[en] The capability to perform thermal-hydraulic analyses of an SP-100 space reactor was demonstrated using the ATHENA computer code. The preliminary General Electric SP-100 design was modeled using Athena. The model simulates the fast reactor, liquid-lithium coolant loops, and lithium-filled heat pipes of this design. Two ATHENA demonstration calculations were performed simulating accident scenarios. A mask for the SP-100 model and an interface with the Nuclear Plant Analyzer (NPA) were developed, allowing a graphic display of the calculated results on the NPA. 22 figs

[en] As part of an independent assessment study to establish code capabilities, comparisons between the RELAP4/MOD6, UPDATE 4 pre-test prediction and experimental data are presented for Semiscale Mod-3 Reflood Test S-07-4. Where deficiencies are indicated in the code-calculated representation of the experiment, methods of code and input selection criteria improvement are developed and appropriate recommendations made. Additional studies include the effect of updating test conditions from planned to actual and of sensitivity to downcomer nodalization

[en] Comparisons between RELAP4/MOD6, Update 3 code-calculated and experimental gravity-feed reflood systems performance are presented for one FLECHT-SET and one Semiscale MOD-1 test. Independent code verification techniques are now being developed and this study represents a preliminary effort applying structured criteria for developing computer models, selecting code input, and performing base-run analyses. Where deficiencies are indicated in the base-case representation of the experiment, methods of code and criteria improvement are developed and appropriate recommendations are made

[en] The capability to perform thermal-hydraulic analyses of a space reactor using the ATHENA computer code is demonstrated. The fast reactor, liquid-lithium coolant loops, and lithium-filled heat pipes of the preliminary General electric SP-100 design were modeled with ATHENA. Two demonstration transient calculations were performed simulating accident conditions. Calculated results are available for display using the Nuclear Plant Analyzer color graphics analysis tool in addition to traditional plots. ATHENA-calculated results appear reasonable, both for steady state full power conditions, and for the two transients. This analysis represents the first known transient thermal-hydraulic simulation using an integral space reactor system model incorporating heat pipes. 6 refs., 17 figs., 1 tab

[en] A series of thermal-hydraulic analyses were performed to investigate phenomena occurring during small break loss-of-coolant-accident (LOCA) sequences in a RESAR-3S pressurized water reactor. The analysis included simulations of plant behavior using the TRAC-PF1 and RELAP5/MOD2 computer codes. Series of calculations were performed using both codes for different break sizes. The analyses presented here also served an audit function in that the results shown here were used by the US Nuclear Regulatory Commission (NRC) as an independent confirmation of similar analyses performed by Westinghouse Electric Company using another computer code. 10 refs., 62 figs., 14 tabs

[en] Developmental verification studies were performed using RELAP4/MOD6 Update 1 to derive guidelines for selection of input options for reflood analyses. Primary emphasis was placed on determining entrainment and dispersed-flow heat-transfer options that would provide improved comparisons between code calculations and experimental data for the FLECHT forced-feed, low-flooding rate, cosine bundle tests. Preliminary guidelines have been established to assist the code user with input selection based on the effects of fundamental experimental parameters

[en] A recent safety concern for Westinghouse 4-loop pressurized water reactors (PWRs) is that, because of a seismic event, instrument tubes may be broken at the flux mapping seal table, resulting in an uncovering and heatup of the reactor core. This study's purpose was to determine the effects upon findings of a similar 1980 study if certain test variables changed. A 1980 US Nuclear Regulatory Commission (USNRC) analysis of PWR behavior used the RELAP4/MOD7 computer code to determine the effects of breaking instrument tubes at the reactor vessel lower plenum wall. The 1986 study discussed here was performed using RELAP5/MOD2, an advanced best-estimate computer code. Separate effects analyses investigated instrument tube pressure loss, heat loss, and tube nodalization effects on break flow. Systems effects analysis: (1) investigated the effects of changing the break location from the reactor vessel to the seal table, (2) compared RELAP4/MOD7 and RELAP5/MOD2 results for an identical transient, (3) verified a key finding from the 1980 analysis, and (4) investigated instrument tube ruptures in the Zion-1 PWR using best-estimate boundary and initial conditions. The outcome of these analyses permits adjustment of the 1980 analysis findings for instrument tube ruptures at the seal table and indicates the best-estimate response of a Washington PWR to the rupture of 25 small instrument tubes at the seal table

[en] Improved guidelines are developed for the selection of RELAP4/MOD6 reflood heat transfer options. The development, involving modifications to the original guidelines, assessed the effect of those modifications on RELAP4/MOD6 data comparisons using previously analyzed reflood experiments. The report also presents an evaluation of the application of the revised guidelines. Data comparisons between RELAP4/MOD6, using the original and revised guidelines, and experimental data are presented for Semiscale and FLECHT, forced-feed reflood tests and Semiscale and FLECHT-SET gravity-feed reflood tests. Because a general improvement was evident in data comparisons using the revised guidelines, their use is recommended in future calculations

[en] The RELAP4/MOD6 computer code is used for the analysis of the reactor core heat transfer during the reflooding phase of a postulated loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). The code requires the user to specify input parameters for the reflood heat transfer models. Results of previous comparisons of code calculations with experimental data have indicated no single selection of input parameters is adequate for a spectrum of tests and test facilities. This paper presents the development of revised quidelines and assesses the effect of those modifications on RELAP4/MOD6 data comparisons using previously analyzed reflood experiments. The paper also presents an assessment of the revised guidelines and the original guidelines against experimental data significantly different from previously analyzed tests

[en] Computer simulations were performed for an extensive selection of forced- and gravity-feed reflood experiments. This effort was a portion of the assessment procedure for the RELAP4/MOD6 thermal hydraulic computer code. A common set of guidelines, based on recommendations from the code developers, was used in determining the model and user-selected input options for each calculation. The comparison of code-calculated and experimental data was then used to assess the capability of the RELAP4/MOD6 code to model the reflood phenomena. As a result of the assessment, the guidelines for determining the user-selected input options were improved