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Constitutive relations in TRAC-P1A

Description: The purpose of this document is to describe the basic thermal-hydraulic models and correlations that are in the TRAC-P1A code, as released in March 1979. It is divided into two parts, A and B. Part A describes the models in the three-dimensional vessel module of TRAC, whereas Part B focuses on the loop components that are treated by one-dimensional formulations. The report follows the format of the questions prepared by the Analysis Development Branch of USNRC and the questionnaire has been attached to this document for completeness. Concerted efforts have been made in understanding the present models in TRAC-P1A by going through the FORTRAN listing of the code. Some discrepancies between the code and the TRAC-P1A manual have been found. These are pointed out in this document. Efforts have also been made to check the TRAC references for the range of applicability of the models and correlations used in the code. 26 refs., 5 figs., 1 tab.
Date: August 1, 1980
Creator: Rohatgi, U.S. & Saha, P.
Partner: UNT Libraries Government Documents Department

Uncertainties in modelling and scaling of critical flows and pump model in TRAC-PF1/MOD1

Description: The USNRC has established a Code Scalability, Applicability and Uncertainty (CSAU) evaluation methodology to quantify the uncertainty in the prediction of safety parameters by the best estimate codes. These codes can then be applied to evaluate the Emergency Core Cooling System (ECCS). The TRAC-PF1/MOD1 version was selected as the first code to undergo the CSAU analysis for LBLOCA applications. It was established through this methodology that break flow and pump models are among the top ranked models in the code affecting the peak clad temperature (PCT) prediction for LBLOCA. The break flow model bias or discrepancy and the uncertainty were determined by modelling the test section near the break for 12 Marviken tests. It was observed that the TRAC-PF1/MOD1 code consistently underpredicts the break flow rate and that the prediction improved with increasing pipe length (larger L/D). This is true for both subcooled and two-phase critical flows. A pump model was developed from Westinghouse (1/3 scale) data. The data represent the largest available test pump relevant to Westinghouse PWRs. It was then shown through the analysis of CE and CREARE pump data that larger pumps degrade less and also that pumps degrade less at higher pressures. Since the model developed here is based on the 1/3 scale pump and on low pressure data, it is conservative and will overpredict the degradation when applied to PWRs.
Date: January 1, 1987
Creator: Rohatgi, U.S. & Yu, Wen-Shi
Partner: UNT Libraries Government Documents Department

Assessment of TRAC and RELAP5 codes with ORNL POST-CHF tests. [PWR]

Description: Brookhaven National Laboratory is involved in assessing thermohydraulic models in various advanced codes such as TRAC-PF1 (Version 7), TRAC-BD1 (Version 12) and RELAP5/MOD1/CY=14. These codes have two fluid formulations and detailed descriptions of wall heat transfer regimes. These wall heat transfer models and correlations were developed using results from separate effect tests for specific fluid conditions and should be assessed for conditions which may exist in the reactor at normal and abnormal operation. In this paper the effort is concentrated on evaluating the capabilities of these codes in predicting the critical heat flux situation in the rod bundle geometry. The tests selected for this purpose are Oak Ridge Post-CHF tests. Oak Ridge Post-CHF tests consist of a series of high pressure and high temperature steady-state experiments and were conducted with water flowing upward through an 8 x 8 rod bundle with rod diameter and rod pitch typical of PWRs with 17 x 17 fuel assemblies.
Date: January 1, 1984
Creator: Rohatgi, U.S.; Neymotin, L. & Pu, J.
Partner: UNT Libraries Government Documents Department

Physical interpretation of geysering phenomena and periodic boiling instability at low flows

Description: Over 30 years ago, Griffith showed that unstable and periodic initial boiling occurred in stagnant liquids in heated pipes coupled to a cooler or condensing plenum volume. This was called ``geysering``, and is a similar phenomenon to the rapid nucleation and voiding observed in tubes filled with superheated liquid. It is also called ``bumping`` when non-uniformly heated water or a chemical suddenly boils in laboratory glassware. In engineering, the stability and predictability has importance to the onset of bulk boiling in a natural and forced circulation loops. The latest available data show the observed stability and periodicity of the onset of boiling flow when there is a plenum, multiple heated channels, and a sustained subcooling in a circulating loop. We examine the available data, both old and new, and develop a new theory to illustrate the simple physics causing the observed periodicity of the flow. We examine the validity of the theory by comparison to all the geysering data, and develop a useful and simple correlation. We illustrate the equivalence of the onset of geysering to the onset of static instability in subcooled boiling. We also derive the stability boundary for geysering, utilizing turbulent transport analysis to determine the effects of pressure and other key parameters. This new result explains the greater stability region observed at higher pressures. The paper builds on the 30 years of quite independent thermal hydraulic work that is still fresh and useful today. We discuss the physical interpretation of geysering onset with a consistent theory, and show where refinements would be useful to the data correlations.
Date: March 1996
Creator: Duffey, R. B. & Rohatgi, U. S.
Partner: UNT Libraries Government Documents Department

Considerations for realistic ECCS evaluation methodology for LWRs

Description: This paper identifies the various phenomena which govern the course of large and small break LOCAs in LWRs, and affect the key parameters such as Peak Clad Temperature (PCT) and timing of the end of blowdown, beginning of reflood, PCT, and complete quench. A review of the best-estimate models and correlations for these phenomena in the current literature has been presented. Finally, a set of models have been recommended which may be incorporated in a present best-estimate code such as TRAC or RELAP5 in order to develop a realistic ECCS evaluation methodology for future LWRs and have also been compared with the requirements of current ECCS evaluation methodology as outlined in Appendix K of 10CFR50. 58 refs.
Date: January 1, 1985
Creator: Rohatgi, U.S.; Saha, P. & Chexal, V.K.
Partner: UNT Libraries Government Documents Department

Quality assurance of PTS thermal hydraulic calculations at BNL. [Pressurized Thermal Shock]

Description: Rapid cooling of the reactor pressure vessel at high pressure has a potential of challenging the vessel integrity. This phenomenon is called overcooling or Pressurized Thermal Shock (PTS). The Nuclear Regulatory Commission (NRC) has selected three plants representing three types of PWRs in use for detailed PTS study. Oconee-1 (B and W), Calvert Cliffs (C.E.), and H.B. Robinson (Westinghouse). The Brookhaven National Laboratory (BNL) has been requested by NRC to review and compare the input decks developed at LANL and INEL, and to compare and explain the differences between the common calculations performed at these two laboratories. However, for the transients that will be computed by only one laboratory, a consistency check will be performed. So far only Oconee-1 calculations have been reviewed at BNL, and the results are presented here.
Date: January 1, 1983
Creator: Rohatgi, U.S.; Pu, J.; Jo, J. & Saha, P.
Partner: UNT Libraries Government Documents Department

Review of thermal-hydraulic calculations for Calvert Cliffs and H. B. Robinson PTS study. [Pressurized thermal shock]

Description: Thermal-hydraulic transient calculations performed by LANL using the TRAC-PF1 code and by INEL using the RELAP5 code for the USNRC pressurized thermal shock (PTS) study of the Calvert Cliffs and H.B. Robinson Nuclear Power Plants have been reviewed at BNL including the input decks and steady state calculations. Furthermore, six transients for each plant have been selected for the in-depth review. Simple hand calculations based on the mass and energy balances of the entire reactor system, have been performed to predict the temperature and pressure of the reactor system, and the results have been compared with those obtained by the code calculation. In general, the temperatures and pressures of the primary system calculated by the codes have been very reasonable. The secondary pressures calculated by TRAC appear to indicate that the codes have some difficulty with the condensation model and further work is needed to assess the code calculation of the U-tube steam generator pressure when the cold auxiliary feedwater is introduced to the steam generator. However, it is not expected that this uncertainty would affect the transient calculations significantly.
Date: January 1, 1984
Creator: Jo, J.H.; Yuelys-Miksis, C. & Rohatgi, U.S.
Partner: UNT Libraries Government Documents Department

Assessment of similarity of HFBR (High Flux Beam Reactor) with separate effects test

Description: A Separate Effects Test (SET) facility was constructed in 1963 to demonstrate the feasibility of the HFBR design and to determine the core power limits for a safe flow reversal event. The objective of the task reported here is to review the capability of the test to scale the dominant phenomena in the HFBR during a flow reversal event and the applicability of the range of the power level obtained from the test to the HFBR. The conclusion of this report was that the flow during the flow reversal event will not be similar in the two facilities. The causes of the dissimilarity are the differences in the core inlet friction, bypass path friction, the absence of the check valve in the test, and the materials used to represent the fuel plates. The impact of these differences is that the HFBR will undergo flow reversal sooner than the test and will have a higher flow rate in the final Natural Circulation Period. The shorter duration of the flow reversal event will allow less time for the plate to heat up and the larger flow in the Natural Circulation Period will lead to higher critical heat flux limits in the HFBR than in the test. Based on these observations, it was concluded that the HFBR can undergo flow reversal safely for heat fluxes up to 46,700 (BTU/hr ft{sup 2}), the heat flux limit obtained from the 1963 test.
Date: November 1, 1990
Creator: Rohatgi, U.S. & Slovik, G.C.
Partner: UNT Libraries Government Documents Department

RELAP5/MOD2. 5 analysis of the HFBR (High Flux Beam Reactor) for a loss of power and coolant accident

Description: A set of postulated accidents were evaluated for the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory. A loss of power accident (LOPA) and a loss of coolant accident (LOCA) were analyzed. This work was performed in response to a DOE review that wanted to update the understanding of the thermal hydraulic behavior of the HFBR during these transients. These calculations were used to determine the margins to fuel damage at the 60 MW power level. The LOPA assumes all the backup power systems fail (although this event is highly unlikely). The reactor scrams, the depressurization valve opens, and the pumps coast down. The HFBR has down flow through the core during normal operation. To avoid fuel damage, the core normally goes through an extended period of forced down flow after a scram before natural circulation is allowed. During a LOPA, the core will go into flow reversal once the buoyancy forces are larger than the friction forces produced during the pump coast down. The flow will stagnate, reverse direction, and establish a buoyancy driven (natural circulation) flow around the core. Fuel damage would probably occur if the critical heat flux (CHF) limit is reached during the flow reversal event. The RELAP5/MOD2.5 code, with an option for heavy water, was used to model the HFBR and perform the LOPA calculation. The code was used to predict the time when the buoyancy forces overcome the friction forces and produce upward directed flow in the core. The Monde CHF correlation and experimental data taken for the HFBR during the design verification phase in 1963 were used to determine the fuel damage margin. 20 refs., 40 figs., 11 tabs.
Date: May 1, 1990
Creator: Slovik, G.C.; Rohatgi, U.S. & Jo, Jae.
Partner: UNT Libraries Government Documents Department

Natural circulation and stability limits in advanced plants: The Galilean law

Description: In our previous papers natural circulation instability problem have been examined. In particular, we analytically demonstrated the link between critical heart flux (CHF) and instability, and the limitations of the present CHF data. We derived an analytical solution for the static instability in natural circulation. We proceed in this paper to build on these novel ideas and physical principles. We focus on the role of gravity in the analysis and prediction of instability. We derive the non-dimensional correlation of natural circulation flowrate, and link this with our physical analysis of instability in parallel-channel systems. We derive the exact form of the unstable limit, and discuss the important effect of zero gravity on the solution. For static instability, the result is a cubic equation for the quality. We solve for the limiting cases to obtain a new explicit result. The effect of gravity on upflow and downflow is shown to be significant for high subcoolings. By combining the natural-circulation and stability results, we demonstrate a new result which analytically describes the limits on natural-circulation operation. This result is relevant to small breaks and transients, as well as to defining the design envelope for safe operation.
Date: April 1, 1994
Creator: Rohatgi, U. S. & Duffey, R. B.
Partner: UNT Libraries Government Documents Department

A level control model for BWR emergency procedure guidelines

Description: The level control during an Anticipated Transient Without Scram (ATWS) event in a BWR as prescribed in the Emergency Procedure Guidelines (EPG) is a difficult task for the operator for he has to keep the water level at the top of active fuel (TAF) without uncovering the reactor core. Also the computer simulation of EPG level control will require many trial and error calculations with the Emergency Core Cooling System (ECCS). A level control system model has been developed and implemented in the RAMONA-4B code in order to simulate the EPG level control without iterations. The model has been extensively tested and the results demonstrate that the model can simulate the EPG level control strategy. The calculations also show that the level control system will speed up the boron circulation to shut down the reactor sooner than the manual control. Furthermore, the suppression pool temperature is predicted to remain within the Technical Specification limit during a MSIV closure ATWS with the proposed level control strategy. 3 refs., 11 figs., 1 tab.
Date: October 1, 1996
Creator: Cheng, H.S. & Rohatgi, U.S.
Partner: UNT Libraries Government Documents Department

Modeling of two-phase flow instabilities during startup transients utilizing RAMONA-4B methodology

Description: RAMONA-4B code is currently under development for simulating thermal hydraulic instabilities that can occur in Boiling Water Reactors (BWRs) and the Simplified Boiling Water Reactor (SBWR). As one of the missions of RAMONA-4B is to simulate SBWR startup transients, where geysering or condensation-induced instability may be encountered, the code needs to be assessed for this application. This paper outlines the results of the assessments of the current version of RAMONA-4B and the modifications necessary for simulating the geysering or condensation-induced instability. The test selected for assessment are the geysering tests performed by Prof Aritomi (1993).
Date: October 1, 1996
Creator: Paniagua, J.; Rohatgi, U.S. & Prasad, V.
Partner: UNT Libraries Government Documents Department

RAMONA-4B code for BWR systems analysis

Description: The RAMONA-4B code is a coupled thermal-hydraulic, 3D kinetics code for plant transient analyses of a complete Boiling Water Reactor (BWR) system including Reactor Pressure Vessel (RPV), Balance of Plant (BOP) and containment. The complete system representation enables an integrated and coupled systems analysis of a BWR without recourse to prescribed boundary conditions.
Date: December 31, 1996
Creator: Cheng, H. S. & Rohatgi, U. S.
Partner: UNT Libraries Government Documents Department

Power generation costs and ultimate thermal hydraulic power limits in hypothetical advanced designs with natural circulation

Description: Maximum power limits for hypothetical designs of natural circulation plants can be described analytically. The thermal hydraulic design parameters are those which limit the flow, being the elevations, flow areas, and loss coefficients. WE have found some simple ``design`` equations for natural circulation flow to power ratio, and for the stability limit. The analysis of historical and available data for maximum capacity factor estimation shows 80% to be reasonable and achievable. The least cost is obtained by optimizing both hypothetical plant performance for a given output,a nd the plant layout and design. There is also scope to increase output and reduce cost by considering design variations of primary and secondary pressure, and by optimizing component elevations and loss coefficients. The design limits for each are set by stability and maximum flow considerations, which deserve close and careful evaluation.
Date: December 31, 1996
Creator: Duffey, R.B. & Rohatgi, U.S.
Partner: UNT Libraries Government Documents Department

Simulation of SBWR startup transient and stability

Description: The Simplified Boiling Water Reactor (SBWR) designed by General Electric is a natural circulation reactor with enhanced safety features for potential accidents. It has a strong coupling between power and flow in the reactor core, hence the neutronic coupling with thermal-hydraulics is specially important. The potential geysering instability during the early part of a SBWR startup at low flow, low power and low pressure is of particular concern. The RAMONA-4B computer code developed at Brookhaven National Laboratory (BNL) for the SBWR has been used to simulate a SBWR startup transient and evaluate its stability, using a simplified four-channel representation of the reactor core for the thermal-hydraulics. This transient was run for 20,000 sec (5.56 hrs) in order to cover the essential aspect of the SBWR startup. The simulation showed that the SBWR startup was a very challenging event to analyze as it required accurate modeling of the thermal-hydraulics at low pressures. This analysis did not show any geysering instability during the startup, following the startup procedure as proposed by GE.
Date: June 1, 1998
Creator: Cheng, H.S.; Khan, H.J. & Rohatgi, U.S.
Partner: UNT Libraries Government Documents Department

Evaluation of the effects of initial conditions on transients in PUMA

Description: A Simplified Boiling Water Reactor (SBWR) is the latest Boiling Water Reactor (BWR) designed by the General Electric (GE). Major differences between the SBWR and the currently operating BWRs include the use of passive gravity-driven systems in the SBWR for emergency cooling of the vessel and containment. In order to investigate the phenomena expected during a Loss of Coolant Accident (LOCA), Nuclear Regulatory Commission (NRC) has sponsored an integral scaled-test facility, called Purdue University Multidimensional Integral Test Assembly (PUMA). The facility models all the major safety-related components of SBWR. Two PUMA initialization calculations were performed to assist the Purdue University in establishing test initialization procedures. Both calculations were based on the initial conditions obtained from SBWR LOCA simulation. In the base case, a complete separation between vapor and liquid was assumed, with all the water in the lower part of the Reactor Pressure Vessel (RPV) and all the vapor above it. In the sensitivity case, the water inventory was distributed in the vessel in the same way as in the SBWR at 1.034 MPa, which is the initial pressure for PUMA facility. Purdue University plans to initialize the PUMA tests as in the base case. The sensitivity calculation is performed to provide assurance that this mode of initialization is adequate. It also provides information on possible differences in the progress of transients. The paper will discuss the differences in the early part of the transient. The conclusion from this study will also apply to many integral facilities which simulate the reactor transients form the middle of the transient.
Date: June 1, 1996
Creator: Parlatan, Y.; Jo, J. & Rohatgi, U.S.
Partner: UNT Libraries Government Documents Department

An assessment of RELAP5 MOD3.1.1 condensation heat transfer modeling with GIRAFFE heat transfer tests

Description: RELAP5 MOD3.1.1 is being used to simulate Loss of Coolant Accidents (LOCA) for the Simplified Boiling Water Reactor (SBWR) being proposed by General Electric (GE). One of the major components associated with the SBWR is the Passive Containment Cooling System (PCCS) which provides the long-term heat sink to reject decay heat. The RELAP5 MOD3.1.1 code is being assessed for its ability to represent accurately the PCCS. Data from the Phase 1, Step 1 Heat Transfer Tests performed at Toshiba`s Gravity-Driven Integral Full-Height Test for Passive Heat Removal (GIRAFFE) facility will be used for assessing the ability of RELAP5 to model condensation in the presence of noncondensables. The RELAP5 MOD3.1.1 condensation model uses the University of California at Berkeley (UCB) correlation developed by Vierow and Schrock. The RELAP5 code uses this heat transfer coefficient with the gas velocity effect multiplier being limited to 2. This heat transfer option was used to analyze the condensation heat transfer in the GIRAFFE PCCS heat exchanger tubes in the Phase 1, Step 1 Heat Transfer Tests which were at a pressure of 3 bar and had a range of nitrogen partial pressure fractions from 0.0 to 0.10. The results of a set of RELAP5 calculations al these conditions were compared with the GIRAFFE data. The effects of PCCS cell nodings on the heat transfer process were also studied. The UCB correlation, as implemented in RELAP5, predicted the heat transfer to {+-}5% of the data with a three-node model. The three-node model has a large cell in the entrance region which smeared out the entrance effects on the heat transfer, which tend to overpredict the condensation. Hence, the UCB correlation predicts condensation heat transfer in the presence of noncondensable gases with only a coarse mesh. The cell length term in the condensation heat transfer correlation implemented ...
Date: September 1995
Creator: Boyer, B. D.; Parlatan, Y.; Slovik, G. C. & Rohatgi, U. S.
Partner: UNT Libraries Government Documents Department

BWR stability analysis at Brookhaven National Laboratory

Description: 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 are dealt with in this paper. The other four questions address local, out-of-phase oscillations and matters of instrumentation; they fall outside the scope of the work reported here. It was the purpose of the work documented in this report to answer ten of the fourteen NRC-stipulated questions. Nine questions are answered by analyzing the LaSalle-2 instability and related BWR transients with the BNL Engineering Plant Analyzer (EPA) and by performing an uncertainty assessment of the EPA predictions. The tenth question is answered on the basis of first principles. The ten answers are summarized
Date: January 1, 1991
Creator: Wulff, W.; Cheng, H.S.; Mallen, A.N. & Rohatgi, U.S.
Partner: UNT Libraries Government Documents Department

Assessment of engineering plant analyzer with Peach Bottom 2 stability tests

Description: Engineering Plant Analyzer (EPA) has been developed to simulate plant transients for Boiling Water Reactor (BWR). Recently, this code has been used to simulate LaSalle-2 instability event which was initiated by a failure in the feed water heater. The simulation was performed for the scram conditions and for the postulated failure in the scram. In order to assess the capability of the EPA to simulate oscillatory flows as observed in the LaSalle event, EPA has been benchmarked with the available data from the Peach Bottom 2 (PB2) Instability tests PT1, PT2, and PT4. This document provides a description of these tests.
Date: January 1, 1992
Creator: Rohatgi, U.S.; Mallen, A.N.; Cheng, H.S. & Wulff, W.
Partner: UNT Libraries Government Documents Department

Assessment of RAMONA-3B methodology with FRIGG dynamic tests

Description: The computer codes used at Brookhaven National Laboratory to compute BWR safety parameters are the Engineering Plant Analyzer (EPA) and RAMONA-3B/MOD1. Both codes have the same methodology for modeling thermal hydraulic phenomena: drift-flux formulation, two-phase multipliers for the wall friction and form losses calculations, and the momentum integral approach for spatial integration of the loop momentum equations. Both codes use explicit integration methods for solving ordinary differential equations. It is concluded that both the codes are capable of modelling the instability problems for a BWR. The accuracy of thermohydraulics codes predictions was assessed by modelling oscillatory FRIGG tests. Nodalizations studies showed that 24 axial nodes were sufficient for a converged solution, 12 axial nodes produced an error of 4.4% in the gain of the power to flow transfer function. The code predicted consistently the effects of power and inlet subcooling on gain and system resonance frequency. The comparisons showed that the code predicted the peak gains with a mean difference from experiments of 7% {plus minus} 30% for all the tests modeled. The uncertainty in the experimental data is {minus}11% to +12%. The mean difference in the predicted frequency at the peak gain is {minus}6% {plus minus} 14%.
Date: January 1, 1990
Creator: Rohatgi, U.S.; Neymotin, L.Y. & Wulff, W.
Partner: UNT Libraries Government Documents Department

Performance characterization of isolation condenser of SBWR

Description: A systematic study of the performance of the Isolation Condenser (IC) for a conceptual design of SBWR is presented. The objective of the IC is to passively remove heat and control the pressure variation in the Reactor Pressure Vessel (RPV). According to the observed trends, the IC cooling capacity and condensate flow can independently influence the ultimate performance of the IC. The transient pressure profile for the IC reaches different equilibrium values for each of the cases analyzed. The absolute magnitude of these values are a function of the cooling capacity and flow rates. With appropriate control of the liquid flow loss coefficients, the performance of the IC can be well predicted. Due to the lack of useful data, this study is limited to the numerical simulation of the IC.
Date: January 1, 1992
Creator: Khan, H.J. & Rohatgi, U.S.
Partner: UNT Libraries Government Documents Department

Prediction of BWR performance under the influence of Isolation Condenser-using RAMONA-4 code

Description: The purpose of the Boiling Water Reactor (BWR) Isolation Condenser (IC) is to passively control the reactor pressure by removing heat from the system. This type of control is expected to reduce the frequency of opening and closing of the Safety Relief Valves (SRV). A comparative analysis is done for a BWR operating with and without the influence of an IC under Main Steam Isolation Valve (MSIV) closure. A regular BWR, with forced flow and high thermal power, has been considered for analysis. In addition, the effect of ICs on the BWR performance is studied for natural convection flow at lower power and modified riser geometry. The IC is coupled to the steam dome for the steam inlet flow and the Reactor Pressure Vessel (RPV) near the feed water entrance for the condensate return flow. Transient calculations are performed using prescribed pressure set points for the SRVs and given time settings for MSIV closure. The effect of the IC on the forced flow is to reduce the rate of pressure rise and thereby decrease the cycling frequency ofthe SRVS. This is the primary objective of any operating IC in a BWR (e.g. Oyster Creek). The response of the reactor thermal and fission power, steam flow rate, collapsed liquid level, and core average void fraction are found to agree with the trend of pressure. The variations in the case of an active IC can be closely related to the creation of a time lag and changes in the cycling frequency of the SRVS. An analysis for natural convection flow in a BWR indicates that the effect of an IC on its transient performance is similar to that for the forced convection system. In this case, the MSIV closure, has resulted in a lower peak pressure due to the magnitude of reduced ...
Date: January 1, 1992
Creator: Khan, H.J.; Cheng, H.S. & Rohatgi, U.S.
Partner: UNT Libraries Government Documents Department

Prediction of BWR performance under the influence of Isolation Condenser-using RAMONA-4 code

Description: The purpose of the Boiling Water Reactor (BWR) Isolation Condenser (IC) is to passively control the reactor pressure by removing heat from the system. This type of control is expected to reduce the frequency of opening and closing of the Safety Relief Valves (SRV). A comparative analysis is done for a BWR operating with and without the influence of an IC under Main Steam Isolation Valve (MSIV) closure. A regular BWR, with forced flow and high thermal power, has been considered for analysis. In addition, the effect of ICs on the BWR performance is studied for natural convection flow at lower power and modified riser geometry. The IC is coupled to the steam dome for the steam inlet flow and the Reactor Pressure Vessel (RPV) near the feed water entrance for the condensate return flow. Transient calculations are performed using prescribed pressure set points for the SRVs and given time settings for MSIV closure. The effect of the IC on the forced flow is to reduce the rate of pressure rise and thereby decrease the cycling frequency ofthe SRVS. This is the primary objective of any operating IC in a BWR (e.g. Oyster Creek). The response of the reactor thermal and fission power, steam flow rate, collapsed liquid level, and core average void fraction are found to agree with the trend of pressure. The variations in the case of an active IC can be closely related to the creation of a time lag and changes in the cycling frequency of the SRVS. An analysis for natural convection flow in a BWR indicates that the effect of an IC on its transient performance is similar to that for the forced convection system. In this case, the MSIV closure, has resulted in a lower peak pressure due to the magnitude of reduced ...
Date: December 31, 1992
Creator: Khan, H. J.; Cheng, H. S. & Rohatgi, U. S.
Partner: UNT Libraries Government Documents Department

Independent code assessment at BNL in FY 1982. [TRAC-PF1; RELAP5/MOD1; TRAC-BD1]

Description: Independent assessment of the advanced codes such as TRAC and RELAP5 has continued at BNL through the Fiscal Year 1982. The simulation tests can be grouped into the following five categories: critical flow, counter-current flow limiting (CCFL) or flooding, level swell, steam generator thermal performance, and natural circulation. TRAC-PF1 (Version 7.0) and RELAP5/MOD1 (Cycle 14) codes were assessed by simulating all of the above experiments, whereas the TRAC-BD1 (Version 12.0) code was applied only to the CCFL tests. Results and conclusions of the BNL code assessment activity of FY 1982 are summarized below.
Date: January 1, 1982
Creator: Saha, P.; Rohatgi, U.S.; Jo, J.H.; Neymotin, L.; Slovik, G. & Yuelys-Miksis, C.
Partner: UNT Libraries Government Documents Department