ML23089A229

Pre-Application Ipwr Pirt Report for NuScale-like Design

ML23089A229
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Issue date:	06/13/2011
From:	Peter Yarsky NRC/RES/DSA
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ML110130538
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NuScale proprietary information has been redacted from this document Pre-Application Phenomena Identification and Ranking Table (PIRT) Report for NuScale-like Integral Pressurized Water Reactors (iPWRs)

(Summary Report for Task 3 of iPWR User Need Response [ML110130538])

Manuscript completed: June 13, 2011 Division of Systems Analysis Office of Nuclear Regulatory Research 1

Pre-Application PIRT Report for NuScale-like Integral Pressurized Water Reactors (iPWRs)

Thermal-Hydraulic PIRT Panel Members:

D. Diamond (BNL)

S. Bajorek (NRC) - Panel Chair P. Griffith (MIT)

W. Krotiuk (NRC)

P. Lien (NRC)

B. Parks (NRC)

J. Staudenmeier (NRC)

Neutronics PIRT Panel Members:

S. Bajorek (NRC)

D. Diamond (BNL) - Panel Chair S. Frankl (NRC)

P. Griffith (MIT)

N. Hudson (NRC)

P. Yarsky (NRC)

Manuscript Preparations:

K. Tene (NRC)

AJ Nosek (NRC)

Project Manager: JR Skarda (NRC) 2

Executive Summary This report describes two pre-application Plant Phenomena Identification and Ranking Tables (PIRTs) developed for a NuScale-like design. The purpose of the PIRTs is to identify phenomena and process that will need to be considered when assessing safety related events for Integral Pressurized Water Reactors (iPWRs). Two events were considered for the design; a Thermal-Hydraulic Driven Event and a Neutronic Driven Event. Because the PIRTs have been developed early in the pre-application phase, neither final design data nor license application submittal data are available. Consequently the PIRTs have been based on best available information that has been obtained from potential applicants and liberal use engineering/expert judgment that has been applied by the PIRT panel members. A revision based on final design information is vital to the development and review of thermal-hydraulic and neutronic codes for licensing decisions.

The accident scenario considered for the thermal-hydraulic PIRT is the inadvertent opening of a reactor recirculation valve. An inadvertent control rod bank withdrawal from hot zero power (HZP) was the scenario considered for the neutronics PIRT. Both PIRTs identify phenomena of high importance, those that must be modeled accurately or bounded in an appropriate manner in a safety analysis to adequately assure margin to regulatory limits.

For the thermal-hydraulic PIRT, the figures-of-merit (FOMs) were the Critical Heat Flux (CHF) ,

mixture level in the vessel, and containment vessel pressure. The panel in general, did not feel that limits on the departure from nucleate boiling ratio (DNBR) would be challenged except possibly in the early part of the transient if the core flow were to stagnate while at or near full power operation. The mixture level was selected as a FOM because the safety systems are designed to prevent core uncovery and heatup. In the NuScale design, a conventional containment is replaced by a relatively small high-pressure containment vessel. During blowdown and early in the transient, high pressure within the containment vessel is a safety concern and therefore included as a FOM.

Of particular interest are those phenomena that are highly ranked in importance, but are poorly understood as indicated by a low knowledge level. For the NuScale-like design, the major concerns appear to be with the Containment Heat Removal System (CHRS), for which there were several phenomena the panel identified as having high importance and less than satisfactory knowledge level. Phenomena in this category include condensation heat transfer (both with and without noncondensables), and mixing in the containment as it affects the liquid pool and the gas space. There were additional concerns with choked flow from the vessel through the reactor vent valves and the reactor recirculation valves. These are mainly with the prediction with flow to the break and choked flow from the primary.

The neutronic PIRT considered a scenario whereby a control bank was inadvertently continuously withdrawn from hot zero power conditions. This would be expected to result in a power increase that would be offset initially by Doppler feedback and then by reactor trip.

Multiple FOMs were used to cover as many phenomena as possible. The FOMs were chosen by taking into account what would be the acceptance criteria for this event. Hence, the FOMs 3

were: DNBR, reactor vessel pressure, fuel centerline temperature, and pellet-clad interaction (PCI) / pellet-clad mechanical interaction (PCMI)

The panel listed 35 phenomena as having special significance in modeling this event. Most of these relate to either the neutronic initial conditions or neutron kinetics. The results show that 22 of the phenomena are of high (H) importance (having significant or dominant influence on one or more FOMs), 8 are of medium (M) importance, and 5 are of low (L) importance (having only a small influence on the FOMs). There are no phenomena that are of high importance for which the knowledge level is low (an H, L ranking). The next level of interest is for phenomena that have either a high importance with a medium knowledge level (H,M) or a medium importance with low knowledge level (M,L). They are listed below with their ranking (importance, knowledge level).

• Shutdown bank speed (H,M)

• Assembly Interaction (H,M)

• Axial/radial reflector representation (H,M)

• Pellet burnup distribution (H,M)

• Core pin-by-pin burnup distribution (H,M)

• Gap conductance (H,M)

• Fuel conductivity and density (H,M)

• CHF correlations (H,M)

• Detector response (M,L)

The deficiencies in knowledge level for these phenomena are summarized as follows:

• Shutdown bank speed, which partially defines the termination of the scenario, is determined, in addition to Technical Specifications, by bowing of guide tubes, and this bowing is difficult to predict.

• Assembly interaction is always an important consideration in modeling and becomes more uncertain when, as in the NuScale design, the core is small.

• Axial and radial reflector representation is more important in the NuScale design than in large PWRs, and the neutronics methods that are suitable for the large PWRs are not expected to be sufficient for the iPWRs.

• Pellet burnup distribution is important in determining the neutronic properties of the fuel and, because of self-shielding effects which are difficult to model, has a significant uncertainty.

• Core pin-by-pin burnup distribution is a factor in determining the thermo-mechanical properties of the fuel. Although power reconstruction methods exist, modeling is still uncertain.

• Gap conductance and fuel conductivity and density determine heat transfer to the coolant, but these properties may be uncertain due to uncertainties in the burnup of a given fuel rod and because of uncertainties in those properties as a function of burnup.

CHF correlations are directly related to one of the FOMs, but their use for short fuel assemblies is uncertain without measurements to confirm correlation applicability to these geometries. Detector response to changes in core power and delays between 4

signal receipt and the movement of control banks is uncertain because of a lack of design information rather than because the phenomenon is difficult to model.

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Table of Contents 1 INTRODUCTION ................................................................................................................. 8 2 PLANT DESCRIPTION........................................................................................................ 8 2.1 Plant Overview ............................................................................................................. 8 2.2 Reactor Core.............................................................................................................. 13 2.3 Pressurizer ................................................................................................................. 13 2.4 Helical Coil Steam Generator ..................................................................................... 13 2.5 High-Pressure Containment ....................................................................................... 14 2.6 Decay Heat Removal System .................................................................................... 15 2.7 Containment Heat Removal System........................................................................... 16 2.8 Emergency Core Cooling System .............................................................................. 17 3 PIRT METHODOLOGY ..................................................................................................... 17 3.1 General Approach ...................................................................................................... 17 3.2 Application to Preliminary Designs ............................................................................. 18 3.3 Ranking Scales .......................................................................................................... 18 3.3.1 Phenomena Importance Ranking ........................................................................ 18 3.3.2 Knowledge Level Ranking ................................................................................... 19 4 THERMAL-HYDRAULIC PIRT........................................................................................... 19 4.1 PIRT Panel Members ................................................................................................. 19 4.2 Scenario ..................................................................................................................... 20 4.3 Figures of Merit .......................................................................................................... 21 4.4 Systems, Subsystems and Components .................................................................... 22 4.5 Phenomena Ranking and Discussion ......................................................................... 23 4.5.1 Core System Phenomena ................................................................................... 23 4.5.2 Primary System Phenomena ............................................................................... 24 4.5.3 Secondary System Phenomena .......................................................................... 27 4.5.4 CHRS System Phenomena ................................................................................. 27 4.6 Phenomena / Knowledge Level Issues....................................................................... 32 4.6.1 Blowdown ............................................................................................................ 32 4.6.2 RVV Depressurization ......................................................................................... 32 4.6.3 Long-Term Cooling .............................................................................................. 32 5 NEUTRONICS PIRT .......................................................................................................... 33 5.1 PIRT Panel Members ................................................................................................. 33 6

5.2 Scenario ..................................................................................................................... 33 5.3 Figures of Merit .......................................................................................................... 34 5.4 Results and Discussion .............................................................................................. 34 5.4.1 Phenomena ......................................................................................................... 34 5.4.2 Ranking ............................................................................................................... 37 6.0

SUMMARY

AND CONCLUSIONS ..................................................................................... 40 6.1 Thermal-hydraulic PIRT ................................................................................................ 40 6.2 Neutronic PIRT .............................................................................................................. 40

7.0 REFERENCES

.............................................................................................................. 42 7

1 INTRODUCTION The U.S. Nuclear Regulatory Commission (NRC) has received notification of intent to submit design certification applications for integral pressurized water reactors (iPWRs) in 2012. The NRC has been requested to engage in pre-application review activities for iPWR designs. The Office of New Reactors (NRO) is proceeding with early efforts to identify and examine key technical and policy issues important to licensing of small- or medium-sized reactors designs.

The Office of Nuclear Regulatory Research (RES) is supporting NRO with these early efforts.

RES assistance focuses on the identification of unique and important phenomena, development of phenomena identification and ranking tables (PIRTs), and assessment of code capabilities that can provide insights to the aforementioned issues and provide better preparation for review of design certification applications. Results of this task will be used as the principal basis for a gap analysis and code applicability assessment of NRC thermal-hydraulic and neutronic computational tools.

In April 2011, PIRT Panels were convened to develop four iPWR PIRTs, two for a NuScale-like design and two for an mPower-like design. Two events or transients were considered for each design, a thermal-hydraulic driven event and a neutronic driven event. For the NuScale-like design, the thermal-hydraulic and the neutronics PIRT meetings were held on April 11, 2011 and April 13, 2011 respectively. During these meetings PIRTs were developed. However, these activities have been performed early in the pre-application phase when neither final design data nor license application submittal data are available. Consequently the PIRTs have been based on best available information obtained from potential applicants and that use of engineering/expert judgment has been applied by the PIRT panel members. The applicability of the PIRTs to subsequent iPWR application reviews is dependent on the degree of similarity of the preapplication design information to that of the final design information submitted for a license application. This report describes the two PIRTS developed for a NuScale-like design.

The PIRT panels were assisted by Dr. Raymond Skarda as Project Manager for the exercise and by Ms. Kimberly Tene in preparation of the final report.

An overview and general description of the NuScale-like plant design considered for this PIRT is provided in Chapter 2. Chapter 3 describes the methodology used for development of the PIRTs. Chapter 4 describes the PIRT for the thermal hydraulic driven event, and the neutronic driven event is discussed in Chapter 5. Conclusions are provided in Chapter 6.

2 PLANT DESCRIPTION 2.1 Plant Overview The information and figures contained in this section were drawn from references 1 as shown in the reference section.

The NuScale conceptual module, illustrated in Figures 2-1 through 2-4, is an integrated light water reactor (LWR) with passive safety features and a power rating of approximately 45 MWe 8

(160 MWt). The pressurizer, steam generator, hot leg, cold leg, and core are all housed in a shared reactor pressure vessel. A relatively small steel containment (approximately 60' x 15' or 18 m x 4.6 m) compared with that of a conventional PWR, envelopes the reactor pressure vessel. The containment vessel is partially evacuated during power operation and is capable of relatively high pressures during accident conditions. The entire module and containment are submerged in a pool of water. The NuScale plant is intended to be of modular design with up to 12 NuScale modules and containments at one site. Each module in the pool is supported by walls separating it from the other modules in the reactor pool. The reactor pool is a stainless steel-lined concrete pool shared by all of the operating modules. Each module is covered by an individual concrete impact shield, and all of the modules and pool are enclosed in a single confinement building.

Containment Reactor Vessel Containment Trunnion Helical Coil Steam Generator Nuclear Core Figure 2-1 Two views of the NuScale module. The view on the left also depicts a portion of the walls supporting the containment in the reactor pool 9

Generator Steam Turbine Condenser Water-Filled Pool Below Ground Reactor Module and Containment Figure 2 A perspective view of a portion of the plant with the water filled pool and one reactor module and containment 10

Figure 2-3 Planar view of a plant, with maximum modules Figure 2-4 Cross sectional view of plant with elevations 11

The NuScale conceptual design relies on passive safety systems and incorporates all large piping paths into the reactor vessel. The vendor postulates that the use of passive safety systems for decay heat removal, emergency core cooling, and containment cooling will eliminate external power requirements under accident cond itions. The NuScale modules, control room, and spent fuel pool are all located below grade and housed in controlled-access buildings.

The primary side flow path is shown in Figure 2-5. The core is located inside a shroud connected to the hot leg riser. Subcooled water enters the core, where it is heated and then flows vertically into the riser section. Circulation continues as hot water exits the riser into the upper plenum and then turns downward into the annulus housing the steam generators. Hot water in the annulus between the riser and the inside wall of the reactor vessel is cooled by the steam generator tubes. The cooled , d enser water descends through the downcomer into the lower plenum, then re-enters the core.

CONTAINMENT REACTOR Control Rod PRESSURE Drives VESSEL Reactor Vent _

Valves Control Rods Upper Plenum Steam SG Header Annulus/

Cold Leg __ Steam Generator Tubes Feed Header Hot Leg Riser Sump Reclr Downcomer Valves Core Shroud -

Figure 2-5 Primary Side Flow Path 12

2.2 Reactor Core The NuScale module's nuclear core is shown in Figure 2-6 and consists of 37 fuel assemblies arranged in a 17 x 17 square array, -. The core includes 16 control rod clusters.

Each fuel assembly includes 264 fuel pins, 24 control rods, and one instrument tube.

Figure 2-6 Core Configuration 2.3 Pressurizer The pressurizer in the NuScale module's Nuclear Steam Supply System provides reactor coolant system pressure control. A baffle region is located above the steam generator region to provide a barrier between the saturated fluid within the pressurizer and the subcooled reactor coolant system fluid. This baffle region limits the temperature of fluid that may surge into or out of the pressurizer region by mixing and heating the fluid as it moves about this region.

2.4 Helical Coil Steam Generator The NuScale steam generator is a helical-coil, once-through heat exchanger located in the annular space between the hot leg riser and the reactor vessel's inside wall. Feedwater enters the tubes at the bottom, and superheated steam exits at the top. Two independent sets of steam generator tube banks occupy the steam generator region as observed in Figure 2-7.

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Figure 2-7 Helical Coil Steam Generator 2.5 High-Pressure Containment NuScales containment vessel is dry and partially evacuated under normal operating conditions.

The vendor expects this configuration to eliminate moisture problems that could cause component corrosion and impact the reliability of instrumentation and other systems within containment. The partial vacuum reduces convection heat transfer without the use of direct-contact reactor vessel insulation. Due to a lack of appreciable amounts of air, the vendor also expects the vacuum to enhance steam condensation rates during reactor vessel blowdowns and prevents the formation of combustible concentrations of hydrogen mixtures in the event of a severe accident. The containment vessel is roughly cylindrical in shape approximately 60 x 15 (18m length x 4.6m diameter).

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Figure 2-8 High Pressure Containment 2.6 Decay Heat Removal System NuScale's decay heat removal system (DHRS) removes residual heat from the reactor core due to decay heat generation. The DHRS provides cooling for the core during normal shutdowns, station blackouts, and/or transients that result in a loss of normal feedwater. It has two independent piping trains, each intended to be capable of passively removing a sufficient fraction of the post-trip core power to prevent damage due to system heat-up.

During DHRS operation, cold water from the containment cooling pool is drawn from the inlet screen and sent to the steam generator tubes, where it transfers heat from the primary fluid and is evaporated. This steam is then vented and condensed in the containment pool. The steam generator removes heat from the reactor coolant in the reactor vessel annulus, creating a density difference between the hotter, lower-density coolant inside the riser and the cooler, higher-density coolant outside the riser. This density difference creates natural circulation of the reactor coolant in the same manner as during normal operation, but at a reduced flow rate. The check valves at various points in the DHRS prevent reverse flow.

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Each DHRS train has an inlet screen, an inlet line that connects the cooling pool to the main feedwater line, an inlet valve, an inlet isolation valve, an outlet line, an outlet isolation valve, and a vent sparger on the outlet. In addition, each feedwater line includes a pre-pressurized, water filled accumulator to provide continual feedwater flow during natural circulation startup of the DHRS.

Spargers Decay Heat Removal Sump -

Line Figure 2-9 Decay Heat Removal System

2. 7 Containment Heat Removal System Following a postulated loss-of-coolant accident (LOCA), the containment heat removal system (CHRS) rapidly reduces the containment pressure and temperature, consistent with the functioning of other associated systems, and maintains them at acceptably low levels for extended periods of time. The CHRS, an engineered safety feature, is classified as a "system,"

even though it only consists of the containment cooling pool water and containment vessel. The containment cooling pool consists of a large, below-grade concrete pool that is designed to provide stable, ample cooling for the containment for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following any LOCA without any active heat removal from the pool. Following a postulated break in the primary system, steam released into the containment would be condensed on the inside surface of the containment wall, which, in turn, would be passively cooled by conduction and convection heat transfer with the cooling pool. Because the containment would be evacuated during normal operation, a low level of noncondensable gases would be present inside the containment, and condensation heat transfer rates would increase.

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Sump Reclr Valvos (Inlet)

Figure 2-10 Containment Heat Removal System 2.8 Emergency Core Cooling System NuScale's emergency core cooling system (ECCS) consists of two independent reactor vent valves (RWs), two independent reactor recirculation valves (RRVs), and the CHRS. The ECCS provides a means of core decay heat removal in the event of loss of the main feedwater flow in conjunction with the loss of both trains of the DHRS. Long-term cooling is established through reactor recirculation cooling via the ECCS flow path in the event of a LOCA. The ECCS is initiated by opening the RWs and RRVs. Opening these valves creates a path by which water condensed on the inside surface of the containment flows into the reactor coolant system via the RRVs. Opening the RWs establishes a natural circulation path whereby water that is boiled in the core leaves through the RWs, is condensed and collected in the containment, and then is reintroduced into the downcomer through the RRVs.

3 PIRT METHODOLOGY 3.1 General Approach Development of a Phenomena Identification and Ranking Table (PIRT) for a new reactor design can be characterized as a systematic, multi-step process to determine which physical 17

phenomena are of most importance to a particular scenario. Because of the size and complexity of the reactor system, and its expected response to various boundary conditions and system interactions, the system needs to be evaluated in terms of its individual components such as the core and steam generator. The overall objective of the PIRT is to define requirements and capabilities of an Evaluation Model (EM). Specific steps that are part of the PIRT process are:

1. Definition of Objectives: The first step is to specify the purpose of the analysis required by the EM and to identify the type of reactor system for which the PIRT applies.

2. Identification of Systems and Subcomponents: A reactor coolant system often contains several safety-related components. These components and interacting systems should be identified as part of the PIRT ranking process.

3. Scenario Selection: The accident type and general scenario are identified. The scenario can be sub-divided into several periods as phenomena evolve or as various sub-systems become activated or de-activated.

4. Specify Figure(s) of Merit: Figures of merit are quantitative standards of acceptance for the safety analysis. While the General Design Criteria (GDC) in Appendix A of 10CFR50.46 describe general requirements for safe reactor operation, specific parameters that are outputs of the EM are desirable as figures of merit for a PIRT.

5. Identification of Phenomena and Importance Ranking: Physical phenomena and processes that affect each component are identified and ranking in terms of importance with respect to the figure(s) of merit.

6. Knowledge Level Ranking: The supporting knowledge base and uncertainties associated with processes and phenomena should be characterized. The ranking of the knowledge base should be based on the state-of-the-art ability to understand and model a process, rather than a statement on whether the process is modeled acceptably in a specific thermal-hydraulic code.

7. Documentation: The documentation of a PIRT should include the phenomena and their rankings, along with the rationale by which the panel arrived at its ranking.

3.2 Application to Preliminary Designs The PIRT process described in the preceding section was applied to the plant as reported in Section 2. Additional information was reviewed by the panel in References 1-3. When the licensing basis design is available, assumptions, results, and overall applicability of this PIRT should be re-evaluated.

3.3 Ranking Scales 3.3.1 Phenomena Importance Ranking The panel determined the importance of a particular phenomenon based on expected effect the phenomena may have on the Figure(s) of Merit (FOMs) for the scenario. A summary of the importance rankings is provided in Table 3-1. Phenomena ranked high (H) are considered likely to have a dominant role in the transient. Phenomena with a high (H) rank would be expected to be modeled with a high degree of accuracy, or be addressed through sensitivity studies to determine the resulting uncertainty in the FOM. Phenomena receiving lower importance rankings [medium (M) or low (L)] are expected to have relatively less impact on the transient, and it may be possible to model such processes with lesser accuracy or lesser concern with the uncertainty in calculating the FOM. The inactive (I) ranking is reserved for those processes that either do not or cannot occur during parts of the transient.

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Table 3-1 Phenomena Importance Ranking Table Ranking Description High (H) Significant or dominant influence on one or more FOM.

Medium (M) Moderate influence on one or more FOM Low (L) Small influence on one or more FOM Inactive (I) Phenomena not present or possible 3.3.2 Knowledge Level Ranking The panel also commented on the state of knowledge for the processes receiving an importance ranking. The knowledge level ranking is meant to serve as an assessment of technical information available for the phenomenon as it may occur in the full-scale prototype during the transient. The knowledge level ranking is an assessment of the ability to model and simulate the phenomena accurately. It is not an assessment on the adequacy of how well the phenomenon is modeled and simulated by a particular analysis code. That type of an assessment should be made in a Code Applicability Report where the codes models and correlations are considered relative to the needs defined by the PIRT for a given transient scenario. Table 3-2 summarizes the knowledge level ranking criteria for this PIRT.

Table 3-2 Knowledge Level Ranking Criteria Ranking Description High (H) Phenomenon is well understood, and can be accurately modeled.

Medium (M) Phenomenon is understood, however may only be modeled with moderate uncertainty.

Low (L) Phenomenon is not well understood. Modeling the phenomena is currently either not possible or is possible only with large uncertainty.

4 THERMAL-HYDRAULIC PIRT 4.1 PIRT Panel Members The following individuals participated on the panel to develop a thermal-hydraulic PIRT for the NuScale reactor:

• Dr. Stephen M. Bajorek (Chair), Senior Advisor for Thermal-Hydraulics, NRC/RES/DSA

• Dr. David Diamond (Facilitator), Brookhaven National Laboratory

• Dr. Peter Griffith, Professor Emeritus, MIT

• Dr. William Krotiuk, NRC/RES/DSA/RSAB

• Dr. Peter Lien, Senior Reactor Engineer, NRC/RES/DSA/RSAB 19

• Dr. Benjamin Parks, Senior Reactor Engineer, NRC/NRR/DSS/SRXB

• Dr. Joseph Staudenmeier, Senior Reactor Engineer, NRC/RES/DSA/CDB 4.2 Scenario No large-break LOCA is possible in the NuScale reactor due to its integral design and absence of large penetrations in the reactor vessel. Thus, only small-break LOCAs need to be considered. Based on information on the NuScale design, there are a limited number of vessel penetrations that may result in significant inventory loss. The inadvertent opening of valves at the top of the pressurizer would result in a rapid depressurization, but the flow out of the vessel would be primarily steam. The lowest possible penetration to the reactor vessel is the reactor recirculation valve (RRV). An inadvertent opening of one of these valves would immediately result in the discharge of subcooled liquid, which would rapidly deplete vessel inventory and have a faster depressurization of reactor and faster pressurization of the containment vessel than a release from the top of the vessel. Therefore, the thermal-hydraulic accident scenario considered for this PIRT is the inadvertent opening of a reactor recirculation valve.

The assumed scenario for the inadvertent opening of a RRV follows. The LOCA is divided into three distinct periods, as outlined in Table 4-1.

Table 4-1 Periods Considered in NuScale Thermal-Hydraulic PIRT Period Period Description Start 1 Blowdown Reactor recirculation valve opens.

2 RVV Depressurization Reactor vent valves open on low primary pressure (6 MPa) 3 Long Term Cooling Primary and containment pressure equilibrate.

Period 1, the blowdown phase begins with the inadvertent opening of one of the reactor recirculation valves. Subcooled critical flow is discharge from the vessel to containment during this initial period of the transient as the vessel depressurizes. Almost immediately, a reactor protection system trip signal occurs on loss of containment vacuum. The containment vessel begins to flood, the turbine and main feedwater pumps trip, and both the main steam line and the main feedwater lines are isolated. The pressurizer water level decreases and eventually reaches the level of the baffle plate. When the pressurizer pressure decreases to 6 MPa (870 psia), the reactor vent valves begin to open.

Period 2, the reactor vent valve depressurization period begins when valves at the top of the pressurizer open and release steam into the containment vessel. This increases the depressurization rate and during this period inventory is lost to containment from both the reactor vent valves and the inadvertently open reactor recirculation valve. Steam condenses on the containment vessel wall and drains into the pool of water collecting in containment. This containment pool level increases as level within the reactor vessel decreases. Heat is also removed by the helical coiled steam generators, from inventory held inside the tubes which eventually boil dry since the feedwater pumps have tripped. The second reactor recirculation 20

valve opens near the end of this period following a timed delay and the pressures in the reactor vessel and containment near equilibrium.

Period 3, the final period, is long-term cooling. This period begins once both RRVs are open and the pressure has equilibrated between reactor vessel and the containment vessel. The water level in containment is expected to be above the RRVs and higher than the level in the reactor vessel. The hydrostatic head will force water into the RRVs, while steam continues to be released from the RVVs. Steam entering containment continues to be condensed on the containment vessel wall and drain into the containment water. Heat is removed from the containment to the water in the containment vessel cooling pool which surrounds that vessel.

4.3 Figures of Merit The Figure(s) of Merit (FOM) for an accident scenario refer to the parameters by which success or failure of the safety systems is determined. Because the accident scenario can have several periods, the FOM may change and depend on the specific period. For small modular reactors, the FOM criteria as specified by 10 CFR 50.46 (peak cladding temperature and the maximum local clad oxidation) are not expected to be challenged. The success of the safety systems is therefore judged on parameters that ensure a wide margin to the 10 CFR 50.46 regulatory criteria.

A loss-of-coolant accident, such as the inadvertent opening of a RRV or RVV, is not expected to result in core uncovery. To determine if cladding heat up occurs as a result of uncovery, the two-phase mixture level is selected as a figure of merit. The mixture level refers to the minimum level in the core or vessel where sufficient liquid is present to prevent dryout. As long as the mixture level, with uncertainty, remains above the top of the core, there is little likelihood that cladding temperatures will exceed the local saturation temperature. The mixture level is considered important during the RVV depressurization and long-term cooling periods of a LOCA transient in the NuScale-like design. Because the blowdown period is short and the water level remains in the pressurizer, the mixture level was not considered important in that early period.

If a significant temperature increase in the fuel and cladding occurs during a LOCA, it is considered more likely to be the result of high-power locations exceeding the critical heat flux (CHF). The critical heat flux is that heat flux at which there is sufficient vapor blanketing or bubble clotting on a fuel rod such that heat is not effectively removed and a rapid temperature rise occurs. Since this phenomenon was considered possible only when the power was high, CHF was selected as a FOM only during the first two periods of the scenario.

NuScale transients, with coolant discharge from the reactor vessel to the containment vessel, are considered tightly coupled. That is, the transient in the reactor vessel can be strongly influenced by phenomena and conditions in the containment. The containment pressure is expected to increase rapidly during a RRV or RVV opening, and flow to the reactor vessel depends on when the pressures between the two vessels equilibrate. The NuScale containment vessel is also an important barrier to the release of fission products, and it must be demonstrated that pressures and temperatures remain within acceptable design limits.

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Therefore, during the blowdown and RVV depressurization periods, containment pressure is considered an important FOM.

For the inadvertent opening of an RVV in a NuScale system, the FOMs by period are listed in Table 4-2.

Table 4-2 Figures of Merit for NuScale T/H PIRT Containment Period CHF Mixture Level Pressure Blowdown X X RVV Depressurization X X X Long-Term Cooling X 4.4 Systems, Subsystems and Components The panel divided the NuScale design into systems, subsystems and components in order to better identify significant phenomena that influence the FOM. The major systems were considered to be the Core and the Primary System, which is composed of the riser, upper plenum, pressurizer, steam generator annulus, downcomer, and the lower plenum. The reactor vent valves and reactor recirculation valves were also considered to be parts of the primary system. The steam generator secondary side was considered to be an individual system. The containment heat removal system (CHRS) is composed of the containment vessel and containment vessel cooling pool. Table 4-3 summarizes the main systems, subsystems and components considered in the NuScale PIRT.

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Table 4-3 Systems and Subsystems System Subsystems Core Fuel rods Coolant Subchannel Barrel/Baffle Control Rods/ Guide Tube Corewide flow Primary Riser Upper plenum Pressurizer Reactor vent valves Steam generator annulus Downcomer Reactor recirculation valves Lower plenum Secondary Steam generator tubes (outside)

Containment Heat Removal Containment vessel System (CHRS) Containment cooling pool 4.5 Phenomena Ranking and Discussion This section presents the ranking of each phenomenon, and provides a brief rationale on how or why the panel arrived at the ranking. Each system, core, primary, secondary, and CHRS, is discussed. Table 4-4 lists the phenomena, along with the importance rank and the knowledge level assessment.

4.5.1 Core System Phenomena 4.5.1.1 Fuel Rods The fuel in the NuScale core is conventional PWR reactor fuel. However, the core and rods are approximately one-half height compared to most operating PWRs. The panel considered the fuel to be composed of the UO2 fuel pellets, the gap between the pellets and the cladding, and the zircalloy-based cladding. Some of the design parameters, such as initial internal pressure, were not available at the time this PIRT was developed.

The panel considered Decay Heat to be a highly ranked process throughout the transient due to its effect on critical heat flux and mixture level. The knowledge level for Decay Heat was considered to be high. Other fuel-related parameters and phenomena such as stored energy, gap conductance and fuel thermal conductivity, were not considered to be of much importance since the core is covered with liquid and most of the initial energy will be transferred to the coolant early in the event. The total peaking factor, both magnitude and location, was given an M importance rank during blowdown because of its potential impact on CHF. Core design methods were considered sufficiently accurate such that the value and location of the peak power location would be known and therefore given a high knowledge level ranking. Because 23

the cladding is expected to remain covered and to remain near saturation temperature, cladding oxidation was not considered an active process and therefore was assigned an I ranking.

4.5.1.2 Coolant Subchannel Coolant subchannel processes and phenomena are those that occur in the fluid. These include single- and two-phase pressure drop and fluid mixing. During the blowdown period, processes that were considered more important were those that determined the velocity and void distributions as the core flow stagnates early in the event. Two-phase pressure drop, natural circulation, CHF (correlations), flow regime transition, cross flow / mixing, and void distribution were assigned an M importance rank as these were considered processes that would determine the onset of CHF. If CHF, by bubble clotting or vapor blanketing, were to occur in an RRV or RVV inadvertent opening in a NuScale reactor, it is much more likely to occur during blowdown as opposed to later periods. Therefore, most of the phenomena important for blowdown CHF were considered of low importance during the RVV Depressurization and Long-Term Cooling periods. Natural circulation retained an M importance ranking during the RVV Depressurization period and Interfacial Drag was ranked M for both the RVV Depressurization and Long Term Cooling periods due to their importance on mixture level. The knowledge level of most processes was considered M or H throughout the transient. The CHF (correlations) were given a low knowledge level ranking because available correlations were considered likely be out of range at the low flow rates expected during RVV Depressurization and Long-Term Cooling.

4.5.1.3 Barrel/Baffle The Barrel/Baffle subsystem refers to the region surrounding the core. This region may result in a bypass of coolant from the core, and large structures may contribute to heat release during the event. The panel considered processes associated with this region to be unimportant because the heat transfer is insignificant relative to other components, and assigned an L importance ranking. The knowledge level of the processes was considered high.

4.5.1.4 Control Rods / Guide Tube The effect of the control rods and guide tubes on the flow distribution and heat release by the structures was considered by the panel. In each case the potential impacts were considered small, and each process was assigned a low importance ranking. Knowledge level was considered high.

4.5.1.5 Corewide Flow Corewide flow refers to the distribution of fluid velocity laterally across the core. The panel considered this as a potential contributor to flow-induced stability, but decided that a low importance ranking was appropriate.

4.5.2 Primary System Phenomena 4.5.2.1 Riser The hot leg riser contains flow that goes from the core to the upper plenum. Since this fluid is at Thot, flashing will occur in this component before it occurs elsewhere in the vessel. The panel considered flashing to be a process of high importance for the riser due to its effect on void formation. Primary natural circulation, and its effect on the bulk coolant flow, was considered as 24

a process of medium importance in the riser during the initial blowdown period but of low importance for later periods as the level dropped below the top of the riser and the path for natural circulation was broken. Interfacial drag between the phases, and two-phase level swell were considered to be of medium importance during the RVV Depressurization and Long-Term Cooling periods as these processes contribute to the determination of the mixture level.

Vertical and radial natural circulation refers to natural circulation internal to the riser itself in which hot fluid in the center of the riser flows up while cooler fluid flows down along the riser wall. Since a large temperature difference exists across the riser wall during most of the transient, this process was given a low importance ranking by the panel. Inlet flow and temperature distribution refers to variations in velocity and temperature entering the riser from the core. These were not considered significant or to have a major effect on the mixture level in the riser and therefore were given a low importance ranking.

4.5.2.2 Upper Plenum The upper plenum in the NuScale vessel is the relatively small region above the top of the riser but below the pressurizer. Flow in this region is at Thot, and the flow is turning towards the annulus between the riser and the vessel outer surface.

4.5.2.3 Pressurizer The pressurizer provides reactor coolant system pressure control. The pressurizer is separated from the rest of the vessel by a baffle that serves as a barrier between the saturated fluid within the pressurizer and the subcooled reactor coolant system fluid. This baffle region limits the temperature of fluid that may surge into or out of the pressurizer region by mixing and heating the fluid as it moves about this region.

The liquid level in the pressurizer and the vessel pressure are important setpoints that determine the opening of the reactor vent valves. Therefore, during blowdown, flashing was considered a highly ranked phenomenon, and phase separation and interfacial drag/relative motion between the phases were both considered to be of medium importance. During the Reactor Vent Valve Depressurization period, these processes remain important, but because the rate of depressurization slows somewhat, flashing has lesser impact than in blowdown.

During the Reactor Vent Valve Depressurization period flow into the pressurizer to the vent valves occurs, and two-phase pressure drop through the pressurizer is considered to have medium importance. Flooding at the baffle plate is expected to occur, limiting the drain rate of the pressurizer, and is considered to be of high importance.

When the Long-Term Cooling period occurs, the pressurizer is expected to be empty and there will be little flow though the pressurizer to the vent valves. All pressurizer processes at this time were therefore considered to have low importance.

4.5.2.4 Reactor Vent Valves (RVV)

The reactor vent valves are located at the top of the pressurizer, and provide a relief path for steam to exit the vessel and condense on the walls of the containment vessel. The vent valves are inactive during the blowdown period in an inadvertent RRV opening event, but affect the transient once they open on a low pressure or low water level signal. During the RVV 25

Depressurization and Long-Term Cooling periods, the RVVs act as a second break to the vessel. Early in time, the flow is expected to be choked. As the vessel level decreases and the pressure becomes low, the flow may become unchoked and single phase. The panel therefore considered choked flow and single phase pressure drop to be of high importance during the RVV Depressurization period, and then having less importance during Long-Term Cooling as the importance ranks were decreased to L and M respectively.

4.5.2.5 Steam Generator Annulus The steam generator annulus refers to phenomena that occur in the fluid space outside of the helical coiled tubes. Because the steam generator is isolated early in the transient and later boils dry, only several processes in the blowdown period were considered to have much importance on the FOMs. In particular, processes that affect natural circulation through the vessel, flashing, two-phase pressure drop, and natural circulation flow were considered to have medium importance. The feed header and steam header are potentially important resistances to flow, and these were therefore also given an M importance ranking. Heat removal during the blowdown period is expected to be by convective heat transfer to the SG tubes. Since the rate of heat transfer may remain significant during blowdown, this process was given a medium importance ranking.

Following the blowdown period, flow through the steam generator annulus is expected to be very low and heat transfer to and from the steam generator also very low. Therefore, all steam generator processes were considered to have low importance for the RVV Depressurization and Long-Term Cooling Periods.

4.5.2.6 Downcomer The downcomer is the annular region between the core barrel and the vessel inner surface that is below the helical coil steam generator. Flashing, two-phase level swell, interfacial drag/relative motion between phases, and the effect of this region on the natural circulation of bulk flow in the vessel were considered to have medium importance due to their effect on flow to the core and the resulting effect on CHF during the blowdown period. During the RVV Depressurization period two-phase level swell, interfacial drag/relative motion between phases and stored energy release from the vessel wall were considered to have medium importance due to their effect on the distribution of void and fluid temperatures due to the potential impact of flow to the reactor recirculation valves. During Long-Term Cooling, only stored energy release from the vessel wall was considered to have medium importance. All other processes considered were assigned low importance rankings.

4.5.2.7 Reactor Recirculation Valves (RRV)

The reactor recirculation valves (RRV) act as the break for the inadvertent opening event.

During blowdown, single- and two-phase pressure drop, as well as choked flow, were all considered phenomena of high importance due to their impact on all three FOMs. After the RVVs open, flow out the RRVs is expected to remain choked and was therefore considered along with the two-phase pressure drop to have high importance. During the Long-Term Cooling period, flow in and out of the RRVs is expected to be single-phase and unchoked, so single-phase pressure drop was considered to have medium importance with other processes ranked low.

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4.5.2.8 Lower Plenum Processes in the lower plenum that may have an impact on the FOMs included flashing, single-phase pressure drop, and the effect of that region on natural circulation bulk flow rate. Each of these phenomena was assigned a medium importance rank based on the possibility that these could affect the core flow rate and therefore the CHF. During the RVV Depressurization and Long Term Cooling periods, energy release from the vessel wall was considered a process of medium importance.

4.5.3 Secondary System Phenomena For an inadvertent opening of an RRV, the secondary of the NuScale design has only a limited involvement. Following loss of containment vacuum, the main feed pump trips, and the main feed line is isolated. Inventory within the steam generator tubes is boiled off. Eventually the inside of the tubes is empty, and the generator ceases to remove heat from the vessel.

Assuming that most of the inventory boils away during blowdown or shortly thereafter, the phenomena of flashing, single-phase convection heat transfer and two-phase convection heat transfer were given a medium importance ranking during blowdown and a low importance ranking for later periods. Other phenomena associated with the secondary side were considered to be of low importance.

4.5.4 CHRS System Phenomena The containment heat removal system (CHRS) is intended to maintain the containment environment (pressure and temperature) at acceptable levels throughout the entire transient and for extended periods of time. The CHRS is classified as a system, even though it only consists of the containment vessel itself and the containment cooling pool that surrounds it.

The containment cooling pool consists of a large, below-grade concrete pool that is designed to provide stable, ample cooling for the containment for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following any LOCA without any active heat removal from the pool.

4.5.4.1 Containment Vessel Following a postulated break in the primary system, steam released into the containment would be condensed on the inside surface of the containment wall, which, in turn, would be passively cooled by conduction and convection heat transfer with the cooling pool. Because the containment would be evacuated during normal operation, a low level of noncondensable gases would be present inside the containment.

Since condensation heat transfer is critical to removing heat from the vessel, phenomena that affect condensation rates were considered of high importance to the determination of the containment pressure FOM. During all three periods of the transient, processes of high importance were condensation heat transfer on the wall of the containment vessel, natural circulation flow in the gas-space of the containment vessel, and the effect of non-condensables.

While the NuScale design asserts that a high vacuum will be maintained during normal operation, the panel considered it unlikely that all noncondensables could be eliminated or prevented during a LOCA. Non-condensables may be present from dissolved nitrogen coming out of solution from the coolant, for example. During blowdown, the effect of thermal properties for fluid below atmospheric pressure was assigned a medium importance ranking due to the potential uncertainties it may create in an analysis. Conduction heat transfer through the 27

reactor vessel wall and the containment vessel wall were considered processes of medium importance due to the resistance to heat transfer that they represent.

By the start of the Long-Term Cooling period, a liquid level will have developed in the containment vessel. Thermal stratification, mixing by internal natural circulation, and heat transfer between the pool and the gas space will affect the containment pressure as well as the temperature of fluid returning to the reactor vessel through the RRVs. Therefore, the panel assigned a medium importance rank to stratification, natural circulation, and interfacial heat transfer at the pool surface for this period.

4.5.4.2 Containment Cooling Pool The containment cooling pool was assumed to contain a large volume of water that would require a very long period of time to heat significantly. The panel assigned low importance rankings to all processes for the blowdown and RW Depressurization periods since natural convection heat transfer should be sufficient to remove heat from the containment vessel. Late in the transient, during the Long-Term Cooling period, some other heat transfer processes may become significant and important to the FOMs. Therefore, in Long-Term Cooling, nucleate boiling, single phase convective heat transfer, natural circulation in the cooling pool, and thermal stratification in the pool were considered to be of medium importance.

Table 4-4 Phenomena and Knowledge Level Rankings RVV System Component Process/Phenomena BLD LTC KL AP CORE FUEL RODS Decay Heat H H H H CORE FUEL RODS Fission Power L L L H CORE FUEL RODS Stored Energy L L L M CORE FUEL RODS Gap Conductance L L L M CORE FUEL RODS Fuel Conductivity L L L M CORE FUEL RODS Initial Gap Pressure L L L M CORE FUEL RODS Cladding Conductivity L L L M CORE FUEL RODS Cladding Oxidation (CORRELATIONS) I I I H CORE FUEL RODS Total Peaking Factor M L L H CORE FUEL RODS Burnup Distribution L L L H CORE FUEL RODS Boron Precipitation L L L M RVV System Component Process/Phenomena BLD LTC KL AP CORE SUBCHANNEL Single Phase Pressure Drop L L L H CORE SUBCHANNEL Two Phase Pressure Drop M L L M CORE SUBCHANNEL Flashing L L L H CORE SUBCHANNEL Natural Circulation M M L M CORE SUBCHANNEL lnterfacial Drag L M M M CORE SUBCHANNEL Single Phase Convection L I I H CORE SUBCHANNEL Two Phase Convection L L L H CORE SUBCHANNEL CHF (correlations) M L L L CORE SUBCHANNEL Flow regime transition M L L M Grid spacer effects CORE SUBCHANNEL (entrainment/deentrainment) I I I M 28

CORE SUBCHANNEL Grid spacer effects (heat transfer) L L L M CORE SUBCHANNEL Cross flow / mixing M L L M CORE SUBCHANNEL Clad ballooning L L L M CORE SUBCHANNEL Void distribution M L L M CORE SUBCHANNEL Turbulent mixing L L L L CORE SUBCHANNEL Boron blockage in subchannels L L L M RVV System Component Process/Phenomena BLD LTC KL AP CORE Barrel/Baffle Stored energy L L L H CORE Barrel/Baffle Bypass flow L L L H RVV System Component Process/Phenomena BLD LTC KL AP CORE Control rods/GT Effects on flow L L L M CORE Control rods/GT Effects on heat transfer L L L H RVV System Component Process/Phenomena BLD LTC KL AP CORE COREWIDE FLOW Stability L L L M RVV System Component Process/Phenomena BLD LTC KL AP PRIMARY Hot leg riser Flashing H H L H PRIMARY Hot leg riser Two-phase level swell L M M M PRIMARY Hot leg riser Two-phase pressure drop L L L M PRIMARY Hot leg riser Primary natural circulation flow/bulk flow M L L M PRIMARY Hot leg riser lnterfacial drag/relative motion of phases L M M M PRIMARY Vertical/radial natural circulation L L L L Hot leg riser PRIMARY Inlet flow/temp distribution L L L M Hot lee riser PRIMARY Hot leg riser Convection heat transfer to shroud/riser L L L H PRIMARY Stored energy release/conduction of Hot leg riser shroud/riser L L L H PRIMARY Hot leg riser Radiation heat transfer from shroud/riser L L L H PRIMARY Control rod drives/supports structures affect Hot leg riser on flow L L L H PRIMARY Riser Bypass Flow L L L H Hot lee riser PRIMARY Hot leg riser Mixing L L L H RVV System Component Process/Phenomena BLD LTC KL AP PRIMARY Upper plenum Flashing H L L H PRIMARY Upper plenum Two-phase level swell L L L M PRIMARY Upper plenum Two-phase pressure drop L L L M PRIMARY Upper plenum Single-phase pressure drop L L L H PRIMARY Upper plenum Primary natural circulation flow/bulk flow M L L M PRIMARY Upper plenum lnterfacial drag/relative motion of phases L L L M PRIMARY Upper plenum Vertical/radial natural circulation L L L M PRIMARY Upper plenum Convection heat transfer to reactor vessel L L L H PRIMARY Stored energy release/conduction of vessel Upper plenum wall L L L H PRIMARY Radiation heat transfer from reactor vessel to Upper plenum containment vessel L L L H RVV System Component Process/Phenomena BLD LTC KL AP PRIMARY PRESSURIZER Flashing H M L H 29

PRIMARY PRESSURIZER Phase separation M M L M PRIMARY PRESSURIZER Flooding at baffle plate L H L M PRIMARY PRESSURIZER Two-phase pressure drop L M L M PRIMARY PRESSURIZER Single-phase pressure drop L L L H PRIMARY PRESSURIZER lnterfacial drag/relative motion of phases M M L M PRIMARY PRESSURIZER Primary natural circulation flow/bulk flow L L L H PRIMARY PRESSURIZER Convection heat transfer to vessel L L L H RVV System Component Process/Phenomena BLD LTC KL AP PRIMARY Reactor vent valves Two-phase pressure drop I M L M PRIMARY Reactor vent valves Single-phase pressure drop I H M M PRIMARY Reactor vent valves Choked flow I H L M RVV System Component Process/Phenomena BLD LTC KL AP PRIMARY Flashing M L L H SGANNULUS PRIMARY SGANNULUS Two-phase level swell L L L M PRIMARY SGANNULUS Two-phase pressure drop M L L M PRIMARY SGANNULUS Single-phase pressure drop L L L H PRIMARY SGANNULUS Primary natural circulation flow/ bulk flow M L L M PRIMARY SGANNULUS lnterfacial drag/relative motion of phases L L L M PRIMARY SGANNULUS Vertical/radial natural circulation L L L L PRIMARY SGANNULUS Convection heat transfer to SG tubes M L L M PRIMARY SGANNULUS Convection heat transfer to vessel L L L H PRIMARY SGANNULUS Convection heat transfer to riser L L L H PRIMARY SGANNULUS Stored energy in the steam generator tubes &

fluid L L L M PRIMARY SGANNULUS Stored energy release/conduction of vessel wall L L L H PRIMARY SGANNULUS Stored energy release/conduction of riser L L L H PRIMARY SGANNULUS Radiation heat transfer from vessel L L L H PRIMARY SGANNULUS Radiation heat transfer from shroud/riser L L L H PRIMARY SGANNULUS Tube bypass flow M L L M PRIMARY SGANNULUS Feed header effect on flow M L L M PRIMARY SGANNULUS Feed header stored energy M L L M PRIMARY SGANNULUS Steam header effect on flow M L L M PRIMARY SGANNULUS Steam header stored energy M L L M RVV System Component Process/Phenomena BLD LTC KL AP PRIMARY DOWNCOMER Flashing M L L H PRIMARY DOWNCOMER Two-phase level swell M M L M PRIMARY DOWNCOMER Two-phase pressure drop L L L M PRIMARY DOWNCOMER Single-phase pressure drop L L L H PRIMARY DOWNCOMER Primary natural circulation flow/ bulk flow M L L M PRIMARY DOWNCOMER lnterfacial drag/relative motion of phases M M L M PRIMARY DOWNCOMER Vertical/radial natural circulation L L L L PRIMARY DOWNCOMER Convection heat transfer to vessel L L L H PRIMARY DOWNCOMER Convection heat transfer to shroud/riser L L L H PRIMARY Stored energy release/conduction of vessel DOWNCOMER wall L M M H 30

PRIMARY DOWNCOMER Stored energy release/conduction of riser L L L H PRIMARY DOWNCOMER Radiation heat transfer from vessel L L L H PRIMARY DOWNCOMER Radiation heat transfer from shroud/riser L L L H RW System Component Process/Phenomena BLD LTC KL AP PRIMARY Reactor Recirc.

Valves Single-phase pressure drop H L M H PRIMARY Reactor Recirc.

Valves Choked flow H H L M PRIMARY Reactor Recirc.

Valves Two-phase pressure drop H H L M RW System Component Process/Phenomena BLD LTC KL AP PRIMARY Lower plenum Flashing M L L H PRIMARY Lower plenum Two-phase level swell L L L M PRIMARY Lower plenum Two-phase pressure drop L L L M PRIMARY Lower plenum Single-phase pressure drop M L L H PRIMARY Lower plenum Primary natural circulation flow/bulk flow M L L H PRIMARY Lower plenum lnterfacial drag/relative motion of phases L L L M PRIMARY Lower plenum Vertical/radial natural circulation L L L L PRIMARY Lower plenum Convection heat transfer to vessel L L L H PRIMARY Lower plenum Stored energy release/conduction of vessel L M M H RW System Component Process/Phenomena BLD LTC KL AP SECONDARY SG tubes Void distribution L L L L SECONDARY SG tubes Single-phase convection heat transfer M L L M SECONDARY SG tubes Two-phase convection heat transfer M L L M SECONDARY SG tubes Flashing M L L H SECONDARY SG tubes Stored energy of tubes L L L H RW System Component Process/Phenomena BLD LTC KL AP CHRS Containment vessel Condensation heat transfer H H H M Vertical/radial natural circulation flow (gas CHRS Containment vessel space) H H H M CHRS Containment vessel Single-phase convection heat transfer L L L H Conduction heat transfer (vessel and CHRS Containment vessel containment) M M M H CHRS Containment vessel Effect of noncondensables H H H L CHRS Containment vessel Fluid properties for pressures< 0.10132 MPa M I I H CHRS Containment vessel Thermal stratification L L M L Vertical/radial natural circulation flow (liquid CHRS Containment vessel space) L L M L CHRS Containment vessel lnterfacial heat / mass transfer (pool to gas) L L M L RW System Component Process/Phenomena BLD LTC KL AP Containment CHRS cooling pool Nucleate boiling L L M H Containment CHRS cooling pool Single-phase convection heat transfer L L M H Containment CHRS cooling pool Vertical/radial natural circulation L L M L Containment CHRS cooling pool Thermal stratification L L M L 31

4.6 Phenomena / Knowledge Level Issues Of particular importance are the processes and phenomena that are considered important, yet are also considered poorly understood. This combination can lead to difficulties in analysis of the transient and/or high uncertainty in the predicted results. This section summarizes the PIRT rankings.

4.6.1 Blowdown The panel considered 120 processes and phenomena during the Blowdown period for the NuScale design. Of this 120, 11 (9%) were considered of high ranking importance. Only two were considered of high or medium importance, but with low knowledge level. These were CHF (correlations) in the core and the effect of non-condensables in the containment vessel. The lack of knowledge for CHF in the core is due to the relatively poor understanding of CHF in part-length rod bundles with low velocity coolant. The panel did not expect this to become a major safety issue because the likelihood of exceeding CHF seemed to be remote and conservative estimates of CHF would probably be sufficient. However, the M knowledge level is an acknowledgement of a spare database at conditions of interest.

During blowdown, the major concern may be that of the effect of non-condensables on condensation heat transfer in the containment vessel. While the NuScale design intends to eliminate this as a factor, the panel was concerned with the effect of trace amounts on the FOMs. Because of the difficulty in determining condensation rates with the presence of non-condensables, the panel identified this as a process with high importance but low knowledge level.

4.6.2 RVV Depressurization For the RVV Depressurization period, 125 processes and phenomena were considered. Of these 10 (8%) were considered of high importance. Of these, only two were considered to be of high or medium importance, but with low knowledge level. These were the effect of non-condensables on containment condensation, and the effect of stored energy heat release in the lower plenum. The effect of non-condensables is due to its effect on the containment pressure FOM in this period. The effect of stored energy release from the lower plenum was due to its impact on the CHF FOM and the difficulty in determining the heat transfer from a complex geometry.

4.6.3 Long-Term Cooling During the Long-Term Cooling period, only 4 (3%) of 122 active processes were considered to be of high importance. However 6 processes were identified as having high or medium importance with a low knowledge level. The effect of non-condensables on containment condensation rate was found to be important with low knowledge level. There were 5 processes in the CHRS with medium importance and low knowledge level. These were thermal stratification and internal natural circulation (in both the containment liquid and the cooling pool),

and interfacial heat transfer between the containment liquid and the gas space. As these processes are associated with the ultimate heat sink for the NuScale design, the panel concluded that this is an area of possible high uncertainty.

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5 NEUTRONICS PIRT 5.1 PIRT Panel Members The panel consisted of experts with an understanding of computational neutronics and thermal-hydraulics, and reactor and plant systems and their response to various upsets. They met for one day to carry out the PIRT. Several of the panel members had also participated in the thermal-hydraulic PIRT. The panel members and their affiliations were:

• Stephen Bajorek (Facilitator), NRC/RES

• David Diamond (Panel Chair), Brookhaven National Laboratory

• Istvan Frankl, NRC/RES

• Peter Griffith, Consultant

• Nathanael Hudson, NRC/RES

• Peter Yarsky, NRC/RES Other NRC staff, from the Office of New Reactors (NRO) were also present to provide their support when needed. In addition, a telecom with Kent Welter from NuScale Power was held to help the panel understand how reactor startup would take place and to answer other general questions about the design.

5.2 Scenario The focus of this PIRT was on an event in which the neutronics was important. The design-basis reactivity accident for PWRs has usually been the control rod ejection accident. However, in the specific case of NuScale it is expected that the frequency of occurrence of this event will be so small that it will no longer be a design-basis accident. Hence, the scenario considered by the panel was an inadvertent control rod bank withdrawal from hot zero power (HZP) condition at any time during the fuel cycle (expected to be a multi-batch, two-year cycle). This scenario would result in a power excursion mitigated first by Doppler reactivity feedback, and then by reactor trip. It was assumed that the reactor would be tripped by a high power trip setpoint (e.g.,

118% of nominal power), rather than by any low power or period trips that might be present, in order to make the power excursion as severe as possible for the purposes of the PIRT.

The initial conditions at HZP for a NuScale-type reactor are expected to be achieved by Decay heat would be determined by how long the fuel had been operating and by taking into account at least one day of shutdown before startup. The pressure of the vessel at HZP would be at its nominal value for normal operation as the result of using pressurizer heaters. Coolant temperature would be determined by those heaters and any decay heat that might be present.

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5.3 Figures of Merit It was decided that multiple figures of merit (FOMs) would be used to cover as many phenomena as possible. The FOMs were chosen by taking into account the acceptance criteria for this event. Hence, the FOMs were:

• Departure-from-nucleate-boiling ratio (DNBR)

• Reactor vessel pressure

• Fuel centerline temperature

• Pellet-clad interaction (PCI) / Pellet-clad mechanical interaction (PCMI)

For this event, the acceptance criteria are that DNBR remain above the 95/95 (i.e. 95%

probability at 95% confidence level) DNBR limit everywhere in the core, peak (pressurizer) pressure must be maintained below 110% of the design value, and the fuel centerline temperature must remain below the melting point.

The difference between PCI and PCMI is subtle, and it is sometimes difficult to differentiate the two types of failure. PCI is generally caused by stress-corrosion cracking due to fission product (iodine) embrittlement of the cladding, while PCMI is primarily a stress-driven failure. Generally, transient safety analysis considers a one percent uniform (elastic and inelastic) strain acceptance criterion to preclude PCI/PCMI failure.

Gaseous release and fuel thermal expansion are responsible for cladding strains at moderate to high exposures. Therefore, PCI or PCMI analyses of cladding strain for transients and accidents should take into consideration: fuel thermal expansion, gaseous fuel swelling models, and irradiated cladding properties 5.4 Results and Discussion 5.4.1 Phenomena The results of the PIRT panels deliberations are shown in Table 5.1. The list of phenomena was developed by the panel taking into account the important neutronic and fuel thermo-mechanical effects during this event as well as thermal-hydraulic phenomena. The thermal-hydraulic phenomena had been found to be relevant to the PIRT for a LOCA event. Note, however, that many fewer phenomena need to be taken into account in the neutronic PIRT relative to the thermal-hydraulic PIRT.

The phenomena are listed in the table according to the system or component in which they are found and according to what general behavior they affect. For this PIRT the three systems/components being considered are the core, reactor coolant system, and instrumentation and control system. Most of the phenomena occur in the core, as the bank withdrawal scenario is primarily a localized event. The focus groups are primarily in the core and include neutron kinetics (recognizing that some phenomena affect initial conditions rather than dynamic effects), fuel thermo-mechanical behavior, and subchannel thermal-hydraulics.

The pressurizer is also present as pressurizer pressure is one of the FOMs. Lastly, instrumentation and controls are part of the reactor protection system and have a direct effect on the neutronic response to the bank withdrawal scenario.

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Because this was a limited PIRT, there was no attempt to define in writing all of the phenomena listed, a practice which is commonly a part of a PIRT. It was assumed that the reader would understand what was meant by the phenomenon. However, there were several instances where clarification is useful and is given below.

• Shutdown bank speed is listed as a phenomenon; it is not simply something determined by design (as for example assumed for the control bank being withdrawn). The speed is primarily expected to be determined by a Technical Specification that specifies that banks must be able to be inserted a specified distance in a specified time. However, the speed beyond that point may be determined by bowing of guide tubes and this may be quite uncertain. It is, therefore, considered a phenomenon which must be considered and possibly taken into account in some way in the modeling of the event.

• Moderator density and temperature feedback are separated because they are accounted for differently in modeling the neutronics. The former refers to the effect of changes in density due to a change in temperature, while the latter refers to the change in scattering properties as a result of changes in moderator temperature.

• Fuel temperature feedback generally takes into account the Doppler effect and the change in scattering by oxygen within the pin. However, it generally does not also take into account the expansion/contraction of the pin with increasing/decreasing temperature.

• Assembly interaction and axial/radial reflector representation are two phenomena that are related. The small size of the integral PWR core necessitates consideration of neutronic phenomena that are generally of low or moderate importance for large commercial PWR cores. In particular, the small size of the reactor and high leakage fraction increases the importance of the axial and radial reflector, and in particular, the treatment of the reflector region.

Generally, PWR cores are analyzed using a two-step method whereby detailed transport calculations are performed for fuel assembly lattices to calculate homogenized, few group, nuclear cross-sections to be used in downstream nodal codes. The lattice calculations are generally performed presuming that the lattice may be depleted in an infinite array of like lattices to generate the exposure dependent nuclear parameters.

However, for high leakage cores, the inter-assembly coupling and coupling to the reflector may be of sufficient importance that this approach may not be fully applicable.

For example, for assemblies near the core periphery, which is a large fraction of the total number of fuel assemblies, the influence of the spectral thermalization of the return current in the reflector is likely to introduce strong gradients in the sub-assembly spectrum for the edge assemblies. Under such conditions, use of the standard two-step approach without modification may result in substantial bias in the calculated power distribution and kinetic behavior.

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• Pin-to-pin burnup distribution and gap conductance and fuel conductivity are related to each other.

There is interplay between the important phenomena during the transient such that it would be difficult to demonstrate that in all cases any one particular phenomenon could be treated conservatively. In particular, one can consider the influence of the average heat resistance of the fuel. The heat resistance here refers to a combination of the fuel pellet conductivity and gap conductance (both of these parameters are a function of the fuel pin exposure). The FOMs include the peak fuel centerline temperature as well as DNBR. Consider a bias in the heat resistance such that the event is analyzed with a higher, average fuel heat resistance. In this case, the transient increase in cladding heat flux is dampened by the higher heat resistance and this arguably will result in the prediction of a greater margin to DNBR. However, the higher heat resistance will result in a greater propensity to hold heat up in the fuel, which will result in a more rapid fuel temperature increase, and thus a more rapid arrest of the power excursion by Doppler worth. This may or may not be conservative when then considered against the resultant increase in fuel temperature, which would reduce predicted margin to the fuel centerline melt FOM. Given this interplay, those parameters affecting the accurate prediction of the average fuel heat resistance were considered highly important.

The average fuel heat resistance, however, is a function of the fuel rod exposures (pin by-pin burnup distribution) and the instantaneous power distribution. The power distribution apportions the power across a variety of different heat resistances and thus influences the dynamics of transient cladding heat flux, average fuel temperature increase, and the overall transient event progression.

• Time in fuel cycle refers to the fact that feedback and bank worth change with burnup. It is necessary to do the analysis with the most conservative initial conditions.

• The pellet burnup distribution refers to the radial distribution of actinides and fission products and any burnable poison within the pellet. The fact that fission power peaks at the surface of the pellet has an effect on this distribution.

• Stored energy is a phenomenon that determines the initial fuel temperature.

• Cladding conductivity includes the effect of any oxide layer.

• Core pressure drop and loop pressure drop are important for calculating the flow rate through the core.

Table 5-1 PIRT Results for NuScale Neutronic Event System/ Focus Process / Phenomena Importance Knowledge Component Level CORE Neutron kinetics Withdrawn control rod bank worth H H CORE Neutron kinetics Shutdown bank worth H H 36

CORE Neutron kinetics Shutdown bank speed H M CORE Neutron kinetics Moderator density feedback H H CORE Neutron kinetics Moderator temperature feedback M H CORE Neutron kinetics Fuel temperature feedback H H CORE Neutron kinetics Assembly interaction (e.g., through ADFs) H M CORE Neutron kinetics Soluble boron reactivity feedback H H CORE Neutron kinetics Initial nodal power distribution H H CORE Neutron kinetics Transient nodal power distribution H H CORE Neutron kinetics Axial/radial reflector representation H M CORE Neutron kinetics Pin to pin power distribution H H CORE Neutron kinetics Time in fuel cycle H H CORE Neutron kinetics Pellet burnup distribution H M CORE Neutron kinetics Delayed neutrons H H CORE Neutron kinetics Decay heat M H CORE Neutron kinetics Xe/Sm Concentrations L H CORE Neutron kinetics Pellet/structure/coolant direct energy M H CORE Fuel Thermo- Pellet radial power distribution M M CORE Fuel Thermo- Core pin-by-pin burnup distribution H M CORE Fuel Thermo- Stored energy L H CORE Fuel Thermo- Fuel heat capacity H H CORE Fuel Thermo- Gap Conductance H M CORE Fuel Thermo- Fuel conductivity and density H M CORE Fuel Thermo- Cladding conductivity L H CORE Fuel Thermo- Cladding strain H H CORE Subchannel Subcooled boiling M H CORE Subchannel Natural Circulation L H CORE Subchannel Convection H H CORE Subchannel Core pressure drop M H CORE Subchannel CHF (correlations) H M RCS Loop pressure drop M H RCS Pressurizer Water expansion H H RCS Steam Generator Tubes Heat transfer L M I&C Reactor Protection Detector Response M L Abbreviations: RCS - reactor coolant system; ADF - assembly discontinuity factor; I&C - instrumentation and control; CHF - critical heat flux 5.4.2 Ranking The results show that 22 of the phenomena are of high importance (having significant or dominant influence on one or more FOMs), 8 are of medium importance, and only 5 are of low importance (having only a small influence on the FOMs). There are no phenomena that are of high importance for which the knowledge level is low (an H,L ranking). A low knowledge level 37

ranking indicates that the phenomenon is not well understood and that modeling the phenomenon is currently either not possible or is possible only with large uncertainty. If there had been highly important phenomena with low knowledge level, these would have the highest priority with respect to assuring that computer codes could model them properly.

The next level of interest is with two other scoring combinations. The phenomena at this level have either a high importance with a medium knowledge level (H, M) or a medium importance with low knowledge level (M,L). They are listed below with their ranking (importance, knowledge level) and an explanation of each with respect to its importance and knowledge level follows.

• Shutdown bank speed (H,M)

• Assembly Interaction (H,M)

• Axial/radial reflector representation (H,M)

• Pellet burnup distribution (H,M)

• Core pin-by-pin burnup distribution (H,M)

• Gap conductance (H,M)

• Fuel conductivity and density (H,M)

• CHF correlations (H,M)

• Detector response (M,L)

The shutdown bank speed, along with its differential worth, determines the shutdown reactivity as a function of time and is directly responsible for terminating the bank withdrawal scenario.

Hence, bank speed is ranked as highly important. As mentioned above, the speed is for the most part expected to be determined by a Technical Specification requirement that specifies that banks must be able to be inserted a specified distance in a specified time. However, the speed beyond that point may be determined by bowing of guide tubes. This may be quite uncertain and is therefore ranked with medium knowledge level.

Assembly interaction is always an important consideration in modeling and becomes more uncertain when, as in the NuScale design, the core is small. Hence, it is ranked of high importance. The small reactor size introduces strong spatial and spectral flux gradients, resulting in a higher degree of uncertainty in the analysis of such reactors. Assembly interaction is usually accounted for with a synthesis approach in which the assembly is represented without any interaction with neighbors in order to obtain homogenized properties to represent the assembly within a core representation. The isolated representation also provides assembly discontinuity factors used to partially correct for the fact that the assembly is not part of an infinite array of identical assemblies. This synthesis approach adds to the uncertainty, and the phenomenon is ranked with a medium knowledge level.

The core of this reactor is expected to have only 37 fuel assemblies, with appearance similar to 17x17 PWR assemblies currently in use except for a reduced height. The reduced radial and axial dimensions mean that the leakage in this core will be high relative to that expected in a typical large PWR. Hence, axial and radial reflector representation will be highly important. For example, significant spectral gradients may be present between the core and reflector, and 38

given a large leakage fraction, the influence of this spectral gradient on other significant parameters (e.g. power distribution) should be accounted for in the overall analysis method.

Standard approaches which generate nuclear parameters for assembly lattices may not fully capture the significant spectral interaction between the core and reflector. Therefore, the panel ranked the knowledge level here as only medium. Although current modeling of reflectors for large PWRs has been satisfactory, there is little experience to fully understand how successful these models will be when leakage is more significant The pellet burnup distribution determines neutronic properties of the fuel and the power distribution within the pellet and is ranked as highly important. The fuel management may either be for a single 24-month fuel cycle or for multi-batch reloads. If gadolinia is used in the fuel, the burnup distribution becomes even more important because self-shielding will cause the gadolinium to burn preferentially at the surface of the fuel rod. There are challenges in modeling the burnup distribution and indeed, some analysis in the past has ignored the effect altogether. These challenges reduce the knowledge level to a medium ranking.

The core pin-by-pin burnup distribution is a factor in determining the thermo-mechanical properties of the fuel. Hence, it is important in determining fuel behavior and the way in which heat gets from the fuel to the coolant and is ranked as highly important. Frequently, only assembly-average burnup is used to determine properties and this introduces uncertainty into the analysis. In particular the dynamic cladding heat flux is sensitive to the combination of the radial power distribution at the sub-assembly level as well as the distribution of heat resistance, which is determined according to pin-by-pin burnup. Pin-by-pin burnup, and the associated influence on fuel thermo-mechanical properties (such as pellet conductivity and gap conductance) can be obtained via an online flux reconstruction methodology with a core depletion calculation. Nevertheless, the pin-by-pin burnup distribution is still considered to have only a medium knowledge level.

Gap conductance and fuel conductivity and density determine heat transfer to the coolant.

Since the bank withdrawal event may lead to a power-cooling mismatch, the heat flux at the clad surface is a key parameter and these phenomena are ranked as highly important. The fuel thermal properties are uncertain because of uncertainties in the burnup of a given fuel rod and because of uncertainties in the properties as a function of burnup and hence, the knowledge level is considered medium.

Although DNBR was identified as a FOM, the onset of DNB is not expected as any safety analysis would need to demonstrate margin to the onset. The calculated margin is based on the use of CHF correlations. CHF correlations are directly related to one of the FOMs and, are therefore, ranked as highly important. They are somewhat uncertain for short fuel assemblies as the measurements that are available have been done with geometries typical of large PWRs.

Hence, the ranking of knowledge level is medium.

Detector response has been ranked as having medium importance. The signal for reactor trip is expected to come from neutron detectors which may be placed within the containment. The detector response to changes in core power and delays between signal receipt and the movement of control banks is uncertain because of a lack of design information. Hence, 39

although this phenomenon has been ranked as having a low knowledge level, this ranking is expected to be easily upgraded when more information is available from the vendor.

6.0

SUMMARY

AND CONCLUSIONS This report documents two PIRTs for the NuScale-like design that is still preliminary. The PIRT panels made use of information that was available as April 2011 and based the rankings on that information. Because the design is preliminary and is expected to change before submittal for Design Certification, it is important that the PIRTs in this report also be viewed as preliminary. A revision, based on final design information is vital to the development and review of thermal-hydraulic and neutronic codes for licensing decisions.

6.1 Thermal-hydraulic PIRT The accident scenario considered for the thermal-hydraulic PIRT is the inadvertent opening of a reactor recirculation valve. An inadvertent control rod bank withdrawal from hot zero power (HZP) was the scenario considered for the neutronics PIRT. Both PIRTs identify phenomena of high importance, those that must be modeled accurately or bounded in an appropriate manner in a safety analysis to adequately assure margin to regulatory limits.

For the thermal-hydraulic PIRT, the figures-of-merit (FOM) were the Critical Heat Flux (CHF),

mixture level in the vessel, and containment vessel pressure. The panel in general, did not believe it likely that limits on the departure from nucleate boiling ration (DNBR) would be challenged except possibly in the early part of the transient if the core flow were to stagnate while at or near full power operation. The mixture level was selected as a FOM because the safety systems are designed to prevent core uncovery and heatup. In the NuScale design, a conventional containment is replaced by a relatively small high pressure containment vessel.

During blowdown and early in the transient, high pressure within the containment vessel is a safety concern and therefore included as a FOM.

Of particular interest are those phenomena that are highly ranked in importance, but are poorly understood as indicated by a low knowledge level. For the NuScale-like design, the major concerns appear to be with the Containment Heat Removal System (CHRS), which is where there were several phenomena in which the panel identified as having high importance and less than satisfactory knowledge level. Phenomena in this category include condensation heat transfer (both with and without noncondensables), and mixing in the containment as it affects the liquid pool and the gas space. There were additional concerns with choked flow from the vessel through the reactor vent valves and the reactor recirculation valves. These are mainly with the prediction with flow to the break and choked flow from the primary.

6.2 Neutronic PIRT The neutronic PIRT considered a scenario whereby a control bank was inadvertently continuously withdrawn from hot zero power conditions. This would be expected to result in a power increase that would be offset initially by Doppler feedback and then by reactor trip.

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Multiple figures-of-merit (FOMs) were used to cover as many phenomena as possible. The FOMs were chosen by taking into account what would be the acceptance criteria for this event.

Hence, the FOMs were:

• Departure-from-nucleate-boiling ratio (DNBR)

• Reactor vessel pressure

• Fuel centerline temperature

• Pellet-clad interaction (PCI) / Pellet-clad mechanical interaction (PCMI)

There were only 35 phenomena that the panel listed as having special significance in modeling this event. Most of these relate to either the neutronic initial conditions or neutron kinetics. In addition, there were phenomena that relate to fuel thermo-mechanical properties and subchannel thermal-hydraulics.

The results show that 22 of the phenomena are of high (H) importance (having significant or dominant influence on one or more FOMs), 8 are of medium (M) importance, and only 5 are of low (L) importance (having only a small influence on the FOMs). There are no phenomena that are of high importance for which the knowledge level is low (an H,L ranking). A low knowledge level ranking indicates that the phenomenon is not well understood and that modeling the phenomenon is currently either not possible or is possible only with large uncertainty. If there had been highly important phenomena with low knowledge level, these would have the highest priority with respect to assuring that computer codes could model them properly.

The next level of interest is with two other scoring combinations. The phenomena at this level have either a high importance with a medium knowledge level (H,M) or a medium importance with low knowledge level (M,L). They are listed below with their ranking (importance, knowledge level), and a detailed explanation of each phenomenon with respect to its importance and knowledge level is found in Section 5.

• Shutdown bank speed (H,M)

• Assembly Interaction (H,M)

• Axial/radial reflector representation (H,M)

• Pellet burnup distribution (H,M)

• Core pin-by-pin burnup distribution (H,M)

• Gap conductance (H,M)

• Fuel conductivity and density (H,M)

• CHF correlations (H,M)

• Detector response (M,L)

The deficiencies in knowledge level for these phenomena are summarized as follows:

• Shutdown bank speed, which partially defines the termination of the scenario, is determined, in addition to Technical Specifications, by bowing of guide tubes and this bowing is difficult to predict.

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• Assembly interaction is always an important consideration in modeling and becomes more uncertain when, as in the NuScale design, the core is small.

• Axial and radial reflector representation is more important in the NuScale design than in large PWRs and the neutronics methods that are suitable for the large PWRs are not expected to be sufficient for the iPWRs.

• Pellet burnup distribution is important in determining the neutronic properties of the fuel and, because of self-shielding effects which are difficult to model, has a significant uncertainty.

• Core pin-by-pin burnup distribution is a factor in determining the thermo-mechanical properties of the fuel but although power reconstruction methods exist, modeling is still uncertain.

• Gap conductance and fuel conductivity and density determine heat transfer to the coolant, but these properties may be uncertain due to uncertainties in the burnup of a given fuel rod and because of uncertainties in the properties as a function of burnup.

CHF correlations are directly related to one of the FOMs, but they are somewhat uncertain for short fuel assemblies as measurements that are available have been done with geometries typical of large PWRs. Detector response to changes in core power and delays between signal receipt and the movement of control banks is uncertain because of a lack of design information rather than because the phenomenon is difficult to model.

7.0 REFERENCES

1. NuScale Preliminary Loss-of-Coolant Accident (LOCA) Thermal-Hydraulic and Neutronics Phenomena Identification and Ranking Table, NuScale PIRT Licensing Topical Report, NuScale Power, Inc. NR-TR0610-289, June 2010.

2. Vasavada, S., Zavisca, M., and Khatib-Rahbar, M., Modeling and Analysis of Selected Loss of Cooling Accidents in an Integrated Light Water Reactor Using the TRACE Computer Code, ERI/NRC 210-203, Energy Research, Inc., Rockville, MD, May 2010.

3. Diamond, David, NuScale Reactor Overview, Presentation to NRC Staff, Rockville, MD, August 11, 2010.

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