NOC-AE-12002784, Summary of the South Texas Project Risk-Informed Approach to Resolve Generic Safety Issue (GSI)-191

Summary of the South Texas Project Risk-Informed Approach to Resolve Generic Safety Issue (GSI)-191

ML12023A056
Person / Time
Site:	South Texas
Issue date:	01/11/2012
From:	Crenshaw J W South Texas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
GSI-191, NOC-AE-12002784, STI: 33215081
Download: ML12023A056 (20)

Text

Nuclear Operating Company South TTas Pmect Electnc Generatin$

Staton PO. Bar 289 Wadsworth.

Txas 77483 , January 11,2012 NOC-AE- 12002784 STI: 33215081 File: G25 U. S. Nuclear Regulatory Commission Attention:

Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MID 20852-2738 South Texas Project Units 1 and 2 Docket Nos. STN 50-498, STN 50-499 Summary of the South Texas Project Risk-Informed Approach to Resolve Generic Safety Issue (GSI)-191 A summary of South Texas Project's methodology towards a risk-informed resolution of GSI-191, "Assessment of Debris Accumulation on PWR Sump Performance" is provided in the Enclosure.

The summary includes the initial quantification and project results to date.There are no regulatory commitments in this letter.Should you have any questions regarding this letter, please contact either Jamie Paul, Licensing, (361) 972-7344, or me at (361) 972-7074.J.W. Crenshaw VP New Plant Deployment

/ Special Projects JLP

Enclosure:

Summary of GSI-191 Risk-Informed Closure Pilot Project 2011: Initial Quantification NOC-AE- 12002784 Page 2 cc: (paper copy)(electronic copy)Regional Administrator, Region IV U. S. Nuclear Regulatory Commission 612 East Lamar Blvd, Suite 400 Arlington, Texas 76011-4125 Balwant K. Singal Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North (MS 8B13)11555 Rockville Pike Rockville, MD 20852 Senior Resident Inspector U. S. Nuclear Regulatory Commission P. O. Box 289, Mail Code: MN116 Wadsworth, TX 77483 C. M. Canady City of Austin Electric Utility Department 721 Barton Springs Road Austin, TX 78704 Stewart Bailey Branch Chief, Safety Issues Resolution U. S. Nuclear Regulatory Commission One White Flint North (MS 011 F01)11555 Rockville Pike Rockville, MD 20852 Donnie Harrison Branch Chief, PRA U. S. Nuclear Regulatory Commission One White Flint North (MS 011 F01)11555 Rockville Pike Rockville, MD 20852 A. H. Gutterman, Esquire Morgan, Lewis & Bockius, LLP John Ragan Chris O'Hara Jim von Suskil NRG South Texas LP Kevin Polio Richard Pena City Public Service Peter Nemeth Crain Caton & James, P.C.C. Mele City of Austin Richard A. Ratliff Alice Rogers Texas Department of State Health Services Balwant K. Singal Stewart Bailey Donnie Harrison U. S. Nuclear Regulatory Commission Summary of GSI-191 Risk-Informed Closure Pilot Project 2011: Initial Quantification South Texas Project, Wadsworth, TX NOC-AE-12002784 January 11, 2012 Enclosure NOC-AE-12002784 STP Nuclear Operating Company's NRC/Industry.

GSI-191 Closure Pilot Project Executive Summary The main objective of the Risk-Informed GSI-191 Closure Project is, "Through a risk-informed approach, establish a technical basis that demonstrates that the current design is sufficient to gain NRC approval to close the safety issues related to GSI-191 by the end of 2013." The results presented in this summary are the joint work of STP Nuclear Operating Company (STPNOC) Risk Management, Los Alamos National Laboratory, The University of Texas at Austin, Texas A&M University, Alion Science and Technology, ABS Consulting, ScandPower, Mike Golay, The University of New Mexico, and KnF Consulting Services, LLC.In the risk-informed approach, STPNOC would seek exemption from certain requirements of 10 CFR 50.46 if the risk associated with the fibrous insulation in STPNOC's containment buildings is not risk significant.

If STPNOC determines this insulation to pose a significant risk, STPNOC is committed to investigating plant modifications including insulation removal and other measures to preserve sufficient margins for nuclear safety.The 2011 preliminary results show that the change in risk for fibrous insulation in contain-ment is less than 1.OE-06 in core damage frequency (CDF) and less than 1.OE-07 for large early release frequency (LERF), that is, very small per RG 1.174.This result represents the uncertainties of more than 20 input parameters and the complementary execution of the physics-based (CASA Grande) model, thermal-hydraulics model, and PRA models. Although previous realistic testing has shown that chemicals are unlikely to affect the head loss in STPNOC debris beds (sump strainers and fuel assemblies), the Pilot Project has used an initial methodology that adds pessimistic head-loss estimates from chemicals.

Including these estimates is believed to fully addresses NRC concerns raised in pre-licensing meetings related to sump chemistry.

This preliminary assessment of the CDF and LERF gives us confidence that the issues asso-ciated with fibrous insulation in the STPNOC reactor containment buildings will be further shown to be non-risk significant with adequate defense-in-depth and safety margins such that STPNOC will be able to provide a sufficient basis for GSI-191 closure by the end of calendar year 2013.STPNOC will expand upon the project's technical contributions by including the uncertainties of more than 50 input parameters and the seamless integration of CASA Grande, a new jet formation model, un-certainty propagation in the thermal-hydraulics models, and PRA analysis that takes into account detailed operational conditions.

The resulting framework will provide STPNOC the ability to assess future issues on risk-informed basis as they may arise.The methodologies and results of the pilot project in 2011 are presented in the following documents:

analysis of results from the physical process solver, CASA Grande and RELAP5 thermal-hydraulic analyses (Letellier, 2011); LOCA Frequency analysis (Fleming et al., 2011); Uncertainty quantification methodologies and illustrative examples (Popova and Galenko, 2011); Jet formation research (Schneider et al., 2011); and Chemical effects research and experimental design (Sande et al., 2011).i Enclosure NOC- AE- 12002784 STP Nuclear Operating Company's NRC/Industry CSI-191 Closure Pilot Project Contents 1 Introduction 1 1.1 Previous efforts ..........

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1 1.2 Risk-informed

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1 1.3 Pilot Project ..........

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2 2 Major assumptions 2 3 Findings in 2011 3 3.1 Chemical effects ...............

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4 3.2 Strainer and downstream performance

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5 3.3 Destruction zone ...............

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5 4 CASA Grande 7 5 Plant configuration 7 6 PRA 8 7 Uncertainty Quantification (UQ) 10 8 LOCA frequency 11 9 Quality Assurance 11 10 Licensing 11 11 Conclusions 13 List of Figures 1 Illustration of the core damage (CDF, ACDF) risk associated with as-designed fibrous insu-lation in the STPNOC containment on the Regulatory Guide 1.174 decision-making Region map ...........

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4 2 Illustration showing how decreasing decay heat removal flow requirement over time reduces cooling flow requirements that would result in significant reductions in head loss across debris beds. Experiments of in-core performance with chemical effects show significantly greater tolerance (higher mass loadings) for fully loaded debris beds at lower flow rates ...........

6 3 Illustration of a CASA Grande (complementary) cumulative distribution and the method used to develop input for the PRA from it ........................................

8 4 High level illustration of the inputs and outputs overlaid on a schematic diagram of the PRA event tree structure.

The event tree structure, with the exception of the long term cooling top event, already exists in the typical nuclear power plant PRA .....................

9 5 Uncertainty quantification for complex computer models .....................

10 6 Illustration of the major elements of the STPNOC quality assurance plan for risk-informed closure of GSI-191 ..........

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12 ii Enclosure NOC-AE-12002784 STP Nuclear Operating Company's NRC/Industr.

GS1-191 Closure Pilot Project 1 Introduction The purpose of this document is to summarize the initial quantification performed for the risk-informed closure path in STPNOC's NRC/Industry risk-informed GSI-191 pilot project (Pilot Project).In this document, significant findings from Pilot Project work in calendar year 2011 are summarized and the overall program approach is described.

While the risk-informed approach is not new to investiga-tion of GSI-191 closure, the approach in the Pilot Project is more comprehensive than previously con-templated.

The closure process developed is gener-ically applicable with site-specific information sup-plied. Each primary area of investigation in the Pilot Project is summarized in this document.1.1 Previous efforts Since 2001 GSI-191: Assessment of Debris Accumu-lation on PWR Sump Performance has eluded reso-lution despite significant efforts by industry and the NRC. Although recent thought has been given to risk quantification (see for example Teolis et al., 2009) and early recognition of the need for risk evaluation was identified, (e.g. Darby et al., 2000), serious investiga-tion into risk quantification has not been undertaken until now. Instead, resolution has followed a classical deterministic approach.

It is STPNOC's view, follow-ing an initial 2011 PRA quantification, that a risk-informed resolution path should be pursued in prefer-ence to a deterministic approach, thereby quantifying the safety margins and identifying any scenarios that pose significant risk in GSI-191.The primary issue here is that in a deterministic approach, examination of the uncertainty required to evaluate risk of plant operation is abandoned and an appeal to a hypothetical "worst case" scenario (as-sumed to encompass all uncertainty) is made. By as-suming a sufficient set of nonphysical processes along with assumed equipment performance scenarios, ex-amination of the full range of possible scenarios is avoided. Although this approach can be effective in some cases, there are shortcomings associated with it (NEI, 2009) and, relative to GSI-191 closure, have become untenable.

That is, the "worst case" is so limiting that it would indicate no design would be ac-ceptable, especially for in-vessel effects (Baier, 2011).Even the presence of tramp debris (regardless of the insulation design) may not be tolerated.

Unless com-promised by incorporating the time-evolution of acci-dent phenomena and accommodation of realistic be-havior, it may be difficult to provide a satisfactory closure path using a deterministic approach.1.2 Risk-informed In keeping with the agency's commitment to move to-wards increasing risk informed regulation, the NRC directed the staff (Vietti-Cook, 2010) to consider GSI-191 resolution strategies including an Option 2: "The staff should take the time needed to consider all options to a risk-informed, safety conscious resolution to GSI-191.While they have not fully resolved this is-sue, the measures taken thus far in response to the sump-clogging issue have contributed greatly to the safety of U.S. nuclear power plants. Given the vastly enlarged advanced strainers installed, compensatory measures already taken, and the low probability of challenging pipe breaks, adequate defense-in-depth is currently being maintained.

The operative words for Option 2 are in-novation and creativity.

The staff should fully explore the policy and technical impli-cations of all available alternatives for risk informing the path forward. These alterna-tives include, but are not limited to, how 50.-46a might impact this issue, and how the application of a no-transition-break-size ap-proach might work." It is worth reemphasizing that a risk-informed anal-ysis includes all scales of postulated accidents and the full spectrum of possible plant responses.

Ideally, there should be no exclusive focus on "bounding" as-sumptions, no exclusive focus on "design basis" or"beyond design basis" assumptions, and no exclusive focus on "best estimate" assumptions.

Every factor in the accident analysis should be described as accu-rately as possible by a statistical distribution that is 1 of 15 Enclosure NOC-AE-12002784 STP Nuclear Operating Company's NRC/Industry GSI-191 Closure Pilot Project consistent with available data, physical models, or ex-pert opinion, and so, factors with limited or no data necessarily have bigger uncertainties.

Furthermore, the methods used to sample and propagate these un-certainties should be unbiased.In our initial quantification numerous conser-vatisms are retained to maintain consistency with regulatory assumptions while developing a robust toolset and exploring parameter sensitivities.

1.3 Pilot Project STPNOC is the plant working with the staff to de-velop risk-informed closure strategies while preparing a site-specific licensing submittal.

Over the 2011 cal-endar year, several public meetings were conducted to inform the NRC staff of the modeling approach and to solicit feedback on the applicability and use of the approach for resolving GSI-191. The several meetings included supporting material so that mem-bers of the public, especially other plants could be informed as well: Rosenburg (2011), Thadani (2011), Singal (2011c,a,b,d,e,f,g).

All materials provided have been posted by the NRC on their website for public access.In design of the Pilot Project, STPNOC has been careful to make the proposed implementation at other plants as straightforward as possible.

In particular, the physical models that typically would be folded into the event trees and fault trees of the PRA have been extracted into a flexible modeling tool and held outside the PRA. In the Pilot Project this part of the toolset is called CASA Grande (Letellier, 2011).This way, the methodology is made more generic and only simple changes should be required in the site-specific PRA. The existing PRA LLOCA, MLOCA, and SLOCA branches are maintained intact and only two relatively simple changes are required in the PRA itself: the sump strainer demand failure likelihood is replaced by a value that CASA Grande produces as an output; and a top event must be added for long term cooling relative to in-vessel effects.The rest of this report summarizes the GSI-191 risk-informed closure process and methods that have been developed over the past eight months. Several assumptions were adopted in the initial quantifica-tion in order to ensure the Pilot Project met 2011 schedule commitments.

Section 2 provides a sum-mary of the major assumptions made in the initial quantification.

Section 3 summarizes some of the sig-nificant insights obtained from the Pilot Project work so far. Section 4 describes the several physical mod-els that have been developed in 2011, others will be further refined and developed in the second project calendar year (2012). A new integration framework has been developed and is described as well in Sec-tion 4. A containment CAD model and toolset used in the project are briefly described in Section 5. Sec-tion 6 is a short description of the PRA modeling required to obtain the change in risk needed for Reg-ulatory Guide 1.174 decision-making.

In Section 7 the methodology developed in calendar year 2011 for uncertainty quantification is described.

The status of the Pilot Project LOCA frequency analysis is pre-sented Section 8. An overview of the Quality As-surance Plan is provided in Section 9. The licensing approach is designed around Regulatory Guide 1.174 (Section 10). Conclusions are in Section 11.2 Major assumptions Several assumptions had to be adopted in calendar year 2011 to accomplish a quantification result within the aggressive Pilot Project schedule.

In some cases, the assumptions bias the resulting CDF and LERF higher, in some cases the assumptions bias the results lower. The Pilot Project has tried to adopt a policy regarding assumptions such that on balance, the CDF and LERF results would be biased higher than with all assumptions relaxed.The following list summarizes of the major assump-tions in calendar year 2011.ZOI: The zone of influence is assumed to be 17D for all fiberglass targets regardless of configuration or the presence of piping restraints.

ZOI is not truncated by the presence of compartment walls.Chemical precipitants:

All chemical precipitants (formed based on the available material in con-tainment) are formed and introduced in bulk at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the LOCA regardless of the 2 of 15 Enclosure NOC-AE-12002784 STP Nuclear Operating Company's NRC/Industry GSI-191 Closure Pilot Project operation of containment spray. In-vessel pre-cipitants are assumed to be of the same form and amount used by Bajer (2011). The head loss due to precipitants on the strainer debris bed doubles the head loss through the bed. All chemicals are passed through the strainer if it is not entirely covered with a 1/8 inch-thick uni-form debris bed.Debris generation:

60% of the debris is fines, 40%is large. 1% of large debris is eroded to fines with 100% spray exposure.Transport of fine debris: 100% go to the sump pool, 4% are retained on the strainer bed (per strainer), 13% go to dead volumne (other sumps, elevator shaft, dead volumes inside the bioshield wall).Transport of large debris: 1% is eroded to fine debris, 99% retained on gratings.Upstream effects: Upstream effects are assumed to have insignificant contribution.

Strainer configuration:

The strainer meets design at the start of the LOCA transient (open bypass paths' contribution from damage for example, is assumed negligible).

Scenario event timing: The scenario progression and associated event timing are assumed to follow nominal values in each LOCA category (small, medium, large LOCA). This also applies to Operator actions with the exception of con-tainment spray operation.

In the case of contain-ment spray, the Operator is successful in stop-ping the single train (when three trains have started) as required but fails to terminate con-tainment spray operation (the remaining two pump trains) after meeting termination criteria.Containment spray is never terminated when less than three trains have started.LOCA frequency:

The existing STPNOC PRA LOCA branch frequencies for small, medium, and large LOCA are preserved in the calcula-tion of location-specific frequency calculations (to avoid over-estimating or under-estimating the LOCA frequency in the source term). That is, in each category of LOCA, the location-specific values for that category closely sum to its PRA initiating event frequency.

For the pur-pose of in-vessel hydraulic effects on core flow and head loss calculations, the LOCA frequency is split evenly between hot and cold leg locations.

Qualified coatings:

At time zero (at the time of the LOCA), 33 lb of Epoxy and 553 lb of IOZ coating particulate is assumed to arrive in bulk to the sump pool. This amount is present in all break scenarios (regardless of break size).Unqualified coatings:

At 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 247 lb of Alkyd, 843 lb of IOZ, 268 lb of enamel, and 294 lb of epoxy coating particulate arrives in the sump pool.Latent debris: 170 lb of particulate and 12.5 ft 3 of fiber arrive at the sump pool at the start of the LOCA (time 0).Strainer bypass: Strainer bypass is a function of face velocity and the number of operating trains only (Zigler, 2011). All debris bypassed is de-posited in the core and evenly spread over the fuel assemblies.

Strainer head loss: Head loss follows the behavior described by Zigler et al. (1995). The strainers do not collapse from overload due to differential pressure.3 Findings in 2011 In calendar year 2011, a primary objective was to obtain an initial PRA quantification of the risk as-sociated with fibrous insulation in containment.

The results of the initial quantification were included in the project plan as a criteria for going forward or terminating the project and taking aggressive correc-tive action to correct any high risk scenarios revealed in the risk assessment.

A major finding of the 2011 3 of 15 Enclosure NOC-AE- 12002784 STP Nuclear Operating Company's NRC,/Industr'y GSI-191 Closure Pilot Project 1 Illustration of 2011 "Initial Quantification" on RG 1.174 Regions*...........

..Region I" No changes allowed Region II.Small changes LL* Track cumulative impacts Region III< -.Very small changes.More flexibility with respect to baseline CDF.Track cumulative impacts t-5 W&.i:: :... .*.". U Region II 10-6 Fibirou's aisuiation risk at Sip (conservative)'.Region III 10-6 10-5 10 4 CDF Figure 1: Illustration of the core damage (CDF, ACDF) risk associated with as-designed fibrous insulation in the STPNOC containment on the Regulatory Guide 1.174 decision-making Region map.quantification is the risk is "very small" in the lan-guage of Regulatory Guide 1.174 as illustrated in Fig-ure 1. Regulatory Guide 1.174 is concerned with the need to make changes guided by risk.The risk of ongoing operation is maintained in the site-specific PRA average risk model for CDF and LERF. The STPNOC PRA meets the ASME/ANS PRA Standard as Capability Category II and has successfully provided the technical basis for several risk-informed applications at STPNOC, for example RMTS (Yilmaz et al., 2009; EPRI, 2008). A change to the plant may increase risk (CDF and/or LERF)and is expressed as ACDF and ALERF. Clearly, if a change results in a large change in risk, it could affect the plant average risk. But if the ACDF is less than 1.0X 10-6 and the ALERF is less than 1.0x 10', the change in risk is considered to be very small and the cumulative risk to CDF and LERF do not need to be considered in decision-making.

On the other hand, the effect on plant average risk must be considered if the changes are not very small. In this case, based on the "very small" risk evaluation, STPNOC man-agement has authorized the project to go forward as communicated in the calendar year 2011 Pilot Project plan.3.1 Chemical effects Based on early prototypical experiments performed by Dallman et al. (2006) and (much later) based on studies by Sande et al. (2011), STPNOC has main-tained that in the initial quantification, chemical ef-fects could be minimized.

STPNOC understood from the start of the project that any assertions about chemical effects would require experimental verifi-cation. However, based on the feedback from the NRC staff during the Pilot Project meetings, (Sin-gal, 2011g), chemical effects were added to the initial quantification.

The approach was modified to include the results found by Baier (2011) in experiments in-tended for deterministic evaluations.

4 of 15 Enclosure NOC-AE- 12002784 STP Nuclear Operating Company's NRC/Industrgy GSI-191 Closure Pilot Project With the inclusion of Baier's chemical effects us-ing mass of fiber as the success criteria, the ACDF increases from extremely low to roughly 5 x 10'.The ALERF remains unchanged.

Based on this re-sult and based on earlier NRC staff Pilot Project feedback, an important finding is that chemical ef-fects on debris bed head loss needs to be better un-derstood.

STPNOC has accelerated and expanded the risk-informed chemical effects experimental pro-gram in order to better understand chemical effects on head losses in ECCS strainer and fuel assembly debris beds. The chemical effects experiments will also be designed to produce results that can be used to develop correlations for realistic physical models of chemical performance for use in CASA Grande.3.2 Strainer and downstream perfor-mance As mentioned earlier, the deterministic approach uses a hypothetical "worst case" scenario assumed to en-compass all uncertainty assuming nonphysical mod-els and scenarios.

Although it does create a bound-ing case in some sense, the deterministic approach has assumed chemicals as precipitants are produced effectively at the start of the LOCA and during a hy-pothetical equipment alignment that produces max-imum ECCS flow rate through the strainer.

None of these conditions would ever exist simultaneously in the unlikely event of a LOCA requiring recirculation switchover with fibrous debris present.Debris beds in the core and on the strainers in-crease resistance to flow as the bed builds up thick-ness, but more importantly (based on experiment re-sults for deterministic evaluations), the resistance in-creases more when chemical precipitants collect on them also. Generally, head loss is higher order (nor-mally second order) with flow in a closed channel sys-tem.Zigler et al. (1995) utilized porous media correla-tions where head loss, H, equals to H = aQ + bQ 2 , where Q is the flow, and the coefficients a and b are determined by the fluid and debris bed conditions.

Referring to Figure 2, one can imagine how dra-matically the head loss would increase at constant flow rate as the coefficients a and b increase (caus-ing the system characteristic to steepen) due to fiber loading and chemical precipitant addition.

In fact, NRC staff has stated in Pilot Project review meetings (Singal, 2011e,f) this behavior has been observed dur-ing experiments as chemical precipitants were added at constant flow and the head loss increased signif-icantly (Baier, 2011). However, and again as illus-trated in the figure, this behavior might not be re-alized in an actual scenario because in the unlikely event of a LOCA that produces an in-vessel debris bed, the required flow decreases dramatically (lin-early with decay heat which is effectively an expo-nential function).

Additionally, chemical precipitants simply can't be created while the required flows are high (say within the first day to one week).A significant result of the risk-informed analysis that takes into account the evolution of the physi-cal processes is that: there are no ECCS failures due to strainer blockage (this needs to be examined for some infrequent potential sequences that, for exam-ple may cause mechanical collapse of the strainer);

all hypothesized small LOCA events succeed; almost all medium LOCA events succeed; and most large LOCA events succeed. All scenarios resulting in core damage were associated with in-vessel effects, not strainer blockage.

That is, in a realistic setting that retains significant conservative assumptions, chemi-cal effects turn out to be less important than previ-ously thought. For example, a 15 grn/FA acceptance criteria would cause failure for essentially all cases based on the latent debris alone. But in a realis-tic setting, it was demonstrated (again with conser-vatisms) that most hypothesized LOCA events would not cause core blockage.3.3 Destruction zone A somewhat surprising result from the initial quan-tification is the insensitivity to the "zone of influence" or ZOI for use in for example, ANSI (1988). A great deal of effort in experimentation and study by the in-dustry has been devoted to understanding if the ZOI could be reduced to much lower volumes of damage than the widely accepted deterministic value defined by a sphere of 17D, where D is the break diameter.In fact Schneider et al. (2011) have been studying jet 5 of 15 Enclosure NOC-AE-12002784 STP Nuclear Operating Company's NRC/Industry GSI-191 Closure Pilot Project Earlier, when debris/chemical luads ae small, and the cooling flow requirement is high, the system characteristic is flatter: lower head losses at higher flows.Later, increasing debrisichemical loads steepens system characteristic:

however, the cooling flow requirement decreases rapidly as well.500 400 Flow rate needed to cool the core (gpml o 0 700 600 Sao 400 300 200 100 0 4 g V_610 rr1 9 11 261 gpm~1'46 gpm 4 f 12 16 20 Time (hr)24 708 712 716 720 Figure 2: Illustration showing how decreasing decay heat removal flow requirement over time reduces cooling flow requirements that would result in significant reductions in head loss across debris beds. Experiments of in-core performance with chemical effects show significantly greater tolerance (higher mass loadings) for fully loaded debris beds at lower flow rates.formation for use in the risk-informed methodology.

The sense is that these destruction volumes are much too large. For example, a 31 inch pipe break would create a sphere of destruction about 44 feet in diam-eter. A commonly accepted practice is to limit the ZOI such that it can't extend beyond the concrete walls of containment compartments.

In this initial risk-informed quantification, the 17D destruction volume was assumed in all break loca-tions without restriction for compartment walls. A sensitivity study was conducted to see the effect of re-ducing the ZOI to about 1/2 17D. While some reduc-tion was observed, this change did not significantly affect the risk. This surprising result can be explained by understanding the scenarios leading to core dam-age. Because the STPNOC strainer design was mod-ified to install very large strainers compared to the original strainer design, risk for loss of ECCS NPSH is effectively eliminated.

However, the sump perfor-mance gain comes at a price for core damage risk, especially when deterministic-based chemical effects are used for success criteria.6 of 15 Enclosure NOC-AE- 12002784 STP Nuclear Operating Company's NRC/Industry GSI-191 Closure Pilot Project The amount of fiber arriving on the fuel is almost entirely governed by the strainer face velocity at the time of recirculation switchover and for a few min-utes following recirculation switchover.

Except for the smallest breaks, the fiber load is governed by the number of ECCS trains running and the train config-uration. That is, the core fiber load (which as men-tioned above is limiting) is (almost) independent of break size and ZOI.4 CASA Grande The risk analysis examines the full spectrum of LOCA accident conditions ranging from predomi-nant, but small accidents that are easily managed up to and including extreme, but unlikely accidents that challenge the design basis. The analysis framework, CASA Grande, ultimately develops cumulative distri-bution functions (cdf) for exceedance of a particular value used in success criteria (for example, fiber load-ing per fuel assembly) for each of the standard PRA initiating event LOCA categories, LLOCA, MLOCA, and SLOCA, and for each of the possible equipment configurations as analyzed in the PRA. Once de-veloped, the failure likelihood for each scenario can be determined knowing the associated value as illus-trated in Figure 3. Note that if a value coming from a bounding deterministic experiment is available, then below that amount, success is assured.For example, if the amount of fiber is the crite-ria value, and say 75 gm/FA is the amount of fiber from an LLOCA, two available ECCS trains, and that value falls below the intersection of the distribution on the abscissa, conceptually one would draw a verti-cal line from that value to the distribution curve and look up the split fraction value (the failure probabil-ity) on the ordinate.

It is important to understand that thousands of samples for different scenarios pro-duce the complementary distribution and it is pos-sible that the same likelihood would come from dif-ferent scenarios.

For example, a small break with a lot of fiber nearby may result in the same value (say fiber loading) as a larger break having less damage opportunity.

From CASA Grande, each scenario can be traced from where the break occurs to, for exam-ple, the strainer loading and the downstream (core FA) loading.The inputs to CASA Grande are a combination of conservative, yet realistic, treatments of accident phenomenology and decision criteria based on pre-cursors to possible system damage. This methodol-ogy enables risk-informed insights without compro-mising the traditional safety basis. In the initial risk quantification for calendar year 2011, the plant per-formance metrics of head loss across the recirculation strainers and fiber deposition per fuel assembly were used to assess the risk of flow blockage leading to core damage during recirculation scenarios.

Detailed de-scription of the models and methods that constitute CASA Grande can be found in Letellier (2011).5 Plant configuration The basis for the insulation source term and locations for all welds and plant components in containment is the current STPNOC plant design drawings and design configuration database.

Although other for-mats may be equally useful, the Pilot Project chose to exploit the availability of computer aided design tools (CAD) as the method to capture and integrate the spatial information required to accurately calcu-late the weld LOCA locations, debris quantity (het-erogeneously distributed insulation) and type of de-bris. Additionally, specialized CAD interface tools were developed that efficiently and reliably interro-gate the CAD model. The output of the tool set re-duces the spatial information contained in the CAD to a database accessible by CASA Grande such that the debris source term can be calculated accurately for all locations, break orientations, and break sizes.Although not credited in the initial quantifica-tion, compartment wall information is included in the database so that when wall truncation is imple-mented in 2012, the ZOI will be properly shaped and limited by their presence.

A configuration database is required because CASA Grande samples all welds at every location in each replication of the total calcula-tion. For each LOCA category and at each location, 10 to 15 samples, random in size and direction, are taken for the source term calculation.

Each sample 7 of 15 Enclosure NOC-AE-12002784 STP Nuclear Operating Company's NRC/lndustry GSI-191 Closure Pilot Project 1.0 0 0 If the threshold of concern value increases, the exceedance ea, valueiincreaseseorese(smaller the efcodnc likeihoo folow alng pcm lativ frato o ailurev fo xape At sufficiently high values the likelihood goes to 0.0------- ------ -- ---U.uI Performance measure (e.g., core fiber mass)Figure 3: Illustration of a CASA Grande (complementary) cumulative distribution and the method used to develop input for the PRA from it.always includes the largest possible break (double-ended guillotine, DEGB) at each location.

A full spherical ZOI shape is used for the DEGB sample.Otherwise, a full hemispherical ZOI shape is used.The configuration database is designed to support au-tomatic calculation of the debris source terms (thou-sands of calculations in each CASA Grande repeti-tion).6 PRA The typical PRA will have initiating events for LLOCA, MLOCA, and SLOCA. It is interesting to understand that before GSI-191 became an issue, each of the LOCA initiating event branches already contemplated a sump screen plugging event (most likely as in the STPNOC PRA) represented as a demand failure of the containment sump screens at the time of recirculation.

Additionally, in each of these branches, there are the preexisting scenarios for the different combinations of ECCS equipment de-pending on support system and component successes or failures.

The Pilot Project exploits this existing structure by replacing failure likelihoods based on the results of CASA Grande as described in Section 4 in-stead of the simplistic demand failure likelihood.

As described in Section 1.3, the physical mod-els have been extracted out of the PRA to enhance portability, simplify incorporation of different plant information and characteristics, and facilitate adop-tion of the methodology.

Another inherent complex-ity fundamental to typical PRA design is overcome by performing the uncertainty quantification in CASA Grande. In the Pilot Project, time-dependent and multivariate uncertainty distributions (described in Section 7) have been identified in the physical models 8 of 15 Enclosure NOC-AE- 12002784 STP Nuclear Operating Company's NRC/Industry GSI-191 Closure Pilot Project Distributions:

Scenariol Scenario2

...LLO L LLii MLOC L L~The recirculation failure event is actually a top that includes other failures such as valves failing to transfer.

However, all the failure scenarios include the sump blockage failure.Figure 4: High level illustration of the inputs and outputs overlaid on a schematic diagram of the PRA event tree structure.

The event tree structure, with the exception of the long term cooling top event, already exists in the typical nuclear power plant PRA.required to understand GSI-191. CASA Grande also provides the platform to propagate the time-dependent and multivariate uncertainties outside the PRA. In fact, this is a basic requirement of a risk-informed analysis because current commercial PRA methodology is not fully capable of propagating the required uncertainties.

Referring to Figure 4, the preexisting event tree structure and LOCA initiating event frequencies are not changed by the Pilot Project methodology.

The distributions from CASA Grande described in Sec-tion 4 are used in the recirculation switchover event to substitute specific scenario likelihoods for the (in-variant) simplified demand recirculation likelihood (a single basic event in the preexisting model). A new top event was required in the STPNOC PRA to ad-dress the possibility of core flow blockage (long term cooling).

This is the only structural change required in the PRA and is considered to be a relatively minor change.A calculation can be performed (at any particu-lar plant) by summing the current frequencies as-sociated with each LOCA initiating event that go through recirculation to success. With each of these initiating event frequencies in hand, one can multiply each by its associated failure likelihood from CASA Grande. The resulting frequency sum is the new suc-cess frequency and the difference between the origi-nal frequency (the unchanged frequencies) sum from the new frequency (multiplied by the CASA Grande 9 of 15 Enclosure NOC-AE-12002784 STP Nuclear Operating Company's NRC/Industry GSI-191 Closure Pilot Project A Figure 5: Uncertainty quantification for complex computer models likelihoods) sum is a close approximation to the new ACDF due to having fibrous insulation in contain-ment.7 Uncertainty (UQ)Quantification faced by the Pilot Project during the 2011 were the choice of probability distributions for the input pa-rameters of CASA Grande and how to sample from them.One of the inputs to CASA Grande is the LOCA frequency table, Fleming et al. (2011). The proba-bilities associated with LLOCA are extremely small and simple Monte Carlo sampling approach will al-most never produce that observation.

In order to have the LLOCA in all of the scenarios, nonuniform Latin Hypercube Sampling was applied, Helton and Davis (2003). This methodology generates observa-tions from the whole space (i.e. the LLOCA is al-ways included) and weights them by the correspond-The modeling and propagation of uncertainties for the GSI-191 project involves several steps. Figure 5 shows the collection of information flow, methodolo-gies and challenges for UQ of computer models that approximate a complex real system like the one being considered in this pilot project. The main challenges 10 of 15 Enclosure NOC-AE-12002784 STP Nuclear Operating Company 's NRC/Industry GSI- 191 Closure Pilot Project ing probabilities.

The Pilot Project will expand upon this approach in 2012 to be able to generate corre-lated samples and estimate quantiles using the data generated by CASA Grande.Another major task during the 2012 will be the ver-ification and validation of the methodology applied to solve the GSI-191. The verification part will make sure that all the code written is "bug-free" and repre-sents correctly the theory and methods implemented.

The validation part is challenging since it will have to be shown that the approximation produced is very close to the real system. Popova and Galenko (2011)provide a detailed description and illustrative exam-ples of the UQ methodology applied to the initial quantification.

8 LOCA frequency In order to explore all possible scenarios, the risk-informed analysis requires knowledge of the likeli-hood of a pressure boundary failure for all possible locations and possible sizes of failure. Per design, significant rupture failures of Class 1 piping in do-mestic light water reactors have not been observed.Obtaining the appropriate likelihoods where there is no evidence presents a challenging problem for the GSI-191 risk-informed closure investigator.

In the initial quantification, Fleming et al. (2011)performed a substantial study designed to build upon the established EPRI risk-informed In-Service Inspection program (EPRI, 1999). EPRI (1999)methodology was used as primary basis to develop the size and location-specific rupture frequencies for the quantification.

In the Pilot Project study, Flem-ing et al. (2011) showed the total frequency in the standard PRA methodology (SLOCA, MLOCA, and LLOCA) were preserved.

Although the over-all methodology is considered to be sound based on peer review (Mosleh, 2011), and reasonableness of the values obtained, NRC review feedback in the Pilot Project has resulted in further review of the approach.

In 2012 an alternative to the LOCA fre-quency methodology will be performed on a new basis to fully address NRC concerns.9 Quality Assurance A quality assurance plan has been developed in cal-endar year 2011. The plan includes regularly sched-uled (nominally weekly) technical review teleconfer-ences supplemented at critical product development steps with on-site review. The STPNOC PRA ana-lyst (Technical Team Lead) is responsible for review and verification of the PRA inputs developed.

How-ever, the STPNOC PRA analyst review is supple-mented by independent critical peer review intended to help disclose any overlooked technical gaps that would compromise results and also help ensure that the overall product is academically defensible even though it is developed for the industrial setting. In-dependent technical oversight was a part of the STP-NOC 2011 efforts and was used to further focus anal-ysis efforts.The overall quality assurance plan is illustrated in Figure 6 as a flow chart. Due to the diverse technol-ogy required to be implemented in the GSI-191 scope, the PRA inputs originate with products developed by experts in their respective field. The CASA Grande integrating framework uses the inputs to generate the two main inputs to the PRA, the sump demand fail-ure likelihood and the in-vessel cooling failure like-lihood (for each category of LOCA and all possible equipment configurations).

These elements are doc-umented by the vendor and the normal vendor doc-ument review process is followed to assure they are suitable for use as input to the PRA. The overall Pilot Project quality assurance methodology is expected to be similar to most utilities' processes for PRA activ-ities although the details are most likely different.

10 Licensing STPNOC will rigorously quantify the risk contribu-tion to core damage and large early release of recir-culation scenarios that are encompassed by GSI-191.The licensing strategy is based on looking at the dif-ference in risk between a hypothetical containment design that has no fibrous insulation and the existing STPNOC design. The expectation is that the hy-pothetical containment design would have lower risk 11 of 15 Enclosure NOC-AE- 12002784 STP Nuclear Operating Company's NRC/lIndustry GSI-191 Closure Pilot Project Responsibility:

Contracted service organization Process: Local quality program Input development CASA Framework I I I I I I Responsibility:

LANL Process: Local Quality Program I Ga 30 CL Responsibility STP ContracrTechnical Coordinator, Project Technical Lead ProcessaOTP Technical Document Review Process Procedure OPGPO4-ZA-0 108 Approved/Distributed Vendor Document Control Process'Internal review supplemented and supported by outside Peer Revuew Input to PRA PRA Quantification/Output PRA Application License Amendment Request Inputs to PRA Verified/Reviewed Responsibility:

ABS Consulting Process: IOCFR5O Appendix B Program I Responsibility STP Contract Technical Coordinator, ProjectTechnical Lead Process' STP PRA Assessoment Process Procedure OPcSP04-ZA-r)6O'rProbabiiisitc Risk Assessment Program'I Responsibility.

STP Licensing Engineer Process: 5fTr License Amendment Process Procedure to Licensing Basis Documents and Amend.ments to the Operating License-Figure 6: Illustration of the major elements of the STPNOC quality assurance plan for risk-informed closure of GSI-191.than the existing design. Clearly an ideal outcome would be the difference in risk between the two de-signs to be roughly zero.The difference between a comparison of the risk analysis to the Regulatory Guide 1.174, Region III limit of ACDF< 10-6 yr-1 and ALERF< 10-7 yr-1 will provide a basis for GSI-191 closure via exemp-tion to certain requirements in 10 CFR 50.46. The 12 of 15 Enclosure NOC-AE-12002784 STP Nuclear Operating Company's NRC/Industry CSI-191 Closure Pilot Project terms ACDF and ALERF in the context of this work mean the difference in CDF and LERF between a hy-pothetically perfect containment building having no fibrous insulation and the existing plant design. The purpose of the initial quantification in calendar year 2011 was to understand if the risk associated with fiber insulation in containment would exceed the Re-gion III limits.If the final analysis (calendar year 2013) continues to support recirculation failure as a Risk Region-Ill event, preventative measures (safety margin, defense in depth) will be identified to address contributing factors that carry the largest potential impacts in the analysis.

Regardless of the quantitative findings, STPNOC will use the risk analysis to prioritize spe-cific actions and determine the degree of remediation that may be required.

Thus, it is essential that all parties fully understand the theory, implementation methods and interpretation of a risk-informed deci-sion process. STPNOC intends to continue to com-municate regularly and openly with the NRC staff and industry in calendar years 2012 and 2013, and continually refine and enhance communication and communication tools as the Pilot Project evolves.11 Conclusions The calendar year 2011 initial quantification in the Pilot Project has demonstrated the viability of a risk-informed closure path to GSI-191. Even at the early stage, the usefulness of the risk-informed approach has been demonstrated in initial findings related to ZOI, chemical effects, time-dependency, and LOCA frequency.

In a short amount of time, a comprehen-sive understanding of the GSI-191 important physical phenomena has been accomplished and has been used by the Pilot Project to develop a new, robust, risk-informed framework for study and closure of GSI-191.The Pilot Project has worked closely and effec-tively with the NRC to incorporate feedback and in-form the NRC staff of progress and the technology developed.

In some cases, NRC feedback has redi-rected the efforts of the Pilot Project and acceler-ated the schedule (for example, chemical effects and LOCA frequency).

The interaction and feedback is appreciated by the Pilot Project due to the aggressive schedule and complexity of the issues.The Pilot Project is designed generically so that other plants can easily take advantage of it. The design provides for simple integration of site-specific models in a method that extracts complexities out of the PRA and integrates them in a flexible model-ing tool, CASA Grande. A straightforward licensing approach is proposed and generally accepted quality assurance methods can be applied.Even with conservative margins retained, the Pilot Project in the initial quantification has shown that the risk from fibrous insulation in containment at STPNOC is very small (in Region III) based on the decision-making criteria of Regulatory Guide 1.174.This helps assure both the NRC and STPNOC that defense against ECCS failures can be easily identified and reasonable measures can be undertaken to avoid excessive risk in a risk-informed closure path.References ANSI (1988). ANSI/ANS-58.2-1988:

Design Basis for Protection of Light Water Nuclear Power Plants Against the Effects of Postulated Pipe Rupture.Design Standard 58.2-1988, ANSI/ANS, Washing-ton, DC.Baier, S. L. (2011, Septmeber).

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Sande, T., K. Howe, and J. Leavitt (2011, October).Expected Impact of Chemical Effects on GSI-191 Risk-Informed Evaluation for South Texas Project.White Paper ALION-REP-STPEGS-8221-02, Re-vision 0, Jointly, Alion Science and Technology and Univiersity of New Mexico, Albuquerque, NM.Schneider, E., J. Day, and W. Gurecky (2011, De-cember). Simulation Modeling of Jet Formation Progress Report, August -December 2011. In-ternal Report Revision 0, University of Texas at Austin, Austin, TX.Singal, B. K. (2011a, .June). FORTHCOMING CON-FERENCE CALL WITH STP NUCLEAR OP-ERATING COMPANY (TAC NOS. ME5358 and ME5359). Memorandum.

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