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295 records · Page 17

Development of F and B Operation Strategy with LPSI and H-SIT

In the current nuclear power plants (NPPs), most of the critical safety functions are provided by many active safety systems. However, in order to meet the futuristic goals of safety, relying on these active safety systems alone does not seem to be viable. As passive system has a strength on the mitigation of station black out accident and it also contributes to increase the diversity of mitigation way, application of passive system becomes a paramount in the nuclear industry to enhance the nuclear plant safety. Recently, many researches for developing passive systems and operation strategies have been carried out in the nuclear field. General passive systems and strategies, however, have a critical limitation, especially for passive injection systems. As these systems use gravity as their driving force, it is difficult to inject coolant in a passive way in a high-pressure condition thus depressurization has to be preferentially performed to do passive injection. The problem is that depressurization inevitably causes the loss of core inventory thus it is not preferred in order to secure plant safety. Hybrid safety injection tank (H-SIT) system was newly suggested a few years ago to passively inject the coolant into the reactor coolant system (RCS) without depressurization in a high-pressure condition. The H-SIT system is planned to be integrated into advanced power reactor plus (APR+). Generally, the H-SIT system can inject water using the pressure from nitrogen gas as a conventional safety-injection tank in low-pressure accidents, such as large and medium-break loss of coolant accidents (LOCA). Additionally, the H-SIT system can also inject water using gravitational force in high-pressure accidents. If a high pressure accident occurs, the pressure of each H-SIT is equalized with pressure of pressurizer through equalizing pipes, thus allowing the H-SITs to inject water by gravitational force. In the target plant, four H-SITs are planned to be installed. Each has a 2.5 in diameter equalizing pipe and it can be operated manually. Long-term cooling of core is an ultimate goal of all mitigation actions for plant safety and feed and bleed (F and B) operation strategy is one of the long-term cooling strategies in conventional pressurized water reactor (PWR). The important point of F and B operation is that injection for feed operation is performed by only high pressure safety injection (HPSI) pump because operation of low pressure safety injection (LPSI) pump needs the depressurization and it causes core dry out. As H-SIT is passively operated in an accident situation, it was originally designed for mitigation of station black out accident. Besides, operator can use the H-SIT system same as high pressure injection pump in any accident situations. Thus, function of high pressure injection by H-SIT can be used for feed operation when HPSI pumps fail. If it is possible, H-SIT can make up the core during depressurization then LPSI pumps can be used for F and B operation. In this study, F and B operation strategy with LPSI and HSIT is developed. This is a new concept for the long-term cooling operation. If this strategy is applied, low pressure injection pump can be successfully used for F and B operation thus operator has the additional mitigation way. As this strategy make plant safe even though HPSI and PAFS are both failed, it can effectively enhance the plant safety. For this strategy two RCGVSs and two POSRVs are needed as a depressurization system for bleed operation and only one LPSI is enough for feed operation. H-SIT operation is also needed to make up core inventory during bleed operation. For this operation, four H-SITs have to be used to make up core safely. Based on the risk analysis using PSA method, if this strategy is applied, core damage frequency is 1.868 e-6 which declined 7 percent from original model. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Systems Engineering (SE) Approach of Developing A Passive Safety Injection System (PSIS) to Cope with APR1400 Extended Station Black-Out (SBO) Accident Using RELAP5

On 11 March 2011 the power supply and cooling of the three Fukushima Daiichi reactors were disabled by a major earthquake followed by a relatively high tsunami waves. The earthquake damaged the electrical grid which is the normal power supply for the reactors, then the tsunami waves disabled the emergency diesel generators of the Fukushima Daiichi reactors, which left the plant with no alternating current (AC) source. This condition is called 'Station Black-Out'. The reactors were tripped immediately, but due to the nuclear physics phenomena, the reactor cores were still producing thermal power, which is generated by the decay of radioactive fission products in the reactor core, known as decay heat. The decay heat had not been circulated to an external heat sink and eventually the accumulative heat produced in the core led to damage to the reactor cores. After the Fukushima Daiichi accident, Design Extension Condition (DEC) such as total loss of feed water (TLOFW), and prolonged Station Black-Out (SBO) attracted the attention of nuclear experts. The attention focuses on the defense in depth concept, redundancy and sustainability of the systems, and development of the passive safety systems, which rely on natural physics laws. The topic of this paper is to discuss the possibility of attaching a Passive Safety Injection System (PSIS) to the APR1400. The PSIS is a safety system which provides adequate cooling water to the steam generators after the loss of the Auxiliary Feed Water System (AFWS) during an extended SBO accident. The PSIS will be available while the power supply of the plant is being recovered and the active safety systems take over the functionality. The anticipated PSIS will be automatically actuated to enhance the engineered safety feature and to reduce the human errors. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

SIT Makeup Effect on the Loss of Residual Heat Removal during Mid-loop Operation

For the loss of cooling during the mid-loop operation, the plant needs to have a sufficient time for operator action without active makeup measures such as a loss of alternating current (AC) power. This concern is raised after the Fukushima accident. Especially, in addition to the loss of residual heat removal (LORHR), if the AC power or active makeup measures are unavailable, the passive makeup using gravity feed or pressurized feed will provide sufficient time before core uncovery. This paper investigates means of the passive makeup measures using safety injection tank (SIT) after the LORHR event to prevent or delay the core uncovery and fuel failure for the typical advanced power reactor 1400 (APR1400). Unfortunately, the refueling water storage tank of APR1400 cannot be used for gravity makeup since it is located on the bottom of the containment with a lower elevation than the reactor vessel. So, the SIT makeup and the effect on the core uncovery are investigated. The results show that gravity makeup of a SIT provides insufficient means but the slightly pressurized SIT is required for sufficient operator action time for the LORHR with the station black out (SBO) during the mid-loop operation for the APR1400. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Analytical Models for Terry Turbine Nozzle

As part of the efforts to understand the unexpected 'self-regulating' mode of the RCIC (Reactor Core Isolation Cooling) systems in Fukushima accidents and extend BWR RCIC and PWR AFW (Auxiliary Feed Water) operational range and flexibility, mechanistic models for the Terry turbine should be developed and implemented in reactor system analysis or severe accident codes. The Terry turbine has been used for all the BWR RCIC systems and for PWR AFW (Auxiliary Feed Water) systems. The Terry turbine is essentially a solid cylindrical wheel with multiple machined semi-circular 'buckets' that are shaped into the body of the wheel. Fixed nozzles and reversing chambers surround the wheel inside the turbine casing. High pressure steam is accelerated to supersonic flow inside the turbine nozzle. The kinetic energy is then converted to shaft work by the impulse force on the turbine buckets. A Fully implicit and strongly coupled RCIC system model has been developed in the RELAP-7 code and used for simplified BWR SBO (Station Black-Out) simulations in the past. In this model, a generic turbine model was used to conserve mass and energy while turbine operation characteristic curves were used to obtain non-dimensional mass flow rate and thermal efficiency. This model could be used for simulating RCIC off-design behavior if off-design operation characteristic curves were available. However, No such curves currently exist for the Terry turbine system due to its unique pure impulse design. A set of mechanistic Terry turbine models suitable for system code implementation has recently been developed by Sandia National Labs. In the Sandia model, the turbine bucket inlet velocity is provided according to a reduced-order model which was obtained from a large number of CFD simulations. In this paper, we propose an alternative method, using an under-expanded jet model to obtain the velocity and thermodynamic conditions for the turbine bucket inlet. These sets of analytical models are simple and generic, and suitable for use in system analysis codes. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Development of Accident Tolerant Fuel Options for Near Term Applications: Fuel Performance Modeling under Transient / Severe Accidents by MELCOR: PART I: Benchmark for SURRY: Short Term Station Black Out (STSBO)

Currently, existing LWR fuel technology is near its inherent limit of performance without improvement in fuel materials. As a result of the great earthquake of east Japan, along with the catastrophic tsunami, the Fukushima Daiichi boiling water plant was severely damaged. The light water reactors (LWRs) fuel gave rise to a serious discussion in the United States. In order to provide the development of Accident Tolerant Fuel (ATF) options for present day applications, MIT is leading a DOE project to search for enhanced accident tolerant fuels. Group members include MIT, UW-Madison, TAMU, AREVA, and Westinghouse. As part of the work, fuel performance modeling under transient / severe accidents: MELCOR is being used to analyze different kinds of materials under several limited accident conditions. While various research groups are obtaining the parameters of the experimental ATF materials, benchmarks for SURRY are being done by use of MELCOR in the University of Wisconsin-Madison. We chose version 1.8.6 for the edition of materials, and INL provided the uncertainty analysis between version 1.8.6 and version 2.1.0. Comparison will be done between potential ATF materials, which include FeCrAl, SiC, Cr, ZrSi, TiAlC, TiSiC and so on. Furthermore, zircaloy is used as the verification. Among all the parameters which influence the materials' performance, the temperature history and hydrogen generation will be the key factors in the future. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Korean Localization Program for Performance Verification Test of APR1400's RCP

The Korean government together with the Korean nuclear society completed the development of technologies for the design, manufacturing, and performance verification test of the APR1400's Reactor Coolant Pump (RCP) as one of the national programs for the localization of deficient nuclear technology. The RCP localization program was driven by KAERI, which was responsible for the development of the performance verification and evaluation technology, and by DOOSAN heavy industrial company, which was in charge of the design and manufacturing of the RCP. The test facility for the localized APR1400's RCP was constructed at the KAERI site at the end of 2012. The thermal hydraulic and electric capability of the RCP test facility (RCPTF) covers up to 18.5 MPa, 343 deg. C, 11.7 m{sup 3}/s, and 14.0 MW for the design pressure, temperature, flow rate, and maximum electric power, respectively. In 2013, a commissioning test was performed to verify its designed capability, followed by several modifications in the RCPTF including signal processing and control logic to enhance the verification and evaluation capability of the RCP performance. After finishing the commissioning and modification of the RCPTF, a type test for the new-type of RCP was performed successfully. In 2014 and 2015, eight production tests for the localized RCPs, which will be installed in a commercial nuclear power plant in Korea, were successfully completed. As a final stage of the localization program of the RCP, KAERI will perform a seal LOCA test to verify the leakage characteristics and produce experimental data on the localized seal assembly for a safety analysis during a station black out (SBO) condition. In this summary, several technical issues developed through a series of research processes for a performance verification test of the localized RCP are described.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗