ISSN:0029-5450    

DOI:10.13182/NT87-A33901

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

The U.S. Nuclear Regulatory Commission (NRC) has encouraged utilities to perform their own safety analyses to support reload applications, technical specification amendments, etc., to significantly improve their understanding of plant behavior. During the meetings of the Advisory Committee on Reactor Safeguards on the Wolf Creek project, Kansas Gas and Electric Company (KG&E) was urged to develop“a strong in-house capability for analyzing and understanding the nuclear-thermal-hydraulic behavior and systems performance.”KG&E fully intends to develop a strong in-house analytical capability and responded as such to the NRC. Part of this in-house analytical capability will be provided through the Safety Analysis Section. The development and application of system analysis is an integral part of the Safety Analysis Section.The development phases of achieving in-house system analysis capability are discussed. They include intermediate and long-term goals, a technical review of all non-loss-of-coolant accident transients performed by Westinghouse in Chap. 15 of the Final Safety Analysis Report, and the development of RETRAN system analysis models. Applications of system analysis are also discussed. Applications include an analysis of the plant loss-of-flow startup test to relax the acceptance criteria and a joint effort with Union Electric to reanalyze the steam generator tube rupture event for the NRC licensing commitment.

ISSN:0029-5450    

DOI:10.13182/NT87-A33902

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

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.

ISSN:0029-5450    

DOI:10.13182/NT87-A33903

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

A detailed RETRAN model of the secondary side of a Westinghouse model E steam generator has been developed to predict steam generator water level and primary side exit (cold-leg) temperature during startup testing and operational transients. These two parameters were identified as important in measuring the behavior of this steam generator RETRAN model. A nodalization study was performed to determine the minimum number of nodes (or control volumes) required in the secondary side to model the response of these two parameters for the transients of interest.The nodalization study began with a relatively detailed base model that represented each of the major secondary side regions. Included on the secondary side are the preheater region, upper and lower downcomer regions, primary steam separators, and leakage flow paths to account for the recirculation flow and flow branching. Eight modifications were developed from the base model to identify the sensitivity of various regions of the steam generator secondary side.Six transients were used as forcing functions to generate the response of the two steam generator parameters for each nodalization. The six transients represented a spectrum of secondary side initiated transients for which this model will be used. The impact on steam generator water level and cold-leg temperature due to a change in nodalization was evaluated for each transient.The nodalization study has identified the importance of the preheater region and the recirculation loop on the steam generator model performance. As long as secondary side water level remained above the tube bundle and below the steam dome, the two parameters of interest were insensitive to the nodalization of the upper tube bundle, lower downcomer, and steam dome regions.

ISSN:0029-5450    

DOI:10.13182/NT87-A33904

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

The ability to model and predict the steady-state and transient thermal-hydraulic behavior of the Westinghouse model D steam generator is an important prerequisite for performing safety and licensing analyses of Commonwealth Edison’s Byron and Braidwood nuclear power plants. A RETRAN model using ten volumes to represent the primary side and ten volumes to represent the secondary side is developed. The model is validated over a range of steady-state conditions and is used to calculate the pressure and level response to a main steam isolation valve closure using operational data to drive the transient and a basis for comparison. Sensitivity studies and a sample reload licensing calculation are performed to further determine the model’s capabilities. The results of the model development show that the RETRAN model is a viable tool for analysis of the model D steam generator’s steady-state and transient behavior. Examination of the model’s behavior during rapid secondary depressurization events and confirmation of the carryover behavior is recommended.

ISSN:0029-5450    

DOI:10.13182/NT87-A33905

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

Susquehanna Steam Electric Station Unit 2 (SSES-2) experienced three main turbine trips on high moisture separator drain tank level during initial startup testing in 1984. The SSES-2, a 3293-MW(thermal) boiling water reactor-4 with Mark II containment, uses two parallel nonreheating moisture separators between the high- and low-pressure turbine stages. Two of the main turbine trips and subsequent scrams occurred due to the high level in the“B”moisture separator drain tank during combined intermediate valve testing. The third trip was also initiated on the same signal, but during a recirculation pump run-back event.A task group was created to determine the cause of the level excursions and to make recommendations to reduce the severity of these transients. The RETRAN-02 computer code was used to evaluate the dynamic response of both the A and B moisture separator drain systems to determine the cause of the events, including why the level excursions only occurred in the B system. RETRAN was also used to evaluate the systems’dynamic response to several proposed corrective plant modifications.Based on the recommendations of the task group, modifications were made to SSES-2 during the precommercial operation outage. Startup testing following the outage proved the success of the modifications.

ISSN:0029-5450    

DOI:10.13182/NT87-A33906

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

As a part of the U.S. Nuclear Regulatory Commission’s unresolved safety issue A-45 decay heat removal program, the Los Alamos National Laboratory (LANL) performed a TRAC-PF1 simulation of the Calvert Cliffs Unit 1 pressurized water reactor in a cooldown to residual heat removal (RHR) entry conditions after a loss of off-site power (LOSP). A detailed four-loop TRAC model developed for the A-49 pressurized thermal shock program was used. The LANL results indicated an inability to both cool down and depressurize the primary system sufficiently to meet RHR entry conditions using only the atmospheric dump valves and auxiliary pressurizer spray.A RETRAN-02/MOD3 analysis was performed for the same transient, using assumptions consistent with those in the LANL analysis. A fast-running one-loop RETRAN model was selected because of the inherent symmetry of the transient. The RETRAN results compared well with sensitivity analyses indicating that the pressurizer model dominates the transient signatures.A best estimate RETRAN analysis of the cooldown was performed using a more accurate set of assumptions to better understand actual plant operational responses. These results indicate that RHR entry could be achieved after an LOSP using only existing plant equipment and procedures.

ISSN:0029-5450    

DOI:10.13182/NT87-A33907

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

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 relaxed setpoints. The steam generator low level, reactor trip on a turbine trip, and thermal margin/low pressure trip setpoints were successfully relaxed.

ISSN:0029-5450    

DOI:10.13182/NT87-A33908

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

Current boiling water reactors (BWRs) use active safety systems comprised of pumps, valves, motors, and diesel generators. These active safety systems (a) are a major cause of controversy in licensing power reactors because of reliability questions, (b) have high capital costs, and (c) require constant maintenance. An advanced BWR concept with passive emergency core cooling systems (PECOS) offers a basic alternative approach to reactor safety. In the PECOS-BWR, passive emergency core cooling is provided for the first 24 h by gravity flow of stored water released through fluidic valves that have no moving parts. Natural-draft air cooling removes heat from the core for longer periods.

ISSN:0029-5450    

DOI:10.13182/NT87-A33909

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

ISSN:0029-5450    

DOI:10.13182/NT87-A33910

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor