Standby Liquid Control System Research Articles

Fuel integrity evaluation method for BWR-6 plant Kuosheng during ATWS events

Abstract Anticipated Transients without Scram (ATWS) are evaluated to demonstrate that during ATWS the fuel’s integrity prevents fission product release, and that the safety features the BWR-6 plant is equipped with mitigate the ATWS event. For analyzing this scenario, a RETRAN-3D system model and hot channel model of Kuosheng NPP have been developed. From all possible events those are analyzed where the main steam isolation valves close (MSIVC) when the reactor is operating at rated power conditions. For cases with low core flows this results in higher peak cladding temperatures and thicker oxidation thicknesses. All results confirm the ATWS acceptance criteria. For sensitivity cases, increasing the boron concentration of standby liquid control system (SLCS) or injecting boron of SLCS earlier were investigated. The results show, that these measures cannot decrease the PCT value immediately, but can shorten the period of peak cladding temperature over 850 °C which results in smaller oxidation thickness.

Assessment of passive safety system performance under gravity driven cooling system drain line break accident

A generation III+ Boiling Water Reactor (BWR) which relies on natural circulation has evolved from earlier BWR designs by incorporating passive safety features to improve safety and performance. Natural circulation allows the elimination of emergency injection pump and no operator action or alternating current (AC) power supply. The generation III+ BWR's passive safety systems include the Automatic Depressurization System (ADS), the Suppression Pool (SP), the Standby Liquid Control System (SLCS), the Gravity Driven Cooling System (GDCS), the Isolation Condenser System (ICS) and the Passive Containment Cooling System (PCCS). The ADS is actuated to rapidly depressurize the reactor leading to the GDCS injection. The large amount of water in the SP condenses steam from the reactor. The SLCS provides makeup water to the reactor. The GDCS injects water into the reactor by gravity head and provides cooling to the core. The ICS and the PCCS are used to remove the decay heat from the reactor. The objective of this paper is to analyze the response of passive safety systems under the Loss of Coolant Accident (LOCA). A GDCS Drain Line Break (GDLB) test has been conducted in the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) which is scaled to represent the generation III+ BWR. The main results of PUMA GDLB test were that the reactor coolant level was well above the Top of Active Fuel (TAF) and the reactor containment pressure has remained below the design pressure. In particular, the containment maximum pressure (266 kPa) was 36% lower than the safety limit (414 kPa). The minimum collapsed water level (1.496 m) before the GDCS injection was 8% lower than the TAF (1.623 m) but it was ensured that two-phase water level was higher than the TAF with no core uncovery.

CFD predictions of standby liquid control system mixing in lower plenum of a BWR

During an anticipated transient without scram (ATWS) scenario in certain boiling water reactor (BWR) systems, a standby liquid control system (SLCS) is used to inject a sodium pentaborate solution into the reactor system in order to quickly shut down (scram) the reactor without the use of the control rods. Some BWR designs utilize a SLCS that injects through a set of nozzles on a vertical pipe in the peripheral region of the lower plenum of the reactor vessel. During the scenario, system water levels are reduced and natural circulation flow rates down through the jet pump nozzles and up into the core are a small fraction of the rated system flow. It is during this period that the SLCS flows are considered. This work outlines some initial scoping studies completed by the staff at the Nuclear Regulatory Commission (NRC).An attempt at benchmarking the computational fluid dynamics (CFD) approach using a set of available test data from a small facility is outlined. Due to our lack of information related to specific details of the facility geometry along with the limited data available from the test, the benchmark exercise produced only a qualitative basis for selecting turbulence models and mesh density.A CFD model simulating a full-scale reactor system is developed for the lower plenum of a representative BWR/4 design and SLCS flows and mixing are studied under a range of flow conditions. The full-scale BWR simulation builds upon the lessons learned from the benchmark exercise. One challenge for this work is the large size of the domain and the relatively small size of the geometric details such as flow passages and gaps. The geometry is simplified to make meshing feasible by eliminating some of the small features. The model includes the lower plenum, control rod drive tubes, basic support structures, and SLCS piping. A series of simulations are completed and modeling choices are discussed in light of best practice guidance. The predictions shed light on the applicability of previous test programs and provide a clearer understanding of the lower plenum flows and mixing.

Assessment of passive safety system performance under main steam line break accident

A generation III+Boiling Water Reactor (BWR) which relies on natural circulation has evolved from earlier BWR designs by incorporating passive safety features which require no emergency injection pump and no operator action or Alternating Current (AC) power supply. The generation III+BWR’s passive safety systems include the Automatic Depressurization System (ADS), the Suppression Pool (SP), the Standby Liquid Control System (SLCS), the Gravity Driven Cooling System (GDCS), the Isolation Condenser System (ICS), and the Passive Containment Cooling System (PCCS). The ADS is actuated to rapidly depressurize the reactor leading to the GDCS injection. The large amount of water in the SP condenses steam from the reactor. The SLCS provides makeup water to the reactor. The GDCS injects water into the reactor by gravity head and provides cooling to the core. The ICS and the PCCS are used to remove the decay heat from the reactor. The objective of this paper is to analyze the response of passive safety systems under the Loss of Coolant Accident (LOCA). A Main Steam Line Break (MSLB) test has been conducted in the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) which is scaled to represent the generation III+BWR. The main results of PUMA MSLB test were that the reactor coolant level was well above the Top of Active Fuel (TAF) and the reactor containment pressure has remained below the design pressure. In particular, the minimum water level (1.706m) was 5% higher than the TAF (1.623m) and the containment maximum pressure (271kPa) was 35% lower than the safety limit (414kPa), respectively.

Failure analysis of the standby liquid control system for a boiling water reactor with fuzzy cognitive maps

A fuzzy cognitive maps (FCM) application is proposed as a simple method to determine failure modes and effects of the standby liquid control system (SLC) during anticipated transient without scram (ATWS) in a boiling water reactor (BWR). The SLC has an important contribution to the total core damage frequency in a BWR. This is the first step in the development of an expert system that could involve many emergency systems of a BWR to simulate accident sequences, through the knowledge representation and reasoning with FCM designs in order to automate the decision making process. A simplified model of the SLC 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, in order to see that both techniques show similar results even if the approaches are different.

Analysis of Reactivity Measurements in Core Physics Experiments on Full-MOX BWR

A core physics experimental program FUBILA has been performed to study core physics characteristics of full-MOX BWR cores consisting of high Pu-enriched MOX assemblies for high burnups. The program includes the measurement of reactivity worth, which is essential in operating BWR cores. The reactivity worth is due to the reactivity caused by (1) changes in the in-channel void fraction of assemblies, (2) the insertion of a B4C control blade, (3) Gd2O3-UO2 rods and UO2 rods in assemblies, and (4) the mixing of boron in a moderator related to a stand-by liquid control system. The reactivity worth was measured by the modified neutron source multiplication method using the reactivity worth of a pilot rod as a reference worth. The measured reactivity worth was determined by processing count rates of neutron detectors taking into account the detector efficiency and effective neutron source intensity analyzed by three-dimensional transport calculations. Comparing the measured reactivity worth with the results obtained by deterministic calculation methods and a Monte Carlo calculation method, the accuracy of the calculations was evaluated. The calculated results generally well reproduce the measurements, except for the boron reactivity worth in the moderator, for which the calculations overestimate the reactivity worth.

Analysis of Reactivity Measurements in Core Physics Experiments on Full-MOX BWR

A core physics experimental program FUBILA has been performed to study core physics characteristics of full-MOX BWR cores consisting of high Pu-enriched MOX assemblies for high burnups. The program includes the measurement of reactivity worth, which is essential in operating BWR cores. The reactivity worth is due to the reactivity caused by (1) changes in the in-channel void fraction of assemblies, (2) the insertion of a B4C control blade, (3) Gd2O3-UO2 rods and UO2 rods in assemblies, and (4) the mixing of boron in a moderator related to a stand-by liquid control system. The reactivity worth was measured by the modified neutron source multiplication method using the reactivity worth of a pilot rod as a reference worth. The measured reactivity worth was determined by processing count rates of neutron detectors taking into account the detector efficiency and effective neutron source intensity analyzed by three-dimensional transport calculations. Comparing the measured reactivity worth with the results obtained by deterministic calculation methods and a Monte Carlo calculation method, the accuracy of the calculations was evaluated. The calculated results generally well reproduce the measurements, except for the boron reactivity worth in the moderator, for which the calculations overestimate the reactivity worth.

Safety of Super LWR, (I) Safety System Design

This paper describes design concept of safety system of the high-temperature supercritical pressure light water cooled reactor with downward-flow water rods (Super LWR). Since this reactor is once-through cooling system without water level and coolant circulation, the fundamental safety requirement is keeping core coolant flow rate while that of light water reactors (LWR) is keeping coolant inventory. “Coolant supply from cold-leg” and “coolant outlet at hot-leg” are needed for it. The advantage of the once-through cooling system is that reactor depressurization induces core coolant flow and cools the core. The downward-flow water rod system enhances this effect because the top dome and the water rods supply its water inventory to the core like an “in-vessel accumulator.” The safety system of the Super LWR is designed referring to those of LWR in consideration of its characteristics and safety principle. “Coolant supply” is kept by high-pressure auxiliary feedwater system and low-pressure core injection system. “Coolant outlet” is kept by safety relief valves and automatic depressurization system. The Super LWR is equipped with two independent shutdown systems: reactor scram system and standby liquid control system. The capacities and the actuation conditions determined in this study are to be used in safety analysis.

Safety of Super LWR, (I) Safety System Design

This paper describes design concept of safety system of the high-temperature supercritical pressure light water cooled reactor with downward-flow water rods (Super LWR). Since this reactor is once-through cooling system without water level and coolant circulation, the fundamental safety requirement is keeping core coolant flow rate while that of light water reactors (LWR) is keeping coolant inventory. “Coolant supply from cold-leg” and “coolant outlet at hot-leg” are needed for it. The advantage of the once-through cooling system is that reactor depressurization induces core coolant flow and cools the core. The downward-flow water rod system enhances this effect because the top dome and the water rods supply its water inventory to the core like an “in-vessel accumulator.” The safety system of the Super LWR is designed referring to those of LWR in consideration of its characteristics and safety principle. “Coolant supply” is kept by high-pressure auxiliary feedwater system and low-pressure core injection system. “Coolant outlet” is kept by safety relief valves and automatic depressurization system. The Super LWR is equipped with two independent shutdown systems: reactor scram system and standby liquid control system. The capacities and the actuation conditions determined in this study are to be used in safety analysis.

Anticipated Transient Without Scram Analysis of the Simplified Boiling Water Reactor Following Main Steam Isolation Valve Closure with Boron Injection

The simplified boiling water reactor (SBWR) operating in natural circulation is designed with many passive safety features. An anticipated transient without scram (ATWS) initiated by inadvertent closure of the main steam isolation valve (MSIV) in an SBWR has been analyzed using the RAMONA-4B code of Brookhaven National Laboratory. This analysis demonstrates the predicted performance of the SBWR during an MSIV closure ATWS, followed by shutdown of the reactor through injection of boron into the reactor core from the standby liquid control system.

A Long-Term MAAP 3.0B Analysis of a Severe Anticipated Transient Without Scram Accident in a Boiling Water Reactor

This paper reports that the long-term responses of the reactor coolant system and the containment of a boiling water reactor (BWR) during an anticipated transient without scram (ATWS) initiated by an inadvertent closure of the main steam isolation valves (MSIVs) are analyzed with the MAAP 3.0B computer code. The plant used in the analysis is the Kuosheng nuclear power station, a General Electric BWR6 reactor with a Mark III containment. A comparison of near-term system responses to an MSIV closure ATWS event shows that the MAAP 3.0B code fails to reproduce the detailed thermal-hydraulic results of the sophisticated RETRAN-02 code. The MAAP code, however, makes a reasonable prediction of the timing of major events and core power in the ATWS accident analyzed. The results of a long-term MAAP ATWS analysis show that if the standby liquid control system fails in an MSIV closure ATWS event, the ATWS event might develop into a core melt accident with the containment failing before the core melts.

Effect of Flushing Boron During a Boiling Water Reactor Anticipated Transient Without Scram

This paper reports on an accident sequence in a boiling water reactor in which there is a large reactivity insertion caused by the flushing of borated water from the core. This sequence can occur during an anticipated transient without scram after the injection of borated water from the standby liquid control system. The boron shuts down the power, but if there is a rapid depressurization of the vessel (e.g. because of the inadvertent actuation of the automatic depressurization system), large amounts of low-pressure, relatively cold, unborated water enters the vessel causing a rapid dilution and cooling. This study determines whether the reactivity addition caused by this flushing could lead to a power excursion that is sufficient to cause catastrophic fuel damage.

ATWS scoping calculations for a BWR plant performed using RELAP5/MOD2 and contain codes

In the context of the licensing process for Alto Lazio nuclear power plant (a BWR/6, MARK III design) some calculatio have been performed for evaluating the plant response during an ATWS transient. Specifically, the ATWS related safety systems were simulated in order to evaluate their influence on the plant behaviour. Furthermore, the containment behavior and the effectiveness of hypothetical containment venting have been evaluated. Two best-estimate codes have been used for this study: • - RELAP5/MOD2 for the Reactor Cooling System (RCS) behaviour calculation; • - CONTAIN 1.1 for the containment response calculation. The initiating event chosen was a Main Steam Isolation Valve (MSIV) closure because the associated ATWS was found, by an Alto Lazio specific Probabilistic Risk Assessment study, to have the highest core melt frequency. In order to assess the corresponding effect, three different Stand-by Liquid Control System (SLCS) actuation modes have been considered: 1. (a) NOSLCS case: the SLCS is assumed to be unavailable; 2. (b) SLCS2 case: the SLCS starts two minutes after the initiating event; 3. (c) SLCS10 case: the SLCS is assumed to start 10 min after the initiating event. The paper contains the calculation results and the related analysis of the above-mentioned cases.