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Doubled-ended breaks in reactor primary piping. [Guillotine breaks]

Description: Results indicate that the probability of double-ended guillotine break (DEGB) in the reactor coolant loop piping of Westinghouse and Combustion Engineering plants is extremely low. It is recommended that the NRC seriously consider eliminating DEGB as a design basis event for reactor coolant loop piping in Westinghouse plants. Pipe whip restraints on reactor coolant loop piping could then be excluded or removed, and the requirement to design supports to withstand asymmetric blowdown loads could be eliminated. It is also recommended that the current requirement to couple safe shutdown earthquake (SSE) and DEGB be eliminated. Recognizing however that seismically induced support failure is the weak link in the DEGB evaluation, it is recommended that the strength of component supports, currently designed for the combination of SSE plus DEGB, not be reduced. The study indicates that the probability of DEGB in reactor coolant loop piping is sufficiently low under all plant conditions, including seismic events, to justify eliminating it entirely as a basis for plant design.
Date: October 1, 1984
Creator: Holman, G.S.
Partner: UNT Libraries Government Documents Department

PSEPLOT: a controller for plotting data from the Mark I Boiling Water Reactor Pressure Suppression Experiment. [BWR]

Description: PSEPLOT is a computer routine that was developed for the Lawrence Livermore Laboratory Octopus computer system to generate several thousand plots of engineering data in a consistent format for referencing and comparison. The time-dependent engineering data were recorded during each of 25 tests of the Mark I Pressure Suppression Experiment (PSE). Although PSEPLOT is restricted to PSE, its concept is applicable to any similar data management task.
Date: May 10, 1978
Creator: Holman, G.S.
Partner: UNT Libraries Government Documents Department

US NRC/LLL liaison with the Federal Republic of Germany for the GKSS-PSS steam condensation tests. Progress report No. 2

Description: This second progress report for the USNRC/LLL liaison program with the Federal Republic of Germany regarding boiling water reactor containment multivent steam condensation tests being conducted by GKSS addresses program activity during the period of July-August, 1979. During this period, the first digital data, video tapes, and complete report for test VM1 were received, together with various computer software used by GKSS for data reduction. Document handling procedures were finalized in order to protect the proprietary nature of informaion received from GKSS and several translations were obtained.
Date: September 13, 1979
Creator: Holman, G.S.
Partner: UNT Libraries Government Documents Department

Component Fragility Research Program: Phase 1 component prioritization

Description: Current probabilistic risk assessment (PRA) methods for nuclear power plants utilize seismic ''fragilities'' - probabilities of failure conditioned on the severity of seismic input motion - that are based largely on limited test data and on engineering judgment. Under the NRC Component Fragility Research Program (CFRP), the Lawrence Livermore National Laboratory (LLNL) has developed and demonstrated procedures for using test data to derive probabilistic fragility descriptions for mechanical and electrical components. As part of its CFRP activities, LLNL systematically identified and categorized components influencing plant safety in order to identify ''candidate'' components for future NRC testing. Plant systems relevant to safety were first identified; within each system components were then ranked according to their importance to overall system function and their anticipated seismic capacity. Highest priority for future testing was assigned to those ''very important'' components having ''low'' seismic capacity. This report describes the LLNL prioritization effort, which also included application of ''high-level'' qualification data as an alternate means of developing probabilistic fragility descriptions for PRA applications.
Date: June 1, 1987
Creator: Holman, G.S. & Chou, C.K.
Partner: UNT Libraries Government Documents Department

Assessment of value-impact associated with the elimination of postulated pipe ruptures from the design basis for nuclear power plants

Description: The US Nuclear Regulatory Commission is proposing to amend the regulations that currently require that the design basis for nuclear power plants include the postulation of dynamic effects from loss of coolant accidents up to and including the double-ended rupture of the largest pipe in the reactor coolant system. Proposed modifications would allow analyses to serve as a sufficient basis for excluding dynamic effects, including but not necessarily limited to pipe whip and jet impingement, associated with specific pipe ruptures. Only dynamic effects would be impacted; current design requirements for containment sizing and discharge capacity of emergency core cooling systems would remain unchanged. This report presents a detailed analysis of value-impact associated with the proposed amendment for PWR reactor coolant loop piping and for BWR recirculation loop piping. The effect of extending application of the proposed rule change to other piping systems is also assessed in a less quantitative manner.
Date: March 29, 1985
Creator: Holman, G.S. & Chou, C.K.
Partner: UNT Libraries Government Documents Department

Computer-aided digitization of graphical mass flow data from the 1/5-scale Mark I BWR pressure suppression experiment

Description: Periodically in the analysis of engineering data, it becomes necessary to use graphical output as the solitary source of accurate numerical data for use in subsequent calculations. Such was our experience in the extended analysis of data from the 1/5-scale Mark I boiling water reactor pressure suppression experiment (PSE). The original numerical results of extensive computer calculations performed at the time of the actual PSE tests and required for the later extended analysis program had not been retained as archival records. We were, therefore, required to recover the previously calculated data, either by a complete recalculation or from available computer graphics records. Time constraints suggested recovery from the graphics records as the more viable approach. This report describes two different approaches to recovery of digital data from graphics records. One, combining hard and software techniques immediately available to us at LLL, proved to be inadequate for our purposes. The other approach required the development of pure software techniques that interfaced with LLL computer graphics to unpack digital coordinate information directly from graphics files. As a result of this effort, we were able to recover the required data with no significant loss in the accuracy of the original calculations.
Date: July 1, 1979
Creator: Holman, G.S. & McCauley, E.W.
Partner: UNT Libraries Government Documents Department

Effects of torus wall flexibility on forces in the Mark I BWR pressure suppression system under SRV T-quencher loading

Description: This report describes a series of extended analyses requested by the US Nuclear Regulatory Commission to provide qualified understanding of possible fluid/structure interaction (FSI) effects for SRV teequencher test results. Three input pulses with total impulses varying by up to a factor of five are applied to two-dimensional finite-element models of the Mark I torus with shell diameter-to-thickness ratios of 0, 300, and 600. The results of these analyses support earlier conclusions that increased wall flexibility enhances attenuation of hydrodynamic loads and furthermore indicate that the magnitude of the attenuation is only weakly affected by the total impulse of the bubble pressure time-history.
Date: January 15, 1980
Creator: Holman, G.S. & Lu, S.C.
Partner: UNT Libraries Government Documents Department

Preliminary planning study for safety relief valve experiments in a Mark III BWR pressure suppression system

Description: In response to a request from the Water Reactor Safety Research Division of the US NRC, a preliminary study is provided which identifies key features and consideration involved in planning a comprehensive in-plant Safety Relief Valve experimental program for a Mark III containment design. The report provides identification of program objectives, measurement system requirements, and some details quantifying expected system response. In addition, a preliminary test matrix is outlined which involves a supporting philosophy intended to enhance the usefulness of the experimental results for all members of the program team: experimentalists, analysts, and plant operator.
Date: April 21, 1980
Creator: McCauley, E.W. & Holman, G.S.
Partner: UNT Libraries Government Documents Department

Equipment fragility testing

Description: Current probabilistic risk assessment (PRA) methods for nuclear power plants utilize component fragilities which are for the most part based on a limited data base and engineering judgement. The seismic design of components is based on code limits and NRC requirements that do not reflect the actual capacity of a component to resist failure. In order to improve the present component fragility data base and establish component seismic design margins, the NRC has commissioned a projected three-year program to compile existing fragilities data and at the same time independently perform fragilities tests on selected mechanical and electrical components. This paper presents the planning and technical approach being taken by LLNL in the NRC Component Fragility Program.
Date: January 1, 1985
Creator: Holman, G.S.; Chou, C.K. & Cummings, G.E.
Partner: UNT Libraries Government Documents Department

Applicability of flat plate methods in determining fluid/structure interaction effects in BWR pressure suppression systems

Description: Flat plate chord tests are one experimental method used to investigate how fluid/structure interaction (FSI) effects modify the impulsive loading in nuclear reactor pressure suppression pools. This analytical study examines the applicability of using a flexible flat plate in an otherwise rigid shell to model dynamic pool wall response in a flexible shell pressure suppression torus. Bubble pressures varying by a factor of seven are used as input to two-dimensional finite-element models. Boundary response to various plate and shell thicknesses are compared on the basis of equivalent natural frequency. Results indicate the qualitative flat plate response compares well with the flexible shell but absolute pressures vary significantly and nonconservatively.
Date: March 5, 1979
Creator: Holman, G.S.; McCauley, E.W. & Lu, S.C.H.
Partner: UNT Libraries Government Documents Department

Three-dimensional linear analysis of fluid-structure interaction effects in the Mark I BWR pressure suppression torus

Description: Most analytical and experimental approaches to the evaluation of fluid-structure interaction (FSI) effects in the General Electric Mark I BWR pressure suppression system treat the torus shell as rigid when the shell in real systems is flexible. This report describes linear three-dimensional finite-element analyses of one torus bay that investigated the qualitative effect of torus wall flexibility on hydrodynamic loads induced by a nominal safety relief valve (SRV) discharge. The results of these analyses support the general conclusion drawn from earlier two-dimensional analyses. The report also discusses finite-element analyses of a 3-D representation of the earlier 2-D plane-strain model of the torus shell.
Date: January 15, 1980
Creator: Holman, G.S.; McCauley, E.W. & Lu, S.C.
Partner: UNT Libraries Government Documents Department

Effect of torus wall flexibility on hydro-structural interaction in BWR containment system

Description: The MARK I boiling water reactor (BWR) containment system is comprised of a light-bulb-shaped reactor compartment connected through vent pipes to a torus-shaped and partially water-filled pressure suppression chamber, or the wetwell. During either a normally occurring safety relief valve (SRV) discharge or a hypothetical loss-of-coolant accident (LOCA), air or steam is forced into the wetwell water pool for condensation and results in hydrodynamically induced loads on the torus shell. An analytical program is described which employs the finite element method to investigate the influence of torus wall flexibility on hydrodynamically induced pressure and the resultant force on the torus shell surface. The shell flexibility is characterized by the diameter-to-thickness ratio which is varied from the perfectly rigid case to the nominal plant condition. The general conclusion reached is that torus wall flexibility decreases both the maximum pressure seen by the shell wall and the total vertical load resulted from the hydrodynamically induced pressure.
Date: April 25, 1979
Creator: Lu, S.C.H.; McCauley, E.W. & Holman, G.S.
Partner: UNT Libraries Government Documents Department

Component Fragility Research Program: Phase 1, Demonstration tests: Volume 2, Appendices

Description: Appendices are presented which contain information concerning: details of controller and relay installation; resonance search transmissibility plots; time-history data from runs 17, 31, 46, and 56; and response spectra from runs 17, 31, 46, and 56. (JDB)
Date: August 1, 1987
Creator: Holman, G.S.; Chou, C.K.; Shipway, G.D. & Glozman, V.
Partner: UNT Libraries Government Documents Department

Component Fragility Research Program: Phase 1, Demonstration tests: Volume 1, Summary report

Description: This report describes tests performed in Phase I of the NRC Component Fragility Research Program. The purpose of these tests was to demonstrate procedures for characterizing the seismic fragility of a selected component, investigating how various parameters affect fragility, and finally using test data to develop practical fragility descriptions suitable for application in probabilistic risk assessments. A three-column motor control center housing motor controllers of various types and sizes as well as relays of different types and manufacturers was subjected to seismic input motions up to 2.5g zero period acceleration. To investigate the effect of base flexibility on the structural behavior of the MCC and on the functional behavior of the electrical devices, multiple tests were performed on each of four mounting configurations: four bolts per column with top bracking, four bolts per column with no top brace, four bolts per column with internal diagonal bracking, and two bolts per column with no top or internal bracking. Device fragility was characterized by contact chatter correlated to local in-cabinet response at the device location. Seismic capacities were developed for each device on the basis of local input motion required to cause chatter; these results were then applied to develop probabilistic fragility curves for each type of device, including estimates of the ''high-confidence low probability of failure'' capacity of each.
Date: August 1, 1987
Creator: Holman, G.S.; Chou, C.K.; Shipway, G.D. & Glozman, V.
Partner: UNT Libraries Government Documents Department

Physical model of lean suppression pressure oscillation phenomena: steam condensation in the light water reactor pressure suppression system (PSS)

Description: 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.
Partner: UNT Libraries Government Documents Department

Failure probability of PWR reactor coolant loop piping. [Double-ended guillotine break]

Description: This paper describes the results of assessments performed on the PWR coolant loop piping of Westinghouse and Combustion Engineering plants. For direct double-ended guillotine break (DEGB), consideration was given to crack existence probability, initial crack size distribution, hydrostatic proof test, preservice inspection, leak detection probability, crack growth characteristics, and failure criteria based on the net section stress failure and tearing modulus stability concept. For indirect DEGB, fragilities of major component supports were estimated. The system level fragility was then calculated based on the Boolean expression involving these fragilities. Indirect DEGB due to seismic effects was calculated by convolving the system level fragility and the seismic hazard curve. The results indicate that the probability of occurrence of both direct and indirect DEGB is extremely small, thus, postulation of DEGB in design should be eliminated and replaced by more realistic criteria.
Date: February 1, 1984
Creator: Lo, T.; Woo, H.H.; Holman, G.S. & Chou, C.K.
Partner: UNT Libraries Government Documents Department

Component fragility research program

Description: To demonstrate how high-level'' qualification test data can be used to estimate the ultimate seismic capacity of nuclear power plant equipment, we assessed in detail various electrical components tested by the Pacific Gas Electric Company for its Diablo Canyon plant. As part of our Phase I Component Fragility Research Program, we evaluated seismic fragility for five Diablo Canyon components: medium-voltage (4kV) switchgear; safeguard relay board; emergency light battery pack; potential transformer; and station battery and racks. This report discusses our Phase II fragility evaluation of a single Westinghouse Type W motor control center column, a fan cooler motor controller, and three local starters at the Diablo Canyon nuclear power plant. These components were seismically qualified by means of biaxial random motion tests on a shaker table, and the test response spectra formed the basis for the estimate of the seismic capacity of the components. The seismic capacity of each component is referenced to the zero period acceleration (ZPA) and, in our Phase II study only, to the average spectral acceleration (ASA) of the motion at its base. For the motor control center, the seismic capacity was compared to the capacity of a Westinghouse Five-Star MCC subjected to actual fragility tests by LLNL during the Phase I Component Fragility Research Program, and to generic capacities developed by the Brookhaven National Laboratory for motor control center. Except for the medium-voltage switchgear, all of the components considered in both our Phase I and Phase II evaluations were qualified in their standard commercial configurations or with only relatively minor modifications such as top bracing of cabinets. 8 refs., 67 figs., 7 tabs.
Date: November 1, 1989
Creator: Tsai, N.C.; Mochizuki, G.L.; Holman, G.S. (NCT Engineering, Inc., Lafayette, CA (USA) & Lawrence Livermore National Lab., CA (USA))
Partner: UNT Libraries Government Documents Department