High Pressure Core Spray Research Articles

Failure possibilities for nuclear safety assessment by fault tree analysis

Fault tree analysis (FTA) is a deductive tool to assess the safety of nuclear power plants. This analysis can only be implemented if all basic events in the tree have their corresponding failure rates. Therefore, safety analysts have to provide those failure rates well in advance. However, it is often difficult to obtain those failure rates due to insufficient data, changing environment or new components. This paper proposes a failure possibility based FTA approach to overcome the limitation of the conventional FTA for nuclear safety assessment. It utilises the concept of failure possibilities to evaluate basic event failure without historical data, fuzzy numbers to map component failure possibilities into mathematical form and defuzzification algorithms to convert fuzzy numbers into component failure rates. A case study on evaluating a typical high pressure core spray system of a boiling water reactor illustrates the applicability of the proposed approach.

Modeling of the High Pressure Core Spray Systems with fuzzy cognitive maps for operational transient analysis in nuclear power reactors

In this paper the application of fuzzy cognitive maps (FCM) to model a risk scenario for Nuclear Power Plants (NPP) in a Boiling Water Reactor (BWR) is presented, specifically for failure modes and effects analysis of High Pressure Core Spray System (HPCS) during loss of reactor coolant inventory transients. A simplified model of the HPCS is analyzed with the fault tree analysis technique in order to compare this results with those obtained with the FCM and show consistency with the results, although this process is not a validation of the FCM techniques. The decision making in an NPP is a complex process, because of the numerous elements involved in its operation, and the permanent attention demanded by its maintenance. This is the first step in the development of an expert system that will help in the decision making process, through the design of the knowledge representation and the design of reasoning with FCM to automate the decision making process.

Validation of RCS Bleed-and-Feed Strategy for the SAMG of the Kuosheng Nuclear Power Plant

The severe accident management guideline (SAMG) was developed and implemented at the Kuosheng nuclear power plant (NPP) in Taiwan at the end of 2003. The Kuosheng NPP is a Mark-III boiling water reactor (BWR)-6 located in the north of Taiwan. The SAMG of the Kuosheng NPP is developed based on the BWR Owners Group Emergency Procedure and Severe Accident Guidelines. In this paper, MELCOR 1.8.5 is used to investigate the effectiveness of the Kuosheng SAMG and to analyze the reactor coolant system (RCS) bleed-and-feed strategy in a postulated station blackout (SBO) accident since this accident is a dominant sequence that can induce core damage. These simulation results can be applied for the typical BWR-6 NPPs. Based on the simulation results, the major events of an SBO accident without any operator actions, including core uncovery, cladding oxidation, hydrogen generation, fuel relocation, vessel failure, and containment failure, are well presented. The RCS bleed-and-feed strategy can cool down the hot core and bring the vessel into the stable condition in the SBO accident with high-pressure core spray injection according to the Kuosheng SAMG. The reactor pressure vessel depressurization before the reactor liquid level reaches one-fourth of the core liquid level can prevent damage to the core fuel. However, the water temperature of the suppression pool will reach saturation temperature, and the containment pressure may challenge the containment integrity. These phenomena were not considered while developing the SAMG.

On the need for dynamic methodologies in risk and reliability studies

Abstract Whenever there is complex interaction between the system components and process variables, the impact of uncertainties in the process physics on the predicted Top Event properties may be very difficult to account for without failure modeling techniques based on quantitative, time-dependent plant models (dynamic methodologies). New quantitative justification is given as to why dynamic methodologies may be needed in risk and reliability studies by demonstrating that these impacts can be large. The process under consideration is the feed-bleed cooling of BRWs following a small-break loss-of-coolant-accident (SBLOCA) through the intermittent operation of the high pressure core spray system (HPCS) and one safety relief valve (SRV). The dynamic methodology models the process by regarding pressure and water level evolution in the reactor vessel (RV) following the SBLOCA as transitions between the computational cells in the process variable space. It is shown that the predicted probability of demand for the low pressure core injection system during one hour following the SBLOCA due to the malfunction of the SRV/HPCS system can differ by more than one order of magnitude depending on: (a) the SBLOCA break size, and (b) whether the stochasticity in the HPCS and SRV demand rates arising from their interaction through the process variables is accounted for or not.

Investigation of effect of pressure control system on BWR LOCA phenomena using ROSA-III test facility.

A pressure control system failure test series was conducted at the Rig of Safety Assessment (ROSA)-III test facility to evaluate the effect of the pressure control system on thermal-hydraulic phenomena during a small break loss-of-coolant accident (LOCA) of a boiling water reactor (BWR). The break was assumed at the recirculation pump suction line. The pressure control system had no effect for breaks greater than 5%. For breaks less than 5%, if the pressure control system was inactive, the core was uncovered temporarily because of the bubble collapse due to pressure rise by the closure of the main steam isolation valve (MSIV), and the fuel rod surface temperature rose high during this period. However, the peak cladding temperature (PCT), which occurred mainly during the later core uncovering by boil-off, was lower in a LOCA with pressure control system failure than in a corresponding LOCA with intact pressure control system. This is because the emergency core cooling systems (ECCSs) were actuated earlier in a LOCA with pressure control system failure due to lower system pressure. The PCT was well below the present safety criteria of 1,473K even if the pressure control system and high pressure core spray system (HPCS) (the severest single failure in ECCS) were assumed to be inactive.

Investigation of BWR LOCA at ROSA-III — Effect of break configuration on system transients

The ROSA-III test facility is a volumetrically scaled ( 1 424 ) BWR/6 system with an electrically heated core to study the thermal-hydraulic response during a postulated loss-of-coolant accident (LOCA). Six loss-of-coolant experiments with a break area of 15%, 50% or 200% at the main recirculation pump inlet line were conducted at the ROSA-III test facility with a high pressure core spray failure. A sharp-edged orifice or a long throat nozzle was used as a break plane. It was found in the experiments that the break flow differences between the orifice and the nozzle break configurations with the same flow area were observed only in the subcooled break flow region. Subcooled break flow rate through the orifice was much larger than that through the nozzle. The break configuration difference had little influence on the other system responses, especially on the peak cladding temperature. The applicability of the test results to a BWR/6 has been confirmed through analyses of the 15% break ROSA-III LOCA experiments and BWR/6 LOCAs by using RELAP4/MOD6/U4/J3 code. The experimental results of the ROSA-III LOCA experiments were calculated well by the code, and the same trends were calculated in the BWR analyses.

Investigation of break location effects on thermal-hydraulics during intermediate break loss-of-coolant accident experiments at ROSA-III.

The rig of safety assessment (ROSA)-III facility is a volumetrically scaled (1/424) boiling water reactor (BWR/6) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) tests. Break location effects on thermal-hydraulics during intermediate LOCAs were investigated by using four experiments at the ROSA-III, the 15 and 25% main recirculation pump suction line break (MRPS-B) experiments, the 21% single-ended jet pump drive line break (JPD-B) experiment and the 15% main steam line break (MSL-B) experiment. Water injection from the high pressure core spray (HPCS) was not used in any of the experiments. Failure of ECCS actuation by the high containment pressure was also assumed in the tests. In the MRPS-B experiments, the discharge flow turned from low quality fluid to high quality fluid when the downcomer water level dropped to the main recirculation line outlet elevation, which suppressed coolant loss from the vessel and the core. In the JPD-B experiment, the jet pump drive nozzle was covered with low quality fluid and low quality fluid discharge continued even after the downcomer water level reached the jet pump suction elevation. Low quality fluid discharge ceased after the ADS actuation. It suggestes that the JPD-B LOCA has the possibility of causing larger and more severe core dryout and cladding temperature excursion than the MRPS-B LOCA. The MSL-B LOCA was characterized by mixture level swell in the downcomer and the core. The core mixture level swell resulted in the much later core dryout initiation than that in the MRPS-B LOCA, however, ECCS actuation was also delayed because of slow downcomer water level drop.

Analyses of ROSA-III Break Area Spectrum Experiments on a Boiling Water Reactor Loss-of-Coolant Accident

The ROSA-III program has conducted system effects tests on the thermal-hydraulic response of a boiling water reactor during a loss-of-coolant accident. The performance of the emergency core cooling systems was of particular interest. As part of this program, ten tests were conducted with (a) a simulated pipe rupture located at the recirculation pump suction line, (b) a spectrum of break area ranging from 0 to 200% of scaled pipe cross-sectional area, and (c) an unavailable high-pressure core spray (HPCS) system. In these tests the pressure vessel depressurized (a) due to the actuation of the automatic depressurization system for scaled break areas of 50%, and (c) due to both for the intermediate break areas between 5 and 50%. Vessel depressurization enabled the injection of emergency core coolant from the low-pressure core spray and low-pressure coolant injection system and thus led to safe recovery without HPCS. The behaviors of the vessel pressure, core mixture level, and core temperatures were fairly well reproduced by the THYDE-B1 code, based on simplified lumpedparameter models for the wide spectrum of break areas investigated.

Experimental analysis of the power curve sensitivity test series at ROSA-III

The rig of safety assessment (ROSA)-III facility is a volumetrically scaled (1/424) boiling water reactor (BWR/6) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) tests. Seven recirculation pump suction line break LOCA experiments were conducted at the ROSA-III facility in order to examine the effect of the initial stored heat of a fuel rod on the peak cladding temperature (PCT). The break size was changed from 200% to 5% in the test series and a failure of a high pressure core spray (HPCS) diesel generator was assumed. Three power curves which represented conservative, realistic and zero initial stored heat, respectively, were used. In a large break LOCA such as 200% or 50% breaks, the initial stored heat in a fuel rod has a large effect on the cladding surface temperature because core uncovery occurs before all the initial stored heat is released, whereas in a small break LOCA such as a 5% break little effect is observed because core uncovery occurs after the initial stored heat is released. The maximum PCTs for the conservative initial stored heat case was 925 K, obtained in the 50% break experiment, and that for the realistic initial stored heat case was 835 K, obtained in the 5% break experiment.

Five percent break BWR LOCA/ECC test at ROSA-III without HPCS actuation - two-dimensional core thermal-hydraulic phenomena

The ROSA (Rig of Safety Assessment)-III facility is a volumetrically scaled (1/424) simulated boiling water nuclear reactor (BWR) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) tests. A recirculation pump suction line break test with a five percent break area was conducted with the assumption of high pressure core spray system (HPCS) failure. The simulated peripheral fuel rods facing the channel box wall had a tendency to be rewetted temporarily at the upper part of the core by falling water from the upper plenum before low pressure core spray system (LPCS) actuation, while the rods in the central region were not rewetted but quenched mainly from the bottom of the core after low pressure coolant injection system (LPCI) actuation. Therefore, the peak cladding temperatures of the simulated high power fuel rods were limited to lower values since they were located in the peripheral region and the temporary rewetting before LPCS actuation occurred mainly in the peripheral region. The ROSA-III five percent break test and a BWR counterpart were analyzed with the RELAP5/MOD1 (cycle 018) code. Similarity between the ROSA-III small break test and a BWR small break LOCA has been confirmed through comparison of the calculated results.

BWR LOSS of Coolant Integral Tests with Two Bundle Loop, (II) Effect of ECCS Activation Modes on Thermal-Hydraulic Characteristics

Simulation tests of a BWR loss-of-coolant accident (LOCA) by a postulated full size, double-ended break of a recirculation suction line were conducted using the Two Bundle Loop (TBL) with two full size, electrically heated bundles. In the tests, single failure of emergency core cooling systems (ECCSs) was assumed. The following results were obtained from the tests: (1) The rods heated up early in one of the two bundles, while heatup initiations were delayed in the other. The difference was observed for any ECCS activation mode. (2) The peak cladding temperature (PCT) of each bundle was expressed as a linear function of integrated power during the heatup period. (3) The highest PCT in the tests was 861 K under the assumption of an HPCS (high pressure core spray) failure, which was well below the licensing limitation of 1,473 K.

BWR loss of coolant integral tests with two bundle loop, (II) effect of ECCS activation modes on thermal-hydraulic characteristics.

Simulation tests of a BWR loss-of-coolant accident (LOCA) by a postulated full size, double-ended break of a recirculation suction line were conducted using the Two Bundle Loop (TBL) with two full size, electrically heated bundles. In the tests, single failure of emergency core cooling systems (ECCSs) was assumed. The following results were obtained from the tests: (1) The rods heated up early in one of the two bundles, while heatup initiations were delayed in the other. The difference was observed for any ECCS activation mode. (2) The peak cladding temperature (PCT) of each bundle was expressed as a linear function of integrated power during the heatup period. (3) The highest PCT in the tests was 861 K under the assumption of an HPCS (high pressure core spray) failure, which was well below the licensing limitation of 1,473 K.

BWR recirculation line break simulation tests at two bundle loop.

Integral blowdown refill/reflood tests have been made at the Two Bundle Loop (TBL) to clarify the BWR safety margin for the Emergency Core Cooling System (ECCS) against the recirculation line break. The tests were based on the scaling considerations of the blowdown process as affected by stored heat release from the vessel structures. The discussion led to a scaling compromise of the nozzle diameter increment of the automatic depressurization system to compensate for the excess stored heat release in the TBL. Test results, taking the break size as a parameter, were given of a single-ended recirculation line break under a single failure of the high pressure core spray system. The comparison of the results with the previous study indicated that the uncovery period of the bundle during blowdown had a peak in the small break test of A B /V=1.0×10−5m−1 (A B : Break area, V: Vessel volume). Based on the scaling principle developed, the uncovery period of heater bundles was corrected. From the estimated uncovery...

ROSA-III Double-Ended Break Test Series for a Loss-of-Coolant Accident in a Boiling Water Reactor

The Rig of Safety Assessment (ROSA) III facility is a volumetrically scaled (1/424) boiling water reactor (BWR) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency-core-cooling-system (ECCS) tests. Experimental results obtained so far confirm that the severest single failure assumption in ECCS is the high-pressure core spray system failure even in a large-break LOCA in a BWR. The measured peak cladding temperature was well below the present safety criterion of 1473 K, even with the single failure assumption in ECCS, and the effectiveness of ECCS for core cooling during a double-ended-break LOCA has been confirmed. The overall agreement between the results calculated by the RELAP4/MOD6/U4/J3 computer code and the experimental results is good. The similarity between the ROSA-III test and a BWR LOCA has been confirmed through the comparison of calculated results for the ROSA-III facility and a BWR system.

Simulation experiment of five percent small break loss-of-coolant accident of boiling water reactor.

Five-percent small break tests have been conducted at the ROSA-HI test facility with various emergency core cooling system (ECCS) actuation conditions. Break was assumed at the main recirculation pump inlet line. The ROSA-HI facility simulated a BWR/6-251 by the volume scaling ratio of 1/424 and a 5% break area in a ROSA-IE test means 5% of 1/424 of the BWR recirculation line pipe cross section. No dryout of fuel surface was observed in the tests with high pressure core spray system (HPCS) actuation irrespective of the conditions of the other ECCS. Without HPCS the whole core was uncovered to the steam environment; however, the peak cladding temperature (PCT) was limited to 1,022 K well below the present safety criteria of 1,473 K. The results of the test with the highest PCT without HPCS actuation were analyzed by the RELAP4/Mod6 computer code. The calculated fuel surface temperatures agreed well with measured results in the blowdown phase. The calculated PCT was 930 K, 92 K below the measured PCT of 1,0...