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Analysis of the thermal-hydraulic system response of a PWR during loss of coolant accident and core blockage scenarios

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Introduction

The containment building of a typical Light Water Reactor (LWR) is designed to both contain the radioactive materials released during an accident and facilitate the core cooling during a loss-of-coolant accident (LOCA) event. Under such postulated accidents, the Emergency Core Cooling System (ECCS) provides the coolant flow which removes the decay heat from the reactor core and brings the system to a cold shutdown condition. During the first, blowdown, phase of the LOCA, the ECCS uses the cold water (nominal 30 °C) contained in a storage tank located inside the containment. The same water is discharged directly into the containment via Containment Sprays System (CSS) to keep its pressure beneath desired limits. Further, when the water of the tank is depleted, the cooling process continues using the water discharged from the break and which has sprayed into the reactor containment and collected in a sump. This phase is often identified as long-term cooling. In the event of a LOCA within the containment of a LWR the piping thermal insulation and other materials (e.g., coatings and concrete) in the vicinity of the break may be damaged and dislodged by the pipe break and the consequent impact of steam/water flow. A fraction of this material may be transported to the containment floor by the water flows induced by the break and by the containment sprays. To protect the ECCS components from possible damage induced by the debris eventually transported into the recirculation sump, a set of screens are typically installed for each ECCS train. The debris, which is accumulated on the sump screen, forms a bed that acts as a filter and could impact the plant’s capability to provide adequate long-term cooling water to the ECCS and CSS pumps. Small size debris could also pass through the debris bed and sump strainers and be injected into the primary system, reducing the core cooling capabilities (downstream effects)1.

The study of the physical phenomena occurring in the reactor containment during such events requires a detailed understanding and knowledge of the reactor system response during the different phases of the accident. In the following bullets, the important parameters used to analyze the thermal-hydraulic phenomena of interest in such analysis are listed, together with the phenomena that each parameter may affect:

 Break flow rate/integral mass flow defines the boundary conditions for the jet of water from the break, which in turn may affect the amount of debris produced for a given break location. It plays a role in the water velocity field in the reactor containment;

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 Break enthalpy flow defines the amount of energy transferred from the primary system to the containment, playing an important role in the containment pressure response;

 Primary Pressure imposes the boundary condition of the water discharged from the break as well as the injection flow rates through the high and low pressure injection pumps of the ECCS trains;

 Total ECCS Flow Rate affects rate of depletion of the water in the storage tank and the tank-to-sump switchover time. During the long-term cooling phase, this parameter affects the debris accumulation and bypass through the sump strainers and the pressure drop through the strainers.

System codes can be used for both the evaluation of the thermal-hydraulic parameters listed above and the analysis of the overall behavior of the reactor system during postulated events. Such codes can provide the user a large flexibility in modeling the entire reactor system, including ECCS and other safety features, with a relatively short computing time and low hardware requirements. RELAP5-3D2, one of the most used thermal-hydraulic system codes for best estimate simulations of normal operation and postulated transients (including LOCA) in LWRs, has been selected to perform an analysis of the reactor systems under selected LOCA scenarios of particular interest. RELAP5-3D code is the latest of the RELAP5 codes, developed by Idaho National Laboratory under sponsorship of the U.S. Department of Energy and the U.S. Nuclear Regulatory Commission. It comprises all the features of the RELAP5 code family plus the capability to include multi-dimensional components in the model, allowing the user to more accurately model the multi-dimensional flow behavior that can be exhibited in particular components or regions of a LWR. An input deck of a typical Westinghouse four-loop PWR plant was developed to support the thermal-hydraulic calculations of this activity. To account for the multi-dimensional flow behavior, the main regions of the reactor vessel were simulated using multi-dimensional components. These include the downcomer, the lower plenum and the portion of the upper plenum where the hot legs are connected, where more complex coolant flow paths are expected to occur during the different phases of the accident, considering the strong asymmetry of the flow due to the different injection locations available (hot or cold legs, intact or broken loops) and the relative position from the break location. The remaining parts of the vessel are modeled with one-dimensional components.

The phases of the LOCA scenarios in two of the main break locations of interest, such as the cold and the hot leg, were simulated and analyzed, with a particular interest on the long-term cooling phase. To provide a complete spectrum of the system behavior, the simulations were repeated for three different break sizes (2

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and 6 inches, and Double-Ended Guillotine), as representative cases for small, medium and large break respectively.

1.1 Historical background

In 1979, the NRC opened Unresolved Safety Issue (USI) A-433, “Containment Emergency Sump Performance” to study the effects of LOCA-generated insulation debris resulting from a pipe-break jet that is then transported to the sump debris screens and consequently blocks the screens, reducing NPSH margin below that required for recirculation pumps to maintain long-term cooling. This concern was particularly interesting since none of the recirculation systems’ strainers or sump screens were designed with regard to differential pressure loss and consequential strainer/screen structural integrity. The collection of a porous, uniform debris bed over the screens/strainers, and their ability to withstand a differential pressure caused by water flowing through this debris bed, was not a previously considered design and operational issue.

To support the resolution of USI A-43, documented in Generic Letter (GL) 85-224, “Potential for Loss of Post-LOCA Recirculation Capability Due to Insulation Debris Blockage”, the NRC undertook an extensive research program. It concluded that each plant should be evaluated on a plant-specific basis for debris-blockage potential and that the original 50-percent blockage assumption (under which most nuclear power plants had been licensed) identified in Regulatory Guide (RG) 1.825, “Sumps for Emergency Core Cooling and Containment Spray Systems”, may result in a non-conservative analysis for screen blockage effects. It was also updated the NRC’s regulatory guidance, including Section 6.2.2 of the Standard Review Plan6 and RG 1.82 to reflect the USI A-43 technical findings.

Following the resolution of USI A-43 in 1985, several events challenged the conclusion that no new requirements were necessary to prevent the clogging of ECCS strainers at operating BWRs:

 On July 28, 1992, at Barsebäck Unit 2, a Swedish BWR7, the spurious opening of a pilot-operated relief valve led to the plugging of two containment vessel spray system suction strainers with mineral wool and required operators to shut down the spray pumps and backflush the strainers.

 In 1993, at Perry Unit 18, two events occurred during which ECCS strainers became plugged with debris. On January 16, ECCS strainers were plugged with suppression pool particulate matter, and on April 14, an ECCS strainer was plugged with glass fiber from ventilation filters that had fallen into the

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suppression pool. On both occasions, the affected ECCS strainers were deformed by excessive differential pressure created by the debris plugging.

 On September 11, 1995, at Limerick Unit 19, following a manual scram due to a stuck-open safety/relief valve, operators observed fluctuating flow and pump motor current on the A loop of suppression pool cooling. The licensee later attributed these indications to a thin mat of fiber and sludge which had accumulated on the suction strainer.

In response to these ECCS suction strainer plugging events, the NRC requested that BWR licensees implement appropriate procedural measures, maintenance practices, and plant modifications to minimize the potential for the clogging of ECCS suction strainers by debris accumulation following a loss-of-coolant accident (LOCA)10, to which they responded by installing new large-surface-area suction strainers.

However, the findings from research to resolve the BWR strainer plugging issue in the late 1990s raised questions concerning the adequacy of PWR sump designs by confirming what the aforementioned BWR strainer plugging events had earlier indicated. These BWR research findings, which may also affect the performance of PWR sumps, prompted the NRC to open Generic Safety Issue (GSI) 19111, “Thermal-Hydraulic Response of PWR Reactor Coolant System and Containments to Selected Accident Sequences.”

The scope of GSI-191 addresses a variety of concerns associated with the operation of the ECCS and the CSS in the recirculation mode. These concerns include debris generation associated with a postulated high-energy line break (HELB), debris transport to the containment sump when the ECCS is realigned to operate in the recirculation mode, and the effects of ingestion of debris through the sump screens. In addition to debris resulting from the action of the jet from the postulated pipe break, there is also the potential for generation of chemical products from the reaction of containment materials and coolant that may also be transported to and through the sump screen.

A parametric evaluation12 was performed as part of the GSI-191 study to demonstrate the credibility of recirculation sump clogging for operating PWRs. Each of the 69 domestic PWRs was modeled in the evaluation using a mixture of generic and plant-specific data. The minimum amount of debris accumulation on the sump screen needed to exceed the required NPSH margin for the ECCS and CSS pumps was determined for each of the 69 representative models. Further, both completed and ongoing GSI-191 PWR research, as well as existing BWR research, were used to support the development of these models and the input to these models13. The evaluation considered small, medium, and large LOCAs using both favorable and unfavorable assumptions, relative to the plant, to a number of parameters. The results of the

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parametric evaluation formed a credible technical basis for making the determination that sump blockage was a credible concern. A risk study that supported the parametric evaluation14 was performed to estimate the amount by which the core damage frequency (CDF) would increase if failure of PWR ECCS recirculation cooling resulting from debris accumulation on the sump screen were accounted for in a manner that reflects the results of recent experimental and analytical work. Further, the estimate was made in a manner that reflected the total population of U.S. PWR plants. Results suggest that the conditional probability of recirculation sump failure, given a demand for recirculation cooling, is sufficiently high at many U.S. plants to cause an increase in the total CDF of an order of magnitude or more. However, the parametric evaluation had a number of limitations; the most notable were attributed to the extremely limited specific data available to the study. The need for more accurate plant-specific assessments of the adequacy of the recirculation function of the ECCS and CSS to be performed for each operating PWR was indicated clearly.

The Pressurized Water Reactor Owners Group (PWROG) undertook a program15 to provide analyses and information on the effect of debris and chemical products on core cooling for PWRs when the ECCS is realigned to recirculate coolant from the containment sump. The objective of the program is to demonstrate that there is reasonable assurance that sufficient long-term core cooling is achieved for PWRs to satisfy the requirements of 10 CFR 50.46 with debris and chemical products that are postulated to be transported to the reactor vessel and core post-LOCA by the coolant recirculating from the containment sump. Concerns have been raised about the potential for debris ingested into the ECCS to affect long-term core cooling when recirculating coolant from the containment sump. During operation of the ECCS to recirculate coolant from the containment sump, debris in the recirculating fluid that passes through the sump screen may collect on the bottom surface of the fuel assembly bottom nozzle, causing resistance to flow through this path. The collection of sufficient debris on the fuel assembly bottom nozzle is postulated to impede flow into the fuel assemblies and core. Other concerns have been raised with respect to the collection of debris and post-accident chemical products within the core itself. Specifically, the debris has been postulated to either form blockages or adhere to the cladding, thereby reducing the ability of the coolant to remove decay heat from the core. Similarly, chemical precipitants have been postulated to plate-out on fuel cladding, again resulting in a reduction of the ability of the coolant to remove decay heat from the core.

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1.2 Objective of the activity

The present work provides the results of the analysis of the simulations performed for a LOCA in two of the main break locations of interest, such as the cold and the hot leg, of a typical Westinghouse four-loop PWR plant. The main objectives of this analysis are:

 To develop and validate a RELAP5-3D nodalization of the reference nuclear power plant. Multi-dimensional components have been used to provide a more detailed estimation of the flow paths inside the reactor vessel and through the core during long-term cooling, when strong asymmetries are expected due to the different combinations of the break-injection location.

 To analyze the response of the plant during the selected accident scenarios and to provide boundary conditions for complementary analysis such as the break jet simulation and the reactor containment calculations.

 To investigate the behavior of the reactor system during a hypothetical core blockage, due to the accumulation of small size debris at the core inlet that could preclude the passage of water flow into the core.

The simulations were performed for three different break sizes (2 and 6 inches, and Double-Ended Guillotine16,17 as representative cases for small, medium and large break respectively) in order to analyze a complete spectrum of the system behavior. In addition, all the safety features of the reactor were assumed to be available during the transient and the boundary conditions defined in the input deck (including initial operating conditions, scram, injection and other setup points, ECCS flow rates, auxiliary injection rate and temperature, etc.) were assumed to be best estimate18.

1.3 Structure of the thesis

The present work consists of five chapters including this one (chapter 1), that provides the overall context in which the present activity arises, the purpose of the work and the organization of the thesis.

In chapter 2 is reported an overview of the LOCA scenarios and of the various phases that occurs during the transients. A distinction was made between large and small break LOCA, focusing on the phenomena involved during the course of events, for both scenarios. Then the safety concerns related to the downstream effects of debris ingested into the RCS during the long-term core cooling phase are discussed. In this section are specified the parameters that influence the debris behavior and the mechanisms by which these debris may interact with the internal components of the

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vessel affecting the core cooling capability. Finally, on the basis of these considerations, the selected scenarios are shown.

The chapter 3 describes the Nuclear Power Plant (NPP) used as reference for this activity. This includes a general overview of a typical Westinghouse four-loop PWR and a detailed description of the full RELAP5-3D model, component by component, developed for the simulations required. This nodalization includes also multi-dimensional components in those regions of the vessel where more complex flow paths are expected, like downcomer, lower plenum and upper plenum (3D Vessel – 1D Core model). Then the achievement of the steady-state conditions and the boundary conditions of the simulations are described, with respect to the nominal value of the plant. In addition a finer RELAP5-3D nodalization is provided, in which the one-dimensional component that simulates the core have been replaced with four multi-dimensional components (3D Vessel – 3D Core model) that allow a more detailed calculation of thermal-hydraulic processes that occur inside the core. This has been done mainly to study the reactor system behavior under hypothetical core blockage scenarios due the accumulation of small sizes debris at the core inlet that could bypass the sump screens and preclude the passage of water flow into the core.

The analysis of the results obtained for the selected scenarios is provided in the

chapter 4. First, the results obtained with the 3D Vessel – 1D Core model for LOCA

scenarios are described, focusing on the long-term cooling phase. Then scenarios that could lead to potential core damage, under conservative assumptions of full core inlet and core bypass blockage at the beginning of the long-term cooling phase were identified. Finally, the medium cold leg break LOCA scenario, which was identified to be one of the critical cases by the simulations performed with the 1D Core model, has been selected to be analyzed with 3D Vessel - 3D core model. A simulation of the long-term cooling scenario was repeated to compare the results obtained with the 3D Vessel – 1D Core model. Then a sensitivity analysis was performed to study the core cooling capability under different core blockage configuration, with a lower level of conservatism.

The discussion of the results obtained and the conclusions of the activity are provided in the chapter 5.

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