[en] RELAP5 is a pressurized water reactor system transient simulation code for use in nuclear power plant safety analysis. The latest version, MOD2, may be used to simulate and study a wide variety of abnormal events, including loss-of-coolant accidents, operational transients, and transients in which the entire secondary system must be modeled. In this paper, a basic overview of the code is given, its assessment and application illustrated, and progress toward its use as a high fidelity simulator described. 7 refs., 7 figs

[en] M1 transition rates are strongly structure dependent and may therefore reveal the structure in a quasicontinuum of nuclear excited states. The theoretical interpretations of low-energy stretched dipole bumps are briefly reviewed. Furthermore, it is suggested that there are higher-energy M1 bumps with a signficant unstretched component, and that the related shell effects influence the cooling which feeds the yrast cascade

[en] Flow oscillations between the downcomer and the electrically heated core during reflood have been observed in reflood experimental facilities such as Semiscale and FLECHT-SET (Westinghouse). The purpose of the study described was to investigate the primary mechanism responsible for the initiation and continuation of the oscillations during reflood using the RELAP4/MOD6 code. The study produced significant results concerning the accuracy of RELAP in predicting the dynamic features of density-wave and pressure drop oscillations, and was conducted on the German three-loop PRIMAR KREISLAUF (PLK) Reflood Experimental System. RELAP4/MOD6 is an analytical computer code which can be used for best estimate analysis of PWR or BWR reactor system blowdown and reflood response to a postulated LOCA

[en] RELAP4/MOD6 is an analytical computer code which can be used for best-estimate analysis of LWR reactor system blowdown and reflood response to a postulated LOCA. In this study, flow oscillations in the PKL reflood test K5A were investigated using RELAP4/MOD6. Both calculated and measured oscillations exhibited transient characteristics of density-wave and pressure-drop oscillations. The calculated average core mixture level rising rate agrees closely with the test data. Several mechanisms which appear to be responsible for initiation and continuation of calculated or experimental reflood flow oscillations are (a) the coupling between the vapor generation in the core channel and the U-tube geometrical arrangement of a downcomer and a heated core; (b) the inherent low core inlet resistance and the high system outlet resistance; (c) the dependence of heat transfer rate on mass flow rate especially in the dispersed flow ially in the dispersed flow regime; (d) the amount of the liquid entrainment fraction of the heated core channel

[en] This report documents, as of September, 1986, the investigation of the failure or degradation of some of the advanced two-phase flow instruments supplied by the United States Nuclear Regulatory Commission (USNRC) to the German Upper Plenum Test Facility (UPTF). These instruments include Tie-Plate Drag Bodies (DBs), Breakthrough Detectors (BTDs), Loop Drag Disc (DD) paddles, Fluid Distribution Grid (FDG) sensors, and Liquid Level Detector (LLD) sensors. The exact causes for these instrument degradations or failures are not known, but several potential causes have been identified. For DBs and BTDs, the primary mechanism for the degradation appears to be a leakage in the Inconel 600 strain gage encapsulation and the subsequent burnout of the strain gage elements. Excessive loads appear to be the cause of the degradation or failure of the drag discs. The degradation cause for most of the FDGs and LLDs may be either steam/water erosion or mechanical abrasion of the sapphire sensor tips. However, some of the FDG tips were found to be cracked also. The corrective actions are being directed towards identification of the primary causes for the instrument degradation or failure and methods of preventing recurrance and toward minimizing the impact on the test program. All possible action items are being reviewed to arrange them in terms of priority and the likelihood of success so that the best results can be obtained under the constraints of a fixed amount of resources and limited time

[en] RELAP4 is a computer code which can be used for the transient thermal hydraulic analysis of light water reactors and related systems. RELAP4/MOD6 includes many new analytical models which were developed primarily for the analysis of the reflood phase of a PWR loss-of-coolant accident (LOCA) transient. The key feature forming the basis for the MOD6 reflood calculation is a unique moving finite differenced heat conductor. The development and application of the moving heat conductor mesh for use in reflood analysis are described

[en] By injecting coolant with a high pressure emergency core cooling system, and removing the heated/vaporized fluid by way of the pressurizer power operated relief valve, primary feed and bleed cooling denotes an operation whereby reactor core cooling is maintained. This paper presents the results from an experimental and analytical study that includes a simplified analysis of mass and energy balances associated with the feed and bleed, examination of test data from the Semiscale system, RELAP5 code analyses of both Semiscale and a four-loop Westinghouse plant, and the primary coolant system behavior for a transient that leads to the need for feed and bleed. Examination of the parameters that govern a stable feed and bleed operation identifies four key parameters such as: (a) core decay heat, (b) cooling water injection capacity, (c) power operated relief valve (PORV) energy removal rate, and (d) PORV mass removal rate. A simplified analytical approach to determining if stable feed and bleed is feasible, has been developed and corroborated by experimental data and computer code calculations

[en] The attenuation factor of the Coriolis off diagonal matrix element has been found to be very sensitive to the pairing strength. About 10% extra reduction of the attenuation factor coming from the recoil term may be expected in the pairing plus recoil model for some Coriolis matrix elements, if a reasonable pairing gap parameter is taken. (orig.)

[en] An analytical method using a critical flow correlation has been developed for small break loss-of-coolant accident (LOCA) analyses. To assess the proposed method, variety of experimental data obtained from both sharp-edged orifices and tubes were used. The maximum discrepancy in most cases was found to be within 7%. It is recognized, however, that if the proposed analytical approach is to become fully applicable to a typical pressurized water reactor plant under LOCA conditions, more test data are definitely needed especially at high system pressure (>9 MPa) and under subcooled flow conditions

[en] RELAP4/MOD6 is a computer code developed specifically to predict the transient thermal-hydraulic behavior of a PWR and related experimental reactor systems during reflood phase of postulated LOCA conditions. A ''blind'' test prediction for the German PKL reflood Test K5A was conducted using RELAP4/MOD6. The results of the prediction were in good agreement with experimental data indicating that RELAP4/MOD6 is capable of predicting transient reflood phenomena in the 200 percent cold-leg break test configuration of the PKL reflood facility