A major objective in reactor design is to provide the capability to withstand a wide range of postulated events without exceeding specified fuel cladding temperatures; thereby maintaining fuel integrity and, ultimately, limiting fission product releases to the environment. Such events include small, medium, and large break loss- of-coolant accidents, and operational transients involving loss-off-inventory, multiple system failure, and power transients. After the TMI and Chernobyl accidents, research emphasis shifted toward analyzing the behavior of the plant during events involving extensive interplay between the various plant processes. Often, this analysis required the development of sophisticated computer programs to accurately model the neutronic and thermal-hydraulic behavior these highly coupled systems. The following programs, funded jointly by the NRC, EPRI, and GE, investigated such phenomena.
BWR Blowdown Heat Transfer Program (BDHT): 1972-1975
This program investigated heat transfer and system response during the blowdown phase of the LOCA using the Two-Loop Test Apparatus (TLTA). The TLTA was a full-pressure scale model of a BWR using a single, full- size, electrically simulated fuel bundle. This facility was scaled from a BWR/4 system with a 7x7 fuel bundle. The tests investigated thermal-hydraulic response only through the blowdown phase and did not include the injection of simulated emergency core coolant.
BWR Refill/Reflood Program: 1979-1983
This program expanded the focus of LOCA research to include full scale experimental evaluations of multidimensional and multichannel affects during system refill. In addition, the program experimentally investigated reflood heat transfer and distribution of ECC spray above the core. The program also made major contributions to the development of the TRAC-BWR thermal hydraulic code.
Full Integral Simulation Test (FIST): 1981-1985
This program used the TLTA to investigate system thermal hydraulic and bundle heat transfer responses over a wide range of simulated BWR LOCA conditions. It allowed more realistic simulation of LOCA's from break initiation through reflood as well as simulation of transient events involving loss of inventory multiple system failures, and power transients, than either of the previous two tests. The FIST Program test results were not intended to directly provide a demonstration of BWR performance. The primary use of the data related to assessment of advanced best estimate methods (computer models - TRAC-BWR) for predicting BWR response.
This was a test to collect data on choked flow phenomena during the blowdown phase of a transient. The blowdown phase of a transient refers to the first stage of a loss of coolant accident and since this is believed \ to be a credible transient data was needed on the effects induced during this part of the transient. It was performed to establish choked flow rate data for a large scale nozzle with subcooled and low quality water conditions at the nozzle inlet.
The development and assessment of large computer programs capable of predicting both local and system-wide behavior of a nuclear reactor under normal, transient and accident conditions has played a significant role in U.S. reactor safety research throughout the 70's, 80's, and 90's. Several of the programs outlined here have played major roles in this effort as sources of experimental data upon which code results could be compared to ensure that they were working properly. Some of the largest and most widely used codes are
RELAP - a light water reactor transient analysis code developed for use in rulemaking, licensing audit calculations, evaluation of operator guidelines, and as a basis for a nuclear plant analyzer. (3) It is a generic pressurized water reactor (PWR) code capable of modeling: large and small break loss-of-coolant accidents, operational transients, and transients in which the entire balance-of-plant must be modeled. (3) Developed at INEL under sponsorship of the NRC.
TRAC-PWR - an advanced best-estimate systems code designed primarily for analysis of large-break loss-of-coolant accidents in pressurized water reactors although its versatility allows for the analysis of a wide range of scenarios. Developed at Los Alamos National Laboratory under sponsorship of the Nuclear Regulatory Commission.
TRAC-BWR - an advanced best-estimate systems code designed primarily for analysis of large-break loss-of coolant accidents in boiling water reactors although its versatility allows for the analysis of a wide range of scenarios. Developed at Idaho National Engineering Laboratories under sponsorship of the Nuclear Regulatory Commission.
RAMONA - Developed at Brookhaven National Laboratory under the sponsorship of the NRC for analyzing BWR system transients. Until recently, RAMONA was the only best-estimate BWR system transient code capable of predicting three dimensional power in the core coupled with the fuel and cladding temperature and vessel thermal hydraulic phenomena.
RETRAN - a best estimate transient thermal hydraulic analysis computer program (sponsored by EPRI) designed to provide analysis capabilities for BWR and PWR transients, small break loss of coolant accidents, balance of plant modeling, and anticipated transients without scram.
Following the completion of the Reactor Safety Study, the NRC initiated research programs to improve the staff's ability to assess the risks of severe accidents in light water reactors. Development began on advanced methods for assessing the frequencies of accidents. Improved means for the collection and use of plant operational data were put into place, and advanced methods for assessing the impacts of human errors and other common-cause failures were developed. In addition, research was begun on key severe accident physical processes identified in the WASH-1400 report, such as the interaction of molten core material with concrete.
By the mid-1980's, the technology for analyzing the physical processes of severe accidents had evolved to the point that a new computational model of severe accident physical processes had been developed - the Source Term Code Package - and subjected to peer review. General procedures for performing PRAs were also developed.
The NUREG-1150 report represents one element of the NRC's effort to close the book on severe accident issues on the set of U.S. operating nuclear power plants. It updates the work summarized by WASH-1400 and provides a "snapshot" (in time) of the estimated plant risks for five commercial nuclear power plants of different design.
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