functions between spatially distributed noise sources and a (neutron or gamma) detector. In particular, the spatial dependence of these functions explains the 'local' and 'global' effects in BWR noise measurements. By including thermal hydraulic feedback effects in the adjoint analysis, it is shown that the common idea of a dominant global effect at low frequencies which should result in point kinetic behaviour, is erroneous. The same analysis provides a method for nonperturbing on-line measurement of the reactor transfer function, which is demonstrated by results from measurements on a BWR in the Netherlands. In the final part of the paper some ideas are given for further research in the field of BWR noise. (author)

Interpretation of incore noise measurements in BWR's

International Nuclear Information System (INIS)

Dam, H. van

1983-01-01

A survey is given of the main incentives for power reactor noise research, and the differences and similarities of noise in power and zero power systems are shown. After a short outline of historical developments the basic characteristics of the adjoint method in reactor noise theory are dealt with. The detector adjoint functions describe the transfer functions between spatially distributed noise sources and a (neutron or gamma) detector. In particular, the spatial dependence of these functions explains the 'local' and 'global' effects in BWR noise measurements. By including thermal hydraulic feedback effects in the adjoint analysis, it is shown that the common idea of a dominant global effect at low frequencies, which should result in point kinetic behaviour, is erroneous. The same analysis provides a method for nonperturbing on-line measurements on a BWR in The Netherlands. In the final part of the paper some ideas are given for further research in the field of BWR noise. (author)

Final air test results for the 1/5-scale Mark I boiling water reactor pressure suppression experiment

International Nuclear Information System (INIS)

Collins, E.K.; Lai, W.

1977-01-01

A loss-of-coolant accident (LOCA) in a boiling-water reactor (BWR) power plant has never occurred. However, because this type of accident is particularly severe, it is used as a principal basis for design. During a hypothetical LOCA in a Mark I BWR, air followed by steam is injected from a drywell into a toroidal wetwell about half-filled with water. A series of consistent, versatile, and accurate air-water tests simulating LOCA conditions was completed in the Lawrence Livermore Laboratory 1/5-Scale Mark I BWR Pressure Suppression Experimental Facility. Results from this test series were used to quantify the vertical loading function and to study the associated fluid dynamic phenomena. Detailed histories of vertical loads on the wetwell are shown. In particular, variations of hydrodynamic-generated vertical loads with changes in drywell pressurization rate, downcomer submergence, and the vent-line loss coefficient are established. Initial drywell overpressure, which partially preclears the downcomers of water, substantially reduces the peak vertical loads. Scaling relationships, developed from dimensional analysis and verified by bench-top experiments, allow the 1/5-scale results to be applied to a full-scale BWR power plant. This analysis leads to dimensionless groupings which are invariant. These groupongs show that if water is used as the working fluid, the magnitude of the forces in a scaled facility is reduced by the cube of the scale factor; the time when these forces occur is reduced by the square root of the scale factor

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-07-01

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

Assessment of the Prony's method for BWR stability analysis

International Nuclear Information System (INIS)

Ortiz-Villafuerte, Javier; Castillo-Duran, Rogelio; Palacios-Hernandez, Javier C.

2011-01-01

Highlights: → This paper describes a method to determine the degree of stability of a BWR. → Performance comparison between Prony's and common AR techniques is presented. → Benchmark data and actual BWR transient data are used for comparison. → DR and f results are presented and discussed. → The Prony's method is shown to be a robust technique for BWR stability. - Abstract: It is known that Boiling Water Reactors are susceptible to present power oscillations in regions of high power and low coolant flow, in the power-flow operational map. It is possible to fall in one of such instability regions during reactor startup, since both power and coolant flow are being increased but not proportionally. One other possibility for falling into those areas is the occurrence of a trip of recirculation pumps. Stability monitoring in such cases can be difficult, because the amount or quality of power signal data required for calculation of the stability key parameters may not be enough to provide reliable results in an adequate time range. In this work, the Prony's Method is presented as one complementary alternative to determine the degree of stability of a BWR, through time series data. This analysis method can provide information about decay ratio and oscillation frequency from power signals obtained during transient events. However, so far not many applications in Boiling Water Reactors operation have been reported and supported to establish the scope of using such analysis for actual transient events. This work presents first a comparison of decay ratio and frequency oscillation results obtained by Prony's method and those results obtained by the participants of the Forsmark 1 and 2 Boiling Water Reactor Stability Benchmark using diverse techniques. Then, a comparison of decay ratio and frequency oscillation results is performed for four real BWR transient event data, using Prony's method and two other techniques based on an autoregressive modeling. The four

External Cooling of the BWR Mark I and II Drywell Head as a Potential Accident Mitigation Measure - Scoping Assessment

International Nuclear Information System (INIS)

Robb, Kevin R.

2017-01-01

This report documents a scoping assessment of a potential accident mitigation action applicable to the US fleet of boiling water reactors with Mark I and II containments. The mitigation action is to externally flood the primary containment vessel drywell head using portable pumps or other means. A scoping assessment of the potential benefits of this mitigation action was conducted focusing on the ability to (1) passively remove heat from containment, (2) prevent or delay leakage through the drywell head seal (due to high temperatures and/or pressure), and (3) scrub radionuclide releases if the drywell head seal leaks.

Method of estimating thermal power distribution of core of BWR type reactor

International Nuclear Information System (INIS)

Sekimizu, Koichi

1982-01-01

Purpose: To accurately and rapidly predict the thermal power of the core of a BWR they reactor at load follow-up operating time. Method: A parameter value corrected from a correction coefficient deciding unit and a xenon density distribution value predicted and calculated from a xenon density distributor are inputted to a thermal power distribution predicting devise, the status amount such as coolant flow rate or the like predetermined at this and next high power operating times is substituted for physical model to predict and calculate the thermal power distribution. The status amount of a nuclear reactor at the time of operating in previous high power corresponding to the next high power operation to be predicted is read from the status amount of the reactor stored in time series manner is a reactor core status memory, and the physical model used in the prediction and calculation of the thermal power distribution at the time of next high power operation is corrected. (Sikiya, K.)

Visualization of neutron flux and power distributions in TRIGA Mark II reactor as an educational tool

International Nuclear Information System (INIS)

Snoj, Luka; Ravnik, Matjaz; Lengar, Igor

2008-01-01

Modern Monte Carlo computer codes (e.g. MCNP) for neutron transport allow calculation of detailed neutron flux and power distribution in complex geometries with resolution of ∼1 mm. Moreover they enable the calculation of individual particle tracks, scattering and absorption events. With the use of advanced software for 3D visualization (e.g. Amira, Voxler, etc.) one can create and present neutron flux and power distribution in a 'user friendly' way convenient for educational purposes. One can view axial, radial or any other spatial distribution of the neutron flux and power distribution in a nuclear reactor from various perspectives and in various modalities of presentation. By visualizing the distribution of scattering and absorption events and individual particle tracks one can visualize neutron transport parameters (mean free path, diffusion length, macroscopic cross section, up-scattering, thermalization, etc.) from elementary point of view. Most of the people remember better, if they visualize the processes. Therefore the representation of the reactor and neutron transport parameters is a convenient modern educational tool for the (nuclear power plant) operators, nuclear engineers, students and specialists involved in reactor operation and design. The visualization of neutron flux and power distributions in Jozef Stefan Institute TRIGA Mark II research reactor is treated in the paper. The distributions are calculated with MCNP computer code and presented using Amira and Voxler software. The results in the form of figures are presented in the paper together with comments qualitatively explaining the figures. (authors)

Damage by radiation in structural materials of BWR reactor vessels; Dano por radiacion en materiales estructurales de vasijas de reactores BWR

Energy Technology Data Exchange (ETDEWEB)

Robles, E.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E. [Departamento de Sintesis y Caracterizacion de Materiales, Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

2002-07-01

The structural materials which are manufactured the pressure vessels of the BWR reactors undergo degradation in their mechanical properties mainly due to the damage produced by the fast neutrons (E> 1 MeV) coming from the reactor core. The mechanisms of neutron damage in this type of materials are experimentally studied, through the irradiation of vessel steel in experimental reactors for a quickly ageing. Alternately the neutron damage through steel irradiation with heavy ions is simulated. In this work the first results of the damage induced by irradiation of a similar steel to the vessel of a BWR reactor are shown. The irradiation was performed with fast neutrons (E> 1 MeV, fluence of 1.45 x 10{sup 18} n/cm{sup 2}) in the TRIGA Mark III Salazar reactor and separately with Ni{sup +3} ions in a Tandetrom accelerator (E= 4.8 MeV and an ion flux rank of 0.1 to 53 ions/A{sup 2}). (Author)

Seismic risk assessment of a BWR: status report

International Nuclear Information System (INIS)

Chuang, T.Y.; Bernreuter, D.L.; Wells, J.E.; Johnson, J.J.

1985-02-01

The seismic risk methodology developed in the US NRC Seismic Safety Margins Research Program (SSMRP) was demonstrated by its application to the Zion nuclear power plant, a pressurized water reactor (PWR). A detailed model of Zion, including systems analysis models (initiating events, event trees, and fault trees), SSI and structure models, and piping models was developed and analyzed. The SSMRP methodology can equally be applied to a boiling water reactor (BWR). To demonstrate its applicability, to identify fundamental differences in seismic risk between a PWR and a BWR, and to provide a basis of comparison of seismic risk between a PWR and a BWR when analyzed with comparable methodology and assumptions, a seismic risk analysis is being performed on the LaSalle County Station nuclear power plant

Standard Technical Specifications, General Electric Plants, BWR/6

International Nuclear Information System (INIS)

1992-09-01

This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/4, and documents the positions of the Nuclear Regulatory Commission based on the BWR Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3, contains the Bases for Sections 3.4--3.10 of the improved STS

BWR Steam Dryer Alternating Stress Assessment Procedures

Energy Technology Data Exchange (ETDEWEB)

Morante, R. J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Hambric, S. A. [Brookhaven National Lab. (BNL), Upton, NY (United States); Ziada, S. [Brookhaven National Lab. (BNL), Upton, NY (United States)

2016-12-01

This report presents an overview of Boiling Water Reactor (BWR) steam dryer design; the fatigue cracking failures that occurred at the Quad Cities (QC) plants and their root causes; a history of BWR Extended Power Uprates (EPUs) in the USA; and a discussion of steam dryer modifications/replacements, alternating stress mechanisms on steam dryers, and structural integrity evaluations (static and alternating stress).

Cost reduction and safety design features of ABWR-II. Annex 5

International Nuclear Information System (INIS)

Koh, F.; Moriya, K.; Anegawa, T.

2002-01-01

The ABWR-II, which is aimed to be the next generation reactor following the latest BWR: Advanced Boiling Reactor (ABWR), is now under development jointly by the Japanese BWR utilities, General Electric Company, Hitachi Limited, and Toshiba Corporation. The key objectives of ABWR-II development include improvement in economics and further sophistication in safety for commercialization in the late 2010's and after. This paper summarizes the current status of ABWR-II development focusing on economics and safety. Plant power rating, fuel size, CRD rationalization and outage period are discussed from a cost reduction perspective. In terms of safety, the features such as diversification in emergency power sources and passive system application against severe accidents are being introduced. (author)

Simulation of a scenario of total loss of external and internal power (Sbo) for different vent pressures of the containment of a BWR-5; Simulacion de un escenario de perdida total de potencia externa e interna (SBO) para distintas presiones de venteo de la contencion de un reactor BWR-5

Energy Technology Data Exchange (ETDEWEB)

Cardenas V, J.; Mugica R, C. A.; Godinez S, V., E-mail: Jaime.cardenas@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

2014-10-15

The simulation of a Station Black Out (Sbo) was realized with intervention of the vent containment by means of a rigid vent coming from the dry-well and that discharges directly to the atmosphere, with the MELCOR code version 2.1. This scenario was carried out for a BWR-5 and containment type Mark II, with a thermal power of 2317 MWt similar to the reactor of nuclear power plant of Laguna Verde. For this scenario was considered as only available system for coolant injection to the reactor to the Reactor Core Isolation Cooling (Rcic), which remained operating 4 hours with batteries bank. The Security and Relief Valves (SR V) were considered functional (by simplicity) and that they mechanically do not exceed their capacity to liberate pressure due to the performances in their safety way. The operator maneuver to perform the SR V and to de pressurize the vessel until the pressure (13 kg/cm{sup 2}) to operate the low pressure systems was modeled. The results cover approximately 48 hours (172000 seconds), time in which was observed the behavior of the level and pressure in the vessel. Also the scenario evolution was analyzed to different vent pressures of the primary containment (2.0, 3.0, 4.5, 6.0, and 10.0 kg/cm{sup 2}), the temperature profiles of the dry-well, the hydrogen accumulation in the containment, the radio-nuclides liberation through rigid vent to the atmosphere and the inventory of these. In this work an analysis of the pressure behavior in the primary containment is presented, with the purpose of minimizing liberated fission products to the environment. (Author)

Maximum thermal loading test of BWR fuel assembly

International Nuclear Information System (INIS)

Nakajima, Yoshitaka; Yoshimura, Kunihiro; Nakamura, Satoshi; Ishizuka, Takao.

1987-01-01

Various proving tests on the reliability of nuclear power plants have been conducted at the Nuclear Power Engineering Test Center and at the Japan Power Plant Engineering and Inspection Corporation. The tests were initiated at the request of the Ministry of International Trade and Industry (MITI). Toshiba undertook one of the proving tests on the reliability of nuclear fuel assembly; the maximum thermal loading test of BWR fuel assembly from the Nuclear Power Engineering Test Center. These tests are part of the proving tests mentioned above, and their purpose is to confirm the reliability of the thermal hydraulic engineering techniques. Toshiba has been engaged for the past nine years in the design, fabrication and testing of the equipment. For the project, a test model fuel assembly was used to measure the critical power of the BWR fuel assembly and the void and fluidity of the coolant. From the test results, it has been confirmed that the heat is transferred safely from the fuel assembly to the coolant in the BWR nuclear power plant. In addition, the propriety and reliability of the thermal hydraulic engineering techniques for the fuel assembly have been proved. (author)

Results of modeling advanced BWR fuel designs using CASMO-4

International Nuclear Information System (INIS)

Knott, D.; Edenius, M.

1996-01-01

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

Metallurgical factors that contribute to cracking in BWR piping

International Nuclear Information System (INIS)

Weeks, J.R.

1975-01-01

During the fall of 1974 and early winter of 1975, cracks have been discovered in the 4 in. bypass lines of several Boiling Water Reactors (BWR's) in the United States. Further, similar cracks were discovered at two BWR's in Japan during the same period. More recently, cracks have been discovered in the core spray piping and in a furnace-sensitized ''safe end'' and adjacent ''dutchman'' at the Dresden Nuclear Power Station, Unit No. 2. Although inspections at all other U.S. BWR's have not disclosed further instances of cracking in core spray piping, leaking cracks have been found in the core spray piping of two BWR's overseas. Metallurgical examinations of these cracks are not yet complete. The following observations have been made to date. All cracks (except those in the furnace-sensitized safe end and dutchman) occurred in seamless type 304 stainless steel piping or in elbows fabricated from such piping, in the outer heat affected zone of either field or shop welds, in lines isolated from the main primary coolant flow during full power operation, except for the not yet examined cracks in the Monticello bypass lines. The cracks are exclusively intergranular, and occur in metal that has been lightly sensitized by the welding process, with only intermittent grain boundary carbides. They developed in the areas of peak axial residual stresses from welding rather than in the most heavily sensitized areas. No fatigue striations have been found on the fracture surfaces. The evidence received to date strongly indicates that these cracks were caused by intergranular stress corrosion of weld-sensitized stainless steel by BWR water containing greater than 0.2 ppM oxygen. The possible role of fatigue or alternating stresses in this corrosion is not clear. Further, not all the cracks detected to date necessarily have occurred by the same mechanism

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

International Nuclear Information System (INIS)

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

2004-01-01

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

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

International Nuclear Information System (INIS)

Adamsson, Carl; Le Corre, Jean-Marie

2011-01-01

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

Identification and assessment of containment and release management strategies for a BWR Mark II containment

International Nuclear Information System (INIS)

Lin, C.C.; Lehner, J.R.

1992-06-01

Accident management strategies that have the potential to maintain containment integrity and control or mitigate the release of radioactivity following a severe accident at a boiling water reactor with a Mark 2 type of containment are identified and evaluated. The strategies are referred to as containment and release strategies. Using information available from probabilistic risk assessments and other existing severe accident research, and employing simplified containment and release event trees, this report identified the challenges a Mark 2 containment may encounter during a severe accident, the mechanisms behind these challenges, and the strategies that could be used to mitigate the challenge. By means of a safety objective tree, the strategies are linked to the general safety objectives of containment and release management. As part of the assessment process, the strategies are applied to certain severe accident sequence categories deemed important to a Mark 2 containment. These sequence categories exhibit one or more of the following characteristics: high probability of core damage, high consequences, lead to a number of challenges, and involve the failure of multiple systems. The Limerick Generating Station is used as a representative Mark 2 plant to illustrate plant specifics in this report

Contain calculations of debris conditions adjacent to the BWR Mark I drywell shell during the later phases of a severe accident

International Nuclear Information System (INIS)

Hyman, C.R.

1988-01-01

Best estimate CONTAIN calculations have recently been performed by the BWR Severe Accident Technology (BWRSAT) Program at Oak Ridge National Laboratory to predict the primary containment response during the later phases of an unmitigated low-pressure Short Term Station Blackout at the Peach Bottom Atomic Power Station. Debris pour conditions leaving the failed reactor vessel are taken from the results of best estimate BWRSAR analyses that are based upon an assumed metallic debris melting temperature of 2750/degree/F (1783 K) and an oxide debris melting temperature of 4350/degree/F (2672 K). Results of the CONTAIN analysis for the case without sprays indicate failure of the drywell seals due to the extremely hot atmospheric conditions extant in the drywell. The maximum calculated temperature of the debris adjacent to the drywell shell is less than the melting temperature of the shell, yet the sustained temperatures may be sufficient to induce primary containment pressure boundary failure by the mechanism of creep-rupture. It is also predicted that a significant portion of the reactor pedestal wall is ablated during the period of the calculation. Nevertheless, the calculated results are recognized to be influenced by large modeling uncertainties. Several deficiencies in the application of the CORCON module within the CONTAIN code to BWR severe accident sequences are identified and discussed. 5 refs., 9 figs., 4 tabs.,

Full-scale mark II CRT program data report, (5)

International Nuclear Information System (INIS)

Kukita, Yutaka; Namatame, Ken; Yamamoto, Nobuo; Takeshita, Isao; Shiba, Masayoshi

1980-03-01

The Full-Scale Mark II CRT (Containment Response Test) Program was initiated in 1977 to provide a data base for evaluation of the LOCA hydrodynamic loads for the Mark II pressure suppression system. The test facility is 1/18 in volume and has a wetwell which is a fullscale replica of one 20 0 -sector of that of a reference Mark II. This report documents test data obtained from TEST 2101, which is a medium size (74 mm) water break test performed on April 27, 1979. TEST 2101 was designed to roughly approximate an intermediate break accident in which so-called chugging phenomenon associated with low-flux steam condensation is anticipated to continue for a longer duration than in a large break accident. (author)

Standard technical specifications: General Electric plants, BWR/4. Volume 1, Revision 1: Specifications

International Nuclear Information System (INIS)

1995-04-01

This report documents the results of the combined effort of the NRC and the industry to produce improved Standard Technical Specifications (STS), Revision 1 for General Electric BWR/4 Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

Studies of fragileness in steels of vessels of BWR reactors; Estudios de fragilizacion en aceros de vasija de reactores BWR

Energy Technology Data Exchange (ETDEWEB)

Robles, E.F.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

2003-07-01

The structural materials with those that are manufactured the pressure vessels of the BWR reactors, suffer degradation in its mechanical properties mainly to the damage taken place by the fast neutrons (E > 1 MeV) coming from the reactor core. Its are experimentally studied those mechanisms of neutron damage in this material type, by means of the irradiation of steel vessel in experimental reactors to age them quickly. Alternatively it is simulated the neutron damage by means of irradiation of steel with heavy ions. In this work those are shown first results of the damage induced by irradiation from a similar steel to the vessel of a BWR reactor. The irradiation was carried out with fast neutrons (E > 1 MeV, fluence of 1.45 x 10{sup 18} n/cm{sup 2}) in the TRIGA MARK lll reactor and separately with Ni{sup +3} ions in a Tandetrom accelerator, E = 4.8 MeV and range of the ionic flow of 0.1 to 53 iones/A{sup 2}. (Author)

External Cooling of the BWR Mark I and II Drywell Head as a Potential Accident Mitigation Measure – Scoping Assessment

Energy Technology Data Exchange (ETDEWEB)

Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-08-01

This report documents a scoping assessment of a potential accident mitigation action applicable to the US fleet of boiling water reactors with Mark I and II containments. The mitigation action is to externally flood the primary containment vessel drywell head using portable pumps or other means. A scoping assessment of the potential benefits of this mitigation action was conducted focusing on the ability to (1) passively remove heat from containment, (2) prevent or delay leakage through the drywell head seal (due to high temperatures and/or pressure), and (3) scrub radionuclide releases if the drywell head seal leaks.

Pressure and temperature analyses using GOTHIC for Mark I containment of the Chinshan Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Chen, Yen-Shu, E-mail: yschen@iner.org.t [Nuclear Engineering Division, Institute of Nuclear Energy Research, 1000, Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China); Yuann, Yng-Ruey; Dai, Liang-Che; Lin, Yon-Pon [Nuclear Engineering Division, Institute of Nuclear Energy Research, 1000, Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China)

2011-05-15

Research highlights: The Chinshan Mark I containment pressure-temperature responses are analyzed. GOTHIC is used to calculate the containment responses under three pipe break events. This study is used to support the Chinshan Stretch Power Uprate (SPU) program. The calculated peak pressure and temperature are still below the design values. The Chinshan containment integrity can be maintained under SPU condition. - Abstract: Chinshan Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/4 plant with original licensed thermal power (OLTP) of 1775 MWt for each unit. Recently, the Stretch Power Uprate (SPU) program for the Chinshan plant is being conducted to uprate the core thermal power to 1858 MWt (104.66% OLTP). In this study, the Chinshan Mark I containment pressure/temperature responses during LOCA at 105% OLTP (104.66% OLTP + 0.34% OLTP power uncertainty = 105% OLTP) are analyzed using the containment thermal-hydraulic program GOTHIC. Three kinds of LOCA (Loss of Coolant Accident) scenarios are investigated: Recirculation Line Break (RCLB), Main Steam Line Break (MSLB), and Feedwater Line Break (FWLB). In the short-term analyses, blowdown data generated by RELAP5 transient analyses are provided as boundary conditions to the GOTHIC containment model. The calculated peak drywell pressure and temperature in the RCLB event are 217.2 kPaG and 137.1 {sup o}C, respectively, which are close to the original FSAR results (219.2 kPaG and 138.4 {sup o}C). Additionally, the peak drywell temperature of 155.3 {sup o}C calculated by MSLB is presented in this study. To obtain the peak suppression pool temperature, a long-term RCLB analysis is performed using a simplified RPV (Reactor Pressure Vessel) volume to calculate blowdown flow rate. One RHR (Residual Heat Removal) heat exchanger is assumed to be inoperable for suppression pool cooling mode. The calculated peak suppression pool temperature is 93.2 {sup o}C, which is below the pool temperature used for evaluating the

Pressure and temperature analyses using GOTHIC for Mark I containment of the Chinshan Nuclear Power Plant

International Nuclear Information System (INIS)

Chen, Yen-Shu; Yuann, Yng-Ruey; Dai, Liang-Che; Lin, Yon-Pon

2011-01-01

Research highlights: → The Chinshan Mark I containment pressure-temperature responses are analyzed. → GOTHIC is used to calculate the containment responses under three pipe break events. → This study is used to support the Chinshan Stretch Power Uprate (SPU) program. → The calculated peak pressure and temperature are still below the design values. → The Chinshan containment integrity can be maintained under SPU condition. - Abstract: Chinshan Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/4 plant with original licensed thermal power (OLTP) of 1775 MWt for each unit. Recently, the Stretch Power Uprate (SPU) program for the Chinshan plant is being conducted to uprate the core thermal power to 1858 MWt (104.66% OLTP). In this study, the Chinshan Mark I containment pressure/temperature responses during LOCA at 105% OLTP (104.66% OLTP + 0.34% OLTP power uncertainty = 105% OLTP) are analyzed using the containment thermal-hydraulic program GOTHIC. Three kinds of LOCA (Loss of Coolant Accident) scenarios are investigated: Recirculation Line Break (RCLB), Main Steam Line Break (MSLB), and Feedwater Line Break (FWLB). In the short-term analyses, blowdown data generated by RELAP5 transient analyses are provided as boundary conditions to the GOTHIC containment model. The calculated peak drywell pressure and temperature in the RCLB event are 217.2 kPaG and 137.1 o C, respectively, which are close to the original FSAR results (219.2 kPaG and 138.4 o C). Additionally, the peak drywell temperature of 155.3 o C calculated by MSLB is presented in this study. To obtain the peak suppression pool temperature, a long-term RCLB analysis is performed using a simplified RPV (Reactor Pressure Vessel) volume to calculate blowdown flow rate. One RHR (Residual Heat Removal) heat exchanger is assumed to be inoperable for suppression pool cooling mode. The calculated peak suppression pool temperature is 93.2 o C, which is below the pool temperature used for evaluating the

Phenomenology of BWR fuel assembly degradation

Science.gov (United States)

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

2018-03-01

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

Some first results from the mark II at SPEAR

CERN Document Server

Abrams, G S; Blocker, C.A.; Boyarski, A.M.; Breidenbach, Martin; Broll, C.H.; Burke, D.L.; Carithers, W.C.; Chinowsky, William; Coles, M.W.; Cooper, S.; Couchman, B.; Dieterle, W.E.; Dillon, J.B.; Dorenbosch, J.; Dorfan, J.M.; Eaton, M.W.; Feldman, G.J.; Fischer, H.G.; Franklin, M.E.B.; Gidal, G.; Goldhaber, G.; Hanson, G.; Hayes, K.G.; Himel, T.; Hitlin, D.G.; Hollebeek, R.J.; Innes, Walter R.; Jaros, J.A.; Jenni, P.; Johnson, Alfred D.; Kadyk, J.A.; Lankford, A.J.; Larsen, Rudolf R.; Longo, Michael J.; Luke, D.; Luth, V.; Martin, J.F.; Millikan, R.E.; Nelson, M.E.; Pang, C.Y.; Patrick, J.F.; Perl, Martin L.; Richter, Burton; Russell, J.J.; Scharre, D.L.; Schindler, R.H.; Schwitters, R.; Shannon, S.; Siegrist, J.; Strait, J.; Taureg, H.; Telnov, Valery I.; Tonutti, M.; Trilling, G.; Vella, E.; Vidal, R.A.; Videau, I.; Weiss, J.; Zaccone, H.

1979-01-01

Preliminary results are given from the Mark II experiment at SPEAR on radiative decays of the psi '(3684) and on inclusive baryon production from 3.67 to 7.4 GeV center-of-mass energy. A 90% confidence level upper limit of 0.12% is given for BR( psi ' to gamma eta /sub c/' (3455))*BR( eta /sub c/'(3455 to gamma psi )). (10 refs).

An ecological interface design for BWR nuclear power plants

International Nuclear Information System (INIS)

Monta, K.; Itoh, J.

1992-01-01

An ecological interface design was applied to realize the support function for the operator's direct perception and analytical reasoning in the development of an intelligent man-machine system for BWR nuclear power plants. The abstraction-aggregation functional hierarchy representation of the work domain is a base of the ecological interface design. Another base is the concept of the level of cognitive control. The former was mapped into the interface to externalize the operator's normative mental model of the plants, which will reduce his/her cognitive work load and support knowledge-based problem solving. In addition, the same framework can be used for the analytical evaluation of man-machine interfaces. The information content and structure of a prototype interface were evaluated. This approach seems promising from these experiences. (author)

Peach Bottom transient analysis with BWR TRACB02

International Nuclear Information System (INIS)

Alamgir, M.; Sutherland, W.A.

1984-01-01

TRAC calculations have been performed for a Turbine Trip transient (TT1) in the Peach Bottom BWR power plant. This study is a part of the qualification of the BWR-TRAC code. The simulation is aimed at reproducing the observed thermal hydraulic behavior in a pressurization transient. Measured core power is an input to the calculation. Comparison with data show the code reasonably well predicts the generation and propagation of the pressure waves in the main steam line and associated pressurization of the reactor vessel following the closure of the turbine stop valve

EPRI BWR Water Chemistry Guidelines Revision

International Nuclear Information System (INIS)

Garcia, Susan E.; Giannelli, Joseph F.

2014-01-01

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

U.S. Department Of Energy's nuclear engineering education research: highlights of recent and current research-II. 7. Hybrid Reactor Simulation and 3-D Information Display of BWR Out-of-Phase Oscillation

International Nuclear Information System (INIS)

Edwards, Robert; Huang, Zhengyu

2001-01-01

The real-time hybrid reactor simulation (HRS) capability of the Penn State TRIGA reactor has been recently expanded for BWR out-of-phase behavior. Out-of-phase oscillation is a phenomenon that occurs at BWRs. During this kind of event, half of the core can significantly oscillate out of phase with the other half, while the average power reported by the neutronic instrumentation may show a much lower amplitude for the oscillations. The HRS will be used for development and validation of stability monitoring and control techniques as part of an ongoing U.S. Department of Energy Nuclear Engineering Education and Research grant. The Penn State TRIGA reactor is used to simulate BWR fundamental mode power dynamics. The first harmonic mode power, together with detailed thermal hydraulics of boiling channels of both fundamental mode and first harmonic mode, is simulated digitally in real time with a computer. Simulations of boiling channels provide reactivity feedback to the TRIGA reactor, and the TRIGA reactor's power response is in turn fed into the channel simulations and the first harmonic mode power simulation. The combination of reactor power response and the simulated first harmonic power response with spatial distribution functions thus mimics the stability phenomena actually encountered in BWRs. The digital simulations of the boiling channels are performed by solving conservation equations for different regions in the channel with C-MEX S-functions. A fast three-dimensional (3-D) reactor power display of modal BWR power distribution was implemented using MATLAB graphics capability. Fundamental mode, first harmonic, together with the total power distribution over the reactor cross section, are displayed. Because of the large amount of computation for BWR boiling channel simulation and real-time data processing and graph generation, one computer is not sufficient to handle these jobs in the hybrid reactor simulation environment. A new three-computer setup has been

Power level control of the TRIGA Mark-II research reactor using the multifeedback layer neural network and the particle swarm optimization

International Nuclear Information System (INIS)

Coban, Ramazan

2014-01-01

Highlights: • A multifeedback-layer neural network controller is presented for a research reactor. • Off-line learning of the MFLNN is accomplished by the PSO algorithm. • The results revealed that the MFLNN–PSO controller has a remarkable performance. - Abstract: In this paper, an artificial neural network controller is presented using the Multifeedback-Layer Neural Network (MFLNN), which is a recently proposed recurrent neural network, for neutronic power level control of a nuclear research reactor. Off-line learning of the MFLNN is accomplished by the Particle Swarm Optimization (PSO) algorithm. The MFLNN-PSO controller design is based on a nonlinear model of the TRIGA Mark-II research reactor. The learning and the test processes are implemented by means of a computer program at different power levels. The simulation results obtained reveal that the MFLNN-PSO controller has a remarkable performance on the neutronic power level control of the reactor for tracking the step reference power trajectories

Standard Technical Specifications General Electric plants, BWR/4:Bases (Sections 3.4-3.10). Volume 3, Revision 1

International Nuclear Information System (INIS)

1995-04-01

This report documents the results of the combined effort of the NRC and the industry to produce improved Standard Technical Specifications (STS), Revision 1 for General Electric BWR/4 Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the specifications for all chapters and sections of the improved STS. Volume 2 contains he Bases for Chapters 2.0 and 3.0, and Sections 3.1-3.3 of the improved STS. This document, Volume 3, contains the Bases for Sections 3.4-3.10 of the improved STS

Moderator temperature coefficient in BWR core

International Nuclear Information System (INIS)

Naito, Yoshitaka

1977-01-01

Temperature dependences of infinite multiplication factor k sub(infinity) and neutron leakage from the core must be examined for estimation of moderator temperature coefficient. Temperature dependence on k sub(infinity) has been investigated by many researchers, however, the dependence on neutron leakage of a BWR with cruciformed control rods has hardly been done. Because there are difficulties and necessity on calculations of three space dimensional and multi-energy groups neutron distribution in a BWR core. In this study, moderator temperature coefficients of JPDR-II (BWR) core were obtained by calculation with DIFFUSION-ACE, which is newly developed three-dimensional multi-group computer code. The results were compared with experimental data measured from 20 to 275 0 C of the moderator temperature and the good agreement was obtained between calculation and measurement. In order to evaluate neutron leakage from the core, the other two calculations were carried out, adjusting criticality by uniform absorption rate and by material buckling. The former underestimated neutron leakage and the latter overestimated it. Discussion on the results shows that in order to estimate the temperature coefficient of BWR, neutron leakage must be evaluated precisely, therefore the calculation at actual pattern of control rods is necessary. (auth.)

ROSA-III/971, BWR Rig of Safety Assessment LOCA, Loss of Offsite Power Transient

International Nuclear Information System (INIS)

1992-01-01

1 - Description of test facility: ROSA-III is a 1/124 scaled down test facility with electrically heated core designed to study the response of engineered safety features to loss-of-coolant accidents in in commercial BWR. It consists of the following, fully instrumented subsystems: (a) the pressure vessel with a core simulating four half-length fuel assemblies and control rod; (b) steam line and feed water line, which are independent open loops; (c) coolant recirculation system, which consists of two loops provided with a recirculation pump and two jet pumps in each loop; (d) emergency cooling system, including HPCS, LPCS, LPCI, and ADS. 2 - Description of test: Run 971 simulated a BWR LOSS of off-site power transient. The core scram was assumed to occur at 6 seconds after the transient initiated by the turbine trip. HPCS failure was assumed. After ADS started, the upper half of the core was uncovered by steam. The core was re-flooded by LPCS alone

Investigation of BWR stability in Forsmark 2

International Nuclear Information System (INIS)

Oguma, R.; Reisch, F.; Bergdahl, B.G.; Lorenzen, J.; Aakerhielm, F.; Kellner, S.

1988-01-01

A series of noise measurements have been conducted at the Forsmark-2 reactor during its start-up operation after the revision in 1987. The main purpose was to investigate the BWR stability problem based on noise analysis, i.e. the problem of resonant power oscillation with frequency of about 0.5 Hz, which tends to arise at high power and low core flow condition. The noise analysis was performed to estimate the noise source which gives rise to the power oscillation, to evaluate the stability condition of the Forsmark-2 reactor in terms of the decay ratio (DR), as well as to investigate a safety related problem in connection with the BWR stability. The results indicate that the power oscillation is due to dynamic coupling between the neutron kinetics and thermal-hydraulics via void reactivity feedback. The DR reached as high as ≅ 0.7 at 63% of the rated power and 4100 kg/s of the total core flow. An investigation was made for the noise recording which represents a strong pressure oscillation with a peak frequency at 0.33 Hz. The result suggests that such pressure oscillation, if the peak frequency coincided with that of the resonant power oscillation, might become a cause of scram. The present noise analysis indicates the importance of a BWR on-line surveillance system with functions like stability condition monitoring and control system diagnosis. (orig.)

Description of the power plant model BWR-plasim outlined for the Barsebaeck 2 plant

International Nuclear Information System (INIS)

Christensen, P. la Cour.

1979-08-01

A description is given of a BWR power plant model outlined for the Barsebaeck 2 plant with data placed at our disposal by the Swedish Power Company Sydkraft A/B. The basic operations are derived and simplifications discussed. The model is implemented with a simulation system DYSYS which assures reliable solutions and easy programming. Emphasis has been placed on the models versatility and flexibility so new features are easy to incorporate. The model may be used for transient calculations for both normal plant conditions and for abnormal occurences as well as for control system studies. (author)

Mark II containment lead plant program load evaluation and acceptance criteria

International Nuclear Information System (INIS)

1978-10-01

The report, prepared by the Office of Nuclear Reactor Regulation, addresses the portion of the Mark II owner's program that provides a generic methodology for establishing design basis LOCA and safety relief valve loads for the lead Mark II facilities (Zimmer, Shoreham, and LaSalle), i.e., the lead plant program. The report includes an evaluation of the Mark II owner's load methodology, a description of load methodologies that we find acceptable for use in the individual plant unique assessments, and the basis for the stated conclusions

Preliminary neutronic design of TRIGA Mark II Reactor

International Nuclear Information System (INIS)

Sarikaya, B.; Tombakoglu, M.; Cecen, Y.; Kadiroglu, O. K.

2001-01-01

It is very important to analyse the behaviour of the research reactors, since, they play a key role in developing the power reactor technology and radiation applications such as isotope generation for medical treatments. In this study, the neutronic behaviour of the TRIGA MARK II reactor, owned and operated by Istanbul Technical University is analysed by using the SCALE code system. In the analysis, in order to overcome the disadvantages of special TRIGA codes, such as TRIGAP, the SCALE code system is chosen to perform the calculations. TRIGAP and similar codes have limited geometrical (one-dimensional geometry) and cross sectional options (two-group calculations), however, SCALE has the capability of wider range of geometrical modelling capability (three-dimensional modelling is possible) and multi-group calculations are possible

Utility experience with BWR-PSMS

International Nuclear Information System (INIS)

Bond, G.R.

1986-01-01

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

Damage by radiation in structural materials of BWR reactor vessels

International Nuclear Information System (INIS)

Robles, E.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E.

2002-01-01

The structural materials which are manufactured the pressure vessels of the BWR reactors undergo degradation in their mechanical properties mainly due to the damage produced by the fast neutrons (E> 1 MeV) coming from the reactor core. The mechanisms of neutron damage in this type of materials are experimentally studied, through the irradiation of vessel steel in experimental reactors for a quickly ageing. Alternately the neutron damage through steel irradiation with heavy ions is simulated. In this work the first results of the damage induced by irradiation of a similar steel to the vessel of a BWR reactor are shown. The irradiation was performed with fast neutrons (E> 1 MeV, fluence of 1.45 x 10 18 n/cm 2 ) in the TRIGA Mark III Salazar reactor and separately with Ni +3 ions in a Tandetrom accelerator (E= 4.8 MeV and an ion flux rank of 0.1 to 53 ions/A 2 ). (Author)

Studies of fragileness in steels of vessels of BWR reactors

International Nuclear Information System (INIS)

Robles, E.F.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E.

2003-01-01

The structural materials with those that are manufactured the pressure vessels of the BWR reactors, suffer degradation in its mechanical properties mainly to the damage taken place by the fast neutrons (E > 1 MeV) coming from the reactor core. Its are experimentally studied those mechanisms of neutron damage in this material type, by means of the irradiation of steel vessel in experimental reactors to age them quickly. Alternatively it is simulated the neutron damage by means of irradiation of steel with heavy ions. In this work those are shown first results of the damage induced by irradiation from a similar steel to the vessel of a BWR reactor. The irradiation was carried out with fast neutrons (E > 1 MeV, fluence of 1.45 x 10 18 n/cm 2 ) in the TRIGA MARK lll reactor and separately with Ni +3 ions in a Tandetrom accelerator, E = 4.8 MeV and range of the ionic flow of 0.1 to 53 iones/A 2 . (Author)

Siemens Nuclear Power Corporation experience with BWR and PWR fuels

International Nuclear Information System (INIS)

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

1992-01-01

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

Recent operating experience during startup testing at latest 1100 MWe BWR-5 nuclear power plants

International Nuclear Information System (INIS)

Tanabe, Akira; Tateishi, Mizuo; Kajikawa, Makoto; Hayase, Yuichi.

1986-01-01

In June and September 1985, the latest two 1100 Mwe BWR-5 nuclear power plants started commercial operation about ten days earlier than initially expected without any unscheduled shutdown. These latest plants, 2F-3 and K-1, are characterized by an improved core with new 8 x 8 fuel assemblies, highly reliable control systems, advanced control room system and turbine steam full bypass system for full load rejection (2F3). This paper describes the following operating experiences gained during their startup testing. 1) Continuous operation at full load rejection. 2) Stable operation at natural circulating flow condition. 3) 31 and 23 hour short time start up operation. 4) 100-75-100 %, 1-8-1-14 hours daily load following operation. (author)

Mark I containment, short term program. Safety evaluation report

International Nuclear Information System (INIS)

1977-12-01

Presented is a Safety Evaluation Report (SER) prepared by the Office of Nuclear Reactor Regulation addressing the Short Term Program (STP) reassessment of the containment systems of operating Boiler Water Reactor (BWR) facilities with the Mark I containment system design. The information presented in this SER establishes the basis for the NRC staff's conclusion that licensed Mark I BWR facilities can continue to operate safely, without undue risk to the health and safety of the public, during an interim period of approximately two years while a methodical, comprehensive Long Term Program (LTP) is conducted. This SER also provides one of the basic foundations for the NRC staff review of the Mark I containment systems for facilities not yet licensed for operation

Emergency response facility technical data system of Taiwan Power Company

International Nuclear Information System (INIS)

Lin, E.; Liang, T.M.

1987-01-01

Taiwan Power Company (Taipower) has developed its emergency response facility program since 1981. This program is integrated with the following activities to enhance the emergency response capability of nuclear power plants: (1) survey of the plant instrumentation based on the requirements of R.G. 1.97; (2) improvement of plant specific emergency operating procedures based on the emergency response guidelines developed by the Owners group; (3) implementation of the detailed control room design review with the consideration of human engineering and task analysis; and (4) organization, staff and communication of emergency planning of nuclear power plant. The emergency response facility programs of Taipower are implemented in Chinshan (GE BWR4/MARK I), Kuosheng (GE BWR6/MARK III) and Maanshan (W PWR). The major items included in each program are: (1) to establish new buildings for On-Site Technical Support Center, Near-Site Emergency Operation Facility; (2) to establish an Emergency Executive Center at Taipower headquarters; (3) to establish the communication network between control room and emergency response facilities; and (4) to install a dedicated Emergency Response Facility Technical Data System (ERFTDS) for each plant. The ERFTDS provides the functions of data acquisition, data processing, data storage and display in meeting with the requirements of NUREG 0696. The ERFTDS is designed with plant specific requirements. These specific requirements are expected to be useful not only for the emergency condition but also for normal operation conditions

A digital data acquisition and display system for ITU TRIGA Mark II reactor

International Nuclear Information System (INIS)

Can, B.; Omuz, S.

2008-01-01

Full text: In this study, a digital data acquisition and display system realized for ITU TRIGA Mark-II Reactor is described. This system is realized in order to help the reactor operator and to increase reactor console capacity. The system consists of two main units, which are host computers and RTI-815F, analog devices, data acquisition card. RTI-815F is multi-function analog/digital input/output board that plugs into one of the available long expansion slots in the IBM-PC, PC/XT, PC/AT, or equivalent personal computers. It has 16 analog input channels for single-ended input signals or 8 analog input channels for differential input signals. But its channel capacity can be increased to 32 input channels for single-ended input signals or 16 input channels for differential input signals. RTI-815F board contains 2 analog output channels, 8 digital input channels and 8 digital output channels. In the ITD TRIGA Mark-II Reactor, 6 fuel temperature channels, 3 water temperature channels, 3 control rod position channels and 4 power channels are chosen as analog input signals for RTI-815F. Its digital outputs are assigned to cooling tower fan, primary and secondary pump reactor scram, control rod rundown. During operation, data are automatically archived to disk and displayed on screen. The channel selection time and sampling time can be adjusted. The simulated movement and position of control rods in the reactor core can be noted and displayed. The changes of power, fuel temperature and water temperature can be displayed on the screen as a graphic. In this system both period and reactivity are calculated and displayed on the screen. (authors)

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

International Nuclear Information System (INIS)

Cunningham, M.; Scott, H.

1998-01-01

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

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

International Nuclear Information System (INIS)

Harrington, R.M.

1987-01-01

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

BWR Refill-Reflood Program. Final report

International Nuclear Information System (INIS)

Myers, L.L.

1983-09-01

The BWR Refill-Reflood Program is part of the continuing Loss of Coolant Accident (LOCA) research in the United States which is jointly sponsored by the Nuclear Regulatory Commission, the Electric Power Research Institute, and the General Electric Company. The current program expanded the focus of this research to include full scale experimental evaluations of multidimensional and multichannel effects during system refill. The program has also made major contributions to the BWR version of the Transient Reactor Analysis Code (TRAC) which has been developed cooperatively with the Idaho National Engineering Laboratory (INEL) for application to BWR transients. A summary description of the complete program is provided including the principal findings and main conclusions of the program. The results of the program have shown that multidimensional and parallel channel effects have the potential to significantly improve the system response over that observed in single channel tests

Decommissioning of TRIGA Mark II type reactor

Energy Technology Data Exchange (ETDEWEB)

Hwang, Dooseong; Jeong, Gyeonghwan; Moon, Jeikwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-10-15

The first research reactor in Korea, KRR 1, is a TRIGA Mark II type with open pool and fixed core. Its power was 100 kWth at its construction and it was upgraded to 250 kWth. Its construction was started in 1957. The first criticality was reached in 1962 and it had been operated for 36,000 hours. The second reactor, KRR 2, is a TRIGA Mark III type with open pool and movable core. These reactors were shut down in 1995, and the decision was made to decommission both reactors. The aim of the decommissioning activities is to decommission the KRR 2 reactor and decontaminate the residual building structures and site, and to release them as unrestricted areas. The KRR 1 reactor was decided to be preserve as a historical monument. A project was launched for the decommissioning of these reactors in 1997, and approved by the regulatory body in 2000. A total budget for the project was 20.0 million US dollars. It was anticipated that this project would be completed and the site turned over to KEPCO by 2010. However, it was discovered that the pool water of the KRR 1 reactor was leaked into the environment in 2009. As a result, preservation of the KRR 1 reactor as a monument had to be reviewed, and it was decided to fully decommission the KRR 1 reactor. Dismantling of the KRR 1 reactor takes place from 2011 to 2014 with a budget of 3.25 million US dollars. The scope of the work includes licensing of the decommissioning plan change, removal of pool internals including the reactor core, removal of the thermal and thermalizing columns, removal of beam port tubes and the aluminum liner in the reactor tank, removal of the radioactive concrete (the entire concrete structure will not be demolished), sorting the radioactive waste (concrete and soil) and conditioning the radioactive waste for final disposal, and final statuses of the survey and free release of the site and building, and turning over the site to KEPCO. In this paper, the current status of the TRIGA Mark-II type reactor

BWR plant analyzer development at BNL

International Nuclear Information System (INIS)

Cheng, H.S.; Wulff, W.; Mallen, A.N.; Lekach, S.V.; Stritar, A.; Cerbone, R.J.

1985-01-01

Advanced technology for high-speed interactive nuclear power plant simulations is of great value for timely resolution of safety issues, for plant monitoring, and for computer-aided emergency responses to an accident. Presented is the methodology employed at BNL to develop a BWR plant analyzer capable of simulating severe plant transients at much faster than real-time process speeds. Five modeling principles are established and a criterion is given for selecting numerical procedures and efficient computers to achieve the very high simulation speeds. Typical results are shown to demonstrate the modeling fidelity of the BWR plant analyzer

BWR Fuel Assemblies Physics Analysis Utilizing 3D MCNP Modeling

International Nuclear Information System (INIS)

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

2008-01-01

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

BWR Fuel Assemblies Physics Analysis Utilizing 3D MCNP Modeling

Energy Technology Data Exchange (ETDEWEB)

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

2008-07-01

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

Reducing BWR O and M costs through on-line performance monitoring

International Nuclear Information System (INIS)

Jonas, T.; Gross, R.; Logback, F.; Josyula, R.

1995-01-01

Competition in the electric power industry has placed significant emphasis on reducing operating and maintenance (O and M) costs at nuclear facilities. Therefore, on-line performance monitoring to locate power losses for boiling water reactor (BWR) plants is creating tremendous interest. In addition, the ability to automate activities such as data collection, analysis, and reporting increases the efficiency of plant engineers and gives them more time to concentrate on solving plant efficiency problems. This capability is now available with a unique software product called GEBOPS. GE Nuclear Energy, in conjunction with Joint Venture partner Black and Veatch, has undertaken development of the General Electric/Black and Veatch On-line Performance System (GEBOPS), an on-line performance monitoring system for BWR plants. The experience and expertise of GE Nuclear Energy with BWR plants, coupled with the proven on-line monitoring software development experience and capability of Black and Veatch, provide the foundation for a unique product which addresses the needs of today's BWR plants

BWR stability analysis at Brookhaven National Laboratory

International Nuclear Information System (INIS)

Wulff, W.; Cheng, H.S.; Mallen, A.N.; Rohatgi, U.S.

1991-01-01

Following the unexpected, but safely terminated, power and flow oscillations in the LaSalle-2 Boiling Water Reactor (BWR) on March 9, 1988, the Nuclear Regulatory Commission (NRC) Offices of Nuclear Reactor Regulation (NRR) and of Analysis and Evaluation of Operational Data (AEOD) requested that the Office of Nuclear Regulatory Research (RES) carry out BWR stability analyses, centered around fourteen specific questions. Ten of the fourteen questions address BWR stability issues in general and are dealt with in this paper. The other four questions address local, out-of-phase oscillations and matters of instrumentation; they fall outside the scope of the work reported here. It was the purpose of the work documented in this report to answer ten of the fourteen NRC-stipulated questions. Nine questions are answered by analyzing the LaSalle-2 instability and related BWR transients with the BNL Engineering Plant Analyzer (EPA) and by performing an uncertainty assessment of the EPA predictions. The tenth question is answered on the basis of first principles. The ten answers are summarized

Kuosheng BWR/6 containment safety analysis with gothic code

International Nuclear Information System (INIS)

Lin Ansheng; Wang Jongrong; Yuann Rueyyng; Shih Chunkuan

2011-01-01

Kuosheng Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/6 plant, each unit rated at 2894 MWt. In this study, we presented the calculated results of the containment pressure and temperature responses after the main steam line break accident, which is the design basis for the containment system. During the simulation, a power of SPU range (105.1%) was used and a model of the Mark III type containment was built using the containment thermal-hydraulic program GOTHIC. The simulation consists of short and long-term responses. The drywell pressure and temperature responses which display the maximum values in the early state of the LOCA were investigated in the short-term response; the primary containment pressure and temperature responses in the long-term response. The blowdown flow was provided by FSAR and used as boundary conditions in the short-term model; in the long-term model, the blowdown flow was calculated using a GOTHIC built-in homogeneous equilibrium model. In the long-term analysis, a simplifier RPV model was employed to calculate the blowdown flow. Finally, the calculated results, similar to the FSAR results, indicate the GOTHIC code has the capability to simulate the pressure/temperature response of Mark III containment to the main steam line break LOCA. (author)

Operating experience with BWR nuclear power

International Nuclear Information System (INIS)

Bonsdorf, Magnus von.

1986-01-01

The two-unit nuclear power station in Olkiluoto on the western coast of Finland produces about 20 per cent of the electricity consumption of the country. The first unit, TVO-I was first connected to the national grid in September 1978 and TVO-II in February 1980. The original rated power output of each unit was 660 MWe, corresponding to the thermal power of 2000 MW from the reactor. Technical modifications allowed the power to be uprated by 8%. The operating statistics (load factors etc.) are given and the outage experience discussed. The radiological history shows very low radioactivity and dose levels have been maintained at the plant. (UK)

BWR core melt progression phenomena: Experimental analyses

International Nuclear Information System (INIS)

Ott, L.J.

1992-01-01

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

The HAMBO BWR simulator of HAMMLAB

International Nuclear Information System (INIS)

Karlsson, Tommy; Jokstad, Haakon; Meyer, Brita D.; Nihlwing, Christer; Norrman, Sixten; Puska, Eija Karita; Raussi, Pekka; Tiihonen, Olli

2001-02-01

Modernisation of control rooms of the nuclear power plants has been a major issue in Sweden and Finland the last few years, and this will continue in the years to come. As an aid in the process of introducing new technology into the control rooms, the benefit of having an experimental simulator where proto typing of solutions can be performed, has been emphasised by many plants. With this as a basis, the BWR plants in Sweden and Finland decided to fund, in co-operation with the Halden Project, an experimental BWR simulator based on the Forsmark 3 plant in Sweden. The BWR simulator development project was initiated in January 1998. VTT Energy in Finland developed the simulator models with the aid of their APROS tool, while the operator interface was developed by the Halden Project. The simulator was thoroughly tested by experienced HRP personnel and professional Forsmark 3 operators, and accepted by the BWR utilities in June 2000. The acceptance tests consisted of 19 well-defined transients, as well as the running of the simulator from full power down to cold shutdown and back up again with the use of plant procedures. This report describes the HAMBO simulator, with its simulator models, the operator interface, and the underlying hardware and software infrastructure. The tools used for developing the simulator, APROS, Picasso-3 and the Integration Platform, are also briefly described. The acceptance tests are described, and examples of the results are presented, to illustrate the level of validation of the simulator. The report concludes with an indication of the short-term usage of the simulator. (Author)

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

International Nuclear Information System (INIS)

Hida, Kazuki; Yoshioka, Ritsuo

1992-01-01

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

Modernization design of neutron radiography of ITU TRIGA Mark-II reactor

International Nuclear Information System (INIS)

Tugrul, B.; Bilge, A.N.

1988-01-01

ITU TRIGA MARK-II Research and Training Reactor has a power of 250 KW and has three beam tubes. One of them is tangential beam tube used for neutron radiography. In this study, the neutron radiography set in the tangential beam tube is described with its problems for ITU TRIGA Reactor. After that modernization of the system is designed and the applicability of the direct and indirect methods is evaluated. Improving the ratio of length to diameter for the beam tube, elimination the fogging on the film and constructive design for practice and secure application of the technique is developed. (author)

Kuosheng Mark III containment analyses using GOTHIC

Energy Technology Data Exchange (ETDEWEB)

Lin, Ansheng, E-mail: samuellin1999@iner.gov.tw; Chen, Yen-Shu; Yuann, Yng-Ruey

2013-10-15

Highlights: • The Kuosheng Mark III containment model is established using GOTHIC. • Containment pressure and temperature responses due to LOCA are presented. • The calculated results are all below the design values and compared with the FSAR results. • The calculated results can be served as an analysis reference for an SPU project in the future. -- Abstract: Kuosheng nuclear power plant in Taiwan is a twin-unit BWR/6 plant, and both units utilize the Mark III containment. Currently, the plant is performing a stretch power uprate (SPU) project to increase the core thermal power to 103.7% OLTP (original licensed thermal power). However, the containment response in the Kuosheng Final Safety Analysis Report (FSAR) was completed more than twenty-five years ago. The purpose of this study is to establish a Kuosheng Mark III containment model using the containment program GOTHIC. The containment pressure and temperature responses under the design-basis accidents, which are the main steam line break (MSLB) and the recirculation line break (RCLB) accidents, are investigated. Short-term and long-term analyses are presented in this study. The short-term analysis is to calculate the drywell peak pressure and temperature which happen in the early stage of the LOCAs. The long-term analysis is to calculate the peak pressure and temperature of the reactor building space. In the short-term analysis, the calculated peak drywell to wetwell differential pressure is 140.6 kPa for the MSLB, which is below than the design value of 189.6 kPa. The calculated peak drywell temperature is 158 °C, which is still below the design value of 165.6 °C. In addition, in the long-term analysis, the calculated peak containment pressure is 47 kPa G, which is below the design value of 103.4 kPa G. The calculated peak values of containment temperatures are 74.7 °C, which is lower than the design value of 93.3 °C. Therefore, the Kuosheng Mark III containment can maintain the integrity after

A BWR 24-month cycle analysis using multicycle techniques

International Nuclear Information System (INIS)

Hartley, K.D.

1993-01-01

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

Safety analysis of thorium-based fuels in the General Electric Standard BWR

International Nuclear Information System (INIS)

Colby, M.J.; Townsend, D.B.; Kunz, C.L.

1980-06-01

A denatured (U-233/Th)O 2 fuel assembly has been designed which is energy equivalent to and hardware interchangeable with a modern boiling water reactor (BWR) reference reload assembly. Relative to the reference UO 2 fuel, the thorium fuel design shows better performance during normal and transient reactor operation for the BWR/6 product line and will meet or exceed current safety and licensing criteria. Power distributions are flattened and thermal operating margins are increased by reduced steam void reactivity coefficients caused by U-233. However, a (U-233/Th)O 2 -fueled BWR will likely have reduced operating flexibility. A (U-235/Th)O 2 -fueled BWR should perform similar to a UO 2 -fueled BWR under all operating conditions. A (Pu/Th)O 2 -fueled BWR may have reduced thermal margins and similar accident response and be less stable than a UO 2 -fueled BWR. The assessment is based on comparisions of point model and infinite lattice predictions of various nuclear reactivity parameters, including void reactivity coefficients, Doppler reactivity coefficients, and control blade worths

Distributed control and data processing system with a centralized database for a BWR power plant

International Nuclear Information System (INIS)

Fujii, K.; Neda, T.; Kawamura, A.; Monta, K.; Satoh, K.

1980-01-01

Recent digital techniques based on changes in electronics and computer technologies have realized a very wide scale of computer application to BWR Power Plant control and instrumentation. Multifarious computers, from micro to mega, are introduced separately. And to get better control and instrumentation system performance, hierarchical computer complex system architecture has been developed. This paper addresses the hierarchical computer complex system architecture which enables more efficient introduction of computer systems to a Nuclear Power Plant. Distributed control and processing systems, which are the components of the hierarchical computer complex, are described in some detail, and the database for the hierarchical computer complex is also discussed. The hierarchical computer complex system has been developed and is now in the detailed design stage for actual power plant application. (auth)

Predictive uncertainty reduction in coupled neutron-kinetics/thermal hydraulics modeling of the BWR-TT2 benchmark

Energy Technology Data Exchange (ETDEWEB)

Badea, Aurelian F., E-mail: aurelian.badea@kit.edu [Karlsruhe Institute of Technology, Vincenz-Prießnitz-Str. 3, 76131 Karlsruhe (Germany); Cacuci, Dan G. [Center for Nuclear Science and Energy/Dept. of ME, University of South Carolina, 300 Main Street, Columbia, SC 29208 (United States)

2017-03-15

Highlights: • BWR Turbine Trip 2 (BWR-TT2) benchmark. • Substantial (up to 50%) reduction of uncertainties in the predicted transient power. • 6660 uncertain model parameters were calibrated. - Abstract: By applying a comprehensive predictive modeling methodology, this work demonstrates a substantial (up to 50%) reduction of uncertainties in the predicted total transient power in the BWR Turbine Trip 2 (BWR-TT2) benchmark while calibrating the numerical simulation of this benchmark, comprising 6090 macroscopic cross sections, and 570 thermal-hydraulics parameters involved in modeling the phase-slip correlation, transient outlet pressure, and total mass flow. The BWR-TT2 benchmark is based on an experiment that was carried out in 1977 in the NPP Peach Bottom 2, involving the closure of the turbine stop valve which caused a pressure wave that propagated with attenuation into the reactor core. The condensation of the steam in the reactor core caused by the pressure increase led to a positive reactivity insertion. The subsequent rise of power was limited by the feedback and the insertion of the control rods. The BWR-TT2 benchmark was modeled with the three-dimensional reactor physics code system DYN3D, by coupling neutron kinetics with two-phase thermal-hydraulics. All 6660 DYN3D model parameters were calibrated by applying a predictive modeling methodology that combines experimental and computational information to produce optimally predicted best-estimate results with reduced predicted uncertainties. Simultaneously, the predictive modeling methodology yields optimally predicted values for the BWR total transient power while reducing significantly the accompanying predicted standard deviations.

Predictive uncertainty reduction in coupled neutron-kinetics/thermal hydraulics modeling of the BWR-TT2 benchmark

International Nuclear Information System (INIS)

Badea, Aurelian F.; Cacuci, Dan G.

2017-01-01

Highlights: • BWR Turbine Trip 2 (BWR-TT2) benchmark. • Substantial (up to 50%) reduction of uncertainties in the predicted transient power. • 6660 uncertain model parameters were calibrated. - Abstract: By applying a comprehensive predictive modeling methodology, this work demonstrates a substantial (up to 50%) reduction of uncertainties in the predicted total transient power in the BWR Turbine Trip 2 (BWR-TT2) benchmark while calibrating the numerical simulation of this benchmark, comprising 6090 macroscopic cross sections, and 570 thermal-hydraulics parameters involved in modeling the phase-slip correlation, transient outlet pressure, and total mass flow. The BWR-TT2 benchmark is based on an experiment that was carried out in 1977 in the NPP Peach Bottom 2, involving the closure of the turbine stop valve which caused a pressure wave that propagated with attenuation into the reactor core. The condensation of the steam in the reactor core caused by the pressure increase led to a positive reactivity insertion. The subsequent rise of power was limited by the feedback and the insertion of the control rods. The BWR-TT2 benchmark was modeled with the three-dimensional reactor physics code system DYN3D, by coupling neutron kinetics with two-phase thermal-hydraulics. All 6660 DYN3D model parameters were calibrated by applying a predictive modeling methodology that combines experimental and computational information to produce optimally predicted best-estimate results with reduced predicted uncertainties. Simultaneously, the predictive modeling methodology yields optimally predicted values for the BWR total transient power while reducing significantly the accompanying predicted standard deviations.

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

International Nuclear Information System (INIS)

Crawford, W.D.

1977-01-01

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

Power plant design: ESBWR - the latest passive BWR

International Nuclear Information System (INIS)

Arnold, H.; Yadigaroglu, G.; Stoop, P.C.

1997-01-01

When General Electric said it would end development of its 670 MWe SBWR (Simplified Boiling Water Reactor), it was not quite the end of the story. Also on the drawing board at the time was the larger ESBWR (standing for either European or Economic Simplified BWR) whose goal was to provide the improved economic performance that the SBWR could not. (UK)

Hybrid Reactor Simulation and 3-D Information Display of BWR Out-of-Phase Oscillation

International Nuclear Information System (INIS)

Edwards, Robert; Huang, Zhengyu

2001-01-01

The real-time hybrid reactor simulation (HRS) capability of the Penn State TRIGA reactor has been expanded for boiling water reactor (BWR) out-of-phase behavior. During BWR out-of-phase oscillation half of the core can significantly oscillate out of phase with the other half, while the average power reported by the neutronic instrumentation may show a much lower amplitude for the oscillations. A description of the new HRS is given; three computers are employed to handle all the computations required, including real-time data processing and graph generation. BWR out-of-phase oscillation was successfully simulated. By adjusting the reactivity feedback gains from boiling channels to the TRIGA reactor and to the first harmonic mode power simulation, limit cycle can be generated with both reactor power and the simulated first harmonic power. A 3-D display of spatial power distributions of fundamental mode, first harmonic, and total powers over the reactor cross section is shown

Timing criteria for supplemental BWR emergency response equipment

International Nuclear Information System (INIS)

Bickel, John H.

2015-01-01

The Great Tohuku Earthquake and subsequent Tsunami represented a double failure event which destroyed offsite power connections to Fukushima-Daiichi site and then destroyed on-site electrical systems needed to run decay heat removal systems. The accident could have been mitigated had there been supplemental portable battery chargers, supplemental pumps, and in-place piping connections to provide alternate decay heat removal. In response to this event in the USA, two national response centers, one in Memphis, Tennessee, and another in Phoenix, Arizona, will begin operation. They will be able to dispatch supplemental emergency response equipment to any nuclear plant in the U.S. within 24 hours. In order to define requirements for supplemental nuclear power plant emergency response equipment maintained onsite vs. in a regional support center it is necessary to confirm: (a) the earliest time such equipment might be needed depending on the specific scenario, (b) the nominal time to move the equipment from a storage location either on-site or within the region of a nuclear power plant, and (c) the time required to connect in the supplemental equipment to use it. This paper describes an evaluation process for a BWR-4 with a Mark I Containment starting with: (a) severe accident simulation to define best estimate times available for recovery based on the specific scenario, (b) identify the key supplemental response equipment needed at specific times to accomplish recovery of key safety functions, and (c) evaluate what types of equipment should be warehoused on-site vs. in regional response centers. (authors)

Safety Evaluation Report related to the final design approval of the GESSAR II BWR/6 Nuclear Island design, Docket No. 50-447

International Nuclear Information System (INIS)

1983-04-01

The Safety Evaluation Report for the application filed by General Electric Company for the Final Design Approval for the General Electric Standard Safety Analysis Report (GESSAR II FSAR) has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. This report summarizes the results of the staff's safety review of the GESSAR II BWR/6 Nuclear Island Design. Subject to favorable resolution of items discussed in the Safety Evaluation Report, the staff concludes that the facilities referencing GESSAR II, subject to approval of the balance-of-plant design, can conform with the provisions of the Act and the regulations of the Nuclear Regulatory Commission

LWR-plants. Their evolutionary progress in the last half-century. (4) The start of the nuclear power generation in Japan

International Nuclear Information System (INIS)

Ishikawa, Hiroshi

2008-01-01

Evolutionary progress of the LWR plants in the last half-century was reviewed in series. The start of the nuclear power generation in Japan was reviewed in this article. The Japan Atomic Energy Research Institute (JAERI) promoted nuclear power research and development and introduced the Japan Power Demonstration Reactor (JPDR)-a 12.5 MWe natural circulation BWR, which began operation in 1965. In 1969 its power uprate modifications to a forced circulation BWR (JPDR-II) began and attained operation in 1972. During 50% power test, primary coolant leakage was observed at reactor core spray pipes in 1972. In 1975 the operation resumed and faults observed at condenser tubes in 1976. Primary coolant leakage from in-core flux monitor guide tubes at the bottom of reactor pressure vessel in 1979 led to its permanent shutdown. The nuclear ship Mutsu was put into service in 1970 and during rising power test radiation leakage due to fast neutron streaming was observed in 1974. After modifications of shielding experimental voyage was made in 1991. The first commercial nuclear power reactor, Tokai-1-a 166 MWe gas-cooled (Magnox) reactor, began operation in 1966 and continued until 1998. The LWR plants became the mainstay in Japan. (T. Tanaka)

Behaviour of the reactivity for BWR fuel cells; Comportamiento de la reactividad para celdas de combustible BWR

Energy Technology Data Exchange (ETDEWEB)

Gonzalez, J. A.; Alonso, G.; Delfin, A.; Vargas, S. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Del Valle G, E., E-mail: galonso@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, U. P. Adolfo Lopez Mateos, Col. Lindavista, 07738 Mexico D. F. (Mexico)

2011-11-15

In this work the behaviour of the reactivity of a fuel assembly type BWR was studied, the objective is to obtain some expressions that consider the average enrichment of U-235 and the gadolinium concentration like a function of the fuel cells burnt. Also, the applicability of the lineal reactivity model was analyzed for fuel cells type BWR. The analysis was carried out with the CASMO-4 code. (Author)

Large bundle BWR test CORA-18: Test results

International Nuclear Information System (INIS)

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

1998-04-01

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

Water chemistry experience following an extensive power up-rate in Oskarshamn 3 BWR

International Nuclear Information System (INIS)

Wegemar, Boerje; Nilsson, Jimmy; Lejon Johan; Bergfors, Asa; Arnberg, Bo

2012-09-01

The Swedish Oskarshamn 3 BWR plant, operated by OKG, was first connected to the grid in 1985. The plant has been power up-rated in two steps; from the original design, 3020 MWth, to 3300 MWth (109%, 1989) and recently to 3900 MWth (129%, 2009). Westinghouse Electric Sweden AB (former ASEA-Atom, OEM of the plant) was rewarded a major contract in the recently implemented up-rating project, the PULS project. The PULS project is quite unique since no operating experience has to date been reported from a similar major power up-rate in a BWR plant. Water chemistry experience from the first period of operation following the implementation of the PULS project is reported and discussed in the paper. Reported chemistry and radiochemistry measurements in feedwater (FW) and reactor water (RW) include corrosion products, activated corrosion products, dissolved oxygen and impurities like chloride, sulfate etc. Furthermore, a comparison of water quality prior to implementation of the PULS project is included. Several process systems have been modified, one of them being the condensate cleanup system (CCU), a Pre-coat filter system. The design criteria for the CCU system include the filter run-lengths, pressure drop before back-washing and requirements on water chemistry quality. The paper describes in some detail the CCU system modifications being implemented in order to fulfil the design criterion. CCU cleanup efficiency, operating temperature and influence of hydrogen peroxide on the CCU resin are all important issues being covered in the paper. As for the latter, it is well known that oxygen and hydrogen peroxide (from radiolysis in the core region) might cause partial deterioration of CCU standard cation resin resulting in increased RW sulfate concentrations. This aspect is covered in the paper as well. The reactor water cleanup system (RWCU) in Oskarshamn 3 consists of deep bed ion exchange filters (mixed bed filter). The purpose of RWCU is to maintain a low level of

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)

First results from Mark II at SPEAR

International Nuclear Information System (INIS)

Abrams, G.S.; Alam, M.S.; Blocker, C.A.

1979-05-01

First results from the SLAC-LBL Mark II magnetic detector at SPEAR are presented. The performance of the detector is discussed and preliminary results are given on inclusive baryon production R/sub p + anti p/, R/sub Λ + anti Λ/, on decay modes of the D mesons and on two-photon production of eta' mesons

Burnup credit feasibility for BWR spent fuel shipments

International Nuclear Information System (INIS)

Broadhead, B.L.

1990-01-01

Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This ''burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent of fuel casks used for transportation and storage. Analyses 1 have shown the feasibility estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This paper summarizes the extension of the previous PWR feasibility assessments to boiling water reactor (BWR) fuel. As with the PWR analysis, the purpose was not verification of burnup credit (see ref. 2 for ongoing work in this area) but a reasonable assessment of the feasibility and potential gains from its use in BWR applications. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. The method includes characterization of a typical pin-cell spectrum, using a one-dimensional (1-D) model of a BWR assembly. The calculated spectrum allows burnup-dependent few-group material constants to be generated. Point depletion methods were then used to obtain the time-varying characteristics of the fuel. These simple methods were validated, where practical, with multidimensional methods. 6 refs., 1 tab

Angiotensin II Infusion Induces Marked Diaphragmatic Skeletal Muscle Atrophy

Science.gov (United States)

Rezk, Bashir M.; Yoshida, Tadashi; Semprun-Prieto, Laura; Higashi, Yusuke; Sukhanov, Sergiy; Delafontaine, Patrice

2012-01-01

Advanced congestive heart failure (CHF) and chronic kidney disease (CKD) are characterized by increased angiotensin II (Ang II) levels and are often accompanied by significant skeletal muscle wasting that negatively impacts mortality and morbidity. Both CHF and CKD patients have respiratory muscle dysfunction, however the potential effects of Ang II on respiratory muscles are unknown. We investigated the effects of Ang II on diaphragm muscle in FVB mice. Ang II induced significant diaphragm muscle wasting (18.7±1.6% decrease in weight at one week) and reduction in fiber cross-sectional area. Expression of the E3 ubiquitin ligases atrogin-1 and muscle ring finger-1 (MuRF-1) and of the pro-apoptotic factor BAX was increased after 24 h of Ang II infusion (4.4±0.3 fold, 3.1±0.5 fold and 1.6±0.2 fold, respectively, compared to sham infused control) suggesting increased muscle protein degradation and apoptosis. In Ang II infused animals, there was significant regeneration of injured diaphragm muscles at 7 days as indicated by an increase in the number of myofibers with centralized nuclei and high expression of embryonic myosin heavy chain (E-MyHC, 11.2±3.3 fold increase) and of the satellite cell marker M-cadherin (59.2±22.2% increase). Furthermore, there was an increase in expression of insulin-like growth factor-1 (IGF-1, 1.8±0.3 fold increase) in Ang II infused diaphragm, suggesting the involvement of IGF-1 in diaphragm muscle regeneration. Bone-marrow transplantation experiments indicated that although there was recruitment of bone-marrow derived cells to the injured diaphragm in Ang II infused mice (267.0±74.6% increase), those cells did not express markers of muscle stem cells or regenerating myofibers. In conclusion, Ang II causes marked diaphragm muscle wasting, which may be important for the pathophysiology of respiratory muscle dysfunction and cachexia in conditions such as CHF and CKD. PMID:22276172

Angiotensin II infusion induces marked diaphragmatic skeletal muscle atrophy.

Directory of Open Access Journals (Sweden)

Bashir M Rezk

Full Text Available Advanced congestive heart failure (CHF and chronic kidney disease (CKD are characterized by increased angiotensin II (Ang II levels and are often accompanied by significant skeletal muscle wasting that negatively impacts mortality and morbidity. Both CHF and CKD patients have respiratory muscle dysfunction, however the potential effects of Ang II on respiratory muscles are unknown. We investigated the effects of Ang II on diaphragm muscle in FVB mice. Ang II induced significant diaphragm muscle wasting (18.7±1.6% decrease in weight at one week and reduction in fiber cross-sectional area. Expression of the E3 ubiquitin ligases atrogin-1 and muscle ring finger-1 (MuRF-1 and of the pro-apoptotic factor BAX was increased after 24 h of Ang II infusion (4.4±0.3 fold, 3.1±0.5 fold and 1.6±0.2 fold, respectively, compared to sham infused control suggesting increased muscle protein degradation and apoptosis. In Ang II infused animals, there was significant regeneration of injured diaphragm muscles at 7 days as indicated by an increase in the number of myofibers with centralized nuclei and high expression of embryonic myosin heavy chain (E-MyHC, 11.2±3.3 fold increase and of the satellite cell marker M-cadherin (59.2±22.2% increase. Furthermore, there was an increase in expression of insulin-like growth factor-1 (IGF-1, 1.8±0.3 fold increase in Ang II infused diaphragm, suggesting the involvement of IGF-1 in diaphragm muscle regeneration. Bone-marrow transplantation experiments indicated that although there was recruitment of bone-marrow derived cells to the injured diaphragm in Ang II infused mice (267.0±74.6% increase, those cells did not express markers of muscle stem cells or regenerating myofibers. In conclusion, Ang II causes marked diaphragm muscle wasting, which may be important for the pathophysiology of respiratory muscle dysfunction and cachexia in conditions such as CHF and CKD.

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

International Nuclear Information System (INIS)

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

1993-01-01

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

Turbine protecting device in a BWR type nuclear power plant

International Nuclear Information System (INIS)

Kasuga, Hajime; Oka, Yoko.

1984-01-01

Purpose: To prevent highly humid steams from flowing into the turbine upon abnormal reduction in the reactor water level in order to ensure the turbine soundness, as well as in order to trip the turbine with no undesired effect on the reactor. Constitution: A protection device comprising a judging device and a timer are disposed in a BWR type reactor, in order to control a water level signal from a reactor water level gage. If the reactor water level is reduced during rated power operation, steams are kept to be generated due to decay heat although reactor is scramed. When a signal from the reactor water level detector is inputted to the protection device, a trip signal is outputted by way of a judging device after 15 second by means of the timer, when the main steam check valve is closed to trip the turbine. With this delay of time, abrupt increase in the pressure of the reactor due to sudden shutdown can be prevented. (Nakamoto, H)

Decay ratio studies in BWR and PWR using wavelet

International Nuclear Information System (INIS)

Ciftcioglu, Oe.

1996-10-01

The on-line stability of BWR and PWR is studied using the neutron noise signals as the fluctuations reflect the dynamic characteristics of the reactor. Using appropriate signal modeling for time domain analysis of noise signals, the stability parameters can be directly obtained from the system impulse response. Here in particular for BWR, an important stability parameter is the decay ratio (DR) of the impulse response. The time series analysis involves the autoregressive modeling of the neutron detector signal. The DR determination is strongly effected by the low frequency behaviour since the transfer function characteristic tends to be a third order system rather than a second order system for a BWR. In a PWR low frequency behaviour is modified by the Boron concentration. As a result of these phenomena there are difficulties in the consistent determination of the DR oscillations. The enhancement of the consistency of this DR estimation is obtained by wavelet transform using actual power plant data from BWR and PWR. A comparative study of the Restimation with and without wavelets are presented. (orig.)

BWR vessel and internals project (BWRVIP)

International Nuclear Information System (INIS)

Bilanin, W.J.; Dyle, R.L.

1996-01-01

Recent Boiling Water Reactor (BWR) inspections indicate that Intergranular Stress Corrosion Cracking (IGSCC) is a significant technical issue for some BWR internals. IN response, the Boiling Water Reactor Vessel and Internals Project (BWRVIP) was formed by an associated of domestic and international utilities which own and operate BWRs. The project is identifying or developing generic, cost-effective strategies for managing degradation of reactor internals from which each utility can select the alternative most appropriate for their plant. The Electric Power Research Institute manages the technical program, implementing the utility defined programs. The BWRVIP is organized into four technical tasks: Assessment, Inspection, Repair and Mitigation. An Integration task coordinates the work. The goal of the Assessment task is to develop methodologies for evaluation of vessel and internal components in support of decisions for operation, inspection, mitigation or repair. The goal of the Inspection task is to develop and assess effective and predictable inspection techniques which can be used to determine the condition of BWR vessel and internals that are potentially susceptible to service-related SCC degradation. The goal of the Repair task is to assure the availability of cost-effective repair/replacement alternatives. The goal of the Mitigation task is to develop and demonstrate countermeasures for SCC degradation. This paper summarizes the BWRVIP approach for addressing BWR internals SCC degradation and illustrates how utilities are utilizing BWRVIP products to successfully manage the effect of SCC on core shrouds

Characteristics of fluctuating pressure generated in BWR main steam lines

International Nuclear Information System (INIS)

Takahashi, Shiro; Okuyama, Keita; Tamura, Akinori

2009-01-01

The BWR-3 steam dryer in the Quad Cities Unit 2 Nuclear Power Plant was damaged by high cycle fatigue due to acoustic-induced vibration. The dryer failure was as attributed to flow-induced acoustic resonance at the stub pipes of safety relief valves (SRVs) in the main steam lines (MSLs). The acoustic resonance was considered to be generated by interaction between the sound field and an unstable shear layer across the closed side branches with SRV stub pipes. We have started a research program on BWR dryers to develop their loading evaluation methods. Moreover, it has been necessary to evaluate the dryer integrity of BWR-5 plants which are the main type of BWR in Japan. In the present study, we used 1/10-scale BWR tests and analyses to investigate the flow-induced acoustic resonance and acoustic characteristics in MSLs. The test apparatus consisted of a steam dryer, a steam dome and 4 MSLs with 20 SRV stub pipes. A finite element method (FEM) was applied for the calculation of three-dimensional wave equations in acoustic analysis. We demonstrated that remarkable fluctuating pressures occurred in high and low frequency regions. High frequency fluctuating pressures was generated by the flow-induced acoustic resonance in the SRV stub pipes. Low frequency fluctuating pressure was generated in an MSL with the dead leg. The frequency of the latter almost coincided with the natural frequency of the MSL with the dead leg. The amplitude of the fluctuating pressures in the multiple stub pipes became more intense because of interaction between them compared with that in the single stub pipe. Acoustic analysis results showed that the multiple stub pipes caused several natural frequencies in the vicinity of the natural frequency of the single stub pipe and several modes of the standing wave in the MSLs. (author)

Transient behavior during reactivity insertion in the Moroccan TRIGA Mark II reactor using the PARET/ANL code

International Nuclear Information System (INIS)

Boulaich, Y.; Nacir, B.; El Bardouni, T.; Boukhal, H.; Chakir, E.; El Bakkari, B.; El Younoussi, C.

2015-01-01

Highlights: • PARET model for the Moroccan TRIGA MARK II reactor has been developed. • Transient behavior under reactivity insertion has been studied based on PARET code. • Power factors required by PARET code have been calculated by using MCNP5 code. • The dependence on time of the main thermal-hydraulic parameters was calculated. • Results are largely far to compromise the thermal design limits. - Abstract: A three dimensional model for the Moroccan 2 MW TRIGA MARK II reactor has been developed for thermal-hydraulic and safety analysis by using the PARET/ANL and MCNP5 codes. This reactor is located at the nuclear studies center of Mâamora (CENM), Morocco. The model has been validated through temperature measurements inside two instrumented fuel elements located near the center of the core, at various power levels, and also through the power and fuel temperature evolution after the reactor shutdown (SCRAM). The axial distributions of power factors required by the PARET code have been calculated in each fuel element rod by using MCNP5 code. Based on this thermal-hydraulic model, a safety analysis under the reactivity insertion phenomenon has been carried out and the dependence on time of the main thermal-hydraulic parameters was calculated. Results were compared to the thermal design limits imposed to maintain the integrity of the clad

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

International Nuclear Information System (INIS)

Yamamoto, Yasushi; Morooka, Shinichi

1997-01-01

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

Control rod pattern exchange in a BWR/6 utilizing gang mode withdrawal

International Nuclear Information System (INIS)

Auvil, A.B. Jr.; Aldemir, T.; Hajek, B.K.

1986-01-01

The use of checkerboard pattern of alternating inserted and fully withdrawn control rods and the uneven void distribution in boiling water reactor (BWR) cores can cause large burnup gradients even after a short time of operation. To compensate for these effects, power has to be reshaped periodically (typically every two full-power months) by individually manipulating the control rods. During this manipulation process (called the control rod pattern exchange), the core power is reduced to 60% of nominal power by means of flow reduction to limit power swings to tolerable levels and to ensure that fuel thermal limits are not exceeded. A control rod pattern exchange by individual rod manipulation typically takes 4 to 8 h and represents a large cost burden to the utility in terms of reduced system output. The latest generation of BWRs, the BWR/6, possesses the capability to simultaneously move up to four symmetrically located control rods. The rods corresponding to a given gang may have rotational symmetry, mirror symmetry, or a combination of the two. This paper presents a pattern exchange procedure that exploits the capability of gang mode rod withdrawal to reduce the pattern exchange execution time and radial power distribution asymmetry associated with individual rod manipulation. The working model used in the study is the Perry Nuclear Power Plant Unit 1, located in Perry, Ohio, and owned by the Cleveland Electric Illuminating Company

PREDICTIVE METHODS FOR STABILITY MARGIN IN BWR

OpenAIRE

MELARA SAN ROMÁN, JOSÉ

2016-01-01

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

Fuel rod failure due to marked diametral expansion and fuel rod collapse occurred in the HBWR power ramp experiment

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki

1985-12-01

In the power ramp experiment with the BWR type light water loop at the HBWR, the two pre-irradiated fuel rods caused an unexpected pellet-cladding interaction (PCI). One occurred in the fuel rod with small gap of 0.10 mm, which was pre-irradiated up to the burn-up of 14 MWd/kgU. At high power, the diameter of the rod was increased markedly without accompanying significant axial elongation. The other occurred in the rod with a large gap of 0.23 mm, which was pre-irradiated up to the burn-up of 8 MWd/kgU. The diameter of the rod collapsed during a diameter measurement at the maximum power level. The causes of those were investigated in the present study by evaluating in-core data obtained from equipped instruments in the experiment. It was revealed from the investigation that these behaviours were attributed to the local reduction of the coolant flow occurred in the region of a transformer in the ramp rig. The fuel cladding material is seemed to become softened due to temperature increase caused by the local reduction of the coolant flow, and collapsed by the coolant pressure, either locally or wholly depending on the rod diametral gap existed. (author)

Results on two-photon interactions from Mark II at SPEAR

International Nuclear Information System (INIS)

Abrams, G.S.; Alam, M.S.; Blocker, C.A.

1979-10-01

Preliminary results on two-photon interactions from the SLAC-LBL Mark II magnetic detector at SPEAR are presented. The cross section for eta' production by the reaction e + e - → e + e - eta' has been measured over the beam energy range from 2 to 4 GeV. The radiative width GAMMA/sub γγ/(eta') has been determined to be 5.8 +- 1.1 keV (+- 20% systematic uncertainty). Upper limits on the radiative widths of the f(1270), and A 2 (1310) and f'(1515) mesons have been determined

Consideration of severe accident issues for the General Electric BWR standard plant: Chapter 10

International Nuclear Information System (INIS)

Holtzclaw, K.W.

1983-01-01

In early 1982, the U.S. Nuclear Regulatory Commission (NRC) proposed a policy to address severe accident rulemaking on future plants by utilizing standard plant licensing documentation. GE provided appendices to the licensing documentation of its standard plant design, GESSAR II, which address severe accidents for the GE BWR/6 Mark III 238 nuclear island design. The GE submittals discuss the features of the design that prevent severe accidents from leading to core damage or that mitigate the effects of severe accidents should core damage occur. The quantification of the accident prevention and mitigation features, including those incorporated in the design since the accident at Three Mile Island (TMI), is provided by means of a comprehensive probabilistic risk assessment, which provides an analysis of the probability and consequences of postulated severe accidents

Parametric study of the potential for BWR ECCS strainer blockage due to LOCA generated debris. Final report

International Nuclear Information System (INIS)

Zigler, G.; Brideau, J.; Rao, D.V.; Shaffer, C.; Souto, F.; Thomas, W.

1995-10-01

This report documents a plant-specific study for a BWR/4 with a Mark I containment that evaluated the potential for LOCA generated debris and the probability of losing long term recirculation capability due ECCS pump suction strainer blockage. The major elements of this study were: (1) acquisition of detailed piping layouts and installed insulation details for a reference BWR; (2) analysis of plant specific piping weld failure probabilities to estimate the LOCA frequency; (3) development of an insulation and other debris generation and drywell transport models for the reference BWR; (4) modeling of debris transport in the suppression pool; (5) development of strainer blockage head loss models for estimating loss of NPSH margin; (6) estimation of core damage frequency attributable to loss of ECCS recirculation capability following a LOCA. Elements 2 through 5 were combined into a computer code, BLOCKAGE 2.3. A point estimate of overall DEGB pipe break frequency (per Rx-year) of 1.59E-04 was calculated for the reference plant, with a corresponding overall ECCS loss of NPSH frequency (per Rx-year) of 1.58E-04. The calculated point estimate of core damage frequency (per Rx-year) due to blockage related accident sequences for the reference BWR ranged from 4.2E-06 to 2.5E-05. The results of this study show that unacceptable strainer blockage and loss of NPSH margin can occur within the first few minutes after ECCS pumps achieve maximum flows when the ECCS strainers are exposed to LOCA generated fibrous debris in the presence of particulates (sludge, paint chips, concrete dust). Generic or unconditional extrapolation of these reference plant calculated results should not be undertaken

BWR Water Chemistry Guidelines: 1993 Revision, Normal and hydrogen water chemistry

International Nuclear Information System (INIS)

Karlberg, G.; Goddard, C.; Fitzpatrick, S.

1994-02-01

The goal of water chemistry control is to extend the operating life of the reactor and rector coolant system, balance-of-plant components, and turbines while simultaneously controlling costs to safeguard the continued economic viability of the nuclear power generation investment. To further this goal an industry committee of chemistry personnel prepared guidelines to identify the benefits, risks, and costs associated with water chemistry in BWRs and to provide a template for an optimized water chemistry program. This document replaces the BWR Normal Water Chemistry Guidelines - 1986 Revision and the BWR Hydrogen Water Chemistry Guidelines -- 1987 Revision. It expands on the previous guidelines documents by covering the economic implications of BWR water chemistry control

Flux and power distributions in BWR multi-bundle fuel arrays

International Nuclear Information System (INIS)

Cheng, H.S.

1976-02-01

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

Calibration and performance of the MARK II drift chamber vertex detector

International Nuclear Information System (INIS)

Durrett, D.; Ford, W.T.; Hinshaw, D.A.; Rankin, P.; Smith, J.G.; Weber, P.

1990-05-01

We have calibrated and studied the performance of the MARK II drift chamber vertex detector with cosmic ray tracks collected with the chamber inside the MARK II detector at the SLC. The chamber achieves 30 μm impact parameter resolution and 500 μm track-pair resolution using CO 2 /C 2 H 6 H 6 (92/8) at 2 atmospheres pressure. The chamber has successfully recorded Z 0 decays at the SLC, and resolved tracks in dense hadronic jets with good efficiency and high accuracy. 5 refs., 13 figs

Simulation of TRIGA Mark II Benchmark Experiment using WIMSD4 and CITATION codes

International Nuclear Information System (INIS)

Dalle, Hugo Moura; Pereira, Claubia

2000-01-01

This paper presents a simulation of the TRIGA Mark II Benchmark Experiment, Part I: Steady-State Operation and is part of the calculation methodology validation developed to the neutronic calculation of the CDTN's TRIGA IPR - R1 reactor. A version of the WIMSD4, obtained in the Centro de Tecnologia Nuclear, in Cuba, was used in the cells calculation. In the core calculations was adopted the diffusion code CITATION. Was adopted a 3D representation of the core and the calculations were carried out at two energy groups. Many of the experiments were simulated, including, K eff , control rods reactivity worth, fuel elements reactivity worth distribution and the fuel temperature reactivity coefficient. The comparison of the obtained results, with the experimental results, shows differences in the range of the accuracy of the measurements, to the control rods worth and fuel temperature reactivity coefficient, or on an acceptable range, following the literature, to the K eff and fuel elements reactivity worth distribution and the fuel temperature reactivity coefficient. The comparison of the obtained results, with the experimental. results, shows differences in the range of the accuracy of the measurements, to the control rods worth and fuel temperature reactivity coefficient, or in an acceptable range, following the literature, to the K eff and fuel elements reactivity worth distribution. (author)

Selected results from the Mark II at SPEAR

International Nuclear Information System (INIS)

Scharre, D.L.

1980-06-01

Recent results on radiative transitions from the psi(3095), charmed meson decay, and the Cabibbo-suppressed decay tau → K* ν/sub tau/ are reviewed. The results come primarily from the Mark II experiment at SPEAR, but preliminary results from the Crystal Ball experiment on psi radiative transitions are also discussed

Mark III LOCA-related hydrodynamic load definition. Generic technical activity B-10

International Nuclear Information System (INIS)

1984-02-01

This report, prepared by the staff of the Office of Nuclear Reactor Regulation and its consultants at the Brookhaven National Laboratory, provides a discussion of LOCA-related suppression pool hydrodynamic loads in boiling water reactor (BWR) facilities with the Mark III pressure-suppression containment design. Its issuance completes NRC Generic Technical Activity B-10, Behavior of BWR Mark III Containment. On the basis of certain large-scale tests conducted between 1973 and 1979, the General Electric Company developed LOCA-related hydrodynamic load definitions for use in the design of the standard Mark III containment. The staff and its consultants have reviewed these load definitions and their bases conclude that, with a few specified changes, the proposed load definitions provide conservative loading conditions. The staff-approved acceptance criteria for LOCA-related hydrodynamic loads are provided in Appendix C of this report

BWR shutdown analyzer using artificial intelligence (AI) techniques

International Nuclear Information System (INIS)

Cain, D.G.

1986-01-01

A prototype alarm system for detecting abnormal reactor shutdowns based on artificial intelligence technology is described. The system incorporates knowledge about Boiling Water Reactor (BWR) plant design and component behavior, as well as knowledge required to distinguish normal, abnormal, and ATWS accident conditions. The system was developed using a software tool environment for creating knowledge-based applications on a LISP machine. To facilitate prototype implementation and evaluation, a casual simulation of BWR shutdown sequences was developed and interfaced with the alarm system. An intelligent graphics interface for execution and control is described. System performance considerations and general observations relating to artificial intelligence application to nuclear power plant problems are provided

Calculation analysis of TRIGA MARK II reactor core composed of two types of fuel elements

International Nuclear Information System (INIS)

Ravnik, M.

1988-11-01

The most important properties of mixed cores are treated for TRIGA MARK II reactor, composed of standard (20% enriched, 8.5w% U content) and FLIP (70% enriched, 8.5w% U content) fuel elements. Large difference in enrichment and presence of burnable poison in FLIP fuel have strong influence on the main core characteristics, such as: fuel temperature coefficient, power defect, Xe and Sm worth, power and flux distributions, etc. They are significantly different for both types of fuel. Optimal loading of mixed cores therefore strongly depends on the loading pattern of both types of fuel elements. Results of systematic calculational analysis of mixed cores are presented. Calculations on the level of fuel element are performed with WIMSD-4 computer code with extended cross-section library. Core calculations are performed with TRIGAP two-group 1-D diffusion code. Results are compared to measurements and physical explanation is provided. Special concern is devoted to realistic mixed cores, for which optimal in-core fuel management is derived. Refs, figs and tabs

Mark-II Data Acquisition and Trigger system

International Nuclear Information System (INIS)

Breidenbach, M.

1984-06-01

The Mark-II Data Acquisition and Trigger system requirements and general solution are described. The solution takes advantage of the synchronous crossing times and low event rates of an electron positron collider to permit a very highly multiplexed analog scheme to be effective. The system depends on a two level trigger to operate with acceptable dead time. The trigger, multiplexing, data reduction, calibration, and CAMAC systems are described

Fuel burnup analysis of the TRIGA Mark II reactor at the University of Pavia

International Nuclear Information System (INIS)

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

2016-01-01

Highlights: • A fuel evolution model for a TRIGA Mark II reactor has been developed. • Reproduction of nearly 50 years of reactor operation. • The model was used to predict the best reactor reconfiguration. • Reactor life was extended without adding fresh fuel elements. - Abstract: A time evolution model was developed to study fuel burnup for the TRIGA Mark II reactor at the University of Pavia. The results were used to predict the effects of a complete core reconfiguration and the accuracy of this prediction was tested experimentally. We used the Monte Carlo code MCNP5 to reproduce system neutronics in different operating conditions and to analyze neutron fluxes in the reactor core. The software that took care of time evolution, completely designed in-house, used the neutron fluxes obtained by MCNP5 to evaluate fuel consumption. This software was developed specifically to keep into account some features that differentiate low power experimental reactors from those used for power production, such as the daily ON/OFF cycle and the long fuel lifetime. These effects can not be neglected to properly account for neutron poison accumulation. We evaluated the effect of 48 years of reactor operation and predicted a possible new configuration for the reactor core: the objective was to remove some of the fuel elements from the core and to obtain a substantial increase in the Core Excess reactivity value. The evaluation of fuel burnup and the reconfiguration results are presented in this paper.

Seismic PRA of a BWR plant

International Nuclear Information System (INIS)

Nishio, Masahide; Fujimoto, Haruo

2014-01-01

Since the occurrence of nuclear power plant accidents in the Fukushima Daichi nuclear power station, the regulatory framework on severe accident (SA) has been discussed in Japan. The basic concept is to typify and identify the accident sequences leading to core/primary containment vessel (PCV) damage and to implement SA measures covering internal and external events extensively. As Japan is an earthquake-prone country and earthquakes and tsunami are important natural external events for nuclear safety of nuclear power plants, JNES performed the seismic probabilistic risk assessment (PRA) on a typical nuclear power plant and evaluated the dominant accident sequences leading to core/PCV damage to discuss dominant scenarios of severe accident (SA). The analytical models and the results of level-1 seismic PRA on a 1,100 MWe BWR-5 plant are shown here. Seismic PRA was performed for a typical BWR5 plant. Initiating events with large contribution to core damage frequency are the loss of all AC powers (station blackout) and the large LOCA. The top of dominant accident sequences is the simultaneous occurrence of station blackout and large LOCA. Important components to core damage frequency are electric power supply equipment. It needs to keep in mind that the results are influenced on site geologic characteristic to a greater or lesser. In the process of analysis, issues such as conservative assumptions related to damages of building or structure and success criteria for excessive LOCA are left to be resolved. These issues will be further studied including thermal hydric analysis in the future. (authors)

The BWR owners' group planning guide for life extension

International Nuclear Information System (INIS)

Smith, S.K.; Lehnert, D.F.; Locke, R.K.

1991-01-01

Extending the operating life of a commercial nuclear power plant has been shown to be economically beneficial to both the utility and the electric customer. As such, many utilities are planning and implementing plant life extension (PLEX) programs. A document has been developed which provides guidance to utilities in formulating a PLEX program plant for one or more boiling water reactor (BWR) plants. The guide has been developed by the BWR Owners' Group Plant Life Extension Committee. The principal bases for this guide were the BWR Pilot and Lead Plant Programs. These programs were used as models to develop the 'base plan' described in this guide. By formulating their program plant utilizing the base plan, utilities will be able to maximize the use of existing evaluations and results. The utility planner will build upon the base plan by adding any tasks or features that are unique to their programs. (author)

Global vibrations in the wetwell condensation process caused by LOCA in BWR plants

International Nuclear Information System (INIS)

Bjoerndahl, O.; Andersson, Magnus

1998-12-01

During the last years a substantial part of third part review work related to dynamical loadings has been review of loading specifications dealing with vibrations in containment building related to so called LOCA-events in Swedish BWR plants. Compared to other loading categories characterised as global vibrations these secondary effects of LOCA-events are complex to analyse. One experience from the review work at SAQ up to now is that it is not fully clear what prediction methods and what model idealisations are the most adequate for structural integrity verification on mechanical systems as pressure vessels and piping under such loading conditions. At SAQ Teknik a project work has been carried out to investigate the general status of the methodology used today in Sweden and a work to in the long term develop simplified prediction models and methods for the loading categories condensation oscillations (CO) and chugging (CH). The work was initially concentrated on a study of the background of the methodology which was developed for these type of loading in American BWR-containments of the Mark-II design. The methodology was developed by General Electric, GE, in cooperation with the Mark-II plant owners. The methodology used in Sweden to predict vibrations in BWR containments of this design is with some minor modifications very close to technique developed by GE. The methodology developed by GE is the only accepted by USNRC for the Mark-II design and could be found as reference in Standard Review Plan 6.2.1.1.C, Rev 6 - August 1984. Based on identical physical assumptions about the dynamic behaviour of the building structure and the water in the suppression pool, mathematical models are derived in this report for predictions of secondary structure response spectra for loading conditions as global vibrations during CO and CH. Based on parameters identified by so called one pipe experiments responses my be predicted. By use of these derived mathematical models as a

Review of international solutions to NEACRP benchmark BWR lattice cell problems

International Nuclear Information System (INIS)

Halsall, M.J.

1977-12-01

This paper summarises international solutions to a set of BWR benchmark problems. The problems, posed as an activity sponsored by the Nuclear Energy Agency Committee on Reactor Physics, were as follows: 9-pin supercell with central burnable poison pin, mini-BWR with 4 pin-cells and water gaps and control rod cruciform, full 7 x 7 pin BWR lattice cell with differential U 235 enrichment, and full 8 x 8 pin BWR lattice cell with water-hole, Pu-loading, burnable poison, and homogenised cruciform control rod. Solutions have been contributed by Denmark, Japan, Sweden, Switzerland and the UK. (author)

The use of the partial coherence function technique for the investigation of BWR noise dynamics

International Nuclear Information System (INIS)

Kostic, Lj.

1983-01-01

The extensive experimental investigations, at the last time, indicate that the partial coherence function technique can be a powerful method of the investigation of BWR noise dynamics. Symple BWR noise dynamics model for the global noise study, based on different noise phenomena, is proposed in this paper. (author)

Comparison of the CORA-12, 13, 17 experiments and B{sub 4} effect on the flooding behavior of BWR bundles; Vergleich der Flutexperimente CORA-12, 13, 17 und der Einfluss des B{sub 4}C auf das Flutverhalten von SWR-Buendeln

Energy Technology Data Exchange (ETDEWEB)

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

1995-08-01

The CORA quench experiments 12, 13 (PWR) and 17 (BWR) are in agreement with LOFT 2 and TMI: Flooding of hot Zircaloy clad fuel rods does not result in an immediate cooldown of the bundle, but produces remarkable temporary temperature increase, connected to a strong peak in hydrogen production. The PWR tests CORA 12 and CORA 13 are of the same geometrical arrangement and test conduct, with the exception of the shorter time between power shutdown and quench initiation for CORA 13. A higher temperature of the bundle at start of quenching was the consequence. BWR test CORA 17 - with B{sub 4}C absorber and additional Zircaloy channel box walls - was in respect to the delay-time between power shutdown and start of quenching similar to test CORA 12. All tests showed during the quench phase the temporary temperature increase, correlated to a hydrogen peak. The CORA 17 test resulted immediately after quenching in a modest increase for 20 s and changed then in a steep increase, resulting in the highest temperature and hydrogen peaks of the three tests. CORA 17 also showed a temperature increase in the lower part of the bundle, in contrast to CORA 12 and CORA 13 with temperature increase only in the upper half of the bundle. We interpret this earlier starting and stronger reaction due to the influence of the boron carbide, the absorber material of the BWR test. B{sub 4}C has an exothermic reaction rate 4 to 9 times larger than Zry and produces 5 to 6,6 times more hydrogen. Probably the hot remained columns of B{sub 4}C (seen in the non-quench test CORA 16) react early in the quench process with the increased upcoming steam. The bundle temperature raised by this reaction increases the reaction rate (exponential dependency) of the remaining metallic Zry. Due to the larger amount of Zry in the BWR bundle (channel box walls) and the smaller steam input during the heatup phase (2 g/s instead of 6 g/s) more metallic Zry can have survived oxidation during the heatup phase. (orig./HP)

Standard Technical Specifications General Electric plants, BWR/4: Bases (Sections 2.0-3.3). Volume 2, Revision 1

International Nuclear Information System (INIS)

1995-04-01

This report documents the results of the combined effort of the NRC and the industry produce improved Standard Technical Specifications (STS), Revision 1 for General Electric BWR/4 Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved ST or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume I contains the Specifications for all chapters and sections of the improved STS. This document, Volume 2, contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1-3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4-3.10 of the improved STS

The status of the SLAC Linear Collider and of the Mark II detector

International Nuclear Information System (INIS)

Lankford, A.J.

1987-10-01

At SLAC we are currently involved in the exciting challenge of commissioning the first example of a new type of colliding beam accelerator, the SLAC Linear Collider, or SLC. The goals of the SLC are two-fold. It will explore the concept of linear colliders, and it will allow the study of physics on the Z 0 resonance. It accomplishes these goals by exploiting the existing SLAC linac and the large visible cross-section of approximately thirty nanobarns of the Z 0 . The MARK II detector will have the opportunity to be first to explore the physics in this regime. This paper briefly reports the status of the SLC and of the MARK II as of early October 1987, at which time commissioning efforts were interrupted in order to place the MARK II detector at the collision point and to incorporate some improvements to the SLC. The first portion of this report highlights some of the milestones achieved in the SLC commissioning and some of the problems encountered. The last portion outlines improvements made to the MARK II for physics at the SLC. 10 refs., 12 figs., 1 tab

A calorimeter software trigger for the Mark II detector at SLC [Stanford Linear Collider

International Nuclear Information System (INIS)

Briggs, D.; Glanzman, T.; Grosse-Wiesmann, P.; Tinsman, J.; Holmgren, S.; Schaad, M.W.

1989-04-01

A new FASTBUS-based calorimeter software trigger for the upgraded Mark II at the Stanford Linear Collider (SLC) is presented. The trigger requirements for SLC and a short description of the hardware used for this purpose are given, followed by a detailed description of the software. Some preliminary results are presented. 9 refs., 4 figs

User’s guide for MapMark4GUI—A graphical user interface for the MapMark4 R package

Science.gov (United States)

Shapiro, Jason

2018-05-29

MapMark4GUI is an R graphical user interface (GUI) developed by the U.S. Geological Survey to support user implementation of the MapMark4 R statistical software package. MapMark4 was developed by the U.S. Geological Survey to implement probability calculations for simulating undiscovered mineral resources in quantitative mineral resource assessments. The GUI provides an easy-to-use tool to input data, run simulations, and format output results for the MapMark4 package. The GUI is written and accessed in the R statistical programming language. This user’s guide includes instructions on installing and running MapMark4GUI and descriptions of the statistical output processes, output files, and test data files.

Complete BWR--EM LOCA analysis using the WRAP--EM system

International Nuclear Information System (INIS)

Beckmeyer, R.R.; Gregory, M.V.; Buckner, M.R.

1979-01-01

The Water Reactor Analysis Package, Evaluation Model (WRAP--EM), provides a complete analysis of postulated loss-of-coolant accidents (LOCA's) in light--water nuclear power reactors. The system is being developed at the Savannah River Laboratory (SRL) for use by the Nuclear Regulatory Commission (NRC) to interpret and evaluate reactor vendor, evaluation model (EM) analyses. The initial version of the WRAP--EM system for analysis of boiling water reactors (BWR's) is operational. To demonstrate the complete capability of the WRAP--BWR--EM system, a LOCA analysis has been performed for the Hope Creek Plant

Analysis of BWR out-of-phase instabilities in the frequency domain

International Nuclear Information System (INIS)

Farawila, Y.M.; Pruitt, D.W.; Kreuter, D.

1992-01-01

During startup or because of an inadvertent recirculation pump trip, a boiling water reactor (BWR) may operate at relatively low flow and high power conditions. At these conditions, a BWR is susceptible to coupled flow and power oscillations that could result in undesirable reactor scram unless appropriate countermeasures are taken. This contribution to analytical methods has been developed to address in part a general industrywide and regulatory concern about BWR stability initiated by the LaSalle 2 instability event in March 1988. This work is designed to extend the capability of the one-dimensional parallel channel frequency domain code STAIF to predict the regional oscillation decay ratio. The basic theory follows that developed by March-Leuba and Blakeman, where the oscillation mechanism is identified as the excitation of a subcritical neutronic mode with a constant core pressure drop boundary condition. The improvements to the basic theory include applying the theory to one-dimensional neutronics instead of point kinetics and taking account of the actual three-dimensional harmonic flux distribution

Comparison of the General Electric BWR/6 standard plant design to the IAEA NUSS codes and guides

International Nuclear Information System (INIS)

D'Ardenne, W.H.; Sherwood, G.G.

1985-01-01

The General Electric BWR/6 Mark III standard plant design meets or exceeds current requirements of published International Atomic Energy Agency (IAEA) Nuclear Safety Standards (NUSS) codes and guides. This conclusion is based on a review of the NUSS codes and guides by General Electric and by the co-ordinated US review of the NUSS codes and guides during their development. General Electric compared the published IAEA NUSS codes and guides with the General Electric design. The applicability of each code and guide to the BWR/6 Mark III standard plant design was determined. Each code or guide was reviewed by a General Electric engineer knowledgeable about the structures, systems and components addressed and the technical area covered by that code or guide. The results of this review show that the BWR/6 Mark III standard plant design meets or exceeds the applicable requirements of the published IAEA NUSS codes and guides. The co-ordinated US review of the IAEA NUSS codes and guides corroborates the General Electric review. In the co-ordinated US review, the USNRC and US industry organizations (including General Electric) review the NUSS codes and guides during their development. This review ensures that the NUSS codes and guides are consistent with the current US government regulations, guidance and regulatory practices, US voluntary industry codes and standards, and accepted US industry design, construction and operational practices. If any inconsistencies are identified, comments are submitted to the IAEA by the USNRC. All US concerns submitted to the IAEA have been resolved. General Electric design reviews and the Final Design Approval (FDA) issued by the USNRC have verified that the General Electric BWR/6 Mark III standard plant design meets or exceeds the current US requirements, guidance and practices. Since these requirements, guidance and practices meet or exceed those of the NUSS codes and guides, so does the General Electric design. (author)

Operational experience of human-friendly control and instrumentation systems for BWR nuclear power plants

International Nuclear Information System (INIS)

Makino, M.; Watanabe, T.; Suto, O.; Asahi, R.

1987-01-01

In recent BWR nuclear power plants in Japan, an advanced centralized monitoring and control system PODIA (Plant Operation by Displayed Information and Automation), which incorporates many operator aid functions, has been in operation since 1985. Main functions of the PODIA system as a computerized operator aid system are as follows. CRT displays for plant monitoring. Automatic controls and operation guides for plant operation. Stand-by status monitoring for engineered safety features during normal operation. Surveillance test procedure guides for engineered safety features. Integrated alarm display. The effectiveness of these functions have been proved through test and commercial operation. It has been obtained that operators have preferred PODIA much more than conventional monitoring and control systems

BWR [boiling water reactor] core criticality versus water level during an ATWS [anticipated transient without scram] event

International Nuclear Information System (INIS)

Sehgal, B.R.; Peng, C.M.; Maly, J.

1988-01-01

The BWR [boiling water reactor] emergency procedures guidelines recommend management of core water level to reduce the power generated during an anticipated transient without scram (ATWS) event. BWR power level variation has traditionally been calculated in the system codes using a 1-D [one-dimensional] 2-group neutron kinetics model to determine criticality. This methodology used also for calculating criticality of the partially covered BWR cores has, however, never been validated against data. In this paper, the power level versus water level issues in an ATWS severe accident are introduced and the accuracy of the traditional methodology is investigated by comparing with measured data. It is found that the 1-D 2-group treatment is not adequate for accurate predictions of criticality and therefore the system power level for the water level variations that may be encountered in a prototypical ATWS severe accident. It is believed that the current predictions for power level may be too high

Advanced methods for BWR transient and stability analysis

Energy Technology Data Exchange (ETDEWEB)

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

2008-07-01

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

Advanced methods for BWR transient and stability analysis

International Nuclear Information System (INIS)

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

2008-01-01

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

Advanced BWR core component designs and the implications for SFD analysis

International Nuclear Information System (INIS)

Ott, L.J.

1997-01-01

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

MARK II end cap calorimeter electronics

International Nuclear Information System (INIS)

Jared, R.C.; Haggerty, J.S.; Herrup, D.A.; Kirsten, F.A.; Lee, K.L.; Olson, S.R.; Wood, D.R.

1985-10-01

An end cap calorimeter system has been added to the MARK II detector in preparation for its use at the SLAC Linear Collider. The calorimeter uses 8744 rectangular proportional counter tubes. This paper describes the design features of the data acquisition electronics that has been installed on the calorimeter. The design and use of computer-based test stands for the amplification and signal-shaping components is also covered. A portion of the complete system has been tested in a beam at SLAC. In these initial tests, using only the calibration provided by the test stands, a resolution of 18%/√E was achieved

Effects of a hypothetical loss-of-coolant accident on a Mark I Boiling Water Reactor pressure-suppression system

International Nuclear Information System (INIS)

Pitts, J.H.; McCauley, E.W.

1977-01-01

A loss-of-coolant accident (LOCA) in a boiling-water-reactor (BWR) power plant has never occurred. However, because this type of accident could be particularly severe, it is used as a principal theoretical basis for design. A series of consistent, versatile, and accurate air-water tests that simulate LOCA conditions has been completed on a 1 / 5 -scale Mark I BWR pressure-suppression system. Results from these tests are used to quantify the vertical-loading function and to study the associated fluid dynamics phenomena. Detailed histories of vertical loads on the wetwell are shown. In particular, variation of hydrodynamic-generated vertical loads with changes in drywell-pressurization rate, downcomer submergence, and the vent-line loss coefficient are established. Initial drywell overpressure, which partially preclears the downcomers of water, substantially reduces the peak vertical loads. Scaling relationships, developed from dimensional analysis and verified by bench-top experiments, allow the 1 / 5 -scale results to be applied to a full-scale BWR power plant. This analysis leads to dimensionless groupings that are invariant. These groupings show that, if water is used as the working fluid, the magnitude of the forces in a scaled facility is reduced by the cube of the scale factor and occurs in a time reduced by the square root of the scale factor

Development and Evaluation of cooperative control system for an HVDC transmission system connected with an isolated BWR power plant

International Nuclear Information System (INIS)

Horiuchi, Susumu; Hara, Tsukusi; Matori, Iwao; Hirayama, Kaiichirou.

1987-01-01

In the cooperative control system developed for an HVDC transmission system connected with an isolated BWR power plant, the equilibrium state between power plant output and DC transmission power is examined by way of the detection of the generator frequency. And, thereby start-up and shutdown of the DC system and controlling of the transmission power are made, so that the signal transmission with the power plant becomes unnecessary, enabling the easy cooperative operation. In order to investigate validity of this control system, various digital simulation and simulator test with the control system were carried out. In this way, behavior of the power plant and stability of the DC transmission system were evaluated in the connection to the DC system at power plant start-up, follow of the transmission power in change of the power plant output and in various system failures. (Mori, K.)

The probability of Mark-I containment failure by melt-attack of the liner

International Nuclear Information System (INIS)

Theofanous, T.G.; Yan, H.; Podowski, M.Z.

1993-11-01

This report is a followup to the work presented in NUREG/CR-5423 addressing early failure of a BWR Mark I containment by melt attack of the liner, and it constitutes a part of the implementation of the Risk-Oriented Accident Analysis Methodology (ROAAM) employed therein. In particular, it expands the quantification to include four independent evaluations carried out at Rensselaer Polytechnic Institute, Argonne National Laboratories, Sandia National Laboratories and ANATECH, Inc. on the various portions of the phenomenology involved. These independent evaluations are included here as Parts II through V. The results, and their integration in Part I, demonstrate the substantial synergism and convergence necessary to recognize that the issue has been resolved

Trade reforms, mark-ups and bargaining power of workers: the case ...

African Journals Online (AJOL)

Ethiopian Journal of Economics ... workers between 1996 and 2007, a model of mark-up with labor bargaining power was estimated using random effects and LDPDM. ... Keywords: Trade reform, mark-up, bargaining power, rent, trade unions ...

In-core failure of the instrumented BWR rod by locally induced high coolant temperature

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki

1985-12-01

In the BWR type light water loop instrumented in HBWR, a current BWR type fuel rod pre-irradiated up to 5.6 MWd/kgU was power ramped to 50 kW/m. During the ramp, the diameter of the rod was expanded significantly at the bottom end. The behaviour was different from which caused by pellet-cladding interaction (PCI). In the post-irradiation examination, the rod was found to be failed. In this paper, the cause of the failure was studied and obtained the followings. (1) The significant expansion of the rod diameter was attributed to marked oxidation of cladding outer diameter, appeared in the direction of 0 0 -180 0 degree with a shape of nodular. (2) The cladding in the place was softened by high coolant temperature. Coolant pressure, 7MPa intruded the cladding into inside chamfer void at pellet interface. (3) At the place of the significant oxidation, an instrumented transformer was existed and the coolant flow area was very little. The reduction of the coolant flow was enhanced by the bending of the cladding which was caused in pre-irradiation stage. They are considered to be a principal cause of local closure of coolant flow and resultant high temperature in the place. (author)

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

International Nuclear Information System (INIS)

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

2010-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2010-07-01

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

BWR zinc addition Sourcebook

International Nuclear Information System (INIS)

Garcia, Susan E.; Giannelli, Joseph F.; Jarvis, Alfred J.

2014-01-01

Boiling Water Reactors (BWRs) have been injecting zinc into the primary coolant via the reactor feedwater system for over 25 years for the purpose of controlling primary system radiation fields. The BWR zinc injection process has evolved since the initial application at the Hope Creek Nuclear Station in 1986. Key transitions were from the original natural zinc oxide (NZO) to depleted zinc oxide (DZO), and from active zinc injection of a powdered zinc oxide slurry (pumped systems) to passive injection systems (zinc pellet beds). Zinc addition has continued through various chemistry regimes changes, from normal water chemistry (NWC) to hydrogen water chemistry (HWC) and HWC with noble metals (NobleChem™) for mitigation of intergranular stress corrosion cracking (IGSCC) of reactor internals and primary system piping. While past reports published by the Electric Power Research Institute (EPRI) document specific industry experience related to these topics, the Zinc Sourcebook was prepared to consolidate all of the experience gained over the past 25 years. The Zinc Sourcebook will benefit experienced BWR Chemistry, Operations, Radiation Protection and Engineering personnel as well as new people entering the nuclear power industry. While all North American BWRs implement feedwater zinc injection, a number of other BWRs do not inject zinc. This Sourcebook will also be a valuable resource to plants considering the benefits of zinc addition process implementation, and to gain insights on industry experience related to zinc process control and best practices. This paper presents some of the highlights from the Sourcebook. (author)

Crud deposition modeling on BWR fuel rods

International Nuclear Information System (INIS)

Kucuk, Aylin; Cheng, Bo; Potts, Gerald A.; Shiralkar, Bharat; Morgan, Dave; Epperson, Kenny; Gose, Garry

2014-01-01

Deposition of boiling water reactor (BWR) system corrosion products (crud) on operating fuel rods has resulted in performance-limiting conditions in a number of plants. The operational impact of performance-limiting conditions involving crud deposition can be detrimental to a BWR operator, resulting in unplanned or increased frequency of fuel inspections, fuel failure and associated radiological consequences, operational restrictions including core power derate and/or forced shutdowns to remove failed fuel, premature discharge of individual bundles or entire reloads, and/or undesirable core design restrictions. To facilitate improved management of crud-related fuel performance risks, EPRI has developed the CORAL (Crud DepOsition Risk Assessment ModeL) tool. This paper presents a summary of the CORAL elements and benchmarking results. Applications of CORAL as a tool for fuel performance risk assessment are also discussed. (author)

BWR-stability investigation at Forsmark 1

International Nuclear Information System (INIS)

Bergdahl, B.G.; Reisch, F.; Oguma, R.; Lorenzen, J.; Aakerhielm, F.

1988-01-01

A series of noise measurements have been conducted at Forsmark 1 during start-up operation after the revision summer '87. The main purpose was to investigate BWR-stability problems, i.e. resonant power oscillations of 0.5 Hz around 65% power and 4100 kg/s core flow, which tend to arise at high power and low core flow conditions. The analysis was performed to estimate the noise source which gives rise to the oscillation, to evaluate the measure of stability, i.e. the Decay Ratio (Dr) as well as to investigate other safety related problems. The result indicates that the oscillation is due to the dynamic coupling between the neutron kinetics and thermal hydraulics via void reactivity feedback. The Dr ranged between values of 0.7 and > 0.9, instead of expected 0.6 (Dr=1 is defined as instability). These high values imply that the core cannot suppress oscillations fast enough and a small perturbation can cause scram. Further it was found that the entire core is oscillating in phase (LPRM's) with varying strength where any connection to the consequences of different fuel (8x8, 9x9) being present simultaneously cannot be excluded. This report elucidates the importance of an on-line BWR-stability surveillance system with functions like stability condition monitoring and control system diagnosis. (orig.)

The BWR Stability Issue

International Nuclear Information System (INIS)

D'Auria, F.

2008-01-01

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

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

International Nuclear Information System (INIS)

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

1994-04-01

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

Improvement technique of sensitized HAZ by GTAW cladding applied to a BWR power plant

International Nuclear Information System (INIS)

Tujimura, Hiroshi; Tamai, Yasumasa; Furukawa, Hideyasu; Kurosawa, Kouichi; Chiba, Isao; Nomura, Keiichi.

1995-01-01

A SCC(Stress Corrosion Cracking)-resistant technique, in which the sleeve installed by expansion is melted by GTAW process without filler metal with outside water cooling, was developed. The technique was applied to ICM (In-Core Monitor) housings of a BWR power plant in 1993. The ICM housings of which materials are type 304 Stainless Steels are sensitized with high tensile residual stresses by welding to the RPV (Reactor Pressure Vessel). As the result, ICM housings have potential of SCC initiation. Therefore, the improvement technique resistant to SCC was needed. The technique can improve chemical composition of the housing inside and residual stresses of the housing outside at the same time. Sensitization of the housing inner surface area is eliminated by replacing low-carbon with proper-ferrite microstructure clad. High tensile residual stresses of housing outside surface area is improved into compressive side. Compressive stresses of outside surface are induced by thermal stresses which are caused by inside cladding with outside water cooling. The clad is required to be low-carbon metal with proper ferrite and not to have the new sensitized HAZ (Heat Affected Zone) on the surface by cladding. The effect of the technique was qualified by SCC test, chemical composition check, ferrite content measurement and residual stresses measurement etc. All equipment for remote application were developed and qualified, too. The technique was successfully applied to a BWR plant after sufficient training

Mark III LOCA-related hydrodynamic load definition. Generic technical activity B-10. Final report

International Nuclear Information System (INIS)

Fields, M.B.; Kudrick, J.A.

1984-08-01

This report, prepared by the staff of the Office of Nuclear Reactor Regulation and its consultants at the Brookhaven National Laboratory, provides a discussion of LOCA-related suppression pool hydrodynamic loads in boiling water reactor (BWR) facilities with the Mark III pressure-suppression containment design. Its issuance completes NRC Generic Technical Activity B-10, Behavior of BWR Mark III Containment. On the basis of certain large-scale tests conducted between 1973 and 1979, the General Electric Company developed LOCA-related hydrodynamic load definitions for use in the design of the standard Mark III containment. The staff and its consultants have reviewed these load definitions and their bases and conclude that, with a few specified changes, the proposed load definitions provide conservative loading conditions. The staff approved acceptance criteria for LOCA-related hydrodynamic loads are provided in an appendix

Report on the BWR owners group radiation protection/ALARA Committee

International Nuclear Information System (INIS)

Aldrich, L.R.

1995-01-01

Radiation protection programs at U.S. boiling water reactor (BWR) stations have evolved during the 1980s and early 1990s from a regulatory adherence-based endeavor to a proactive, risk-based radiation protection and prevention mission. The objectives are no longer to merely monitor and document exposure to radiation and radioactive materials. The focus of the current programs is the optimization of radiation protection of occupational workers consistent with the purpose of producing cost-effective electric power. The newly revised 10 CFR 20 defines the term ALARA (as low as reasonably achievable) to take into account the state of technology, the economics of improvements in relation to the state of the technology, and the benefits to the public health and safety. The BWR Owners Group (BWROG) initially formed the Radiation Protection/ALARA Committee in January 1990 to evaluate methods of reducing occupational radiation exposure during refueling outages. Currently, twenty U.S. BWR owner/operators (representing 36 of the operational 37 domestic BWR units), as well as three foreign BWR operators (associate members), have broadened the scope to promote information exchange between BWR radiation protection professionals and develop good practices which will affect optimization of their radiation protection programs. In search of excellence and the challenge of becoming open-quotes World Classclose quotes performers in radiation protection, the BWROG Radiation Protection/ALARA Committee has recently accepted a role in assisting the member utilities in improving radiation protection performance in a cost-effective manner. This paper will summarize the recent activities of this Committee undertaken to execute their role of exchanging information in pursuit of optimizing the improvement of their collective radiation protection performance

Report on the BWR owners group radiation protection/ALARA Committee

Energy Technology Data Exchange (ETDEWEB)

Aldrich, L.R. [Commonwealth Edison Co., Downers Grove, IL (United States)

1995-03-01

Radiation protection programs at U.S. boiling water reactor (BWR) stations have evolved during the 1980s and early 1990s from a regulatory adherence-based endeavor to a proactive, risk-based radiation protection and prevention mission. The objectives are no longer to merely monitor and document exposure to radiation and radioactive materials. The focus of the current programs is the optimization of radiation protection of occupational workers consistent with the purpose of producing cost-effective electric power. The newly revised 10 CFR 20 defines the term ALARA (as low as reasonably achievable) to take into account the state of technology, the economics of improvements in relation to the state of the technology, and the benefits to the public health and safety. The BWR Owners Group (BWROG) initially formed the Radiation Protection/ALARA Committee in January 1990 to evaluate methods of reducing occupational radiation exposure during refueling outages. Currently, twenty U.S. BWR owner/operators (representing 36 of the operational 37 domestic BWR units), as well as three foreign BWR operators (associate members), have broadened the scope to promote information exchange between BWR radiation protection professionals and develop good practices which will affect optimization of their radiation protection programs. In search of excellence and the challenge of becoming {open_quotes}World Class{close_quotes} performers in radiation protection, the BWROG Radiation Protection/ALARA Committee has recently accepted a role in assisting the member utilities in improving radiation protection performance in a cost-effective manner. This paper will summarize the recent activities of this Committee undertaken to execute their role of exchanging information in pursuit of optimizing the improvement of their collective radiation protection performance.

Study on reactor vessel replacement (RVR) for 1100 MW class BWR plants in Japan

International Nuclear Information System (INIS)

Mizutani, J.; Kawamura, S.; Aoki, M.; Mori, T.

2001-01-01

Plant Life Management (PLM) is being studied in Japan, and reactor vessel replacement (RVR) is being considered as one option. Since reactor internals, except for reusable parts, and the reactor pressure vessel (RPV) are replaced, the RVR provides an effective technology for extending the service life of nuclear power plants substantially. At ICONE 7, we reported on the technical viability of the RVR for BWR4-type 800 MWe class plants. This time, we rationalized the RVR method through a study for BWR5-type 1100 MWe class plants to reduce the RVR duration and evaluated the technical viability and the economic efficiency of the method. In addition, we discuss how to dispose of the RPV to complete a scenario of the process from the RVR to its final disposal. (author)

On the Decay Ratio Determination in BWR Stability Analysis by Auto-Correlation Function Techniques

International Nuclear Information System (INIS)

Behringer, K.; Hennig, D.

2002-11-01

A novel auto-correlation function (ACF) method has been investigated for determining the oscillation frequency and the decay ratio in BWR stability analyses. The neutron signals are band-pass filtered to separate the oscillation peak in the power spectral density (PSD) from background. Two linear second-order oscillation models are considered. These models, corrected for signal filtering and including a background term under the peak in the PSD, are then least-squares fitted to the ACF of the previously filtered neutron signal, in order to determine the oscillation frequency and the decay ratio. Our method uses fast Fourier transform techniques with signal segmentation for filtering and ACF estimation. Gliding 'short-term' ACF estimates on a record allow the evaluation of uncertainties. Numerical results are given which have been obtained from neutron data of the recent Forsmark I and Forsmark II NEA benchmark project. Our results are compared with those obtained by other participants in the benchmark project. The present PSI report is an extended version of the publication K. Behringer, D. Hennig 'A novel auto-correlation function method for the determination of the decay ratio in BWR stability studies' (Behringer, Hennig, 2002)

Evaluation of containment failure modes and fission product releases during core meltdown accidents in a BWR with a Mark III containment

International Nuclear Information System (INIS)

Ludewig, H.; Yu, W.S.; Jaung, R.; Pratt, W.T.

1985-01-01

An assessment is described of potential failure modes and fission product releases for a large number of postulated core meltdown accidents in a BWR with a Mark III containment. For this containment design, the most important failure mode was found to be due to hydrogen related phenomena. A one-dimensional lumped parameter computer code has been developed and used to determine the probability of various hydrogen phenomena for a range of postulated core meltdown sequences. Potential containment loads have been estimated and compared against the containment capacity to determine the probability of containment failure. The fission product release assessment began by using the MARCH/CORRAL system of codes with key input parameters varied over a reasonable range. The parameters relate to primary system retention, re-emission, pool scrubbing, and fission product release in-vessel vs ex-vessel. The final step used more mechanistic calculations based on the system of codes recently developed under sponsorship of the Accident Source Term Program Office, NRC, and compares these predictions with the range of releases calculated in the sensitivity study

Validations of BWR nuclear design code using ABWR MOX numerical benchmark problems

International Nuclear Information System (INIS)

Takano, Shou; Sasagawa, Masaru; Yamana, Teppei; Ikehara, Tadashi; Yanagisawa, Naoki

2017-01-01

BWR core design code package (the HINES assembly code and the PANACH core simulator), being used for full MOX-ABWR core design, has been benchmarked against the high-fidelity numerical solutions as references, for the purpose of validating its capability of predicting the BWR core design parameters systematically from UO 2 to 100% MOX cores. The reference solutions were created by whole core critical calculations using MCNPs with the precisely modeled ABWR cores both in hot and cold conditions at BOC and EOC of the equilibrium cycle. A Doppler-Broadening Rejection Correction (DCRB) implemented MCNP5-1.4 with ENDF/B-VII.0 was mainly used to evaluate the core design parameters, except for effective delayed neutron fraction (β eff ) and prompt neutron lifetime (l) with MCNP6.1. The discrepancies in the results between the design codes HINES-PANACH and MCNPs for the core design parameters such as the bundle powers, hot pin powers, control rod worth, boron worth, void reactivity, Doppler reactivity, β eff and l, are almost within target accuracy, leading to the conclusion that HINES-PANACH has sufficient fidelity for application to full MOX-ABWR core design. (author)

Performances on nuclear activation analysis by TRIGA Mark II reactor

International Nuclear Information System (INIS)

Capannesi, G.; Rosada, A.

1986-01-01

Progresses in methodological research and connected applications in the field of activation analysis are introduced. Some peculiar characteristics on the TRIGA MARK II reactor have enabled the possibility of obtaining interesting results. The particular, the rotating radiation device Lazy Susan, with a capability of 40 positionings, permits homogeneity in neutron flux and energy spectrum stability within 15%. High level of precision and accuracy are obtained in analytic. Applications of major interest have been: - reference material certification; - forensic applications; - electrolytic cell productivity evaluation. The TRIGA MARK II reactor is equipped with a thermal column throughout a D 2 O diaphragm with a thickness of 70 cm. The available neutron flux has no fast and epithermal components. Via this facility a method has been tested for the instrumental determination of Al in Si metal of solar and electronic degree. (author)

Decontamination of TRIGA Mark II reactor, Indonesia

International Nuclear Information System (INIS)

Suseno, H.; Daryoko, M.; Goeritno, A.

2002-01-01

The TRIGA Mark II Reactor in the Centre for Research and Development Nuclear Technique Bandung has been partially decommissioned as part of an upgrading project. The upgrading project was carried out from 1995 to 2000 and is being commissioned in 2001. The decommissioning portion of the project included disassembly of some components of the reactor core, producing contaminated material. This contaminated material (grid plate, reflector, thermal column, heat exchanger and pipe) will be sent to the Decontamination Facility at the Radioactive Waste Management Development Centre. (author)

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

International Nuclear Information System (INIS)

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

1993-09-01

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

SCORPIO-BWR: status and future plans

International Nuclear Information System (INIS)

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

2004-01-01

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

SCORPIO-BWR: status and future plans

Energy Technology Data Exchange (ETDEWEB)

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

2004-07-01

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

Investigation of power oscillation mechanisms based on noise analysis at Forsmark-1 BWR

International Nuclear Information System (INIS)

Oguma, Ritsuo

1996-01-01

Noise analysis has been performed for stability test data collected during reactor start-up in January 1989 at the boiling water reactor (BWR) Forsmark unit 1. A unique instrumentation to measure local coolant flow in this reactor allowed investigation of dynamic interactions between neutron flux and coolant flow noise signals at different radial positions in the core. The causal relationship for these signals was evaluated based on a method called signal transmission path (STP) analysis with the aim of identifying the principal mechanism of power oscillations in this reactor. The results of the present study indicated that large amplitude power oscillations were induced by two instability mechanisms concurrent in the core. The first is the global void reactivity feedback effect which played the most significant role to power oscillations at a resonant frequency of about 0.53 Hz. The second is the thermal-hydraulics coupling with neutron kinetics, inducing resonant oscillations at about 0.45 Hz. The latter was found to be active only in a certain core region. A peculiar phenomenon of amplitude modulations observed in some local power range monitor (LPRM) signals was also examined. It was interpreted to occur as the consequence of these two resonant power oscillations, the frequencies of which lie close to each other. The noise analysis technique applied in the present study is expected to be useful to get a deeper understanding of the power oscillation mechanism which is active in the reactor under evaluation. The technique may be applicable to BWRs with instruments to measure local channel flow together with in-core neutron detectors. (Author)

Containment performance improvement program

International Nuclear Information System (INIS)

Beckner, W.; Mitchell, J.; Soffer, L.; Chow, E.; Lane, J.; Ridgely, J.

1990-01-01

The Containment Performance Improvement (CPI) program has been one of the main elements in the US Nuclear Regulatory Commission's (NRC's) integrated approach to closure of severe accident issues for US nuclear power plants. During the course of the program, results from various probabilistic risk assessment (PRA) studies and from severe accident research programs for the five US containment types have been examined to identify significant containment challenges and to evaluate potential improvements. The five containment types considered are: the boiling water reactor (BMR) Mark I containment, the BWR Mark II containment, the BWR Mark III containment, the pressurized water reactor (PWR) ice condenser containment, and the PWR dry containments (including both subatmospheric and large subtypes). The focus of the CPI program has been containment performance and accident mitigation, however, insights are also being obtained in the areas of accident prevention and accident management

Scaling and uncertainty in BWR instability problems

International Nuclear Information System (INIS)

Di Auria, F.; Pellicoro, V.

1995-01-01

This paper deals with a critical review of activities, performed at the DCMN of Pisa University, in relation to the thermo-hydraulic oscillations in two-phase systems. Stability analyses, including model development and achievement of experimental data, are generally performed for BWRs in order to achieve the following objectives: to reach a common understanding in relation to the predictive capabilities of system codes and to the influence of various parameters on the instability; to establish a data base for the qualification of the analytical tools already or becoming available; to set-up qualified tools (code/models + nodalization + user assumption) suitable for predicting the unstable behaviour of the nuclear plants of interest (current BWR, SBWR, ABWR and RBMK). These considerations have been the basis for the following researches: 1) proposal of the Boiling Instability Program (BIP) (1) 2) evaluation of stability tests in PIPER-ONE apparatus (2) 3) coupled thermal-hydraulic and neutronic instabilities in the LaSalle-2 BWR plant (3) 4) participation to the NEA-OECD BWR Benchmark (4) The RELAP/MOD2 and RELAP5/MOD3 codes have been used. (author)

Differential Signature of the Centrosomal MARK4 Isoforms in Glioma

Directory of Open Access Journals (Sweden)

Ivana Magnani

2011-01-01

Full Text Available Background: MAP/microtubule affinity-regulating kinase 4 (MARK4 is a serine-threonine kinase expressed in two spliced isoforms, MARK4L and MARK4S, of which MARK4L is a candidate for a role in neoplastic transformation. Methods: We performed mutation analysis to identify sequence alterations possibly affecting MARK4 expression. We then investigated the MARK4L and MARK4S expression profile in 21 glioma cell lines and 36 tissues of different malignancy grades, glioblastoma-derived cancer stem cells (GBM CSCs and mouse neural stem cells (NSCs by real-time PCR, immunoblotting and immunohistochemistry. We also analyzed the sub-cellular localisation of MARK4 isoforms in glioma and normal cell lines by immunofluorescence. Results: Mutation analysis rules out sequence variations as the cause of the altered MARK4 expression in glioma. Expression profiling confirms that MARK4L is the predominant isoform, whereas MARK4S levels are significantly decreased in comparison and show an inverse correlation with tumour grade. A high MARK4L/MARK4S ratio also characterizes undifferentiated cells, such as GBM CSCs and NSCs. Accordingly, only MARK4L is expressed in brain neurogenic regions. Moreover, while both MARK4 isoforms are localised to the centrosome and midbody in glioma and normal cells, the L isoform exhibits an additional nucleolar localisation in tumour cells. Conclusions: The observed switch towards MARK4L suggests that the balance between the MARK4 isoforms is carefully guarded during neural differentiation but may be subverted in gliomagenesis. Moreover, the MARK4L nucleolar localisation in tumour cells features this MARK4 isoform as a nucleolus-associated tumour marker.

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

International Nuclear Information System (INIS)

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

1992-01-01

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

RELAP4/MOD5: a computer program for transient thermal-hydraulic analysis of nuclear reactors and related systems. User's manual. Volume I. RELAP4/MOD5 description. [PWR and BWR

Energy Technology Data Exchange (ETDEWEB)

1976-09-01

RELAP4 is a computer program written in FORTRAN IV for the digital computer analysis of nuclear reactors and related systems. It is primarily applied in the study of system transient response to postulated perturbations such as coolant loop rupture, circulation pump failure, power excursions, etc. The program was written to be used for water-cooled (PWR and BWR) reactors and can be used for scale models such as LOFT and SEMISCALE. Additional versatility extends its usefulness to related applications, such as ice condenser and containment subcompartment analysis. Specific options are available for reflood (FLOOD) analysis and for the NRC Evaluation Model.

Burn-up TRIGA Mark II benchmark experiment

International Nuclear Information System (INIS)

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

1998-01-01

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Test results of a jet impingement from a 4 inch pipe under BWR LOCA conditions

International Nuclear Information System (INIS)

Isozaki, Toshikuni; Yano, Toshikazu; Miyazaki, Noriyuki; Kato, Rokuro; Kurihara, Ryoichi; Ueda, Shuzo; Miyazono, Shohachiro

1982-09-01

Hypothetical instantaneous pipe rupture is now considered to be one of the design basis accidents during the operation of the light water reactor. If a pipe rupture accidnet occurs, the pipe will start moving with the sudden discharge of internal fluid. So, the various apparatus such as pipe whip restraints and jet deflectors are being installed near the postulated break location to protect the nuclear power plants against the effect of postulated pipe rupture. Pipe whipping test and jet discharge test are now being conducted at the Division of Reactor Safety of the Japan Atomic Energy Research Institute. This report describes the test results of the jet discharge from a 4 inch pipe under BWR LOCA condition. In front of the pipe exit the target disk of 1000 mm in diameter was installed. The distance between the pipe exit and the target was 500 mm. 13 pressure transducers and 13 thermocouples were mounted on the target disk to measure the pressure and temperature increase due to jet impingement on the target. (author)

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

International Nuclear Information System (INIS)

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

2004-01-01

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

An A BWR demonstration simulator for training and developing technical staff

International Nuclear Information System (INIS)

Powers, J.; Yonezawa, H.; Aoyagi, Y.; Kataoka, K.

2015-09-01

The US-Advanced Boiling Water Reactor (A BWR), certified by the US NRC, is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3-4) Combined License Application (Cola). Nuclear Innovation North America (Nina) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. Toshiba has developed a Demonstration Simulator of the A BWR control room that provides a realistic experience for training and education on BWR principles and operations fundamentals. The Demonstration Simulator is located in the Toshiba America Nuclear Energy (Tane) office in Charlotte, North Carolina and is composed of standard office computer equipment set up in a specific arrangement that is representative of the layout of an A BWR control room. The Demonstration Simulator is not intended for licensed operator training, but can provide a framework for encouraging entry level technically oriented nuclear workers to enter the operations field; strengthening the linkage between university energy field curricula and real-life application of theory; and, improving understanding of integrated plant operations for developing station technical staff. This paper describes the A BWR Demonstration Simulator and its applications for training and educating future nuclear workers. (Author)

An A BWR demonstration simulator for training and developing technical staff

Energy Technology Data Exchange (ETDEWEB)

Powers, J. [Toshiba America Nuclear Energy, Charlotte, North Carolina (United States); Yonezawa, H.; Aoyagi, Y.; Kataoka, K., E-mail: jim.powers@toshiba.com [Toshiba Corporation, Kawasaki, Kanagawa (Japan)

2015-09-15

The US-Advanced Boiling Water Reactor (A BWR), certified by the US NRC, is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3-4) Combined License Application (Cola). Nuclear Innovation North America (Nina) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. Toshiba has developed a Demonstration Simulator of the A BWR control room that provides a realistic experience for training and education on BWR principles and operations fundamentals. The Demonstration Simulator is located in the Toshiba America Nuclear Energy (Tane) office in Charlotte, North Carolina and is composed of standard office computer equipment set up in a specific arrangement that is representative of the layout of an A BWR control room. The Demonstration Simulator is not intended for licensed operator training, but can provide a framework for encouraging entry level technically oriented nuclear workers to enter the operations field; strengthening the linkage between university energy field curricula and real-life application of theory; and, improving understanding of integrated plant operations for developing station technical staff. This paper describes the A BWR Demonstration Simulator and its applications for training and educating future nuclear workers. (Author)

Z0 physics from the Mark II at the SLC [SLAC Linear Collider

International Nuclear Information System (INIS)

Abrams, G.S.

1989-06-01

The MARK II detector has started to take data at the new SLAC Linear Collider. The novel aspects of the accelerator and of the MARK II are briefly described. Displays of event pictures from some of the early-on data are presented to illustrate the quality of the data. A first presentation of the results of an energy scan near the Z 0 mass that is currently in progress shows the expected resonant enhancement near 91 GeV. 2 refs., 23 figs., 1 tab

High Fidelity BWR Fuel Simulations

Energy Technology Data Exchange (ETDEWEB)

Yoon, Su Jong [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-08-01

This report describes the Consortium for Advanced Simulation of Light Water Reactors (CASL) work conducted for completion of the Thermal Hydraulics Methods (THM) Level 3 milestone THM.CFD.P13.03: High Fidelity BWR Fuel Simulation. High fidelity computational fluid dynamics (CFD) simulation for Boiling Water Reactor (BWR) was conducted to investigate the applicability and robustness performance of BWR closures. As a preliminary study, a CFD model with simplified Ferrule spacer grid geometry of NUPEC BWR Full-size Fine-mesh Bundle Test (BFBT) benchmark has been implemented. Performance of multiphase segregated solver with baseline boiling closures has been evaluated. Although the mean values of void fraction and exit quality of CFD result for BFBT case 4101-61 agreed with experimental data, the local void distribution was not predicted accurately. The mesh quality was one of the critical factors to obtain converged result. The stability and robustness of the simulation was mainly affected by the mesh quality, combination of BWR closure models. In addition, the CFD modeling of fully-detailed spacer grid geometry with mixing vane is necessary for improving the accuracy of CFD simulation.

An application of the process computer and CRT display system in BWR nuclear power station

International Nuclear Information System (INIS)

Goto, Seiichiro; Aoki, Retsu; Kawahara, Haruo; Sato, Takahisa

1975-01-01

A color CRT display system was combined with a process computer in some BWR nuclear power plants in Japan. Although the present control system uses the CRT display system only as an output device of the process computer, it has various advantages over conventional control panel as an efficient plant-operator interface. Various graphic displays are classified into four categories. The first is operational guide which includes the display of control rod worth minimizer and that of rod block monitor. The second is the display of the results of core performance calculation which include axial and radial distributions of power output, exit quality, channel flow rate, CHFR (critical heat flux ratio), FLPD (fraction of linear power density), etc. The third is the display of process variables and corresponding computational values. The readings of LPRM, control rod position and the process data concerning turbines and feed water system are included in this category. The fourth category includes the differential axial power distribution between base power distribution (obtained from TIP) and the reading of each LPRM detector, and the display of various input parameters being used by the process computer. Many photographs are presented to show examples of those applications. (Aoki, K.)

Physical model of nonlinear noise with application to BWR stability

International Nuclear Information System (INIS)

March-Leuba, J.; Perez, R.B.

1983-01-01

Within the framework of the present model it is shown that the BWR reactor cannot be unstable in the linear sense, but rather it executes limited power oscillations of a magnitude that depends on the operating conditions. The onset of these oscillations can be diagnosed by the decrease in stochasticity in the power traces and by the appearance of harmonics in the PSD

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

International Nuclear Information System (INIS)

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

1975-01-01

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

Introduction to nuclear power reactors and their health physics systems

International Nuclear Information System (INIS)

Brtis, J.S.

1982-01-01

This paper provides an introduction to: (1) the major systems of Boiling Water Reactors (BWR's) and Pressurized Water Reactors (PWR's), (2) the production and distribution of radiation sources in BWR's and PWR's, (3) the regulatory and functional requirements for nuclear power reactor design from a health physics standpoint, (4) the health physics systems provided to meet such requirements, and (5) a bibliography of documents germane to power reactor health physics design

BWR recirculation pump diagnostic expert system

International Nuclear Information System (INIS)

Chiang, S.C.; Morimoto, C.N.; Torres, M.R.

2004-01-01

At General Electric (GE), an on-line expert system to support maintenance decisions for BWR recirculation pumps for nuclear power plants has been developed. This diagnostic expert system is an interactive on-line system that furnishes diagnostic information concerning BWR recirculation pump operational problems. It effectively provides the recirculation pump diagnostic expertise in the plant control room continuously 24 hours a day. The expert system is interfaced to an on-line monitoring system, which uses existing plant sensors to acquire non-safety related data in real time. The expert system correlates and evaluates process data and vibration data by applying expert rules to determine the condition of a BWR recirculation pump system by applying knowledge based rules. Any diagnosis will be automatically displayed, indicating which pump may have a problem, the category of the problem, and the degree of concern expressed by the validity index and color hierarchy. The rules incorporate the expert knowledge from various technical sources such as plant experience, engineering principles, and published reports. These rules are installed in IF-THEN formats and the resulting truth values are also expressed in fuzzy terms and a certainty factor called a validity index. This GE Recirculation Pump Expert System uses industry-standard software, hardware, and network access to provide flexible interfaces with other possible data acquisition systems. Gensym G2 Real-Time Expert System is used for the expert shell and provides the graphical user interface, knowledge base, and inference engine capabilities. (author)

Design guideline to prevent the pipe rupture by radiolysis gases in BWR steam piping

International Nuclear Information System (INIS)

Inagaki, T.; Miyagawa, M.; Ota, T.; Sato, T.; Sakata, K.

2009-01-01

In late 2001, pipe rupture accidents due to fast combustion of radiolysis gas occurred in Japan and elsewhere's BWR power plants. TENPES began to set up the guideline as action to such a new problem to prevent accumulation and combustion of radiolysis gas in BWR steam piping. And then, the first edition of guideline was published in October 2005. Afterwards, the experimental study about combustion/detonation of radiolysis gas have been continued. And in March 2007, TENPES published a revised edition of the guideline. This is the report of the revised edition of that guideline. According to this guideline, it became possible to design BWR's steam piping to prevent accumulation of radiolysis gas. (author)

Neutronics analysis of TRIGA Mark II research reactor

Directory of Open Access Journals (Sweden)

Haseebur Rehman

2018-02-01

Full Text Available This article presents clean core criticality calculations and control rod worth calculations for TRIGA (Training, Research, Isotope production-General Atomics Mark II research reactor benchmark cores using Winfrith Improved Multi-group Scheme-D/4 (WIMS-D/4 and Program for Reactor In-core Analysis using Diffusion Equation (PRIDE codes. Cores 133 and 134 were analyzed in 2-D (r, θ and 3-D (r, θ, z, using WIMS-D/4 and PRIDE codes. Moreover, the influence of cross-section data was also studied using various libraries based on Evaluated Nuclear Data File (ENDF/B-VI.8 and VII.0, Joint Evaluated Fission and Fusion File (JEFF-3.1, Japanese Evaluated Nuclear Data Library (JENDL-3.2, and Joint Evaluated File (JEF-2.2 nuclear data. The simulation results showed that the multiplication factor calculated for all these data libraries is within 1% of the experimental results. The reactivity worth of the control rods of core 134 was also calculated with different homogenization approaches. A comparison was made with experimental and reported Monte Carlo results, and it was found that, using proper homogenization of absorber regions and surrounding fuel regions, the results obtained with PRIDE code are significantly improved.

Mark I 1/5-scale boiling water reactor pressure suppression experiment. Quick-look report for test numbers 1.0(a) and 1.0(b) performed on March 4 and 8, 1977

International Nuclear Information System (INIS)

McCauley, E.W.; Pitts, J.H.

1977-01-01

The experimental results obtained from pressure suppression experiment numbers 1.0(a) and 1.0(b) that were performed on the Lawrence Livermore Laboratory's 1 / 5 -scale boiling water reactor (BWR) Mark I pressure suppression experimental facility are summarized

Material operating behaviour of ABB BWR control rods

International Nuclear Information System (INIS)

Rebensdorff, B.; Bart, G.

2000-01-01

The BWR control rods made by ABB use boron carbide (B 4 C and hafnium as absorber material within a cladding of stainless steel. The general behaviour under operation has proven to be very good. ABB and many of their control rod customers have performed extensive inspection programs of control rod behaviour. However, due to changes in the material properties under fast and thermal neutron irradiation defects may occur in the control rods at high neutron fluences. Examinations of irradiated control rod materials have been performed in hot cell laboratories. The examinations have revealed the defect mechanism Irradiation Assisted Stress Corrosion Cracking (IASCC) to appear in the stainless steel cladding. For IASCC to occur three factors have to act simultaneously. Stress, material sensitization and an oxidising environment. Stress may be obtained from boron carbide swelling due to irradiation. Stainless steel may be sensitized to intergranular stress corrosion cracking under irradiation. Normally the reactor environment in a BWR is oxidising. The presentation focuses on findings from hot cell laboratory work on irradiated ABB BWR control rods and studies of irradiated control rod materials in the hot cells at PSI. Apart from physical, mechanical and microstructural examinations, isotope analyses were performed to describe the local isotopic burnup of boron. Consequences (such as possible B 4 C washout) of a under operation in a ABB BWR, after the occurrence of a crack is discussed based on neutron radiographic examinations of control rods operated with cracks. (author)

Evaluation of thermal margin during BWR neutron flux oscillation

International Nuclear Information System (INIS)

Takeuchi, Yutaka; Takigawa, Yukio; Chuman, Kazuto; Ebata, Shigeo

1992-01-01

Fuel integrity is very important, from the view point of nuclear power plant safety. Recently, neutron flux oscillations were observed at several BWR plants. The present paper describes the evaluations of the thermal margin during BWR neutron flux oscillations, using a three-dimensional transient code. The thermal margin is evaluated as MCPR (minimum critical power ratio). The LaSalle-2 event was simulated and the MCPR during the event was evaluated. It was a core-wide oscillation, at which a large neutron flux oscillation amplitude was observed. The results indicate that the MCPR had a sufficient margin with regard to the design limit. A regional oscillation mode, which is different from a core-wide oscillation, was simulated and the MCPR response was compared with that for the LaSalle-2 event. The MCPR decrement is greater in the regional oscillation, than in the core wide -oscillation, because of the sensitivity difference in a flow-to-power gain. A study was carried out about regional oscillation detectability, from the MCPR response view point. Even in a hypothetically severe case, the regional oscillation is detectable by LPRM signals. (author)

Pipe rupture test results; 4 inch pipe whip tests under BWR operational condition-clearance parameter experiments

International Nuclear Information System (INIS)

Ueda, Syuzo; Isozaki, Toshikuni; Miyazaki, Noriyuki; Kurihara, Ryoichi; Kato, Rokuro; Saito, Kazuo; Miyazono, Shohachiro

1981-05-01

The purpose of pipe rupture studies in JAERI is to perform the model tests on pipe whip, restraint behavior, jet impingement and jet thrust force, and to establish the computational method for analyzing these phenomena. This report describes the experimental results of pipe whip on the pipe specimens of 4 inch in diameter under BWR condition on which the pressure is 6.77 MPa and the temperature is 285 0 C. The pipe specimens were 114.3 mm (4 inch) in diameter and 8.6 mm in thickness and 4500 mm in length. Four pipe whip restraints used in the tests were the U-bar type of 8 mm in diameter and fabricated from type 304 stainless steel. The experimental parameter was the clearance (30, 50 and 100 mm). The dynamic strain behavior of the pipe specimen and the restraints was investigated by strain gages and their residual deformation was obtained by measuring marking points provided on their surface. The Pressure-time history in the pipe specimens was also obtained by pressure gages. The maximum pipe strain is caused near the restraints and increases with increase of the clearance. The experimental results of pipe whip tests indicate the effectiveness of pipe whip restraints. The ratio of absorbed strain energy of the pipe specimen to that of the restraints is nearly constant for different clearances at the overhang length of 400 mm. (author)

Nine years of operation of ITU-TRR TRIGA Mark-II reactor

International Nuclear Information System (INIS)

Yavuz, H.; Bayuelken, A.R.; Yavuz, M.A.

1988-01-01

ITU-TRR TRIGA Mark-II reactor in Istanbul with a steady state power of 250 kW and a pulsing capability up to 1200 MW has been operating since March 11,1979 with an energy release of 107.5 MWh and a total of 72 pulses. During this nearly nine years, the reactor was in operation without any major undesired shut down. One of the major problems was faced when the instrumented fuel element in position 9 of the F ring went totally out of order. Secondly, the cooling tower of the secondary cooling system could not be operated properly during the hot summer days, and also we had a tar leakage problem with the radial beam port connection to the tank. During the regular maintenance work in this summer, the measurements of changes in nuclear and physical parameters of the reactor fuel and dummy elements have also proceeded. (author)

Probabilistic Safety Assessment Of It TRIGA Mark-II Reactor

International Nuclear Information System (INIS)

Ergun, E; Kadiroglu, O.S.

1999-01-01

The probabilistic safety assessment for Istanbul Technical University (ITU) TRIGA Mark-II reactor is performed. Qualitative analysis, which includes fault and event trees and quantitative analysis which includes the collection of data for basic events, determination of minimal cut sets, calculation of quantitative values of top events, sensitivity analysis and importance measures, uncertainty analysis and radiation release from fuel elements are considered

Neutron noise analysis of BWR using time series analysis

International Nuclear Information System (INIS)

Fukunishi, Kohyu

1976-01-01

The main purpose of this paper is to give more quantitative understanding of noise source in neutron flux and to provide a useful tool for the detection and diagnosis of reactor. The space dependent effects of distributed neutron flux signals at the axial direction of two different strings are investigated by the power contribution ratio among neutron fluxes and the incoherent noise spectra of neutron fluxes derived from autoregressive spectra. The signals are measured on the medium sized commercial BWR of 460 MWe in Japan. From the obtained results, local and global noise sources in neutron flux are discussed. This method is indicated to be a useful tool for detection and diagnosis of anomalous phenomena in BWR. (orig./RW) [de

RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - models and correlations

Energy Technology Data Exchange (ETDEWEB)

Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.

1998-03-01

This document describes the major modifications and improvements made to the modeling of the RAMONA-3B/MOD0 code since 1981, when the code description and assessment report was completed. The new version of the code is RAMONA-4B. RAMONA-4B is a systems transient code for application to different versions of Boiling Water Reactors (BWR) such as the current BWR, the Advanced Boiling Water Reactor (ABWR), and the Simplified Boiling Water Reactor (SBWR). This code uses a three-dimensional neutron kinetics model coupled with a multichannel, non-equilibrium, drift-flux, two-phase flow formulation of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients and instability issues. Chapter 1 is an overview of the code`s capabilities and limitations; Chapter 2 discusses the neutron kinetics modeling and the implementation of reactivity edits. Chapter 3 is an overview of the heat conduction calculations. Chapter 4 presents modifications to the thermal-hydraulics model of the vessel, recirculation loop, steam separators, boron transport, and SBWR specific components. Chapter 5 describes modeling of the plant control and safety systems. Chapter 6 presents and modeling of Balance of Plant (BOP). Chapter 7 describes the mechanistic containment model in the code. The content of this report is complementary to the RAMONA-3B code description and assessment document. 53 refs., 81 figs., 13 tabs.

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

International Nuclear Information System (INIS)

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

2014-01-01

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

Description of steam-condensation phenomena during the loss-of-coolant accident

International Nuclear Information System (INIS)

McCauley, E.W.; Holman, G.S.; Aust, E.; Schwan, H.; Vollbrandt, J.; Fuerst, H.

1980-01-01

The development and verification of advanced computer models which describe the boiling water reactor (BWR) pressure suppression process for a hypothetical loss-of-coolant accident (LOCA) require a clear description of basic steam condensation phenomena. The GKSS Research Center, in coordination with interested institutions of West Germany and the United States, is currently conducting a test program for such basic research on a multivent BWR-related pressure suppression system. The Lawrence Livermore National Laboratory (LLNL) acts as the principal US NRC liaison for this test program, with particular emphasis on development of GKSS data for confirmatory use regarding US Mark II nuclear power plants as well as to advanced code development. The multivent test facility, placed in operation in February 1979, is a three-pipe full-scale vent system modelling main features of both the West German KWU and United States G.E. Mk II BWR pressure suppression systems. The test facility and testing programs are described

BWR full integral simulation test (FIST) pretest predictions with TRACBO2

International Nuclear Information System (INIS)

Sutherland, W.A.; Alamgir, M.

1984-01-01

The Full Integral Simulation Test program is a three pronged approach to the development of best-estimate analysis capability for BWR systems. An analytical method development program is underway to extend the BWR-TRAC computer code to model reactor kinetics and major interfacing systems, including balance-of-plant, to improve application modeling flexibility, and to reduce computer running time. An experimental program is underway in a new single bundle system test facility to extend the large break loss-of-coolant accident LOCA data base to small breaks and operational transients. And a method qualification program is underway to test TRACBO2 against experiments in the FIST facility. The recently completed Phase 1 period included a series of LOCA and power transient tests, and successful pretest analysis of the large and small break LOCA tests with TRACBO2. These comparisons demonstrate BWR-TRAC capability for small and large break analysis, and provide detailed understanding of the phenomena

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Verification of a BWR code package by gamma scan measurements

International Nuclear Information System (INIS)

Nakajima, Tsuyoshi; Iwamoto, Tatsuya; Kumanomido, Hironori

1996-01-01

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

Recent BWR fuel management reactor physics advances

International Nuclear Information System (INIS)

Crowther, R.L.; Congdon, S.P.; Crawford, B.W.; Kang, C.M.; Martin, C.L.; Reese, A.P.; Savoia, P.J.; Specker, S.R.; Welchly, R.

1982-01-01

Improvements in BWR fuel management have been under development to reduce uranium and separative work (SWU) requirements and reduce fuel cycle costs, while also maintaining maximal capacity factors and high fuel reliability. Improved reactor physics methods are playing an increasingly important role in making such advances feasible. The improved design, process computer and analysis methods both increase knowledge of the thermal margins which are available to implement fuel management advance, and improve the capability to reliably and efficiently analyze and design for fuel management advances. Gamma scan measurements of the power distributions of advanced fuel assembly and advanced reactor core designs, and improved in-core instruments also are important contributors to improving 3-d predictive methods and to increasing thermal margins. This paper is an overview of the recent advances in BWR reactor physics fuel management methods, coupled with fuel management and core design advances. The reactor physics measurements which are required to confirm the predictions of performance fo fuel management advances also are summarized

BWR internals life assurance

International Nuclear Information System (INIS)

Herrera, M.L.; Stancavage, P.P.

1988-01-01

Boiling water reactor (BWR) internal components play an important role in power plant life extension. Many important internals were not designed for easy removal and changes in material properties and local environmental effects due to high radiation makes stress corrosion cracking more likely and more difficult to correct. Over the past several years, operating experience has shown that inspection, monitoring and refurbishment can be accomplished for internal structures with existing technology. In addition, mitigation techniques which address the causes of degradation are available to assure that life extension targets can be met. This paper describes the many considerations and aspects when evaluating life extension for reactor vessel internals

BWR stability using a reducing dynamical model; Estabilidad de un BWR con un modelo dinamico reducido

Energy Technology Data Exchange (ETDEWEB)

Ballestrin Bolea, J M; Blazquez Martinez, J B

1990-07-01

BWR stability can be treated with reduced order dynamical models. When the parameters of the model came from dynamical models. When the parameters of the model came from experimental data, the predictions are accurate. In this work an alternative derivation for the void fraction equation is made, but remarking the physical structure of the parameters. As the poles of power/reactivity transfer function are related with the parameters, the measurement of the poles by other techniques such as noise analysis will lead to the parameters, but the system of equations is non-linear. Simple parametric calculation of decay ratio are performed, showing why BWRs become unstable when they are operated at low flow and high power. (Author)

Reactor physics tests of TRIGA Mark-II Reactor in Ljubljana

International Nuclear Information System (INIS)

Ravnik, M.; Mele, I.; Trkov, A.; Rant, J.; Glumac, B.; Dimic, V.

2008-01-01

TRIGA Mark-II Reactor in Ljubljana was recently reconstructed. The reconstruction consisted mainly of replacing the grid plates, the control rod mechanisms and the control unit. The standard type control rods were replaced by the fuelled follower type, the central grid location (A ring) was adapted for fuel element insertion, the triangular cutouts were introduced in the upper plate design. However, the main novelty in reactor physics and operational features of the reactor was the installation of a pulse rod. Having no previous operational experience in pulsing, a detailed and systematic sequence of tests was defined in order to check the predicted design parameters of the reactor with measurements. The following experiments are treated in this paper: initial criticality, excess reactivity measurements, control rod worth measurement, fuel temperature distribution, fuel temperature reactivity coefficient, pulse parameters measurement (peak power, prompt energy, peak temperature). Flux distributions in steady state and pulse mode were measured as well, however, they are treated only briefly due to the volume of the results. The experiments were performed with completely fresh fuel of 12 w% enriched Standard Stainless Steel type. The core configuration was uniform (one fuel element type, including fuelled followers) and compact (no irradiation channels or gaps), as such being particularly convenient for testing the computer codes for TRIGA reactor calculations. Comparison of analytical predictions, obtained with WIMS, SLXTUS, TRIGAP and PULSTRI codes to measured values showed agreement within the error of the measurement and calculation. The paper has the following contents: 1. Introduction; 2. Steady State Experiments; 2.1. Core loading and critical experiment; 2.2. Flux range determination for tests at zero power; 2.3. Digital reactivity meter checkout; 2.4. Control rod worth measurements; 2.5. Excess reactivity measurement; 2.6. Thermal power calibration; 2

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

International Nuclear Information System (INIS)

Ott, L.J.

1994-01-01

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

A simplified spatial model for BWR stability

International Nuclear Information System (INIS)

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

2012-01-01

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

BWR type reactor core

International Nuclear Information System (INIS)

Tatemichi, Shin-ichiro.

1981-01-01

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

dE/dx electronics for MARK II experiment at SLAC

International Nuclear Information System (INIS)

Bernstein, D.; Boyarski, A.; Coupal, D.; Feldman, G.; Paffrath, L.

1985-10-01

This paper describes a 100 MHz pulse digitizer for dE/dx measurements on the MARK II drift chamber at SLAC. The electronics provides the read-out of the detector's 5832 sense based on a 16-channel FASTBUS module. The basic element of the module is the TRW 6-bit Flash-ADC

Sophistication of operator training using BWR plant simulator

International Nuclear Information System (INIS)

Ohshiro, Nobuo; Endou, Hideaki; Fujita, Eimitsu; Miyakita, Kouji

1986-01-01

In Japanese nuclear power stations, owing to the improvement of fuel management, thorough maintenance and inspection, and the improvement of facilities, high capacity ratio has been attained. The thorough training of operators in nuclear power stations also contributes to it sufficiently. The BWR operator training center was established in 1971, and started the training of operators in April, 1974. As of the end of March, 1986, more than 1800 trainees completed training. At present, in the BWR operator training center, No.1 simulator of 800 MW class and No.2 simulator of 1100 MW class are operated for training. In this report, the method, by newly adopting it, good result was obtained, is described, that is, the method of introducing the feeling of being present on the spot into the place of training, and the new testing method introduced in retraining course. In the simulator training which is apt to place emphasis on a central control room, the method of stimulating trainees by playing the part of correspondence on the spot and heightening the training effect of multiple monitoring was tried, and the result was confirmed. The test of confirmation on the control board was added. (Kako, I.)

Crud removal with deep bed type condensate demineralizer in Tokai-2 BWR

International Nuclear Information System (INIS)

Abe, Ayumi; Takiguchi, Hideki; Numata, Kunio; Saito, Toshihiko

1996-01-01

The major objective and functions for the installation of the deep bed type condensate polishers in BWR power plants is to remove both ionic impurities caused by sea water leakage and suspended impurities called crud mainly consisting of metal oxides which are produced from metal corrosion. In considering the reduction of occupational radiation exposure level, it is extremely important to remove the crud effectively. In recent Japanese BWR power plants, condensate pre-filters with powdered ion exchange resins or with hollow fiber membrane have been installed to remove the crud at the upper stream of the deep bed polishers. In such plants, the crud removal is conventionally the secondary objective for the deep bed polishers. The Japan Atomic Power Company has introduced the small particle ion exchange resin and a soak regeneration method since April 1985, and then applied the low cross-linked resin since July 1995 at Tokai-2 Power Station, to improve the crud removal performance by using only deep bed type condensate demineralizer, and as a result condensate demineralizer outlet iron level has been kept below 1 ppb since 1991

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

International Nuclear Information System (INIS)

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

2002-01-01

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

The Mark II detector for the SLC

International Nuclear Information System (INIS)

Abrams, G.; Baden, A.R.; Boyer, J.; Butler, F.; Drell, P.S.; Fay, J.; Gidal, G.; Goldhaber, G.; Haggerty, J.; Harr, R.; Hearty, C.; Herrup, D.; Holmgren, S.O.; Jaffre, M.; Juricic, I.; Kadyk, J.A.; Kral, J.F.; Levi, M.E.; Lynch, G.R.; Richman, J.D.; Rouse, F.R.; Schaad, M.W.; Schmidke, W.B.; Schumm, B.A.; Trilling, G.H.; Wood, D.R.; Akerlof, C.; Bonvicini, G.; Chapman, J.; Chmeissani, M.; Frey, R.; Gero, E.; Hong, S.J.; Koska, W.; Nitz, D.; Petradza, M.; Thun, R.; Tschirhart, R.; Veltman, H.; Alexander, J.P.; Ballam, J.; Barklow, T.; Bartelt, J.; De Boer, W.; Boyarski, A.; Braune, K.; Bulos, F.; Burke, D.L.; Cords, D.; Coupal, D.P.; Destaebler, H.C.; Dorfan, J.M.; Feldman, G.J.; Fernandes, D.; Field, R.C.; Fordham, C.; Fujino, D.; Gan, K.K.; Glanzman, T.; Grosse-Wiesmann, P.; Hanson, G.; Hayes, K.; Himel, T.; Hutchinson, D.; Innes, W.R.; Jacobsen, R.G.; Jaros, J.A.; Jung, C.K.; Karlen, D.; Klein, S.R.; Koetke, D.; Komamiya, M.; Kowalski, L.A.; Kozanecki, W.; Lankford, A.J.; Larsen, R.R.; Lueth, V.; Mattison, T.; Moffeit, K.C.; Mueller, L.; Munger, C.T.; Nash, J.; Ong, R.A.; O'Shaughnessy, K.F.; Perl, J.; Perl, M.L.; Perrier, F.; Petersen, A.; Pitthan, R.; Riles, K.; Swartz, M.; Taylor, R.E.; Van Kooten, R.; Voruganti, P.; Weigend, A.; Woods, M.; Wormser, G.; Wright, R.; Alvarez, M.; Calvino, F.; Fernandez, E.; Ford, W.T.; Hinshaw, D.A.; Rankin, P.; Smith, J.G.; Wagner, S.R.; Weber, P.; White, S.L.; Averill, D.; Blockus, D.; Brabson, B.; Brom, J.M.; Murray, W.N.; Ogren, H.; Rust, D.R.; Snyder, A.; Yurko, M.; Barish, B.C.; Hawkes, C.M.; Hoenk, M.; Kuhlen, M.; Li, Z.; McKenna, J.A.; Milliken, B.D.; Nelson, M.E.; Peck, C.; Porter, F.C.; Soderstrom, E.; Stroynowski, R.; Weinstein, A.J.; Weir, A.J.; Wicklund, E.; Wolf, R.C.; Wu, D.Y.; Barnett, B.A.; Boswell, C.; Dauncey, P.; Drewer, D.C.; Harral, B.; Hylen, J.; Matthews, J.A.J.; Stoker, D.P.; Vejcik, S.; Breakstone, A.; Cence, R.J.; Gong, X.; Harris, F.A.; Koide, A.; Parker, S.I.; Green, A.; Lawrence Berkeley Lab., CA; California Univ., Berkeley

1989-01-01

The Mark II detector has been upgraded in preparation for its role as the first detector to take data at the Stanford Linear Collider. The new detector components include the central drift chamber, the time-of-flight system, the coil, the endcap electromagnetic calorimeters and the beam energy and luminosity measuring devices. There have also been improvements in detector hermeticity. All of the major components were installed for a test run at the PEP storage ring (√s=29 GeV) in 1985. This paper describes the upgraded detector, including its trigger and data acquisition systems, and gives performance figures for its components. Future improvements are also discussed. (orig.)

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

International Nuclear Information System (INIS)

Anstine, L.D.

1983-05-01

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

Synchrotron power supply of TARN II

International Nuclear Information System (INIS)

Watanabe, Shin-ichi.

1991-07-01

The construction and performance of synchrotron power supply of TARN II are described. The 1.1 GeV synchrotron-cooler TARN II has been constructed at Institute for Nuclear Study, University of Tokyo. Constructed power supply for the dipole magnets is 600 V, 2500 A operated in the mode of trapezoid wave form with the repetition cycle of 0.1 Hz. The stability of magnetic field within 10 -3 and tracking error of 10 -4 have been attained with the aid of computer control system. First trial of synchrotron acceleration of He 2+ beam has been done up to 600 MeV in April, 1991. (author)

Prony's method application for BWR instabilities characterization

Energy Technology Data Exchange (ETDEWEB)

Castillo, Rogelio, E-mail: rogelio.castillo@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera México-Toluca s/n, La Marquesa, Ocoyoacac, Estado de México 52750 (Mexico); Ramírez, J. Ramón, E-mail: ramon.ramirez@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera México-Toluca s/n, La Marquesa, Ocoyoacac, Estado de México 52750 (Mexico); Alonso, Gustavo, E-mail: gustavo.alonso@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera México-Toluca s/n, La Marquesa, Ocoyoacac, Estado de México 52750 (Mexico); Instituto Politecnico Nacional, Unidad Profesional Adolfo Lopez Mateos, Ed. 9, Lindavista, D.F. 07300 (Mexico); Ortiz-Villafuerte, Javier, E-mail: javier.ortiz@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera México-Toluca s/n, La Marquesa, Ocoyoacac, Estado de México 52750 (Mexico)

2015-04-01

Highlights: • Prony's method application for BWR instability events. • Several BWR instability benchmark are assessed using this method. • DR and frequency are obtained and a new parameter is proposed to eliminate false signals. • Adequate characterization of in-phase and out-of-phase events is obtained. • The Prony's method application is validated. - Abstract: Several methods have been developed for the analysis of reactor power signals during BWR power oscillations. Among them is the Prony's method, its application provides the DR and the frequency of oscillations. In this paper another characteristic of the method is proposed to determine the type of oscillations that can occur, in-phase or out-of-phase. Prony's method decomposes a given signal in all the frequencies that it contains, therefore the DR of the fundamental mode and the first harmonic are obtained. To determine the more dominant pole of the system a normalized amplitude W of the system is calculated, which depends on the amplitude and the damping coefficient. With this term, it can be analyzed which type of oscillations is present, if W of the fundamental mode frequency is the greater, the type of oscillations is in-phase, if W of the first harmonic frequency is the greater, the type of oscillations is out-of-phase. The method is applied to several stability benchmarks to assess its validity. Results show the applicability of the method as an alternative analysis method to determine the type of oscillations occurred.

Prony's method application for BWR instabilities characterization

International Nuclear Information System (INIS)

Castillo, Rogelio; Ramírez, J. Ramón; Alonso, Gustavo; Ortiz-Villafuerte, Javier

2015-01-01

Highlights: • Prony's method application for BWR instability events. • Several BWR instability benchmark are assessed using this method. • DR and frequency are obtained and a new parameter is proposed to eliminate false signals. • Adequate characterization of in-phase and out-of-phase events is obtained. • The Prony's method application is validated. - Abstract: Several methods have been developed for the analysis of reactor power signals during BWR power oscillations. Among them is the Prony's method, its application provides the DR and the frequency of oscillations. In this paper another characteristic of the method is proposed to determine the type of oscillations that can occur, in-phase or out-of-phase. Prony's method decomposes a given signal in all the frequencies that it contains, therefore the DR of the fundamental mode and the first harmonic are obtained. To determine the more dominant pole of the system a normalized amplitude W of the system is calculated, which depends on the amplitude and the damping coefficient. With this term, it can be analyzed which type of oscillations is present, if W of the fundamental mode frequency is the greater, the type of oscillations is in-phase, if W of the first harmonic frequency is the greater, the type of oscillations is out-of-phase. The method is applied to several stability benchmarks to assess its validity. Results show the applicability of the method as an alternative analysis method to determine the type of oscillations occurred

Collimator and shielding design for boron neutron capture therapy (BNCT) facility at TRIGA MARK II reactor

International Nuclear Information System (INIS)

Mohd Rafi Mohd Solleh; Abdul Aziz Tajuddin; Abdul Aziz Mohamed; Eid Mahmoud Eid Abdel Munem; Mohamad Hairie Rabir; Julia Abdul Karim; Yoshiaki, Kiyanagi

2011-01-01

The geometry of reactor core, thermal column, collimator and shielding system for BNCT application of TRIGA MARK II Reactor were simulated with MCNP5 code. Neutron particle lethargy and dose were calculated with MCNPX code. Neutron flux in a sample located at the end of collimator after normalized to measured value (Eid Mahmoud Eid Abdel Munem, 2007) at 1 MW power was 1.06 x 10 8 n/ cm 2 / s. According to IAEA (2001) flux of 1.00 x 10 9 n/ cm 2 / s requires three hours of treatment. Few modifications were needed to get higher flux. (Author)

Seismic risk assessment of a BWR

International Nuclear Information System (INIS)

Wells, J.E.; Bernreuter, D.L.; Chen, J.C.; Lappa, D.A.; Chuang, T.Y.; Murray, R.C.; Johnson, J.J.

1987-01-01

The simplified seismic risk methodology developed in the USNRC Seismic Safety Margins Research Program (SSMRP) was demonstrated by its application to the Zion nuclear power plant (PWR). The simplified seismic risk methodology was developed to reduce the costs associated with a seismic risk analysis while providing adequate results. A detailed model of Zion, including systems analysis models (initiating events, event trees, and fault trees), SSI and structure models, and piping models, was developed and used in assessing the seismic risk of the Zion nuclear power plant (FSAR). The simplified seismic risk methodology was applied to the LaSalle County Station nuclear power plant, a BWR; to further demonstrate its applicability, and if possible, to provide a basis for comparing the seismic risk from PWRs and BWRs. (orig./HP)

Background sources and masks for Mark II detector at PEP

International Nuclear Information System (INIS)

Kadyk, J.

1981-06-01

The shielding masks currently at use in several of the current experiments at PEP are the result of an early organized effort to understand the sources of particle background expected at PEP, followed by the evolution of the conceptual designs into actual hardware. The degree and kind of background particle loading which could be tolerated was expected to differ significantly among the different experiments, and several designs emerged from the common study. Qualitatively, the types of radiations studied were, Synchrotron Radiation (SR), Beam Gas Bremsstrahlung (BGB), and, to a limited extent others, e.g., Electroproduction (EP). Calculations will be given of predicted occupancies in the pipe counter and other sensitive elements at small radius, since these will be most susceptible to the SR and BGB backgrounds. The calculations presented in this note are specific to the Mark II detector. Some general statements will be made first about the character of each of the various types of backgrounds considered, then some detailed calculations made for application to the Mark II detector

Facility of BWR type nuclear power plant

International Nuclear Information System (INIS)

Kubo, Mitsuji

1998-01-01

A condensate filtering device for cleaning condensate flown from a low pressure turbine and a condensate desalting device are connected by way of a condensate pipeline. Control rod drives (CRD) are disposed to the lower portion of BWR. A CRD pump and one end of a CRD feedwater pipeline are connected in series to the upstream of CRD. The other end of the CRD feedwater pipeline is connected to a CRD water taking pipeline branched from the condensate pipeline. Water is taken to the CRD from downstream of the condensate filtering device and upstream of a connecting portion between a low pressure heater drain pipeline and the condensate pipeline. Flow of impurities leached out of the condensate desalting device to the reactor can be suppressed, and rising of temperature of CRD water by the low pressure heater drain water is prevented. In addition, flowing of dissolved oxygen to the CRD system can be suppressed. (I.N.)

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

International Nuclear Information System (INIS)

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

2007-01-01

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

BWR level estimation using Kalman Filtering approach

International Nuclear Information System (INIS)

Garner, G.; Divakaruni, S.M.; Meyer, J.E.

1986-01-01

Work is in progress on development of a system for Boiling Water Reactor (BWR) vessel level validation and failure detection. The levels validated include the liquid level both inside and outside the core shroud. This work is a major part of a larger effort to develop a complete system for BWR signal validation. The demonstration plant is the Oyster Creek BWR. Liquid level inside the core shroud is not directly measured during full power operation. This level must be validated using measurements of other quantities and analytic models. Given the available sensors, analytic models for level that are based on mass and energy balances can contain open integrators. When such a model is driven by noisy measurements, the model predicted level will deviate from the true level over time. To validate the level properly and to avoid false alarms, the open integrator must be stabilized. In addition, plant parameters will change slowly with time. The respective model must either account for these plant changes or be insensitive to them to avoid false alarms and maintain sensitivity to true failures of level instrumentation. Problems are addressed here by combining the extended Kalman Filter and Parity Space Decision/Estimator. The open integrator is stabilized by integrating from the validated estimate at the beginning of each sampling interval, rather than from the model predicted value. The model is adapted to slow plant/sensor changes by updating model parameters on-line

An optimized BWR fuel lattice for improved fuel utilization

International Nuclear Information System (INIS)

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

1984-01-01

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

1100 MW BWR power station

International Nuclear Information System (INIS)

Anon.

1983-01-01

Now, the start-up test of No. 2 plant in Fukushima No. 2 Nuclear Power Station is smoothly in progress, and the start of its commercial operation is scheduled at the beginning of 1984. Here, the main features of No. 2 plant including piping design are explained. For No. 2 plant, many improving measures were adopted as the base plant of the improvement and standardization project of the Ministry of International Trade and Industry, such as the adoption of Mark-2 improved PCV, the adoption of an intermediate loop in the auxiliary cooling system, one-body forging of the lower end cover of the reactor pressure vessel, the adoption of many curved pipes, the adoption of large one-body structural components in reactor recirculation system piping and so on, which are related to the reduction of radiation exposure and the improvement of plant reliability in operation and regular inspection. Also, in order to do general adjustment in the arrangement of equipment and piping and in route design, and to establish the rational construction work plan, model engineering was adopted. In No. 2 plant, a remote-controlled automatic and semiautomatic ultrasonic flaw detection system was adopted to reduce radiation exposure in in-service inspection. Automatic welding was adopted to improve the quality. (Kako, I.)

Best-estimate analysis development for BWR systems

International Nuclear Information System (INIS)

Sutherland, W.A.; Alamgir, M.; Kalra, S.P.; Beckner, W.D.

1986-01-01

The Full Integral Simulation Test (FIST) Program is a three pronged approach to the development of best-estimate analysis capability for BWR systems. An experimental program in the FIST BWR system simulator facility extends the LOCA data base and adds operational transients data. An analytical method development program with the BWR-TRAC computer program extends the modeling of BWR specific components and major interfacing systems, and improves numerical techniques to reduce computer running time. A method qualification program tests TRAC-B against experiments run in the FIST facility and extends the results to reactor system applications. With the completion and integration of these three activities, the objective of a best-estimate analysis capability has been achieved. (author)

Crud separation from equipment drain of BWR atomic power station

International Nuclear Information System (INIS)

Hayashi, Masaru; Yamaguchi, Hisashi; Moriya, Yasuhiro; Koshiba, Yukihiko; Ota, Yoshiharu.

1977-01-01

In the primary cooling systems of BWR nuclear power stations, radioactive crud is generated and accumulates in reactors and circulating systems, which causes the radiation exposure of workers at the time of the inspection and maintenance of reactors. The chemical composition and grain size distribution of crud differ largely according to the construction of primary systems, the operational conditions of reactors, and the process of operation. The study on the application of nuclear pore membrane filter NPMF to the separation of crud in the waste water from equipment drain systems has been carried out. With the NPMF, clarified filtrate can be obtained without any filter aid, therefore the secondary waste of filter sludge is not generated. When the filter is clogged, the filtration capability can be regenerated by reverse flow washing, and continuous filtration is possible actually because the regeneration takes only short time. The NPMF is the polycarbonate membrane of about 10 μm thick, to which charged particles are irradiated vertically, and the flight tracks are etched with alkali solution, thus the required pore treatment is applied. The basic investigation of waste liquid, the endurance test of actual filters, the filtration test with the pilot apparatus, the demonstration test with an actual equipment, and the design of the actual equipment have been carried out for three years. (Kako, I.)

TLTA/6431, Two-Loop-Test-Apparatus, BWR/6 Simulator, Small-Break LOCA

International Nuclear Information System (INIS)

1992-01-01

1 - Description of test facility: The Two-Loop-Test-Apparatus (TLTA) is a 1:624 volume scaled BWR/6 simulator. It was the predecessor of the better-scaled FIST facility. The facility is capable of full BWR system pressure and has a simulated core with a full size 8 x 8, full power single bundle of indirect electrically heated rods. All major BWR systems are simulated including lower plenum, guide tube, core region (bundle and bypass), upper plenum, steam separator, steam dome, annular downcomer, recirculation loops and ECC injection systems. The fundamental scaling consideration was to achieve real-time response. A number of the scaling compromises present in TLTA were corrected in the FIST configuration. These compromises include a number of regional volumes and component elevations. 2 - Description of test: 64.45 sqcm small break LOCA with activation of the full emergency core cooling system, but without activation of the automatic decompression system

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

Energy Technology Data Exchange (ETDEWEB)

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

2006-07-01

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

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

International Nuclear Information System (INIS)

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

2006-01-01

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

BWR AXIAL PROFILE

International Nuclear Information System (INIS)

Huffer, J.

2004-01-01

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

Parameter identification of a BWR nuclear power plant model for use in optimal control

International Nuclear Information System (INIS)

Volf, K.

1976-02-01

The problem being considered is the modeling of a nuclear power plant for the development of an optimal control system of the plant. Current system identification concepts, combining input/output information with a-priori structural information are employed. Two of the known parameter identification methods i.e., a least squares method and a maximum likelihood technique, are studied as ways of parameter identification from measurement data. A low order state variable stochastic model of a BWR nuclear power plant is presented as an application of this approach. The model consists of a deterministic and a noise part. The deterministic part is formed by simplified modeling of the major plant dynamic phenomena. The moise part models the effects of input random disturbances to the deterministic part and additive measurement noise. Most of the model parameters are assumed to be initially unknown. They are identified using measurement data records. A detailed high order digital computer simulation is used to simulate plant dynamic behaviour since it is not conceivable for experimentation of this kind to be performed on the real nuclear power plant. The identification task consists in adapting the performance of the simple model to the data acquired from this plant simulation ensuring the applicability of the techniques to measurement data acquired directly from the plant. (orig.) [de

European BWR R and D cluster for innovative passive safety systems

International Nuclear Information System (INIS)

Hicken, E.F.; Lensa, W. von

1996-01-01

The main technological innovation trends for future nuclear power plants tend towards a broader use of passive safety systems for the prevention, mitigation and managing of severe accident scenarios. Several approaches have been undertaken in a number of European countries to study and demonstrate the feasibility and charateristics of innovative passive safety systems. The European BWR R and D Cluster combines those experimental and analytical efforts that are mainly directed to the introduction of passive safety systems into boiling water reactor technology. The Cluster is grouped around thermohydraulic test facilities in Europe for the qualification of innovative BWR safety systems, also taking into account especially the operating experience of the nuclear power plant Dodewaard and other BWRs, which already incorporated some passive safety features. The background, the objectives, the structure of the project and the work programme are presented in this paper as well as an outline of the significance of the expected results. (orig.) [de

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

International Nuclear Information System (INIS)

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

1992-01-01

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

Development of methodology for early detection of BWR instabilities

International Nuclear Information System (INIS)

Alessandro Petruzzi; Shin Chin; Kostadin Ivanov; Asok Ray; Fan-Bill Cheung

2005-01-01

Full text of publication follows: The objective of the work presented in this paper research, which is supported by the US Department of Energy under the NEER program, is to develop an early anomaly detection methodology in order to enhance safety, availability, and operational flexibility of Boiling Water Reactor (BWR) nuclear power plants. The technical approach relies on suppression of potential power oscillations in BWRs by detecting small anomalies at an early stage and taking appropriate prognostic actions based on an anticipated operation schedule. The model of coupled (two-phase) thermal-hydraulic and neutron flux dynamics, based on the US NRC coupled code TRACE/PARCS, is being utilized as a generator of time series data for anomaly detection at an early stage. The concept of the methodology is based on the fact that nonlinear systems show bifurcation, which is a change in the qualitative behavior as the system parameters vary. Some of these parameters may change on their own accord and account for the anomaly, while certain parameters can be altered in a controlled fashion. The non-linear, non-autonomous BWR system model considered in this research exhibits phenomena at two time scales. Anomalies occur at the slow time scale while the observation of the dynamical behavior, based on which inferences are made, takes place at the fast time scale. It is assumed that: (i) the system behavior is stationary at the fast time scale; and (ii) any observable non-stationary behavior is associated with parametric changes evolving at the slow time scale. The goal is to make inferences about evolving anomalies based on the asymptotic behavior derived from the computer simulation. However, only sufficient changes in the slowly varying parameter may lead to detectable difference in the asymptotic behavior. The need to detect such small changes in parameters and hence early detection of an anomaly motivate the utilized stimulus-response approach. In this approach, the model

BWR stability using a reduced dynamical model

International Nuclear Information System (INIS)

Ballestrin Bolea, J.M.; Blazquez, J.B.

1990-01-01

BWR stability can be treated with reduced order dynamical models. When the parameters of the model came from experimental data, the predictions are accurate. In this work an alternative derivation for the void fraction equation is made, but remarking the physical struct-ure of the parameters. As the poles of power/reactivity transfer function are related with the parameters, the measurement of the poles by other techniques such as noise analysis will lead to the parameters, but the system of equations in non-linear. Simple parametric calculat-ion of decay ratio are performed, showing why BWRs become unstable when they are operated at low flow and high power. (Author). 7 refs

Power and power-to-flow reactivity transfer functions in EBR-II [Experimental Breeder Reactor II] fuel

International Nuclear Information System (INIS)

Grimm, K.N.; Meneghetti, D.

1989-01-01

Reactivity transfer functions are important in determining the reactivity history during a power transient. Overall nodal transfer functions have been calculated for different subassembly types in the Experimental Breeder Reactor II (EBR-II). Steady-state calculations for temperature changes and, hence, reactivities for power changes have been separated into power and power-to-flow-dependent terms. Axial nodal transfer functions separated into power and power-to-flow-dependent components are reported in this paper for a typical EBR-II fuel pin. This provides an improved understanding of the time dependence of these components in transient situations

Present Services at the TRIGA Mark II Reactor of the JSI

International Nuclear Information System (INIS)

Smodiš, B.; Snoj, L.

2013-01-01

The TRIGA Mark II research reactor of the Jožef Stefan Institute has been continuously operating since the year 1966. The currently offered services include: (1) Neutron activation analysis in both instrumental and radiochemical modes; (2) neutron irradiation of various kinds of materials intended to be used for research and applicative purposes; (3) training and education of university students as well as on-job training of staff working in public and private institutions, (4) verification of computer codes and nuclear data, comprising primarily criticality calculations and neutron flux distribution studies and (5) testing and development of a digital reactivity meter. The paper briefly describes the aforementioned activities and shows that even such small reactors are still indispensable in nuclear science and technology. (author)

RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - user's manual

International Nuclear Information System (INIS)

Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.

1998-03-01

This document is the User's Manual for the Boiling Water Reactor (BWR), and Simplified Boiling Water Reactor (SBWR) systems transient code RAMONA-4B. The code uses a three-dimensional neutron-kinetics model coupled with a multichannel, nonequilibrium, drift-flux, phase-flow model of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients. Chapter 1 gives an overview of the code's capabilities and limitations; Chapter 2 describes the code's structure, lists major subroutines, and discusses the computer requirements. Chapter 3 is on code, auxillary codes, and instructions for running RAMONA-4B on Sun SPARC and IBM Workstations. Chapter 4 contains component descriptions and detailed card-by-card input instructions. Chapter 5 provides samples of the tabulated output for the steady-state and transient calculations and discusses the plotting procedures for the steady-state and transient calculations. Three appendices contain important user and programmer information: lists of plot variables (Appendix A) listings of input deck for sample problem (Appendix B), and a description of the plotting program PAD (Appendix C). 24 refs., 18 figs., 11 tabs

Safety analysis calculations for a mixed and full FLIP core in a TRIGA Mark II

International Nuclear Information System (INIS)

Ringle, John C.; Hornyik, K.; Robinson, A.H.; Anderson, T.V.; Johnson, A.G.

1976-01-01

The Oregon State TRIGA Reactor will be reloading with FLIP fuel in August 1976. As we are the first Mark II TRIGA with a circular grid pattern and graphite reflector to utilize FLIP fuel, the safety analysis calculations performed at other facilities using FLIP were only of limited use to us. A multigroup, multiregion, one-dimensional diffusion theory code was used to calculate power densities in six different operational cores - mixed to full FLIP. Pulsing characteristics were obtained from a computer code based on point kinetics, with adiabatic heating of the fuel, linear temperature dependence of the specific heat, and prompt fuel temperature feedback coefficient. The results of all pertinent calculations will be presented. (author)

Investigation of decreasing reactor coolant inventory as a mechanism to reduce power during a BWR ATWS

International Nuclear Information System (INIS)

Peterson, C.E.; Chexal, V.K.; Layman, W.; Hentzen, R.D.; Gose, G.C.

1985-01-01

A best-estimate analysis was performed to evaluate the technique of intentionally reducing reactor coolant inventory in order to reduce power during a BWR ATWS event. The ATWS was initiated by closure of the main steam isolation valves. The analysis was performed with the RETRAN-02 computer code utilizing the one-dimensional kinetics option. The one-dimensional cross sections were developed using the SIMULATE-E and SIMTRAN-E computer codes. The MSIV closure transient provides some of the more severe conditions following a postulated failure to scram. In this transient, the only mechanism for removing energy from the vessel is through the safety relief valve system which results in a heating up of the suppression pool fluid. Consequently, the reactor power must be reduced so that the suppression pool temperature limits are not exceeded. Under the proposed emergency procedure guidelines for the ATWS event, the reactor vessel water level will be lowered to reduce system power. This analysis evaluated the dynamic response of the system as the water level was lowered to the top of active fuel evaluation. Correlating the system power and flow patterns to water level was of particular interest in the analysis. Under natural circulating conditions, the system flows, core power, and pressure responses are extremely tightly coupled and the analysis results proved to be very sensitive to the modeling of downcomer, upper plenum, and core regions

A detailed BWR recirculation loop model for RELAP

Energy Technology Data Exchange (ETDEWEB)

Araiza-Martínez, Enrique, E-mail: enrique.araiza@inin.gob.mx; Ortiz-Villafuerte, Javier, E-mail: javier.ortiz@inin.gob.mx; Castillo-Durán, Rogelio, E-mail: rogelio.castillo@inin.gob.mx

2017-01-15

Highlights: • A new detailed BWR recirculation loop model was developed for RELAP. • All jet pumps, risers, manifold, suction and control valves, and recirculation pump are modeled. • Model is tested against data from partial blockage of two jet pumps. • For practical applications, simulation results showed good agreement with available data. - Abstract: A new detailed geometric model of the whole recirculation loop of a BWR has been developed for the code RELAP. This detailed model includes the 10 jet pumps, 5 risers, manifold, suction and control valves, and the recirculation pump, per recirculation loop. The model is tested against data from an event of partial blockage at the entrance nozzle of one jet pump in both recirculation loops. For practical applications, simulation results showed good agreement with data. Then, values of parameters considered as figure of merit (reactor power, dome pressure, core flow, among others) for this event are compared against those from the common 1 jet pump per loop model. The results show that new detailed model led to a closer prediction of the reported power change. The detailed recirculation loop model can provide more reliable boundary condition data to a CFD models for studies of, for example, flow induced vibration, wear, and crack initiation.

BWR power uprate

International Nuclear Information System (INIS)

Berry, K.K.

2004-01-01

This paper discusses the program developed by GE Nuclear Energy (GE) to increase the power output of Boiling Water Reactors (BWRs). For the implementation of power uprate, this unique approach reduces the cost, the uncertainty and the level of effort for both the utility and the licensing authority. (author)

NUPEC BWR Full-size Fine-mesh Bundle Test (BFBT) Benchmark. Volume II: uncertainty and sensitivity analyses of void distribution and critical power - Specification

International Nuclear Information System (INIS)

Aydogan, F.; Hochreiter, L.; Ivanov, K.; Martin, M.; Utsuno, H.; Sartori, E.

2010-01-01

This report provides the specification for the uncertainty exercises of the international OECD/NEA, NRC and NUPEC BFBT benchmark problem including the elemental task. The specification was prepared jointly by Pennsylvania State University (PSU), USA and the Japan Nuclear Energy Safety (JNES) Organisation, in cooperation with the OECD/NEA and the Commissariat a l'energie atomique (CEA Saclay, France). The work is sponsored by the US NRC, METI-Japan, the OECD/NEA and the Nuclear Engineering Program (NEP) of Pennsylvania State University. This uncertainty specification covers the fourth exercise of Phase I (Exercise-I-4), and the third exercise of Phase II (Exercise II-3) as well as the elemental task. The OECD/NRC BFBT benchmark provides a very good opportunity to apply uncertainty analysis (UA) and sensitivity analysis (SA) techniques and to assess the accuracy of thermal-hydraulic models for two-phase flows in rod bundles. During the previous OECD benchmarks, participants usually carried out sensitivity analysis on their models for the specification (initial conditions, boundary conditions, etc.) to identify the most sensitive models or/and to improve the computed results. The comprehensive BFBT experimental database (NEA, 2006) leads us one step further in investigating modelling capabilities by taking into account the uncertainty analysis in the benchmark. The uncertainties in input data (boundary conditions) and geometry (provided in the benchmark specification) as well as the uncertainties in code models can be accounted for to produce results with calculational uncertainties and compare them with the measurement uncertainties. Therefore, uncertainty analysis exercises were defined for the void distribution and critical power phases of the BFBT benchmark. This specification is intended to provide definitions related to UA/SA methods, sensitivity/ uncertainty parameters, suggested probability distribution functions (PDF) of sensitivity parameters, and selected

ASTM standards associated with PWR and BWR power plant licensing, operation and surveillance

International Nuclear Information System (INIS)

McElroy, W.N.; McElroy, R.J.; Gold, R.; Lippincott, E.P.; Lowe, A.L. Jr.

1994-01-01

This paper considers ASTM Standards that are available, under revision, and are being considered in support of Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) Nuclear Power Plant (NPP) licensing, regulation, operation, surveillance and life attainment. The current activities of ASTM Committee E10 and its Subcommittees E10.02 and current activities of ASTM Committee E10 and its Subcommittees E10.02 and E10.05 and their Task Groups (TG) are described. A very important aspect of these efforts is the preparation, revision, and balloting of standards identified in the ASTM E706 Standard on Master Matrix for Light Water Reactor (LWR) Pressure Vessel (PV) Surveillance Standards. The current version (E706-87) of the Master Matrix identifies 21 ASTM LWR physics-dosimetry-metallurgy standards for Reactor Pressure Vessel (RPV) and Support Structure (SS) surveillance programs, whereas, for the next revision 34 standards are identified. The need for national and international coordination of Standards Technology Development, Transfer and Training (STDTT) is considered in this and other Symposium papers that address specific standards related physics-dosimetry-metallurgy issues. 69 refs

Stability Analysis of the EBR-I Mark-II Core Meltdown Accident

Energy Technology Data Exchange (ETDEWEB)

Lim, Jae-Yong; Kang, Chang Mu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

The purpose of this paper is to analyze the stability of the EBR-I core meltdown accident using the NuSTAB code. The result of NuSTAB analysis is compared with previous stability analysis by Sandmeier using the root locus method. The Experimental Breeder Reactor I (EBR-1) at Argonne National Laboratory was designed to demonstrate fast reactor breeding and to prove the use of liquid-metal coolant for power production and reached criticality in August 1951. The EBR-I reactor was undergoing a series of physics experiments and the Mark-II core was melted accidentally on Nov. 29, 1955. The experiment was going to increase core temperature to 500C to see if the reactor loses reactivity, and scram when the power reached 1500 kW or doubling of fission rate per second. However the operator scrammed with a slow moving control and missed the shutdown by two seconds and caused the core meltdown. The NuSTAB code has an advantage of analyzing space-dependent fast reactors and predicting regional oscillations compared to the point kinetics. Also, NuSTAB can be useful when the coupled neutronic-thermal-hydraulic codes cannot be used for stability analysis. Future work includes analyses of the PGSFR for various operating conditions as well as further validation of the NuSTAB calculations against SFR stability experiments when such experiments become available.

BWR stability using a reducing dynamical model

International Nuclear Information System (INIS)

Ballestrin Bolea, J. M.; Blazquez Martinez, J. B.

1990-01-01

BWR stability can be treated with reduced order dynamical models. When the parameters of the model came from dynamical models. When the parameters of the model came from experimental data, the predictions are accurate. In this work an alternative derivation for the void fraction equation is made, but remarking the physical structure of the parameters. As the poles of power/reactivity transfer function are related with the parameters, the measurement of the poles by other techniques such as noise analysis will lead to the parameters, but the system of equations is non-linear. Simple parametric calculation of decay ratio are performed, showing why BWRs become unstable when they are operated at low flow and high power. (Author)

Cooperative control scheme for an HVDC system connected to an isolated BWR nuclear power plant

International Nuclear Information System (INIS)

Sakurai, T.; Goto, K.; Kawai, T.; Matori, I.; Nakao, T.; Watanabe, A.

1983-01-01

This paper describes a cooperative control system to achieve stable operation of an isolated BWR nuclear plant linked to an HVDC system. In the proposed control system, under normal conditions the power plant is controlled according to the generating power reference and the generator frequency deviation is adjusted by converter power control. Such frequency control is also effective in the case of AC-DC system faults. In addition to the frequency control, an overload control is provided with the HVDC system, where the DC transmission power in the sound poles is increased due to a fault detection signal from the faulty pole. Effects of the above mentioned control systems were studied using digital dynamic programs. The sets of simulation results confirmed that in the case of a DC single pole fault, the plant is able to continue operation without any use of the turbine speed control units even for a restarting failure in the faulty pole. In case of a DC two pole fault, the plant is able to continue operation, being assisted by turbine speed control units when restarting in the faulty poles succeeds. In case of an AC three-line to ground fault near the AC terminal of the converter at the sending or receiving end, the system is able to continue stable operation, being supplemented by the turbine control unit when the faulty section of the AC system is isolated by a main or back-up relaying system

Data base formation for important components of reactor TRIGA MARK II

Energy Technology Data Exchange (ETDEWEB)

Jordan, R; Mavko, B; Kozuh, M [Inst. Jozef Stefan, Ljubljana (Slovenia)

1992-07-01

The paper represents specific data base formation for reactor TRIGA MARK II in Podgorica. Reactor operation data from year 1985 to 1990 were collected. Two groups of collected data were formed. The first group includes components data and the second group covers data of reactor scrams. Time related and demand related models were used for data evaluation. Parameters were estimated by classical method. Similar data bases are useful everywhere where components unavailabilities may have severe drawback. (author) [Slovenian] V referatu smo prikazali raziskavo, v okviru katere smo za raziskovalni reaktor TRIGA MARK II v Podgorici izoblikovali specificno bazo podatkov. Zbrali smo podatke obratovanja reaktorja od leta 1985 do 1990. Rezultate raziskave dogodkov smo razdelili v dve glavni skupini. V prvo spadajo obratovalni podatki o komponentah, v drugo skupino pa spadajo zagoni oz. zaustavitve reaktorja. Podatke smo ovrednotili z modelom v casovnem prostoru in z modelom na zahtevo. Parametre modelov smo dolocili s klasicno metodo. Opisane baze podatkov so uporabne povsod, kjer so lahko posledice nezanesljivega delovanja sistemov velike. [author].

A nonlinear 3D real-time model for simulation of BWR nuclear power plants

International Nuclear Information System (INIS)

Ercan, Y.

1982-02-01

A nonlinear transient model for BWR nuclear power plants which consists of a 3D-core (subdivided into a number of superboxes, and with parallel flow and subcooled boiling), a top plenum, steam removal and feed water systems and main coolant recirculation pumps is given. The model describes the local core and global plant transient situation as dependent on both the inherent core dynamics and external control actions, i.e., disturbances such as motions of control rod banks, changes of mass flow rates of coolant, feed water and steam outlet. The case of a pressure-controlled reactor operation is also considered. The model which forms the basis for the digital code GARLIC-B (Er et al. 82) is aimed to be used on an on-site process computer in parallel to the actual reactor process (or even in predictive mode). Thus, special measures had to be taken into account in order to increase the computational speed and reduce the necessary computer storage. This could be achieved by - separating the neutron and power kinetics from the xenon-iodine dynamics, - treating the neutron kinetics and most of the thermodynamics and hydrodynamics in a pseudostationary way, - developing a special coupling coefficient concept to describe the neutron diffusion, calculating the coupling coefficients from a basic neutron kinetics code, - combining coarse mesh elements into superboxes, taking advantage of the symmetry properties of the core and - applying a sparse matrix technique for solving the resulting algebraic power equation system. (orig.) [de

Simplified compact containment BWR plant

International Nuclear Information System (INIS)

Heki, H.; Nakamaru, M.; Tsutagawa, M.; Hiraiwa, K.; Arai, K.; Hida, T.

2004-01-01

The reactor concept considered in this paper has a small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. The Compact Containment Boiling Water Reactor (CCR), which is being developed with matured BWR technologies together with innovative systems/components, is expected to prove attractive in the world energy markets due to its flexibility in regard to both energy demands and site conditions, its high potential for reducing investment risk and its safety features facilitating public acceptance. The flexibility is achieved by CCR's small power output of 300 MWe class and capability of long operating cycle (refueling intervals). CCR is expected to be attractive from view point of investment due to its simplification/innovation in design such as natural circulation core cooling with the bottom located short core, internal upper entry control rod drives (CRDs) with ring-type dryers and simplified ECCS system with high pressure containment concept. The natural circulation core eliminates recirculation pumps and the maintenance of such pumps. The internal upper entry CRDs reduce the height of the reactor vessel (RPV) and consequently reduce the height of the primary containment vessel (PCV). The safety features mainly consist of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), passive auto catalytic recombiner and in-vessel retention (IVR) capability. The large inventory increases the system response time in the case of design-base accidents, including loss of coolant accidents. The IC suppresses PCV pressure by steam condensation without any AC power. The recombiner decreases hydrogen concentration in the PCV in the case of a severe accident. Cooling the molten core inside the RPV if the core should be damaged by loss of core coolability could attain the IVR. The feasibility of CCR safety system has been confirmed by LOCA

AREVA solutions to licensing challenges in PWR and BWR reload and safety analysis

Energy Technology Data Exchange (ETDEWEB)

Curca-Tivig, Florin [AREVA GmbH, Erlangen (Germany)

2016-05-15

Regulatory requirements for reload and safety analyses are evolving: new safety criteria, request for enlarged qualification databases, statistical applications, uncertainty propagation.. In order to address these challenges and access more predictable licensing processes, AVERA is implementing consistent code and methodology suites for PWR and BWR core design and safety analysis, based on first principles modeling and extremely broad verification and validation data base. Thanks to the high computational power increase in the last decades methods' development and application now include new capabilities. An overview of the main AREVA codes and methods developments is given covering PWR and BWR applications in different licensing environments.

BWR stability analysis

International Nuclear Information System (INIS)

Valtonen, K.

1990-01-01

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

Advances in BWR water chemistry

International Nuclear Information System (INIS)

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

2012-09-01

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

RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - user`s manual

Energy Technology Data Exchange (ETDEWEB)

Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.

1998-03-01

This document is the User`s Manual for the Boiling Water Reactor (BWR), and Simplified Boiling Water Reactor (SBWR) systems transient code RAMONA-4B. The code uses a three-dimensional neutron-kinetics model coupled with a multichannel, nonequilibrium, drift-flux, phase-flow model of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients. Chapter 1 gives an overview of the code`s capabilities and limitations; Chapter 2 describes the code`s structure, lists major subroutines, and discusses the computer requirements. Chapter 3 is on code, auxillary codes, and instructions for running RAMONA-4B on Sun SPARC and IBM Workstations. Chapter 4 contains component descriptions and detailed card-by-card input instructions. Chapter 5 provides samples of the tabulated output for the steady-state and transient calculations and discusses the plotting procedures for the steady-state and transient calculations. Three appendices contain important user and programmer information: lists of plot variables (Appendix A) listings of input deck for sample problem (Appendix B), and a description of the plotting program PAD (Appendix C). 24 refs., 18 figs., 11 tabs.

System control model of a turbine for a BWR; Modelo del sistema de control de una turbina para un BWR

Energy Technology Data Exchange (ETDEWEB)

Vargas O, Y. [Universidad del Valle de Mexico, Campus Toluca, Av. Las Palmas No. 136, Col. San Jorge Pueblo Nuevo, 52140 Metepec, Estado de Mexico (Mexico); Amador G, R.; Ortiz V, J.; Castillo D, R.; Delfin L, A. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)], e-mail: rodolfo.amador@inin.gob.mx

2009-10-15

In this work is presented a design of a control system of a turbine for a nuclear power plant with a BWR like energy source. The model seeks to implement later on at thermal hydraulics code of better estimate RELAP/SCDAPSIM. The model is developed for control and protection of turbine, and the consequent protection to the BWR, considering that the turbine control could be employed for one or several turbines in series. The quality of present designs of control pattern of turbine it is that it considers the parameters more important in the operation of a turbine besides that is has incorporated at control the secondary parameters that will be activated originally as true when the turbine model is substituted by a model more detailed. The development of control model of a turbine will be good in short and medium term to realize analysis about the operation of turbine with different operation conditions, of vapor extraction specific steps of turbine to feed other equipment s, besides analyzing the separate effect and integrated effect. (Author)

TRACE/PARCS analysis of the OECD/NEA Oskarshamn-2 BWR stability benchmark

Energy Technology Data Exchange (ETDEWEB)

Kozlowski, T. [Univ. of Illinois, Urbana-Champaign, IL (United States); Downar, T.; Xu, Y.; Wysocki, A. [Univ. of Michigan, Ann Arbor, MI (United States); Ivanov, K.; Magedanz, J.; Hardgrove, M. [Pennsylvania State Univ., Univ. Park, PA (United States); March-Leuba, J. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Hudson, N.; Woodyatt, D. [Nuclear Regulatory Commission, Rockville, MD (United States)

2012-07-01

On February 25, 1999, the Oskarshamn-2 NPP experienced a stability event which culminated in diverging power oscillations with a decay ratio of about 1.4. The event was successfully modeled by the TRACE/PARCS coupled code system, and further analysis of the event is described in this paper. The results show very good agreement with the plant data, capturing the entire behavior of the transient including the onset of instability, growth of the oscillations (decay ratio) and oscillation frequency. This provides confidence in the prediction of other parameters which are not available from the plant records. The event provides coupled code validation for a challenging BWR stability event, which involves the accurate simulation of neutron kinetics (NK), thermal-hydraulics (TH), and TH/NK. coupling. The success of this work has demonstrated the ability of the 3-D coupled systems code TRACE/PARCS to capture the complex behavior of BWR stability events. The problem was released as an international OECD/NEA benchmark, and it is the first benchmark based on measured plant data for a stability event with a DR greater than one. Interested participants are invited to contact authors for more information. (authors)

Lujan Mark-4

Energy Technology Data Exchange (ETDEWEB)

Mocko, Michael Jeffrey [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Zavorka, Lukas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Koehler, Paul E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-11-13

This is a review of Mark-IV target neutronics design. It involved the major redesign of the upper tier, offering harder neutron spectra for upper-tier FPs; a redesign of the high-resolution (HR) moderator; and a preservation of the rest of Mark-III features.

Standard Technical Specifications, General Electric plants, BWR/6

International Nuclear Information System (INIS)

1992-09-01

This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. This document Volume 2, contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

Standard Technical Specifications, General Electric Plants, BWR/6

International Nuclear Information System (INIS)

1992-09-01

This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. This document Volume 1, contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-07-01

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

Alternative cooling water flow path for RHR heat exchanger and its effect on containment response during extended station blackout for Chinshan BWR-4 plant

Energy Technology Data Exchange (ETDEWEB)

Yuann, Yng-Ruey, E-mail: ryyuann@iner.gov.tw

2016-04-15

Highlights: • Motivating alternative RHR heat exchanger tube-side flow path and determining required capacity. • Calculate NSSS and containment response during 24-h SBO for Chinshan BWR-4 plant. • RETRAN and GOTHIC models are developed for NSSS and containment, respectively. • Safety relief valve blowdown flow and energy to drywell are generated by RETRAN. • Analyses are performed with and without reactor depressurization, respectively. - Abstract: The extended Station Blackout (SBO) of 24 h has been analyzed with respect to the containment response, in particular the suppression pool temperature response, for the Chinshan BWR-4 plant of MARK-I containment. The Chinshan plant, owned by Taiwan Power Company, has twin units with rated core thermal power of 1840 MW each. The analysis is aimed at determining the required alternative cooling water flow capacity for the residual heat removal (RHR) heat exchanger when its tube-side sea water cooling flow path is blocked, due to some reason such as earthquake or tsunami, and is switched to the alternative raw water source. Energy will be dissipated to the suppression pool through safety relief valves (SRVs) of the main steam lines during SBO. The RETRAN model is used to calculate the Nuclear Steam Supply System (NSSS) response and generate the SRV blowdown conditions, including SRV pressure, enthalpy, and mass flow rate. These conditions are then used as the time-dependent boundary conditions for the GOTHIC code to calculate the containment pressure and temperature response. The shaft seals of the two recirculation pumps are conservatively assumed to fail due to loss of seal cooling and a total leakage flow rate of 36 gpm to the drywell is included in the GOTHIC model. Based on the given SRV blowdown conditions, the GOTHIC containment calculation is performed several times, through the adjustment of the heat transfer rate of the RHR heat exchanger, until the criterion that the maximum suppression pool temperature

Perturbation analysis of the TRIGA Mark II reactor Vienna

Energy Technology Data Exchange (ETDEWEB)

Khan, R. [Pakistan Institute of Engineering and Applied Sciences (PIEAS), Islamabad (Pakistan); Villa, M.; Stummer, T.; Boeck, H. [Vienna Univ. of Technology (Austria). Atominstitut; Saeedbadshah [International Islamic Univ., Islamabad (Pakistan)

2013-04-15

The safety design of a nuclear reactor needs to maintain the steady state operation at desired power level. The safe and reliable reactor operation demands the complete knowledge of the core multiplication and its changes during the reactor operation. Therefore it is frequently of interest to compute the changes in core multiplication caused by small disturbances in the field of reactor physics. These disturbances can be created either by geometry or composition changes of the core. Fortunately if these changes (or perturbations) are very small, one does not have to repeat the reactivity calculations. This article focuses the study of small perturbations created in the Central Irradiation Channel (CIC) of the TRIGA mark II core to investigate their reactivity influences on the core reactivity. For this purpose, 3 different kinds of perturbations are created by inserting 3 different samples in the CIC. The cylindrical void (air), heavy water (D2O) and Cadmium (Cd) samples are inserted into the CIC separately to determine their neutronics behavior along the length of the core. The Monte Carlo N-Particle radiation transport code (MCNP) is applied to simulate these perturbations in the CIC. The MCNP theoretical predictions are verified by the experiments performed on the current reactor core. The behavior of void in the whole core and its dependence on position and water fraction is also presented in this article. (orig.)

Investigation of control rod worth and nuclear end of life of BWR control rods

International Nuclear Information System (INIS)

Magnusson, Per

2008-01-01

This work has investigated the Control Rod Worth (CRW) and Nuclear End of Life (NEOL) values for BWR control rods. A study of how different parameters affect NEOL was performed with the transport code PHOENIX4. It was found that NEOL, expressed in terms of 10 B depletion, can be generalized beyond the conditions for which the rod is depleted, such as different power densities and void fractions, the corresponding variation in the NEOL will be about 0.2-0.4% 10 B. It was also found that NEOL results for different fuel types and different fuel enrichments have a variation of about 2-3% in 10 B depletion. A comparative study on NHOL and CRW was made between PHOENIX4 and the stochastic Monte Carlo code MCNP. It was found that there is a significant difference, both due to differences in the codes and to limitations in the geometrical modeling in PHOENIX4. Since MCNP is considered more physically correct, a methodology was developed to calculate the nuclear end of life of BWR control rods with MCNP. The advantages of the methodology are that it does not require other codes to perform the depletion of the absorber material, it can describe control rods of any design and it can deplete the control rod absorber material without burning the fuel. The disadvantage of the method is that is it time-consuming

The noise analysis and the BWR operation map

International Nuclear Information System (INIS)

Blazquez, J.; Ballestrin, J.

1996-01-01

An analytical expression for the Decay Ratio is obtained: DR = exp(-bW / P 1/2 ). The physics behind is also explained. It applies to a commercial BWR Operation Map, on the vicinity of the power instability. This functional form seems fitting to the structure of the Operation map. The power P and the coolant flow are measured straightforward; the Decay Ratio is obtained by neutron noise analysis techniques. The parameter b, depending on the void reactivity coefficient, is then calculated on line during the Reactor Operation. New DR value is now predicted for each new displacement on the Map, so unexpected instability events are more likely avoided. (authors)

The LaSalle probabilistic safety analysis

International Nuclear Information System (INIS)

Frederick, L.G.; Massin, H.L.; Crane, G.R.

1987-01-01

A probabilistic safety analysis has been performed for LaSalle County Station, a twin-unit General Electric BWR5 Mark II nuclear power plant. A primary objective of this PSA is to provide engineers with a useful and useable tool for making design decisions, performing technical specification optimization, evaluating proposed regulatory changes to equipment and procedures, and as an aid in operator training. Other objectives are to identify the hypothetical accident sequences that would contribute to core damage frequency, and to provide assurance that the total expected frequency of core-damaging accidents is below 10 -4 per reactor-year in response to suggested goals. (orig./HSCH)

Experience with advanced driver fuels in EBR-II

International Nuclear Information System (INIS)

Lahm, C.E.; Koenig, J.F.; Pahl, R.G.; Porter, D.L.; Crawford, D.C.

1992-01-01

The Experimental Breeder Reactor II (EBR-II) is a complete nuclear power plant, incorporating a pool-type liquid-metal reactor (LMR) with a fuel-power thermal output of 62.5 MW and an electrical output of 20 MW. Initial criticality was in 1961, utilizing a metallic driver fuel design called the Mark-I. The fuel design has evolved over the last 30 yr, and significant progress has been made on improving performance. The first major innovations were incorporated into the Mark-II design, and burnup then increased dramatically. This design performed successfully, and fuel element lifetime was limited by subassembly hardware performance rather than the fuel element itself. Transient performance of the fuel was also acceptable and demonstrated the ability of EBR-II to survive severe upsets such as a loss of flow without scram. In the mid 1980s, with renewed interest in metallic fuels and Argonne's integral fast reactor (IFR) concept, the Mark-II design was used as the basis for new designs, the Mark-III and Mark-IV. In 1987, the Mark-III design began qualification testing to become a driver fuel for EBR-II. This was followed in 1989 by the Mark-IIIA and Mark-IV designs. The next fuel design, the Mark-V, is being planned to demonstrate the utilization of recycled fuel. The fuel cycle facility attached to EBR-II is being refurbished to produce pyroprocessed recycled fuel as part of the demonstration of the IFR

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-06-06

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

Effects of Void Uncertainties on Pin Power Distributions and the Void Reactivity Coefficient for a 10X10 BWR Assembly

International Nuclear Information System (INIS)

Jatuff, F.; Krouthen, J.; Helmersson, S.; Chawla, R.

2004-01-01

A significant source of uncertainty in Boiling Water Reactor physics is associated with the precise characterisation of the axially-dependent neutron moderation properties of the coolant inside the fuel assembly channel, and the corresponding effects on reactor physics parameters such as the lattice neutron multiplication, the neutron migration length, and the pin-by-pin power distribution. In this paper, the effects of particularly relevant void fraction uncertainties on reactor physics parameters have been studied for a BWR assembly of type Westinghouse SVEA-96 using the CASMO-4, HELIOS/PRESTO-2 and MCNP4C codes. The SVEA-96 geometry is characterised by the sub-division of the assembly into four different sub-bundles by means of an inner bypass with a cruciform shape. The study has covered the following issues: (a) the effects of different cross-section data libraries on the void coefficient of reactivity, for a wide range of void fractions; (b) the effects due to a heterogeneous vs. homogeneous void distribution inside the sub-bundles; and (c) the consequences of partly inserted absorber blades producing different void fractions in different sub-bundles. (author)

LCLS-II high power RF system overview and progress

Energy Technology Data Exchange (ETDEWEB)

Yeremian, Anahid Dian

2015-10-07

A second X-ray free electron laser facility, LCLS-II, will be constructed at SLAC. LCLS-II is based on a 1.3 GHz, 4 GeV, continuous-wave (CW) superconducting linear accelerator, to be installed in the first kilometer of the SLAC tunnel. Multiple types of high power RF (HPRF) sources will be used to power different systems on LCLS-II. The main 1.3 GHz linac will be powered by 280 1.3 GHz, 3.8 kW solid state amplifier (SSA) sources. The normal conducting buncher in the injector will use four more SSAs identical to the linac SSAs but run at 2 kW. Two 185.7 MHz, 60 kW sources will power the photocathode dual-feed RF gun. A third harmonic linac section, included for linearizing the bunch energy spread before the first bunch compressor, will require sixteen 3.9 GHz sources at about 1 kW CW. A description and an update on all the HPRF sources of LCLS-II and their implementation is the subject of this paper.

Analysis of risk-dominant sequences by MAAP3.0 for Kuosheng Nuclear Power Plant

International Nuclear Information System (INIS)

Lin, J.D.; Chieng, C.C.; Wang, T.K.; Hsiue, R.K.

1987-01-01

Kuosheng Nuclear Power Plant is the first operating model-6/Mark III boiling water reactor (BWR6/Mark III) in the world, and a probabilistic risk assessment (PRA) has been performed to determine the likely frequencies of core melt accidents and the magnitude, composition, and fraction of fission products released in these accidents. The final report of this PRA indicates that categories 8 and 15 are ranked No. 1 by risk index (the product of release frequency and release fraction) and release frequency, respectively. The dominant contributors of these two categories are frequent earthquakes and typhoons

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

International Nuclear Information System (INIS)

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

2007-01-01

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

Quantitative evaluation for training results of nuclear plant operator on BWR simulator

International Nuclear Information System (INIS)

Sato, Takao; Sato, Tatsuaki; Onishi, Hiroshi; Miyakita, Kohji; Mizuno, Toshiyuki

1985-01-01

Recently, the reliability of neclear power plants has largely risen, and the abnormal phenomena in the actual plants are rarely encountered. Therefore, the training using simulators becomes more and more important. In BWR Operator Training Center Corp., the training of the operators of BWR power plants has been continued for about ten years using a simulator having the nearly same function as the actual plants. The recent high capacity ratio of nuclear power plants has been mostly supported by excellent operators trained in this way. Taking the opportunity of the start of operation of No.2 simulator, effort has been exerted to quantitatively grasp the effect of training and to heighten the quality of training. The outline of seven training courses is shown. The technical ability required for operators, the items of quantifying the effect of training, that is, operational errors and the time required for operation, the method of quantifying, the method of collecting the data and the results of the application to the actual training are described. It was found that this method is suitable to quantify the effect of training. (Kako, I.)

Photographic and video techniques used in the 1/5-scale Mark I boiling water reactor pressure suppression experiment

International Nuclear Information System (INIS)

Dixon, D.; Lord, D.

1978-01-01

The report provides a description of the techniques and equipment used for the photographic and video recordings of the air test series conducted on the 1/5 scale Mark I boiling water reactor (BWR) pressure suppression experimental facility at Lawrence Livermore Laboratory (LLL) between March 4, 1977, and May 12, 1977. Lighting and water filtering are discussed in the photographic system section and are also applicable to the video system. The appendices contain information from the photographic and video camera logs

Method of operating BWR type reactors

International Nuclear Information System (INIS)

Sekimizu, Koichi

1980-01-01

Purpose: To enable reactor control depending on any demanded loads by performing control by the insertion of control rods in addition to the control by the regulation of the flow rate of the reactor core water at high power operation of a BWR type reactor. Method: The power is reduced at high power operation by decreasing the flow rate of reactor core water from the starting time for the power reduction and the flow rate is maintained after the time at which it reaches the minimum allowable flow rate. Then, the control rod is started to insert from the above time point to reduce the power to an aimed level. Thus, the insufficiency in the reactivity due to the increase in the xenon concentration can be compensated by the withdrawal of the control rods and the excess reactivity due to the decrease in the xenon concentration can be compensated by the insertion of the control rods, whereby the reactor power can be controlled depending on any demanded loads without deviating from the upper or lower limit for the flow rate of the reactor core water. (Moriyama, K.)

Report on the phase II R and D program of magneto-hydro-dynamics (MHD) electrical power generation. Prompt report by Electrotechnical Laboratory; Denji ryutai (MHD) hatsuden no dainiki kenkyu kaihatsu ni kansuru hokokusho. Densoken kenkyu sokuho

Energy Technology Data Exchange (ETDEWEB)

NONE

1989-07-28

This report summarizes results of the phase II R and D program of MHD electrical power generation (FY 1976 - 1983), which has been now completed. The phase II R and D efforts were concentrated on development of the durable power generation channels, where the designs and manufacture of the Mark II system were started, and the elementary techniques were simultaneously studied for, e.g., phenomena occurring around the electrodes, seed condensation and its effects on the electrode phenomena, and electrode and insulator materials for the power generation channels. The power generation channel was tested for its durability for a total of 430 hours, after it was incorporated in the Mark II system. The MHD power generation can incorporate direct combustion of coal, and will hold a dominant position in coal-fired power generation, which is expected to grow in the future. For this reason, the basic research schedules were revised in March, 1983, and the Mark II system was operated by firing a mixed fuel of kerosene and finely divided coal in a kerosene combustor, in line with the revised project, to understand the basic power generation characteristics with the combustion gases containing coal slag. (NEDO)

Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in BWR Environments

International Nuclear Information System (INIS)

Chen, Y.; Chopra, O. K.; Gruber, Eugene E.; Shack, William J.

2010-01-01

The internal components of light water reactors are exposed to high-energy neutron irradiation and high-temperature reactor coolant. The exposure to neutron irradiation increases the susceptibility of austenitic stainless steels (SSs) to stress corrosion cracking (SCC) because of the elevated corrosion potential of the reactor coolant and the introduction of new embrittlement mechanisms through radiation damage. Various nonsensitized SSs and nickel alloys have been found to be prone to intergranular cracking after extended neutron exposure. Such cracks have been seen in a number of internal components in boiling water reactors (BWRs). The elevated susceptibility to SCC in irradiated materials, commonly referred to as irradiation-assisted stress corrosion cracking (IASCC), is a complex phenomenon that involves simultaneous actions of irradiation, stress, and corrosion. In recent years, as nuclear power plants have aged and irradiation dose increased, IASCC has become an increasingly important issue. Post-irradiation crack growth rate and fracture toughness tests have been performed to provide data and technical support for the NRC to address various issues related to aging degradation of reactor-core internal structures and components. This report summarizes the results of the last group of tests on compact tension specimens from the Halden-II irradiation. The IASCC susceptibility of austenitic SSs and heat-affected-zone (HAZ) materials sectioned from submerged arc and shielded metal arc welds was evaluated by conducting crack growth rate and fracture toughness tests in a simulated BWR environment. The fracture and cracking behavior of HAZ materials, thermally sensitized SSs and grain-boundary engineered SSs was investigated at several doses (3 dpa). These latest results were combined with previous results from Halden-I and II irradiations to analyze the effects of neutron dose, water chemistry, alloy compositions, and welding and processing conditions on IASCC. The

Steady-state thermal-hydraulic analysis of the Moroccan TRIGA MARK II reactor by using PARET/ANL and COOLOD-N2 codes

International Nuclear Information System (INIS)

Boulaich, Y.; Nacir, B.; El Bardouni, T.; Zoubair, M.; El Bakkari, B.; Merroun, O.; El Younoussi, C.; Htet, A.; Boukhal, H.; Chakir, E.

2011-01-01

Research highlights: → The COOLOD/N2 and PARET/ANL codes were used for a steady-state thermal-hydraulic and safety analysis of the 2 MW TRIGA MARK II reactor located at the Nuclear Studies Center of Maamora (CENM), Morocco. → The main objective of this study is to ensure the safety margins of different safety related parameters by steady-state calculations at full power level (2 MW). → The most important conclusion is that all obtained values of DNBR, fuel center and surface temperature, cladding surface temperature and coolant temperature across the hottest channel are largely far to compromise safety of the reactor. - Abstract: The COOLOD/N2 and PARET/ANL codes were used for a steady-state thermal-hydraulic and safety analysis of the 2 MW TRIGA MARK II reactor located at the Nuclear Studies Center of Maamora (CENM), Morocco. In order to validate our PARET/ANL and COOLOD-N2 models, the fuel center temperature as function of core power was calculated and compared with the corresponding experimental values. The comparison indicates that the calculated values are in satisfactory agreement with the measurement. The main objective of this study is to ensure the safety margins of different safety related parameters by steady-state calculations at full power level (2 MW). Therefore, we have calculated the departure from nucleate boiling ratio (DNBR), fuel center and surface temperature, cladding surface temperature and coolant temperature profiles across the hottest channel. The most important conclusion is that all obtained values are largely far to compromise safety of the reactor.

Accident management information needs for a BWR with a MARK I containment

Energy Technology Data Exchange (ETDEWEB)

Chien, D.N.; Hanson, D.J. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

1991-05-01

In support of the US Nuclear Regulatory Commission Accident Management Research Program, information needs during severe accidents have been evaluated for Boiling Water Reactors (BWRs) with MARK 1 containments. This evaluation was performed using a methodology that identifies plant information needs necessary for personnel to: (a) diagnose that an accident is in progress, (b) select and implement strategies to prevent or mitigate the accident, and (c) monitor the effectiveness of these strategies. The information needs and capabilities identified are intended to form a basis for more comprehensive information needs assessments. The assessments will be performed during the analysis and development of specific strategies, which will be used in accident management prevention and mitigation. 3 refs., 4 figs., 2 tabs.

Accident management information needs for a BWR with a MARK I containment

International Nuclear Information System (INIS)

Chien, D.N.; Hanson, D.J.

1991-05-01

In support of the US Nuclear Regulatory Commission Accident Management Research Program, information needs during severe accidents have been evaluated for Boiling Water Reactors (BWRs) with MARK 1 containments. This evaluation was performed using a methodology that identifies plant information needs necessary for personnel to: (a) diagnose that an accident is in progress, (b) select and implement strategies to prevent or mitigate the accident, and (c) monitor the effectiveness of these strategies. The information needs and capabilities identified are intended to form a basis for more comprehensive information needs assessments. The assessments will be performed during the analysis and development of specific strategies, which will be used in accident management prevention and mitigation. 3 refs., 4 figs., 2 tabs

Decommissioning of Swedish nuclear power reactors. Technology and costs

International Nuclear Information System (INIS)

1994-06-01

The main topics discussed are planning, technology and costs of decommissioning nuclear power reactors. Oskarshamn-3 (BWR) and Ringhals-4 (PWR) have been used as reference reactors. 29 refs, figs, tabs

Radioactive waste management plan during the TRIGA Mark II and III decommissioning

International Nuclear Information System (INIS)

Jung, K.J.; Park, S.K.; Geong, G.H.; Lee, K.W.; Chung, U.S.; Paik, S.T.

2001-01-01

The decontamination and decommissioning (D and D) project of TRIGA Mark-I and Mark-II (KRR 1 and 2) was started in January 1997 and will be completed by December 2002. In the first year of the project, work was performed in preparation of the decommissioning plan, start of the environmental impact assessment and setup licensing procedure and documentation for the project with cooperation of the Korea Institute of Nuclear Safety (KINS). In the second year, Hyundai Engineering Company (HEC) with British Nuclear Fuels pie (BNFL) as technical assisting partner was designated as the contractor to do design and licensing documentation for the D and D of both reactors. After pre-design, a hazard and operability (HAZOP) study checked each step of the work. At the end of 1998, the decommissioning plan documentation including environmental impact assessment report was finished and submitted to the Ministry of Science and Technology (MOST) for licensing. It is expected to be issued by the end of September 1999. Practical work will then be started around the end of 1999. The safe treatment and management of the radioactive waste arising from the D and D activities is of utmost importance for successful completion of the practical dismantling work. This paper summarizes general aspects of radioactive waste treatment and management plan for the TRIGA Mark-I and II decommissioning work. (author)

Experiences with monitoring and control of microbiological growth in the standby service water system of a BWR nuclear power plant

International Nuclear Information System (INIS)

Zisson, P.S.; Whitaker, J.M.; Neilson, H.L.; Mayne, L.L.

1995-01-01

In 1989, the Unites States Nuclear Regulatory Commission formally recognized the potential for nuclear accidents resulting from microbiological causes. Such causes range from loss of heat transfer due to microbiological fouling, to loss of system integrity caused by microbiologically influenced corrosion (MIC). As a result of these potential problems, monitoring, mitigation, and control procedures must be developed by all regulated plants. In developing a control and mitigation strategy for the standby service water system of a boiling water reactor (BWR) nuclear power plant, numerous monitoring techniques were employed to evaluate effectiveness. This paper describes the monitoring techniques that were evaluated, and those that ultimately proved to be effective

Characteristic evaluations of BWR uprate method based on heat balance shift concept

International Nuclear Information System (INIS)

Kitou, Kazuaki; Aoyama, Motoo; Shiina, Kouji; Sasaki, Hiroshi; Yoshikawa, Kazuhiro

2007-01-01

Reactor power uprate of nuclear power plants is an efficient plant operating method. Most BWR plants need the exchange of high pressure turbines when plant thermal power increases over 5% because main steam flow rate exceeds the limitation of inlet steam flow rate of a high pressure turbine. Therefore, the new power uprate method named heat balance shift power uprate method has been developed. This method decreases feedwater temperature with increasing plant thermal power not to increase main steam flower rate. This study clarified that the heat balance shift method could increase plant electric power up to 2.8% compared with conventional power uprate method without the exchange of a high pressure turbine. (author)

Generic BWR-4 degraded core in-vessel study. Status report

International Nuclear Information System (INIS)

1984-11-01

Original intent of this project was to produce a phenomenological study of the in-vessel degradation which occurs during the TQUX and TQUV sequences for a generic BWR-4 from the initiation of the FSAR Chapter 15 operational transient through core debris bed formation to the failure of the primary pressure boundary. Bounding calculations were to be performed for the two high pressure and low pressure non-LOCA scenarios to assess the uncertainties in the current state of knowledge regarding the source terms for containment integrity studies. Source terms as such were defined in terms of hydrogen generation, unreacted metal, and coolant inventroy, and in terms of the form, sequencing and mode of dispersal through the primary vessel boundary. Fission product release was not to be considered as part of this study. Premature termination of the project, however, led to the dicontinuation of work on an as is basis. Work on the in-core phase from the point of scram to core debris bed formation was largely completed. A preliminary scoping calculation on the debris bed phase had been initiated. This report documents the status of the study at termination

Advanced Construction of Compact Containment BWR

International Nuclear Information System (INIS)

Takahashi, M.; Maruyama, T.; Mori, H.; Hoshino, K.; Hijioka, Y.; Heki, H.; Nakamaru, M.; Hoshi, T.

2006-01-01

The reactor concept considered in this paper has a mid/small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. Compact Containment BWR (CCR) is being developed with matured BWR technologies together with innovative systems/components, will provide attractiveness for the energy market in the world due to its flexibility in energy demands as well as in site conditions, its high potential in reducing investment risk and its safety feature facilitating public acceptance. The flexibility is achieved by CCR's mid/small power output of 400 MWe class and capability of long operating cycle (refueling intervals). The high investment potential is expected from CCR's simplification/innovation in design such as natural circulation core cooling with the bottom located short core, top mounted upper entry control rod drives (CRDs) with ring-type dryers and simplified safety system with high pressure resistible primary containment vessel (PCV) concept. The natural circulation core eliminates recirculation pumps as well as needs for maintenance of such pumps. The top mounted upper entry CRDs enable the bottom located short core in RPV. The safety feature mainly consists of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), high pressure resistible PCV and in-vessel retention (IVR) capability. The large inventory increases the system response time in case of design base accidents including loss of coolant accidents. The IC suppresses PCV pressure by steam condensation without any AC power. Cooling the molten core inside the RPV if the core should be damaged by loss of core coolability could attain the IVR. CCR's specific self-standing steel high pressure resistible PCV is designed to contain minimum piping and valves inside with reactor pressure vessel (RPV), only 13 m in diameter and 24 m in height. This compact PCV makes it possible to

BWR Services maintenance training program

International Nuclear Information System (INIS)

Cox, J.H.; Chittenden, W.F.

1979-01-01

BWR Services has implemented a five-phase program to increase plant availability and capacity factor in operating BWR's. One phase of this program is establishing a maintenance training program on NSSS equipment; the scope encompasses maintenance on both mechanical equipment and electrical control and instrumentation equipment. The program utilizes actual product line equipment for practical Hands-on training. A total of 23 formal courses will be in place by the end of 1979. The General Electric Company is making a multimillion dollar investment in facilities to support this training. These facilities are described

HVDC transmission from nuclear power plant

International Nuclear Information System (INIS)

Yoshida, Yukio; Takenaka, Kiyoshi; Ichikawa, Takemi; Ueda, Kiyotaka; Machida, Takehiko

1979-01-01

The HVDC transmission directly from nuclear power plants is one of the patterns of long distance and large capacity HVDC transmission systems. In this report, the double pole, two-circuit HVDC transmission from a BWR nuclear power plant is considered, and the dynamic response characteristics due to the faults in dc line and ac line of inverter side are analyzed, to clarify the dynamic characteristics of the BWR nuclear power plant and dc system due to system faults and the effects of dc power control to prevent reactor scram. (1) In the instantaneous earthing fault of one dc line, the reactor is not scrammed by start-up within 0.8 sec. (2) When the earthing fault continues, power transmission drops to 75% by suspending the faulty pole, and the reactor is scrammed. (3) In the instantaneous ground fault of 2 dc lines, the reactor is not scrammed if the faulty dc lines are started up within 0.4 sec. (4) In the existing control of dc lines, the reactor is scrammed when the ac voltage at an ac-dc connection point largely drops due to ac failure. (J.P.N.)

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Criticality calculation in TRIGA MARK II PUSPATI Reactor using Monte Carlo code

International Nuclear Information System (INIS)

Rafhayudi Jamro; Redzuwan Yahaya; Abdul Aziz Mohamed; Eid Abdel-Munem; Megat Harun Al-Rashid; Julia Abdul Karim; Ikki Kurniawan; Hafizal Yazid; Azraf Azman; Shukri Mohd

2008-01-01

A Monte Carlo simulation of the Malaysian nuclear reactor has been performed using MCNP Version 5 code. The purpose of the work is the determination of the multiplication factor (k e ff) for the TRIGA Mark II research reactor in Malaysia based on Monte Carlo method. This work has been performed to calculate the value of k e ff for two cases, which are the control rod either fully withdrawn or fully inserted to construct a complete model of the TRIGA Mark II PUSPATI Reactor (RTP). The RTP core was modeled as close as possible to the real core and the results of k e ff from MCNP5 were obtained when the control fuel rods were fully inserted, the k e ff value indicates the RTP reactor was in the subcritical condition with a value of 0.98370±0.00054. When the control fuel rods were fully withdrawn the value of k e ff value indicates the RTP reactor is in the supercritical condition, that is 1.10773±0.00083. (Author)

Core followup studies of the Tarapur Reactors with the three dimensional BWR simulator COMTEG

Energy Technology Data Exchange (ETDEWEB)

Dwivedi, S. R.; Jagannathan, V.; Mohanakrishnan, P.; Srinivasan, K. R.; Rastogi, B. P.

1976-07-01

Both the units of the Tarapur Atomic Power Station started operation in the year 1969. Since then, these units have completed three cycles. For efficient operation and fuel management of these reactors, a three dimensional BWR simulator COMETG has been developed. The reactors are closely being followed using the simulator. The detailed analyses for cycle 3/4 operation of both the units are described in the paper. The results show very good agreement between calculated and measured values. It is concluded that reactor core behaviour could be predicted in a satisfactory manner with the core simulator COMETG.

Logical model for the control of a BWR turbine;Modelo logico para el control de una turbina de un BWR

Energy Technology Data Exchange (ETDEWEB)

Vargas O, Y. [Universidad del Valle de Mexico, Campus Toluca, Av. Las Palmas No. 136, Col. San Jorge Pueblo Nuevo, 52140 Metepec, Estado de Mexico (Mexico); Amador G, R.; Ortiz V, J.; Castillo D, R., E-mail: yonaeton@hotmail.co [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2009-07-01

In this work a design of a logical model is presented for the turbine control of a nuclear power plant with a BWR like energy source. The model is sought to implement later on inside the thermal hydraulics code of better estimate RELAP/SCDAPSIM. The logical model is developed for the control and protection of the turbine, and the consequent protection to the BWR, considering that the turbine control will be been able to use for one or several turbines in series. The quality of the present design of the logical model of the turbine control is that it considers the most important parameters in the operation of a turbine, besides that they have incorporated to the logical model the secondary parameters that will be activated originally as true when the turbine model is substituted by a detailed model. The development of the logical model of a turbine will be of utility in the short and medium term to carry out analysis on the turbine operation with different operation conditions, of vapor extraction, specific steps of the turbine to feed other equipment s, in addition to analyze the separate and the integrated effect. (Author)

Time-dependent coolant velocity measurements in an operating BWR

International Nuclear Information System (INIS)

Luebbesmeyer, D.; Crowe, R.D.

1980-01-01

A method to measure time-dependent fluid velocities in BWR-bundle elements by noise analysis of the incore-neutron-detector signals is shown. Two application examples of the new method are given. The time behaviour of the fluid velocity in the bundle element during a scheduled power excursion of the plant. The change of power was performed by changing the coolant flow through the core The apparent change of the fluid velocity due to thermal elongation of the helix-drive of the TIP-system. A simplified mathematical model was derived for this elongation to use as a reference to check the validity of the new method. (author)

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Corrosion problem in the CRENK Triga Mark II research reactor

International Nuclear Information System (INIS)

Kalenga, M.

1990-01-01

In August 1987, a routine underwater optical inspection of the aluminum tank housing the core of the CRENK Triga Mark II reactor, carried out to update safety condition of the reactor, revealed pitting corrosion attacks on the 8 mm thick aluminum tank bottom. The paper discuss the work carried out by the reactor staff to dismantle the reactor in order to allow a more precise investigation of the corrosion problem, to repair the aluminum tank bottom, and to enhance the reactor overall safety condition

Instrumentation and testing of a prestressed concrete containment vessel model

International Nuclear Information System (INIS)

Hessheimer, M.F.; Pace, D.W.; Klamerus, E.W.

1997-01-01

Static overpressurization tests of two scale models of nuclear containment structures - a steel containment vessel (SCV) representative of an improved, boiling water reactor (BWR) Mark II design and a prestressed concrete containment vessel (PCCV) for pressurized water reactors (PWR) - are being conducted by Sandia National Laboratories for the Nuclear Power Engineering Corporation of Japan and the U.S. Nuclear Regulatory Commission. This paper discusses plans for instrumentation and testing of the PCCV model. 6 refs., 2 figs., 2 tabs

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

International Nuclear Information System (INIS)

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

2012-09-01

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

BWR Servicing and Refueling Improvement Program: Phase I summary report

Energy Technology Data Exchange (ETDEWEB)

Perry, D.R.

1978-09-01

Under the U.S. Department of Energy sponsorship, General Electric Co. (GE) undertook a study of boiling water reactor (BWR) refueling outages for the purpose of recommending the development and demonstration of critical path time savings improvements. The Tennessee Valley Authority (TVA) joined the study as a subcontractor, providing monitoring assistance and making the Browns Ferry Site available for improvement demonstrations. Agreement was also reached with Georgia Power Co., Power Authority of the State of New York, and Commonwealth Edison Co. for monitoring and data collection at Hatch 1, FitzPatrick, and Quad Cities 1 nuclear plants, respectively. The objective was to identify, develop, and demonstrate improved refueling, maintenance, and inspection procedures and equipment. The improvements recommended in this study are applicable to BWR nuclear plants currently in operation as well as those in the design and construction phases. The recommendations and outage information can be used as a basis to plan and conduct the first outages of new plants and to improve the planning and facilities of currently operating plants. Many of the recommendations can readily be incorporated in plants currently in the design and construction phases as well as in the design of future plants. Many of these recommended improvements can be implemented immediately by utilities without further technical development.

BWR Servicing and Refueling Improvement Program: Phase I summary report

International Nuclear Information System (INIS)

Perry, D.R.

1978-09-01

Under the U.S. Department of Energy sponsorship, General Electric Co. (GE) undertook a study of boiling water reactor (BWR) refueling outages for the purpose of recommending the development and demonstration of critical path time savings improvements. The Tennessee Valley Authority (TVA) joined the study as a subcontractor, providing monitoring assistance and making the Browns Ferry Site available for improvement demonstrations. Agreement was also reached with Georgia Power Co., Power Authority of the State of New York, and Commonwealth Edison Co. for monitoring and data collection at Hatch 1, FitzPatrick, and Quad Cities 1 nuclear plants, respectively. The objective was to identify, develop, and demonstrate improved refueling, maintenance, and inspection procedures and equipment. The improvements recommended in this study are applicable to BWR nuclear plants currently in operation as well as those in the design and construction phases. The recommendations and outage information can be used as a basis to plan and conduct the first outages of new plants and to improve the planning and facilities of currently operating plants. Many of the recommendations can readily be incorporated in plants currently in the design and construction phases as well as in the design of future plants. Many of these recommended improvements can be implemented immediately by utilities without further technical development

The possibility of gamma ray sterilization by using ITU TRIGA Mark II reactor

International Nuclear Information System (INIS)

Bilge, A.N.; Tugrul, B.; Yavuz, H.

1988-01-01

Gamma rays are one of the effective method for sterilization which is preferred for a long time. Generally Co-60 radioisotope sources betatrons or accelerators are used for the sterilization. In this work, it was aimed to find the possibilities of the sterilization by gamma rays obtained in ITU TRIGA Mark-II radial tube. Radiation dosages are measured in the radial tube and several medical products are irradiated. Irradiation is arranged according to the desired dosages. Irradiated sterilized goods (mainly medical products) tested and checked at the Governmental Medical Health Center and results compared with literature. It can be seen that this kind of irradiation is a good tool for sterilization. Unfortunately, because of the stability of the radial tube and impracticality of the system it is rather difficult to compete with industrial system using Co-60 and accelerators. Nevertheless, this type of irradiation is also applicable for the purpose of the sterilization by using ITU TRIGA Mark II. (author)

Residual stress analysis in BWR pressure vessel attachments

International Nuclear Information System (INIS)

Dexter, R.J.; Leung, C.P.; Pont, D.

1992-06-01

Residual stresses from welding processes can be the primary driving force for stress corrosion cracking (SCC) in BWR components. Thus, a better understanding of the causes and nature of these residual stresses can help assess and remedy SCC. Numerical welding simulation software, such as SYSWELD, and material property data have been used to quantify residual stresses for application to SCC assessments in BWR components. Furthermore, parametric studies using SYSWELD have revealed which variables significantly affect predicted residual stress. Overall, numerical modeling techniques can be used to evaluate residual stress for SCC assessments of BWR components and to identify and plan future SCC research

Using TRIGA Mark II research reactor for irradiation with thermal neutrons

Energy Technology Data Exchange (ETDEWEB)

Kolšek, Aljaž, E-mail: aljaz.kolsek@gmail.com; Radulović, Vladimir, E-mail: vladimir.radulovic@ijs.si; Trkov, Andrej, E-mail: andrej.trkov@ijs.si; Snoj, Luka, E-mail: luka.snoj@ijs.si

2015-03-15

Highlights: • Monte Carlo N-Particle Transport Code was used to design and perform calculations. • Characterization of the TRIGA Mark II ex-core irradiation facilities was performed. • The irradiation device was designed in the TRIGA irradiation channel. • The use of the device improves the fraction of thermal neutron flux by 390%. - Abstract: Recently a series of test irradiations was performed at the JSI TRIGA Mark II reactor for the Fission Track-Thermoionization Mass Spectrometry (FT-TIMS) method, which requires a well thermalized neutron spectrum for sample irradiation. For this purpose the Monte Carlo N-Particle Transport Code (MCNP5) was used to computationally support the design of an irradiation device inside the TRIGA model and to support the actual measurements by calculating the neutron fluxes inside the major ex-core irradiation facilities. The irradiation device, filled with heavy water, was designed and optimized inside the Thermal Column and the additional moderation was placed inside the Elevated Piercing Port. The use of the device improves the ratio of thermal neutron flux to the sum of epithermal and fast neutron flux inside the Thermal Column Port by 390% and achieves the desired thermal neutron fluence of 10{sup 15} neutrons/cm{sup 2} in irradiation time of 20 h.

Decontamination and decommissioning project status of the TRIGA Mark II and III in Korea

International Nuclear Information System (INIS)

Paik, S.T.; Park, S.K.; Chung, K.W.; Chung, U.S.; Jung, K.J.

1999-01-01

TRIGA Mark-II, the first research reactor in Korea, has operated since 1962, and the second one, TRIGA Mark-III since 1972. Both of them had their operation phased out in 1995 due to their lives and operation of the new research reactor, HANARO (High-flux Advanced Neutron Application Reactor) at the Korea Atomic Energy Institute (KAERI) in Taejon. Decontamination and decommissioning (D and D) project of TRIGA Mark-II and Mark-III was started in January 1997 and will be completed in December 2002. The first year of the project, work was performed in preparation of the decommissioning plan, start of the environmental impact assessment and setup licensing procedure and documentation for the project with cooperation of Korea Institute of Nuclear Safety (KINS). Hyundai Engineering Company (HEC) is the main contractor to do design and licensing documentation for the D and D of both reactors. British Nuclear Fuels plc (BNFL) is the technical assisting partner of HEC. The decommissioning plan document was submitted to the Ministry of Since and Technology (MOST) for the decommissioning license in December 1998, and it expecting to be issued a license in mid 1999. The goal of this project is to release the reactor site and buildings as an unrestricted area. This paper summarizes current status and future plan for the D and D project. (author)

Thermohydraulic stability coupled to the neutronic in a BWR

International Nuclear Information System (INIS)

Calleros M, G.; Zapata Y, M.; Gomez H, R.A.; Mendez M, A.; Castlllo D, R.

2006-01-01

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

System control model of a turbine for a BWR

International Nuclear Information System (INIS)

Vargas O, Y.; Amador G, R.; Ortiz V, J.; Castillo D, R.; Delfin L, A.

2009-10-01

In this work is presented a design of a control system of a turbine for a nuclear power plant with a BWR like energy source. The model seeks to implement later on at thermal hydraulics code of better estimate RELAP/SCDAPSIM. The model is developed for control and protection of turbine, and the consequent protection to the BWR, considering that the turbine control could be employed for one or several turbines in series. The quality of present designs of control pattern of turbine it is that it considers the parameters more important in the operation of a turbine besides that is has incorporated at control the secondary parameters that will be activated originally as true when the turbine model is substituted by a model more detailed. The development of control model of a turbine will be good in short and medium term to realize analysis about the operation of turbine with different operation conditions, of vapor extraction specific steps of turbine to feed other equipment s, besides analyzing the separate effect and integrated effect. (Author)

A FASTBUS flash ADC system for the Mark II vertex chamber

International Nuclear Information System (INIS)

Barker, L.

1988-10-01

This is a description of a flash ADC system built for the Mark II experiment at the Stanford Linear Accelerator Center (SLAC). This system was designed for use in the experiment's vertex chamber where signals could occur over a relatively long time, approximately 10 microseconds. This long time, coupled with fast cable amplifiers, necessitated an alternate design approach than was used with a dE/dX FASTBUS flash ADC design. 1 ref., 6 figs

3-D flux distribution and criticality calculation of TRIGA Mark-II

International Nuclear Information System (INIS)

Can, B.

1982-01-01

In this work, the static calculation of the (I.T.U. TRIGA Mark-II) flux distribution has been made. The three dimensional, r-θ-z, representation of the core has been used. In this representation, for different configuration, the flux distribution has been calculated depending on two group theory. The thermal-hydraulics, the poisoning effects have been ignored. The calculations have been made by using the three dimensional and multigroup code CAN. (author)

Study of heavy quark production with the Mark II at PEP

International Nuclear Information System (INIS)

Abrams, G.; Amidei, D.; Baden, A.

1983-10-01

The methods adopted by the Mark II collaboration to study heavy quark production at PEP are described. Two complementary techniques are used: D* tagging using the decay chain D* + . D 0 π + , D 0 → K - π + , and inclusive lepton tagging using the characteristic p/sub T/ distributions to distinguish contributions from b and c quarks. These techniques are used to derive information about heavy quark fragmentation and about the weak coupling of heavy quarks

A study of heat capacity temperature limit of BWR

International Nuclear Information System (INIS)

Wang, Shih-Jen; Chen, Jyh-Jun; Chien, Chun-Sheng; Teng, Jyh-Tong

2012-01-01

Highlights: ► The purpose of this study is to verify the HCTL. ► MAAP4 was used as code to generate a realistic and convenient HCTL. ► The current HCTL curve causes confusing in reading data. ► The revised HCTL curves developed in this study. ► Users can obtain important parameters from the revised HCTL without confusion and interpolation. - Abstract: Heat capacity temperature limit (HCTL) is an important parameter for operation of BWR. Current version of the HCTL was derived, based on simple model of computation aids (CA) of BWR owners’ group (BWROG). However, some parts of the current HCTL are confusing to the users in reading data. The purpose of this study is to verify the HCTL by applying the MAAP4 code to the field of emergency operating procedure (EOP). The trends of HCTL generated by MAAP4 code are consistent with those obtained from CA. A series of revised HCTL evaluated at various times after scram are provided and the confusing part is eliminated.

A study of heat capacity temperature limit of BWR

Energy Technology Data Exchange (ETDEWEB)

Wang, Shih-Jen, E-mail: sjenwang@iner.gov.tw [Institute of Nuclear Energy Research (INER), 1000, Wunhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China); Chen, Jyh-Jun [Department of Mechanical Engineering, Chung Yuan Christian University, 200, Chung Pei Rd., Chung Li City, Taoyuan County 32023, Taiwan (China); Chien, Chun-Sheng [Institute of Nuclear Energy Research (INER), 1000, Wunhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China); Teng, Jyh-Tong [Department of Mechanical Engineering, Chung Yuan Christian University, 200, Chung Pei Rd., Chung Li City, Taoyuan County 32023, Taiwan (China)

2012-02-15

Highlights: Black-Right-Pointing-Pointer The purpose of this study is to verify the HCTL. Black-Right-Pointing-Pointer MAAP4 was used as code to generate a realistic and convenient HCTL. Black-Right-Pointing-Pointer The current HCTL curve causes confusing in reading data. Black-Right-Pointing-Pointer The revised HCTL curves developed in this study. Black-Right-Pointing-Pointer Users can obtain important parameters from the revised HCTL without confusion and interpolation. - Abstract: Heat capacity temperature limit (HCTL) is an important parameter for operation of BWR. Current version of the HCTL was derived, based on simple model of computation aids (CA) of BWR owners' group (BWROG). However, some parts of the current HCTL are confusing to the users in reading data. The purpose of this study is to verify the HCTL by applying the MAAP4 code to the field of emergency operating procedure (EOP). The trends of HCTL generated by MAAP4 code are consistent with those obtained from CA. A series of revised HCTL evaluated at various times after scram are provided and the confusing part is eliminated.

Comparative study of the hydrogen generation during short term station blackout (STSBO) in a BWR

International Nuclear Information System (INIS)

Polo-Labarrios, M.A.; Espinosa-Paredes, G.

2015-01-01

Highlights: • Comparative study of generation in a simulated STSBO severe accident. • MELCOR and SCDAP/RELAP5 codes were used to understanding the main phenomena. • Both codes present similar thermal-hydraulic behavior for pressure and boil off. • SCDAP/RELAP5 predicts 15.8% lower hydrogen production than MELCOR. - Abstract: The aim of this work is the comparative study of hydrogen generation and the associated parameters in a simulated severe accident of a short-term station blackout (STSBO) in a typical BWR-5 with Mark-II containment. MELCOR (v.1.8.6) and SCDAP/RELAP5 (Mod.3.4) codes were used to understand the main phenomena in the STSBO event through the results comparison obtained from simulations with these codes. Due that the simulation scope of SCDAP/RELAP5 is limited to failure of the vessel pressure boundary, the comparison was focused on in-vessel severe accident phenomena; with a special interest in the vessel pressure, boil of cooling, core temperature, and hydrogen generation. The results show that at the beginning of the scenario, both codes present similar thermal-hydraulic behavior for pressure and boil off of cooling, but during the relocation, the pressure and boil off, present differences in timing and order of magnitude. Both codes predict in similar time the beginning of melting material drop to the lower head. As far as the hydrogen production rate, SCDAP/RELAP5 predicts 15.8% lower production than MELCOR

Fuel performance experience at TVO nuclear power plant

International Nuclear Information System (INIS)

Patrakka, E.T.

1985-01-01

TVO nuclear power plant consists of two BWR units of ASEA-ATOM design. The fuel performance experience extending through six cycles at TVO I and four cycles at TVO II is reported. The experience obtained so far is mainly based on ASEA-ATOM 8 x 8 fuel and has been satisfactory. Until autumn 1984 one leaking fuel assembly had been identified at TVO I and none at TVO II. Most of the problems encountered have been related to leaf spring screws and channel screws. The experience indicates that satisfactory fuel performance can be achieved when utilizing strict operational rules and proper control of fuel design and manufacture. (author)

Simplified distributed parameters BWR dynamic model for transient and stability analysis

International Nuclear Information System (INIS)

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

2006-01-01

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

Summary report of seismic PSA of BWR model plant

International Nuclear Information System (INIS)

1999-05-01

This report presents a seismic PSA (Probabilistic Safety Assessment) methodology developed at the Japan Atomic Energy Research Institute (JAERI) for evaluating risks of nuclear power plants (NPPs) and the results from an application of the methodology to a BWR plant in Japan, which is termed Model Plant'. The seismic PSA procedures developed at JAERI are to evaluate core damage frequency (CDF) and have the following four steps: (1) evaluation of seismic hazard, (2) evaluation of realistic response, (3) evaluation of component capacities and failure probabilities, and (4) evaluation of conditional probability of system failure and CDF. Although these procedures are based on the methodologies established and used in the United States, they include several unique features: (1) seismic hazard analysis is performed with use of available knowledge and database on seismological conditions in Japan; (2) response evaluation is performed with a response factor method which is cost effective and associated uncertainties can be reduced with use of modern methods of design calculations; (3) capacity evaluation is performed with use of test results available in Japan in combination with design information and generic capacity data in the U.S.A.; (4) systems reliability analysis, performed with use of the computer code SECOM-2 developed at JAERI, includes identification of dominant accident sequences, importance analysis of components and systems as well as the CDF evaluation with consideration of the effect of correlation of failures by a newly developed method based on the Monte Carlo method. The effect of correlation has been recognized as an important issue in seismic PSAs. The procedures was used to perform a seismic PSA of a 1100 MWe BWR plant. Results are shown as well as the insights derived and future research needs identified in this seismic PSA. (J.P.N.)

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.)

ECP measurements in the BWR-1 water loop relative to water composition changes

Energy Technology Data Exchange (ETDEWEB)

Kus, P.; Vsolak, R.; Kysela, J., E-mail: ksp@ujv.cz [Nuclear Research Inst. Rez plc, Husinec - Rez (Czech Republic); Hanawa, S.; Nakamura, T.; Uchida, S., E-mail: hanawa.satoshi@jaea.go.jp [Japan Atomic Energy Agency, Tokai-mura, Ibaraki (Japan)

2010-07-01

The goal of this study is to investigate the usage of ECP sensors in nuclear power plants. ECP sensors were tested using the LVR-15 reactor at the Nuclear Research Institute Rez plc (NRI) in the Czech Republic. The experiment took place on the BWR-1 loop, which was designed for investigating the behaviour of structural materials and radioactivity transport under BWR conditions. The BWR-1 loop facilitates irradiation experiments within a wide range of operating parameters (max. pressure of 10 MPa, max. temperature of 573 K and a neutron flux of 1.0* 10{sup 18} n/m{sup 2}s). This study involves the measurement of electrochemical potential (ECP). Corrosion potential is the main parameter for monitoring of water composition changes in nuclear power plants (NPP). The electrochemical potentials of stainless steel were measured under high temperatures in a test loop (BWR-1) under different water composition conditions. Total neutron flux was ∼10{sup -3} to ∼10{sup 12} n/cm{sup 2}s (>0.1 MeV) at a temperature of 560K, neutral pH, and water resistivity of 18.2 MOhm. ECP sensor response related to changes in water composition was monitored. Switching from NWC (normal water conditions) to HWC (hydrogen water conditions) was controlled using oxygen dosage. Water chemistry was monitored approx. 50 meters from the active channel. The active channel temperature was maintained within a range of 543 - 561 K from the start of irradiation for the entire duration of the experiment. A total of 24 reference electrodes composed of platinum (Pt), silver/silver chloride (Ag/AgCl) and a zircon membrane containing silver oxide (Ag{sub 2}O) powder were installed inside the active channel of the LVR-15 test reactor. The active channel (Field tube) was divided into four zones, with each zone containing six sensors. A mathematical radiolysis code model was created in cooperation with the Japan Atomic Energy Agency. (author)

Data base formation for important components of reactor TRIGA MARK II

International Nuclear Information System (INIS)

Jordan, R.; Mavko, B.; Kozuh, M.

1992-01-01

The paper represents specific data base formation for reactor TRIGA MARK II in Podgorica. Reactor operation data from year 1985 to 1990 were collected. Two groups of collected data were formed. The first group includes components data and the second group covers data of reactor scrams. Time related and demand related models were used for data evaluation. Parameters were estimated by classical method. Similar data bases are useful everywhere where components unavailabilities may have severe drawback. (author) [sl

A FASTBUS flash ADC system for the Mark II vertex chamber

Energy Technology Data Exchange (ETDEWEB)

Barker, L.

1988-10-01

This is a description of a flash ADC system built for the Mark II experiment at the Stanford Linear Accelerator Center (SLAC). This system was designed for use in the experiment's vertex chamber where signals could occur over a relatively long time, approximately 10 microseconds. This long time, coupled with fast cable amplifiers, necessitated an alternate design approach than was used with a dE/dX FASTBUS flash ADC design. 1 ref., 6 figs.

The theological significance of the Isaiah citation in Mark 4:12 ...

African Journals Online (AJOL)

The well-known passage Mark 4:1–34 is no stranger to New Testament scientific scrutiny, not to even mention the hotly debated phrases in Mark 4:10–12. To avoid repetition, the aim with this article is to determine the extent of the impact the Isaiah 6:9–10 citation in Mark 4:12 might have had on the interpretation and ...

BWR water chemistry impurity studies

International Nuclear Information System (INIS)

Ljungberg, L.G.; Korhonen, S.; Renstroem, K.; Hofling, C.G.; Rebensdorff, B.

1990-03-01

Laboratory studies were made on the effect of water impurities on environmental cracking in simulated BWR water of stainless steel, low alloy steel and nickel-base alloys. Constant elongation rate tensile (CERT) tests were run in simulated normal water chemistry (NWC), hydrogen water chemistry (HWC), or start-up environment. Sulfate, chloride and copper with chloride added to the water at levels of a fraction of a ppM were found to be extremely deleterious to all kinds of materials except Type 316 NG. Other detrimental impurities were fluoride, silica and some organic acids, although acetic acid was beneficial. Nitrate and carbon dioxide were fairly inoccuous. Corrosion fatigue and constant load tests on compact tension specimens were run in simulated normal BWR water chemistry (NWC) or hydrogen water chemistry (HWC), without impurities or with added sulfate or carbon dioxide. For sensitized Type 304 SS in NWC, 0.1 ppM sulfate increased crack propagation rates in constant load tests by up to a factor of 100, and in fatigue tests up to a factor of 10. Also, cracking in Type 316 nuclear grade SS and Alloy 600 was enhanced, but to a smaller degree. Carbon dioxide was less detrimental than sulfate. 3 figs., 4 tabs

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

Energy Technology Data Exchange (ETDEWEB)

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

1995-09-01

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

Application of the coupled code Athlet-Quabox/Cubbox for the extreme scenarios of the OECD/NRC BWR turbine trip benchmark and its performance on multi-processor computers

International Nuclear Information System (INIS)

Langenbuch, S.; Schmidt, K.D.; Velkov, K.

2003-01-01

The OECD/NRC BWR Turbine Trip (TT) Benchmark is investigated to perform code-to-code comparison of coupled codes including a comparison to measured data which are available from turbine trip experiments at Peach Bottom 2. This Benchmark problem for a BWR over-pressure transient represents a challenging application of coupled codes which integrate 3-dimensional neutron kinetics into thermal-hydraulic system codes for best-estimate simulation of plant transients. This transient represents a typical application of coupled codes which are usually performed on powerful workstations using a single CPU. Nowadays, the availability of multi-CPUs is much easier. Indeed, powerful workstations already provide 4 to 8 CPU, computer centers give access to multi-processor systems with numbers of CPUs in the order of 16 up to several 100. Therefore, the performance of the coupled code Athlet-Quabox/Cubbox on multi-processor systems is studied. Different cases of application lead to changing requirements of the code efficiency, because the amount of computer time spent in different parts of the code is varying. This paper presents main results of the coupled code Athlet-Quabox/Cubbox for the extreme scenarios of the BWR TT Benchmark together with evaluations of the code performance on multi-processor computers. (authors)

Calculation of neutron fluxes in biological shield of the TRIGA Mark II reactor

International Nuclear Information System (INIS)

Bozic, M.; Zagar, T.; Ravnik, M.

2001-01-01

The complete calculation of neutron fluxes in biological shield and verification with experimental results is presented. Calculated results are obtained with TORT code (TORT-Three Dimensional Oak Ridge Discrete Ordinates Neutron/Photon Transport Code). Experimental results used for comparison are available from irradiation experiment with selected type of concrete and other materials in irradiation channel 4 in TRIGA Mark II reactor. These experimental results were used as a benchmark. Homogeneous type of problem (without inserted irradiation channel) and problem with asymmetry (inserted beam port 4, filled with different materials) were of interest for neutron flux calculation. Deviation from material data set up as original parameters is also considered (first of all presence of water in concrete and density of concrete) for type of concrete in biological shield and for selected type of concrete in irradiation channel. BUGLE-96 (47 neutron energy groups) library is used. Excellent agreement between calculated and experimental results for reaction rate is received.(author)

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

International Nuclear Information System (INIS)

Pantoja C, R.

2010-01-01

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

Strategies of operation cycles in BWR type reactors

International Nuclear Information System (INIS)

Molina, D.; Sendino, F.

1996-01-01

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

Simulation of a large break loss of coolant (LBLOCA), without actuation of the emergency injection systems (ECCS) for a BWR-5; Simulacion de un escenario de perdida de refrigerante grande (LBLOCA), sin actuacion de los sistemas de inyeccion de emergencia (ECCS) para un reactor BWR-5

Energy Technology Data Exchange (ETDEWEB)

Cardenas V, J.; Mugica R, C. A.; Lopez M, R., E-mail: jaime.cardenas@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col. Narvarte, 03020 Ciudad de Mexico (Mexico)

2015-09-15

In this paper the analysis of scenario for the loss of coolant case was realized with break at the bottom of a recirculation loop of a BWR-5 with containment type Mark II and a thermal power of 2317 MWt considering that not have coolant injection. This in order to observe the speed of progression of the accident, the phenomenology of the scenario, the time to reach the limit pressure of containment venting and the amount of radionuclides released into the environment. This simulation was performed using the MELCOR code version 2.1. The scenario posits a break in one of the shear recirculation loops. The emergency core cooling system (ECCS) and the reactor core isolation cooling (Rcic) have not credit throughout the event, which allowed achieve greater severity on scenario. The venting of the primary containment was conducted via valve of 30 inches instead of the line of 24 inches of wet well, this in order to have a larger area of exhaust of fission products directly to the reactor building. The venting took place when the pressure in the primary containment reached the 4.5 kg/cm{sup 2} and remained open for the rest of the scenario to maximize the amount released of radionuclides to the atmosphere. The safety relief valves were considered functional they do not present mechanical failure or limit their ability to release pressure due to the large number of performances in safety mode. The results of the analysis covers about 48 hours, time at which the accident evolution was observed; behavior of level, pressure in the vessel and the fuel temperature profile was analyzed. For progression of the scenario outside the vessel, the pressure and temperature of the primary containment, level and temperature of the suppression pool, the hydrogen accumulation in the container and the radionuclides mass released into the atmosphere were analyzed. (Author)

Improvement for BWR operator training, 3

International Nuclear Information System (INIS)

Noji, Kunio; Toeda, Susumu; Saito, Genhachi; Suzuki, Koichi

1990-01-01

BWR Operator Training Center Corporation (BTC) is conducting training for BWR plant operators using Full-scope Simulators. There are several courses for individual operators and one training course for shift crew (Family Training Course) in BTC. Family Training is carried out by all members of the operating shift-crew. BTC has made efforts to improve the Family Training in order to acquire more effective training results and contribute to up-grade team performance of all crews. This paper describes some items of our efforts towards Family Training improvement. (author)

Assessment of two BWR accident management strategies

International Nuclear Information System (INIS)

Hodge, S.A.; Petek, M.

1991-01-01

Candidate mitigative strategies for management of in-vessel events during the late phase (after core degradation has occurred) of postulated BWR severe accidents were considered at Oak Ridge National Laboratory (ORNL) during 1990. The identification of new strategies was subject to the constraint that they should, to the maximum extent possible, make use of the existing equipment and water resources of the BWR facilities and not require major equipment modifications or additions. As a result of this effort, two of these candidate strategies were recommended for additional assessment. The first is a strategy for containment flooding to maintain the core and structural debris within the reactor vessel in the event that vessel injection cannot be restored to terminate a severe accident sequence. The second strategy pertains to the opposite case, for which vessel injection would be restored after control blade melting had begun; its purpose is to provide an injection source of borated water at the concentration necessary to preclude criticality upon recovering a damaged BWR core. Assessments of these two strategies have been performed during 1991 under the auspices of the Detailed Assessment of BWR In-Vessel Strategies Program. This paper provides a discussion of the motivation for and purpose of these strategies and the potential for their success. 33 refs., 9 figs

Signal analysis of acoustic and flow-induced vibrations of BWR main steam line

Energy Technology Data Exchange (ETDEWEB)

Espinosa-Paredes, G., E-mail: gepe@xanum.uam.mx [División de Ciencias Básicas e Ingeniería, Universidad Autónoma Metropolitana-Iztapalapa, México, D.F. 09340 (Mexico); Prieto-Guerrero, A. [División de Ciencias Básicas e Ingeniería, Universidad Autónoma Metropolitana-Iztapalapa, México, D.F. 09340 (Mexico); Núñez-Carrera, A. [Comisión Nacional de Seguridad Nuclear y Salvaguardias, Doctor Barragán 779, Col. Narvarte, México, D.F. 03020 (Mexico); Vázquez-Rodríguez, A. [División de Ciencias Básicas e Ingeniería, Universidad Autónoma Metropolitana-Iztapalapa, México, D.F. 09340 (Mexico); Centeno-Pérez, J. [Instituto Politécnico Nacional, Escuela Superior de Física y Matemáticas Unidad Profesional “Adolfo López Mateos”, Av. IPN, s/n, México, D.F. 07738 (Mexico); Espinosa-Martínez, E.-G. [Departamento de Sistemas Energéticos, Universidad Nacional Autónoma de México, México, D.F. 04510 (Mexico); and others

2016-05-15

Highlights: • Acoustic and flow-induced vibrations of BWR are analyzed. • BWR performance after extended power uprate is considered. • Effect of acoustic side branches (ASB) is analyzed. • The ASB represents a reduction in the acoustic loads to the steam dryer. • Methodology developed for simultaneous analyzing the signals in the MSL. - Abstract: The aim of this work is the signal analysis of acoustic waves due to phenomenon known as singing in Safety Relief Valves (SRV) of the main steam lines (MSL) in a typical BWR5. The acoustic resonance in SRV standpipes and fluctuating pressure is propagated from SRV to the dryer through the MSL. The signals are analyzed with a novel method based on the Multivariate Empirical Mode Decomposition (M-EMD). The M-EMD algorithm has the potential to find common oscillatory modes (IMF) within multivariate data. Based on this fact, we implement the M-EMD technique to find the oscillatory mode in BWR considering the measurements obtained collected by the strain gauges located around the MSL. These IMF, analyzed simultaneously in time, allow obtaining an estimation of the effects of the multiple-SRV in the MSL. Two scenarios are analyzed: the first is the signal obtained before the installation of the acoustic dampers (ASB), and the second, the signal obtained after installation. The results show the effectiveness of the ASB to damp the strong resonances when the steam flow increases, which represents an important reduction in the acoustic loads to the steam dryer.