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in subjects: "Excursions"

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Design of a full scale model fuel assembly for full power production reactor flow excursion experiments

A novel full scale production reactor fuel assembly model was designed and built to study thermal-hydraulic effects of postulated Savannah River Site (SRS) nuclear reactor accidents. The electrically heated model was constructed to simulate the unique annular concentric tube geometry of fuel assemblies in SRS nuclear production reactors. Several major design challenges were overcome in order to produce the prototypic geometry and thermal-hydraulic conditions. The two concentric heater tubes (to… more

Date: January 1, 1990

Creator: Nash, C. A. (Westinghouse Savannah River Co., Aiken, SC (United States)); Blake, J. E. & Rush, G. C. (Babcock and Wilcox Co., Alliance, OH (United States))

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Final report of fuel dynamics Test E7

Test data from an in-pile failure experiment of high-power LMFBR-type fuel pins in a simulated $3/s transient-overpower (TOP) accident are reported and analyzed. Major conclusions are that (1) a series of cladding ruptures during the 100-ms period preceding fuel release injected small bursts of fission gas into the flow stream; (2) gas release influenced subsequent cladding melting and fuel release (there were no measurable FCI's (fuel-coolant interactions), and all fuel motion observed by the … more

Date: April 1, 1977

Creator: Doerner, R. C.; Murphy, W. F.; Stanford, G. S. & Froehle, P. H.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Outlet plenum mixing for transient overpower conditions of a one-exit nozzle LMFBR

Two types of transient tests were employed to model a one-exit nozzle LMFBR outlet plenum. Water was used as a test fluid in the simulation of constant flowrates, Transient Overpower (TOP) conditions. In the first test, simulated fuel flow was 85% and blanket flow was 15%, whereas in the second test, the fuel flow was 100%. This allowed the assessment of the mitigating effects of blanket flow upon the exit nozzle temperature transient. The flow field was clearly three-dimensional, and a less ac… more

Date: April 1, 1978

Creator: Howard, P. A.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Prism reactor system design and analysis of postulated unscrammed events

Key safety characteristics of the PRISM reactor system include the passive reactor shutdown characteristic and the passive shutdown heat removal system, RVACS. While these characteristics are simple in principle, the physical processes are fairly complex, particularly for the passive reactor shutdown. It has been possible to adapt independent safety analysis codes originally developed for the Clinch River Breeder Reactor review, although some limitations remain. In this paper, the analyses of p… more

Date: August 1, 1991

Creator: Van Tuyle, G.J. & Slovik, G.C.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Benchmarking of flowtran with Mark-22 mockup flow excursion test data from Babcock Wilcox

Version 16.2 of the FLOWTRAN code with a Savannah River Site (SRS) working criterion (St=0.00455) for the onset of significant void (OSV) was benchmarked against power and flow excursion data derived from tests at the Babcock Wilcox Alliance Research Center test facility. This document presents analyses which show that FLOWTRAN accurately predicts the mockup test assembly thermal-hydraulic behavior during the steady state and LOCA transient conditions, and that FLOWTRAN with a Savannah River Si… more

Date: June 1, 1990

Creator: Chen, Juo-Fu.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Simulation of LMFBR excursion models by means of ICECO

Test data from experiments dealing with LMFBR containment studies are compared with containment code solutions. The features of fluid spillage, which is a subject of some concern in experiments, is modeled in the ICECO code and is stressed. Several variations of spillage solution were arrived at gaining more insight to this problem and thus serve in reducing some of the uncertainties connected with fluid spillage. Other features of the analytical models are utilized and described.

Date: January 1, 1977

Creator: Marchertas, A. H.; Wang, Y. C. & Fistedis, S. H.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Reliability of fast reactor mixed-oxide fuel during operational transients

Results are presented from the cooperative DOE and PNC Phase 1 and 2 operational transient testing programs conducted in the EBR-2 reactor. The program includes second (D9 and PNC 316 cladding) and third (FSM, AST and ODS cladding) generation mixed-oxide fuel pins. The irradiation tests include duty cycle operation and extended overpower tests. the results demonstrate the capability of second generation fuel pins to survive a wide range of duty cycle and extended overpower events. 15 refs., 9 f… more

Date: July 1, 1991

Creator: Boltax, A.; Neimark, L.A.; Tsai, Hanchung (Argonne National Lab., IL (United States)); Katsuragawa, M. & Shikakura, S. (Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center)

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Analysis of fluid-structure interaction and structural response of Chernobyl-4 reactor

On April 26, 1986, an accident occurred at the Chernobyl-4 Nuclear Power Plant in the Soviet Union. A post accident meeting was held in Vienna during the week of August 25, 1986. In mid-July 1986, the DOE formed a team to analyze the accident, including experts from the national laboratories such as Argonne National Laboratory, Brookhaven National Laboratory, Oak Ridge National Laboratory, and Pacific Northwest Laboratory. The goal was to assess the information's plausibility, provided analytic… more

Date: January 1, 1989

Creator: Wang, C. Y.; Pizzica, P. A.; Gvildys, J. & Spencer, B. W.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Finite-element method for above-core structures. [LMFBR]

Three-dimensional finite-element models for the treatment of the nonlinear, transient response of a fast breeder reactor's above-core structures are described. For purposes of treating arbitrarily large rotations, node orientations are described by unit vectors and the deformable elements are treated by a corotational formulation in which the coordinate system is embedded in the elements. Deformable elements may be connected either to nodes directly or through rigid bodies. The time integration… more

Date: December 1, 1979

Creator: Kennedy, J. M. & Belytschko, T. B.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

BWR stability analysis at Brookhaven National Laboratory

Following the unexpected, but safely terminated, power and flow oscillations in the LaSalle-2 Boiling Water Reactor (BWR) on March 9, 1988, the Nuclear Regulatory Commission (NRC) Offices of Nuclear Reactor Regulation (NRR) and of Analysis and Evaluation of Operational Data (AEOD) requested that the Office of Nuclear Regulatory Research (RES) carry out BWR stability analyses, centered around fourteen specific questions. Ten of the fourteen questions address BWR stability issues in general and a… more

Date: January 1, 1991

Creator: Wulff, W.; Cheng, H.S.; Mallen, A.N. & Rohatgi, U.S.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Performance of fast reactor mixed-oxide fuels pins during extended overpower transients

The Operational Reliability Testing (ORT) program, a collaborative effort between the US Department of Energy and the Power Reactor and Nuclear Fuel Development Corp. (PNC) of Japan, was initiated in 1982 to investigate the behavior of mixed-oxide fuel pin under various slow-ramp transient and duty-cycle conditions. In the first phase of the program, a series of four extended overpower transient tests, with severity sufficient to challenge the pin cladding integrity, was conducted. The objectiv… more

Date: February 1, 1991

Creator: Tsai, H.; Neimark, L.A. (Argonne National Lab., IL (USA)); Asaga, T. & Shikakura, S. (Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan))

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Design inputs document: Boiling behavior during flow instability

The coolant flow in a nuclear reactor core under normal operating conditions is kept as a subcooled liquid. This coolant is evenly distributed throughout the multiple flow channels with a uniform pressure profile across each coolant flow channel. If the coolant flow is reduced, the flow through individual channels will also decrease. A decrease in coolant flow will result in higher coolant temperatures if the heat flux is not reduced. When flow is significantly decreased, localized boiling may … more

Date: January 1, 1991

Creator: Coutts, D. A.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Verification and Validation of Corrected Versions of RELAP5 for ATR Reactivity Analyses

Two versions of the RELAP5 computer code, RELAP5/MOD2.5 and RELAP5/MOD3 Version 3.2.1.2, are used to support safety analyses of the Advanced Test Reactor (ATR). Both versions of RELAP5 contain a point reactor kinetics model that has been used to simulate power excursion transients at the ATR. Errors in the RELAP5 point kinetics model were reported to the RELAP5 code developers in 2007. These errors had the potential to affect reactivity analyses that are part of the ATR’s safety basis. Conseque… more

Date: November 1, 2008

Creator: Davis, Cliff B.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Analysis of postulated events for the revised ALMR/PRISM design

The Nuclear Regulatory Commission (NRC) is continuing a pre-application review of the 471 MWt, Liquid Metal Reactor, PRISM by General Electric, with Brookhaven National Laboratory providing technical support. The revised design has been evaluated, using the SSC code, for an unscrammed loss of heat sink (ULOHS), an unscrammed loss of flow (ULOF) with and without the Gas Expansion Modules (GEMs), and a 40{cents} unscrammed transient overpower (UTOP) event. The feedback effects for U-27Pu-10Zr met… more

Date: January 1, 1991

Creator: Slovik, G.C. & Van Tuyle, G.J.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Elastic and plastic properties of uranium dioxide from 5 to 330 GPa

We have measured the shock-compression parameters for UO/sub 2/ to 330 GPa. The Hugoniot elastic limit was found to be 5.7 GPa. Evidence for a shock-induced phase transition was observed at about 54 GPa.

Date: July 11, 1979

Creator: Gust, W.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 Special Power Excursion Reactor Test (SPERT), and Japanese Nuclear Safety Research Reactor (NSRR) tests. 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 and NSRR test programs investigated the behavior of si… more

Date: January 1, 1979

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

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Prompt burst energetics (PBE) studies

These studies involve an in-pile experimental program and complementary analytical investigation of the energetic response of important reactor fuel-clad-coolant systems subjected to energy deposition conditions associated with super-prompt critical excursions. The objectives of the present program include identification and characterization of the phenomena which dominate the conversion of thermal energy to work, development of models which accurately predict the energetics associated with suc… more

Date: January 1, 1979

Creator: Reil, K.O. & Young, M.F.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Effect of power skew on the FTR TOP

An enduring program of studies, both experimentally and analytically, on the unprotected Transient Overpower Hypothetical Core Disruptive Accident (TOP-HCDA) of the Liquid Metal Fast Breeder Reactor (LMFBR) has been in progress for a decade or so. The data available from the experiments and the computational methodology improvements have made the recent TOP safety assessments more realistic. The evidences from the in-pile tests in the TREAT facility, such as the axial location of fuel pins' fai… more

Date: March 1, 1978

Creator: Yung, S.C. & Wilburn, N.P.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Transient-overpower test E8 on FFTF-type low-power-irradiated fuel

Test E8 simulated a hypothetical $3/s transient overpower accident in an LMFBR using seven (Pu, U)O/sub 2/ fuel elements of the FTR type. The test elements were preirradiated in the PNL-10 assembly in EBR-II to 5 at. % burnup at 30 kW/m. The preirradiation in EBR-II caused a fuel-restructuring range characteristic of a low-to-moderate power microstructure for FTR. Failure predictions indicated that fuel with this microstructural characteristic would fail at a lower energy deposition than fuel i… more

Date: December 1, 1977

Creator: Simms, R.; Lo, R. K.; Murphy, W. F.; Stanford, G. S. & Rothman, A. B.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Improvement and Verification of Fast Reactor Safety Analysis Techniques. Progress Report, 1 July 1977--30 September 1977

A calorimeter was designed and built to measure reaction power in the range where liquid is boiling. Measurements made with the new calorimeter at a concentration just lower than what would produce boiling (3m) was about 10% lower than those measured using the Precision Calorimeter. A heat loss of about 40% due to a non-adiabatic vessel was determined by comparing the reaction power in solution of the reaction vessel to that of the calorimeter. A closed reaction vessel was designed to observe t… more

Date: January 1, 1977

Creator: Barker, D. H. & Wheeler, P. A.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Fuel vapor pressure (FVAPRS). [BWR; PWR]

A subcode (FVAPRS) is described which calculates fuel vapor pressure. This subcode was developed as part of the fuel rod behavior modeling task performed at EG and G Idaho, Inc. The fuel vapor pressure subcode (FVAPRS), is presented and a discussion of literature data, steady state and transient fuel vapor pressure equations and estimates of the standard error of estimate to be expected with the FVAPRS subcode are included.

Date: April 1, 1979

Creator: Mason, R. E.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Causes of Instability at LaSalle and Consequences From Postulated Scram Failure

The March 9, 1988 instability event at the LaSalle County-2 BWR power plant was simulated on the BNL Engineering Plant Analyzer (EPA) in order to assess its ability to simulate oscillatory transients, to identify the causes of the instability, to determine the maximum power amplitude of limit-cycle oscillations after postulated scram failure and to rank leading modeling parameters which affect instability. The EPA was found capable of simulating the LaSalle instability, which was found to have … more

Date: 1990~

Creator: Wulff, W.; Cheng, H. S. & Mallen, A. N.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Low sodium void cores

To avoid high energy releases in LMFBR TUC accidents which are accompanied by a failure to scram with a regular shutdown system, various devices have been proposed which would add negative reactivity to the core by either bringing poison material into the core or by creating negative reactivity feedbacks coming from the thermal expansion of the core. While inherent shutdown systems (ISSs) show promise for enhancing safety by adding poison to the reactor, the trigger mechanism and the geometry o… more

Date: January 1, 1978

Creator: Barthold, W. P.; Beitel, J. C.; Lam, P. S. K.; Orechwa, Y.; Su, S. F. & Turski, R. B.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Fault tree analysis of the EBR-II reactor shutdown system

As part of the level I Probabilistic Risk Assessment of the Experimental Breeder Reactor II (EBR-II), detailed fault trees for the reactor shutdown system are developed. Fault tree analysis is performed for two classes of transient events that are of particular importance to EBR-II operation: loss-of-flow and transient-overpower. In all parts of EBR-II reactor shutdown system, redundancy has been utilized in order to reduce scram failure probability. Therefore, heavy emphasis is placed in the f… more

Date: January 1, 1992

Creator: Kamal, S.A. & Hill, D.J.

Item Type: Article

Partner: UNT Libraries Government Documents Department