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in subjects: "Reactivity Insertions"

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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

Assessment of molten debris freezing in a severe RIA in-pile test. [PWR; BWR]

An understanding of the freezing of molten debris on cold core structures following a hypothetical core meltdown accident in a light water reactor (LWR) is of importance to reactor safety analysis. The purpose of the present investigation was to analyze the transient freezing of the molten debris produced in a severe reactivity initiated accident (RIA) scoping test, designated RIA-ST-4, which was performed in the Power Burst Facility and simulated a BWR control rod drop accident. In the RIA-ST-… more

Date: January 1, 1980

Creator: El-Genk, Mohamed S. & Moore, Richard L.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Effects of subcooling and rod drop speed on the BWR rod drop accident

The techniques and models used in the analysis of the control rod drop accident (CRDA) in a BWR have ranged from approximate conservative methods with a simple feedback model to detailed representations of the thermal-hydraulic and neutronic mechanisms. In a recent paper Cheng and Diamond presented a detailed evaluation of the CRDA and the effects of varying a number of important accident parameters. Their calculations performed with the BNL-TWIGL core dynamics code, have shown that the effect … more

Date: January 1, 1982

Creator: Cokinos, D.; Carew, J. & Aronson, A.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Sodium Loop Safety Facility W-2 experiment fuel pin rupture detection system. [LMFBR]

The objective of the Sodium Loop Safety Facility (SLSF) W-2 experiment is to characterize the combined effects of a preconditioned full-length fuel column and slow transient overpower (TOP) conditions on breeder reactor (BR) fuel pin cladding failures. The W-2 experiment will meet this objective by providing data in two technological areas: (1) time and location of cladding failure, and (2) early post-failure test fuel behavior. The test involves a seven pin, prototypic full-length fast test re… more

Date: May 1, 1980

Creator: Hoffman, M. A.; Kirchner, T. L. & Meyers, S. C.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Failure of latch mechanism for motion control of safety rods

During safety rod tests in K-reactor prior to startup, one safety rod could not be lifted because the button'' broke off and became lodged in the mechanism. Examination of the failed latch assembly along with other assemblies from both K-Area and L-Area revealed several missing buttons as well as severely deformed jaw hanger extensions.'' We participated in the investigation of the damage by request of the Reactor Restart Section. Based on our study of the latch mechanism, the modifications to … more

Date: January 16, 1992

Creator: Yau, W. W. F. & Leader, D. R.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

ORNL analyses of AVR performance and safety

Because of the high interest in modular High Temperature Reactor performance and safety, a cooperative project has been established involving the Oak Ridge National Laboratory (ORNL), Arbeitsgemeinschaft Versuchs Reaktor GmbH (AVR), and Kernforschungsanlage Juelich GmbH (KFA) in reactor physics, performance and safety. This paper presents initial results of ORNL's examination of a hypothetical depressurized core heatup accident and consideration of how a depressurized core heatup test might be … more

Date: January 1, 1985

Creator: Cleveland, J.C.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Xenon changes under power-burst conditions. [BWR]

Under ordinary operating conditions the xenon concentration in a reactor core can change significantly in times on the order of hours. Core transients of safety significance are much more rapid and hence calculations are done with xenon concentration held constant. However, in certain transients (such as reactivity initiated accidents) there is a very large power surge and the question arises as to whether under these circumstances the xenon concentration could change. This would be particularl… more

Date: January 1, 1983

Creator: Diamond, D. J.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Analysis of reactivity-insertion accidents in the TREAT Upgrade reactor

The expansion of the experimental capabilities of the TREAT Upgrade (TU) reactor also tends to increase the potential risks associated with off-normal reactivity insertion incidents compared to the TREAT reactor. To provide adequate prtection for the public and the facility, while meeting experimenter's requirements, a specialized Reactor Trip System (RTS) with energy-dependent scram trips on reactor power and period has been developed. With this protection strategy, the consequences of reactiv… more

Date: January 1, 1983

Creator: Rudolph, R.R. & Bhattacharyya, S.K.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Physics of reactor safety. Quarterly report, January--March 1977. [LMFBR]

This report summarizes work done on reactor safety, Monte Carlo analysis of safety-related critical assembly experiments, and planning of DEMI safety-related critical experiments. Work on reactor core thermal-hydraulics is also included.

Date: June 1, 1977

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Comparison of measured and calculated LWR fuel behavior during a hypothetical reactivity initiated accident

Comparisons of measured and calculated LWR fuel rod responses during a reactivity initiated accident test are presented. The results indicate that the computer code, FRAP-T5, adequately calculates the fuel rod behavior up to the time at which the gap closes and provides a good thermal solution up to the time of gross fuel and cladding relocation. Three areas have been identified for further model development: (a) development of a fuel swelling model for near molten fuel temperatures; (b) incorp… more

Date: January 1, 1980

Creator: Fukuda, S. K.; MacDonald, P. E. & Garner, R. W.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Simulation of the response of the Fort St. Vrain high temperature gas cooled reactor system to a postulated rod withdrawal accident

Transients resulting from the accidental withdrawal of a control rod pair from the Fort St. Vrain reactor core from 100% power conditions have been analyzed with the ORTAP nuclear steam supply system simulation. This analysis was done as part of an ongoing effort to obtain an independent assessment of the HTGR system response to several postulated accidents. Results are presented and discussed.

Date: January 1, 1977

Creator: Cleveland, J. C.; Hedrick, R. A.; Ball, S. J.; Delene, J. G. & Conklin, J. C.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Uncertainty in BWR power during ATWS events

A study was undertaken to improve our understanding of BWR conditions following the closure of main steam isolation valves and the failure of reactor trip. Of particular interest was the power during the period when the core had reached a quasi-equilibrium condition with a natural circulation flow rate determined by the water level in the downcomer. Insights into the uncertainity in the calculation of this power with sophisticated computer codes were quantified using a simple model which relate… more

Date: January 1, 1986

Creator: Diamond, D. J.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Analysis of the SL-1 Accident Using RELAPS5-3D

On January 3, 1961, at the National Reactor Testing Station, in Idaho Falls, Idaho, the Stationary Low Power Reactor No. 1 (SL-1) experienced a major nuclear excursion, killing three people, and destroying the reactor core. The SL-1 reactor, a 3 MW{sub t} boiling water reactor, was shut down and undergoing routine maintenance work at the time. This paper presents an analysis of the SL-1 reactor excursion using the RELAP5-3D thermal-hydraulic and nuclear analysis code, with the intent of simulat… more

Date: November 8, 2007

Creator: Francisco, A.D. and Tomlinson, E. T.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Interim Status Report on Lead-Cooled Fast Reactor (Lfr) Research and Development.

This report discusses the status of Lead-Cooled Fast Reactor (LFR) research and development carried out during the first half of FY 2008 under the U.S. Department of Energy Generation IV Nuclear Energy Systems Initiative. Lead-Cooled Fast Reactor research and development has recently been transferred from Generation IV to the Reactor Campaign of the Global Nuclear Energy Partnership (GNEP). Another status report shall be issued at the end of FY 2008 covering all of the LFR activities carried ou… more

Date: March 31, 2008

Creator: Tzanos, C. P.; Sienicki, J. J.; Moisseytsev, A.; Smith, C. F.; de Caro, M.; Halsey, W. G. et al.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Analysis of an Earthquake-Initiated-Transient in a PBR

One of the Design Basis Accidents (DBA) for a Pebble Bed Reactor has been identified as the “Safe shutdown earthquake with core conduction cooling to passive mode of Reactor Cavity Cooling System.” A new methodology to analyze this particular DBA has been developed at the Idaho National Laboratory (INL). During the seismic event the reactor core experiences the densification of the pebbles, which produce small reactivity insertions due to the effective fuel densification. In addition, a decreas… more

Date: May 1, 2009

Creator: Ougouag, A. M.; Ortensi, J. & Hiruta, H.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Light water reactor fuel response during reactivity initiated accident experiments

Experimental results from six recent Power Burst Facility (PBF) reactivity initiated accident (RIA) tests are compared with data from previous SPERT, TREAT and NSRR programs. The RIA fuel behavior experimental program recently started in the PBF is being conducted with coolant conditions typical of hot-startup conditions in a commercial boiling water reactor. The SPERT, TREAT and NSRR test programs investigated the behavior of single or small clusters of light water reactor (LWR) type fuel rods… more

Date: January 1, 1979

Creator: MacDonald, P. E.; McCardell, R. K.; Martinson, Z. R.; Hobbins, R. R.; Seiffert, S. L. & Cook, B. A.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

PARET code and the analysis of the SPERT I transients

The PARET code has been adapted for the testing of methods and models and for subsequent use in the analysis of transient behavior in research reactors. Comparisons with the experimental results from the SPERT-I transients are provided. The code has also been applied to the analysis of the IAEA 10 MW benchmark cores for protected and unprotected transients. The present version of the PARET code provides a convenient means of assessing the various models and correlations proposed for use in the … more

Date: January 1, 1982

Creator: Woodruff, W.L.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Re-evaluation of Savannah River reactor transient reactivity coefficient tests: The effect of delayed neutron constants and spatial variations

Transient reactivity test conducted in one of the Savannah River production reactors in 1962 have been re-evaluated. A significantly lower (more negative) coolant temperature coefficient is now ascribed to that test; {minus}1 pcm/Deg-C vs the previously obtained +2 pcm/Deg-C. The change from the previous value is because of revisions to delayed neutron constants and accounting for spatial effects. The new value is in reasonable agreement with the currently calculated value of {minus}2 pcm/Deg-C… more

Date: January 1, 1991

Creator: Burnett, T.W.T. (Westinghouse Electric Corp., Pittsburgh, PA (United States). Nuclear and Advanced Technology Div.) & Graves, W.E. (Westinghouse Savannah River Co., Aiken, SC (United States))

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Fission-product-release signatures for LWR fuel rods failed during PCM and RIA transients

This paper discusses fission product release from light-water-reactor-type fuel rods to the coolant loop during design basis accident tests. One of the tests was a power-cooling-mismatch test in which a single fuel rod was operated in film boiling beyond failure. Other tests discussed include reactivity initiated accident (RIA) tests, in which the fuel rods failed as a result of power bursts that produced radial-average peak fuel enthalpies ranging from 250 to 350 cal/g. One of the RIA tests us… more

Date: January 1, 1981

Creator: Osetek, D. J.; King, J. J. & Croucher, D. W.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Comparative analysis of the CRDA using BNL-TWIGL and RAMONA-3B. [BWR]

A comparative analysis of the BWR control rod drop accident (CRDA) using BNL-TWIGL and RAMONA-3B has been performed as part of the BNL/NRC evaluation of methods currently used to analyze BWR CRDA events. A principal objective of this analysis was to test the two-dimensional neutronics model used in BNL-TWIGL aganist the full three-dimensional model in RAMONA-3B. Additionally, the results of analyzing the identical transient with the two codes were expected to help evaluate other approximate mod… more

Date: June 1, 1983

Creator: Neogy, P. & Carew, J.F.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Scoping studies of vapor behavior during a severe accident in a metal-fueled reactor

Scoping calculations have been performed examining the consequences of fuel melting and pin failures for a reactivity-insertion type accident in a sodium-cooled, pool-type reactor fueled with a metal alloy fuel. The principal gas and vapor species released are shown to be Xe, Cs,and bond sodium contained within the fuel porosity. Fuel vapor pressure is insignificant, and there is no energetic fuel-coolant interaction for the conditions considered. Condensation of sodium vapor as it expands into… more

Date: April 15, 1985

Creator: Spencer, B.W. & Marchaterre, J.F.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Effect of thermal-hydraulic feedback on the BWR rod drop accident

An important design-basis accident for boiling water reactors (BWR's) is the rod drop accident (RDA). This accident is defined to be a rapid reactor transient caused by an accidental drop (out of the core) of the highest-worth control rod at various conditions ranging from cold start-up to about 10% of rated power. For most BWR designs the highest worth rod is normally situated at the center of the core. Despite the fact that the chance of a RDA in extremely unlikely, the consequence of the RDA… more

Date: January 1, 1979

Creator: Cheng, H. S. & Diamond, D. J.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

A theoretical prediction of critical heat flux in saturated pool boiling during power transients

Understanding and predicting critical heat flux (CHF) behavior during steady-state and transient conditions is of fundamental interest in the design, operation, and safety of boiling and two-phase flow devices. Presented within this paper are the results of a comprehensive theoretical study specifically conducted to model transient CHF behavior in saturated pool boiling. Thermal energy conduction within a heating element and its influence on the CHF are also discussed. The resultant theory prov… more

Date: January 1, 1987

Creator: Pasamehmetoglu, K.O.; Nelson, R.A. & Gunnerson, F.S.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Response of FFTF core to protected reactivity addition transients

The response of the FFTF core to protected reactivity insertion events was evaluated. Reactivity addition transients ranging from .05 cents/s to 3$/s have been considered. The evaluation method is based on a calculational model which predicts cladding strain from modified fuel-cladding differential thermal expansion. The results show that for all ramp rates considered, the Plant Protection System (PPS) controls consequences to required limits. Comparisons made between predicted fuel damage and … more

Date: August 1, 1979

Creator: Yee, A. K.; Baars, R. E. & Stepnewski, D. D.

Item Type: Article

Partner: UNT Libraries Government Documents Department