Jason Chao

Benchmark Calculations for the Doppler Coefficient of Reactivity

This paper reports on the Doppler coefficient of reactivity that is a crucial parameter in the evaluation of several transients in light water reactors (LWRs). It is relatively small in magnitude and cannot be measured directly in operating reactors. Doppler coefficients are presented for slightly idealized pressurized water reactor pin cells. These coefficients were calculated with the MCNP-3A continuous-energy Monte Carlo code using data taken directly from the ENDF/B-V nuclear data library. This combination represents the most rigorous analytical tool and the best nuclear data available. Consequently, these results comprise a set of numerical benchmarks that may be used to evaluate the accuracy of LWR lattice physics codes in predicting Doppler behavior at operating conditions. An example of one such evaluation, using the CELL-2 code, is included.

Pressurized Thermal Shock — An Integrated Analysis

The potential for long term neutron embrittlement of reactor vessels has been recognized for a number of years. Reactor vessel thermal shock is not a new concern, but with a growing number of plants approaching their mid-lives, it is a concern that must be understood and dealt with. Recent attention has focused on the performance of vessels during overcooling transients. This concern was designated as Unresolved Safety Issue A-49 by the Nuclear Regulatory Commission in December 1981. The USNRC staff has identified eight overcooling events of concern in U.S. PWRs (Table 1). The concern is currently limited to Pressurized Water Reactors. The Electric Power Research Institute (EPRI) has supported research on reactor vessel integrity for a number of years and has supported an extensive effort on reactor vessel pressurized thermal shock (PTS) over the last three years. In addition, EPRI has developed a linked set of computer codes to simulate the pressurized thermal shock transients and assess the integrity of the nuclear reactor vessels for various overcooling transients. This paper focuses on the integrated analysis approach being used by EPRI in performing such analysis.

Reducing scram frequency by modifying reactor setpoints for a Westinghouse four-loop plant

Several scram setpoints were analyzed for the purpose of reducing scram frequency in a Westinghouse four-loop plant. The results showed that the low-low steam generator (SG) level setpoint can be eliminated when reactor power is 50% or less during a loss of heat sink (LOHS) event. (The LOHS is the basis of this setpoint.) Without this setpoint, the reactor can still scram safely on either high pressurizer pressure or high pressurizer level without lifting the safety valves. The scram signal on the low SG level in coincidence with the signal from a mismatch of steam flow and feedwater flow can also be removed with no adverse effect on safety. This setpoint has never been included in the safety analysis. The results also showed that the power level above which the reactor should be scrammed when there is a turbine trip can be raised from its current value of 10% power to 50% when the condenser is available.

Simple mixing model for pressurized thermal shock applications

Abstract The phenomenon of fluid/thermal mixing in the cold leg and downcomer of a Pressurized Water Reactor (PWR) has been a critical issue related to the concern of pressurized thermal shock. The question of imperfect mixing arises when the possibility of cold emergency core cooling water contacting the vessel wall during an overcooling transient could produce thermal stresses large enough to initiate a flaw in a radiation embrittled vessel wall. The temperature of the fluid in contact with the vessel wall is crucial to a determination of vessel integrity since temperature affects both the stresses and the material toughness of the vessel material. A simple mixing model is described which was developed as part of the EPRI pressurized thermal shock program for evaluation of reactor vessel integrity.

An isoperibolic calorimeter to study electrochemical insertion of deuterium into palladium

The design and the operational characteristics of a new isoperibolic calorimeter that is developed to study the electrochemical insertion of deuterium into palladium are described. The design is simple and involves inexpensive materials to build. It possesses a number of distinct advantages that makes it suitable for thermal measurements in other electrochemical systems. It is insensitive to the nature and the location of the heat source within the electrochemical cell. The calibration constant is found to be stable with [+-]0.5% uncertainty over a wide range of input power levels up to 22 W. It also has the capability of operating over a wide temperature range. In principle, the calorimeter can be used up to 600[degrees]C, provided that the electrochemical cell design and materials are chosen appropriately. The design also provides flexibility to adjust the sensitivity of the calorimeter according to the needs of the system under study. 25 refs., 11 figs.

Analysis of a pressurized water reactor natural circulation transient at beginning of life

A pressurized water reactor (PWR) with a positive moderator temperature coefficient of reactivity is potentially susceptible to a severe overheating transient. This study identifies a scenario in which such a transient could occur and is similar in some respects to the accident at Chernobyl Unit 4. The scenario so identified is a natural circulation test at beginning of life under the assumption that all scrams are disabled. The results obtained demonstrate that a runaway power excursion does not occur and that the domestically designed PWR that was analyzed displays inherently safe behavior for the chosen scenario. The analysis is performed using two codes in tandem over three sequential stages of the analysis.

An empirical mixing model for pressurized thermal shock applications

Empirical correlations are developed for the local temperature and velocity distributions in the pressurized water reactor downcomer for pressurized thermal shock scenarios. The correlation is based on Creare test data and has been validated with Science Applications, Inc., experiments and COMMIX code calculations. It provides good agreement under pump flow and natural circulation conditions and gives a conservative estimate under stagnation conditions.

Analysis of Safety Injection Fluid Mixing in the Downcomer and Cold Leg of Pressurized Water Reactors

Four PWR designs were analyzed to obtain detailed three dimensional temperature and velocity distributions in the downcomers for the various pressurized thermal shock conditions.

RETRAN-02 and DYNODE-P analyses of a steam generator tube break transient

The RETRAN-02 and DYNODE-P thermal-hydraulic codes were compared against actual Prairie Island plant data from a steam generator tube break incident that occurred on October 2, 1979. The predictions from the code calculations compare well with actual plant behavior. The time of the break in the Prairie Island incident was found to be about 260 s prior to scram with an initial break flow of 625 gal/min. Discharge coefficients are recommended for the calculations of critical flow from the break with extended Henry-Fauske and Moody critical flow models. In addition, a linear correlation was developed to predict the break flow with a given system depressurization rate for a Westinghouse two-loop plant.

Safety Analyses Using RETRAN-02 with Relaxed Trip Setpoints on Combustion Engineering Reactors

The reactor protection system (both analog- and digital-based) setpoints of Combustion Engineering nuclear steam supply systems were examined to determine the feasibility of scram reduction by relaxing these setpoints. Representative safety analyses, using RETRAN-02, were performed to demonstrate that acceptable results were obtained with the relaxes setpoints. The steam generator low level, reactor trip on a turbine trip, and thermal margin/low pressure trip setpoints were successfully relaxed.