ML14357A171

Request for Additional Information, Round 2, Request for Exemptions and Revised Pilot License Amendment Request for a Risk-Informed Approach to Resolve Generic Safety Issue (GSI)-191 (TAC Nos. MF2400 Through MF2409)

ML14357A171
Person / Time
Site:	South Texas
Issue date:	03/03/2015
From:	Lisa Regner Plant Licensing Branch IV
To:	Koehl D South Texas
Singal B
References
TAC MF2400, TAC MF2401, TAC MF2402, TAC MF2403, TAC MF2404, TAC MF2405, TAC MF2406, TAC MF2407, TAC MF2408, TAC MF2409
Download: ML14357A171 (35)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 3, 2015 Mr. Dennis L. Koehl President and CEO/CNO STP Nuclear Operating Company South Texas Project P.O. Box 289 VVadsworth,TX 77483

SUBJECT:

SOUTH TEXAS PROJECT, UNITS 1 AND 2 - REQUEST FOR ADDITIONAL INFORMATION RELATED TO REQUEST FOR EXEMPTIONS AND LICENSE AMENDMENT FOR USE OF A RISK-INFORMED APPROACH TO RESOLVE THE ISSUE OF POTENTIAL IMPACT OF DEBRIS BLOCKAGE ON EMERGENCY RECIRCULATION DURING DESIGN-BASIS ACCIDENTS AT PRESSURIZED-VVATER REACTORS (TAC NOS. MF2400 THROUGH MF2409)

Dear Mr. Koehl:

By letter dated June 19, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML131750250), as supplemented by letters dated October 3, October 31, November 13, November 21, and December 23, 2013 (two letters); and January 9, February 13, February 27, March 17, March 18, May 15 (two letters), May 22, June 25, and July 15, 2014 (ADAMS Accession Nos. ML13295A222, ML13323A673, ML13323A128, ML13338A165, ML14015A312, ML14015A311, ML14029A533, ML14052A053, ML14072A076, ML14086A383, ML14087A126, ML14149A353, ML14149A354, ML14149A434, ML14178A481, and ML14202A045, respectively), STP Nuclear Operating Company (STPNOC, the licensee) submitted exemption requests accompanied by a license amendment request for a risk-informed approach to resolve the issue of potential impact of debris blockage on emergency recirculation during design-basis accidents Generic Safety Issue (GSl)-191) for South Texas Project, Units 1 and 2. The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the information provided in your application and your response to NRC staff request for additional information (RAI) dated April 15, 2014 (ADAMS Accession No. ML14087A075) and determined that additional information, as described in the enclosure to this letter, is required to complete review of your application.

A draft copy of the enclosed RAI was provided to Mr. VVayne Harrison of your staff via e-mail on November 18, 2014. The RAls were discussed with your staff during public meetings held from December 1 to December 3, 2014, and February 4, 2015. It was agreed that STPNOC will provide response to the requested information by March 13, 2015.

D. Koehl If you have any questions, please contact me at 301-415-1906 or via e-mail at Lisa.Regner@nrc.gov.

Lisa M. Regner, Senior Project Manager Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-498 and 50-499

Enclosure:

Request for Additional Information cc w/encl: Distribution via Listserv

REQUEST FOR ADDITIONAL INFORMATION EXEMPTION REQUESTS AND LICENSE AMENDMENT REQUEST RISK-INFORMED APPROACH TO RESOLVE THE ISSUE OF POTENTIAL IMPACT OF DEBRIS BLOCKAGE ON EMERGENCY RECIRCULATION DURING DESIGN-BASIS ACCIDENTS AT PRESSURIZED-WATER REACTORS STP NUCLEAR OPERATING COMPANY SOUTH TEXAS PROJECT, UNITS 1AND2 DOCKET NOS. 50-498 AND 50-499 By letter dated June 19, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML131750250), as supplemented by letters dated October 3, October 31, November 13, November 21, and December 23, 2013 (two letters); and January 9, February 13, February 27, March 17, March 18, May 15 (two letters}, May 22, June 25, and July 15, 2014 (ADAMS Accession Nos. ML13295A222, ML13323A673, ML13323A128, ML13338A165, ML14015A312, ML14015A311, ML14029A533, ML14052A053, ML14072A076, ML14086A383, ML14087A126, ML14149A353, ML14149A354, ML14149A434, ML14178A481, and ML14202A045, respectively), STP Nuclear Operating Company (STPNOC, the licensee) submitted exemption requests accompanied by a license amendment request (LAR) for a risk-informed approach to resolve the issue of potential impact of debris blockage on emergency recirculation during design-basis accidents Generic Safety Issue (GSl)-191) for South Texas Project (STP), Units 1 and 2. The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the information provided in your application and your response to NRC staff request for additional information (RAI) dated April 15, 2014 (ADAMS Accession No. ML14087A075},

and determined that additional information, as described in the enclosure to this letter, is required to complete review of your application.

A draft copy of the enclosed RAI was provided to Mr. Wayne Harrison of your staff via e-mail on November 18, 2014. The RAls were discussed with your staff during public meetings held from December 1 to December 3, 2014, and February 4, 2015. It was agreed that STPNOC will provide response to the requested information by March 13, 2015. to the licensee's letter dated November 13, 2013, contained Enclosures 4-1, 4-2, and 4-3 (defined as Volumes 1, 2, and 3, respectively). Also, Enclosure 5 was defined as Volume 6.2. The RAls make multiple references to Volumes 1, 2, 3, and 6.2. Please note the following, as it relates to these references:

• "Volume 1" refers to Enclosure 4-1 of the LAR dated November 13, 2013, Project Summary

• "Volume 2" refers to Enclosure 4-2 of the LAR dated November 13, 2013, Probabilistic Risk Assessment (PRA)

Enclosure

• "Volume 3" refers to Enclosure 4-3 of the LAR dated November 13, 2013, Engineering (CASA [Containment Accident Stochastic Analysis] Grande)

Analysis

• "Volume 6.2" refers to Enclosure 5 of the LAR dated November 13, 2013, Responses to NRC Request for Supplemental Information on the 2013 Submittal REQUEST FOR ADDITIONAL INFORMATION (RAI)

Probabilistic Risk Assessment (PRA) Licensing Branch (APLA)

NOTE: The numbering system has been retained from the first set of RAls issued by letter dated April 15, 2014, for consistency. Additional RAls have been assigned new numbers.

Project Quality Assurance Regulatory Guide (RG) 1.174, Revision 2, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant Specific Changes to the Licensing Basis,"

May 2011 (ADAMS Accession No. ML100910006), Section 5, "Quality Assurance," provides the NRC staff's position on quality assurance (QA) requirements for risk-informed changes to the licensing basis. Specifically, this section contains several provisions that should be met when a licensee elects to use PRA information to enhance or modify activities affecting the safety-related functions of structures, systems, and components (SSCs). When referring to QA, the term "activities" is typically interpreted to mean designing, purchasing, fabricating, handling, shipping, storing, cleaning, erecting, installing, inspecting, testing, operating, maintaining, repairing, refueling, and modifying. Therefore, the proposed decision not to remove problematic insulation represents a modification to several activities affecting the safety-related functions of SSCs, namely the Emergency Core Cooling System (ECCS) and Containment Spray (CS) systems.

1. Please describe how "PRA information" that is used to justify not removing problematic insulation including but not limited to the PRA, CASA Grande, and supporting analyses meets the following provisions in RG 1.174, Section 5:

• Use personnel qualified for the analysis.

• Use procedures that ensure control of documentation, including revisions, and provide for independent review, verification, or checking of calculations and information used in the analyses.

• Provide(s) documentation and maintain(s) records in accordance with the guidelines Section 6 of RG 1.174.

• Use(s) procedures that ensure that appropriate attention and corrective actions are taken if assumptions, analyses, or information used in previous decisionmaking are changed (e.g., licensee voluntary action) or determined to be in error.

2. The LAR, Volume 1 describes some quality assurance activities that were implemented in support of the LAR but states that CASA Grande "is a proprietary MATLAB application, which was unavailable to the [quality] oversight team." Therefore, please

provide a brief summary of the software QA (SQA) program for CASA Grande and the anticipated date when the CASA Grande software will become compliant with that SQA program. Describe any standards and upon which the SQA is based.

3. Identify any QA programs that were employed for any traditional engineering analyses/calculations performed in support of the LAR and state whether these programs meet 10 CFR 50, Appendix B requirements.

4. Describe the QA program employed by each vendor or contractor that performed calculations or analyses used to support the LAR. Explain whether vendor QA programs were assessed by STPNOC for compliance with applicable QA requirements.

Treatment of Unanalyzed Plant Conditions

1. RG 1.200, Revision 2, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2 (ADAMS Accession No. ML090410014), Section 1.4, "PRA Development, Maintenance, and Upgrade," states that plant information used in the Probabilistic Risk Assessment (PRA)

(e.g., expected thermal-hydraulic plant response to different states of equipment) should be as realistic as possible. Verify that the conditional split fraction values (i.e., failure probabilities used by PRA) for sump failure and in-vessel failure are based on CASA Grande simulations that represent accurately plant conditions for each accident sequence relevant to the LAR, or justify that the chosen failure probabilities are upper bounds for any plant conditions that might occur for a given scenario. For example, for plant conditions where a simulation is impractical, unnecessary, or not performed for any other reason, a split fraction value of 1.0 should be assigned or a qualitative argument should be made to select an existing CASA Grande result as bounding. Based on information provided in the LAR Volumes 2 and 3, this approach is already employed for pump states. Each of the 64 pump states identified in the LAR were assigned conditional split fraction values for sump and in-vessel failure that were based on:

• CASA Grande simulations (pump states 1, 22, 9, 26, 43)

• Qualitative arguments as to why existing CASA Grande results are bounding (the 11 bounded states)

• Assigned a conditional core damage probability of 1.0 (48 other pump states)

A similar verification that assigned failure probabilities are realistic or bounding should be applied to all other unanalyzed plant conditions including but not limited to:

• Number of containment fan coolers not equal to 6

• Failure of containment isolation

• Failure of operators to secure one train of CS early

• Failure of operators to secure remaining trains of CS late

• Failure to switch to hot leg injection prior to securing CS trains

• Failure to swap to hot leg (HL) recirculation

• Failure of a running pump following a successful start

• Failure of one or more residual heat removal system heat exchangers Therefore, provide the results of a systematic review of all accident sequences containing a top event corresponding to one of the seven GSl-191 failure modes. For each sequence, provide one of the following:

a. Confirmation that the split fractions assigned to sump and in-vessel failure were derived from CASA Grande simulations that are consistent with the specific plant conditions associated with the sequence (i.e., availability of plant equipment, success/failure of operator actions, etc.).

b. Technical basis for concluding that the existing CASA Grande simulation provides results that are applicable or bounding

c. Confirmation that the conditional split fraction value for sump or in-vessel failure were set to 1.0 for non-analyzed cases.

Human Reliability Analysis NOTE: Round 2 RA/ question numbers begin with the next sequential number from the April 15, 2014, RA/ for this section.

7. RG 1.174, Sections 2.3.1 and 2.3.2 state that the scope and level of detail of the PRA model must be sufficient to model the impact of the proposed change. NRC letter dated April 15, 2014 (ADAMS Accession No. ML14087A075), includes a number of RAls related to the human reliability analysis (HRA) used in the risk assessment and STPNOC responses to RAls describe a number of human actions that are important during a loss-of-coolant-accident (LOCA). Please describe how the dependency among multiple human actions (both those in the "clean plant" and "debris" models) in the same sequence was assessed for the debris PRA model.

Key Assumptions/Key Sources of Uncertainty

1. RG 1.200 defines a "key" source of uncertainty as an issue where no consensus approach or model exists and where the choice of approach or model is known to have an effect on the risk profile (e.g., CDF [core damage frequency], LERF [large early release frequency), LlCDF [delta CDF], LlLERF [delta LERF]) 1 . RG 1.174 and NUREG-1855, Revision 1, "Guidance on the Treatment of Uncertainties with PRAs in Risk-Informed Decisionmaking," March 2013 (ADAMS Accession No. ML13093A346),

state that "consensus" refers to an approach or model that has a publically available published basis and has been peer reviewed and widely adopted by an appropriate stakeholder group. In addition, widely accepted PRA practices may be regarded as consensus models. Examples include the use of the constant probability of failure on demand model and the Poisson model for initiating events. Finally, models that the NRC 1

The NRC staff's position is that cases where a consensus model does exist, but the licensee chooses an alternate model also represent key sources of model uncertainty if they have an effect on the risk profile.

has utilized or accepted for the specific application in question can also be considered "consensus."

RG 1.200 defines a key assumption as one that is made in response to a key source of model uncertainty where a different reasonable alternative assumption would change the plant's risk profile.

RG 1.200 states that "for each application that calls upon this regulatory guide, the applicant identifies the key assumptions and approximations relevant to that application.

This will be used to identify sensitivity studies as input to the decision-making associated with the application."

Therefore, please provide a table or other structured response that lists key sources of uncertainty. For each key source of uncertainty, please identify the key assumption(s) that were made to address it and provide either a sensitivity study in terms of GDF, LERF, 6CDF, and 6LERF or use a qualitative discussion as to why a different reasonable alternative assumption would not cause the risk acceptance guidelines in RG 1.174 to be exceeded. This response should address:

a. L* approach for chemical effects

b. Head loss correlation

c. Success criteria for fuel blockage and boron precipitation (7.5 grams per fuel assembly (g/FA))

d. Fiber penetration model for sump strainer

e. The use of geometric, rather than arithmetic mean aggregated values from NUREG-1829, "Estimating Loss-of-Coolant Accident (LOCA) Frequencies Through the Elicitation Process," April 2008 (Volumes 1 and 2: ADAMS Accession Nos. ML082250436 and ML081060300)

f. The continuum break model (vs. double ended guillotine break (DEGB) only model)

g. The quantity and release rate of unqualified coatings The response should evaluate each of these areas one-at-a-time and should include an aggregate analysis that quantifies the integrated impact on GDF, LERF, 6CDF, and 6LERF from the sensitivity studies that were performed.

Validity of Assumption on Pump Configurations

1. In response to question 3, "Plant Configuration," of the April 15, 2014, RAI, STPNOC analyzed different pump configurations for Case 22 to verify Assumption 2b of Volume 3, which stated that a combination of pumps failing in the same train would result in a bounding failure probability compared to other combinations with the same number of each type of pump (i.e., high head, low head, and CS).

The results of this sensitivity showed that the assumption was false for in-vessel failure probabilities. Therefore, non-conservative failure probabilities were assigned to PRA model top events for certain scenarios. This approach may result in an underestimation of the risk of debris.

Failure of the selected pump configuration (Case 22) to uphold assumption 2b calls into question the combinations of the other cases used to simplify the risk assessment.

Therefore:

• Please determine whether assumption 2b provides realistic or bounding failure probabilities for each pump state that is assigned a non-unity failure probability.

• Please provide CDF, LERF, llCDF, and llLERF using realistic or bounding failure probabilities for all possible pump configuration.

CASA Grande to PRA Interface

7. The licensee's response to question 5 of the April 15, 2014, RAI, contains a figure showing that the smallest observed break size leading to debris-induced core damage was approximately 17 inches. This appears to conflict with the response to question 1, "Success Criteria," which stated that "the largest break size below which no failures related to either the sump or vessel performance were recorded during the CASA Grande runs was a DEGB in a 5.189 diameter (D) inch pipe." Please clarify these contradictory statements.

Fidelity between RELAP Simulations and CASA Grande

1. Volume 6.2 describes the RELAP simulations that were used to determine whether core cooling could be accomplished with partial or complete blockage. Page 123 states that "all the safety systems were assumed to be available throughout the transient."

Therefore, it would appear that Table 2.5.39, "Core Blockage Scenarios Summary,"

(Volume 6.2) would only apply to scenarios where all ECCS and CS pumps are available (i.e., Case 1). The response to question 1, "Success Criteria," of the April 15, 2014, RAI, states that "analyses performed in support of the LAR included consideration of a 6 inch hot leg break with only one train of ECCS available." [emphasis added]. Please clarify if this refers to an analysis performed subsequent to the LAR. Provide additional details on this or any other analyses that are used to justify applying the results of Table 2.5.39 to pump states other than Case 1. Include a description on the quality assurance of these analyses in relationship to question 1, "Success Criteria."

State-of-Knowledge Correlation

1. RG 1.174 Section 2.5.2 states that the state-of-knowledge correlation should be accounted for unless it can be shown to be unimportant. In question 5, "Uncertainty Analysis, of the April 15, 2014, RAI, the NRC staff requested the licensee to clarify why the state-of-knowledge correlation was not applied to the LOCA frequencies used by the PRA and CASA Grande. STP's response stated that " ... dependence of the PRA and

CASA Grande on different parameters of the LOCA break frequencies is sufficient so as not to warrant correlation between the PRA and CASA Grande."

This answer may not be inaccurate because the choice of LOCA frequency percentile affects both the absolute LOCA frequency (used by the PRA) and the shape of LOCA frequency versus break size curve (used by CASA Grande). Therefore, both the PRA and CASA Grande rely on the same underlying parameter and the state of knowledge correlation applies. This position was communicated to the licensee by the Advisory Committee on Reactor Safeguards (ACRS) during the meeting on September 3, 2014 (ADAMS Accession No. ML14266A510}, and by the NRC staff during the audit conducted from September 15-17, 2014. Please revise your analysis by correlating the LOCA frequencies used by the PRA and CASA Grande. Please also provide updated CDF, LERF, f1CDF, and f1LERF based on mean values resulting from the parametric uncertainty calculation that properly considers the correlation between the initiating event frequencies and the failure probabilities (sump and in-vessel) for debris-related events.

Selection of Johnson Parameters

1. In the response to question 4, "Uncertainty Analysis," of the April 15, 2014, RAI, STPNOC stated that it was not possible to approximate the NUREG-1829 mean frequencies with different selections of the parameter "A. NRC independent analyses indicate otherwise; for example, the following alternative fits yields means that are relatively close to those tabulated in Table 2.2.2 of the Volume 3 submittal. Please evaluate the sensitivity of the CDF and LERF on different selections of bounded Johnson distribution fits, such as the alternative fit in the table below.

Median Mean 95th 1

Size {in) 5 h {1/yr) {1/yr) {l/yr) {1/yr) y 8 I; A.

0.5 0.000068 0.00063 0.001853 0.0071 4.962538 0.671235 1.49E-05 1 1.625 5E-06 8.9E-05 0.000408 0.0016 4.551311 0.568039 8.48E-08 0.268427 2 3.69E-06 6.57E-05 0.000301 0.00118 4.594371 0.568322 5.61E-08 0.212914 3 2.lE-07 3.4E-06 1.59E-05 6.lE-05 6.024348 0.568377 2.31E-08 0.135431 6 6.3E-08 1.08E-06 5.16E-06 1.98E-05 6.194246 0.56465 4.59E-09 0.062491 7 1.46E-08 3.04E-07 1.67E-06 6.34E-06 6.529987 0.541377 1.4E-11 0.052616 14 4.lE-10 1.2E-08 1.94E-07 5.8E-07 6.142561 0.422624 1.69E-10 0.024278 31 3.5E-11 1.2E-09 3.21E-08 8.lE-08 6.207166 0.389148 1.77E-11 0.01

Mechanical and Civil Engineering Branch (EMCB)

NOTE: Round 2 RA/ question numbers begin with the next sequential number from the April 15, 2014, RA/ (Round 1 RA/) for the EMCB sections.

2. In a letter dated December 13, 2013, the licensee explains that the strainers were analyzed for two load cases. Case 1 corresponds to the maximum temperature and a low differential pressure which occurs early following a loss-of-coolant accident, while debris loading is low. Case 2 corresponds to a maximum differential pressure, which occurs later when debris loading is at a maximum, and corresponds to a lower temperature. In both cases the interaction ratios are maintained below 1. However, it is unclear to the U.S. Nuclear Regulatory Commission staff that these two load cases represent the most limiting loading conditions, and bound all other possible temperature and pressure combinations. There could be a case where the differential pressure due to debris loading increases at a faster rate than the yield stress of the material increases due to the temperature drop.

Explain the basis for concluding that Case 1 and Case 2 are the bounding post-accident loading conditions for the strainers (i.e., they bound all other pressure and temperature combinations).

Steam Generator Tube Integrity and Chemical Engineering Branch (ESGB)

NOTES:

a. Round 2 RA/ question numbers begin with the next sequential number from the April 15, 2014, RA/ (Round 1 RA/) for the ESGB sections.

b. Follow-up questions from the STPNOC responses to the Round 1 RA/ questions refer to the Round 1 RA/ number for the ESGB sections.

c. In all cases, "Enclosure 1" refers to Enclosure 1 to Attachment 5, "Quantification of Chemical Head Loss Epistemic Uncertainty; Basis for Incremental Chemical Head Loss Epistemic Uncertainty" contained in the licensee's letter dated July 15, 2014 (ADAMS Accession No. ML14202A045).

Chemical Effects

23. During the NRC staff audit in September 2014, representatives from STPNOC stated that the chemical effects evaluation model was being changed from the chemical "bump-up" factor multiplier discussed in the licensee's submittal dated November 13, 2013 (ADAMS Accession No. ML13323A190), to an alternate chemical model that uses an additive chemical head loss factor determined from the chemical loading term "L*." Assuming the bump-up factor approach is no longer being pursued, the NRC staff has reconsidered previous chemical effects related RAls and determined that the following April 15, 2014, RAI questions are no longer relevant to the new chemical model: 1a-d, 3*, 4, 5, 9, 17, and 18a-c. Please confirm that the staff's understanding is correct.

24. The NRC staff has reviewed the overview of an alternate chemical effects approach contained in Enclosure 1 to Attachment 5, "Quantification of Chemical Head Loss Epistemic Uncertainty; Basis for Incremental Chemical Head Loss Epistemic Uncertainty," contained in the licensee's letter dated July 15, 2014 (ADAMS Accession No. ML14202A045). This enclosure provides an overview of the alternate chemical effects method.

a. Please provide a detailed description of this chemical head loss model and its application to the STP plant-specific chemical effects analysis such that the NRC staff can perform a thorough review and evaluation.

b. As part of the detailed description and based on CASA Grande realizations, please provide a histogram showing chemical head loss (feet) on the x-axis and number of occurrences on the y-axis for the medium-break LOCA (MBLOCA) and large-break (LBLOCA) categories. Please ensure the bin selections allow the NRC staff to discriminate different outcomes that result in acceptable head loss.

c. Please discuss whether the chemical head loss determined from the L* method is independent of the debris bed or in some way correlated with the debris bed.

d. Please describe in detail how the new chemical model will account for uncertainties. Some examples of uncertainties include: variability in chemical head loss behavior (e.g., an approximate 40 percent difference in head loss resulting from a change to the precipitate addition sequence in Enclosure 1, Figure 9), variability in head loss across different debris beds for the same type and quantity of precipitate, differences in corrosion/leaching behavior between test materials and plant materials, variability in temperature or pH compared to testing, other post-LOCA conditions (e.g., radiological) not present during testing.

25. The NRC staff has several questions related to Figure 14 in the aforementioned Enclosure 1 to Attachment 5.

a. Given the head loss response to chemical precipitate addition shown earlier in Figures 1 and 2, it seems more appropriate to model head loss in a non-linear manner. Please discuss any plans to further develop the model.

b. The NRC staff is of the opinion that the 4th "Bahn" data point placement in this plot is not appropriate given that the test loop was shut down at this point since the test loop head loss limit had been reached. Please discuss a plausible range of head loss for this test had it not been stopped and how that would affect the chemical head loss correlation.

c. Without consideration of item (b), the NRC staff calculated a greater chemical head loss (CHL) value (approximately 0.7 feet) when scaling a 13 feet of water result to the STP strainer test conditions according to Equation 2. Please provide a copy of the calculation showing the scaled value is approximately 0.4 feet.

d. While the NRC staff agrees that comparison of chemical effects testing may provide insight, the relationship between flow and chemical head loss may be more complex than as shown by Equation 2. Please provide a basis for this scaling equation or discuss the limitations that may exist when extrapolating data over more than an order of magnitude in flow rates.

26. Figure 25 in Enclosure 1 to Attachment 5 of letter dated July 15, 2014, shows new aluminum release equations that appear to be based on experiments run for Southern Nuclear Operating Company.

a. Please provide a copy of the Reference 17 (CHLE-SNC-005 Bench Test) Report that contains this data so that the NRC staff may understand how these tests were performed.

b. Confirm that the orange line in Figure 25 represents the 1600 series tests.

c. The aluminum release model appears to be predicting the same data as in Figure 24, which was used to develop the model. Please clarify if any additional data was used to develop the model.

27. A limited release of aluminum during chemical effects testing is one of the key items STP is relying on for concluding STP has relatively minor chemical effects. In ESGB question 13.b. of the April 15, 2014, RAI, the NRC staff asked if the two parts of scaffolding had been tested to compare their aluminum release. The licensee's response provided scanning electron microscope images along with energy dispersive spectroscopy (EDS) and x-ray photoelectron spectrometry (XPS) results.

Given that the two parts of scaffolding were used in different test conditions, that they were visually observed to have different texture and appearance, and that the Table 1 elemental compositions indicate potentially significant differences in key elements (e.g.,

Al, 0, P), the NRC staff thinks it is important to verify that the corrosion behavior of the two scaffolding parts is similar. For example, one way to verify similitude would be to run a direct comparison of aluminum release in bench tests at higher post-LOCA temperatures to determine if the aluminum release was reasonably similar. Please provide a comparison of the corrosion behavior of the two parts of scaffolding.

28. Since multiple tests suggest aluminum corrosion will be inhibited by phosphate after a relatively short time into the post-LOCA ECCS mission time, understanding the corrosion behavior of aluminum at elevated temperatures becomes very important. Recent aluminum corrosion testing by another licensee (see ADAMS Accession No. ML141848509, Slide 18) showed that for their plant-specific conditions, significantly longer test durations at 195 degrees Fahrenheit (°F) did not release an equivalent quantity of aluminum as shorter time at higher temperatures. Please discuss the relevance of these results to the STP chemical effects approach for aluminum release at higher temperatures. Please include in that discussion the range of postulated plant-LOCA temperature profiles relative to the CHLE test MBLOCA and LBLOCA profiles and if any adjustments are needed to the aluminum release rates at temperatures greater than 185 °F.

29. In Section 2.1.1 ("Zinc Phosphate") of Enclosure 1 to Attachment 5 of the licensee's letter dated July 15, 2014, the discussion states the following:

When zinc corrosion materials were included in the STP risk-informed tests, head loss response was observed during the initial hour of testing; however, additional tests indicated that the head loss response to the zinc product was likely the result of initial dissolution of a surface layer and not from transport of a continuously generated zinc corrosion product (Zn3(P04)2*4 H20). Therefore, the initial zinc product release is treated as a particulate source and not considered a zinc chemical product.

Since Zn3(P04)2*4 H20 is unlikely to transport to the strainer and, given that Option 1 CHL is intended to produce conservative or overestimated CHL response to identified precipitate loads, Zn3(P04)2*4 H20 generation is ignored in the CHL correlation development.

Given the significant quantities of zinc present, the NRC staff finds it would be appropriate for a chemical model to account for zinc. Dissolution of galvanized steel or inorganic zinc coatings may occur at the lower pH before the trisodium phosphate (TSP) buffer fully dissolves to adjust the pH to an alkaline value. Dissolved zinc would then be available to react with the phosphate. In addition, some percentage of the galvanized surfaces could be susceptible to having zinc corrosion product knocked off by water falling from the pipe break, drains, etc. The NRC staff recognizes it may be appropriate to model zinc products separately from amorphous aluminum hydroxide type precipitates if warranted by the head loss response across a debris bed representative of a sump strainer bed. Please provide the quantity of zinc that is included in the "particulate source," how this amount of zinc affects head loss and if an additional zinc product should be included in the model.

30. Figure 21 in Enclosure 1 to Attachment 5 of the licensee's letter dated July 15, 2014, implies the WCAP-16530 release rate equations are being incorporated into CASA Grande which is not the case for aluminum. Please clarify which, if any, WCAP 16530-NP-A, "Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSl-191, March 2008 (ADAMS Accession No. ML081150379),

release equations will be used with the alternate chemical head loss model approach.

31. The chemical head loss is determined based on chemical precipitate loading per strainer (grams per meter square (g/m 2)). Please describe how the plant-specific incorporation of this model accounts for the greater chemical head loading for the cases where less than three trains operate following a LOCA.

32. It is unclear to the NRC staff how STPNOC's response to ESGB question 14.a. of the April 15, 2014, RAI, evaluated the radiation effects on precipitates. Explain how uncertainties from the radiation effects on precipitate formation are considered in the STP chemical effects analysis.

33. In the response to ESGB question 21 of the April 15, 2014, RAI, the mass of 24 pounds for a CRUD release following a LOCA is based upon the Electric Power Research Institute (EPRI) Boron-Induced Offset Anomaly estimates of fuel deposits that would

affect a CRUD induced power shift (CIPS). While this may be an adequate prediction for CIPS susceptibility, it does not assess the total available transient CRUD layer in the primary coolant system. The fuel surface area is approximately 30 percent of available reactor coolant system (RCS) surface with other surfaces such as piping and Steam Generator tubing making up most of the remaining surface areas.

The EPRI Pressurized-Water Reactor (PWR) Primary Water Chemistry Guidelines state, in part:

Core flow transients should be minimized to minimize particulate entrainment which will increase dose rates and particulate contamination levels in low flow regions. Wall shear, which is approximately proportional to the square of the coolant velocity, is the primary factor promoting particulate releases subsequent to shutdown. A smooth transition to one pump operation is considered appropriate to reduce shear and minimize particulate releases during the shutdown transient.

During a reactor trip following a LOCA there is no "smooth transition" with liquid and gaseous flow plus solids entrainment. Thermal, hydraulic and chemical transients are all present, simultaneously. One of the most significant chemical changes is the presence of both hydrogen and oxygen in the water flowing to the sump as well as being recirculated back through the reactor core. This uncontrolled chemistry condition leads to both reductive and oxidative processes occurring simultaneously leading to particulate formation.

The EPRI PWR Primary Water Chemistry Guidelines (Table 3-5 of Section 3.8) identifies analyses to be performed by Chemistry during a normal shutdown, including filterable and non-filterable: radioactive corrosion products, elemental nickel and iron. Therefore, the Chemistry department may have this information related to normal shutdowns and transient shutdowns.

Therefore, the NRC staff requests that the licensee determine if historical information is available concerning crud release from normal shutdowns and unplanned trips and to re-evaluate the crud release estimate based on any additional information, including release from all RCS sources during a LOCA.

34. Please clarify the difference between the fiber amounts shown in the Table 2 and Figure 3 in Enclosure 1 of the licensee's letter dated July 15, 2014.

Coatings

8. With respect to question 1 of the April 15, 2014, RAI, the response does not seem consistent with the current NRC staff position on debris characteristics for unqualified coatings. The testing you referenced is not applicable to unqualified coatings. This position is described in the NRC staff review guidance available at ADAMS Accession No. ML080230462. Please provide a revised analysis for the unqualified epoxy coatings in question.

9. The licensee's response in question 2 of the April 15, 2014, RAI, stated that a Zone of Influence (ZOI) of 4D (4 Diameter) was used for inorganic zinc coatings. This position is inconsistent with the current NRC staff position. Based on the latest test data available the ZOI for inorganic zinc coatings should be 1OD. A description of this position is available at ADAMS Accession No. ML100960495. Please provide a revised analysis for the ZOI of inorganic zinc coatings.

10. With respect to question 6 of the April 15, 2014, RAI, the reductions credited for debris generated by unqualified coatings in upper containment is inconsistent with the current NRC staff position. Both the treatment of failure percentages and failure timing are based on EPRI testing that the staff has previously issued positions on. Staff guidance found at ADAMS Accession No. ML080230462 describes the staff's position with respect to this testing. In addition the NRC staff concerns regarding the failure timing being based on filter data (as described in the original questions 6b and 6c) are not adequately addressed by your responses. Please provide a revised analysis for the unqualified coatings in upper containment.

Containment and Ventilation Branch (SCVB)

NOTE: Round 2 RA/ question numbers begin with the next sequential number from the April 15, 2014, RA/ for this section.

10.

Background:

The response to question 3.a of the April 15, 2014, RAI, does not appear to provide adequate justification for not revising the Updated Final Safety Analysis Report (UFSAR) description of the containment heat removal analysis. The response to question 3.c refers to a proposed UFSAR description of the risk assessment given in Enclosure 3, Attachment 2 of the licensee's letter dated November 13, 2013, which does not provide a revised licensing basis description of the containment heat removal analysis.

The licensee's response to question 4.a of the April 15, 2014, RAI, does not provide adequate justification for not revising the UFSAR description of the fission product removal analysis. The response to question 4c of the April 15, 2014, RAI, refers to a proposed UFSAR description of the risk assessment given in Enclosure 3, Attachment 2 of the licensee's letter dated November 13, 2013, which does not provide a revised licensing basis description of the revised fission product removal analysis.

Please refer to the following excerpt taken from the licensee's response to question 3.b of the April 15, 2014, RAI:

As described in the LAR, the proposed exemptions from General Design Criteria (GDC)-35, "Emergency Core Cooling", GDC-38, "Containment Heat Removal", and GDC-41, "Containment Atmosphere Cleanup" are for approval of a risk-informed approach for addressing GSl-191 and responding to Generic Letter (GL) 2004-02 for STP Units 1 and 2 as the pilot plants for other licensees pursuing a similar approach. As further described, STPNOC seeks NRC approval based on a determination that the risk informed approach and the risk associated with the postulated

failure mechanisms due to GSl-191 concerns meets the guidance, key principles for risk-informed decision making, and the acceptance guidelines in RG 1.174.

STP is not proposing to apply the risk-informed approach to revise the licensing basis for containment design described in the UFSAR. The proposed risk assessment evaluates a spectrum of Loss of Coolant Accident (LOCA) scenarios to quantify the amount of debris of various types that might be generated and transported to the emergency sumps, and how that debris might affect available NPSH [net positive suction head] for Emergency Core Cooling System (ECCS) and Containment Spray System (CSS) pumps taking suction from the sumps in the recirculation mode. It also evaluates potential transport of debris to the reactor core. It calculates failure probabilities that are fed to the STP PRA.

Concern: The staff agrees that the currently licensed design and configuration of the CSS and ECCS as described in the UFSAR will not be impacted by the risk-informed resolution to GSl-191 except for the change in the sump strainer design. However, the NRC staff is not in agreement that the UFSAR description of the licensing basis containment heat removal analysis. which uses CSS; the licensing basis containment fission product removal analysis. which also uses CSS; and the licensing basis 10 CFR 50.46 analysis. which uses ECCS, will not be impacted by the risk-informed resolution to GSl-191. For breaks that produce less or no debris, the licensing basis analysis should be based on the deterministic approach without taking exemption from GDCs 35, 38, and 41. For breaks that produce large amount of debris and without taking exemptions from the GDCs (for example exemption from assuming single failure) it is not possible to meet the acceptance criteria for peak cladding temperature and containment heat and fission product removal, the risk-informed approach may be used and exemption from the GDCs may be requested for these specific breaks only.

The NRC staff has developed the flow chart shown in Figure 1 (on page 19 of this RAI) for defining the LOCA containment NPSH licensing basis analysis (which is the most significant part of containment heat removal analysis) for deterministic and risk-based GSl-191 resolution. The staff suggests the licensee to develop similar flow charts defining the deterministic and risk-based fission product removal and ECCS licensing basis analysis.

Question: RG 1.174 requires that the licensee should identify those aspects of the plant's licensing basis that may be affected by the proposed change, including but not limited to rules and regulations, UFSAR, technical specifications, licensing conditions, and licensing commitments. NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," (SRP) Chapter 19.2, "Review of Risk Information Used to Support Permanent Plant-specific Changes to the Licensing Basis:

General Guidance," Section 111.1 also requires that the changes in the plant licensing basis should be appropriately reflected in licensing documents such as technical specification (TSs), license conditions (LCs), and UFSAR. Therefore, the current licensing basis for the containment heat removal described in UFSAR Chapters 6 and 15

must be revised by including the description for the breaks for which partial or complete exemption from GDCs 35, 38, and 41 is requested.

a. Provide UFSAR revisions of Chapters 6 and 15 for the description of revised licensing basis analysis of the containment heat removal for the breaks for which exemption from GDC-38 is requested.

b. Provide UFSAR revisions of Chapter 6 for the description of revised licensing basis of the analysis of the containment spray system - iodine removal for the breaks for which exemption from GDC-41 is requested.

c. Provide UFSAR revision of Section 6.3 for the description of revised licensing basis analysis of the ECCS for the breaks for which exemption from GDC-35 is requested.

11. Please note that the use of risk-based approach for resolution of GSl-191 requires a change in the licensing basis for the CSS operating in the presence of debris. RG 1.174 describes an acceptable approach for assessing the nature and impact of proposed licensing basis changes. This RG requires that the licensee should identify all SSCs, procedures, and activities that are covered by the licensing basis change being evaluated.

The response to question 1.a of the April 15, 2014, RAI, states that the CSS is the only system for which the exemption from GDC-38 is requested. Note that the CSS has associated supporting systems such as the safety-related electrical, Emergency Diesel Generator (EOG), instrumentation and control (l&C), and cooling water systems.

Therefore, as required by RG 1.174, please identify all the associated SSCs, procedures and activities that support the operation of the CSS for containment heat removal in the presence of debris.

12. The response to question 1.b of the April 15, 2014, RAI, does not state which requirements of GDC-38 will not be met. The key GDC-38 requirements to be met for the CSS system design, concurrent with functioning of associated systems are as follows:

(1) Perform the safety function of containment heat removal, and rapidly reduce the containment pressure and temperature and maintain them at acceptably low level.

(2) Safety function (1) shall be performed following any LOCA.

(3) Safety function (1) shall be performed in the presence or absence of Loss of Offsite Power (LOOP).

(4) Safety function (1) shall be performed in the presence of a worst single failure.

Note that requirement (2) covers all postulated LOCAs of any break size, including the most limiting from debris generation, containment peak pressure, and containment peak temperature standpoint.

Please provide the following information:

a. Is full exemption from the GDC-38 requirements (2), (3), and (4) requested? If so, irrespective of the break size, break location, or quantity of debris generation, all CSS trains along with their supporting system may be used. Please provide justification for the proposal of a full exemption from these requirements.

b. Is a partial exemption from GDC-38 requirement (2) requested (i.e., for specific LOCAs only and full exemption from requirements (3) and (4))? If so, specify the LOCAs in terms of location, break size, and debris generation rate for which the exemption is requested from meeting requirement# (3) and (4), and provide justification for the exemption request.

13. The response to question 2.a of the April 15, 2014, RAI, states that the CSS is the only system for which the exemption from GDC-41 is requested. Note that the CSS also has associated supporting systems to which GDC-41 may apply. Please list all the associated systems that support the operation of the CSS; such as the safety-related electrical, EOG, l&C, and cooling water systems. Therefore as required by RG 1.174, please identify all the associated SSCs, procedures and activities that support the operation of the CSS for fission product removal in the presence of debris.

14. The response to question 2.b of the April 15, 2014, RAI, does not state which requirements of GDC-41 will not be met. The key GDC-41 requirements to be met for the CSS system design, concurrent with functioning of associated systems are as follows:

(1) Please list systems required to perform the safety function of controlling fission products, hydrogen, oxygen, and other substances that may be released into the reactor containment to reduce, consistent with the functioning of other associated systems, the concentration and quality of fission products released to the environment and to control the concentration of hydrogen and oxygen and other substances in the containment atmosphere to assure that containment integrity is maintained.

(2) Safety function (1) shall be performed following all postulated accidents.

(3) Safety function (1) shall be performed by providing suitable redundancy in components and features, suitable interconnections, leak detection and isolation, and containment capabilities.

(4) Safety function (1) shall be performed in the presence or absence of LOOP.

(5) Safety function (1) shall be performed in the presence of a worst single failure.

Please provide the following information:

a. Is full exemption from the GDC-41 requirements (2), (3), (4), and (5) requested?

If so, than irrespective of the break size, break location, or quantity of debris generation, all CSS trains along with their supporting system may be used.

Please provide justification for the proposal of a full exemption from these requirements.

b. Is a partial exemption from GDC-41 requirement (2) requested (i.e., for specific LOCAs only, and full exemption from requirements (3), (4), and (5))? If so, specify the LOCAs in terms of location, break size, and debris generation rate for which the exemption is requested from meeting requirements (3), (4), and (5),

and provide justification for the exemption request.

15. The response to question 9.a of the April 15, 2014, RAI, states that the ECCS is the only system for which the exemption from GDC-35 is requested. Please note that the ECCS whose subsystems are High Head Safety Injection (HHSI) and the Low Head Safety Injection (LHSI) systems are not the only ones for which the proposed exemption to GDC-35 would apply. List all of the supporting system that support the operation of the HHSI and LHSI subsystems; for example the safety-related electrical, EDG, l&C, and cooling water systems. Therefore as required by RG 1.174, please identify all the associated SSCs, procedures and activities that support the operation of the HHSI and LHSI systems in the presence of debris.

16. The response to question 9b of the April 15, 2014, RAI, does not state which requirements of GDC-35 will not be met. The key GDC-35 requirements to be met for the ECCS design, concurrent with functioning of associated systems are as follows:

( 1) Perform the safety function of transferring heat from reactor core at a rate such that (a) fuel and clad damage that could interfere with continued effective core cooling is prevented and (b) clad metal-water reactor is limited to negligible amounts.

(2) Safety function (1) shall be performed following any LOCA.

(3) Safety function (1) shall be performed by providing suitable redundancy in components and features, suitable interconnections, leak detection and isolation, and containment capabilities.

(4) Safety function (1) shall be performed in the presence or absence of LOOP.

(5) Safety function (1) shall be performed in the presence of a worst single failure.

Note that requirement (2) covers all postulated LOCAs of any break size, including the most limiting from debris generation or peak clad temperature standpoint. Please provide the following information:

a. Is full exemption from the GDC-35 requirements (2), (3), (4), and (5) requested?

If so, irrespective of the break size, break location, or quantity of debris generation, all ECCS trains along with their supporting system may be used for performing safety function (1). Please provide justification for requesting a full exemption from these requirements.

b. Is a partial exemption from GDC-35 requirement (2) requested (i.e., for specific LOCAs only and full exemption from requirements (3), (4), and (5))? If so, specify the LOCAs in terms of location, break size, and debris generation rate for which the exemption is requested from meeting requirement# (3), (4), and (5),

and provide justification for the exemption request.

17. In question 7 of the April 15, 2014, RAI, the NRC staff requested the licensee to provide the equivalent of UFSAR Section 6.2.1.5, which should describe the licensing basis of the minimum containment pressure analysis for performance capability of ECCS in the presence of debris for the risk-based analysis. Successful functioning of the LHSI, HHSI systems and the CSS in the presence of debris requires exemption from GDC-35 and GDC-38. Therefore, in the presence of debris during LOCAs, the description of the minimum containment pressure analysis for performance capability should be different from what is described in the UFSAR Section 6.2.1.5. The licensee's response to question 7 did not describe the proposed containment analysis, including assumptions and inputs, performed for the calculation of minimum containment pressure input for the ECCS analysis that calculates the peak cladding temperature for risk-informed GSl-191.

Please justify that the inputs and assumptions are conservative for the purpose.

18. Please provide the following additional information with respect to your response to question 3.b of the April 15, 2014, RAI:

a. Refer to the table on page 9 of Attachment 3 to the licensee's letter dated June 25, 2014 (ADAMS Accession No. ML14178A481), of major qualitative differences, for the subject "Sump Pool Treatment," please explain what is meant by: "No decay heat added. Mass and energy subtracted from the pool based on RELAD-30 instructions."

b. Refer to the table referenced in item a) for the subject "Pipe break mass/energy source," please explain what is meant by: "Communicated from RELAP5-3D via coupling interface as problem time progresses. The source is split by MELCOR into part liquid water, part steam, and part 'fog'."

c. Refer to the table under the heading "Summary Comparison of Main Parameter Values," on page 10 of Attachment 3 to the licensee's letter dated June 25, 2014, please provide the basis for selecting the RELAP-30/MELCOR values of the parameters in the table below and how are they determined:

RELAP-30/MELCOR VALUE Initial atmosphere temperature 119.93 °F Initial containment pressure 14.94 psia Initial relative humidity, partial pressure of water vapor 70%1 1, 184 psia

Initial RWST temperature 85 °F Spray actuation times 15 s delay after setpoint, linear ramp to full flow Fan cooler actuation times 15 s delay after setpoint

d. Refer to the table referenced in item c) for the CONTEMPT and RELAP-30/MELCOR analysis, please provide the basis for using different values of (1) thermal conductivity of concrete, (2) thermal conductivity of stainless steel, (3) specific heat capacity of concrete, (4) specific heat capacity of stainless steel, and (5) density of stainless steel.

Using DETERMINISTIC APPROACH Acronvms perform conservative LOCA NPSH analysis for the entire break spectrum CAP Containment accident considering debris (Notes 1 & 5)

A shall be the licensing basis analysis for the Al CDF pressure Core damage entire break spectrum ~

I GDC frequency General Design using deterministic NPSHA ~ NPSHRett I Criterion (Note 2) NPSHA < NPSHRett HI Hydraulic Institute approach (Note 3) LERF Large early release frequency Perform conservative LOCA LOCA Loss of coolant NPSH analysis without debris for accident LOOP Loss of offsite power breaks in which NPSHA < NPSH Net positive suction NPSHRett in A (Note 5). B head No NPSHA < NPSHRett NPSHA NPSH available licensing NPSHRett 'NPSH required' basis including uncertainty PRA Probabilistic Risk NPSHA ~ NPSHRett Assessment (Note4)

NPSHA < NPSHR3% Use PRA APPROACH for breaks analyzed in B. 1. Analysis shall be in compliance with GDC-38

2. Criteria shall be satisfied for the entire break spectrum considering debris.

3. Criterion is satisfied for some breaks Perform realistic LOCA NPSH cases considering debris.

Perform realistic analysis for break cases analyzed 4. Criteria shall be satisfied for all break LOCA NPSH in B while considering debris cases analyzed without considering analysis for debris break cases in C in (Note 6). c debris.

NPSHA < NPSHR3% 5. Conservative LOCA NPSH analysis shall which NPSHA < (Note 3) be based on conservative input NPSHR3% assuming parameters and assumptions to minimize no single failure & NPSHA while assuming single failure and no LOOP. LOOP.

0 NPSHA ~ NPSHR3% 6. Realistic LOCA NPSH analysis shall be (Notes 3 & 10) based on nominal input parameters and assumptions while assuming single failure and LOOP.

NPSHA ~ NPSHR3% 7. HI definition of NPSHR3% is the NPSH (Note 11) corresponding to a decrease in the pump total dynamic head of 3% for a given flow.

8. NPSHRett = NPSHR3%+ Uncertainty

9. CAP guidance in ADAMS document Verify if ~GDF and ~LERF are acceptable + other requirements ML13015A437 shall be followed for determining uncertainty.

10. Partial exemption from GDC-38 is required (because of not meeting the requirement for entire break spectrum with debris) for break cases analyzed in C that meet Not acceptable N 1--~~~~~~~~~- 0 NPSHA ~ NPSHR3%*

licensing 11. Partial exemption from GDC-38 is required Acceptable basis (because of not meeting the single failure criteria, LOOP, and the entire break spectrum with debris) for breaks cases analyzed in D

• C shall be the licensing basis for breaks analyzed in C in which NPSHA ~ NPSHR3'fo.using PRA approach.

• D shall be the licensing basis for breaks analyzed in D in which NPSHA ~ NPSHR3% using PRA approach.

• A shall be the licensing basis for breaks analyzed in A in which NPSHA ~ NPSHRett using deterministic approach.

FIGURE 1: Flow Chart for Defining the LOCA Containment NPSH Licensing Basis Analvsis for Deterministic and Risk-Based GSl-191 Resolution

Nuclear Performance and Code Review Branch (SNPB)

NOTE: Round 2 RA/ question numbers begin with the next sequential number from the April 15, 2014, RA/ for this section.

The NRC staff has reviewed the STP RELAP5-3D with the 1-D core analyses, entitled "Core Blockage Thermal-Hydraulic Analysis." The NRC staff recognizes that these analyses have the objective of demonstrating that under the blocked core inlet cases, sufficient water can match boil-off and maintain coolability and the RELAP5 analyses have shown the conditions for which this is true. However, the staff also recognizes that while sufficient water addition to the core is to be justified to match/exceed boil-off, precipitation of boric acid in the core with various blockages also needs to be addressed. As such, the analyses only address the first critical issue for long-term cooling, but would require an evaluation of precipitation to be able to state that long-term cooling has been demonstrated. Without the precipitation evaluation, long-term cooling cannot be justified. It is noted that the RELAP5-3d code tracks the boron solute concentration, however it does not include boric acid build-up on the liquid density and the static head term in the momentum equation. As such, flow rates and thermal hydraulic behavior may be of concern. Also, transport properties with increased boric acid concentrations is also omitted in RELAP5-3D. The NRC staff requests the following additional information:

6. For the small 2-inch cold leg break of Table 2, while water fills the steam generator cold sides spilling over to the hot side and refilling the core to keep it cooled, the question of precipitation could be an issue that represents failure for this case. That is, with the core totally blocked there is no means of flushing the boric acid build-up in the core that begins upon initiation of boiling. If it assumed no water can pass through the blocked region from cold side injection then switching to hot side injection should not flush the boric acid build-up from the core. It would be instructive to perform a precipitation calculation to show the timing for precipitation once the core begins to boil. Since the RCS pressure is fairly high the precipitation limit will be likewise higher, but it is not clear that the precipitation limit will not be reached. It appears that with the core totally blocked, precipitation cannot be avoided. Please explain and provide an evaluation of precipitation timing for this case.

7. The cases with one assembly unblocked (center and periphery) presented in Figure 32 shows adequate water enters the core to match boil-off. However, as boric acid builds up in the core, the density increases degrading the flow into the core. Given that the downcomer level is fixed due to the break, flow would be expected to decrease as the density in the core increases. As such, calculation of the precipitation timing and mixing in the core needs to be evaluated. Since there is only one unblocked assembly bottom location, it is not obvious that the switch to simultaneous injection can flush the boric acid from the core that builds-up prior to the switch to preclude precipitation.

Furthermore with only the one open assembly inlet path to the core regions, locations near the periphery can trap boric acid and cause local build-up of concentration that may not be flushed out with hot side injection. It is not clear that precipitation can be precluded for these blocked cases. Please provide a detailed explanation.

8. The case in Figure 32 with the bypass free shows adequate water enters the core for cooling. Please identify the elevations above the bottom of the core where these bypass

paths are located. If the first bypass is located above the bottom elevation of the core, this region of the core below the first bypass path will trap boric acid and build-up to potentially reach precipitation. It is not clear how the downward and then upward flow can flush the boric acid from this lower isolated region. If the bypass is located at the core bottom elevation it is still not clear if simultaneous injection can arrest the build-up of boric acid and flush the core through the bypass region. Please explain how precipitation is prevented and demonstrate that RELAP5-3D can predict the correct flows to flush the core under these unusual flow path configurations. Since the RELAP5-3D code does not include the density increases with boric acid concentration, please explain and demonstrate that the flow and mixing behavior in the core can be correctly calculated. What validation calculations have been performed to show that the omission in the momentum equation do not provide excessive flow and mixing behavior, noting that the transport properties are also omitted in the code.

9. Please describe how the advection term in RELAP5-3D is numerically expressed and demonstrate that numerical diffusion does not produce erroneous or excessive flow behavior that could change the conclusions of this analysis. Since advection and diffusion can play key roles in affecting the calculated liquid and steam velocities in the core, please demonstrate that RELAP5-3D can properly model these effects. It may be advantageous to solve the transport equation with advection and diffusion in a 1-D pipe and 3-D volume using the same numerical approximation in RELAP5-3D for the advection and the second order viscous diffusion terms. Please show that a step function density wave or concentration wave moving down the pipe does not suffer from numerical diffusion characteristic of the 1-D upwind differencing scheme that has been employed in RELAP5 code versions.

10. The review indicates that the switch to simultaneous injection for some of the cases occurs at different times for the various breaks evaluated. For example, Figure 8 shows the switch time at about 32,000 seconds for the 2-inch hot leg break while Figure 27 shows about 22,000 seconds for the switch for the double ended guillotine hot leg break.

Typically the switch time is an Emergency Operating Procedure action and occurs at one time that is sufficiently early enough that assures all break sizes are flushed prior to reaching the precipitation limit for the limiting case. These differences should have no impact on the analysis conclusions but please explain the basis and verify that the use of different timing has no impact on the results and conclusions and does not impact the Emergency Operating Procedure guidance for the operators.

Safety Issue Resolution Branch (SSIB)

NOTES:

a. Round 2 RA/ question numbers begin with the next sequential number from the April 15, 2014, RA/ (Round 1) for this section.

b. Follow-up questions from the STPNOC responses to the Round 1 RA/ questions refer back to the Round 1 RA/ number from this section unless otherwise specified.

43. In question 2 of the April 15, 2014, RAI, the licensee stated that the values for size distributions for the fibrous insulation are documented in Reference 46. Reference 46 was not included in the submittal. Please provide a summary of the relevant size information from Reference 46 in the form of a table including the size distributions within the postulated ZOls.

44. In question 4 of the April 15, 2014, RAI, that NRC staff's review indicates that it is likely that the STP methodology discussed in the response may be acceptable and may provide conservative transport results when considered within the probabilistic framework. However, low density fiberglass (LDFG) congestion may not be the metric that dominates the likelihood of debris reaching the strainer based on break location.

Although the use of the steam generator compartment transport fraction may be moderately conservative as claimed, the NRC staff was unable to verify this assumption.

It was also not clear to the staff that the measure of fiber congestion within specifically defined volumes in containment provide the most important measure of debris amounts that may be generated or the probability that debris would be generated within those volumes. If one location is congested, but the fibrous debris in that area cannot be damaged by a break it is not relevant. Please verify that the methodology results in overall realistic or conservative transport fractions considering the possible break locations and the LDFG congestion.

45. For question 6.a. of the April 15, 2014, RAI, the NRC staff finds that the licensee did not provide an adequate response to the question. The Drywell Debris Transport Study (DOTS) states that if gratings do not cover an entire transport path that they may not be as effective in debris capture as noted in the test metrics. Simply using a ratio of open area to total area may not provide a realistic or conservative estimate of debris capture.

The grated area is likely to have higher resistance to flow that will increase as it collects debris. Debris is generally assumed to be homogeneously distributed throughout the blowdown flow. If less volume of blowdown passes through the grating due to flow resistance, less debris is available to pass through or collect on the grating. The Nuclear Energy Institute (NEI) baseline guidance assumes that small fines are debris that will pass through gratings, so no holdup of small or fine fibrous material is assumed using baseline methodology. The baseline further assumes small fines to be the basic constituent of the debris for transport purposes. There were no refinements regarding crediting gratings to reduce transport found in either NEI 04-07 or the NRC SE on 04-07.

Therefore, the licensee should justify the assumption that the amount of debris captured by gratings in pathways that are not fully covered can be estimated using a simple ratio of the open area to total area. Please provide a justification that the debris capture metrics used in the evaluation are realistic considering the issue identified above.

46. The licensee's response to question 7.b. of the April 15, 2014, RAI, stated that the significantly longer washdown periods at STP, compared to the length of the DOTS washdown tests are inconsequential to the STP evaluation. The conclusion is based on a portion of the NEI guidance document, NEI 04-07, that found the erosion of fibrous debris by containment spray is less than one percent. The RAI aimed at the erosion of fibrous debris by containment spray, but requested for clarification if the washdown of fibrous debris through gratings would increase above that found during the DOTS and if the washdown time is significantly increased? The NRC staff is specifically interested in the small fiber washdown transport fractions provided in Table 2.2.23 of Volume 3 of the licensee's submittal dated November 13, 2014. These values are currently listed as 7-19 percent washed down in the annulus and 21-27 percent washed down inside the secondary shield wall. These do not appear to be fibrous erosion values. Please provide justification that the washdown values from a 30-minute test are applicable to the STP condition considering the clarification provided.

47. In question 14 of the April 15, 2014, RAI, the NRC staff requested the basis for the use of 1/16 inch as the value below which a filtering bed is assumed not to occur. The licensee's response to the question is based on NRC staff's acceptance of the head loss correlation and a sensitivity study that showed no change in CDF if the criterion is reduced to zero inches. Because neither has been accepted at this time, the acceptability of the response to RAI 14 is indeterminate. Additionally, the use of a 1/16-inch criterion below which chemical effects cannot occur is not supported by some industry tests that had 1/16 inch of fiber or less added. Some tests had measureable increases in head loss with less than 1/16-inch theoretical fiber on the strainer after chemicals were added to the loop. The NRC staff agrees that it is unlikely that a head loss great enough to result in strainer failure will occur with such a low fiber load.

However, the potential for the head loss to result in flashing or additional deaeration was not addressed by the licensee. The sensitivity study was also conducted before corrections to pool level and clean strainer head loss (CSHL) values were implemented.

As stated above, the NRC staff has not accepted the head loss correlation used to perform the sensitivity study. The licensee is requested to provide revised response to RAI 14 considering the information discussed above.

48. Questions 15, 16, 17, 18, 21, and 22 of the April 15, 2014, RAI, requested additional information regarding the licensee's use of a correlation to calculate debris head loss.

The NRC staff has established a position that correlations may not be used to calculate head loss unless the correlation is validated, under plant-specific conditions, for the range of conditions to which the results will be applied. This position was discussed with the licensee before the formal submittal. The NRC staff does not consider the responses to be adequate since the licensee's use of correlations were not validated under plant-specific conditions. The licensee is requested to provide a revised response consistent with the NRC staff position.

49. The licensee's response to question 27 of the April 15, 2014, RAI, stated that the use of 0.220 ft as the CSHL value was an error. The licensee performed sensitivity studies to determine the effect of using the correct value of 1.952 ft on overall CDF. The licensee stated that the change in CDF would be about 18 percent when the correct value is

used. The licensee also stated that a more accurate CSHL value would be used. Does it mean that 1.952 ft is the "more accurate" value of CSHL? If not, please provide the "more accurate" value of CSHL that will be used in future calculations.

50. The licensee's response to question 28 of the April 15, 2014, RAI, stated that the use of a head loss correlation is essential to the risk-informed method because it provides understanding of subtle interactions between variable parameters considered in the analysis. However, the need to apply a 5X safety factor to bound uncertainties in the correlation indicates that confidence in the method is relatively low and that evaluation of interactions between the parameters may be significantly skewed. These relationships may be further mischaracterized by resorting to a limiting packing factor for the debris bed. The response provides a sensitivity study for safety factor values around the 5X value used in the evaluation. However, the response does not provide a basis for the values used in the study. The NRC staff believes that because there are uncertainties in many aspects of the model and that many of these are significant, that the 5X multiplier may not envelope these uncertainties. The RAI response does not appear to address two significant issues, the uncertainty caused by non-homogeneous beds and the lack of testing to validate the model for plant-specific conditions that lead to model uncertainty.

Other uncertainties inherent to the use of correlations for head loss should also be addressed including statistical uncertainties arising from the use of test data, uncertainties arising from the use of the correlation, and uncertainties introduced by assuming that test conditions are representative of the plant. Please provide an evaluation of how the individual uncertainties within the model are accounted for and provide an estimate of the total uncertainty created by use of the model.

51. The licensee's response to question 31 of the April 15, 2014, RAI, described a calculation that evaluates the potential for the collection of gas bubbles in the STP strainer. The licensee cites Reference 56 of Volume 3, TDl-6005-07, "Vortex Air Ingestion and Void Fraction South Texas Project Units 1 and 2," Revision 3, November 17, 2008, which evaluates the transport of gas voids in the piping between strainer and ECCS and CS pumps. Neither a copy of the referenced document nor any applicable details from the reference were provided to the NRC staff. It was also not described how it was determined that voids would not collect in the strainer. Please provide a summary of the relevant sections of Reference 56 describing how it was determined that voids would not collect in the pump suction piping. Additionally, please provide information that evaluates whether voids can collect within the strainer, and if they do, how the effect was evaluated.

52. The licensee's response to question 33 of the April 15, 2014, RAI, stated that the CASA Grande model overestimates the water level compared to computer aided design calculated levels. The licensee stated that the error will be corrected so that future submittals contain accurate pool levels. However, the licensee also needs to verify that strainer submergence is adequate and that vortexing, deaeration, and flashing evaluations adequately reflect the corrected levels and that transport is not affected due to higher pool velocities. Please provide information that justifies that these areas are not adversely affected.

53. The licensee's response to question 34 of the April 15, 2014, RAI, stated that total CSS flow is determined by multiplying the random pump flow rate by the number of operable CSS pumps. These flow rates are randomly selected from between the maximum calculated flow rate and some minimum value. It was not clear that using random values is appropriate and how the minimum values were calculated. The licensee also stated that for all two and three train cases that CASA uses the higher two train flow, since it is conservative. The licensee included reference to Reference 42, Volume 3, 5N109MB01024, "Design Basis Document Containment Spray," Revision 3, November 17, 2004. Neither a copy of the referenced document nor any applicable details from the reference were provided to the NRC staff. Please summarize the relevant information from Reference 42, provide the methodology used to determine the minimum flow rates, or provide the basis for using random flow rates for each event instead of calculating event specific flow rates.

54. The licensee's response to question 36 of the April 15, 2014, RAI, states that strainer buckling is the limiting failure criterion when compared to NPSH. It was not clear that flashing was considered as a failure mode for the strainer in the STP submittal. Please state how flashing across the strainer is evaluated by CASA Grande since this failure mode may be more limiting than strainer buckling when the fluid temperature is high.

55. For question 41.c. of the April 15, 2014, RAI, the NRC staff has accepted the use of mitigative measures to address defense-in-depth. The licensee credited backwash of the strainers as a mitigative measure. However, it was further stated that the mitigative measures for backwash of the ECCS strainers have not been proceduralized. Please describe the procedural requirements that are in place to initiate ECCS strainer backwash or revise the submittal to remove its credit.

55. CASA Grande uses a distribution for the temperature at which chemical effects are assumed to occur and a distribution for the conventional head loss bump up factor.

Volume 3 states that chemical effects are assumed to occur below 140°F and that the conventional head loss bump-up factor is 5. Please state which methodology is intended to be used and update the documentation or the model to reflect the intended methodology.

56. CASA Grande does not implement the bed compression aspects of the NUREG-6224 Correlation. Volume 3, equations 33-38 imply that the compression function is implemented in CASA Grande. The NRC staff understands that this issue was addressed by implementing a limiting bed compression for all debris head loss calculations. Please verify that this has been accomplished and provide updated results based on the updated method. Please provide the basis for the assumption that the limiting bed compression chosen is appropriate.

57. The NRC staff has several concerns with the model used for fiber penetration through the strainer. Considering the issues described in this RAI, the NRC staff does not have high level of confidence that the debris penetration model accurately represents the expected debris penetration and in-vessel fiber accumulation that could occur in the plant. Please provide information that justifies that the CASA Grande calculations for fiber penetration are meaningful and represent the plant conditions:

a. Assumptions and modeling techniques regarding debris arrival timing and filtration may result in non-conservative bypass results. In the response to question 6.b. of the April 15, 2014, RAI, the licensee stated that early arrival of debris at the strainer resulted in higher filtration and lower total bypass. The response to RAI 11 b stated that debris transported during pool fill is placed directly on the strainer at the initiation of the LOCA. This is also related to the "non-intuitive results found during a sensitivity study provided to the NRC staff for review. The NRC staff believes that the result is non-intuitive because it is non-physical. Debris arrival timing should not have a significant effect on filtration, if realistic timing is used. The staff understands that placing fiber on the strainer at the start of recirculation may be conservative with respect to head loss, but may be non-conservative with respect to strainer penetration. The NRC staff has determined that assuming homogeneous mixing of fiber in the pool at the start of recirculation rather than assuming that some fiber transports to the strainer prior to recirculation is likely to be more conservative. The NRC staff understanding is based on the relatively short time during which significant bypass occurs and the longer time over which head loss becomes more risk significant. Please provide information that justifies the STP approach is conservative or incorporate a methodology that is more appropriate.

b. If the existing model or a model that results in penetration being highly dependent on arrival timing early in the event is maintained, please justify why the model is not more correlated to the amount of debris arriving at the strainer regardless of timing. If debris arriving at the strainer at the initiation of the LOCA affects the calculated bypass amount please justify this model behavior. Does the model assume that early arriving material can pass through the strainer? If not, please justify the assumption. Also, please provide justification that less debris would bypass the strainer in the plant if debris arrives at the strainer earlier in the scenario, that is, the model accurately reflects plant performance.

c. How are the uncertainties resulting from applying bypass test results to the plant condition accounted for in the model? Are there conditions potentially present in the plant that would result in more bypass than occurred in the relatively controlled test conditions? Please explain.

d. How are uncertainties associated with the strainer bypass calculation accounted for? The calculation appears to be very sensitive to arrival timing. Also, how are uncertainties that arise from testing and the translation of test results into bypass models accounted for? Please explain.

e. The NRC staff noted that changing the time step in the CASA Grande debris penetration model has a significant effect on the output (amount of debris reaching and accumulating in the core). CASA Grande uses a relatively inaccurate method to integrate the mass balance equations for debris accumulated in the core, especially early in the accident sequence after initiation of recirculation. Please describe how the licensee determined that the time step interval and integration method provide appropriate results (ideally, the

conditional probability of failure by exceedance of the cold-leg break fiber limit should be independent of the computational time steps).

f. The NRC staff noted that one input parameter to the CASA Grande code to compute the filtration efficiency is one order of magnitude more than that determined by testing and documented in Volume 3 of the submittal (Table 2.2.28, parameter m1e51 : the upper bound is 0.0003723 1/g; instead a value of 0.003723 1/g was apparently used in the CASA Grande computations in support of license submittals). The result of this error is overestimation of the filtration efficiency, which causes underestimation in the amount of fiber penetration and in-vessel accumulation. Sensitivity studies suggest that this error would underestimate the cold-leg break in-vessel fiber limit failure contribution to the CDF by about an order of magnitude. Please explain and include comparisons of filtration efficiencies and shedding rates computed by Monte Carlo sampling to test data in your response.

58. The NRC staff reviewed the relationship between break size, and CASA Grande failure predictions. The results of the review indicate that there may be discontinuities in the results that suggest that failures due to certain break sizes are not predicted or are much less likely to occur than would be expected. For example one break sized at about 5 inches results in a failure. With respect to break size, no additional failures occur until the break size reaches about 10 inches. This behavior appears to be non-physical.

Please discuss this observation and provide an evaluation of whether this behavior affects the results of the analysis.

59. It is the NRC staff 's understanding that the computer-aided design (CAD) model used to determine debris generation amounts was developed under a 10 CFR 50, Appendix B program, and therefore treats the output to be accurate. However, it may not be the case for the debris generation values used in CASA Grande. Please describe the methodology used to import the CAD values into CASA Grande and provide information that describes how the debris generation amounts used by CASA Grande were validated to be accurate. Please include information that demonstrates how the interfaces between the CAD model or its input to CASA Grande were validated to be correctly implemented and describe whether raw CAD values were validated to be the same as those used in CASA Grande.

60. Section 5.4.3 of the submittal dated November 13, 2013, indicates that almost 100 percent of the break scenarios generate less than 10 ft 3 of fiberglass debris (the probability of generating more than 10 ft 3 is smaller than 10-12 ). Using a density of 2.4 lb/ft3, the equivalent mass of 10 ft 3 of fiber is 24 lbs (10.89 kgs). The NRC staff review of the CASA Grande program indicates that there may be a significant number of cases that generate much more than 10 ft 3 of fiberglass (hundreds and up to one-thousand kg).

Please clarify the meaning of Figure 5.4.5 in Volume 3, which appears to imply that the probability of generating more than 10.89 kg of fiberglass is smaller than 10-12 , and clarify if this information was used in the CASA Grande model. Please clarify if the information in Figure 5.4.5 in Volume 3 also includes latent fiber.

61. For breaks that are not DEGB and are assumed to have a hemispherical ZOI, please explain how are the robust barriers treated? For example, if the break is on a pipe near the floor and occurs on the bottom of the pipe, is the potential for damage from a reflected jet accounted for?

62. During review of the licensee's response to question 9 of the April 15, 2014, RAI, the NRC staff developed an additional question regarding the treatment of debris in the head loss calculation. Please clarify if the small and fine debris are treated as if they have the same properties in the head loss calculation (correlation)? In your response, please clearly explain how each debris size is treated?

63. Based on the licensee's response to ESGB question 1.b. of the April 15, 2014, RAI, it appears that some large breaks, many medium breaks, and all small breaks do not generate enough debris to result in a 1/16-inch bed when distributed over 3 strainer trains. Please provide the following information:

• Distribution of low density fiberglass (LDFG) debris mass reaching the strainers for small, medium, and large breaks separately.

• The amount of latent fibrous debris that reaches the strainers for each break category and if it varies, provide the distribution and methodology used to determine the amounts.

• The range of the mass of fine fibrous debris and small piece fibrous debris generated for each of the break categories.

• The range of the masses of these fiber categories that transport to the strainer.

64. Based on the review of the licensee's response to question 2 of the April 15, 2014, RAI, the staff has identified the following concern: what causes the variability in the head loss calculation performed by the correlation? For example, scenarios that contain apparently similar debris loads (CASA Grande Case 1 in the RAI response) may have significantly different calculated head losses. The head losses from the referenced tests represent CASA Grande values in the 99th percentile, indicating that almost all head losses predicted by CASA Grande are lower than the test results. The limiting CASA Grande Head loss calculation was 8.2 feet for conventional debris and 161.9 feet for total head loss, which is much higher than the test results. These maximum values seem higher than could possibly occur, at least for the total head loss. Explain if these maximum values realistic or are they non-physical predictions. Explain why CASA Grande predicts lower head losses than the test results over 99 percent of the time.

65. RG 1.174 states that licensees are expected to evaluate "whether sufficient safety margins would be maintained if the proposed licensing basis change were to be implemented." The NRC staff recognizes that safety margin cannot be characterized by a single number for a time-dependent analysis with multiple failure modes. Instead, it can be represented by an equation or relationship that represents the safety margin as a function of time for each of the seven GSl-191 failure modes. For example, the safety margin with respect to strainer mechanical collapse can be represented as:

Sm = 9.35 ft - ~P(t)

=

Where Sm margin with respect to strainer mechanical collapse

~P(t) = differential pressure across strainer as a function of time Please describe whether CASA Grande calculates success with respect to each of the seven GSl-191 failure modes in a manner that is consistent with RG 1.174 guidance on safety margins. Specifically, please identify the failure threshold (worst allowable value) for each failure mode and state whether it is consistent with existing licensing basis calculations.

66. Assumption 1j of Volume 3 states that "switchover to hot leg injection would occur between 5. 75 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the start of the event." Assumption 11 a states "the current STP design basis evaluation methodology used to calculate the required hot leg switchover timing is appropriate with the exception of GSl-191 related phenomenon."

When analyzing boric acid precipitation in regards to post-LOCA long-term core cooling, the mixing volume and percentage of voids in the core used in the analyses need to be justified. Improper modeling could result in non-conservative liquid volume after a LOCA. Ultimately, this could impact the hot-leg switchover time in a plant's emergency operating procedures. STP's calculation for hot-leg switchover time following a LOCA (NC-7136, Revision 1) was provided in response to SNPB RAI 4. An input for this calculation is liquid volume in the RCS. Please provide the mixing volume and percentage of voids in the core for STP licensing basis calculations used to determine the liquid volume in the RCS for hot leg switchover timing in the calculation to validate assumptions 1j and 11 a. Please justify the use of these numbers and any assumptions made. The licensee can refer to NRG-approved methods, as appropriate.

Technical Specification Branch (STSB)

NOTE: Round 2 RA/ question numbers begin with the next sequential number from the April 15, 2014, RA/ for this section.

4.

Background:

In response to NRC staff comment/question 2.4 (page 6 of 179, Volume 6.2), the licensee stated the following:

A description of how the proposed change will affect the technical specifications is provided in Regulatory Evaluation Section 4.1.3 in the LAR provided in Enclosure 3. As discussed in more detail in Enclosure 3, no changes to operability requirements for affected systems and no changes to the existing technical specification Action Statements are proposed. Proposed changes to the technical specification bases that conform to the changes in the licensing and design bases are included in Attachment 3 to Enclosure 3 for staff information.

Page 1 of 1 of Attachment 3 to Enclosure 3 of the licensee's letter dated November 13, 2013, "Technical Specifications Bases Page Markups," states, in part:

UFSAR Appendix 6A provides a risk-informed approach that addresses the potential of debris blockage concluding that long-term core cooling following a design basis loss of coolant accident is assured with high probability. UFSAR Appendix 6A also provides guidance for assessing the potential impact on Operability due to unexpected material such as loose debris discovered in containment that may contribute to debris loading on the strainers.

Page 15 of 16 of Attachment 2 to Enclosure 3 of the licensee's letter dated November 13, 2013, "STPEGS UFSAR Page Markups" states, in part, the following:

The table provides guidance that may be used to immediately assess the potential impact due to unexpected material discovered in containment that may contribute to debris loading on the strainers. As discussed in Reference 6A-4 [licensee letter dated June 19, 2013], these values are not necessarily the limiting amount of each type as analyzed.

Conservatisms in the reported values are also discussed in Reference 6A-2 [licensee letter dated December 11, 2008]. Therefore, a condition that may exceed the values shown in the table does not preclude reasonable expectation of operability.

Input Parameter Value Minimum Debris Type (Reference 6A-6) Margin Latent debris, consisting of:

Dirt and/or dust 200 lbm (Total) 170 lbm (1.0 cubic ft) 40 lbm (Total) 34 lbm (0.2 cubic ft)

• Fiber, e.g., fibrous insulation 30 lbm (12.5 cubic ft) 6 lbm (2.5 cubic ft)

Miscellaneous debris, including but not 100 sq-ft 10 sq-ft limited to unqualified tags and labels Unqualified coatings Table 6.1-4 100 sq-ft Due to the complexity of the analysis, the potential exists for conditions to be discovered which may not be represented by the values in the table, and for which evaluations would be required to evaluate the impacts [ ].

Page 26 of 31 of Enclosure 3 to a letter from STP dated November 13, 2013, states, in part, the following:

When warranted, an immediate operability determination will be followed by a prompt operability determination that will apply additional information and supporting analyses to confirm the immediate operability determination. Evaluations may consider additional information provided in the inputs to the CASA Grande analysis as well as the identified conservatisms associated with the categories of major assumptions in the CASA Grande analysis, Section 3 of Volume 3 (Enclosure 4-3).

For a discovered condition that potentially affects debris quantities in containment, the applicable CASA Grande input parameters and assumptions provide a means for immediate operability determinations and follow-up determinations, as warranted, to evaluate the impact on containment sump performance.

Concern: It is the NRC staff's position that when evaluating operability of an SSC, the use of risk assessment or probabilities of occurrence of accidents or events is unacceptable. The definition of operability is that the SSCs must be capable of performing their specified safety function or functions. This inherently assumes that the event occurs and that the safety function or functions can be performed. Operability is not indeterminate. An SSC required to be operable must be able to perform its specified safety function or it is inoperable.

The NRC staff is concerned that the CASA Grande design inputs (parameters and assumptions) referred to in Reference 6A-5 of the UFSAR Markup includes probability aspects the licensee proposes to be acceptable to be used during an operability determination if a condition is discovered that potentially affects debris quantities in containment and the need arises for evaluating the impact on containment sump performance.

Request: provide the following additional information:

1) An explanation of how the assumptions referred to in Reference 6A-5 of the UFSAR Markup and discussed in Section 2.2 of Volume 3 of the same document will be used during an operability determination. Please include an example to the extent practical.

2) If the licensee is proposing to allow the use of risk information in the assessment of operability, then:

a. Please provide a description demonstrating the relationship between probability and operability for each of the assumptions discussed in Section 2.2 of Volume 3.

b. Please explain how would the probability of occurrence of each of the assumptions discussed in Section 2.2 of Volume 3 change to improve or degrade the impact on containment sump performance.

D. Koehl If you have any questions, please contact me at 301-415-1906 or via e-mail at Lisa. Regner@nrc.gov.

Sincerely, IRA/

Lisa M. Regner, Senior Project Manager Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-498 and 50-499

Enclosure:

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