ML15187A119

Comment (1) of Wayne Harrison and Michael P. Murray on Draft Regulatory Guide DG-1322 - Risk-Informed Approach for Addressing the Effects of Debris on Post-Accident Long-Term Core Cooling

ML15187A119
Person / Time
Site:	South Texas   (NPF-076, NPF-080)
Issue date:	06/24/2015
From:	Harrison W, Marc-Anthony Murray South Texas
To:	Cindy Bladey Rules, Announcements, and Directives Branch
References
80FR21658 00001, DG-1322, NOC-AE-15003267, NRC-2015-0095
Download: ML15187A119 (47)

Text

6/26/2015 6/262015N RC-2015-0095-DRAFT-0002.html As of: 6/26/15 3:28 PM Received: June 25 201 5 PUBLIC SUBMISSION Status: =Pendin-gPo st Tracking No. ljz-8jme-vcpx Comments Due: July 06, 201 Submission Type: Web Docket: NRC-20 15-0095 Alternate Risk-Informed Approach for Addressing the Effects of Debris on Post-Accident Long-Term Core Cooling Comment On: NRC-2015-0095-0001 Alternate Risk-Informed Approach for Addressing the Effects of Debris on Post-Accident Long-Term Core Cooling; Draft Regulatory Guide for Comment Document: NRC-20 15-0095-DRAFT-0002 Comment on FR Doc # 20 15-08964 15 Submitter Information Name: Wayne Harrison Submitter's Representative: Wayne Harrison (awharrison@stpegs.corn)

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Nuclear Operating Company South Texas I'ro/cct Eklecrlc Cfeneratln& Station PO. BoX 289 Wadsworth, Texas 77483

• v/v June 24, 2015 NOC-AE-1 5003267 10OCFR50.46c Cindy Bladey Office of Administration Mail Stop: OWFN-1.2H08 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001.

South Texas Project Units 1 and 2 Comments on Draft.Regulatory Guide DG-1 322 - Risk-Informed Approach for Addressing the Effects of Debris.on Post-Accident Long-Term Core Cooling Docket ID NRC-201 5-0095-0002 STP Nuclear Operating Company (STPNOC) and the participants in the STP Pilot Risk-Informed GSI-1 91 program have reviewed the subject draft regulatory guide associated with I0CFR50.46c rule change, and the group's comments are attached. The review includes input from representatives of STPNOC, Ameren, Pacific Gas & Electric, NextEra Energy/Florida Power & Light, Southern Company, Exelon and Wolf Creek Nuclear Generating Company.

The attached Comments were made by several reviewers and consequently vary in perspective in some cases. The reviewers believe the regulatory guide will be a valuable tool and add clarity to the risk-informed assessment of debris effects. One common theme among the reviewers is focused on the need for additional clarity in differences in the requirements for the integrated approach as opposed to the simplified approach.

If you should have any questions on the comments, please contact Steve Blossom at (361) 972-7495,.or Wayne Harrison at.(361) 972-8774.

Michael P. Murray Manager, Regulatory Affairs

Attachment:

DG-1 322 with comments STI 34149431

NOC-AE-1 5003267 Page 2 of 2 cc: (electronic)

Steve Frantz, Esquire Morgan, Lewis & Bockius LLP Lisa Regner John Stang U. S. Nuclear Regulatory Commission John Ragan Chris O'Hara Jim von Suskil NRG South Texas LP Kevin PolIo Cris Eugster L. D. Blaylock CPS Energy Peter Nemeth Cramn Caton & James, P.C.

C. Mele City of Austin Steven Blossom South Texas Project Charles Gears NextEra Energy/Florida Power & Light Craig Sellers Exelon Corporation Roger Andreasen Ameren Corporation Phil Grissom Southern Company Rasool Baradaran Pacific Gas & Electric Company Maurice Dingier Wolf Creek Nuclear Generating Station Bruce Montgomery Nuclear Energy Institute

FEDERAL REGISTER Vol. 80

Monday, No. 75 April 20, 2015 Nuclear Regulatory Commission 10 CFR Parts 50 and 52" Alternate Risk-Informed Approach for Addressing the Effects of Debris on Post-Accident Long-Term Core Cooling

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM DG-1 322 LANGUAGE COMMENTS RISK-INFORMED APPROACH FOR ADDRESSING THE EFFECTS OF DEBRIS ON POST-ACCIDENT LONG-TERM CORE COOLING

1) INTRODUCTION Purpose This regulatory guide (RG) describes methods and approaches that the Comment: Is the purpose of the rule to also allow for use of staff of the U.S. Nuclear Regulatory Commission (NRC) considers the risk-informed approach in other GDCs and rules? If so, acceptable for demonstrating compliance with the voluntary, risk-informed should that purpose also be stated in the RG?

alternative for addressing the effects of debris during long-term cooling in 10 CFR 50.46c; "Emergency core cooling system performance during loss-of-coolant accidents (LOCA)" (Ref. 1) 10 CER 50.46c requires that the ECCS have the capability to provide long-term cooling of the reactor core following any successful initial operation of the ECCS. The ECCS must be able to remove decay heat so that the core temperature is maintained at an acceptably low value for the extended period of time required by the long-lived radioactivity remaining in the core. The rule contains a provision in section (e) that allows the voluntary use of a risk-informed approach to address the effects of debris on long-term cooling. The risk-informed approach is an alternative to deterministic approaches for complying with paragraph (d)(2)(iii) of the rule.

2

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM DG-1 322 LANGUAGE COMMENTS This RG describes acceptable methods and approaches for addressing Comment: General comment on RG scope:

paragraph 50.46c(e), "Alternate risk-informed approach for addressing the While it is clear that this RG is focused on debris effects on effects of debris on long-term core cooling," and paragraph (m)(4),

ECCS and containment spray in their long-term cooling role, "Updates to risk-informed consideration of debris in long-term cooling," of teenest ergltr udnefrcnanetha 10 CFR 50.46c. While the general risk-informed approach in this RG may removal that typically relies on CSS operating in conjunction be applied to any reactor design within the scope of 50.46c, many of the with ECCS and for containment atmosphere cleanup, which specific approaches (e.g., WCAP-1 6530-NP-A for chemical effects) and relies on CSS to manage dose, including dose to the control acceptance criteria (e.g. 15 grams per fuel assembly for hot leg break) were ro.Ntsgetn tsol ei hsRol htaR 2

developed for the current fleet of PWRs. Licensees or applicants using this b

rprdt drs t guidance should justify that the application of each approach or method used meets the intent of this guidance.

Applicable Rules and Regulations

• Title 10, Part 50, of the Code of Federal Regulations (10 CFR 50),

Comment: Should GDC 38 and 41 also be included?

"Domestic Licensing of Production and Utilization Facilities" (Ref.

2).

• 10 CFR 50.46c, 'Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors."

• 10 CFR 50 Appendix A, "General Design Criteria for Nuclear 3

Power Plants," Criterion 15, 'Reactor coolant system design" (Ref.

3).

• 10 CFR 50 Appendix A, "General Design Criteria for Nuclear Power Plants," Criterion 35, "Emergency core cooling" (Ref. 4).

3

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM DG-1 322 LANGUAGE COMMENTS Related Guidance The NRC's risk-informed approach includes consideration of risk, defense-in-depth, and safety margins, and the NRC expects licensees to implement performance measurement strategies to ensure these principles continue to be addressed. This RG does not change these principles, but rather builds on existing guidance and provides additional detail for the specific risk-informed analysis of the effects of debris on ECCS long-term cooling performance. The following RGs are relied upon in large measure as set forth in Section C of this RG:

Comment 1: Does this imply that we will need a process to periodically monitor performance and confirm that analysis inputs/assumptions and PRA results are still valid?

Comment 2: Consider adding RG 1.177 if changes to the TS are being proposed as part of the risk-informed approach.

0 RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant Specific Changes to the Licensing Basis" (Ref. 5).

RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities" (Ref. 6).

w

• RG 1.82, 'Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident" (Ref. 7).

4 4

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM DG-1 322 LANGUAGE COMMENTS Purpose of Regulatory Guides The NRC issues RGs to describe to the public methods that the NRC staff No comment (staff) considers acceptable for use in implementing specific parts of the agency's regulations and to provide guidance to licensees and applicants.

Regulatory guides are not substitutes for regulations and compliance with them is not required. Methods and solutions that differ from those set forth in regulatory guides will be deemed acceptable if a basis acceptable~to the NRC for the specific application is provided and it meets the applicable regulatory requirement.

Paperwork Reduction Act This RG contains information collection requirements covered by 10 CFR No comment 50 that the Office of Management and Budget (0MB) approved under 0MB control number 3150-0011. The NRC may neither conduct nor sponsor, 6 and a person is not required to respond to, an information collection request or requirement unless the requesting document displays a currently valid 0MB control number.

2) DISCUSSION Reason for Issuance This guide addresses the risk-informed alternative in 10 CFR 50.46c(e).

No Comment This section of the rule allows licensees to address the effects of debris on long-term core cooling using a risk-informed approach as an alternative to deterministic approaches, which typically rely on plant-specific or generic performance tests that use conservative test protocols and do not allow 7

credit for non-safety-related mitigation capabilities. This guide is intended to describe a risk-informed approach acceptable to the NRC that licensees can use in addressing the effects of debris on long-term core cooling.

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ENCLOSURE 1 GSI-191 Option 2b Comments ITEM DG-1 322 LANGUAGE COMMENTS

~~~~~~~~~~Background 8

The risk-informed alternative for consideration of effects of debris during post-accident long-term core cooling in 10 CFR 50.46c implements Commission direction in the Staff Requirements Memorandum (SRM) related to SECY-12-0093, "Closure Options for Generic Safety Issue -

191, Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance" (Ref. 8) and in the SRM related to SECY-12-0034, "Proposed Rulemaking - 10 CFR 50.46c: Emergency Core Cooling System Performance During Loss-of-Coolant Accidents (RIN 3150-AH42)"

(Ref. 9). Without this alternative, licensees would need to seek exemptions from the rule to implement the risk-informed approach.

Efforts have been focused in the past on ascertaining the reliability of ECCSs in nuclear power plants during design-basis accidents. The performance of sump strainers for recirculation of cooling water could be challenged by the presence of debris - whether already present in the containment or generated as a result of an initiating event such as a LOCA. RG 1.82, "Sumps for Emergency Core Cooling and Containment Spray Systems," Revision 0, (Ref. 10) required licensees to assume a 50-percent blockage for recirculation sump strainers in their analyses.

Generic Letter (GL) 85-22, "Potential for Loss of Post-LOCA Recirculation Capability Due to Insulation Debris Blockage" (Ref. 11), later called for replacement of the 50-percent blockage assumption with a more comprehensive requirement to assess debris effects on a plant-specific basis.

A number of events occurred during the 1990s that motivated re-examination of the reliability of ECCS strainers during accident conditions at operating boiling water reactors (BWRs). The NRC requested that BWR

.licensees implement appropriate procedural measures, maintenance practices, and plant modifications to minimize the potential for the clogging of ECCS suction strainers by debris accumulation following a LOCA. The BWR-related research helped to identify issues related to the adequacy of pressurized water reactor (PWR) strainer designs in general. The BWR research findings demonstrated that the amount of debris generated by a high-energy line break (HELB) in a PWR could be greater, that the debris Comment: 5 th Paragraph Recommend changing the wording of the paragraph to:

In response to GL 2004-02,

• ~u~bie most licensees have implemented major modifications to their plants to ensure adequate recirculation system performance. For example,.

some most licensees have significantly increased the size of strainers, and some have replaced fibrous insulation with reflective metal insulation, the debris of which is considered less likely to reach or impede flow through strainers.

Demonstrating adequate performance of strainers is challenging given the difficulty of testing them such that all conditions (e.g., temperatures, debris amounts and compositions, and operating components of the ECCS and CSS) that might exist during an accident are properly addressed. It is also difficult to develop reasonable, reliable, and validated models for strainer performance operating under complex conditions.

The basis for the proposed change is to more properly characterize the efforts that have been taken by the industry.

6

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM DG-1 322 LANGUAGE COMMENTS could be finer (and thus more easily transportable), and that certain combinations of debris (e.g., fibrous material plus particulate material) could result in a substantially greater head loss through ECCS strainers than an equivalent amount of either type of debris alone. The NRC opened Generic Safety Issue (GSI)-1 91, "Assessment of Debris Accumulation on PWR Sump Performance," to track these issues. The objective of GS1-1 91 is to ensure that post-accident debris blockage will not impede or prevent the operation of the ECCS or containment spray system (CSS) in recirculation mode at PWRs during LOCAs or other HELB accidents for which recirculation is required.

The NRC issued GL 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors," (Ref. 12) requesting holders of operating licenses for PWRs to address GS1-1 91. Specifically, licensees were requested to perform a mechanistic evaluation of the recirculation functions and, as appropriate, to take additional actions, such as plant modifications, to ensure system functionality. From the results of testing and analyses, the NRC identified additional issues, such as the combined effect of chemicals and debris on strainer performance and the effects of debris penetration through the strainer and into the reactor vessel and reactor coolant system.

In response to GL 2004-02, a number of licensees have implemented major modifications to their plants to ensure adequate recirculation system performance. For example, some licensees have significantly increased the size of strainers, and some have replaced fibrous insulation with reflective metal insulation, the debris of which is considered less likely to reach or impede flow through strainers. Demonstrating adequate performance of strainers is challenging given the difficulty of testing them such that all conditions (e.g., temperatures, debris amounts and compositions, and operating components of the EGGS and CSS) that might exist during an accident are properly addressed. It is also difficult to develop reasonable, reliable, and validated models for strainer 7

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM DG-1 322 LANGUAGE COMMENTS performance operating under complex conditions.

The NRC staff prepared two Commission papers (SECY papers) (SECY-12-0093 and SECY 0034) that include risk-informed options to address GSI-191. The Commission issued SRMs for SECY-12-0093 and SECY-12-0034 directing the staff to propose revised regulations in 10 CFR 50.46c to contain a provision allowing GS1-191 to be addressed, on a case-by-case basis, using risk-informed alternatives, without the need for an exemption (e.g., under 10 CFR 50.12, "Specific Exemptions"). The objective of this RG is to provide guidance to licensees that choose to implement the risk-informed approach for addressing the effect of debris on post-accident long-term core cooling. This guidance is consistent with RG 1.174, and it may be used by licensees to support the staff's approval of a risk-informed application.

Harmonization with International Standards The NRC staff reviewed guidance from the international Atomic Energy No Comment Agency, International Organization for Standardization, and International Electrotechnical Commission and did not identify any guidance from these organizations that provided useful information specific to the topic of risk-informed consideration of the effects of debris during post-accident long-term core cooling.

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM DG-1 322 LANGUAGE COMMENTS Documents Discussed in Staff Regulatory Guidance This regulatory guide refers to several industry documents (e.g., topical No Comment reports) that provide information that may be used in the risk-informed analysis of debris. These industry documents are not approved by the staff in this RG, unless this RG expressly indicates approval of the identified industry document. The staff approval may be conditioned, as stated in this RG. The bases for any of these conditions are set forth in this RG. NRC approval of these references, including any limitations or conditions, is 10 contained in the safety evaluation for those specific documents, which is either included in the final version of topical reports or separately referenced in this regulatory guide. These referenced industry documents are provided as examples of approaches that may be used for specific portions of the risk-informed analysis as set forth herein. In the future, other topical reports or industry documents may be reviewed and endorsed by the NRC staff. This regulatory guide neither endorses nor modifies the previous NRC approval of these industry documents.

C. STAFF REGULATORY GUIDANCE 10 CFR 50.46c(e) requires that an application be submitted to the NRC to Comment: Section C.20 includes this footnote defining PRA, request the use of the alternative risk-informed approach for consideration "In this context, "PRA" also includes any complementary of effects of debris during post-accident long-term core cooling. This analyses (e.g., debris evaluation model, human reliability section provides descriptions of the methods, approaches, and data that analysis) that are used to calculate the increase in risk the NRC staff considers acceptable for meeting the requirements of the attributable to debris."

regulations cited in the Introduction. The methods, approaches, or data in Ti per oidct gemn ihteaiiyt efr 11 these regulatory guidance positions are not requirements.

the risk analysis without modifying the PRA model. However this guidance is hidden in the very back, vague in completeness as a lost footnote. This guidance should be front and center in the final regulatory guide and comprehensive allowing the full range of analyses provided that the applicable statement of the ASME Standard and RG 1.200 are not violated.

9

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM jDG-1 322 LANGUAGE

]COMMENTS

1. The risk assessment required by 10 CFR 50.46c(e) should include all relevant initiating events and plant operating modes for all hazard groups for which debris could adversely affect core damage frequency (CDF) or large, early release frequency.(LERF). Therefore, the application should identify and group all scenarios that could be mitigated by the activation of sump recirculation. In this context, the term scenario means an initiating event followed by a plant response (e.g., combination of equipment successes, failures, and human actions) leading to a specified end state (e.g., success, core damage, large early release). These scenarios should be grouped in a logical fashion, for example according to initiating event.

12 Comment 1: Clarify if the intent is to identify/determine the scope of PRA model, in terms of operating modes and hazards, required for the risk analysis. Initial efforts have been focused on calculating CDF/LERF impacts using Internal events PRA only (modes 1, 2). As currently worded, we are required to show, on a plant-specific basis, that initiating events resulting from other hazards (e.g. Seismic) and operating modes would not result in LOCAs requiring recirculation which could be adversely affected by debris.

Comment 2: This statement should be modified to clarify that the only initiating events that need to be considered are those that could generate debris from the effects of a jet from a RCS pipe break. Some plants consider the use of the recirculation mode for cases where a different initiating event could require aligning the recirculation sump to provide core cooling. For example, some plants place reliance on the containment recirculation sump strainers as an alternative source for supplying core cooling during conditions where there is a loss of refueling canal water level in Mode 6 or loss of secondary to heat sink event.

Comment 3: The term "initiating event" needs to be better defined. For purposes of this guidance, an initiating event should be a design basis event described in the FSAR that will result in generation of debris that can adversely affect long-term cooling.

Comment 4: "all relevant initiating events" are not a priority limited to LOCAs. For plants with High Head injection capable of lifting relief valves, transients may have to be considered.

This point is alluded to later in C.14 d (3) 10

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM DG-1 322 LANGUAGE COMMENTS Consistent with RG 1.174, the licensee may exclude hazard groups and Comment: See comment 1 and.2 for item 12.

operating modes from further consideration when the licensee demonstrates, qualitatively or quantitatively, that the corresponding risk contribution of the excluded hazard group or Operating mode would not affect the decision being made or overall conclusion of the risk-informed 13 analysis. Any such screening should be performed on a plant-specific basis and the licensee should document the basis for each hazard or operating mode not being included in the risk-informed analysis. For screening purposes, these scenarios should be grouped in a logical fashion, for example according to initiating event.

An example of screening criteria that could be used for a PWR might be Comment 1: As a minimum, any scenario or group of the following: "As a minimum, any scenario or group of scenarios meeting scenarios meeting all the following four ---. In the second all of the following four inclusion criteria should be included in the risk-paragraph in same sections, 'When the licensee informed analysis:

demonstrates, qualitatively or quantitatively ---"does this

a. The scenario response involves recirculation to provide core cooling; saeetapyt tt ntidprgah"samnmm-

"7

b. The scenario involves the potential for debris inside primary containment that could adversely impact SSCs needed for recirculation; Comment 2: Provide examples of other scenarios not meeting all four criteria that should be considered for inclusion 14

c.

The scenario involves a mechanism that could transport the debris to the in the risk analyses.

sump; and, Comment 3: As discussed in the comment to Item 12, if an

d. The debris is necessary for the scenario to result in core damage oradionlsrengctrawsaddriembws containment failure.

diinlsreigciei a

deo tmb a

modified to consider only those scenarios that have the potential for debris to be generated from the jet resulting from a RCS pipe break, then those scenarios would be more correctly categorized.

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ENCLOSURE 1 GSI-191 Option 2b Comments ITEM DG-1 322 LANGUAGE COMMENTS

2. The licensee should identify the debris-related failure modes for each Comment 1: "The license should identify the debris related SSC whose successful operation helps to mitigate the postulated failure modes for each SSC whose successful operation helps.

scenarios screened as included under Paragraph (0.1 of this RG. For to mitigate the-..."Does this mean SSC as a component or example, it is expected that the ECOS would be identified during this step.

as a system as the example states e.g. pump of the system, The EGCS may fail because of the following debris-related failure modes or the RHR system or the entire ECCS as a component?

(the list is not exhaustive and other failure modes may need to be Also does this mean listing each valve in the system also for considered):

downstream wear?

a. Excessive head loss at the strainer leads to loss of net positive suction head Comment 2: Delete paragraph C.2.g. Boric acid precipitation (NPSH) margin for adequate operation of pumps; is independent of GS1-1 91. Paragraph C.2.e. adequately 15

b. Excessive head loss at the strainer causes mechanical collapse of the strainer; addresses long-term core cooling considerations applicable to GSI-1 91.

c. Excessive head loss at the strainer lowers the fluid pressure, causing release of dissolved gasses (i.e., degassing) and void fractions in excess of pump Comment 3: In the paragraph on page 6 that immediately limits. Vortexing and flashing may also cause pump failure; follows 0.2 g, it seems that the example given for screening

d. Debris in the system exceeds ex-vessel limits (e.g., blocks small passages in does not have to consider chemical effects. Whether or not downstream components or causes excessive wear);

this is the case in NRC's view should be explicit.

e. Debris results in core blockage and core heat transfer limits are exceeded;

f. Debris buildup on cladding exceeds heat transfer limits; and,

g. Debris buildup in the vessel leads to potential excessive boron concentrations within the core caused by reduction of fresh coolant entering the core.

12

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM DG-1t322 LANGUAGE COMMENTS The licensee may exclude debris-related failure modes from further No comment consideration if a bounding analysis shows that maximum credible debris loads under detrimental configurations (e.g., fine and compact debris filling the void space of a fibrous debris bed) would not lead to a given failure mode. For excluded failure modes, the licensee should still consider direct and indirect effects of debris on SSC performance for other parts of the analysis.

For example, an analysis may show that a bounding amount of debris 16 would not completely block flow through the residual heat removal (RHR) heat exchanger, with a maximum loss in heat transfer rate not sufficient to significantly change cooling rates and cause core damage. In this example, exclusion of this failure mechanism (i.e., flow blockage) for that RHR heat exchanger might be justifiable. However, the estimated percent reduction in heat transfer rates would still need to be considered when computing temperatures of water volumes inside containment (i.e., pool temperatures), which may affect other failure modes. As another example, analysis may show that the strainer can function with the calculated amount of debris; however, in-vessel limits may be exceeded.

3. After identifying and screening relevant scenarios and debris-related Comment 1: Clarify if the intent is to permanently incorporate failure modes of SSCs, the licensee should evaluate failure modes into the plant's baseline model of record or just in the PRA identified from Paragraph 0.2 of this RG and identify how to incorporate model used in the analysis to calculate delta CDF and LERF.

these failure modes into the probabilistic risk assessment (PRA) model to Cmet2 o h ipiidapocdsusdltri be used for the risk assessment, which is used to calculate COF and teRtebs R

oe a

o aet emdfe LERF. The "baseline" PRA model for assessing the risk increase 17 attributable to debris is one where the effects of debris are assumed to be snescesi salse ybudn eemnsi negligible. For example, the baseline PRA model might not distinguish evlainadalotrcssaeasudtogtoflr.

between successful actuation of one train of ECCS versus two trains, as Comment 3: "the licensee should evaluate failure modes either would meet the traditional PRA success criterion for a LOCA. When identified from Paragraph 0.2 of this RG and identify how to evaluating the effects of debris, however, the distinction between one and incorporate these failure modes into the probabilistic risk two trains may be important as it may impact the distribution of debris (to assessment (PRA) model to be used for the risk assessment, one versus two strainers) as well as safety injection flow rates and could, which is used to calculate CDF and LERF"

____therefore, affect the probability and frequency of ECCS debris-related 13

ENCLOSURE 1 GSI-191 Option.2b Comments ITEM DG-1 322 LANGUAGE

]COMMENTS failure modes. Changes to the PRA should be clearly described in the application to the NRC. Any operator actions credited with reducing the CDF or LERF attributable to debris should be clearly described.

- incorporate into the PRA model, if applicable. The simplified approach/RoverD assumes a CCDP of 0.0 or 1.0 which precludes the need for use of the PRA model (i.e. plant mitigating systems) for calculating CDF. Need a defined method for calculating LERF for the simplified approach (i.e.

correlating CDF to LERF).

Comment 4:

1. The draft states that the failure modes will be "incorporated" into the PRA which is not necessary for a thorough and RG 1.200 compliant analysis to be completed. An alternative is to use the PRA of Record which meets RG 1.200 and related Peer Review and ASME requirements to generate information illuminating the likelihood and frequency of Recirculation following a postulated event with full fidelity as to the scenario circumstances such as number and type of pumps functioning and flow rates and related phenomena. This can be analyzed to determine the impact of all identified phenomena and potential failure modes with failure representing a delta CDF or LERF with and without consideration of mitigative actions. Such effects are consistent with application of the RG 1.174 criteria.

This paragraph goes on to state, "Changes to the PRA should be clearly described in the application to the NRC." As stated above these may not be "changes" but fully developed analysis of continued mitigation of an alternative end state, namely Recirculation to evaluate the delta COF and LERF.

Current PRA analyses already include in many cases a probability of recirculation failure and the current analyses is actually a refined calculation with all known information of a possibility previously simply estimated. Certainly structural 14

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM DG-1 322 LANGUAGE COMMENTS failure and collapse were envisioned as low probability events in the original analysis of some plants. It is an important distinction that the actual risk relative to the prior PRA values may even go down for some plants.

Comment 5: Section 0.3: perhaps adding 'e.g. HHI in large LOCA response models' after the reference to safety injection flow rates would add clarity.

4. It is anticipated that licensees pursuing a risk-informed approach to Comment: Add the following in the first sentence:

evaluate the effects of debris on the ECCS and CSS functions will utilize

".... utilize integrated models to evaluate strainer and integrated models to evaluate strainer and downstream system downstream system performance, including the following performance, including the following:

submodels for: "

a. debris source term (debris generation mechanisms and debris size distribution);

b. debris transport and accumulation on strainers;

c. strainer head loss and criteria for strainer failure (e.g., available head less than 18 the required net positive suction head, flashing, and deaeration);

d. debris penetrationthrough strainers and downstream effects (such as debris accumulation inside the reactor pressure vessel);

e. chemical effects that could increase flow resistance (for example by the formation of chemical precipitates) and head loss through debris beds on strainers and in the vessel; and, f.effects of safety-related and non-safety-related system activation to mitigate the event (e.g., strainer blockage, in-vessel effects, and ex-vessel downstream effects).

Integrated models should account for uncertainty in parameters and Comment 1: We need more explanation and details of phenornenological models, as well as for the frequency of initiating events expectations for uncertainty analyses. Are distributions of 1 9 (e.g., frequency of large break LOCAs) and intensity of those events (e.g.,

parametric uncertainty required for each submodel that size of the pipe break causing a LOCA). Paragraphs C.5 to C.13 Of this RG generates an input into the analyses? Can sensitivity provide details on guidance of the integrated model components.

analyses which determine the parameters that the analyses 15

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM DG-1 322 LANGUAGE COMMENTS are most sensitive to be used?

Comment 2: For the simplified approach, an integrated model will most likely not be r'equired. Therefore, uncertainty modeling will also not be required. This is supported by the statements in C.6.e and C.7.d.

5. The licensee should develop descriptions of the as-built and as-Comment 1: Specify the criteria to be used to identify operated nuclear power plant (i.e., accounting for the effects of components important to the risk-informed analysis of debris debris) system evaluated by phenomenological, physical, and effects. Component contribution to CDF/LERF is specific to mathematical models identified under Paragraph C3.4 of this RG. The the ECCS equipment operational configurations/alignments of licensee should define the following:

the high likelihood configurations used in the risk evaluations.

a. power plant operating modes and operating components important to the risk-Clarify which integrated model (or models). The integrated informed analysis of debris effects; submodels in C.4 or the modified PRA model?

S20

b. long-term period of performance, including a definition of the safe and stable Comment 2: The wording should be changed for C.5 to end-state of the nuclear power plant (i.e., safe state after mitigation of the require a description of those items described in C.5.a and event) -- the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> mission time typically used in PRAs may not be C.5.b for both the simplified risk-informed approach and a applicable if long term effects (e.g., chemical precipitation) are expected to risk-informed approach that relies upon an integrated model.

occur outside of this time frame; C.5.c and C.5.d are only required for the latter approach.

c. human actions that are part of the accident sequence; and, Comment 3: a. Simplified approach precludes the need for

d. the Set of assumptions and considerations relevant to the development of the use of the PRA model (i.e. plant mitigating systems) for integrated model.

calculating CDF. Need a defined method for calculating LERF for the simplified approach (i.e. correlating CDF to LERF).

6. The licensee should describe the source term for generation of Comment 1: C.6.e - "The licensee should identify relevant debris under a postulated event to be mitigated by activation of the data and model uncertainties, and propagate those recicultionsysemuncertainties into the integrated model to compute the CDF 21 rcruainsse.and LERF for the as-built and as-operated plant (i.e., plant

a. The licensee should describe the postulated accidents and debris gene~ration including debris). The licensee should describe where in its mechanisms (e.g., pipe breakand jet impinging on materials withinanlsscsevtedtrmitcapochs(e.

containment) applicable to the as-built and as-operated plant.

app*roaches that tend to overestimate the CDF and LERF) 16

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM DG-1 322 LANGUAGE ICOMMENTS

b. The licensee should identify the types of debris or materials that could be generated and transported to the strainer and affect its performance, or otherwise affect core or containment cooling. In addition, the licensee should quantify the potential amounts of debris that could be generated by the initiating events and included scenarios identified under Paragraph C.lof this RG.

Licensees may refer to Section C.1.3.3 of RG 1.82 for guidance on identifying debris types and the use of the zone of influence concept to estimate debris amounts. Sections C.1.3.5 and 0.1.3.6 of RG 1.82 provide guidance pertinent to coatings debris and latent debris. The NRC staff review guidance, "NRC Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Coatings Evaluation," provides guidance focused on coatings as a source of debris (Ref. 13).

c. As necessary for quantifying the head loss through a bed of debris on strainers, the licensee should quantify debris characteristics, including material type, size distribution and shape, and density. The licensee should quantify the amount of debris penetration through or bypass around the strainers. The licensee should account for interactions with chemicals in the water when relevant to strainer failure or core damage mechanisms. Safety evaluations in Sections 3.4.3 and 3.5 of the Nuclear Energy Institute (NEI) document, NEI 04-07 (Ref. 14),

provide guidance acceptable to the NRC for quantifying debris characteristics, including latent debris, which the licensee should consider in the development of risk-informed analysis.

d. The licensee should integrate information from Paragraph C.6.a to C.6.c of this RG into a model for quantifying debris amounts after a postulated initiating event. The licensee should verify the validity of the model, relying, for example, on tests and empirical data, analogy to other systems, or comparison with other calculations.

e. The licensee should identify relevant data and model uncertainties, and propagate those uncertainties into the integrated model to compute the ODE and LERF for the as-built and as-operated plant (i.e., plant including debris).

The licensee should describe where in its analysis conservative deterministic approaches (i.e., approaches that tend to overestimate the ODE and LERF) from NRC-approved guidance are used. Doing so will identify areas where propagation of uncertainty is not required. RG 1.174 contains additional guidance on uncertainty quantification and propagation.

from NRC-approved guidance are used. Doing so will identify areas where propagation of uncertainty is not required. RG 1.174 contains additional guidance on uncertainty quantification and propagation." This statement is awkward as it states what is required only to then add exceptions. It would be clearer to state the requirement explicitly as in "For cases in which conservative deterministic values that overestimate the CDF and LERF... "~ Further the requirement should be to illuminate the uncertainty which may be done by propagation of uncertainties or by sensitivity studies.

Comment 2: We need more explanation and details of expectations for uncertainty analyses. Conditional failure probabilities for sump strainer and core blockage. determined based on the analyses in C.6.a through C.6.d will be incorporated into the PRA model used to calculate CDF/LERF impacts. We need to know if the uncertainty analysis program UNCERT (based on the state of knowledge) run as part of PRA model quantification is sufficient to satisfy this requirement.

Comment 3: Embedded within this item is the following statement; Doing so will identify areas where propagation of uncertainty is not required. (C.6.e) It may be beneficial to*

have a separate section in the RG dealing with uncertainty quantification and propagation. Consideration should also be given to the use of sensitivity analysis as a method to demonstrate uncertainty.

Comment 4: 1. C.6.e. should allow for sensitivity studies to be used in lieu of propagating uncertainties into the integrated model. This is consistent with Regulatory Guide 1.174. This statement is also awkward as it states what is required only to then add exceptions. It would be clearer to state the requirement explicitly as in "For cases in which conservative 17

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM DG-1 322 LANGUAGE COMMENTS deterministic values that overestimate the CDF and LERF...

Further the requirement should be to illuminate the uncertainty which may be done by propagation of uncertainties or by sensitivity studies Comment 5: Would the requirement in C.6 e be applicable for a plant following the guidance in C.9?

7. Once the amount and type of debris is characterized, the licensee Comment 1: Same comment as above for C.6.e should describe the mechanism for debris transport to the strainers.

Astireuemnisepadthogut05hogh.1

a. The calculation of debris quantities transported to the ECOS strainers should and elsewhere in this RG, it would be beneficial to have one consider all modes of debris transport, including blowdown, washdown, pool section on uncertainty which addresses all requirements for fill, and recirculation. Section 0.1.3.4 of RG 1.82 provides guidance on the development of deterministic transport analyses and models.

uncertainty and sensitivity analyses. This would include what is required for the integrated submodels, propagation of

b. The licensee should develop a model for debris transport to be used in the integrated model that will be used to calculate CDF and LERF. The transport inputs into the PRA models (including identification of areas model should be consistent with water inventory balance (e.g., safety injection where propagation of uncertainty is not required), and PRA 22 flow rates, containment spray system flow rates) related to the postulated model quantification. Discussion of suitable methods such as event under consideration, those described in 0.14 should be provided.

c. The licensee should evaluate the validity of the transport model, relying, for example, on tests and empirical data, analogy to other systems, or detailed Comment 2: Embedded within this item is the following computational fluid dynamics models.

statement; Doing so will identify areas where propagation of

d. The licensee should identify relevant data and model uncertainties, and uncertainty is not required. (C. 7.d) It may be beneficial to propagate those uncertainties into the integrated model to compute the total have a separate section in the RG dealing with uncertainty ODE and LERF. The licensee should describe where in its analysis quantification and propagation. Consideration should also be conservative deterministic approaches (i.e., approaches that tend togietoheuefsniivyaalissamtodo overestimate the CDE and LERE) from NRC-approved guidance are used.

gietoheuefsniivyaalissamtodo Doing so will identify areas where propagation of uncertainty is not required.

demonstrate uncertainty.

8. The licensee should evaluate the fluid conditions at the strainer, Comment: Recommend providing a reference for acceptable properly accounting for the pool water level, the water volume displaced by use of containment pressure beyond saturation.

hardware, post-accident pressure, and water inventory holdup in upstream 23 paths. Relevant guidance is provided in RG 1.82, Sections C.1.3.1 and C.1.3.7. In general, RG 1.82 recommends conservatively assuming that the containment pressure is equal to the saturation pressure. The licensee should justify use of pressure beyond atmospheric in NPSH computations.

18

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM [DG-1 322 LANGUAGE JCOMMENTS

9. Using this RG, the licensee should perform analyses for each debris-related failure mode identified in Paragraph C.2, accounting for the debris sources in Paragraph C.6 that are assumed to be transported according to mechanisms identified in Paragraph 0.7, with the goal to estimate CDF and LERF. A simplified approach to this part of the analysis may be performed, as described in paragraphs C.9.a through C.9.d below, in which case a licensee could then skip paragraphs C.10 through 0.13.

24 Comment 1: "the licensee may select a simplified "go/no-go" approach by assuming a justifiable range of debris loads and demonstrating through testing that long-term core cooling will be maintained under those debris loads and pertinent conditions of the as-built and as-operated plant. This option would not seek to calculate time-dependent flow conditions at the strainer or in the vessel but would instead compare each scenario to a threshold value, determined by testing.

Scenarios which produce debris exceeding this limit would be assigned a conditional core damage probability (CCDP) of 1.0O. Scenarios which produce less debris than this limit would be bounded by the test results and would be assigned a CCDP of 0."

This is conservative as noted but also ignores the ASME Standard that allows and encourages screening. While CCDP can and should be set to zero for cases that do not create enough debris the cases that do potentially create such debris may still be evaluated to determine potential for mitigation or core damage.

The approach outlined above can be used to define the scenarios and frequencies within which the plant might be Vulnerable to Core Damage (VTCD) in that they do not screen with conservative values but that does not equate to plant damage. Such scenarios are cases for which the plant may be VTCD or'Vulnerable to Large Early Release (VTLER).

Other available features or actions would still need to fail and the conservative assumptions would have to actually occur -

conditions that can be analyzed. Thus the simplified approach may allow screening of insignificant scenarios thereby allowing the analyst to focus more clearly on the non-screened cases.

These can then be analyzed in great detail to generate best estimate CDF and LERE or to by sensitivity studies illuminate the sensitivity.

19

ENCLOSURE 1 GSI-19I Option 2b Comments ITEM DG-1 322 LANGUAGE COMMENTS Comment 2: Paragraph C.9. Allows the user to skip steps 03.10. through (C.13. while paragraph C.9.a. refers the user to paragraph (C.1 1. It is recommended that the reference to paragraph (C.1 1. in paragraph C.9.a. be deleted.

Comment 3: It states that licensee can skip paragraphs C.10 through 0.13 if a simplified approach is used but in paragraph 9a. it sends the licensee to 0.11 and 0.13. Please clarify.

Comment 4: It would improve clarity regarding what is expected from the Staff if Section 0.9, which describes a different approach when compared to the other sections, was set apart from the other sections more explicitly. It is not clear what, if any, requirements from other sections are applicable for a utility adopting this simplified approach. (I call your attention that Section 10 immediately following starts with the phrase "Licensees who do not select the simplified approach..." The arrangement of the sections may cause confusion and possible misinterpretation.

In a simplified approach, acceptance criteria for the 'tests' should be addressed. That is, direction above what important test constituents should be included would add clarity, Comment 1: Section C.9 states four requirements: each For the simplified approach to analyzing debris-related failure modes, the shudloweprainftecnevtistolumae licensee may select a simplified "go/no-go" approach by assuming a justifiable range of debris loads and demonstrating through testing that sensitivity and uncertainty.

longter coe colig wil b mantanedundr thse ebrs ladsand Simplified does not "calculate time-dependent flow conditions log-ermin core cooiong wil bhe masbintained undoerathoedebrntTis loadsiand at the strainer or in the vessel but would instead compare perti nosektcacattiedenent fo conditions of the asbitada-prtdpath stoptione each scenario to a threshold value determined by testing."

wou inoth veseek bto calulantea timedpaendent h

flwcndriotion at thesrainerd This may be used to say what the owners group is doing on 25 arliute, vesselmne but woldistead.

Scompareo eaich srdcenaerio to xctheshold saying chemical production is not for X hours cannot be used valu, dterine bytestng.Scearis wichprodce ebrs eceeing or it can be read as only time dependent debris and flows this limit would be assigned a conditional core damage probability (CCDP).

based on T/H should not be used.

of 1.0. Scenarios which produce less debris than this limit would beLatsnecinfrtpagphayweankp0.0o bounded by the test results and would be assigned a CCDP of 0. If this 0.1abt 9ascn sentence saysisee paragraphsy ecnski 0.11 and apprach s slectd:C.13.

Need to clarify intent.

.* 20

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM JDG-1 322 LANGUAGE COMMENTS

a. The licensee should define a range of loads, debris types, debris combinations, debris arrival sequences, and interactions with chemicals in the fluid (see Paragraphs C. 11 and C.13 of this regulatory guide), where the strainer is not expected to fail. Testing should be conducted per guidance in RG 1.82, Section 1.3.12, to support conclusions of strainer performance.

b. The licensee should determine when an initiating-event could result in debris loads that are predicted to cause head losses greater than those shown acceptable under Paragraph C.9.a of this RG, and assume system failure whenever those conditions are predicted to occur.

c. The licensee should define a range of debris loads, debris types, debris combinations, debris arrival sequences, and interactions with chemicals in the fluid where adequate flow to the core is maintained. Debris load limits should be defined by testing. WCAP-1 6793, Revision 2 (Ref. 15), has been accepted by the NRC staff (with conditions and limitations) as adequately defining in-vessel debris limits. This topical report, or other NRC accepted topical reports or methods, may be used to define in-vessel debris limits. Analysis may be used to show that water can reach the core via alternate flowpaths. The analysis should demonstrate that the alternate flowpaths provide adequate coolant flow and cannot be blocked by debris.

d. The licensee should determine conditions when an initiating event can result in in-vessel debris limits or loads that are greater than those found acceptable under Paragraph C.9.c of this RG, and assume system failure whenever those conditions are predicted to occur.

9.b says we determine the initiating event could results in debris loads that are predicated to cause head loss greater than shown under C.9.a. Does this say we don't have to do this in detail as defined in C.1 and 0.2. 0.3. etc.?

If not PWROG analysis, what would it be? How would 'not blocked" be determined by analysis?

Comment 2:C0.9 - Apparently absent from this section is discussion of debris penetration. Recommend modifying the statement in C.12.a and adding language to this section since this is a consideration for the simplified approach.

0.9 - Apparently absent from this section is discussion of ex-vessel effects. Recommend adding the statement from C.13.a to this section.

C.9.a -Why are we referring to 0.11 and 0.13 if previously told in 0.8 that if the simplified approach is being used that 0.10 through 0.13 can be skipped.

Comment 3:

b. "... assume system failure whenever those conditions are predicted to occur."

What is the definition of "system" used here? Using the simplified approach just assumes a CCDP of 0.0 or 1.0.

Comment 4:

1. This is conservative as noted but also ignores the ASME Standard that allows and encourages screening. While CCDP can and should be set to zero for cases that do not create enough debris the cases that do potentially create such debris may still be evaluated to determine potential for mitigation or core damage.

The approach outlined above can be used to define the 21

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM DG-1322 LANGUAGE COMMENTS scenarios and frequencies within which the plant might be Vulnerable to Core Damage (VTCD) in that they do not screen with conservative values but that does not equate to core damage. Such scenarios are cases for which the core may be VTCD or Vulnerable to Large Early Release (VTLER). Other available features or actions would still need to fail and the conservative assumptions would have to actually occur - conditions that can be analyzed. Thus the simplified approach may allow screening of insignificant scenarios thereby allowing the analyst to focus more clearly on the non-screened cases.

These can then be analyzed in great detail to generate best estimate CDF and LERF or to by sensitivity studies illuminate the sensitivity.

2. Section C.9a-d state four requirements. Each should allow exploration of the conservatisms to illuminate sensitivity and uncertainty.

3. C.9.e. should allow for sensitivity studies to be used in lieu of propagating uncertainties into the integrated model.

This is consistent with Regulatory Guide 1.174. This statement is also awkward as it states what is required only to then add exceptions. It would be clearer to state the requirement explicitly as in "For cases in which conservative deterministic values that overestimate the CDF and LERF... " Further the requirement should be to illuminate the uncertainty which may be done by propagation of uncertainties or by sensitivity studies.

Comment 5: In-vessel debris limits are discussed in Section C.9, but boron precipitation is not introduced until Section 0.13 c. Does a utility following C.9 guidance need to consider boron? (see comment 4 to item 24).

22

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM DG-1 322 LANGUAGE COMMENTS In the consideration of in-vessel effects within the context of 03.9, guidance should be expanded on what processes are acceptable for inferring applicability of the tests to unanalyzed (or untested) plant states. For example, a test that involves n-1 ECOS trains may be interpreted to address the more likely nominal case of n ECCS trains. Acceptance criteria for both strainer failure criteria and in-vessel criteria would be helpful.

Again, this is a case where the guidance in C3.9 may differ than other guidance (e.g., 0.15 f). Is the guidance in 0.12 c applicable if 03.9 is followed?

10. Licensees who do not select the simplified approach described in Comment 1: The staff should be more specific on their Par~agraph 03.9 should develop and implement a model for debris acceptance criteria for correlations that support the model, or accumulation and head loss through the potential debris bed developed on specify limitations on assumptions for phenomena such as strainers. The output of this approach is a calculated head loss value and failure of coatings and transport time. Additional specificity is in-vessel debris load for each scenario.

needed with respect to testing requirements to support

a. Guidance on the development of head loss analyses is provided in the safety moesadcrltin.(pisto.1-3 evaluation of Sections 3.4.3 and 3.5 of NEI 04-07 and in RG 1.82, Section Comment 2: C3.10.d. should allow for sensitivity studies to be C.1.3.11.

used in lieu of propagating uncertainties into the integrated

b. The licensee should develop~a model of head loss through the debris bed on model. This is consistent with Regulatory Guide 1.174. This strainers to be used in the integrated model for the computation of ODE and statement is also awkward as it states what is required only to 26 LERF. The model should represent or bound the broad range of possibilities of then add exceptions. It would be clearer to state the 26 debris loads and compositions, as well as pertinent accident conditions.

requirement explicitly as in "For cases in which conservative

c. The licensee should evaluate the validity of the model, relying, for example, on dtriitcvle htoeetmt h

D n

EF..

tests and empirical data, analogous systems, and use of approved guidance.

Further the requirement should be to illuminate the Section 0.1.3.12 of RG 1.82 defines prototypical head loss testing that could uncertainty which may be done by propagation of be used to support models. The model should be validated for the range of uncertainties or by sensitivity studies.

plant-specific conditions and debris loads to which it is being applied.

Validation should be based on results of prototypical head loss testing using appropriate debris types.

d. The licensee should identify relevant data and model uncertainties, and propagate those uncertainties into the integrated model to compute the total ODE and LERF. The licensee should describe where in its analysis conservative deterministic approaches (i.e., approaches that tend to 23

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM DG-1 322 LANGUAGE COMMENTS overestimate the CDF and LERF) from NRC-approved guidance are used, to identify areas where propagation of uncertainty is not required.

11. The licensee should account for the presence of chemicals in the Comment 1: 11.a - Shouldn't this statement also be provided water and interactions with debris that could change the head loss through in Section 9 since it applies to both the simplified approach debris beds.

and a more extensive risk-informed approach?

a, The Westinghouse topical report, WCAP-16530-NP-A, and the limitations Comment 2: 0.11.d. should allow for sensitivity studies to be discussed in the associated NRC staff safety evaluation (Ref. 16) provide an used in lieu of propagating uncertainties into the integrated acceptable approach for the evaluation of chemical effects that may occur in a model. This is consistent with Regulatory Guide 1.174. This post-accident containment sump pool. The NRC staff review guidance, "NRC statement is also awkward as it states what is required only to Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Areathnadecpis.Iwolbelarrosaete of Plant-Specific Chemical Effect Evaluations," (Ref. 17) provides guidance on requirement explicitly as in "For cases in which conservative plant-specific chemical effect evaluations.

deterministic values that overestimate the CDF and LERF...

b.

The licensee should develop a model of chemical effects on flow resistance Further the requirement should be to illuminate the and head loss through the debris bed on strainers to be used in the integrated uncertainty which may be done by propagation of model for the computation of the CDF and LERF. The model should represent uncertainties 27 or bound the broad range of conditions described in the safety evaluation of WCAP-1 6530-N P-A.

c.

The licensee should evaluate the validity of the model, relying, for example on tests, empirical data, and analogies to other systems. The safety evaluation of topical report WCAP-16530-N P-A defines testing and analyses that could be used to support models of chemical effects. The chemical effects model should be validated for the full range of plant conditions and debris loads to which it is applied.

d.

The licensee should identify relevant data and model uncertainties, and propagate those uncertainties into the integrated model to compute the total ODE and LERF. The licensee should describe where in its analysis conservative deterministic approaches (i.e., approaches that tend to overestimate the ODE and LERF) from NRC-approved guidance are used.

Doing so will identify areas where propagation of uncertainty is not required.

24

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM DG-1 322 LANGUAGE COMMENTS

12. The licensee should evaluate debris penetration through the strainer.

Comment 1: C.12.d. should allow for sensitivity studies to be

a.

The licensee should characterize debris penetration through strainers under usdileuopragtnucranisitohenertd potential accident conditions. The licensee should account for all debrismoe.TiiscnsttwthRglorGud117.hs penetration mechanisms or mechanisms where debris can bypass the statement is also awkward as it states what is required only to

strainer, then add exceptions. It would be clearer to state the requirement explicitly as in "For cases in which conservative

b. The licensee should develop a model to estimate the amount of debrisdermnsivausttoeetmteheOEndLR penetration through strainers, with the goal of evaluating downstream effects.Futethrqiemnsoudbtoluiaete 28

c. The licensee should evaluate the validity of the model, relying, for example, uncertainty which may be done by propagation of on tests, empirical data, and analogies to other systems. Testing to validate uncertainties or by sensitivity studies.

the strainer penetration model should be conducted under conditions that are prototypical or conservative with respect to the as-built and as-operated plant.

d.

The licensee should identify relevant data and model uncertainties, and propagate those uncertainties into an integrated model to compute the total CDF and LERF. The licensee should describe where in its analysis conservative deterministic approaches (i.e., approaches that tend to overestimate the ODE and LERF) from NRC-approved guidance are used.

Doing so will identify areas where propagation of uncertainty is not required.

13. The licensee should evaluate the effects of debris strainer penetration Comment:

inside (in-vessel) and outside (ex-vessel) the reactor vessel.

1. Delete paragraph C.13.c. Boric acid precipitation is

a.

The licensee should evaluate downstream ex-vessel effects of debris (e.g.,

independent of GSI-1 91.

blockage of flowpaths in equipment, and wear and abrasion of surfaces). The 0.1 13.g. should allow for sensitivity studies to be used in lieu of safety evaluation for the Topical Report WCAP-16406-P (Ref. 18) and RG propagating uncertainties into the integrated model. This is 1.82 provide guidance that the licensee may use to evaluate ex-vessel effects consistent with Regulatory Guide 1.174. This statement is of debris.

also awkward as it states what is required only to then add 29

b.

The safety evaluation for the Topical Report WCAP-16793-NP provides exceptions. It would be clearer to state the requirement guidance to evaluate the effect of debris in recirculating fluid on long-term explicitly as in "For cases in which conservative deterministic cooling, including in-vessel effects such as blockage of flow clearances values that overestimate the CDF and LERF.. "Further the through fuel assemblies. The topical report defines in-vessel debris load limits requirement should be to illuminate the uncertainty which may (i.e., 15 grams (g) of fiber per fuel assembly as transported and accumulated be done by propagation of uncertainties or by sensitivity during a hot-leg break) below which testing has demonstrated that long-term studies.

core cooling is not impeded. Licensees may use the debris load limits described in WCAP-1 6793 or other NRC-approved values as acceptance 25

ENCLOSURE 1 GSI-191. Option 2b Comments ITEM DG-1 322 LANGUAGE COMMENTS limits for in-vessel debris loading without additional justification.

c.

The licensee should address the potential for boric acid precipitation in its analysis.

d.

The licensee should develop a model for debris penetration effects, and clearly identify how the model would be used in the estimate of CDF and LERF. In particular, the licensee should properly account for the fraction of the debris-carrying flow passing through the core in mass balance computations for the amount of in-vessel debris accumulation. A fraction of the flow that carries debris may be considered not to contribute to debris buildup inside the pressure vessel, such as the flow discharged through breaks or through the containment spray system. The licensee should provide a technical basis for" any fraction of the flow considered not to contribute to in-vessel debris buildup. However, note that the debris returned to the pool may pass through the strainer again.

e.

Chemical effects should be considered in the evaluation of the effects of debris penetration.

f.The licensee should evaluate the validity of the penetrated debris effects model, relying, for example, on tests, empirical data, and analogies of.

g.

The licensee should identify relevant data and model uncertainties, and propagate those uncertainties into the integrated model to compute the total CDF and LERF. The licensee should describe where in its analysis conservative deterministic approaches (i.e., approaches that tend to overestimate the CDF and LERF) from NRC-approved guidance are used.

Doing so will identify areas where propagation of uncertainty is not required.

Comment 1:* The licensee should combine the submodels for

14. The licensee should combine the submodels for the debris source thdersouctrmdbisraprsrinroel term, debris transport, strainer model, chemical interactions with debris, chemical interactions with debris, debris strainer penetration, debris strainer penetration, and in-vessel effects into the integrated model and in-vessel effects into the integrated model with the goal of 30 with the goal of computing failure probabilities in the modified PRA model computing failure probabilities in the modified PRA model to 30 to evaluate debris effects (implemented in Paragraph C.15 of this RG).

evaluate debris effects. For the simplified approach this These submodels were discussed in this RG's paragraphs C.5 to C.9 for section and the remainder still apply. This should allow for deletion of scenarios screened as described in comment Item the simplified approach and paragraphs C.5 to C.15, excluding C.9, 24 above and should also define the term "model" to include otherwise.

bounding assumptions where appropriate or desired when 26

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM [

DG-1 322 LANGUAGE JCOMMENTS

a.

The integrated model should be structured to allow for propagation of relevant parameter and model uncertainties identified in Paragraphs C.5 to 0.13 of this RG. A model implementing the Monte Carlo method or other suitable sampling approach could be considered for the propagation of parameter uncertainty.

The sampling approach should implement variance reduction techniques, such as stratified sampling, to ensure that relevant distribution tails are properly considered and sampled.

b.

Uncertainties that can be represented as distributions (i.e., probability density functions or. probability mass functions) in input parameters should be identified and properly justified. Effort should be aimed at selecting distributions that properly reflect the state of knowledge of each parameter so that CDF and LERF can be calculated accurately.

c.

Inputs to the integrated model should be consistent with inputs to the modified PRA discussed in Paragraph 0.15 of this RG, and sampled in the integrated model and the modified PRA model consistent with any known statistical correlation or dependence between the uncertainty distributions.

d.

Initiating event frequencies should be represented by distributions to be used as an input to the modified PRA and the integrated model.

(1) Information in NUREG-1829, "Estimating Loss-of-Coolant Accident (LOCA)

Frequencies through the Elicitation Process" (Ref. 19) is acceptable to estimate LOCA frequencies in general for piping and non-piping passive systems in PWRs and BWRs. The licensee should ensure that HELB locations identified in paragraph 0.6 of this RG are consistent with the locations assumed in NUREG-1829. The licensee should confirm that the NUREG-1829 values are applicable to its plant.

(2) NUREG-1 829 provides different summary tables for 0.05, 0.5, and 0.95 LOCA frequency quantiles and mean frequencies derived using different approaches to aggregate elicitations from the individual elicited experts.

NUREG-1 829 does not advocate any specific aggregation method. The licensee should select frequencies that would not underestimate the mean CDF, LERF, ACDF, and ALERF as compared to alternative methods. The NRC finds that the LOCA frequencies from NUREG-1 829 derived using the arithmetic mean aggregation and mixed distribution aggregation methods are acceptable. Use of LOCA frequencies derived from properly defined and characterized.

The remainder of section 0.14 includes five areas for treatment by "propagation of relevant parameter and model uncertainties" and by "treatment of distributions" which ignores the ability to demonstrate through the simplified approach conformance to RG 1.174 and through sensitivities robustness of that conformance without statistical analysis.

This should be corrected to support the simplified approach.

14 would read we have to modify PRA model, Is this true in the simplified approach?

Would a modified PRA be required if the plant PRA is not used to determine ACDF and ALERF?

d.(1) - How to confirm NUREG-1 829 values are applicable to the plant?

Section C.14.d.4 states: "Statistical distributions chosen to represent the uncertainty about parameters in the integrated model and in the modified PRA should preserve the mean values of the initiating event frequencies from original source documents, such as NUREG-1829." The words "if used" should follow "statistical distributions".

Comment 2: See comment for 0.7 about consolidating all uncertainty analysis requirement into a single section in this RG.

Clarify all the requirements in 14.d starting with what is required to confirm NUREG-1 829 values are applicable.

Are the other analyses only required if LOCA frequencies are not derived from NUREG-1 829?

Comment 3: The guidance provided in this section does not apply to a simplified approach. An integrated model may not be necessary to calculate ACDF since this value will be principally derived from the NUREG-1 829 break frequencies.

There will be a submodel evaluation tool (CASA or CASA 27

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM [DG-1 322 LANGUAGE COMMENTS alternative aggregation methods (such as geometric mean aggregation) should be justified by the licensee, and alternatives should be considered in uncertainty analyses. The licensee should demonstrate that conclusions would not be significantly different from the conclusions reached through the use of alternative aggregation methods by comparing the licensee's conclusions to the results obtained using arithmetic or mixture distribution aggregation.

(3) If the information from NUREG-1829 is not used to estimate LOCA frequencies, the licensee should justify LOCA and non-LOCA initiating event frequencies. The licensee should evaluate the impact of alternative selection of frequency ranges or distributions for initiating events on CDF and LERF, and demonstrate that conclusions would not be impacted by

• alternatives.

(4) Statistical distributions chosen to represent the uncertainty about parameters in the integrated model and in the modified PRA should preserve the mean values of the initiating event frequencies from original source documents, such as NUREG-1829.

e.

The licensee should consider failure modes of SSCs, identified in Paragraph C.2 of this RG, and the corresponding failure probabilities. The licensee's analysis should be consistent with the failure of piping and non-piping (e.g.,

valve bodies, pump casings, manways, control rod penetrations, etc.) passive systems considered in NUREG-1 829.

(1) The licensee should provide a technical basis for allocating plant wide LOCA frequencies to individual locations (e.g., pipe weld). One acceptable approach is to use information from the licensee's In-Service Inspection (ISI) program. In general, the LOCA frequency assigned to each location should be informed by the known degradation mechanisms at that location.

(2) Assumptions that are made when allocating plant-wide LOCA frequencies to individual locations (e.g., relative likelihood of a complete rupture of a small pipe compared to an equivalent size opening in a larger pipe) should be identified and their impact on CDF and LERF should be quantified.

type software, or spreadsheet) that will manipulate the inputs and compare them against the failure criteria to establish the bounding quantities of debris that results in success/failure for both the strainer and in-vessel. Once this debris quantity is known, the tool will then determine the break size that results in the success/failure transition point. As stated, this model does not require an integrated model.

14.a - Uncertainty determination and propagation is not necessary for those parameters that are established under conservative and bounding deterministic methodologies.

14.b - Parameter and uncertainty distribution may not be required for the simplified approach.

14.c - The PRA model may not have to be modified for the simplified approach since ACDF and ALERF can be calculated outside the PRA model.

14.d - See previous comments on this section with regard to the "integrated model".

14.e.(2) - As far as I know, there is no technical basis for assuming a smaller break in a larger pipe and then determining the appropriate break frequency. This is outside the basis for the frequencies identified in NUREG-1 829. The overall effort would be better justified, and less complicated, if only DEGBs were considered.

Comment 4: Skipping to here from C.9 for the simplified approach:

C.14 needs to more clearly distinguish between the integrated model and the simplified approach. For instance, the STP RoverD approach does not involve inputs into the plant PRA.

Suggest splitting C.14 into separate sections for the integrated approach and the simplified approach. Having ser~arate sections for simp~lified and intecirated approach

-I 28

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM DG-1 322 LANGUAGE COMMENTS should also be considered.

Comment 5: For simrplified approach, PRA model would not be needed for calculation of CDF. Maybe for LERF though...

Comment 6: This section and the remaining sections, which apply to the simplified approach, should allow for deletion of scenarios screened as described in comment 24 above and should also define the term "model" to include bounding assumptions where appropriate or desired when properly defined and characterized.

The remainder of section 0.14 includes five areas for treatment by "propagation of relevant parameter and model uncertainties" and by "treatment of distributions" which ignores the ability to demonstrate throughthe simplified approach conformance to RG 1.174 and through sensitivities robustness of that conformance without statistical analysis.

This should be corrected to support the simplified approach.

C.14.d.(1) Comments:

a. This item presumes that the PRA model initiating events used in the GS1-191 application are aggregated LOCA frequencies as normally assumed in base-line PRAs and given in NUREG-1 829. If the PRA model initiating events are defined as LOCAs at specific locations, there is no guidance given in NUREG-1 829 how to allocate the LOCA frequencies and their uncertainties to specific components and specific locations. Hence there is interplay between item 6d and 6e in this draft RG. NUREG-1829 acceptability is tied to this modeling approach.

b. It is not clear what is meant by "HELB locations" as this term is not defined or used in 0.6. If HELB is interpreted to apply to feed-line and steam-line breaks, such initiating events are not included in NUREG-1829. Current definitions of HELB would seem to be apply to both LOCA sensitive components as well as feed-water and main steam line components. If feed line and steam line breaks are intended to be included, no criteria are given for 29

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM DG-1 322 LANGUAGE JCOMMENTS initiating event frequency bases.

c. It is not clear how one can confirm that NUREG-1829 values are applicable to a specific plant based on the information available in NUREG-1829. First it is not clear how or whether plant-to -plant variability in LOCA frequencies was considered or accounted for in NUREG-1829. Does the uncertainty presented in the NUREG-1829 LOCA frequencies include both plant-to-plant variability and within plant uncertainty? For example the Base Case analyses that were performed to inform the expert elicitation for PWRs were based on a 3-loop Westinghouse PWR design with a specific configuration.

How different designs such as 2-loop and 4-loop Westinghouse PWRs, CE and B&W designs that employ different loop configurations and piping materials, and differences in pipe sizes, damage mechanisms, etc. from plant to plant were accounted for in the reference which would be important to prepare a justification for a specific plant. Given that situation, how one can in principle justify that the NUREG-1829 values are applicable to any specific plant is difficult to envision. The problem is that NUREG-1 829 provides no justification that the results are applicable to any specific plant.

C.14.d.(2) Comments:

a. Although NUREG-1 829 includes aggregated LOCA frequency results using both the geometric mean and mixture distribution method, in the discussion of these results, NUREG-1 829 appears to give more weight to the geometric mean method. As noted in this discussion, a major problem with the mixture distribution method is that the results are too highly skewed to the input from the expert with the most pessimistic results. In addition, the results of using the geometric mean method are more prominently displayed in the executive summary and in the results section of the report. Hence the statement that NUREG-1 829 does not advocate any specific aggregation method appears to be somewhat at odds with how this issue is presented in the report. Given this, the statement 30

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM DG-1 322 LANGUAGE COMMENTS that the NRC finds the mixture distribution method acceptable and then asks the user of the RG to justify use of geometric mean is questioned. By what basis has the staff shifted it position from NUREG-1 829 which at best was neutral on the issue? Is it just because the NUREG-1829 results for the mixture distribution have larger values?

b. The uncertainty quantiles referred to in NUREG-1 829 are applicable to the uncertainties in the aggregated LOCA frequencies. What is more important for this application is the uncertainty in the LOCA frequencies at specific locations that may be the most important for debris-induced core damage frequency. Uncertainties in location specific LOCA frequencies would be expected to be much greater but these are not available in NUREG-1 829.

Section C.14.d.4 states: "Statistical distributions chosen to represent the uncertainty about parameters in the integrated model and in the modified PRA should preserve the mean values of the initiating event frequencies from original source documents, such as NUREG-1829." The words "if used" should follow "statistical distributions". Also, this requirement is framed in a manner that presumes the PRA model initiating events are aggregated LOCA frequencies and not location dependent LOCA frequencies.

Comment 7: Consider rewording paragraph C.14.d.1 as the term HELB has a very specific definition as a Hazard Event which would fall outside DBA mitigation requirements Comment 1: Clarify if the intent is to permanently incorporate

15. The licensee should estimate the change in risk attributable to debris, into the plant's baseline model of record or just in the PRA

a.

The licensee should make modifications to the baseline PRA model model used in the analysis to calculate delta CDF and LERF.

31 (i.e., the PRA model that assumes any effects of debris are negligible),

See comment for C.7 about consolidating all uncertainty 31

~consistent with Paragraph C.3, to perform the calculation of the risk aayi eurmn noasnl eto nti G

(CDF and LERF) for the as-built and as-operated nuclear power plant.

Comment 2: Changes to the base PRA model may not be

b.

Licensees should use commonly accepted methods and approaches to necessary for the simplified approach. The guidance

_________________________________________________pr'ovided in this section may be necessary for a more in-depth 31

ENCLOSURE 1 GSI-19I Option 2b Comments ITEM DG-1 322 LANGUAGE 3COMMENTS implement changes to the PRA model for the debris risk assessment.

These methods and approaches should be consistent with the guidance in RG 1.200 to the extent applicable. For example, if new operator actions are added to the PRA model to account for debris, the human reliability analysis would typically be the same as is used in the base PRA model. Similarly, event tree and fault tree changes made to account for debris would typically use the same approach as used in the peer-reviewed PRA model. The changes made and methods employed to implement those changes should be well described in the license amendment request.

c.

Changes to the PRA should include revisions of failure frequencies and probabilities and reliability data in general to account for the presence of debris.

d.

New human failure events (HFEs) should be added to the model as appropriate. Debris effects on the HFEs in the PRA model should be determined and human error probabilities adjusted accordingly. The dependency among multiple human errors in the same accident sequence, including new HFEs added to the model to account for debris presence, Should be assessed and accounted for in the quantification of the PRA model.

e.

Inputs to the modified PRA should be consistent with inputs and information used by the integrated model.

(1) Common input distributions should be consistently sampled in the modified PRA and in the integrated model.

(2) Common information of the modified PRA and the integrated model should be consistently treated, including the use of correlations where needed.

f.Nuclear power plant states and configurations not explicitly treated in the modified PRA or in the integrated model, and which are not excluded following the screening procedure in Paragraph C.1 of this RG, should be assumed to lead to core damage. The contribution to risk-informed approach. There should be some distinguishing between the two approaches.

Comment 3: Because the simplified approach approximates CDF with LOCA initiating event frequency, it would not be appropriate to use that value to modify the baseline PRA. The simplified calculation serves only to show that the CDF is within RG 1.174 acceptance criteria, but does not establish an actual LOCA CDF.

Comment 4: For simplified approach, PRA model would not be needed for calculation of CDF. Maybe for LERF though...

Comment 5:

1. This entire section presumes that the work will be done as a modification to a full PRA model. As discussed in comment 17 this may not be necessary or even the best approach. The requirement should relate to the use of baseline PRA information and models in a manner consistent with RG 1.200.

2. Section C.1 5.c states: "Changes to the PRA should include revisions of failure frequencies and probabilities and reliability data in general to account for the presence of debris." This is only required to the extent that such changes are necessary to achieve a high quality analysis of the delta CDF and LERF.

3. Section C.15.g.2 suggests that mean frequencies should be generated from propagation of parametric uncertainties Using the PRA and integrated model for comparison to RG 1.174 risk acceptance guidance.

A statement should be added that sensitivity studies are acceptable in lieu of uncertainty propagation where it can be demonstrated that there would be no change in the conclusion that the acceptance guidance still would be met.

32

ENCLOSURE 1-GSI-191 Option 2b Comments ITEM DG-1 322 LANGUAGE COMMENTS the CDF and LERF for these unaccounted states and configurations should be quantified.

g.

The modified PRA, together with the integrated model, should be used to quantifY the mean values of CDF and LERF, accounting for debris effects, and compared to risk regions in Figures 4 and 5 of RG 1.174.

(1) The ACDF and ALERF should be computed with respect to risk of the plant assuming that debris effects are negligible. The CDF and LERF on the horizontal axis on Figures 4 and 5 of RG 1.174 should be interpreted as the total risk estimates for the plant as described in Section 2.4 of that regulatory guide.

(2) The mean value resulting from a propagation of parametric uncertainty in the PRA model quantification is the appropriate point-estimate for comparison to the RG 1.174 risk acceptance guidelines.

(3) The mean values of CDF/ACDF and LERF/ALERF should meet the risk acceptance guidelines of Figures 4 and 5, respectively, of RG 1.174.

16. The licensee should provide a summary description of the plant Comment: 16.b, 16.c-As discussed previously, an response to debris. The objective of this summary is to ensure that the integrated model may not be required for the simplified overall model produces reasonable results and that there are no counter-approach.

intuitive or non-physical modeling artifacts.

a.

The licensee should identify key aspects of the plant that limit the magnitude of the CDF and LERF when accounting for effects of debris, such as the following:

32 (1) frequency of events that could produce significant amounts of debris amount of debris produced (2) resiliency of strainer system to failure under the presence of debris (3) debris filtration by strainers (4) resiliency of the system to provide adequate core cooling under debris presence (5) resiliency of the system to limit large early releases 33

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM DG-1 322 LANGUAGE COMMENTS (6) alternate flow paths for cooling of the core, if credited

b.

The identification of key aspects of the plant should be complete and consistent with other paragraphs in this RG, including the submodels in Paragraphs 03.5 to 03.13, the integrated model in Paragraph 03.14, and the modified PRAmi Paragraph C3.15.

c.

The licensee should summarize the technical basis of the identified key aspects of the plant.

(1) The extent of the technical basis should be commensurate with the risk significance of the identified key aspects and the associated uncertainties.

(2) The summary technical basis should be consistent with this RG including the submodels in Paragraphs 0.5-0C.13, the integrated model in Paragraph 0.14, the modified PRA in Paragraph C.15,.and their uncertainties.

(3) The technical basis should describe any action taken for prevention (e.g.,

removal of fiber and insulation, increase in strainer areas to capture debris), and how key aspects under Paragraph C.16.a of this RG would mitigate debris effects. Relevant operator actions to reduce CDF and LERF should be described, such as hot leg switchover and securing of containment spray system pumps. The description should account for.

relevant plant states and operating SS~s.

(4) The licensee should summarize key aspects of the analysis where margins are applied to arrive at conservative estimates of the ODE and LERF, such as strainer failure criteria, in-vessel failure criteria, zones of influence to estimate debris sources, and tests conditions that would overestimate detrimental effects.

17. Section C.2.1.1 of RG 1.174 provides seven elements that can be No Comment used to demonstrate consistency with the defense-in-depth philosophy.

For the risk-informed evaluation of debris effects, additional guidance for 33 each element is provided below.

a.

Reasonable balance should be preserved among prevention of core damage, prevention of containment failure or bypass, and mitigation of consequences of an offsite release. The licensee should address the impact of debris-related 34

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM j DG-1 322 LANGUAGE COMMENTS failure modes on the EGGS (prevention of core damage), on the containment systems (prevention of containment failure), and on emergency preparedness (con~sequence mitigation). Examples of defense-in-depth measures can be found in a paper by the Nuclear Energy Institute, "Example Pressurized Water Reactor Defense-in-Depth Measures For GS1-191, PWR Sump Performance,"

(Ref. 20). The assessment should consider the impact of debris on the availability and reliability of each level of defense, as well as the aggregate impact.

b.

There should not be an over-reliance on programmatic activities to compensate for weaknesses in plant design. The licensee should evaluate programmatic activities relevant to the effects of debris including, but not limited to, design controls to limit debris, the ISI program, plant personnel training, the reactor coolant system leak detection program, and containment cleanliness inspection activities.

c.

System redundancy, independence, and diversity should be preserved commensurate with the expected frequency of challenges, consequences of failure of the system, and associated uncertainties in determining these parameters. If systems that could be impacted by debris are modified, added, or removed, the licensee should address the effect of such changes on redundancy, independence, and diversity. Absent such system changes, the licensee may conclude that this element of defense-in-depth is addressed.

Note that common cause failures are addressed by the next element so do not have to be addressed under this one.

d.

Defenses against potential common cause failures should be preserved and the potential for the introduction of new common cause failure mechanisms are assessed and addressed. The licensee should assess the impact of debris on inter-system (e.g., among low-head and high-head injection systems) and intra-system (e.g., among trains of a given system) availability and reliability. The licensee should justify, qualitatively or quantitatively, that any increase in common-cause failure rates across systems impacted by debris is very small compared to other failure rates that are not debris-related.

e.

Independence of barriers should not be degraded. As stated in RG 1.174, a barrier is a layer of defense against core damage, containment failure, or bypass, and not necessarily a physical barrier. The licensee should provide an evaluation describing a realistic plant response to each debris-related 35

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM DG-1 322 LANGUAGE *COMMENTS failure mode identified in Paragraph 0.2 of this RG. This evaluation should assume that a debris-related failure mode has occurred (i.e., a corresponding barrier has failed) and should identify the remaining plant equipment or mitigative measures (i.e., remaining barriers) that can be independently relied upon. For example, if strainer mechanical collapse occurred because of debris, the ECCS may not be sufficient to prevent core damage. The next barrier would be containment structures. Therefore, the licensee should demonstrate, qualitatively or quantitatively, that reasonable confidence exists that the containment would remain as an effective independent barrier for these scenarios.

Examples of defense-in-depth or mitigative measures can be found in "Example Pressurized Water Reactor Defense-in-Depth Measures For GSI-191, PWR Sump Performance,'

Equipment (e.g., containment fan coolers) and operator actions that would not be compromised by this debris-related failure mode should be described and credited as contributing to barrier independence. When performing this step, licensees may take into account how plant conditions vary over time. For example, when evaluating containment performance following assumed strainer structural failure, licensees may assume thermal-hydraulic conditions consistent with the time that strainer failure would reasonably be assumed to occur.

f.Defenses against human errors should be preserved. The licensee should discuss any operator actions for the plant with debris that Would not exist in a debris-free plant. The feasibility of these operator actions and any effect on non-debris operator actions should be discussed (e.g., any impact on crew workload). The licensee should justify that any human errors, in general, will not be significantly more likely compared to the clean plant.

g.

The intent of the plant's design criteria should be maintained. The licensee should confirm that no debris-related failure could completely disable multiple layers of defense between the fission product source term and the public.

36

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM DG-1 322 LANGUAGE COMMENTS

18. The licensee should demonstrate that sufficient safety margins are No Comment maintained when debris is present in the as-built and as-operated plant.

This demonstration may be qualitative or quantitative and should address safety margins associated with both the design-basis aspects (e.g., effect on SS~s, flow rates, temperatures, pressures) as well as with any realistic assumptions used in the integrated analysis. In a fundamental sense, 34 margin is the difference between some limit and a value that may be attained by a parameter. Assumptions about the limit and actual parameter values should be consistent with licensing-basis calculations unless otherwise justified. The demonstration of safety margins should be consistent with guidance at Paragraph C. 1 6.c (4). For example, if the licensing basis calculations use a given value for the required NPSH, then the integrated model should also use this value, or a justification should be provided if a different value is considered.

19. The licensee should ensure that the risk assessment for the evaluation No Comment of debris was performed under a QA program that meets the guidance of RG 1.174, Section 5. The use of the pertinent quality assurance 35 requirements of Appendix B to 10 CFR Part 50 as set forth in RG 1.174 is justified because the risk-informed analysis of debris (including the integrated model and supporting computations) is needed to demonstrate that the design of safety-related SS~s meets NRC requirements.

37

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM [DG-1 322 LANGUAGE JCOMMENTS

20. Licenses selecting the risk-informed alternative in 50.46c(e) must demonstrate that the risk attributable to debris is small and that defense in depth and safety margins are maintained. Consistent with RG 1.174, the licensee should develop an implementation and monitoring program that will ensure the long term validity of these conclusions. This program should provide reasonable assurance that future planned or unplanned changes to the plant (e.g., modifications, discovery of additional latent debris) or changes to the PRA1 (e.g., change in LOCA frequencies) are analyzed to ensure that the original conclusions from the LAR remain valid.

Consistent with application-specific guidance for other risk-informed initiatives, the implementation and monitoring program may be partially or fully comprised of existing licensee programs. Licensees wishing to credit existing programs should describe how these programs are suited (or have been modified) to account for the unique challenges of calculating the portion of CDF and LERF attributable to debris. For example, many programs used to track risk are based on equipment availability and reliability. Such programs would not generally be suited to evaluating -

for example - the discovery of a large quantity of degraded coatings that could contribute to the source term. Licensees should also describe updates to existing programs that will prevent or mitigate addition of known problematic debris sources into containment. For example, if a licensee's evaluation credits the removal of Marinite, the licensee should describe how plant work control practices prevent its future introduction.

Consistent with RG 1.174, the results (e.g. tracking and trending data) of this monitoring program should be retained onsite for inspection.

Consistent with 50.46c, licensees are not required to report these results to the NRC unless it is determined that the acceptance criteria in 50.46c(e)(1) are no longer met. In this case, existing licensee programs for reporting (e.g., 10 CFR 50.72, 10 CFR 50.73) may be used.

1 In this context, "PRA" also includes any complementary analyses (e.g.,

debris evaluation model, human reliability analysis) that are used to calculate the increase in risk attributable to debris.

Comment: Does this imply that we will need a process to periodically monitor performance and confirm that analysis inputs/assumptions and PRA results are still valid?

36 3

ENCLOSURE 1 GSI-19I Option 2b Comments ITEM DG-1 322 LANGUAGE COMMENTS

0. IMPLEMENTATION The purpose of this section is to provide information on how applicants and No Comment licensees2 may use this guide and information regarding the NRC's plans for using this regulatory guide. In addition, it describes how the NRC staff complies with 10 CFR 50.109, "Backfitting" and any applicable finality provisions in 10 CFR Part 52 "Licenses, Certifications, and Approvals for Nuclear Power Plants."

37 2 In this section, "licensees" refers to licensees of nuclear power plants under 10 CFR Parts 50 and 52; and the term "applicants," refers to applicants for licenses and permits for (or relating to) nuclear power plants under 10 CFR Parts 50 and 52, and applicants for standard design approvals and standard design certifications under 10 CFR Part 52.

Use by Licensees Licensees may voluntarily3 use the guidance in this document to No Comment demonstrate compliance with the underlying NRC regulations. Methods or solutions that differ from those described in this regulatory guide may be deemed acceptable if they provide sufficient basis and information for the NRC staff to verify that the proposed alternative demonstrates compliance with the appropriate NRC regulations.

Licensees may use the information in this regulatory guide for actions 38 which do not require NRC review and approval such as changes to a facility design under 10 CFR 50.59, "Changes, Tests, and Experiments,"

that do not require prior NRC review and approval. Licensees may use the information in this regulatory guide or applicable parts to resolve regulatory or inspection issues.

3 In this section, "voluntary" and "voluntarily" means that the licensee is seeking the action of its own accord, without the force of a legally binding 39

ENCLOSURE 1 GSI-191 Option 2b Comments DG-1 322 LANGUAGE COMMENTS action.

Use by NRC Staff 39 The NRC staff does not intend or approve any imposition or backfitting of the guidance in this regulatory guide. The NRC staff does not expect any existing licensee to use or commit to using the guidance in this regulatory guide, unless the licensee makes a change to its licensing basis. The NRC staff does not expect or plan to request licensees to voluntarily adopt this regulatory guide to resolve a generic regulatory issue. The NRC staff does not expect or plan to initiate NRC regulatory action which would require the use of this regulatory guide. Examples of such unplanned NRC regulatory actions include issuance of an order requiring the use of the regulatory guide, requests for information under 10 CFR 50.54(f) as to whether a licensee intends to commit to use of this regulatory guide, generic communication, or promulgation of a rule requiring the use of this regulatory guide without further backfit consideration.

Comment: SECY-12-0093 required all licensees to report to the NRC staff the chosen methodology that the licensee would voluntarily adopt to address the Generic Regulatory Issue described in GS1-191 and GL 2004-02. Please explain how this does not meet "Backfit" requirements per 10OCER50.1I09(a)(1).

During regulatory discussions on plant specific operational issues, the staff No Comment may discuss with licensees various actions consistent with staff positions in this regulatory guide, as one acceptable means of meeting the underlying NRC regulatory requirement. Such discussions would not 40 ordinarily be considered backfitting even if prior versions of this regulatory guide are part of the licensing basis of the facility. However, unless this regulatory guide is part of the licensing basis for a facility, the staff may not represent to the licensee that the licensee's failure to comply with the positions in this regulatory guide constitutes a violation.

40

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM DG-1 322 LANGUAGE COMMENTS If an existing licensee voluntarily seeks a license amendment or change No Comment and (1) the NRC staff's consideration of the request involves a regulatory issue directly relevant to this new or revised regulatory guide and (2) the specific subject matter of this regulatory guide is an essential consideration in the staff's determination of the acceptability of the licensee's request, 41 then the staff may request that the licensee either follow the guidance in this regulatory guide or provide an equivalent alternative process that demonstrates compliance with the underlying NRC regulatory requirements. This is not considered backfitting as defined in 10 CFR 50.109(a)(1) or a violation of any of the issue finality provisions in 1 0 CFR Part 52.

If a licensee believes that the NRC is either using this regulatory guide or No Comment requesting or requiring the licensee to implement the methods or processes in this regulatory guide in a manner inconsistent with the discussion in this Implementation section, then the licensee may file a backfit appeal with the NRC in accordance with the guidance in NRC Management Directive 8.4, "Management of Facility-Specific Backfitting and Information Collection" (Ref. 21), and in NUREG-1409, "Backfitting Guidelines," (Ref. 22).

42 41

ENCLOSURE 1 GSI-191 Option 2b Comments iTEM "DG-1 322 LANGUAGE COMMENTS

• REFERENCES 4 43

1. Title 10, Part 50, Section 46c, of the Code of Federal Regulations (10 CFR 50.46c), "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors."

2. Title 10, Part 50, of the Code of Federal Regulations (10 CFR 50),

"Domestic Licensing of Production and Utilization Facilities."

3.

10 CFR 50 Appendix A, "General Design Criteria for Nuclear Power Plants," Criterion 15, "Reactor coolant system design."

4. 10 CFR 50 Appendix A, "General Design Criteria for Nuclear Power Plants," Criterion 35, "Emergency core cooling."

5.

U.S. Nuclear Regulatory Commission (NRC), Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,"

Washirngton, DC.

6. NRC, RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Washington, DC, March 2009.

7. NRC, RG 1.82, "Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident," Revision 4, March 2012, Washington, DC.

8. NRC, "Staff Requirements - SECY-12-0093 - Closure Options for Generic Safety Issue - 191, Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance." Washington, DC, December, 14, 2012.

9. NRC, "Staff Requirements - SECY-12-0034 - Proposed Rulemaking -

10 CFR 50.46c: Emergency Core Cooling System Performance during Loss-of-Coolant Accidents (RIN 3150- AH42)." Washington, DC, January, 7, 2013.

10. NRC, RG 1.82, "Sumps for Emergency Core Cooling and Containment Spray Systems," Revision 0, June 1974, Washington, DC.

11. NRC, Generic Letter (GL) 85-22, "Potential for Loss of Post-LOCA Comment: Add the following document (cited on under Section B on page 3) to the list of references and please include the appropriate ADAMS number:

NRC, GS1-191, Assessment of Debris Accumulation on PWR Sump Performance, (ADAMS No ML xxxxxxxxx) 42

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM JDG-1 322 LANGUAGE COMMENTS Recirculation Capability Due to Insulation Debris Blockage" Washington, DC, December 3, 1985. (ADAMS No. ML ML031150731)

12. NRC, GL 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors," Washington, DC, September 13, 2004.

13. NRC, "NRC Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Coatings Evaluation," Washington, DC. March 2008.

(ADAMS No. ML080230462)

14. NRC, Report NEI 04-07, "Pressurized Water Reactor Sump Performance Evaluation Methodology. Volume 2 - Safety Evaluation by the Office of Nuclear Reactor Regulation Related to NRC Generic Letter 2004-02, Revision 0, December 6, 2004," Washington, DC, December 2004. (ADAMS No. ML050550156)

• 15. Westinghouse and NRC. "Final Safety Evaluation for Pressurize Water Reactor Owners Group Topical Report WCAP-16793-NP, Revision 2,

'Evaluation of Long-Term Cooling Considering Particulate Fibrous and Chemical Debris in the Recirculating Fluid," Washington, DC. July, 2013.

(ADAMS No. ML13239A111)

16. Westinghouse and NRC. "Final Safety Evaluation by the Office of Nuclear Reactor Regulation, Topical Report WCAP-16530-NP-A

'Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GS1-1 91,"'" Washington, DC. March, 2008. (ADAMS Nos.

ML081150383 and ML101230629)

17. NRC, "NRC Staff Review Guidance Regarding Generic Letter 2004-02*

Closure in the Area of Plant-Specific Chemical Effect Evaluations,"

Washington, DC..March 2008. (ADAMS No. ML080380214)

18. NRC, "Safety Evaluation by the Office of Nuclear Reactor Regulation, Topical Report (TR) WCAP-16406-P, Revision 1, "Evaluation of Downstream Sump Debris Effects in Support of GsM-91" Pressurized Water Reactor Owners Group, Project No. 694. Washington, DC.

December 20, 2007. (ADAMS No. ML073520295)

19. NRC, NUREG-1 829, "Estimating Loss-of-Coolant Accident (LOCA) 43

ENCLOSURE 1 GSI-191 Option 2b Comments ITEM DG-1 322 LANGUAGE COMMENTS Frequencies Through the Elicitation Process," Washington, DC. April 2008.

(ADAMS No. ML ML080630013)

20. Nuclear Energy Institute (NEI), "Example Pressurized Water Reactor Defense-in-Depth Measures For GS1-191, PWR Sump Performance,"

Washington, DC. March 2012. (ADAMS No. ML120730660)

21. NRC, Management Directive 8.4, "Management of Facility-Specific Backfitting and Information Collection," Washington, DC. October 9, 2013.

22. NRC, NUREG-1409, "Backfitting Guidelines," Washington, DC. July 1990.

4 Publicly available NRC-published documents are available electronically through the NRC Library on the NRC's public Web site at http:llwww. nrc.gov/reading-rm/doc-collections/ and through the NRC's Agencywide Documents Access and Management System (ADAMS) at http:/lvwvw.nrc.govfreading-rm/adams.html. The documents can also be viewed online or printed for a fee in the NRC's Public Document Room (PDR) at 11555 Rockville Pike, Rockville, MD. For problems with ADAMS, contact the PDR staff at 301-415-4737 or 800-397-4209; fax 301-41 5-3548; or e-mail to pdr.resource@nrc.gov.___________________________

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