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Licensee contractor and vendor inspection status report: Quarterly report, January--March 1997. Volume 21, Number 1

Description: A fundamental premise of the US Nuclear Regulatory Commission (NRC) licensing and inspection program is that licensees are responsible for the proper construction and safe and efficient operation of their nuclear power plants. The Federal government and nuclear industry have established a system for the inspection of commercial nuclear facilities to provide for multiple levels of inspection and verification. Each licensee, contractor, and vendor participates in a quality verification process in compliance with requirements prescribed by the NRC`s rules and regulations (Title 10 of the Code of Federal Regulations). The NRC does inspections to oversee the commercial nuclear industry to determine whether its requirements are being met by licensees and their contractors, while the major inspection effort is performed by the industry within the framework of quality verification programs. This periodical covers the results of inspections performed by the NRC`s Special Inspection Branch, Vendor Inspection Section, that have been distributed to the inspected organizations during the period from January 1997 through March 1997.
Date: July 1997
Partner: UNT Libraries Government Documents Department

Human factors review for nuclear power plant severe accident sequence analysis

Description: The paper discusses work conducted to: (1) support the severe accident sequence analysis of a nuclear power plant transient based on an assessment of operator actions, and (2) develop a descriptive model of operator severe accident management. Operator actions during the transient are assessed using qualitative and quantitative methods. A function-oriented accident management model provides a structure for developing technical operator guidance on mitigating core damage preventing radiological release.
Date: January 1, 1985
Creator: Krois, P.A. & Haas, P.M.
Partner: UNT Libraries Government Documents Department

Importance of momentum dynamics in BWR neutronic stability: experimental evidence

Description: Momentum dynamics affect the boiling water reactor (BWR) neutronic stability by coupling steam void perturbations and core-inlet coolant flow. Computer simulations have shown that proper modeling of the recirculation loop, which shares the upper and lower plena pressures with the reactor core, is essential for accurate stability calculations. Purpose of this paper is to show experimental evidence, obtained from a recent series of stability tests performed at the Browns Ferry-1 BWR, demonstrating the important role of momentum dynamics in BWR neutronic stability.
Date: January 1, 1985
Creator: March-Leuba, J. & Otaduy, P.J.
Partner: UNT Libraries Government Documents Department

Interim reliability evaluation program, Browns Ferry 1

Description: Probabilistic risk analysis techniques, i.e., event tree and fault tree analysis, were utilized to provide a risk assessment of the Browns Ferry Nuclear Plant Unit 1. Browns Ferry 1 is a General Electric boiling water reactor of the BWR 4 product line with a Mark 1 (drywell and torus) containment. Within the guidelines of the IREP Procedure and Schedule Guide, dominant accident sequences that contribute to public health and safety risks were identified and grouped according to release categories.
Date: January 1, 1981
Creator: Mays, S.E.; Poloski, J.P.; Sullivan, W.H.; Trainer, J.E.; Bertucio, R.C. & Leahy, T.J.
Partner: UNT Libraries Government Documents Department

Recent SCDAP/RELAP5 improvements for BWR severe accident simulations

Description: A new model for the SCDAP/RELAP5 severe accident analysis code that represents the control blade and channel box structures in a boiling water reactor (BWR) has been under development since 1991. This model accounts for oxidation, melting, and relocation of these structures, including the effects of material interactions between B{sub 4}C, stainless steel, and Zircaloy. This paper describes improvements that have been made to the BWR control blade/channel box model during 1994 and 1995. These improvements include new capabilities that represent the relocation of molten material in a more realistic manner and modifications that improve the usability of the code by reducing the frequency of code failures. This paper also describes a SCDAP/RELAP5 assessment calculation for the Browns Ferry Nuclear Plant design based upon a short-term station blackout accident sequence.
Date: December 31, 1995
Creator: Griffin, F.P.
Partner: UNT Libraries Government Documents Department

RAMONA-3B application to Browns Ferry ATWS

Description: This paper discusses two preliminary MSIV clsoure ATWS calculations done using the RAMONA-3B code and the work being done to create the necessary cross section sets for the Browns Ferry Unit 1 reactor. The RAMONA-3B code employs a three-dimensional neutron kinetics model coupled with one-dimensional, four equation, nonhomogeneous, nonequilibrium thermal hydraulics. To be compatible with 3-D neutron kinetics, the code uses parallel coolant channels in the core. It also includes a boron transport model and all necessary BWR components such as jet pump, recirculation pump, steam separator, steamline with safety and relief valves, main steam isolation valve, turbine stop valve, and turbine bypass valve. A summary of RAMONA-3B neutron kinetics and thermal hydraulics models is presented in the Appendix.
Date: January 1, 1984
Creator: Slovik, G.C.; Neymotin, L.; Cazzoli, E. & Saha, P.
Partner: UNT Libraries Government Documents Department

Potential effects of the fire protection system sprays at Browns Ferry on fission product transport

Description: The fire protection system (FPS) sprays within any nuclear plant are not intended to mitigate radioactive releases to the environment resulting from severe core-damage accidents. However, it has been shown here that during certain postulated severe accident scenarios at the Browns Ferry Nuclear Plant, the functioning of FPS sprays could have a significant impact on the radioactive releases. Thus the effects of those sprays need to be taken into account for realistic estimation of source terms for some accident scenarios. The effects would include direct ones such as cooling of the reactor building atmosphere and scrubbing of radioactivity from it, as well as indirect effects such as an altered likelihood of hydrogen burning and flooding of various safety-related pumps in the reactor building basement. Thus some of the impacts of the sprays would be beneficial with respect to mitigating releases to the environment but some others might not be. The effects of the FPS would be very scenario dependent with a wide range of potential effects often existing for a given accident sequence. Any generalization of the specific results presented here for Browns Ferry to other nuclear plants must be done cautiously, as it appears from a preliminary investigation that the relevant physical and operational characteristics of FPS spray systems differ widely among even otherwise apparently similar plants. Likewise the standby gas treatment systems, which substantially impact the effects of the FPS, differ significantly among plants. More work for both Mark I plants and other plants, BWRs and PWRs alike, is indicated so the potential effects of FPS spray systems during severe accidents can be at least ball-parked for more realistic accident analyses.
Date: January 1, 1983
Creator: Niemczyk, S.J.
Partner: UNT Libraries Government Documents Department

BWR severe accident sequence analyses at ORNL - some lessons learned

Description: Boiling water reactor severe accident sequence studies are being carried out using Browns Ferry Unit 1 as the model plant. Four accident studies were completed, resulting in recommendations for improvements in system design, emergency procedures, and operator training. Computer code improvements were an important by-product.
Date: January 1, 1983
Creator: Hodge, S.A.
Partner: UNT Libraries Government Documents Department

Analysis of loss of decay-heat-removal sequences at Browns Ferry Unit One

Description: This paper summarizes the Oak Ridge National Laboratory (ORNL) report Loss of DHR Sequences at Browns Ferry Unit One - Accident Sequence Analysis (NUREG/CR-2973). The Loss of DHR investigation is the third in a series of accident studies concerning the BWR 4 - MK I containment plant design. These studies, sponsored by the Nuclear Regulatory Commission Severe Accident Sequence Analysis (SASA) program, have been conducted at ORNL with the full cooperation of the Tennessee Valley Authority (TVA). The purpose of the SASA studies is to predetermine the probable course of postulated severe accidents so as to establish the timing and the sequence of events. The SASA studies also produce recommendations concerning the implementation of better system design and better emergency operating instructions and operator training. The ORNL studies also include a detailed, best-estimate calculation of the release and transport of radioactive fission products following postulated severe accidents.
Date: January 1, 1983
Creator: Harrington, R.M.
Partner: UNT Libraries Government Documents Department

Interim reliability evaluation program, Browns Ferry fault trees

Description: An abbreviated fault tree method is used to evaluate and model Browns Ferry systems in the Interim Reliability Evaluation programs, simplifying the recording and displaying of events, yet maintaining the system of identifying faults. The level of investigation is not changed. The analytical thought process inherent in the conventional method is not compromised. But the abbreviated method takes less time, and the fault modes are much more visible.
Date: January 1, 1981
Creator: Stewart, M.E.
Partner: UNT Libraries Government Documents Department

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

Description: 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.
Date: September 1, 1986
Creator: Bennett, P.R.; Kolaczkowski, A.M. & Medford, G.T.
Partner: UNT Libraries Government Documents Department

Scram discharge volume break studies accident sequence analysis

Description: This paper is a summary of a report describing the predicted response of Unit 1 at the Tennessee Valley Authority (TVA) Browns Ferry Nuclear Plant to a hypothetical small break loss of coolant accident (SBLOCA) outside of containment. The accident studied would be initiated by a break in the scram discharge volume (SDV) piping when it is pressurized to full reactor vessel pressure as a normal consequence of a reactor scram. If the scram could be reset, the scram outlet valves would close to isolate the SDV and the piping break from the reactor vessel. However, reset is possible only if the conditions that caused the scram have cleared; it has been assumed in this study that the scram signal remains in effect over a long period of time.
Date: January 1, 1982
Creator: Harrington, R.M. & Hodge, S.A.
Partner: UNT Libraries Government Documents Department

Fission product transport analysis in a loss of decay heat removal accident at Browns Ferry

Description: This paper summarizes an analysis of the movement of noble gases, iodine, and cesium fission products within the Mark-I containment BWR reactor system represented by Browns Ferry Unit 1 during a postulated accident sequence initiated by a loss of decay heat removal (DHR) capability following a scram. The event analysis showed that this accident could be brought under control by various means, but the sequence with no operator action ultimately leads to containment (drywell) failure followed by loss of water from the reactor vessel, core degradation due to overheating, and reactor vessel failure with attendant movement of core debris onto the drywell floor.
Date: January 1, 1984
Creator: Wichner, R.P.; Weber, C.F.; Hodge, S.A.; Beahm, E.C. & Wright, A.L.
Partner: UNT Libraries Government Documents Department

Nonlinear dynamics and chaos in boiling water reactors

Description: There are currently 72 commercial boiling water reactors (BWRs) in operation or under construction in the western world, 37 of them in the United States. Consequently, a great effort has been devoted to the study of BWR systems under a wide range of plant operating conditions. This paper represents a contribution to this ongoing effort; its objective is to study the basic dynamic processes in BWR systems, with special emphasis on the physical interpretation of BWR dynamics. The main thrust in this work is the development of phenomenological BWR models suited for analytical studies performed in conjunction with numerical calculations. This approach leads to a deeper understanding of BWR dynamics and facilitates the interpretation of numerical results given by currently available sophisticated BWR codes. 6 refs., 14 figs., 2 tabs.
Date: January 1, 1988
Creator: March-Leuba, J.
Partner: UNT Libraries Government Documents Department

Station blackout calculations for Browns Ferry

Description: This paper presents the results of calculations performed with the ORNL SASA code suite for the Station Blackout Severe Accident Sequence at Browns Ferry. The accident is initiated by a loss of offsite power combined with failure of all onsite emergency diesel generators to start and load. The Station Blackout is assumed to persist beyond the point of battery exhaustion (at six hours) and without DC power, cooling water could no longer be injected into the reactor vessel. Calculations are continued through the period of core degradation and melting, reactor vessel failure, and the subsequent containment failure. An estimate of the magnitude and timing of the concomitant fission product releases is also provided.
Date: January 1, 1985
Creator: Ott, L.J.; Weber, C.F. & Hyman, C.R.
Partner: UNT Libraries Government Documents Department

Interfacing systems LOCAs (Loss of Coolant Accidents) at boiling water reactors

Description: The work presented in this paper was performed by Brookhaven National Laboratory (BNL) in support of Nuclear Regulatory Commission's (NRC) effort towards the resolution of Generic Issue 105 ''Interfacing System Loss of Coolant Accidents (LOCAs) at Boiling Water Reactors (BWRs).'' For BWRs, intersystem LOCA have typically either not been considered in probabilistic risk analyses, or if considered, were judged to contribute little to the risk estimates because of their perceived low frequency of occurrence. However, recent operating experience indicates that the pressure isolation valves (PIVs) in BWRs may not adequately protect against overpressurization of low pressure systems. The objective of this paper is to present the results of a study which analyzed interfacing system LOCA at several BWRs. The BWRs were selected to best represent a spectrum of BWRs in service using industry operating event experience and plant-specific information/configurations. The results presented here include some possible changes in test requirements/practices as well as an evaluation of their reduction potential in terms of core damage frequency (CDF).
Date: January 1, 1987
Creator: Chu, Tsong-Lun; Fitzpatrick, R. & Stoyanov, S.
Partner: UNT Libraries Government Documents Department

Browns Ferry waste heat greenhouse environmental control system design

Description: Oak Ridge National Laboratory, Tennessee Valley Authority and the Environmental Research Laboratory at the University of Arizona cooperated on the design of an experimental greenhouse located at TVA's Browns Ferry Nuclear Generating Station. Two greenhouse zones are heated by waste heat from the plant's condenser effluent. For comparison, a third greenhouse zone is heated conventionally (fossil-fueled burners) as a control. Design specifics for each of the three zones and a qualitative operating evaluation are presented.
Date: March 1, 1980
Creator: Olszewski, M.; Stovall, T.K.; Hicks, N.G.; Pile, R.S.; Burns, E.R. & Waddell, E.L. Jr.
Partner: UNT Libraries Government Documents Department

Severe accident testing of electrical penetration assemblies

Description: This report describes the results of tests conducted on three different designs of full-size electrical penetration assemblies (EPAs) that are used in the containment buildings of nuclear power plants. The objective of the tests was to evaluate the behavior of the EPAs under simulated severe accident conditions using steam at elevated temperature and pressure. Leakage, temperature, and cable insulation resistance were monitored throughout the tests. Nuclear-qualified EPAs were produced from D. G. O'Brien, Westinghouse, and Conax. Severe-accident-sequence analysis was used to generate the severe accident conditions (SAC) for a large dry pressurized-water reactor (PWR), a boiling-water reactor (BWR) Mark I drywell, and a BWR Mark III wetwell. Based on a survey conducted by Sandia, each EPA was matched with the severe accident conditions for a specific reactor type. This included the type of containment that a particular EPA design was used in most frequently. Thus, the D. G. O'Brien EPA was chosen for the PWR SAC test, the Westinghouse was chosen for the Mark III test, and the Conax was chosen for the Mark I test. The EPAs were radiation and thermal aged to simulate the effects of a 40-year service life and loss-of-coolant accident (LOCA) before the SAC tests were conducted. The design, test preparations, conduct of the severe accident test, experimental results, posttest observations, and conclusions about the integrity and electrical performance of each EPA tested in this program are described in this report. In general, the leak integrity of the EPAs tested in this program was not compromised by severe accident loads. However, there was significant degradation in the insulation resistance of the cables, which could affect the electrical performance of equipment and devices inside containment at some point during the progression of a severe accident. 10 refs., 165 figs., 16 tabs.
Date: November 1, 1989
Creator: Clauss, D.B. (Sandia National Labs., Albuquerque, NM (USA))
Partner: UNT Libraries Government Documents Department