Purdue University Multi-Dimensional Integral Test Research Articles

Behaviors of passive safety systems under a Feed Water Line Break LOCA on a Generation III+ Boiling Water Reactor

A postulated Large Break Loss-Of-Coolant-Accident (LB LOCA) on a generic Generation III+ Boiling Water Reactor (BWR) equipped with passive safety systems has been analyzed by conducting integral tests at a scale-down test facility, named Purdue University Multi-Dimensional Integral Test Assembly (PUMA). The integral tests simulate the hypothetical Feed Water Line Break (FWLB) accident caused by a guillotine rupture of the Feed Water Line (FWL) attached to the reactor vessel. This research is motivated to evaluate the behaviors of passive safety systems and to provide experimental data for the Validation and Verification (V&V) of related system codes. The experiment simulates the LOCA transient after the reactor vessel depressurizes to 1 MPa (150 psi) from normal operating pressure. The early high-pressure blowdown process of LOCA transient is simulated by the system code. The plant system conditions at 1.0 MPa calculated by the system code are interpreted as the initial conditions for the integral test with appropriate scaling. The normal FWLB integral test is also repeated with the partial failure of the PCCS to evaluate the safety margin. Experimental data indicates that this LB LOCA is successfully mitigated by the functioning of Emergency Core Cooling System (ECCS) and Passive Containment Cooling system (PCCS), with no occurrence of core uncover or containment overpressure.

Experimental investigation of void distribution in suppression pool over the duration of a loss of coolant accident using steam–water two-phase mixture

Studies are underway to determine if a large amount gas discharged through the downcomer pipes in the pressure suppression chamber during the blowdown of Loss of Coolant Accident (LOCA) can potentially be entrained into the Emergency Core Cooling System (ECCS) suction piping of BWR. This may result in degraded ECCS pumps performance which could affect the ability to maintain or recover the water inventory level in the Reactor Pressure Vessel (RPV) during a LOCA. Therefore, it is very important to understand the void behavior in the pressure suppression chamber during the blowdown period of a LOCA. To address this issue, a set of experiments is conducted using the Purdue University Multi-Dimensional Integral Test Assembly for ESBWR applications (PUMA-E) facility. The geometry of the test apparatus is determined based on the basic geometrical scaling analysis from a prototypical BWR containment (MARK I) with a consideration of downcomer size, downcomer water submergence depth and Suppression Pool (SP) water level. Several instruments are installed in the test facility to measure the required experimental data such as the steam mass flow rate, void fraction, pressure and temperature.In the experiments, sequential flows of air, steam–air mixture and pure steam-each with the various flow rate conditions are injected from the Drywell (DW) through a downcomer pipe in the SP. Eight tests with two different downcomer sizes, various initial gas volumetric fluxes at the downcomer, and two different initial non-condensable gas concentration conditions in the DW are conducted. Three distinct phases, namely, an initial phase, a quasi-steady, and a chugging phase are observed. The maximum void penetration depth is observed in the initial phase. A reduction in the void penetration depth is observed in the quasi-steady phase. As a result of low non-condensable gas concentration, chugging is observed at the tail end of the experiment. Chugging provides renewed void penetrations comparable to those in the initial phase. It is determined that the void distribution and area of void penetration in the SP is governed by the gas volumetric flux at the downcomer and the non-condensable gas concentration in the downcomer.

Experimental investigation of void distribution in Suppression Pool during the initial blowdown period of a Loss of Coolant Accident using air–water two-phase mixture

During the initial blowdown period of a Loss of Coolant Accident (LOCA), the non-condensable gas initially contained in the BWR containment is discharged to the pressure suppression chamber through the blowdown pipes. The performance of Emergency Core Cooling System (ECCS) can be degraded due to the released gas ingestion into the suction intakes of the ECCS pumps. The understanding of the relevant phenomena in the pressure suppression chamber is important in analyzing potential gas intrusion into the suction intakes of ECCS pumps. To obtain the basic understanding of the relevant phenomena and the generic data of void distribution in the pressure suppression chamber during the initial blowdown period of a LOCA, tests with various blowdown conditions were conducted using the existing Suppression Pool (SP) tank of the integral test facility, called Purdue University Multi-Dimensional Integral Test Assembly for ESBWR applications (PUMA-E) facility, a scaled downcomer pipe installed in the PUMA-E SP, and air discharge pipe system. Two different diameter sizes of air injection pipe (0.076 and 0.102m), a range of air volumetric flux (7.9–24.7m/s), initial void conditions in an air injection pipe (fully void, partially void, and fully filled with water) and different air velocity ramp rates (1.0, 1.5, and 2.0s) are used to investigate the impact of the blowdown conditions to the void distribution in the SP. Two distinct phases of experiments, namely, an initial and a quasi-steady phase were observed. The maximum void penetration depth was experienced during the initial phase. The quasi-steady phase provided less void penetration depth and was categorized by oscillations in the void penetration. The initial void conditions in an air injection pipe and air volumetric fluxes were only found to significantly impact on the void distribution in the SP.

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.

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.

Counterpart experimental study of ISP-42 PANDA tests on PUMA facility

A counterpart test to the Passive Nachwärmeabfuhr und Druckabbau Test Anlage (Passive Decay Heat Removal and Depressurization Test Facility, PANDA) International Standard Problem (ISP)-42 test was conducted at the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) facility. Aimed to support code validation on a range of light water reactor (LWR) containment issues, the ISP-42 test consists of six sequential phases (Phases A–F) with separately defined initial and boundary conditions, addressing different stages of anticipated accident scenario and system responses. The counterpart test was performed from Phases A to D, which are within the scope of the normal integral tests performed on the PUMA facility. A scaling methodology was developed by using the PANDA facility as prototype and PUMA facility as test model, and an engineering scaling has been applied to the PUMA facility. The counterpart test results indicated that functions of passive safety systems, such as passive containment cooling system (PCCS) start-up, gravity-driven cooling system (GDCS) discharge, PCCS normal operation and overload function were confirmed in both the PANDA and PUMA facilities with qualitative similarities.

Assessment of performance of BWR passive safety systems in a small break LOCA with integral testing and code simulation

An integral test simulating a postulated Small Break Loss Of Coolant Accident (SB LOCA) has been conducted on a scaled-down test facility, Purdue University Multi-Dimensional Integral Test Assembly (PUMA). This test aims to simulate the hypothetical Bottom Drain Line Break (BDLB) accident initialized by an instantaneous guillotine rupture of the vessel Bottom Drain Line (BDL) in a generic design of natural circulation BWR equipped with passive safety systems. The objective of this research is to evaluate the ability of passive safety systems in this BWR to keep core from uncover and containment from overpressure during a SB LOCA transient by reproducing the LOCA transient in the test facility. The integral test simulates the interaction of reactor components and demonstrates the functionality of passive safety systems, such as Emergency Core Cooling System (ECCS) and Passive Containment Cooling System (PCCS). The experimental results showed that the two most important safety criteria were satisfied: the core coolant level was always higher than the Top of Active Fuel (TAF) and containment pressure did not exceed the design limit during this SB LOCA. The scaling principle of the test facility and RELAP5/Mod3.3 code capability were evaluated through the comparison between the experimental data and the code simulation.

ICONE19-43705 SUPPRESION POOL VOID DISTRIBUTION DURING BLOWDOWN

The void fraction distribution and fluid dynamics of the Boiling Water Reactor (BWR) Suppression Pool (SP) during blowdown is studied using the SP of the Purdue University Multi-Dimensional Integral Test Assembly for ESBWR applications (PUMA-E) facility. A pipe with inner diameter of 0.102 m is installed in the SP to simulate the downcomer pipe in the suppression chamber of the prototypic plant. To simulate the early blowdown stage of Loss Of Coolant Accident (LOCA), the air is injected air with a certain volumetric flow rate through the downcomer submerged in the water of SP. The void fraction distribution, bubble velocity, and bubble chord length around the downcomer exit are measured by the single-sensor and double-sensor conductivity probes mounted on the instrument cage. The sixteen tests which varied with the downcomer void condition, air volumetric flow rate and velocity ramp rate are performed and the test results are obtained. The parametric study reveals that the air volumetric flow rates have effect on the void fraction distribution and void penetration depth during the initial air injection phase and the quasi-steady phase. The initial downcomer void conditions strongly effect the void fraction distribution for the initial air injection phase while the ramp rates of air velocity have minor impact on the void fraction distribution in the both of phases.

Startup transient simulation for natural circulation boiling water reactors in PUMA facility

In view of the importance of instabilities that may occur at low-pressure and -flow conditions during the startup of natural circulation boiling water reactors, startup simulation experiments were performed in the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) facility. The simulations used pressure scaling and followed the startup procedure of a typical natural circulation boiling water reactor. Two simulation experiments were performed for the reactor dome pressures ranging from 55 kPa to 1 MPa, where the instabilities may occur. The experimental results show the signature of condensation-induced oscillations during the single-phase-to-two-phase natural circulation transition. The results also suggest that a rational startup procedure is needed to overcome the startup instabilities in natural circulation boiling water reactor designs.

The three-level scaling approach with application to the Purdue University Multi-Dimensional Integral Test Assembly (PUMA)

The three-level scaling approach was developed for the scientific design of an integral test facility and then it was applied to the design of the scaled facility known as the Purdue University Multi-Dimensional Integral Test Assembly (PUMA). The NRC Technical Program Group for severe accident scaling developed the conceptual framework for this scaling methodology. The present scaling method consists of the integral system scaling, whose components comprise the first two levels, and the phenomenological scaling constitutes the third level of scaling. More specifically, the scaling is considered as follows: (1) the integral response function scaling, (2) control volume and boundary flow scaling, and (3) local phenomena scaling. The first two levels are termed the top-down approach while the third level is the bottom-up approach. This scheme provides a scaling methodology that is practical and yields technically justifiable results. It ensures that both the steady state and dynamic conditions are simulated within each component, and also scales the inter-component mass and energy flows as well as the mass and energy inventories within each component.