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in subjects: "Pwr Type Reactors"

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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

De-entrainment on vertical elements in air droplet cross flow. [PWR]

De-entrainment phenomena on vertical elements in air-water droplet cross flow are generated using a horizontal array of water spray nozzles and a draft-induced wind tunnel. These conditions are used to obtain experimental values of the de-entrainment efficiency of isolated elements (25.4-, 63.5-, and 101.6-mm-diam cylinders and a 76.2-mm-square tube), and of an array of 101.6-mm-diam cylinders. A flow model is developed that extrapolates the de-entrainment efficiency of isolated elements throug… more

Date: January 1, 1980

Creator: Dallman, J.C. & Kirchner, W.L.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

NRC plan for cleanup operations at Three Mile Island Unit 2

This NRC Plan, which defines NRC's functional role in cleanup operations at Three Mile Island Unit 2 and outlines NRC's regulatory responsibilities in fulfilling this role, is the first revision to the initial plan issued in July 1980 (NUREG-0698). Since 1980, a number of policy developments have occurred which will have an impact on the course of cleanup operations. This revision reflects these developments in the area of NRC's review and approval process with regard to cleanup … more

Date: February 1, 1982

Creator: Lo, R. & Snyder, B.

Item Type: Report

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

Analysis of the results of the Midland PRA

This paper presents the results of a limited review of the Midland PRA (MPRA), aimed at facilitating regulatory analyses and at providing insights into safety related plant failures. In particular, stress was laid on the root causes of accident sequences, particularly their failure modes, viz., hardware, human, maintenance, test, and repair. Because this information was difficult to extract in this detailed and sophisticated PRA (and in certain others) a special algorithm was developed to displ… more

Date: January 1, 1986

Creator: Bozoki, G. & Teichmann, T.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Characterization of solids in the Three Mile Island Unit 2 Reactor defueling water: Addendum

Shortly after ORNL/TM-10362 was issued, it was discovered that a series of 31 figures had been inadvertently omitted. These figures, which consist of scanning electron microscope (SEM) photographs and energy-dispersive X-ray fluorescence scans, provide significant information about the results obtained in the tests performed with water sample W3. This Addendum includes these figures. Details of and comments on the SEM photographs may be found in ORNL/TM-10362.

Date: March 1, 1988

Creator: Campbell, D.O.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Thorium fuel cycles for LWRs: fuel diversion assessments and recycle requirements

A number of fuel cycles have been proposed for evaluation in the nonproliferation alternative systems assessment program. Among these systems are light water reactors (LWR) operating on the thorium-uranium cycle or the plutonium-thorium cycle either inside or outside energy centers. These proposals support the President's nuclear power policy of accelerating research into alternative fuel cycles that do not permit direct access to materials usable for nuclear weapons but still retain the benefi… more

Date: January 1, 1978

Creator: Carter, W. L.; Rainey, R. H. & Johnson, D. R.

Item Type: Article

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

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Calculations of thermal-reactor spent-fuel nuclide inventories and comparisons with measurements

Comparisons with integral measurements have demonstrated the accuracy of CINDER codes and libraries in calculating aggregate fission-product properties, including neutron absorption, decay power, and decay spectra. CINDER calculations have, alternatively, been used to supplement measured integral data describing fission-product decay power and decay spectra. Because of the incorporation of the extensive actinide library and the use of ENDF/B-V data, it is desirable to compare the inventory of i… more

Date: January 1, 1982

Creator: Wilson, W. B.; LaBauve, R. J. & England, T. R.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Integrity of PWR pressure vessels during overcooling accidents

The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, there is a class of postulated accidents, referred to as overcooling accidents, that can subject the pressure vessel to severe thermal shock while the pressure is substantial. As a result of such accidents vessels containing high concentrations of copper and nickel, which enhance… more

Date: January 1, 1982

Creator: Cheverton, R.D.; Iskander, S.K. & Whitman, G.D.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Environmental monitoring at major U. S. Energy Research and Development Administration contractor sites: calendar year 1976. Volume 2

The purpose of this compilation is to present, in a central reference document, all of the individual annual reports summarizing the results of the environmental monitoring programs conducted at each of the Energy Research and Development Administration (ERDA) sites having a potential for environmental impact or which release a significant quantity of radioactivity or nonradioactive pollutants. Data on the levels of radioactivity and nonradioactive pollutants in effluents and the environs at ea… more

Date: August 1, 1977

Item Type: Report

Partner: UNT Libraries Government Documents Department

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LOFT system structural response during subcooled blowdown

The Loss-of-Fluid Test (LOFT) facility is a highly instrumented, pressurized water reactor test system designed to be representative of large pressurized water reactors (LPWRs) for the simulation of loss-of-coolant accidents (LOCAs). Detailed structural analysis and appropriate instrumentation (accelerometers and strain gages) on the LOFT system provided information for evaluation of the structural response of the LOFT facility for loss-of-coolant experiment (LOCE) induced loads. In general, th… more

Date: January 1, 1978

Creator: Martinell, J. S.

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

open access

Once-through steam generator (OTSG) materials and water chemistry. [PWR]

Materials and water chemistry research results associated with the development of the Oconee-1 Reactor steam generator are presented. A summary of water chemistry data acquired during preoperational testing and power operation to date is also included. These data confirm the operational practicality of the nuclear once-through concept using volatile water treatment and high purity condensate demineralized feedwater.

Date: January 1, 1974

Creator: Pocock, F.J. & Levstek, D.F.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Insights from an overview of four PRAs

This paper summarizes the findings of an investigation of four probabilistic risk assessments (PRAs), those for Millstone 3, Seabrook, Shoreham, and Oconee 3, performed by Brookhaven National Laboratory (BNL) for the Reliability and Risk Assessment Branch of the US Nuclear Regulatory Commission (NRC). This group of four PRAs was subjected to an overview process with the basic goal of ascertaining what insights might be gained (beyond those already documented within the individual PRAs) by an in… more

Date: January 1, 1986

Creator: Fitzpatrick, R.; Arrieta, L.; Teichmann, T. & Davis, P.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Development of an extended-burnup Mark B design. First semi-annual progress report, July-December 1978. Report BAW-1532-1. [PWR]

The primary objective of this program is to develop and demonstrate an improved PWR fuel assembly design capable of batch average burnups of 45,000-50,000 MWd/mtU. To accomplish this, a number of technical areas must be investigated to verify acceptable extended-burnup fuel performance. This report is the first semi-annual progress report for the program, and it describes work performed during the July-December 1978 time period. Efforts during this period included the definition of a preliminar… more

Date: October 1, 1979

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Effect of reactor coolant pumps following a small break in a pressurized water reactor

Small-break loss-of-coolant accidents were calculated to help determine whether to trip the reactor-coolant pumps early in the accident when the reactor scrams or to delay the pump trip (pump trip times ranged from 450 s to no trip at all). Four-in.-diam (approximate) cold-leg breaks in Babcock and Wilcox (B and W) and Westinghouse (W) pressurized-water reactors were investigated using the Transient Reactor Analysis Code, TRAC-PD2. The results indicated that for a 4-in.-diam cold-leg break the … more

Date: January 1, 1982

Creator: Elliott, J.L.; Lime, J.F. & Willcutt, G.J.E. Jr.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Characterization of radioactive ion exchange media waste generated at Three Mile Island

The March 1979 accident at General Public Utilities Nuclear Corporation (GPUNC) Three Mile Island Nuclear Power Station Unit 2 (TMI-2), resulted in the transfer of more than 1100 m/sup 3/ of contaminated water to the auxiliary and fuel handling building. The principal sources of the water were the makeup and letdown purification system and the containment building sump. The contaminated water was processed through an ion exchange system designated as EPICOR II. The EPICOR-II System is a three-s… more

Date: October 1, 1981

Creator: Runion, T. C.; Holzworth, R. E.; Ogle, R. E.; Burton, H. M. & Bixby, W. W.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Analysis of LOFT pressurizer spray and surge nozzles to include a 450/sup 0/F step transient

This report presents the analysis of the LOFT pressurizer spray and surge nozzles to include a 450/sup 0/F step thermal transient. Previous analysis performed under subcontract by Basic Technology Incorporated was utilized where applicable. The SAASIII finite element computer program was used to determine stress distributions in the nozzles due to the step transient. Computer results were then incorporated in the necessary additional calculations to ascertain that stress limitations were not ex… more

Date: January 18, 1978

Creator: Nitzel, M.E.

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

PNL technical review of pressurized thermal-shock issues. [PWR]

Pacific Northwest Laboratory (PNL) was asked to develop and recommend a regulatory position that the Nuclear Regulatory Commission (NRC) should adopt regarding the ability of reactor pressure vessels to withstand the effects of pressurized thermal shock (PTS). Licensees of eight pressurized water reactors provided NRC with estimates of remaining effective full power years before corrective actions would be required to prevent an unsafe operating condition. PNL reviewed these responses and the r… more

Date: July 1, 1982

Creator: Pedersen, L. T.; Apley, W. J.; Bian, S. H.; Defferding, L. J.; Morgenstern, M. H.; Pelto, P. J. et al.

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

Method for fabricating a seal between a ceramic and a metal alloy

A method of fabricating a seal between a ceramic and an alloy comprising the steps of prefiring the alloy in an atmosphere with a very low partial pressure of oxygen, firing the assembled alloy and ceramic in air, and gradually cooling the fired assembly to avoid the formation of thermal stress in the ceramic. The method forms a bond between the alloy and the ceramic capable of withstanding the environment of a pressurized water reactor and suitable for use in an electrical conductivity sensiti… more

Date: July 24, 1981

Creator: Kelsey, P.V. Jr. & Siegel, W.T.

Item Type: Patent

Partner: UNT Libraries Government Documents Department

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Developments in sensitivity theory

A review of recent developments in sensitivity theory is presented with an emphasis on (a) extensions to new areas such as thermal hydraulics, reactor depletion, multi-constraint and extrema problems, and (b) recent mathematical refinements to and extensions of the basic theory. The diverse new areas of application are discussed from a unified theoretical viewpoint based on nonlinear functional analysis. Several new applications of sensitivity theory are presented for problems in constrained re… more

Date: January 1, 1980

Creator: Cacuci, D. G.; Greenspan, E.; Marable, J. H. & Williams, M. L.

Item Type: Article

Partner: UNT Libraries Government Documents Department