Intentional Depressurization Research Articles

Effectiveness re-evaluation on the intentional primary system depressurization during Zion-like Westinghouse PWR station blackout considering pressure dependence of radionuclides release

Previous studies have successfully resolved the issue of high-pressure melt ejection (HPME) followed by direct containment heating (DCH) during a total station blackout of Zion-like Westinghouse pressurized water reactor (PWR). This resolution is crucial, as the earliest occurrence of hot-leg creep failure can cause a decrease in the pressure difference between the reactor pressure vessel (RPV) and the primary containment vessel (PCV), falling below the cut-off pressure for HPME/DCH. As a recommended accident management strategy, intentional depressurization of the reactor coolant system (RCS) has been proposed. Depressurization leads to a delayed accident progression due to accumulator injection and maintains a pressure difference at the time of RPV failure below the cut-off pressure. In the present analyses using MAAP5, it was observed that depressurization before core heat-up, achieved by opening 2 power-operated relief valves (PORVs), resulted in the most delayed RPV failure, consistent with previous studies. However, present sensitivity analysis considering the pressure-dependent release of radionuclides from the fuel revealed significant changes in both the accident progression and the point of creep failure. In addition, the creep-failure point shifted from the RPV bottom to the RPV sidewall, and the pressure difference at that location exceeded the cut-off pressure. To prevent sidewall failure of the RPV, it may be advisable to employ a depressurization rate higher than that achieved by using 2 PORVs, even if it slightly accelerates the accident progression.

Strategy Evaluation for Cavity Flooding during an ESBO Initiated Severe Accident

Intentional depressurization and cavity flooding are two important measures in current severe accident management guidelines (SAMGs). An extreme scenario of an extended station blackout (ESBO), when electric power is unavailable for more than 24 hours, actually occurred in the Fukushima Daiichi accident and attracted lots of attention. In an ESBO, the containment spray cannot be activated for condensation, and, thus, cavity flooding will generate a large amount of steam, which, ironically, overpressurizes the containment to failure before the reactor vessel is melted through. Therefore, consideration of these conflicting issues and the ways in which plants operate is crucial for strengthening the strategies outlined in SAMGs. In this paper, the effects of intentional depressurization and cavity flooding in an ESBO for a representative 900 MW second-generation pressurized water reactor (PWR) are simulated with MAAP4 code. Diverse scenarios with different starting times of depressurization and water injection are also compared to summarize the positive and negative impacts for accident mitigation. The phenomena associated with creep ruptures, hydrogen combustion, corium stratification, and cavity boiling are also analyzed in detail to strengthen our understanding of severe accident mechanisms. The results point out the facility limitations of second-generation PWRs which can improve existing SAMGs.

Analysis of Fourth Stage of Automatic Depressurization System Failure to Open in AP1000 LOCA

Automatic Depressurization System (ADS) is a very important part of passive core cooling system in passive safety nuclear plant AP1000. ADS have four stages with each stage having two series and only ADS4 utilizes squib valves. During the accident, emergency core injecting is realized by gravity driven passive safety injection system like makeup tank (CMT), accumulator and In-Containment Refueling Water Storage Tank (IRWST). The objective e of this study is to analyze the system response and phenomenon under part of failure of ADS in AP1000 LOCA. The plant model is built by using SCDAP/RELAP5/MOD 3.4 code. The chosen accident scenario is small and medium LOCAs followed by failure of ADS4 to open, whose location is different from the other three stages. The results indicate that long time core cooling from IRWST is postponed greatly through intentional depressurization only by ADS1, 2, 3. In addition, LOCAs with equivalent diameter 25.4 cm and 34.1 cm will not lead to core melt while 5.08 cm break LOCA will. Meanwhile, high water level in the pressurizer will appear during all of three LOCAs.

Compression Behavior of Entrapped Gas in High Pressure Diecasting

Die castings generally contain a large quantity of porosities due to the entrapment of air or gas in molten metal during mold filling. Although the entrapped air or gas is compressed by high casting pressure during pressurization, it will eventually remain in the castings as defects after solidification. Therefore, it is important to clarify the relation between the volume of gas defects and the pressure applied to the molten metal so as to optimize the casting design.In this study, we investigated the compression behavior of entrapped gas during casting. We determined the volume of gas defects and gas content in die castings by density measurement and vacuum fusion extraction method respectively. Then we calculated the gas pressure in the defects from the above volume of defects and gas content, and compared with the die casting pressure. The calculated gas pressure in the defects was found to be not equal to the die casting pressure, but equal to the pressure of the molten metal just before it dropped abruptly due to the complete blocking of the liquid metal channel by solidification. From the experimental results, the behavior of the entrapped gas can be inferred as follows. Immediately after the mold was filled with molten metal, the entrapped gas was instantly compressed. After that, the pressure of molten metal decreased gradually with the progress of solidification of the molten metal channel, and the volume of entrapped gas increased correspondingly until the pressure of the molten metal dropped abruptly. Then the volume of the entrapped gas showed a slight expansion equal to the solidification shrinkage of the enclosed molten metal.The above inference was verified by measuring the volume of the entrapped gas defects in castings made with intentional depressurization carried out at the time when mold filling just finished or halfway through the solidification of the molten metal channel.

Detailed Evaluation of RCS Boundary Rupture during High-Pressure Severe Accident Sequences

A depressurization possibility of the reactor coolant system (RCS) before a reactor vessel rupture during a high-pressure severe accident sequence has been evaluated for the consideration of direct containment heating (DCH) and containment bypass. A total loss of feed water (TLOFW) and a station blackout (SBO) of the advanced power reactor 1400 (APR1400) has been evaluated from an initiating event to a creep rupture of the RCS boundary by using the SCDAP/RELAP5 computer code. In addition, intentional depressurization of the RCS using power-operated safety relief valves (POSRVs) has been evaluated. The SCDAPRELAP5 results have shown that the pressurizer surge line broke before the reactor vessel rupture failure, but a containment bypass did not occur because steam generator U tubes did not break. The intentional depressurization of the RCS using POSRV was effective for the DCH prevention at a reactor vessel rupture.

Evaluation of intentional depressurization strategy in Chinese 600 MWe PWR NPP

Intentional depressurization is one of the effective strategies in preventing high-pressure melt ejection (HPME) and direct containment heating (DCH), which is most feasible for the operating nuclear power plants (NPPs) in China. In order to evaluate this strategy of a Chinese 600 MWe PWR NPP, the plant model is built using SCDAP/RELAP5 code. ATWS, SBO, SGTR and SLOCA are selected as the base cases for analysis of intentional depressurization. The results show that opening safety valves of pressurizer manually when the core exit temperature exceeds 922 K can reduce the RCS pressure effectively and prevent the occurrence of HPME and DCH. Several uncertainties such as the operability of safety valves, ex-vessel failure and the transitory rise of RCS pressure are also analyzed subsequently. The results show that the opening of the safety valves can be initiated normally and that opening three safety valves is a more favorable strategy in the event of possible failure of one or more of the safety valves; the probability of ex-vessel failure is reduced after intentional depressurization is implemented; the transitory rising of reactor coolant system (RCS) pressure when the molten core materials relocate to the lower head of reactor pressure vessel (RPV) will not influence the effect of depressurization.

RELAP5/MOD3 Code Analyses of LSTF Experiments on Intentional Primary-Side Depressurization Following SBLOCAs with Totally Failed HPI

The effectiveness of intentional depressurization of a pressurized water reactor primary system as a means to maintain core cooling during a small-break loss-of-coolant accident (SBLOCA) was studied. The investigation was based on experiments conducted at the Rig of Safety Assessment-V (ROSA-V) Large Scale Test Facility (LSTF) and RELAP5/MOD3 code calculations performed for LSTF geometry, together with single lumped-volume model calculations - all simulating hypothetical total failure of the high-pressure-injection system. For cold-leg breaks ≶2.5% of the leg cross-sectional area, experimental and analytical results have shown that the break discharge depressurizes the primary system to the accumulator (ACC) and low-pressure-injection (LPI) system injection pressures, and thus the core cladding temperature would be maintained below ~1000 K. For break areas ≤1.0%, on the other hand, additional depressurization means are needed to initiate the ACC injection before the core is overheated. RELAP5/MOD3 calculations have shown that steam venting through the pressurizer power-operated relief valves would be effective in depressurizing the primary system to the ACC and LPI pressures. However, for break areas <0.5%, the peak cladding temperature would finally reach the safety criterion of 1473 K.

Core Liquid Level Responses Due to Secondary-Side Depressurization during PWR Small Break LOCA

The effects of the rate of intentional secondary-side depressurization and of the break area on the core liquid level behavior during a PWR small-break LOCA with total failure of the high pressure injection (HPI) systems were studied using experimental data from the Large Scale Test Facility (LSTF) and using analysis results obtained with the REFLA/TRAC code. The LSTF is a 1/48 volumetrically scaled full-height integral model of a Westinghouse-type PWR. The code reproduced the thermal-hydraulic responses observed in the experiment for important parameters such as the primary and secondary-side pressures and core liquid level. The sensitivity of the core minimum liquid level to the depressurization rate and break area was studied by using the code assessed above. It was found that the core liquid level took a local minimum value for a given break area as a function of secondary-side depressurization rate. The core temperature was kept below about 1,000 K when the secondary depressurization rate was greater than—200 K/h in the saturation temperature for break areas smaller than 2.5%.

Intentional Depressurization of Steam Generator Secondary Side during a PWR Small-Break Loss-of-Coolant Accident

The consequence of intentional depressurization of the steam generator (SG) secondary side during a small break loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR) was studied using the ROSA-IV Large Scale Test Facility (LSTF). The LSTF is a 1:48 volumetrically scaled full-height model of a Westinghouse-type PWR. The experiment simulated a 0.5% cold leg small-break LOCA with total failure of high pressure injection (HPI) system. The SG secondary-side atmospheric relief valve (ARV) was latched open after the core top region was uncovered. Then, the primary pressure closely followed the secondary pressure and dropped to the accumulator (ACC) injection pressure of 4.51MPa before the core became severely overheated. A post test analysis, which was performed using the RELAP51 MOD2 code with modifications made by the authors, predicted well the primary and secondary side responses during this experiment.

SCDAP/RELAP5 Analysis of Station Blackout with Pump Seal LOCA in Surry Plant.

During a station blackout of PWR, the pump seal will fail due to loss of the seal cooling. This particular transient-LOCA sequence designated as S3-TMLB' analyzed by SNL with MELPROG/TRAC for Surry plant showed that the depressurization due to the pump seal LOCA would result in early accumulator injection and subsequent core cooling which lead to the delay of reactor pressure vessel (RPV) meltthrough. The present analysis was performed with SCDAP/RELAP5 to evaluate this scenario shown in the MELPROG/TRAC analyses. Addition-ally, the calculated results were compared with the similar experimental studies of JAERI's ROSA-IV program. The present analyses showed that: (1) During S3-TMLB', the loop seal clearing would occur and cause a slight delay of accident progression. (2) It is unlikely that the accumulator injection, which leads to the delay of RPV meltthrough by approximately 60 min, is initiated automatically during S3-TMLB'. Accordingly, an intentional depressurization using PORVs is recommended for th...

Intentional Depressurization of Steam Generator Secondary Side during a PWR Small-Break Loss-of-Coolant Accident.

Intentional Depressurization of Steam Generator Secondary Side during a PWR Small-Break Loss-of-Coolant Accident.

PWR Cold-Leg Small-Break LOCA with Total HPI Failure

Cold-leg small-break loss-of-coolant accident (LOCA) tests were performed at the ROSA-IV Large Scale Test Facility (LSTF), a 1/48 volumetrically-scaled model of a pressurized water reactor (PWR). The tests were conducted for break areas ranging 0.5–10% of the scaled cold leg area, and simulated hypothetical total failure of the high pressure injection (HPI) system. One of the tests, conducted with 1% break area, included an intentional depressurization of the primary system that was initiated after the onset of core dryout. A simple prediction model is proposed for prediction of times of major events, namely, loop seal clearing, core dryout, accumulator (ACC) injection and actuation of low pressure injection (LPI) system. Test data and model calculations show that intentional primary system depressurization with use of the pressurizer power-operated relief valves (PORVs) is effective for break areas of approximately 0.5% or less, is unnecessary for breaks of approximately 5% or more, and might be insuffic...

PWR Cold-Leg Small-Break LOCA with Total HPI Failure. Effect of Break Area on Core Dryout and Intentional Depressurization for Prevention of Excess Core Dryout.

Cold-leg small-break loss-of-coolant accident (LOCA) tests were performed at the ROSA-IV Large Scale Test Facility (LSTF), a 1/48 volumetrically-scaled model of a pressurized water reactor (PWR). The tests were conducted for break areas ranging 0.5–10% of the scaled cold leg area, and simulated hypothetical total failure of the high pressure injection (HPI) system. One of the tests, conducted with 1% break area, included an intentional depressurization of the primary system that was initiated after the onset of core dryout. A simple prediction model is proposed for prediction of times of major events, namely, loop seal clearing, core dryout, accumulator (ACC) injection and actuation of low pressure injection (LPI) system. Test data and model calculations show that intentional primary system depressurization with use of the pressurizer power-operated relief valves (PORVs) is effective for break areas of approximately 0.5% or less, is unnecessary for breaks of approximately 5% or more, and might be insuffic...

Analysis of steam generator U-tube rupture and intentional depressurization in LOBI-MOD2 facility

LOBI-MOD2 results from an experimental investigation of a steam generator tube rupture (SGTR) with no active emergency core cooling system (ECCS) available are presented. The characteristic system response, following a three-tube break occuring during normal full power operation, is analysed. Thermal—hydraulic data for a phenomenological understanding of natural circulation, the reflux condenser heat transport mode and core dry-out at high primary pressure are provided. The importance of the availability of the ECCS for the SGTR mitigation is outlined by comparing the above mentioned data to the results of a second LOBI test on SGTR with the same break area but with the standard ECCS in operation. The effectiveness of a new recovery procedure, starting from highly degraded core scenarios, is also examined. It relies on intentional primary system fast depressurization through the pressurizer relief valves and passive accumulator injection. Primary passive feed and bleed for the long term cooling can improve the safety criteria for the design of nuclear power plants (NPPs).