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The ’80s and ’90s: disasters, reflection, and revisionism
Published in J. Mangano Joseph, Low-Level Radiation and Immune System Damage, 2018
Shortly before 4 a.m., Frederick Scheimann, the Unit 2 shift foreman, went to the basement to see how work on the pipe was progressing. He discussed the problem with (technician Donald) Miller and with Harold Farst, the other technician on duty. Scheimann climbed up on top of a larger pipe so that he could look into the polisher through a glass window. “All of a sudden, I started hearing loud, thunderous noises, like a couple of freight trains,” he said later. He jumped down from the pipe, heard the words “Turbine trip, turbine trip” over a loudspeaker, and rushed to the control room. The maintenance crew working on the polisher had accidentally choked off the flow in the main feedwater system, forcing Unit 2’s generating equipment — its turbine and reactor, which had been operating at ninety-seven percent of full power — to shut down. The equipment was suddenly tripped at thirty seven seconds past 4 a.m.1
Steam Generators for Nuclear Power Plants
Published in Maurizio Cumo, Antonio Naviglio, Thermal Hydraulic Design of Components for Steam Generation Plants, 1991
The upset conditions refer to operations occurring during component malfunctions, not resulting in an operational impairment. These transients, listed in Table 10 together with the expected number of occurrences during the lifetime of the SG, must be considered in the mechanical analysis. As a representative transient of this class, the “loss of load” (or turbine trip) is analyzed. The boundary conditions used in the analysis are shown in Figure 26 (the primary flow rate is assumed constant); the calculation results are presented in Figures 27 and 28, reporting, respectively: The exchanged thermal power (Figure 27)The cold leg temperature (Figure 27)The circulated flow rate (Figure 27)The steam pressure variation (Figure 28)The water level variation (Figure 28)
Loss-of-Cooling Accidents:
Published in Geoffrey F. Hewitt, John G. Collier, Introduction to Nuclear Power, 2018
Geoffrey F. Hewitt, John G. Collier
Phase 1. Turbine Trip (0–6 mm) This phase is illustrated in Figure 5.1. The valves that allow steam to be dumped to the condenser opened as designed and the auxiliary feedwater pumps started. The interruption of the flow of feedwater to the steam generators caused a reduction in heat removal from the primary system. The reactor coolant system responded to the turbine trip in the expected manner. The reactor coolant pumps continued to operate and to maintain coolant flow through the core. The reactor coolant system pressure started to rise because the heat generated by the core—which was still operating—was not being removed from the system at the required rate by the steam generators. This rise in system pressure caused the power-operated relief valve (PORV) on top of the pressurizer (1 in Figure 5.1) to operate to relieve the pressure. However, the opening of this valve was insufficient to reduce the pressure immediately, and the pressure continued to increase. The operation of the valve occurred between 3 and 6 s after the turbine trip, and the pressurization continued until 8 s after the start of the incident, when the primary circuit pressure reached 162 bars. At this point the control rods were automatically driven into the core as a result of a protection system signal’s detecting the overpressurization. This immediately stopped the fission reaction. At this early stage all the automatic protection features had operated as designed, and the reactor had been shut down. However, as we explained in previous chapters (and as indicated in Table 2.2), the decay heat remains significant. Under normal circumstances this can be dealt with straightforwardly by the various coolant systems.
Core Modeling and Simulation of Peach Bottom 2 Turbine Trip Test 2 Using CASMO5/TRACE/PARCS
Published in Nuclear Technology, 2022
In 1977, turbine trip (TT) tests were successfully performed at the Peach Bottom Unit 2 (PB2) nuclear power plant (NPP) at the end of cycle 2 (EOC 2). The Electric Power Research Institute (EPRI) subsequently documented all tests.1,2 The Organisation for Economic Co-operation and Development (OECD)/U.S. Nuclear Regulatory Commission (NRC) boiling water reactor (BWR) TT benchmark3–6 for Turbine Trip Test 2 (TT2) (PB2 TT2) was designed to provide a foundation for validating developed best-estimate codes. It includes data for both thermal-hydraulic and neutronic parameters at the initial steady-state conditions and throughout the PB2 TT2 transient. The transient boundary conditions, decay heat values, and cross-section libraries, as a function of time, were provided by OECD/NRC. However, despite using the common nodewise cross-sectional data set, the participant-submitted benchmark results revealed deviations because of different thermal-hydraulic models in the code-to-code comparisons of the maximum total power following the TT.
A Multiscale and Multiphysics PWR Safety Analysis at a Subchannel Scale
Published in Nuclear Science and Engineering, 2020
H. Y. Yoon, I. K. Park, J. R. Lee, S. J. Lee, Y. J. Cho, S. J. Do, H. K. Cho, J. J. Jeong
A steam line break accident is one of the postulated events of a PWR in which a three-dimensional analysis is needed for the prediction of a local power and temperature increase in the core that is caused by a moderator feedback of cold coolant from the broken loop. The coupled code is applied to the accident analysis where the core power, reactor vessel thermal hydraulics, and system transient are calculated by MASTER, CUPID-RV, and MARS simultaneously. The reactor core kinetics parameters used in the MASTER calculation were determined at the end of the fuel burnup cycle. This is usually assumed in the steam line break analysis because it may cause a return to core power after the reactor scram due to the large negative moderator density feedback at the end of cycle. The number of meshes for the reactor vessel is 5.4 million for a subchannel-scale analysis. Table III shows the sequence of events of the analysis. The transient begins with a main steam line break and the reactor power continues to increase due to moderator feedback until it reaches an overpower reactor trip set point of 121% at 13.5 s. Then reactor scram and turbine trip occur with some time delays. The minimum value of the departure from nucleate boiling ratio (DNBR) is 2.89, which appears at the top of the hottest channel before the scram rod is dropped. The main steam line isolation valves (MSIVs) are closed at 34.3 s after the low steam generator pressure set point of 5.44 MPa is reached at 33.1 s.
Comparison of Two Different Sized Small-Break LOCAs on the Passive Safety Injection Line Using SMART-ITL Data
Published in Nuclear Technology, 2020
Jin-Hwa Yang, Hwang Bae, Sung-Uk Ryu, Byong Guk Jeon, Sung-Jae Yi, Hyun-Sik Park
The reactor trip signal was generated with a 1.1-s delay after the PZR pressure reached the low pressurizer pressure (LPP) set point (A: 744 s versus B: 7723 s). The turbine trip was assumed to start simultaneously with the reactor trip. Feedwater was stopped and the RCP began to coast down. A CMT actuation signal (CMTAS) was also generated with the reactor trip signal by LPP (A: 745 s versus B: 7724 s). The control rod insertion was simulated by reduced core power, as shown in Figs. 3c and 3d with an additional 0.5-s delay (A: 746 s versus B: 7724 s). The three trains of CMT (except for train 4) injection started following a CMTAS with a time delay of 1.1 s by opening the isolation valves installed on the safety ILs of the CMTs (A: 747 s versus B: 7725 s).