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in subjects: "210100 -- Power Reactors, Nonbreeding, Light-Water Moderated, Boiling Water Cooled"

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Assessment of light water reactor fuel damage during a reactivity initiated accident

This paper presents an assessment of LWR fuel damage during a reactivity initiated accident and comments on the adequacy of the present USNRC design requirements. Results from early SPERT tests are reviewed and compared with results from recent computer simulations and PBF tests. A progression of fuel rod and cladding damage events is presented. High strain rate deformation of relatively cool irradiated cladding early in the transient may result in fracture at a radial average peak fuel enthalp… more

Date: January 1, 1980

Creator: MacDonald, P. E.; Seiffert, S. L.; Martinson, Z. R.; McCardell, R. K.; Owen, D. E. & Fukuda, S. K.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Nuclear power in the Soviet Bloc

The growth of Soviet Bloc nuclear power generation to the end of the century is evaluated on the basis of policy statements of objectives, past and current nuclear power plant construction, and trends in the potential for future construction. Central to this study is a detailed examination of individual reactor construction and site development that provides specific performance data not given elsewhere. A major commitment to nuclear power is abundantly clear and an expansion of ten times in nu… more

Date: March 1, 1982

Creator: Davey, W.G.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Health effects models for nuclear power plant accident consequence analysis: Low LET radiation

This report describes dose-response models intended to be used in estimating the radiological health effects of nuclear power plant accidents. Models of early and continuing effects, cancers and thyroid nodules, and genetic effects are provided. Weibull dose-response functions are recommended for evaluating the risks of early and continuing health effects. Three potentially lethal early effects -- the hematopoietic, pulmonary, and gastrointestinal syndromes -- are considered. In addition, model… more

Date: January 1, 1990

Creator: Evans, J. S.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays. [BWR; PWR]

The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-… more

Date: February 1, 1978

Creator: Benedetti, R. L.; Lords, L. V. & Kiser, D. M.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Nuclear plant-aging research on reactor protection systems

This report presents the rsults of a review of the Reactor Trip System (RTS) and the Engineered Safety Feature Actuating System (ESFAS) operating experiences reported in Licensee Event Reports (LER)s, the Nuclear Power Experience data base, Nuclear Plant Reliability Data System, and plant maintenance records. Our purpose is to evaluate the potential significance of aging, including cycling, trips, and testing as contributors to degradation of the RTS and ESFAS. Tables are presented that show th… more

Date: January 1, 1988

Creator: Meyer, L.C.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Commercial US nuclear reactors and waste: the current status

Between March 1 and June 15, 1980, the declared size of the commercial light waste reactor (LWR) nuclear power industry in the US has decreased another 9 GWe. For the presently declared size: the 165 declared reactors will peak at a capacity of 153 GWe in 2001 and will consume about 870,000 MTU as enrichment feed; the theoretical rate of enrichment requirements will peak at about 19,000,000 SWUs/y in the year 2014; as few as two repositories each with capacity equivalent to 100,000 MTU would ho… more

Date: September 1, 1980

Creator: Platt, A.M. & Robinson, J.V.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Fission product release from high gap-inventory LWR fuel under LOCA conditions

Fission product release tests were performed with light water reactor (LWR) fuel rod segments containing large amounts of cesium and iodine in the pellet-to-cladding gap space in order to check the validity of the previously published Source Term Model for this type of fuel. The model describes the release of fission product cesium and iodine from LWR fuel rods for controlled loss-of-coolant accident (LOCA) transients in the temperature range 500 to 1200/sup 0/C. The basis for the model was tes… more

Date: January 1, 1980

Creator: Lorenz, R. A.; Collins, J. L.; Osborne, M. F. & Malinauskas, A. P.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Quarterly Progress Report on Fission Product Behavior in LWRs for the Period October-December 1977

Analysis of release data obtained during High Burnup Fuel Test 10 (HBU-10) has been completed. In this test the fuel rod segment was ruptured by internal pressurization at 900/sup 0/C, at which time the temperature was rapidly increased to 1200/sup 0/C and maintained at this temperature for 10 min. Approximately 0.061% of the total cesium inventory in the rod segment was released; this was accompanied by the release of about 1.69% of the total /sup 85/Kr inventory. Moreover, about 0.022% of the… more

Date: February 1, 1978

Creator: Malinauskas, A. P.; Lorenz, R. A.; Collins, J. L.; Osborne, M. F.; Whatley, S. K. & Towns, R. L.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Thermal reactor safety

Information is presented concerning new trends in licensing; seismic considerations and system structural behavior; TMI-2 risk assessment and thermal hydraulics; statistical assessment of potential accidents and verification of computational methods; issues with respect to improved safety; human factors in nuclear power plant operation; diagnostics and activities in support of recovery; LOCA transient analysis; unresolved safety issues and other safety considerations; and fission product transp… more

Date: June 1, 1980

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Reactor containment analysis for BWR suppression systems. Progress report, April--October 1977

Work during the period included analytic studies of vent clearing and jet-slug dynamics and computer simulations (HEMP and CHAMP codes) of pressure suppression experiments.

Date: December 14, 1977

Creator: Edwards, L.L.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Thermal Reactor Safety

Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

Date: June 1, 1980

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Results of two-phase natural circulation in hot-leg U-bend simulation experiments

In order to study the two-phase natural circulation and flow termination during a small break loss of coolant accident in LWR, simulation experiments have been performed using two different thermal-hydraulic loops. The main focus of the experiment was the two-phase flow behavior in the hot-leg U-bend typical of BandW LWR systems. The first group of experiments was carried out in the nitrogen gas-water adiabatic simulation loop and the second in the Freon 113 boiling and condensation loop. Both … more

Date: January 1, 1987

Creator: Ishii, M.; Lee, S. Y. & Abou El-Seoud, S.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Aging assessment of BWR control rod drive systems

This study examines the aging phenomena associated with boiling water reactor (BWR) control rod drive mechanisms (CRDMs) and assess the merits of various methods of managing this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of Nuclear Plant Reliability Data System (NPRDS) failure cases attributed to… more

Date: January 1, 1991

Creator: Greene, R.H.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Physical Model of Lean Suppression Pressure Oscillation Phenomena: Steam Condensation in the Light Water Reactor Pressure Suppression System (PSS)

Using the results of large scale multivent tests conducted by GKSS, a physical model of chugging is developed. The unique combination of accurate digital data and cinematic data has provided the derivation of a detailed, quantified correlation between the dynamic physical variables and the associated two-phase thermo-hydraulic phenomena occurring during lean suppression (chugging) phases of the loss-of-coolant accident in a boiling water reactor pressure suppression system.

Date: April 1, 1980

Creator: McCauley, E. W.; Holman, G. S.; Aust, E.; Schwan, H. & Vollbrandt, J.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Argonne Code Center: benchmark problem book

This report is a supplement to the original report, published in 1968, as revised. The Benchmark Problem Book is intended to serve as a source book of solutions to mathematically well-defined problems for which either analytical or very accurate approximate solutions are known. This supplement contains problems in eight new areas: two-dimensional (R-z) reactor model; multidimensional (Hex-z) HTGR model; PWR thermal hydraulics--flow between two channels with different heat fluxes; multidimension… more

Date: June 1, 1977

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Experimental results of direct containment heating by high-pressure melt ejection into the Surtsey vessel: The DCH-3 and DCH-4 tests

Two experiments, DCH-3 and DCH-4, were performed at the Surtsey test facility to investigate phenomena associated with a high-pressure melt ejection (HPME) reactor accident sequence resulting in direct containment heating (DCH). These experiments were performed using the same experimental apparatus with identical initial conditions, except that the Surtsey test vessel contained air in DCH-3 and argon in DCH-4. Inerting the vessel with argon eliminated chemical reactions between metallic debris … more

Date: August 1, 1991

Creator: Allen, M.D.; Pilch, M.; Brockmann, J.E.; Tarbell, W.W. (Sandia National Labs., Albuquerque, NM (United States)); Nichols, R.T. (Ktech Corp., Albuquerque, NM (United States)) & Sweet, D.W. (AEA Technology, Winfrith (United Kingdom))

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Comparative assessment of nuclear fuel cycles. Light-water reactor once-through, classical fast breeder reactor, and symbiotic fast breeder reactor cycles

The object of the Alternative Nuclear Fuel Cycle Study is to perform comparative assessments of nuclear power systems. There are two important features of this study. First, this evaluation attempts to encompass the complete, integrated fuel cycle from mining of uranium ore to disposal of waste rather than isolated components. Second, it compares several aspects of each cycle - energy use, economics, technological status, proliferation, public safety, and commercial potential - instead of conce… more

Date: June 1, 1980

Creator: Hardie, R.W.; Barrett, R.J. & Freiwald, J.G.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Transformer failure and common-mode loss of instrument power at Nine Mile Point Unit 2 on August 13, 1991

On August 13, 1991, at Nine Mile Point Unit 2 nuclear power plant, located near Scriba, New York, on Lake Ontario, the main transformer experienced an internal failure that resulted in degraded voltage which caused the simultaneous loss of five uninterruptible power supplies, which in turn caused the loss of several nonsafety systems, including reactor control rod position indication, some reactor power and water indication, control room annunciators, the plant communications system, the plant … more

Date: October 1, 1991

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

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

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 /sup 1///sub 5/-scale Mark I BWR pressure-suppression system. Results from these tests are used to quantify the vertical-loading function and to study the as… more

Date: December 22, 1977

Creator: Pitts, J.H. & McCauley, E.W.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

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

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 t… more

Date: September 1, 1991

Creator: Lin, C. C. & Lehner, J. R.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Severe accident testing of electrical penetration assemblies

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. … more

Date: November 1, 1989

Creator: Clauss, D.B. (Sandia National Labs., Albuquerque, NM (USA))

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Finite-difference methods in multi-dimensional two-phase flow. [BWR; PWR; LMFBR]

In the summer of 1974, the Theoretical Division of the Los Alamos Scientific Laboratory began several research programs in the area of reactor safety for the United States Nuclear Regulatory Commission. Research efforts were started in the Liquid Metal Fast Breeder (LMFBR) and the Light Water Reactor (LWR) safety programs. The character of the Theoretical Division was to develop computer codes for the safety analysis of these reactor systems. The question of whether or not, during the course of… more

Date: January 1, 1977

Creator: Travis, J. R.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Reactor calculation benchmark PCA blind test results

Further improvement in calculational procedures or a combination of calculations and measurements is necessary to attain 10 to 15% (1 sigma) accuracy for neutron exposure parameters (flux greater than 0.1 MeV, flux greater than 1.0 MeV, and dpa). The calculational modeling of power reactors should be benchmarked in an actual LWR plant to provide final uncertainty estimates for end-of-life predictions and limitations for plant operations. 26 references, 14 figures, 6 tables.

Date: January 1, 1980

Creator: Kam, F.B.K. & Stallmann, F.W.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Dose reduction at nuclear power plants

The collective dose equivalent at nuclear power plants increased from 1250 rem in 1969 to nearly 54,000 rem in 1980. This rise is attributable primarily to an increase in nuclear generated power from 1289 MW-y to 29,155 MW-y; and secondly, to increased average plant age. However, considerable variation in exposure occurs from plant to plant depending on plant type, refueling, maintenance, etc. In order to understand the factors influencing these differences, an investigation was initiated to st… more

Date: January 1, 1983

Creator: Baum, J. W. & Dionne, B. J.

Item Type: Article

Partner: UNT Libraries Government Documents Department