Text

Supplement to Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program
	ML15272A443
Person / Time
Site:	Harris
Issue date:	09/29/2015
From:	Waldrep B; Duke Energy Progress
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	HNP-15-083, TAC MF6583
	Download: ML15272A443 (94)
	v • d • e

Benjamin C. Waldrep Vice President Harris Nuclear Plant 5413 Shearon Harris Road New Hill NC 27562-9300 919.362.2000 10 CFR 50.90 September 29, 2015 Serial: HNP-15-083 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Shearon Harris Nuclear Power Plant (HNP), Unit 1 Docket No. 50-400/Renewed License No. NPF-63

Subject:

SUPPLEMENT TO HARRIS NUCLEAR PLANT APPLICATION FOR TECHNICAL SPECIFICATION CHANGE REGARDING RISK-INFORMED JUSTIFICATION FOR THE RELOCATION OF SPECIFIC SURVEILLANCE FREQUENCY REQUIREMENTS TO A LICENSEE CONTROLLED PROGRAM

References:

1. Duke Energy letter, Harris Nuclear Plant, Unit No. 1, Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program, dated August 18, 2015 (ML15236A254).

2. Nuclear Regulatory Commission letter, Shearon Harris Nuclear Power Plant, Unit 1 -

Supplemental Information Needed For Acceptance of Requested Licensing Action for Technical Specification Change Regarding Risk-Informed Justification For the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (TAC No. MF6583), dated September 18, 2015 (ML15259A435)

Ladies and Gentlemen:

In accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR Part 50.90), Application for Amendment of License, Construction Permit, or Early Site Permit, Duke Energy Progress, Inc. (Duke Energy) submitted a request for an amendment to the Technical Specifications (TS) for Shearon Harris Nuclear Power Plant, Unit 1 (HNP) (Reference 1).

Reference 2 specified that the NRC staff has reviewed this application and concluded that the information delineated in the enclosure to Reference 2 is necessary to enable the staff to make an independent assessment regarding the acceptability of the proposed amendment in terms of regulatory requirements and the protection of public health and safety and the environment. In response to the request for supplemental information provided in Reference 2, HNP is submitting the enclosed additional information to support acceptance review of Reference 1.

This information is provided in Enclosure 1.

This document contains no new Regulatory Commitments.

Please refer any questions regarding this submittal to John Caves at (919) 362-2406.

XV QXFOHDUUHJXODWRU\Commission KQS

SDJH

Ldeclare under SHQDOW\of perjury that the foregoing is true and correct. Executed on September 29,

VLQFHUHO\

Benjamin F ZDOGUHS Vice SUHVLGHQWShearon Harris Nuclear SRZHUSODQW HQFORVXUH

1. VXSSOHPHQWDOLQIRUPDWLRQNeeded for Acceptance of Requested Licensing Action cc: Mr. J . D. Austin, NRC VU Resident LQVSHFWRU KQS Mr. W. L. Cox, LLLSection Chief, N.C.

Ms. M. EDULOODVNRC SURMHFWManager, KQS Mr. V. M. Mccree, NRC UHJLRQDOAdministrator, Region LL

U.S. Nuclear Regulatory Commission HNP-15-083 Page 3 bc: Duncan Brewer John Caves David Cummings Jim Eayres Donald Griffith Tanya Hamilton Brian McCabe Sean OConnor Ryan Treadway Ben Waldrep HNP NSRB Paula Hamilton Nuclear Records

HNP-15-083 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400 / RENEWED LICENSE NO. NPF-63 SUPPLEMENT TO HARRIS NUCLEAR PLANT APPLICATION FOR TECHNICAL SPECIFICATION CHANGE REGARDING RISK- INFORMED JUSTIFICATION FOR THE RELOCATION OF SPECIFIC SURVEILLANCE FREQUENCY REQUIREMENTS TO A LICENSEE CONTROLLED PROGRAM Enclosure 1 Supplemental Information Needed for Acceptance of Requested Licensing Action (91 pages)

U.S. Nuclear Regulatory Commission Page 2 of 91 HNP-15-083, Enclosure 1 Shearon Harris Nuclear Power Plant, Unit No. 1 Docket No. 50-400 / Renewed License No. NPF-63 Supplement to Harris Nuclear Plant Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee-Controlled Program Supplemental Information Needed for Acceptance of Requested Licensing Action

1.0 INTRODUCTION

By letter dated September 18, 2015 (Agencywide Documents Access and Management System Accession No. ML15259A435), the U.S. Nuclear Regulatory Commission (NRC) requested that Duke Energy Progress, Inc. (Duke Energy) supplement the Harris Nuclear Plant (HNP) application for relocation of specific surveillance frequency requirements to a licensee-controlled program. Two insufficiencies related to the technical adequacy of the internal events probabilistic risk assessment (PRA) were identified, and supplemental information was requested to enable the NRC staff to begin its detailed review. Duke Energy's response is provided herein with the requested information on the PRA technical adequacy.

2.0 INSUFFICIENCIES The license amendment request (LAR) states that the HNP, Unit 1, internal events probabilistic risk assessment (PRA), "meets the requirements of the American Society of Mechanical Engineers/American Nuclear Society standard as endorsed by Regulatory Guide (RG) 1.200, Revision 2, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," at an appropriate capability category to support the HNP Surveillance Frequency Control Program [SFCP]." Based on the list of peer reviews included in Section 4.1 of Enclosure 2 to the LAR, the most recent peer review of the internal events, not including internal flooding, PRA was performed in 2007 against RG 1.200, Revision 1.

a. Provide the facts and observations (F&Os) from the peer reviews and gap assessments of the internal events PRA which are open, not met, or met a capability category I, and explanation of how the F&Os were dispositioned for this application.

b. Provide an overview of the changes in the internal events PRA that occurred after the 2007 peer review, and clarify whether any of these changes qualify as a PRA upgrade that would require a focused scope peer review.

3.0 DUKE ENERGY RESPONSE Duke Energy's responses to the NRC's request for supplemental information are provided in the following paragraphs. Paragraphs 3.a and 3.b refer to the insufficiencies identified in paragraphs 2.a and 2.b, respectively, as identified above.

a. The F&Os that were considered resolved but open following the 2007 peer review for Internal Events (i.e., the review of record) are provided in Table 1 of this Enclosure. There

U.S. Nuclear Regulatory Commission Page 3 of 91 HNP-15-083, Enclosure 1 are no F&Os that were not met or were met at capability category I. The resolutions for all twelve (12) of the F&Os in Table 1 were provided to and reviewed by the peer review teams for the HNP Fire PRA and the HNP Internal Flooding PRA described in the HNP application for relocation of specific surveillance frequency requirements to a licensee-controlled program. The impact of the disposition of each F&O for the HNP SFCP application is also provided in Table 1.

A summary of the assessment of the HNP Internal Events PRA against each of the eight (8) technical elements (i.e., high level requirements) of ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2, is provided in Tables 2 through 9, respectively, of this Enclosure.

Finally, the historical, closed internal events F&Os from peer reviews that were conducted prior to the 2007 peer review of record are provided in Table 10 of this Enclosure.

b. There have been four (4), formal model of record (MOR) revisions since 2007, and each upgrade was peer reviewed:

2007 Internal Events Update. The 2007 revision to the internal events model of record (MOR) incorporated findings and observations (F&O) resolutions for the April 2006 HNP PRA Self-Assessment in order to meet ASME/ANS Internal Events Standard (Revision 1) for Category II compliance. A peer review was performed to support the NFPA 805 license submittal. Major revisions included expansion of plant-specific data, Human Reliability Analysis (HRA) updates, and addition of new or more detailed heating, ventilation and air conditioning (HVAC) models for CSIP rooms, Switchgear rooms, and Emergency Service Water (ESW) pump rooms. The model revision also included addition of logic to address fire induced multiple spurious failures, developed in conjunction with HNP adoption of the NFPA 805 program for fire induced vulnerabilities. Other general updates to the model included an update to the initiating event frequencies, revision of the station blackout (SBO) induced seal LOCA, and Loss of offsite power (LOOP) recovery. Motor Control center modeling was improved to support the NFPA 805 LAR with the required level of detail. These updates were peer reviewed in the 2007 Industry Peer Review, and the resolved but open resolutions are provided in Table 1.

2010 Internal Events Update. The 2010 revision was to incorporate model changes and the final, as-built alternate seal injection (ASI) system. The major change for the 2010 update was the addition of the Alternate Seal Injection - Dedicated Shutdown Diesel Generator (ASI-DSDG) to the MOR. The installation of the ASI-DSDG modification provided a diverse and redundant power source for alternate seal injection and also to the emergency DC battery chargers, as described in HNP's NFPA 805 LAR. This reduced the effect of the 4-hour coping duration of the batteries by providing a means to supply DC power to the DC busses during SBO. The LOOP initiator was separated into plant, grid, switchyard and weather induced LOOPs, which allowed the model to apply recovery actions to the higher frequency events (plant and switchyard). Other changes related to de-energizing charging pump discharge header cross-connect valves, adding temporary air compressors, and updates from fire model were added to the 2010 model. This was not an upgrade and a peer review was not required for these revisions.

2013 Fire Update. The 2013 revision implemented resolutions for the previously identified conservatisms in the fire model. The main changes involved updating human failure events, dependency analysis, and recovery rule files. Other updates included additional walkdowns to identify fixed and transient ignition sources, crediting of implemented plant modifications,

U.S. Nuclear Regulatory Commission Page 4 of 91 HNP-15-083, Enclosure 1 and updates to fire frequency bin numbers to match the newest version of NUREG/CR-6850. This was not an upgrade and a peer review was not required for these revisions.

2014 Internal Flooding Only Upgrade. The 2014 revision was focused on the internal flooding portion of the HNP PRA model. A comprehensive flooding analysis was performed in order to meet the supporting requirements of the IFPRA portion of the PRA standard. The most noteworthy changes to the flooding model included the addition of spray effects and high energy line breaks (HELB) and their associated impacts on PRA equipment not previously included. The analysis resulted in the identification and quantification of flood-induced scenarios that were incorporated into the model. The IFPRA was peer reviewed in 2014 in accordance with ASME/ANS RA-Sa-2009 as endorsed by RG 1.200, Rev. 2.

Since 2007, the HNP PRA model has also been used to support PRA applications, and to assist in decision making with regard to the design, licensing, operation, and maintenance of the plant. Engineering changes, including equipment modifications, procedure changes, and plant performance (data), have been assessed and incorporated (per Duke Energy procedure) into the HNP model, as appropriate, in order to reflect the as-built and as-operated plant. These PRA maintenance activities do not qualify as PRA upgrades that would require an additional focused peer review. There have been no methodology changes or significant changes in scope or capability (other than the Fire and Internal Flood models described previously in this section) that impact the significant accident progression sequences. Therefore, no additional focused peer reviews of the Internal Events PRA have been required.

U.S. Nuclear Regulatory Commission Page 5 of 91 HNP-15-083, Enclosure 1 Table 1. Resolved but Open F&Os from 2007 Industry Peer Review of the HNP Internal Events PRA.

F&O # Finding Resolution 5b Impact A value of 0.33 was used for The use of a Jeffreys non-informative prior is the generally The use of the 0.33 number of failures when there accepted approach for addressing a lack of specific prior value was were no failures in preparing data when performing a Bayesian update. The referenced previously judged to generic failure data (page 7 of development of a generic database utilized in the HNP not be dispositioned Appendix B- component PSA is not based on a Bayesian approach and is not adequately database development). The related to the plant-specific update process where the (ML101750604). In 0.33 value was not based on generic data is updated based on plant-specific 2015, therefore, the well known, generally accepted experience. The generic database is taken from contractor 0.33 values were all statistical approaches. Use of report (Reference 1) and is an aggregation of available changed to 0.5 as Jefferys non-informative prior, data sources that include both industry raw data and part of a plant-for this case it is equivalent to generic estimations to define a best-estimate generic specific data update assuming 0.5 failures when estimate that reflects both actual industry experience and a that was performed there are no failures, is a more broad range of opinions related to failure rates. The and incorporated rigorous way. aggregation approach is a standard statistical analysis and into the HNP PRA the two sources are first aggregated separately and then working model for DA-C1-01 combined. Since no adjustment to the generic estimations PRA applications.

is made during the process, all failure rates based solely on The 2015 updated generic estimates can be excluded from this discussion data will be used for and no issue exists for resolution. analyses to be performed under the Of the 140 generic failure rates provided in the referenced SFCP when report, only 59 have historical plant data and the remaining implemented. There 81 failure rates do not utilize the estimate factor. These 81 is no additional are correct without any consideration of the selection of the impact to the 5b 0.33 failure approach. Of the remaining 59, only 34 actually application.

applied the 0.33 factor so that an additional 25 failure rates are unaffected.

The remaining 34 did use the 0.33 failure approach for at least one of the plant data records. The appropriateness of this approach for these failure rates is discussed below.

U.S. Nuclear Regulatory Commission Page 6 of 91 HNP-15-083, Enclosure 1 Table 1. Resolved but Open F&Os from 2007 Industry Peer Review of the HNP Internal Events PRA.

F&O # Finding Resolution 5b Impact For the failure events that utilize some measure of historical plant data, the plant data from available sources was used to generate a log-normally-distributed failure rate distribution based on a statistical assessment using each plant data source as an individual data point (also called a plant-to-plant variability). This distribution represents an industry generic value based on actual collected plant data.

This statistical assessment is based on collected data points and is not in any way related to the Jeffreys uninformed prior assumption so the comment is not appropriate when dealing with a lack of failures. In contrast, a failure rate must be estimated by using an upper bound approach. The development of a failure rate for a case with zero failures is documented in Reference 2 and utilizes a Chi-squared distribution assuming two degrees of freedom.

This is typically considered a 95 percentile value and is considered to be very conservative when estimating the mean failure rate.

An alternative approach is to assume that for all plant data sources with no data that 0.33 failures have occurred.

Reference 3 indicates that this is a more realistic estimate for the expected value than the Chi-squared approach. This approach was used for cases when there were no failures indicated by the plant data.

It is important to remember that in this context plant data refers to a historical experience at another plant and is not HNP-specific plant data and it is not representative of a desire to develop a Bayesian updated value for HNP but

U.S. Nuclear Regulatory Commission Page 7 of 91 HNP-15-083, Enclosure 1 Table 1. Resolved but Open F&Os from 2007 Industry Peer Review of the HNP Internal Events PRA.

F&O # Finding Resolution 5b Impact rather to generate an industry failure rate distribution using plant-to-plant variability.

Using the plant data from the various sources the distribution is developed and a mean value is found. It is then combined arithmetically with the other generic sources to arrive at an aggregate value. Therefore, the use of the 0.33 factor has only a small impact on the distribution of the plant data mean value which is then combined with the generic failure rate mean values to define a new generic failure rate.

The use of 0.33 is not associated with a Bayesian updating activity but a statistical assessment of data. The use of the value is based on a documented approach for estimating a failure rate and it is applied appropriately for the analysis being performed. It is only applied for some of the failure rate data and is always combined with other generic estimates to form an aggregated mean value such that the actual selection of the value (0.33 or 0.5) has very little if any impact on the outcome. Therefore, the current approach is valid and no changes are required.

The discussion in Appendix B was changed to eliminate confusion introduced through the one third failure discussion.

References

1. Young, R., Documentation of the RSC Generic Database for PSA Studies, Rev. 1, Ricky Summitt Consulting, Inc.,

RSC 97-90, April 1999.

2. Handbook of Parameter Estimation for Probabilistic Risk

U.S. Nuclear Regulatory Commission Page 8 of 91 HNP-15-083, Enclosure 1 Table 1. Resolved but Open F&Os from 2007 Industry Peer Review of the HNP Internal Events PRA.

F&O # Finding Resolution 5b Impact Assessment, U.S. Nuclear Regulatory Commission, NUREG/CR-6823, September 2003

3. Welker, E. and M. Lipow, Estimating the Exponential Failure Rate from Data with No Failure Events, TRW Systems Group, January 1974.

On page 23 of Appendix B - A review of the P.S. data indicates that this is not a specific Documentation was Component Database example from plant experience but rather a hypothetical updated for clarity.

development, it was stated if a situation that is not consistent with the PSA data collection. There is no impact chiller trips and within an hour It was determined that if a component trips while operating, to the 5b is restarted by the operator, that the trip, regardless of reason, would be categorized as application.

after which it runs successfully, maintenance rule functional failure. It would therefore have DA-C4-01 this would not be considered as been counted as a failure to run for the PSA failure rate a failure This event should be calculation. The only exception, is if the trip was due to a counted as 1 failure and 1 component outside the component boundary that is success event because in an specifically modeled in the PSA. The discussion on this emergency situation there may type of failures has been removed to avoid confusion.

not be an hour available for waiting and/or recovery.

The time that components were The requirement refers to components that use standby The model data and configured in standby was not failures and clearly have standby time and operational time. documentation were estimated. Note that only MOV All components types using plant specific data use demand updated to address failure data update needs this only failure rates except ten infrequently tested MOVs use this finding. There is kind of information for this PRA. standby failures rates. Because functional failures are no impact to the 5b recorded for failure to stroke from the standby position to application.

DA-C8-01 the operational position and from the operational position to the standby position, the only time components are technically not in standby with regard to failure monitoring is during travel. The average fraction of travel time, is less than 0.01%. Applying this time reduction in standby time has no impact on the calculated failure rate at the precision

U.S. Nuclear Regulatory Commission Page 9 of 91 HNP-15-083, Enclosure 1 Table 1. Resolved but Open F&Os from 2007 Industry Peer Review of the HNP Internal Events PRA.

F&O # Finding Resolution 5b Impact calculated and with consideration of the normal uncertainty bounds.

MOV spurious operation was also reviewed because this is a similar calculation to the standby failure to operate calculation. The same rules apply for definitions of functional failures. If a valve transferred from its standby position to its non-standby position, and vice-versa, regardless of function, a failure would be applied.

Therefore, there is no time that a valve is not in standby.

However, there are four valves that were included in the population that have power removed during plant power operations, such that the potential for spurious operation is essentially eliminated so that the standby state of these valves is not representative of the modeled failure mode.

There are four of such valves used in the plant specific data calculation. Their contribution to the total valve exposure time has been removed from the data calculation.

The overall impact is an increase of 3.5% of the valve transferring failure rate and potential increase in CDF of approximately 0.02%.

The following files were updated from this finding.

BayesComponents.xls, HNP_MOV_Data.xls, Appendix B:

HNP-F-PSA-0023, HNP_2007.RR

U.S. Nuclear Regulatory Commission Page 10 of 91 HNP-15-083, Enclosure 1 Table 1. Resolved but Open F&Os from 2007 Industry Peer Review of the HNP Internal Events PRA.

F&O # Finding Resolution 5b Impact It was CC-I because no The finding questions if counted tests completely or The analysis was evidence could be found to partially exposes all elements of the modeled failure reviewed, and prove CC-II is met. It could be modes. The SR for capability categories I, II are the same documentation was CC-II if it can be shown that for counting valid tests. In short, partial tests should not be updated to partial test results were not used as valid tests if all elements of the modeled failures demonstrate counted as valid test (see mode are not tested. CC-II further states that if component CC-II/III. There is no example in Cat II requirements) failures are decomposed into sub-elements, then use tests impact to the 5b that exercise the specific sub-element being modeled. CC- application.

III says to decompose failure modes into sub-elements that are fully tested.

The specific example cited from the standard, ASME RA-SB-2005, is for models that include load sequencers in the EDG component boundary. Typically, the load sequencer is only tested every 18 months instead of monthly with the DA-C10-01 EDG operability test. Therefore, the only valid test should be the 18 month test if the load sequencer is not broken out of the component boundary.

Appendix B (Ref. HNP-F/PSA-0023, Attachment 2) lists the component boundaries and states that the HNP EDG component boundary excludes the load sequencer. The HNP load sequencer is explicitly modeled in the ESFAS system.

A more appropriate example for HNP is for infrequently tested valves. A stroke test for a motor-operated-valve (MOV) may not appropriately verify the movement of a valve body at the same test interval. HNP uses sub-elements for MOVs representing the motor operator and the valve body with the appropriate tests as the basis.

Section 10 (Ref. HNP-F/PSA-0034) provides

U.S. Nuclear Regulatory Commission Page 11 of 91 HNP-15-083, Enclosure 1 Table 1. Resolved but Open F&Os from 2007 Industry Peer Review of the HNP Internal Events PRA.

F&O # Finding Resolution 5b Impact documentation for how HNP sub-divides component failure modes based on testing exposure which is consistent with SR DA-C10 Category III.

Other components, using plant specific data, were reviewed for partial test opportunities, like pumps and motors. Since actual operational and demand data is collected and only test retuning component to operability or placing in service are used, no partial test were found.

(See Appendix B for data collection methods) The specific test for each component failure mode and sub-element are documented in each system notebook.

Because only complete tests for the specific component boundary are counted, and failure modes are decomposed into sub-elements where sub-elements are exposed differently, HNP not only meets CC-II but also meets category CC-III for SR DA-C10.

The component boundary for For plant specific data using maintenance rule data, Data and EDG defined in generic source breakers associated with DGs or pumps are tracked in both documentation were NUREG/CR-5497 (Reference the component and the ac-power monitoring groups. PSA updated to address

24) is not consistent with the systems are considered safety significant for maintenance the finding. There is boundary defined in component rule. Functional failures of the breakers always are applied no impact to the 5b database (Attachment 2 of to the safety significant system. Therefore, functional application.

DA-D6-01 Appendix B - component failures of breakers are within the PSA component data database development). There boundary and within the MR component data boundary.

is no evidence that a systematic For other components, the task included reviewing comparison was performed for NUREG/CR-5497 for boundary consistency, identification other component boundaries. of outliers. The following outliers were identified and Make sure that component corrected as appropriate:

boundaries are consistent

U.S. Nuclear Regulatory Commission Page 12 of 91 HNP-15-083, Enclosure 1 Table 1. Resolved but Open F&Os from 2007 Industry Peer Review of the HNP Internal Events PRA.

F&O # Finding Resolution 5b Impact between component failure

EDG output breakers changed to be included in the data and CCF database EDG component boundary

Battery charger input and output breakers were changed to be included in the component boundary

AFW TDP trip and governor valves were changed to be included in the component boundary.

For the above cases, breakers were modeled uniquely, but no CCF were modeled. The independent failures were set to zero so the events could be retained to account for use in mapping databases associated with EOOS and FRANC.

Their associated system notebooks were updated.

Appendix B definitions on component boundaries were revised and a statement was added indicating that boundaries for CCFs and independent failures were verified to be consistent Appendix B Attachment 2 was updated.

It was not shown that generic The PS data used for the component failure modes in the Documentation on MGL parameters are consistent PSA was reviewed for potential CCFs. No failures were the analysis was with plant experience. It is identified that could be categorized as common cause added to address recommended to show, using a failures. Beta estimators for the plant specific experience the finding. There is proper statistical method, that were produced using a statistical approach. The MGL no impact to the 5b MGL parameters are consistent values used in Appendix B were converted to Beta application.

with plant experience. estimators and compared to the plant specific values. Due DA-D6-02 to a low population of plant specific failure experience, the derived Beta estimators for most of the component types provide no statistical insights for a comparison to the generic MGLs. For component groups experiencing a large number of failures, some insights can be gained. The results were reviewed based on percent difference in the Beta estimators between the generic values and the plant specific values. The following component types had a

U.S. Nuclear Regulatory Commission Page 13 of 91 HNP-15-083, Enclosure 1 Table 1. Resolved but Open F&Os from 2007 Industry Peer Review of the HNP Internal Events PRA.

F&O # Finding Resolution 5b Impact difference in Beta estimators greater than 50%.

MOVs and AOVs - the plant specific Beta value were based on the total MOV (or AOV) failures across multiple systems and was somewhat smaller than values used in the model. However, the model uses specific CCFs for various systems based on data available in NUREG/CR-5497 and from RSC 01-17. If the plant specific failures were broken up by system, then there would be insufficient data for a meaningful comparison. Therefore, the values that are currently in use are considered to be adequate based on a lack of plant experience.

Chillers - There are a large number of chiller failures, many are repetitive failures but no common cause failures were assumed based on time of failure. The Beta factor used in the PSA comes from EPRI TR-100382 and is approximately an order of magnitude greater than the plant experience for failure to run. A detailed review of the plant data may indicate some common cause aspects to reduce the difference. However, currently the CCF of the chillers contributes less than 1% to the CDF using the higher generic value. Therefore, a more detailed assessment and potential reduction of the CCF is not merited at this time.

Air compressors A & B. The generic Beta estimator for CCF to run of compressors is approximately 55% higher than the plant specific experience. The CCF of the compressors contributes less than 1% to the CDF using the higher generic value. Therefore, a more detailed assessment and potential reduction of the CCF is not merited at this time.

U.S. Nuclear Regulatory Commission Page 14 of 91 HNP-15-083, Enclosure 1 Table 1. Resolved but Open F&Os from 2007 Industry Peer Review of the HNP Internal Events PRA.

F&O # Finding Resolution 5b Impact MGL parameter values are HNP uses MGLs representing both staggered and non- Resolution of this different according to testing staggered testing. These MGLs were initially applied F&O accounted for schemes. MGL parameters in without consideration of the impact of component testing staggered and non-NUREG/CR-5497 are based on schemes. The MGLs used from NUREG/CR-5497 are staggered testing.

staggered testing scheme. No based on staggered testing and appropriately represents This data was investigation was made to testing practices for major safety train equipment. For incorporated into check if MGL parameters based components separated by trains that are tested online, this the model and the on staggered testing are finding is not an issue. documentation was applicable to this PRA. Make updated. There is sure that MGL parameters for A review of the specific components using NUREG/CR- no impact to the 5b proper testing scheme are 5497 values was performed. The following components application.

used. were found to inappropriately use the staggered testing assumption: SI to RCS loop check valves, AFW to SG check valves, SG PORVs, PRZ PORVs, PRZ SRVs, and RHR hot leg suction MOVs. Given a non-conservatism DA-D6-03 existed, a sensitivity study was performed that determined the impact of correcting these MGLs to non-staggered testing would result in an increase in CDF of less than 5%.

Therefore, this finding has a minimal impact on the PSA results.

The MGLs used from report RSC 01-17 provide MGLs for generic components. This source is used for the balance of component CCFs not found in NUREG/CR-5497. The MGLs developed in this document are based on the same underlying data used in NUREG/CR-5497, but use the more conservative non-staggered testing assumption. A known conservatism is that pumps that are tested on a staggered basis use this data and therefore produce conservative CDF results. A typical example is the CCW pumps. No sensitivity study was performed to determine

U.S. Nuclear Regulatory Commission Page 15 of 91 HNP-15-083, Enclosure 1 Table 1. Resolved but Open F&Os from 2007 Industry Peer Review of the HNP Internal Events PRA.

F&O # Finding Resolution 5b Impact the impact of correcting these conservative MGL parameters because impact vectors are not readily available to perform the study. However, the current PSA results are known to be conservative.

An action item (NTM 49072-48) is open to correct the CCF events to use MGLs representing the appropriate testing scheme during the next model update. The overall impact on PSA results is expected to be a reduction in CDF due to the important components using the more conservative non-staggered assumption.

Per discussions with HNP staff, The first paragraph does not refer to a finding. Category II Documentation was type codes are assigned to all is met. Comment refers to bringing up the uncertainty updated for basic events representing analysis to Category III. No action is required. clarification. There independent component Appendix B (HNP-F/PSA-0023) provides a listing of plant is no impact to the failures. However, common specific and generic component types using type codes. 5b application.

cause events are not assigned Basic events not falling into those categories of failures use type codes. There is a data probabilities calculated outside of the fault tree. These correlation between common include common cause failures, initiating events, split cause events, e.g., for 3 fractions, operator actions, and recoveries. Although this Charging Pumps (A, B, and C), recommendation requires no action, this information has QU-E3-01 there is a correlation between been added to Appendix B for clarification as section B.5.

the 2 out of 3 combination [A B], [A C], and [B C].

Recommend type codes be assigned to common cause events. In addition, it is recommended to enhance HNP-F/PSA-0001 to include a description of the process for assigning type codes.

U.S. Nuclear Regulatory Commission Page 16 of 91 HNP-15-083, Enclosure 1 Table 1. Resolved but Open F&Os from 2007 Industry Peer Review of the HNP Internal Events PRA.

F&O # Finding Resolution 5b Impact SR HR-D3 requires that each This assumption is justified since the procedures Documentation was detailed evaluation of a pre- performing maintenance and restoring systems from updated to address initiator HFE include an maintenance are step-by-step procedures that include step this finding. There is assessment of the quality of the sign-offs for place keeping and, in many instances, a no impact to the 5b written procedures and the reviewer or verifier sign-off for key steps. The procedures application.

quality of the man-machine also have twenty years of performance with improvements interface. While Harris does added based on the experience gained. The plant has a identify the specific procedures proceduralized labeling program (PLP-610, Equipment for each action, they use a Identification and Numbering System) for plant piping and HR-D3-01 blanket assumption that the HVAC components and the proceduralized work HNP procedures are accurate management program (ADM-NGGC-0104, Work and consistent with the plant Management Process) requires Controlled Wiring configuration. (See Assumption Diagrams in work packages for electrical / I&C work.

6 in section 2.6 of Appendix E.) Added to section 2.6 #6 of HNP-F/PSA-0070 There seems to be no evaluation of the man-machine interface.

Harris uses the HRA Toolbook The spreadsheets in the HRA Toolbox have been changed Resolution of this for quantifying their pre-initiator to use the mean instead of median values to determine the F&O updated the HEPs. For the pre-initiator pre-initiator HEP values. Values have been updated in the spreadsheets in the HEPs, Harris basically uses the calculation and the model. HRA Toolbox and in ASEP approach and treats the the model. There is ASEP Basic HEPs as means no impact to the 5b HR-D6-01 with the associated error application.

factors. However, as defined on page xv of NUREG/CR-4772, the ASEP BHEP values are medians for a log-normal distribution. Thus, the treatment of the BHEP values for the pre-

U.S. Nuclear Regulatory Commission Page 17 of 91 HNP-15-083, Enclosure 1 Table 1. Resolved but Open F&Os from 2007 Industry Peer Review of the HNP Internal Events PRA.

F&O # Finding Resolution 5b Impact initiators is mathematically incorrect. Note that in the HRA TOOLBOOK Users Guide (SAROS 21-16), this issue was specifically identified and evaluated. The contention was that the ASEP values were intended to be bounding, screening values and that they were so conservative that treating them as means still yielded conservative results with respect to an equivalent THERP analysis with the proper conversion from medians to means. While the reviewers appreciate this issue, the treatment is still mathematically incorrect. NOTE: the HEPs generated by using this approach are considered to still be somewhat conservative with respect to a detailed analysis using THERP. An inquiry will be made to the ASME CRNM with respect to this issue.

U.S. Nuclear Regulatory Commission Page 18 of 91 HNP-15-083, Enclosure 1 Table 1. Resolved but Open F&Os from 2007 Industry Peer Review of the HNP Internal Events PRA.

F&O # Finding Resolution 5b Impact The review team could not find MAAP runs were conducted to confirm that the current Resolution of this evidence that sequence specific timing for feed-and bleed initiation used in the HRA was F&O updated timing estimates were used. appropriate for steam generator tube rupture and S1 documentation of The particular case examined LOCA. Four cases were run, one SGTR and three S1 the MAAP runs that was Feed and Bleed. Only one LOCAs cases for the upper, mid, and lower break range. were completed.

Feed and Bleed HEP was Core damage did not occur in any case for feed and bleed There is no impact found and the timing to support initiated at 75 minutes, which is the limiting time used for to the 5b this appeared to not be based transient sequences. Appendix D was updated to reflect application.

on a limiting case. additional MAAP cases that were run.

HR-F2-01 The team was referred to the success criteria in Appendix D of the HNP PRA. The feed and bleed success criteria were based on a transient case and success time for opening a primary PORV. The team believed that other cases such as small LOCA may be more time limiting.

NUREG-1278 contains median The spreadsheets in the HRA Toolbox have been changed Resolution of this values that do not appear to be to use the mean instead of median values to determine the F&O updated the converted to means before post-initiator HEP values. Values have been updated in the spreadsheets in the being used in the HNP PRA. calculation and the model. HRA Toolbox and in For example, spreadsheet the model. There is HR-G9-01 HNP-CP.XLS contains no impact to the 5b individual tables that reference application.

THERP median values. These values are multiplied by factors to account for stress in individual actions, but they are never converted to means.

U.S. Nuclear Regulatory Commission Page 19 of 91 HNP-15-083, Enclosure 1 Table 2:

HNP Assessment of Supporting Requirement (SR) Capability Categories For Initiating Events (IE), ASME/ANS RA-Sa-2009 as endorsed by RG 1.200, Rev. 2 Capability Category HLR SR I II III Met Not Met N/A HLR-IE-A IE-A1 ALL X IE-A2 ALL X IE-A3 ALL X IE-A4 I/II X IE-A5 II X IE-A6 II X IE-A7 ALL X IE-A8 II X IE-A9 II X IE-A10 ALL X HLR-IE-B IE-B1 ALL X IE-B2 ALL X IE-B3 II X IE-B4 ALL X IE-B5 ALL X HLR-IE-C IE-C1 ALL X IE-C2 ALL X IE-C3 ALL X IE-C4 ALL X IE-C5 I/II X IE-C6 ALL X IE-C7 I/II X IE-C8 ALL X IE-C9 ALL X IE-C10 ALL X IE-C11 ALL X IE-C12 ALL X IE-C13 I/II X IE-C14 I/II X IE-C15 ALL X HLR-IE-D IE-D1 ALL X IE-D2 ALL X IE-D3 ALL X

U.S. Nuclear Regulatory Commission Page 20 of 91 HNP-15-083, Enclosure 1 Table 3:

HNP Assessment of Supporting Requirement (SR) Capability Categories For Accident Sequences (AS), ASME/ANS RA-Sa-2009 as endorsed by RG 1.200, Rev. 2 Capability Category HLR SR I II III Met Not Met N/A HLR-AS-A AS-A1 ALL X AS-A2 ALL X AS-A3 ALL X AS-A4 ALL X AS-A5 ALL X AS-A6 ALL X AS-A7 I/II X AS-A8 ALL X AS-A9 II X AS-A10 II X AS-A11 ALL X HLR-AS-B AS-B1 ALL X AS-B2 ALL X AS-B3 ALL X AS-B4 ALL X AS-B5 ALL X AS-B6 ALL X AS-B7 ALL X HLR-AS-C AS-C1 ALL X AS-C2 ALL X AS-C3 ALL X

U.S. Nuclear Regulatory Commission Page 21 of 91 HNP-15-083, Enclosure 1 Table 4:

HNP Assessment of Supporting Requirement (SR) Capability Categories For Success Criteria (SC), ASME/ANS RA-Sa-2009 as endorsed by RG 1.200, Rev. 2 Capability Category HLR SR I II III Met Not Met N/A HLR-SC-A SC-A1 ALL X SC-A2 II/III X SC-A3 ALL X SC-A4 ALL X SC-A5 II/III X SC-A6 ALL X HLR-SC-B SC-B1 II X SC-B2 II/III X SC-B3 ALL X SC-B4 ALL X SC-B5 ALL X HLR-SC-C SC-C1 ALL X SC-C2 ALL X SC-C3 ALL X

U.S. Nuclear Regulatory Commission Page 22 of 91 HNP-15-083, Enclosure 1 Table 5:

HNP Assessment of Supporting Requirement (SR) Capability Categories For System Analysis (SY), ASME/ANS RA-Sa-2009 as endorsed by RG 1.200, Rev. 2 Capability Category Not HLR SR I II III Met Met N/A HLR-SY-A SY-A1 ALL X SY-A2 ALL X SY-A3 ALL X SY-A4 II/III X SY-A5 ALL X SY-A6 ALL X SY-A7 I/II X SY-A8 ALL X SY-A9 ALL X SY-A10 ALL X SY-A11 ALL X SY-A12 ALL X SY-A13 ALL X SY-A14 ALL X SY-A15 ALL X SY-A16 I/II X SY-A17 ALL X SY-A18 ALL X SY-A19 ALL X SY-A20 ALL X SY-A21 ALL X SY-A22 II X SY-A23 ALL X SY-A24 ALL X HLR-SY-B SY-B1 II/III X SY-B2 I/II X SY-B3 ALL X SY-B4 ALL X SY-B5 ALL X SY-B6 ALL X

U.S. Nuclear Regulatory Commission Page 23 of 91 HNP-15-083, Enclosure 1 Table 5, continued:

HNP Assessment of Supporting Requirement (SR) Capability Categories For System Analysis (SY), ASME/ANS RA-Sa-2009 as endorsed by RG 1.200, Rev. 2 Capability Category HLR SR I II III Met Not Met N/A HLR-SY-B SY-B7 II X (contd) SY-B8 ALL X SY-B9 ALL X SY-B10 II/III X SY-B11 ALL X SY-B12 ALL X SY-B13 ALL X SY-B14 ALL X SY-B15 ALL X HLR-SY-C SY-C1 ALL X SY-C2 ALL X SY-C3 ALL X

U.S. Nuclear Regulatory Commission Page 24 of 91 HNP-15-083, Enclosure 1 Table 6:

HNP Assessment of Supporting Requirement (SR) Capability Categories For Human Reliability (HR), ASME/ANS RA-Sa-2009 as endorsed by RG 1.200, Rev. 2 Capability Category Not HLR SR I II III Met Met N/A HLR-HR-A HR-A1 ALL X HR-A2 ALL X HR-A3 ALL X HLR-HR-B HR-B1 II/III X HR-B2 ALL X HLR-HR-C HR-C1 ALL X HR-C2 II/III X HR-C3 ALL X HLR-HR-D HR-D1 ALL X HR-D2 II X HR-D3 II/III X HR-D4 ALL X HR-D5 ALL X HR-D6 ALL X HR-D7 I/II X HLR-HR-E HR-E1 ALL X HR-E2 ALL X HR-E3 II/III X HR-E4 II/III X HLR-HR-F HR-F1 I/II X HR-F2 II X HLR-HR-G HR-G1 II X HR-G2 ALL X HR-G3 II/III X HR-G4 II X HR-G5 II X HR-G6 ALL X HR-G7 ALL X HR-G8 ALL X HLR-HR-H HR-H1 II X HR-H2 ALL X HR-H3 ALL X

U.S. Nuclear Regulatory Commission Page 25 of 91 HNP-15-083, Enclosure 1 Table 6:

HNP Assessment of Supporting Requirement (SR) Capability Categories For Human Reliability (HR), ASME/ANS RA-Sa-2009 as endorsed by RG 1.200, Rev. 2 Capability Category Not HLR SR I II III Met Met N/A HLR-HR-I HR-I1 ALL X HR-I2 ALL X HR-I3 ALL X

U.S. Nuclear Regulatory Commission Page 26 of 91 HNP-15-083, Enclosure 1 Table 7:

HNP Assessment of Supporting Requirement (SR) Capability Categories For Data Analysis (DA), ASME/ANS RA-Sa-2009 as endorsed by RG 1.200, Rev. 2 Capability Category Not HLR SR I II III Met Met N/A HLR-DA-A DA-A1 ALL X DA-A2 ALL X DA-A3 ALL X DA-A4 ALL X HLR-DA-B DA-B1 II X DA-B2 I/II X HLR-DA-C DA-C1 ALL X DA-C2 ALL X DA-C3 ALL X DA-C4 ALL X DA-C5 ALL X DA-C6 ALL X DA-C7 II/III X DA-C8 II/III X DA-C9 I/II X DA-C10 II X DA-C11 ALL X DA-C12 ALL X DA-C13 II/III X DA-C14 ALL X DA-C15 ALL X DA-C16 ALL X HLR-DA-D DA-D1 II X DA-D2 ALL X DA-D3 II X DA-D4 II/III X DA-D5 II X DA-D6 II X DA-D7 ALL X DA-D8 II X

U.S. Nuclear Regulatory Commission Page 27 of 91 HNP-15-083, Enclosure 1 Table 7:

HNP Assessment of Supporting Requirement (SR) Capability Categories For Data Analysis (DA), ASME/ANS RA-Sa-2009 as endorsed by RG 1.200, Rev. 2 Capability Category Not HLR SR I II III Met Met N/A HLR-DA-E DA-E1 ALL X DA-E2 ALL X DA-E3 ALL X

U.S. Nuclear Regulatory Commission Page 28 of 91 HNP-15-083, Enclosure 1 Table 8:

HNP Assessment of Supporting Requirement (SR) Capability Categories For Quantification (QU), ASME/ANS RA-Sa-2009 as endorsed by RG 1.200, Rev. 2 Capability Category Not HLR SR I II III Met Met N/A HLR-QU-A QU-A1 ALL X QU-A2 ALL X QU-A3 II X QU-A4 ALL X QU-A5 ALL X HLR-QU-B QU-B1 ALL X QU-B2 ALL X QU-B3 ALL X QU-B4 ALL X QU-B5 ALL X QU-B6 ALL X QU-B7 ALL X QU-B8 ALL X QU-B9 ALL X QU-B10 ALL X HLR-QU-C QU-C1 ALL X QU-C2 ALL X QU-C3 ALL X HLR-QU-D QU-D1 ALL X QU-D2 ALL X QU-D3 ALL X QU-D4 II/III X QU-D5 ALL X QU-D6 II/III X QU-D7 ALL X HLR-QU-E QU-E1 ALL X QU-E2 ALL X QU-E3 II X QU-E4 ALL X

U.S. Nuclear Regulatory Commission Page 29 of 91 HNP-15-083, Enclosure 1 Table 8:

HNP Assessment of Supporting Requirement (SR) Capability Categories For Quantification (QU), ASME/ANS RA-Sa-2009 as endorsed by RG 1.200, Rev. 2 Capability Category Not HLR SR I II III Met Met N/A HLR-QU-F QU-F1 ALL X QU-F2 ALL X QU-F3 II/III X QU-F4 ALL X QU-F5 ALL X QU-F6 ALL X

U.S. Nuclear Regulatory Commission Page 30 of 91 HNP-15-083, Enclosure 1 Table 9:

HNP Assessment of Supporting Requirement (SR) Capability Categories For LERF Analysis (LE), ASME/ANS RA-Sa-2009 as endorsed by RG 1.200, Rev. 2 Capability Category Not HLR SR I II III Met Met N/A HLR-LE-A LE-A1 ALL X LE-A2 ALL X LE-A3 ALL X LE-A4 ALL X LE-A5 ALL X HLR-LE-B LE-B1 III X LE-B2 II X LE-B3 ALL X HLR-LE-C LE-C1 II X LE-C2 II/III X LE-C3 II/III X LE-C4 II X LE-C5 II X LE-C6 ALL X LE-C7 ALL X LE-C8 ALL X LE-C9 II/III X LE-C10 II X LE-C11 II/III X LE-C12 II X LE-C13 II/III X HLR-LE-D LE-D1 II X LE-D2 II X LE-D3 II X LE-D4 II X LE-D5 II X LE-D6 II X LE-D7 II X HLR-LE-E LE-E1 ALL X LE-E2 II X LE-E3 II X LE-E4 ALL X

U.S. Nuclear Regulatory Commission Page 31 of 91 HNP-15-083, Enclosure 1 Table 9:

HNP Assessment of Supporting Requirement (SR) Capability Categories For LERF Analysis (LE), ASME/ANS RA-Sa-2009 as endorsed by RG 1.200, Rev. 2 Capability Category Not HLR SR I II III Met Met N/A HLR-LE-F LE-F1 II/III X LE-F2 ALL X LE-F3 ALL X HLR-LE-G LE-G1 ALL X LE-G2 ALL X LE-G3 II/III X LE-G4 ALL X LE-G5 ALL X LE-G6 ALL X

U.S. Nuclear Regulatory Commission Page 32 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact Table 3.3 of Section 3 contains The current Table 3.3 had IPE data in it. The WOG PSA Data tables and outdated IPE data for Transient Comparison Database, Revision 4, 2003 was used to documentation Initiating Event Categories. develop a new table. were updated.

There is no impact 01-IE-A3a to the 5b application.

Additional documentation is The plant events occurring at power and during shutdown Documentation was needed to indicate how events were reviewed. Events occurring during shutdown that added per the 02-IE-A5 occurring during shut-down are were applicable to power operations were included in the finding. There is no evaluated. frequency of the initiating events, such as LOSP. A impact to the 5b discussion was added to Section 3.0 application.

No documentation of interviews An LER review was performed to determine any potential Documentation was with plant personnel to missed IEs. LERs affecting initiating events or systems updated to reflect determine if potential initiating were added to the IE analysis and system notebooks the analysis events have been overlooked. respectively. There were no new initiating events identified. performed. There No new interviews were considered necessary because the is no impact to the 03-IE-A6 PSA personnel continually consult plant personnel for 5b application.

various applications and the model periodic updates.

Additionally, the IPE received a Plant Technical Support review and a number of notebooks have undergone system engineer review.

Plant-specific operating An LER review was performed from initial IPE development Documentation was experience for identifying IE through 2006 to determine any potential missed IEs. LERs updated to reflect precursors has apparently not affecting initiating events or systems were added to the IE the analysis 04-IE-A7 been examined since 1995, analysis and system notebooks respectively. There were performed. There including review of LERs and no new initiating events identified. is no impact to the other applicable information. 5b application.

U.S. Nuclear Regulatory Commission Page 33 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact Some EPRI subgroup initiating It was determined that the groupings of the EPRI subgroup Updated the events appear to be in IEs were appropriately matched to the HNP initiating initiating event groupings that behave events. For HNP a loss of condenser or loss of circulating analysis and differently from water would not impact the availability of FW, so these documentation.

other events in the group. For events can be grouped in %T3. The availability of FW or There is no impact example, loss of condenser steam dumps is based on a plant specific split fraction of to the 5b vacuum and loss of circulating the %T3 and %T1 events. Rod drop events would go in the application.

water (vs. say rod drop) - if manual trip category %T1.

there is a loss of condenser vacuum, secondary side As part of this F&O the following were performed:

equipment will not be available.

Therefore, credit cannot be Reviewed the change logs for changes in initiators with given to steam dumps, main emphasis on %T1, T2, T3, T7, T8.

feedwater, etc. These events seem to have different impacts Updated plant specific initiating event spreadsheet using 05-IE-B3 on plant and system response current LER data through 2005 and generic data from than other events in the same NUREG/CR-5750 that was re-categorized to match HNPs grouping which is not in breakdown of IEs. Re-categorized excessive feedwater to accordance with the PRA %T3 versus %T7. Verified that removal of prior IE %T8 was Procedures Guide. appropriate based on NUREG categories and signal inter-dependencies. Additionally, the HNP IE spreadsheet is being converted back to a moment-matching update method (as prior revisions of the HNP PSA used) to gain more realistic data results. The justification for this is taken from an RSC, Inc. engineering calculation provided to HNP in 2002. From this document:

Overall, it appears that the discrete method has several weaknesses that are not present in the moment-matching approach. The discrete method is dependent on the interval selection and does not necessarily provide a single

U.S. Nuclear Regulatory Commission Page 34 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact rule for all cases. In fact, the analysis shows how a more refined interval selection can both increase and decrease the result. The method also consistently penalizes the analysis by generating failure rates in excess of the generic rate although no atypical performance is found due to the need to include a failure contribution to be able to perform the update. This bias could influence the selection of important components and risk ranking. Since the moment-matching method produces consistent results, the risk ranking should not be impacted regardless of the failure rate estimation.

Another possible drawback for the discrete evaluation is that it appears to inherently discount the quantity of plant data collected since it only included the failure rate and variance into the calculation as second order terms. For example, a plant that collected 1 failure in 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months would be treated in a similar manner as a plant with 10 failures in 1000 hours0.0116 days 0.278 hours 0.00165 weeks 3.805e-4 months since the failure rate would be the same. This seems counter to the desire to reflect the inclusion of plant-specific characteristics into the analysis to gain true understanding of plant performance.

Although both methods may be utilized in the industry, the moment-matching approach seems to be more appropriate and should be used in data analyses. Reasons for this selection include:

1. It is statistically valid and maintains distribution consistency during the update

2. It is not impacted by any user-defined function such as

U.S. Nuclear Regulatory Commission Page 35 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact the number of intervals chosen

3. It produces consistent results between failure rate cases such that rank order is not biased

4. It does not produce illogical results for cases where no failure data has occurred and there is no reason to predict poor performance

5. It allows for continued data collection to impact the update even when the plant data indicates that failure rate is unchanged Other tasks associated with the completion of this F&O included:

Developed split fraction for loss of steam dumps.

Updated split fractions for plant trips that feed water was lost and had to be restarted.

Updated split fractions for plant trips with RCS pressure challenges.

Corrected gates #MRST and #QT102 (See notes below)

Update Section 3 for appropriate documentation of the impact of initiators on plant response. Including Updating Tables C.2 and C.3 as necessary to verify that the existing breakdown of initiators (especially T1 and T3) adequately capture the plant response.

Document steam dump/condenser availability spit fraction

U.S. Nuclear Regulatory Commission Page 36 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact in the SG system notebook.

Update Table D.9 Updated the frequency of AC/DC bus failure to use a generic type code frequency as opposed to a published NUREG/CR-5750 frequency or newer data. This was done to be consistent with other bus failures in the model and to fault tree based IE models. Additionally, there is some debate over if the published frequency values are per bus or per plant, so proper use was not certain.

Updated LOSP frequency through 2005 and document.

The HNP IE spreadsheet LOSP calculations were converted back to a combined probability of non-recovery curve method (as prior revisions of the HNP PSA used) to gain more realistic data results. The justification for this is that a single curve type analysis (power fit, log normal, exponential, Weibull fit, etc.) will either overestimate or underestimate the recovery probability for cases where larger component failures occur (such as EDG run failures at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ). A composite curve is required to provide a more accurate solution.

Notes: (Pre-correction)

For HNP, feedwater is not lost on a loss of condenser vacuum; however a split fraction is used to indicate initiators where FW is not available.

Steam dumps would be assumed to be lost on a loss of

U.S. Nuclear Regulatory Commission Page 37 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact condenser vacuum. There is no system notebook containing a complete description of the steam dumps. A number of initiators are modeled to fail the steam dumps.

Loss of instrument air appears to be missing as a failure of the steam dumps. However, that failure also fails feedwater so the omission is not consequential.

Gate #MRST is incorrect. It is allowing OPER-46 to recover feedwater for all transient initiating events. (Some initiating events fail FW by fault tree linking but not all)

HNP INITIATING EVENTS.XLS categorizes %T1 as manual reactor trip and %T3 as automatic reactor trip. This was done to facilitate the W ATWS methodology. The plant specific events appear to be mapped to match those categories, but it is not clear if the underlying generic data is mapped appropriately.

The old model break down was %T1, reactor trip, %T2, reactor trip with pressure challenge, and %T3, turbine trip.

The current PORV challenge split fractions are being applied to %T1, T3 (X-RCSPC) and %T4 (X%4LIFT) only, The application to T4p is missing (basis for denominator).

Table D.9 appears to be correct, except that Reactor trip (both automatic and manual) is mislabeled as Turbine trip.

The main steam notebook needs to be broadened to discuss how steam dumps are credited and what initiators are assumed to fail it.

U.S. Nuclear Regulatory Commission Page 38 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact The system notebooks already include the impact of initiators on the system.

The current breakdown of %T1, %T3 and %T7 are consistent with the data in NUREG/CR-5750. The loss of condenser vacuum will be rolled in to %T3 with an added split fraction to account for plant specific events where condenser steam dumps could not be used to recover feedwater so that only CST makeup to the condenser is credited.

NUREG/CR-5750 only includes spurious ESFAS (group QR9) actuation that accounts for categories 9 and 10 in NUREG/CR-3862. In Table D.13 of 5750, all 22 of the PWR events resulted in MSIV closure. Tables 2 and 6 of 3862 indicated the category 10 causes MSIV closures.

Category 9 was grouped in the same transient group in some of the other PSAs with an expected MSIV closure.

A review of the plant specific design is as follows:

Low PRZ P >> SI signal Low Steam Line P >> SI signal CT Hi 1 (3psi on 2/3 Channels 2-4) >> SI signal CT Hi 2 (3psi on 2/3 channels 2-4) >> MSIV Closure CT Hi 3 (10psi on 2/4 channels 1-4) >> Phase B

U.S. Nuclear Regulatory Commission Page 39 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact SI Signal >> Phase A, ESF actuations and FW isolation Phase A >> CT isolation of NSW, RCP seal return, letdown, sampling and instrument air Phase B >> CT isolation of CCW to RCPs and actuation of CT spray Tech. Specs. allow one of the above CT channels to be failed and placed in test indefinitely (energized in actuated state). Another channel may be tested if the failed channel is bypassed so that a 2/3 or 4 condition is not actuated from the test. The time limit on the bypass is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

The tech spec confirms that it is possible that operator error (failure to bypass), or common cause failures (second failure occurring while first is still in test) could lead to actuation.

Since both SI actuations (group 9) and containment pressure problems (group 10) are documented in the NUREGs as causing MSIV closure in the historical events, it is assumed (without further LER review) that the most probable scenarios are inadvertent actuations at the instrument channel level and not in the logic cabinets. This has been confirmed by plant operations staff. This means that 2 CT channels, if actuated would likely result in satisfying all the set points and result in SI, Phase A, MSIV Isolation, and Phase B. This is consistent with the current modeling approach.

U.S. Nuclear Regulatory Commission Page 40 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact The placement of group 20 (excess feedwater) into the spurious ESFAS class based on the assumption that overcooling the RCS would cause a pressure drop in the PRZ and cause an SI is an incorrect assessment. There is no indication in the NUREG/CR-5750 records of this condition. Plant personnel have confirmed that the more likely outcome is that the overcooling of the SGs would result in a reactivity spike that would manage RCS pressure and prevent the SI. The actual trip mechanism will be a race between feedwater isolation on high SG level or reactor high reactivity. In either event, the outcome is an automatic Reactor trip to be placed in event %T3 instead of

%T7.

ASME SR IE-C11 states For The purpose of this F&O is to be sure for rare events, that Documentation of rare initiating events, USE plant specific aspects that make the occurrence of the rare events was industry generic data and event unique for HNP is included, especially if a rare event improved. There is INCLUDE has occurred at HNP. HNP has experienced no rare no impact to the 5b plant-specific functions." No events. application.

evidence was found to indicate plant-specific functions were Initiating events were reviewed and where appropriate a included for rare events. discussion of plant specific experience or design was considered. The following considerations were made:

06-IE-C11 LOCA: HNP uses 4 break sizes. Actual pipe segment counts were used to split the generic 3 categories into 4:

Discussion was added that no insights would be gained from updating the generic estimate with un-informed plant experience.

SGTR: HNP derived a SGTR frequency based on a generic estimate of individual tube failure multiplied by the number

U.S. Nuclear Regulatory Commission Page 41 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact of SG tubes to address plant specific aspects.

SLB: Secondary Line Break is not a rare event based on generic industry frequency of 1.4E-2, but was included in this discussion for completeness. The generic data appears to be less representative of current industry experience, so plant specific experience of no events at HNP was included by performing a Bayes update of the generic experience to lower the event frequency to a more realistic value.

ATWS: Harris models ATWS as a response to a plant transient. It is not considered an initiating event.

LOSP: By ASME definition, LOSP is not a rare event based on industry experience, but was included in this discussion for completeness. Plant specific aspects were included in the analysis of the industry data for applicability to HNP to consider such things as plant location, administrative control, and design of the switchyard and electrical distribution system. A Bayes update was performed, but the large amount of industry data, dominates the result.

ISLOCA: HNP uses a plant specific analysis of ISLOCA based on generic valve failure rates. Data is acceptable.

Appendix C and Section 3 were updated to note above findings.

U.S. Nuclear Regulatory Commission Page 42 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact The key assumptions and key Seventy-four (74) assumptions regarding IE were identified Resolution of this sources of uncertainty in Section 3 and Appendix C, and were evaluated to F&O included associated with the initiating determine their effect on PRA results. The results of this analysis and 07-IE-D3 event analysis do not appear to evaluation are included in Appendix U, Uncertainty documentation of be documented. Analysis. uncertainties. There is no impact to the 5b application.

Although safety functions and Added table to each event tree development section that Resolution of this success criteria are contained lists safety functions, success criteria, and sequence tops F&O added in the AS notebook, they are events. For SR-A3, the requirement is to identify initiating documentation to not well organized and do not events with special handling. There are no initiating events make clear links have clear links from the that require special handling because the event tree tops from the accident specific accident sequence to model functions and not specific systems. The effects of sequence to the the supporting references. specific initiators are addressed by fault tree linking. For reference. There is SR-A4, the requirement is to identify the operator actions no impact to the 5b required and provide references. The major procedures for application.

the event tree tops were added as well as the associated 08-AS-A2 operator actions. For SR-A5, the requirement to define the accident logical progression was already met. The ESDs already define the accident progression, plus the description associated with the event trees discusses the accident progression. The ATWS event tree section was expanded to discuss the accident progression; no other action is required. For SR-A9, the requirement to reference the SC basis was accomplished by adding cross references in the same tables to the SC sections of other sources.

U.S. Nuclear Regulatory Commission Page 43 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact The list of safety functions in Table 4.1 was expanded to added additional safety Resolution of this AS Table 4.1 should be functions and supporting systems. F&O expanded the expanded to include: referenced table as 09-AS-A2

- reactor pressure control necessary. There is

- reactor coolant inventory no impact to the 5b control. application.

Most event tree descriptions do Added a general discussion in section 4.0 on the basis of Resolution of this not include any discussion of the ordering of event trees. F&O added a the ordering sequence. general discussion of the ordering 10-AS-A6 sequence in event tree descriptions.

There is no impact to the 5b application.

The end state of the Event A general discussion was added to Section 4.0 that defines Resolution of this Trees is not specifically defined core damage as uncovered fuel and success as a safe F&O added a to be stable or core damage. stable state. Specific core damage temperatures were also discussion to the added. A brief discussion of the core damage bins and documentation that 11-AS-A8 early/late classification in the event trees was added. defines core damage and success. There is no impact to the 5b application.

The T-H code requirements for The PSA success criteria and accident sequence Resolution of this acceptability need to be stated. development is based on T/H runs from both RELAP and F&O justified the Duke Energy Progress, Inc., MAAP 3.0b. The MAAP analyses are well documented and acceptability of the 12-AS-A9 uses MAAP for most PSA are considered to be acceptable for PSA purposes. The use of MAAP code.

specific T-H analysis. There is MAAP code was edited and recompiled for HNP specific There is no impact some question from NRC about considerations. No major changes to success criteria are to the 5b MAAP acceptability. expected from an upgrade to MAAP 4.05 or higher version. application.

U.S. Nuclear Regulatory Commission Page 44 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact The impact of initiators on The system notebooks already include a discussion of Resolution of this mitigating systems is not impact of various IEs on the systems. F&O added a specifically discussed in the dependency matrix documentation. An initiator-to-system impact dependency matrix was linking the impact of 13-AS-B1 added to the Section 3.0 report. The matrix indicates the initiators to extent of damage to systems from each initiator. mitigating systems.

There is no impact to the 5b application.

The use and impact of system The existing flag section in each system notebook was The resolution of alignments is not adequately updated to list the alignment flag events affecting the the F&O updated discussed. system. Most alignment flags are tied to protected train the flag section flags. within the system notebook to The plant generally operates with one train in service and address the impact 14-AS-B5a the other in standby. Because it is equally likely that either of system train is in service, alignment flags are quantified with a alignments. There value of 0.5. The contribution difference between the is no impact to the protected train A and protected train B is less than 1%. 5b application.

(F-V of 0.125 vs. 0.129) No insights are identified that need to be added to the accident sequence discussion.

Key Assumptions and Key Ninety-three (93) assumptions were identified for AS/SC in Resolution of this Sources of Uncertainty are not Section 4 and Appendices D, F, and G. They were F&O specifically specifically identified. evaluated to determine their effects on PRA results. The 93 identified the assumptions apply to either accident sequence definition or assumptions for 15-AS-C3 success criteria. An additional 48 assumptions have been AS/SC. There is no identified for Appendices H (ISLOCA) and R (LOSP impact to the 5b Recovery). They were also evaluated for their effects on application.

PRA results. The results of these evaluations are presented in Appendix U, Uncertainty Analysis.

U.S. Nuclear Regulatory Commission Page 45 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact The AS/IE documentation The accident sequences are based on functions rather Resolution of this supporting the modeling of than systems. The combined documentation of Section 3.0 F&O added a accident sequences based on and 4.0 discuss the standard approach of fault tree linking section clarifying systems and operator with small event trees used in the HNP model. The effects the modeling of responses is weak. of initiating events is embedded in the system fault tree accident models. The linking of these models to the functional logic sequences. There ensures that sufficient detail is present to propagate the is no impact to the effects of initiating events through the accident sequences. 5b application.

A short discussion was added to Section 4.1.1 that states 16-AS-A10 the following: The event trees model mitigating system functions. These do not vary with various types of transient initiating events. Additionally, the effects of specific initiating events on systems was added to Section 3.0. The specific operator responses associated with the mitigating functions was documented in Section 4.0 and Appendix D.

The discussion was expanded in Section 4.0 and cross-referenced to procedures and appropriate section of Appendix D.

No discussion of This F&O addresses pre-core damage environmental Resolution of this phenomenological conditions effects on success of components. A review of plant F&O evaluated the which address this SR was systems indicates the following areas of interest: phenomenological noted. conditions for

1. The impact of LOCAs and F&B cooling on containment several areas of equipment include: interest. The 17-AS-B3 a. RHR SD Cooling MOVs documentation was

b. RCS PORVs and BVs updated to reflect

c. Containment Instrumentation the additional

d. Fan Coolers (Level II) analysis. There is

e. Containment Isolation Valves (Level II) no impact to the 5b

f. Containment Sump Valves application.

U.S. Nuclear Regulatory Commission Page 46 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact

2. Sump plugging

3. Spray and humidity concerns from Internal flooding on RAB

4. Impact of MSLB on AFW TDP inlet valves, MSIVs, SG PORVs and Steam Dumps

5. HVAC and room heat-up.

The above items are resolved as follows:

1. Based on FSAR, the design basis accident valve temperature maximum is 245°F for LBLOCA. All components in containment are qualified to operate after reaching this temperature. Generally, MAAP runs indicate that with 1 train of sprays and 1 train of fan coolers, the containment temperature remains below this value. For scenarios without fan coolers and sprays, the temperature typically exceeds 300°F if there is a LOCA or loss of secondary heat removal.

1a: RHR shutdown cooling valves are only modeled for small LOCAs (including transient induced and SGTR) and successful secondary heat removal. Based on report RSC-06-13, containment conditions will not exceed 200°F and will not exceed the evaluated EQ temperatures.

1b: RCS PORVs are required to open following the initial pressure challenge following a plant trip. For the initial pressure challenge, the containment environment is normal and no concerns of environment effects exists. In the event a PORV fails to reclose, the block valve would be closed early in the event before the containment conditions would

U.S. Nuclear Regulatory Commission Page 47 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact appreciably change. The RCS PORVs and Block valves are challenged for feed and bleed cooling for three modeled cases, plant transients, S1 LOCA, and transient induced LOCAs. The PORVs are also challenged for depressurization following a SGTR. MAAP run RSC-CALMAP-2001-0118 indicates for a loss of feedwater, with no containment cooling, the containment upper region temperature reaches 302°F at the time of feed and bleed recirculation. Because feed and bleed are implemented early in the accident, the valves operate before the containment conditions exceed the analyzed limits. If the operator fails to open the PORVs, no credit is given for success. With the fan coolers and sprays operating, the containment conditions remain below the analyzed temperatures. No specific analysis was performed for the long term operability of the PORVs to perform bleed cooling because the probability of failure of the fan coolers or sprays in addition to CSIPs is a small contribution compared to the need of the operator to reset SI and maintain long term instrument air to the containment.

Therefore, containment cooling for PORV operability is considered to be non-risk significant.

1c: Instrumentation in containment provides SG level, pressurizer level and pressure, and containment pressure signals that are important to the PSA. Other signals such as core thermal temperature are not specifically addressed in the PSA. During the early portions of the accident, actuations will occur before significant changes to the containment environment occur. For large and medium LOCAs, long term effects on instrumentation inside

U.S. Nuclear Regulatory Commission Page 48 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact containment are not as critical as the instrumentation for LHSI flow and RWST level which are not affected by the accident. For small LOCAs and transient, if secondary heat removal remains intact, the containment temperatures are analyzed to stay below the EQ analyzed limits even without the containment cooling systems. For loss of secondary heat sink scenarios with feed and bleed cooling underway, the important instrumentation for success involves the RWST level. Steam generator level is not needed.

Instrumentation in containment would remain within EQ analyzed limits until feed and bleed cooling is underway. If pressurizer level and pressure are not available after initiation of feed and bleed cooling, the analyzed success path is to overfill the pressurizer and relieve liquid through the PORVs. For this success path, the availability of pressurizer instrumentation is not significant.

1d: If fan coolers operate, then the containment conditions remain within the EQ analyzed limits during core damage mitigation phase. For severe accident characterized by core damage sequences, the uncertainty of the availability of fan coolers is discussed in Section 8, Containment Response Assessment.

1e: Containment isolation valves operate early in the accident scenarios prior to conditions that would exceed the EQ analyzed limits.

1f: The containment sump valves are outside containment and the post-accident environment is expected to remain within the EQ analyzed limits.

U.S. Nuclear Regulatory Commission Page 49 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact 2: Sump plugging is modeled in the PSA for LOCA and no further action is needed.

3: The internal flooding analysis identified targets of concern. There is no further action required.

4: The equipment in the main steam line tunnel is analyzed in the EQ analysis for design basis main steam line break so that components and instruments of interest, AFWTDP steam emission valves, MSIVs, MS PORVs and SRVS would not be adversely affected by the accident.

5: The PSA includes an HVAC system notebook and supporting room heat-up analysis appendix that identifies those plant areas that require cooling or ventilation for equipment operability concerns. Other F&Os address specific rooms. A discussion of the above concerns has been added to the appropriate system notebooks to close out this F&O.

HNP-F/PSA-0028 R3 Table 4.1 It was determined that combining Appendix D and Resolution included list plant Safety function this Section 4 is not desirable. For clarity of understanding the adding table does not list the Key important information is provided in Section 4. The bases documentation of Safety Functions as defined in and documentation of the bases are provided in the the success criteria 18-SC-A4 the ASME standard. These supporting Appendix D. Modifications to these bases in for key safety 18-SC-C1 safety functions seemed to be Appendix D need only to be addressed in Section 4 if functions to each 18-SC-C2 addressed, but there is not a revised analyses indicate that the information provided in event tree and clear trail. Section 4 is invalidated. This reduces the amount of effort adding MAAP needed to update the model. Changing this approach references. There would mean a major revision to both sections. is no impact to the 5b application.

U.S. Nuclear Regulatory Commission Page 50 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact The resolution for this F&O includes.

For SCA4, In conjunction with AS08 the success criteria for the key safety functions were added for each event tree in Section 4.0. The success criteria of the key safety functions are not unique to each modeled initiating event.

For SC-C1, Actions for AS08 meet this requirement.

For SC-C2, the documentation as updated for the F&Os meets the requirement. Appendix D is improved by adding specific references to the MAAP runs as part of F&O 19.

MAAP analysis is not Specific MAAP runs were cited in Appendix D. Resolution of this referenced or documented F&O is addressed clearly. Appendix D sections were cited in Section 4.0. by providing 19-SC-B1 guidance for the 19-SC-B4 This will reduce the amount of revisions required the next layout of specific 19-SC-C2 time the MAAP runs are updated. The documentation MAAP references.

cross-referencing is as follows: There is no impact to the 5b Section 4 >> App. D >> App. F. application.

Section 4 and Appendix D do EOP references were cited for each safety function along Resolution of this not clearly connect EOP to the with the safety function success criteria in section 4.0. The F&O added EOP success criteria and event trees EOP references are already in Appendix D for the major references for each actions discussed. safety 20-SC-A6 function/success criteria. There is no impact to the 5b application.

U.S. Nuclear Regulatory Commission Page 51 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact Review SGTR and AFW The AFW long term availability is already addressed in Resolution of this success criteria to ensure Section 4.0. The reference to the Appendix G calculation F&O added cross stable end states after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months has been added regarding CST inventory and depletion. references linking An analysis for RWST depletion is already included in relevant analysis.

Appendix D.8.3, A cross reference was added to There is no impact section 4.0. to the 5b application.

The model assumes core damage based on failures during the mission time, regardless of if the core damage occurs after the end of the mission time. Therefore, the mission time measures the time that mitigating functions are available and therefore, stable. This is already addressed 21-SC-A5 in Section 10. No mission times are extended beyond 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to preclude core damage. If RWST inventory runs out prior to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , but core damage occurs after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , core damage is assumed based on a failure to maintain a stable state. However, available time to recovery RWST makeup is based on the available time prior to core damage regardless of the mission time.

Additionally, a calculation reference for success of branch events #XS-CC and #DS-CC was added. The citation is for Appendix G calculation RSC 06-13 HNP RHR Cooling Study Update. This document supersedes RSC-97-15.

Verify that engineering Engineering Judgment was reviewed as part of the Resolution of this judgment is used appropriately. Uncertainty analysis, and the results of the evaluation are F&O updated included in the Uncertainty Analysis appendix. documentation of 22-SC-B2 the uncertainty analysis. There is no impact to the 5b application.

U.S. Nuclear Regulatory Commission Page 52 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact In Appendix D the success A survey of PSA success criteria was solicited from Documentation was criteria is compared to similar Summer, Beaver Valley 1 & 2, Farley 1 & 2, and North updated to resolve plant designs, this is integrated Anna 1 & 2. The survey results for PWRs were reviewed this finding. There throughout the document, but and changes were incorporated into the comparison of is no impact to the the data is from IPE information plants success criteria throughout Appendix D. Tables for 5b application.

23-SC-B5 and should be updated with comparison of S2 and S1 LOCA criteria were added also.

current information.

Comparisons did not result in any HNP PSA model changes.

Uncertainty analysis may need Ninety-three (93) assumptions were identified in Section 4 Resolution of this to be improved. and Appendices D, F, and G. They were evaluated to F&O included determine their effect on PRA results. The 93 assumptions additional analysis apply to either accident sequence definition or success and documentation criteria. Because accident sequences and success criteria of assumption and are interrelated in the PRA document, they were not uncertainty. No separated in the identification and evaluation of further analysis 24-SC-C3 assumptions. needed for the 5b application.

An additional 48 assumptions were identified for Appendices H (ISLOCA) and R (LOSP Recovery). They were evaluated for their effect on PRA results.

Results of the evaluation are included in the Uncertainty Analysis appendix.

Not all of the information System notebooks were reviewed to determine consistency Instrumentation requested by this requirement in detail of required instrumentation for operation of the modeling and was included in the model and system. Variations in level of detail are appropriate for level documentation 25-SY-A3 the notebooks. Need to include of instrumentation required for each system. within system minimum instrumentation notebooks were required for successful control, The BE file was reviewed to identify instrumentation updated, and all

U.S. Nuclear Regulatory Commission Page 53 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact this is going to be helpful for fire modeled but not in the notebooks. New modeled gaps were closed.

risk analysis. No information instrumentation to provide placeholders for multiple hot There is no impact found about component shorts was added to the system notebooks. to the 5b operability and design limits, application.

need to add this to the Tech. Spec.s were reviewed to identify imitating conditions notebooks. of operation on each systems instrumentation to determine modeling deficiencies. No new instrumentation was identified for discussion in the system notebooks.

System Descriptions were reviewed to identify the available control room instrumentation. Information on design basis and limits of instrumentation is omitted from notebooks because PSA assumptions are that systems operate as designed and DBDs are generally cited in the system notebooks reference section. In addition, instrument related trips of components are discussed in system notebooks.

The following changes were made:

A.1 HHSI - Added instrumentation available to identify a loss of letdown cooling event. The discussion on loss of letdown cooling was expanded. Added instrumentation related to letdown LOCA to RCDT and charging with letdown isolation induced PZR PORV LOCA.

A.2 RHR - No change A.3 PSI - Minor correction to control discussion A.4 ESFAs - Added list in instrumentation section of instrumentation modeled including new instrumentation

U.S. Nuclear Regulatory Commission Page 54 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact added for fire concerns.

A.5 AFW - Added SG level instrumentation discussion used to control AFW flow and listed instrumentation. Added discussion on SG pressure instrumentation that can isolate AFW. Same instrumentation was added to the ESFAS system.

A.6 CFC - No change A.7 SG Relief - Added SG level and pressure instrumentation discussion. Added discussion on SG pressure instrumentation that can isolate main steam.

Same instrumentation is modeled in the ESFAS system.

A.8 RCS - Expanded discussion on PZRs level and pressure instrumentation. Discussed which PZR instrumentation is modeled in the ESFAS system.

A.9 ESW - Added instrumentation related to room cooling A.10 CCW - Organized discussion and added instrument loop for letdown cooling valve.

A.11 EDG - No change A.12 VDC - Added sentence stating that there is no control/protective instrumentation for PSA purposes.

A.13 Inst Power - Added sentence stating that there is no control/protective instrumentation for PSA purposes.

U.S. Nuclear Regulatory Commission Page 55 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact A.14 Instr Air - Added to discussion that all instrumentation providing control and trips are within the compressor and dryer component boundaries. Listed trips and instruments.

Added tag numbers to header pressure instrumentation and PORV accumulator instrumentation available in the control room.

A.15 HVAC - Added instrumentation for switchgear room AHU flow switches as required for new switchgear room cooling model.

A.16 CI - No changes A.17 CT - Added modeled and non-modeled instrumentation A.18 DW - Corrected discussion about automatic control of RWMST level.

A.19 AC - Added pointer to ESFAS for undervoltage sensing relays. Added discussion on breaker position indication on MCB and why not modeled.

A.20 MFW - Identified instrumentation that would cause trips of pumps, Listed instrumentation used to control SG level but that did not need to be modeled.

A.21 NVDC - Added sentence stating that there is no control/protective instrumentation for PSA purposes.

U.S. Nuclear Regulatory Commission Page 56 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact A.22 CVCS - Added instrumentation related to boration, letdown, makeup, charging, and seal injection.

A.23 NSW - No change SR says: PERFORM plant Plant walkdowns and interviews have been conducted for Resolution of this walkdowns and interviews with every major development of the PSA. For the IPE, these F&O added system engineers and plant walkdowns included spatial and HRA walkdowns to name a Walkdown operators to confirm that the few. The pertinent information was included in the system documentation to systems analysis correctly analysis and HRA analysis or other relevant analysis allow for the reflects the as-built, as- documentation. For internal flooding, the insights from the information to be operated plant. flooding walkdowns were included in Appendix F. For the more readily IPEEE, walkdown sheets are retained in the historic available. There is No checklists, interview sheets, documentation. no impact to the 5b etc. were found to indicate what application.

was done. Suggest they be In general for applications, such as Maintenance Rule, found or developed. SDP, or MSPI, information obtained by interview or walkdown is documented with the analysis or application.

26-SY-A4 It should be noted that interview and walkdown documentation is not considered to have the authoritative weight as does plant specific design documents and are therefore considered secondary with regard to determining a modeling approach.

In an effort to make walkdown information more readily accessible, Appendix W, Walkdown Documentation, has been created to document information from walkdowns made to meet this requirement and for the fire analysis and other plant walkdowns that can be documented in the future.

U.S. Nuclear Regulatory Commission Page 57 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact Need to augment discussion of A discussion of component boundaries was added to Resolution of this boundaries and data Appendix B. F&O included development to clarify whether adding component and how data were developed There are explicit type codes and data development sheets boundary separately for valve bodies and for generic and plant specific data associated with whole discussion and operators when those two valves and valve operators for both demand and time details regarding elements are modeled dependent standby failures. Plant specific data was permissives and separately. collected for AOVs, HOVs, and MOVs. interlocks to the documentation.

Need to check, augment Section 10 paragraph 5.11 states that interlocks are There is no impact modeling of permissives and modeled explicitly. Interlocks may be pressure/temperature to the 5b interlocks. See for example permissives or valve position limit switches. Failures would application.

pressure interlock for RH-1, include, drifting, failure to open/close, spurious operation, RH-2, containment sump or miscalibration. If the interlock is between two valves, the suction valves, charging recommended approach is to model the valves required to 27-SY-A8 suction, etc. change positions. If the limit switches are unique to the 27-SY-B11 interlock and not to the installed valve operation, then the limit switch needs to be modeled uniquely. This is especially important when the limit switch is exposed at a different frequency than the installed valve.

The RHR hot leg suction valves interlocks include a low pressure permissive (RCS pressure (PT-402, 403), closed valve interlocks for the RWST (1SI-322 and 1SI-323) and CSIP suction (1RH-25 and 63).

The CSIP suction from RHR has the reverse interlocks on RHR hot leg suction and the alternate mini-flows must be closed.

Interlocks between CSIP suction paths to the VCT or

U.S. Nuclear Regulatory Commission Page 58 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact RWST include a permissive for the respective train VCT valve to close once the RWST valve is open.

ESW valves are interlocked with the associated pump.

Failure of a pump to start is a minimal failure. Failure of the valve to change state due the interlock is counted as a valve failure.

SR says: INCLUDE in either the This F&O has been addressed by resolutions for AS-17 Resolution of this system model or accident A section to each notebook for environmental effects on F&O added sequence modeling those equipment has been added. documentation for conditions that cause the environmental system to isolate or trip, or effects. There is no those conditions that once impact to the 5b exceeded cause the system to application.

fail, or SHOW that their exclusion does not impact the results. For example, conditions that isolate or trip a system include:

28-SY-A17 (c) adverse environmental conditions (see SY-A20)

No evidence was found that adverse environmental conditions other than floods were evaluated. For example, it is unclear whether an evaluation was performed to confirm that a primary PORV would work in a post-LOCA

U.S. Nuclear Regulatory Commission Page 59 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact environment (f&B is credited for small LOCAs). Perform and document evaluations.

SR says: IDENTIFY system This F&O is largely addressed by F&O AS-17. Resolution of this conditions that cause a loss of F&O involved desired system function, e.g., The system notebooks already include a discussion of the adding additional excessive heat loads, mechanisms for instrumentation induced trips of the documentation.

excessive electrical loads, systems; no further action is required in that regard. There is no impact excessive humidity, etc. to the 5b Heat loads considerations are addressed in the HVAC application.

No evidence identified that this analysis, and the system notebook system dependencies was performed for excessive include HVAC as a dependency. Switchgear room cooling humidity. Some evidence found was addressed under separate F&O.

that other conditions were evaluated. Update / upgrade Excessive heat loads due to phenomenological issues such 29-SY-A19 evaluations and documentation as MSLB and LOCA are addressed in F&O AS-17.

The effect of excessive heat loads on other equipment due to equipment failures is not considered to be credible because plant areas are generally large or open where there are large motors.

Excessive electrical loads can occur if a motor overloads or if additional loads are aligned in a non-design configuration. For motor overloads, breakers include protective trips that protect the motors and the bus. These protective trips are not typically modeled due to their low failure probability. Motor overloads are modeled in the

U.S. Nuclear Regulatory Commission Page 60 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact component boundary as reliability data. There are no recovery actions that align equipment in an unanalyzed configuration.

To protect the EDG from starting overload, the bus loads shed. This is explicitly modeled in the PSA due to consequence. Bus failure by short circuit is another bus overload mechanism that is modeled. There are no overload conditions identified outside of these existing conditions.

Regarding excessive humidity, the resolutions to F&O AS-17 addresses phenomenological humidity concerns within the containment and main steam tunnel. For LOCA and MSLB, the equipment in those areas are already analyzed for humidity, temperature, and radiation as part of the licensing basis. There are no other areas where excessive humidity are expected for accidents modeled in the fault trees.

For accidents developed by a specific analysis, ISLOCA and Internal flooding, the effects of humidity are bounded by the existing analysis.

The flooding analysis included assessment of the impact of floods and sprays on equipment. Humidity from internal flooding is not a limiting concern due to low temperature water. Water depth or spray is a limiting concern. Piping failures following an accident are not considered credible and are not modeled, so concerns for humidity are excluded.

U.S. Nuclear Regulatory Commission Page 61 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact For ISLOCA, no credit for mitigations of a pipe break in the RAB is given, so no further analysis is required for excessive heat or humidity loads.

The above discussions were added to the system notebooks where appropriate.

A discussion of flooding and humidity effects from RWST pipe break flood was added to the RHR system notebook.

A statement to the effect that the effects of flooding, spray, and humidity are addressed in the internal flooding notebook and was added to system notebooks for ESW, NSW, CSIP, CCW, AFW and ac-power.

SR says: TAKE CREDIT for Resolution for AS-17 addresses containment component Resolution of this system or component operability for LOCAs and main steam tunnel component F&O added the operability only if an analysis operability for MSLB. necessary exists to demonstrate that rated references and or design capabilities are not NPSH issues for small LOCA with no containment heat additional exceeded. removal have been analyzed in MAAP analysis. For calculations. There example, two bounding cases, RSC-CALMAP-2001-1112 is no impact to the 30-SY-A20 Operability considered in sense for LBLOCA and RSC-CALMAP-2001-1116 for S1 LOCA 5b application.

30-SY-B15 of can be used successfully, both were run with no containment cooling and not Tech Spec definition. demonstrated that NPSH was not an issue to the RHR pump suctions. This is verified by absence of the Consider PORV operability, inadequate NPSH flag in the MAAP run summary files.

containment sump level indication, etc. in post-LOCA Another analysis, (RSC-06-13) indicates that with environment. Also, failure of secondary heat removal and no containment cooling, the

U.S. Nuclear Regulatory Commission Page 62 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact containment heat removal HHSI pumps can draw from the RHR pump heat systems (spray, fan coolers) is exchangers with no CCW cooling. The latter of these not modeled as causing calculations is discussed in the success criteria notebook, inoperability of SSCs in Appendix D. These calculations have been referenced in containment. Evaluate SSCs in the RHR and HHSI notebooks.

containment to ensure that they could continue to function in NPSH issues for HHSI on a loss of VCT has been added to this environment. the system notebooks as an assumed failure. A calculation of the time to VCT overheating or loss of level was added See also SY-B17: (may need to to appendix G.

address steamline break environment in some plant RHR suction swaps automatically on RWST level, not spaces) IDENTIFY SSCs that containment sump level, so sump level indication failure may be required to operate in from environmental concerns is not a failure of RHR. A conditions beyond their discussion was added to the RHR notebook regarding environmental qualifications. containment environment effects on RHR system signals.

INCLUDE dependent failures of multiple SSCs that result from Room cooling concerns are addressed by specific operation in these adverse analyses.

conditions. Examples of degraded environments include:

(a) LOCA inside containment with failure of containment heat removal (b) safety relief valve operability (small LOCA, drywell spray, severe accident) (for BWRs)

(c) steam line breaks outside containment (d) debris that could plug

U.S. Nuclear Regulatory Commission Page 63 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact screens/filters (both internal and external to the plant)

(e) heating of the water supply (e.g., BWR suppression pool, PWR containment sump) that could affect pump operability (f) loss of NPSH for pumps (g) steam binding of pumps SR says: IDENTIFY spatial and Plant walkdowns and interviews have been conducted for Resolution of this environmental hazards that every major development of the PSA. For the IPE, these F&O added may impact multiple systems or walkdowns included spatial and HRA walkdowns to name a documentation of redundant components in the few. The pertinent information was included in the system plant walkdowns.

same system , and ACCOUNT analysis and HRA analysis or other relevant analysis There is no impact for them in the system fault tree documentation. For internal flooding, the insights from the to the 5b or the accident sequence flooding walkdowns were included in Appendix F. For the application.

evaluation. IPEEE, walkdown sheets are retained in the historic documentation.

Example: Use results of plant walkdowns as a source of In general for applications, such as Maintenance Rule, 31-SY-B8 information regarding SDP, or MSPI, information obtained by interview or spatial/environmental hazards, walkdown is documented with the analysis or application.

for resolution of spatial /

environmental issues, or It should be noted that interview and walkdown evaluation of the impacts of documentation is not considered to have the authoritative such hazards. weight as does plant specific design documents and are therefore considered secondary with regard to determining No evidence of plants a modeling approach.

walkdowns (checklists, evaluations, etc.) were found. In an effort to make walkdown information more readily Need to develop these. accessible, Appendix W, Walkdown Documentation, has been created to document information for meeting this

U.S. Nuclear Regulatory Commission Page 64 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact requirement and for documenting other walkdowns.

SR says: DO NOT USE Four cases of crediting proceduralized and non- Resolution of this proceduralized recovery actions proceduralized recovery were found to be credited or F&O included as the sole basis for eliminating assumed insignificant contributors. They include: updating the room a support system from the heat-up analysis model; however, INCLUDE 1. Loss of Switchgear Room cooling: The original and documentation.

these recovery actions in the assumption was that operator had time to open doors to There is no impact model quantification. For preclude room heat-up and no model was necessary. A to the 5b example, it is not acceptable to detailed room cooling analysis was performed and the application.

not model a system such as determined requirements for success were included in the HVAC or CCW on the basis model. The potential for procedure changes is in review to that there are procedures for reduce the impact. An operator action with value of 1.0 was dealing with losses of these added to determine the potential risk reduction of potential systems. procedure changes.

32-SY-B13 This was apparently done for 2. Loss of ESW pump room cooling: Existing conservative the switchgear rooms. Need to room heat-up calculations indicate a loss of ventilation will correct. Check if other cause the ESW pumps to overheat. This analysis was instances occur. assumed to be conservative and not best estimate. Based on engineering judgment, room cooling was not modeled.

In the absence of a detailed room heat-up analysis, it was considered prudent to include the requirement in the model. Given there is no procedural guidance or HVAC analysis, there is large uncertainty if success is possible through alternate ventilation methods. An operator action with probability of 1.0 was added to determine the potential importance and benefit of performing further analysis.

3. Loss of letdown cooling: There was an assumption that loss of letdown cooling would be mitigated by operations

U.S. Nuclear Regulatory Commission Page 65 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact and not affect CSIP operation. However, further review indicated this is a potential loss of CSIP suction if CCW is lost to the letdown heat exchanger, resulting in introduction of steam to the VCT. Calculations for available response time show that this is a potentially significant event. The model was updated to model this failure and take credit for existing procedural guidance for manual letdown isolation.

4. Loss of VCT level control. A similar assumption was found for LT-112/115, VCT level. However, spurious high level would result in a flow diversion of VCT makeup to the RHT. If operators failed to manually isolate the flow diversion, this would also fail auto swapover to RWST so a failure of the running CSIP would occur on loss of suction as the VCT depletes. The presence of only one train of level indication on the MCB exacerbates this failure if that is the train that malfunctions to the high level state. Like item 3, the available response time is very limited. The model was updated to model this failure and take credit for existing procedural guidance, as well as simulator training on this scenario.

Documentation was updated in system notebooks and Room heat-up analysis accordingly.

SR says: DOCUMENT the Resolution includes: Resolution of this system functions and boundary, F&O included the associated success criteria, (d) No specific action is required; system boundary and reviewing LERs 33-SY-C2 the modeled components and success criteria are documented. Other F&Os will address and adding failure modes including human operability considerations and assumptions. In general, documentation.

actions, and a description of design basis is adequate. There is no impact modeled dependencies to the 5b

U.S. Nuclear Regulatory Commission Page 66 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact including support system and (e) LERs were reviewed and information was added to application.

common cause failures, system notebooks including the inputs, methods, and results. For example, this (f) There is some component spatial information in the documentation typically system notebooks. The fire walk downs are using a check includes: list that includes spatial information. This will be documented as a separate Appendix W. If considered (d) information and calculations relative to internal or flooding events, additional information to support equipment will be added to the system notebooks.

operability considerations and assumptions (o & p) The current fault tree structure does not support (e) actual operational history system model quantification. The existing system indicating any past problems in notebooks include system level insights. The summary the system operation document includes relative component and system level (j) component spatial importance. No other action is required.

information (o) results of the system model (q) The system notebooks already include a complete list of evaluations references. The check list from the walk downs will be (p) results of sensitivity studies included in a separate Appendix.

(if used)

(q) the sources of the above information, (e.g., completed checklist from walkdowns, notes from discussions with plant personnel)

Several of these elements were not found. Add to system notebooks.

U.S. Nuclear Regulatory Commission Page 67 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact SR says: DOCUMENT the key Four hundred twenty-five (425) assumptions associated Resolution of this assumptions and key sources with the systems analysis (Appendices A.1 through A.23) F&O compiled all uncertainty associated with the were identified. They were evaluated to determine their assumptions and systems analysis. effect on PRA results. added 34-SY-C3 documentation on Key sources of uncertainty Results of the evaluation are included in Appendix U, uncertainty were not identified. Add this Uncertainty Analysis. analysis. There is information. no impact to the 5b application.

The pre-initiator human A new calculation was created to upgrade pre-initiator Resolution of this reliability analysis does not F&Os to ASME/ANS Standard Capability Category II. F&O created a new appear to meet HLR-A in calculation to several aspects. This F&O update pre-initiator applies to all aspects of the pre- HRA to meet the initiator analysis. Standard. There is no impact to the 5b

1. There is no evidence of a application.

rigorous search for pre-initiator alignment errors caused by 52-HR-HLR- maintenance and testing A activities as required by A1.

2. There is no evidence of a search for single activities which can affect two redundant trains as required by A3 and B2.

3. The screening process for mis-calibration errors is not identified as required by A2.

U.S. Nuclear Regulatory Commission Page 68 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact

4. The quantification process for the mis-calibration errors does not provide enough detail to be reproducible. The quantification process does not meet the requirements of D2, D3, D4, D5, D6, D7. Based on these observations, the original pre-initiator analysis does not meet the ASME criteria. A complete pre-initiator analysis should be done from the beginning.

Spreadsheet shows several The NSAC method provides detailed recovery event trees / Resolution of this RHR misalignments that are fault trees that incorporate the individual elements using F&O added quantified from NSAC-154. the SHARP1 and NUREG/CR-1278 methods for the documentation to These are not included in the ISLOCA pre-initiator HRA events. This is sufficient to meet justify the use of 53-HR-D1 final report. NSAC method may the objective of accounting for systematic methods of the NSAC method.

not be acceptable based on the determining pre and post initiator actions. There is no impact SR. to the 5b application.

No evidence of REVIEW of the The methodology used forces consistency given scenario Resolution of this HFEs and their final HEPs context, procedures, operational practices and timed F&O included relative to each other to check responses. The final HFEs and HEPs were reviewed review and their reasonableness given the against each other for consistency. documentation of 54-HR-G6 scenario context, plant history, final HFEs and procedures, operational HEPs. There is no practices, and experience was impact to the 5b performed. application.

U.S. Nuclear Regulatory Commission Page 69 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact From the Table in Annex B of HEP error factors in the CAFTA database were adjusted. Resolution of this Appendix E, it appears the error Calculation was updated to explain special cases for error F&O included factors of the HEP values are factors at end of summary Table E-6. adjusting HEP error estimated at 10 if the HEP is factors in the less than 1E-3 and 5 if the HEP CAFTA database is greater than 1E-3. This is and updating OK< except for some HEP documentation.

greater than .1. In fact, there There is no impact are 2 HEP at .5, with an EF of to the 5b 5, which makes the upper application.

55-HR-G9 bound on a lognormal 3.15.

Some HEPs for Cr actions have EF of 3.

Suggest the EF on HEP > .1 be changed to 3 and EF on HEP>

.5 be changed to 1.0. If these are input into UNCERT with a lognormal distribution, the answers will be incorrect.

Tables E3 and E4 allow a Cr Credit for non-proceduralized recoveries WITHOUT Resolution for this type event to be used if there is JUSTIFICATION was removed from calculation and model. F&O removed all no procedural guidance Element HR-H2(a) allows for justification if procedures or non-proceduralized directing the action. This is in training are lacking for an HRA. Non-proceduralized recoveries without conflict with the SR recoveries OPER-22 and OPER-42 remain with justifications. There 56-HR-H2 requirements for Category II. It justification. is no impact to the is not known if any Cr actually 5b application.

lack a corresponding procedure. This is a comment about the HNPPRA process for identification of Cr actions.

U.S. Nuclear Regulatory Commission Page 70 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact Suggest revision of Table E-3 to eliminate credit for actions with no procedure.

A review of the HEP OPER-7 Comment Resolution: Resolution of this dependency combinations F&O addressed generated several comments Clarify the wording to: issues concerning on the resulting combinations. the individual and Upon the loss of Component Cooling Water, the Reactor combination OPER-7: The use of this HFE Coolant Pumps require seal injection in order to avoid a operator actions.

and the definition of the HFE Reactor Coolant Pump Seal LOCA. The major contributor Documentation was should be made clear, which to this is the spurious closure of the Motor Operated Valves updated to reflect may result in different timing 1CS-165 or 1CS-166. This action recognizes that the the updated and a different numerical result. running CSIP would fail on loss of suction and that a analyses. There is The HFE appears to be a pre- suction flow path would need to be re-established and the no impact to the 5b initiator action. The action alternate CSIP would need to be started in order to application.

57-HR-H2 occurs when a VCT valve preclude a seal LOCA.

spuriously closes to the CSIP suction without the concurrent The event addresses the potential that the flow path to the opening of a RWST valve. The running charging pump can be lost without a corresponding running pump is assumed to fail swap of the CSIP source to the RWST. The VCT valve due to loss of suction. The closes and blocks flow. The loss of suction will fail the operator has 8 minutes to running pump if not acted upon within 5 minutes. The diagnose the loss of suction, failure of the operators to identify the loss of suction before provide alternate suction to the starting the standby CSIP is assumed to result in the failure other pump and start the of both CSIPs. The timing assumes the pump is running.

standby pump. The calc sheet The most critical timing is for the loss of Reactor Coolant states a 13 minute time Pump seal injection with a loss of CCW at the same time.

window, which would imply he In this case the operator has 13 minutes to get a CSIP

U.S. Nuclear Regulatory Commission Page 71 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact has a chance to save the running with a suction source. The response time includes running pump, which is not the review of alarm procedure (APP-ALB-006) and the RCP case. It should also be noted abnormal procedure AOP-0018. Then to determine the that the 13 minute time window cause of the alarm is low flow and determine the cause is is for loss of RCP seal cooling. loss of suction flow path. Thus 5 minute response time is This event only fails seal appropriate as walked through with operators on 11/1/05.

injection flow. CCW to the This AOP-018 directs the operator to Attachment 4.

thermal barrier is not affected Attachment 4 will also direct the establishment of a suction by this event. flow path.

OPER-67: There may be a typo The lack of water to the running pump should result in a in the calculation sheet for this lack of charging flow and a Charging Pump Disch Header event. The write-up lists 3 High-Low Flow alarm (ALB-6-1-1). The response for this procedures APP-ALB-023, alarm is to check high versus low and for low flow (charging AOP-26, AOP-22. AOP-22 relative to letdown), the suction alignment must be should be AOP-26. checked. However, it does not stop the pump and if the OPER-68: This event is not failure is due to valve faults (dominant case), it is used in the model and should reasonable to assume that the running pump will fail.

be removed from the However this alarm procedure will have to suction path documentation to avoid check and provides guidance to restore the suction path confusion. before starting the standby CSIP. Additionally the operators will enter AOP-003, Malfunction of Reactor Makeup OPER-70: This action is for Control, if the isolation of the VCT is due to instrument installation of the spare CSIP malfunction. If the VCT level is being maintained then the pump after failure of the running operator will be directed to recover using the VCT and pump. open the applicable suction valves.

Spare pump installation requires 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The Pc Note that OPER-42 is a closely related action that is calculation uses 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months as currently listed as non-proceduralized action. PATH-1 step the time available for the action 8 may provide the indication before the pumps fail.

to be completed (basis not

U.S. Nuclear Regulatory Commission Page 72 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact explained). In order for this OPER-42 wording:

event to be meaningful in the sequence of events, there must If an SI signal occurs the two supply valves from the be a sequence where the plant volume control tank (VCT) to close and the two supply can be without HHSI for 12 valves from the RWST to open. If the RWST valves do not hours, which is not likely. In open, the VCT will quickly run out of water and the CSIP addition, if the HFE is used pumps will fill with vapor. This can cause the failure of the during a LOCA or feed and CSIP pumps. It is assumed that the pumps can run without bleed, it must be shown that damage for ~5 minutes.

plant personnel have access to the CSIP room during the There are annunciators for pump trouble but that would not LOCA. If this is a pre-initiator occur until after pump failure. Also, Path-1 directs HFE, then it is not necessary, operators to check the valve alignment if insufficient because the unavailability of injection is present but it occurs too late to be effective.

the standby CSIP will include this contribution. The operator action addresses the potential that the operators will diagnose the situation based on valve OPER-30 & OPER-64: These positions for the CVCS valves that are displayed by red HFEs appear to be related and and green lights above each valve switch on the control should appear in the same board. The operators must notice that at least one of four of sequence of events. However, the valve positions are incorrect. It is believed that if they they have different timing, see one mis-positioned valve, they will check and correct procedural direction and the other three valves associated with this action as probabilities. OPER-30 is failure needed. The operators are trained to scan the control to establish long term injection board repeatedly as soon as they get an initiator. However, source to the RWST from the it is not assured that they would select the CVCS panel as BAT system. The HEP is 7E-5. their first selection.

OPER-64 is for opening Demin valves to the RMWST on Low OPER-67 Comment Resolution:

water level. The HEP is 1.7E-3. Typo corrected in documentation.

U.S. Nuclear Regulatory Commission Page 73 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact OPER-D21: Combination of OPER-68 Comment Resolution:

OPER 41 and OPER 30. The OPER-68 removed from the model.

write-up states there is an intervening success of HHSI OPER-70 Comment Resolution:

initiation. But this is not true. Plant has been modified and OPER-70 has been updated The HHSI is initiated very early since comment was made. Comment is no longer relevant.

in response to the SGTR. There Of other note, the execution phase of the recovery has are no obvious successes three more required actions for locking open 1CS-748 and between failure of OPER-41 1CS-749 and locking closed 1CS-747. All of the actions and OPER-30. The time except for closing the transfer switch have independent between these events is long verification resulting in a HEP of 3.7E-3 instead of 7.3E-3.

and may still lead to zero dependence. OPER-30 & OPER-64 Comment Resolution:

The nominal time for the OPER-30 action is 300 minutes OPER-D50: Combination of that would start within about 30 minutes of the SGTR OPER-11 (fail to switch inst. diagnosis; nominal time for OPER-64 is 432 minutes with a bus to back-up supply) and cue time at 399 minutes into the event. There are different OPER-26) fail to control AFW alarms for cues. There are no intervening successes in the from control room. The OPER scenario. The long time between actions could lead to the 26 event is for operation of events being considered independent.

AFW from the main control room. The OPER-11 event OPER-D21 Comment Resolution:

occurs after an instrument bus, The nominal time for the OPER-30 action is 300 minutes; as directed by ALB-015. With nominal time for OPER-41 is 120 minutes. There are no instrument power, the different alarms for cues. There are no intervening operator must go locally to the successes in the scenario. The long time between the AFW pump and manipulate the events results in the events being considered independent.

controls by hand. The correct HEP for this situation should be OPER-D50 Comment Resolution:

OPER-66, which has a OPER-11 and OPER-26 are mutually exclusive. The probability of .012 rather than recoveries using these two events should be given the HEP

U.S. Nuclear Regulatory Commission Page 74 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact 8E-4 (for OPER-26). This is not for the combination of OPER-11 and OPER-66. OPER-D50 an HRA problem, but is more is addressed in rule recovery file using OPER-D20 value of likely a problem in sequence 7.8E-3. Currently neither OPERD50 nor OPER-D20 show development. up in the cutsets at a 1E-11 truncation level.

The OPER-26 HFE should not OPER-T3 Comment Resolution:

be used in the same logic The need to prevent failing the CSIP pumps on loss of combination as the OPER-11 suction occurs before taking manual control of AFW event. outside of the control room. Order is correct.

OPER-T3: Combination OPER- OPER-T5 Comment Resolution:

35 (Fail to Manually Start AFW Agree that it could be removed from rule recovery file, but from control room); OPER-42 keeping in case it shows up under increased truncation or Align CSIP Suction for SI; with components out of service.

OPER Take Local Control of TDAFWP. The timing of OPER-Q11 Comment Resolution:

these events as stated in the The logic is convoluted. The spurious SI results in a seal dependency analysis is not LOCA which eventually requires SI to actuate to prevent correct. The first event is core damage.

OPER-35. The second event should be OPER-66. If both of OPER-Q13 Comment Resolution:

these fail, the next step is feed Recovery has a intervening successful local start of the and bleed whereupon the TDAFW pump OPER-66.

failure of CSIP suction may occur. See OPER-T4 for correct OPER-Q14 Comment Resolution:

order of similar events. The Harris thermal hydraulics analysis has a success path with secondary cooling and just RHR recirculation without OPER-T5: Does not appear in service water.

the RAW base cutsets. Should be removed from documentation.

U.S. Nuclear Regulatory Commission Page 75 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact OPER-Q11:

OPER-38 fail to manually initiate SI OPER-42 fail to align CSIP suction OPER-20S fail to reopen RCP CCW isolation valves OPER-4 fail to open containment sump valves.

This combination of events is strange. The write-up states the initiating event is a plant trip with spurious SI. If so, then there is no reason to initiate SI.

The sequence results in a seal LOCA. Same comment applies to Q-12.

OPER-Q13:

OPER-35 fail to manually start AFW OPER-46 fail to align MFW OPER-26 fail to control SG level OPER-3 fail to start feed and bleed.

This combination does not make sense, unless there is an

U.S. Nuclear Regulatory Commission Page 76 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact intervening success of OPER

66. The first two events fail all AFW, which leaves no place for OPER26.

OPER-Q14:

OPER-10 Fail to start ESW OPER-19 Failure to start NSW OPER-9 Failure to initiate RCS cooldown OPER-17 Fail to establish HHSR This combination appears does not make sense, unless it appears in a non-minimal LERF sequence. The first two events cause loss of all ESW. That means CCW is also lost. A seal LOCA occurs, with no way to remove decay heat.

All valves and other non-major Plant specific data scope was expanded to include: The use of plant-equipment uses generic data specific data was grouped by type. Category II CSIP swing pump, CCW swing pump, motor-operated expanded, and requires consideration for valves, pneumatically-operated valves, hydraulically- additional service conditions. operated valves, pressurizer PORVs, SG PORVs, heat components were 35-DA-B1 exchangers, batteries, battery chargers, instrument air included in the data compressors, startup transformers and axial fans. analysis There is no impact to the 5b Service conditions are accounted for by using plant specific application..

data. For larger populations, such as MOVs, service

U.S. Nuclear Regulatory Commission Page 77 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact conditions were evaluated for cases with raw water and demineralized water. There were no failures associated with raw water and thus, no insights gained by separating the grouping. MOV failures are dominated by motor-operator failure and not valve body failures. In general, the service conditions are best modeled by modeling the exposure time. Both demand failure rates and time dependent failure rates were developed. The criteria for use of the different failure rates were updated in Appendix B, Data, and in the Section 10, Ground Rules.

No discussion of outliers was There were no outliers found with regard to component Resolution of this found in the documentation. unavailability. With regard to exposure times and F&O updated environmental effects, the system notebooks are existing considered the appropriate location to document outliers. documentation with regard to outliers.

The system notebooks include tables with component There is no impact exposure times and reference test or inspections. These to the 5b tables were made current and each notebook includes a application.

discussion of infrequently tested equipment. There are no other outliers identified, other than infrequently tested 36-DA-B2 components.

Components tested at least every refueling cycle are not considered outliers.

The most notable outliner is the ESW to AFW supply line.

These valves are periodically inspected but over a long period of time. Their valve bodies are modeled separately.

No action was required, except to update the current documentation.

U.S. Nuclear Regulatory Commission Page 78 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact HNP uses generic parameter Generic estimates are only used for the ESFAS system. Documentation on estimates for unavailability. These values are well documented. the unavailability estimates used was The ESW headers generic unavailability is subtracted from clarified. There is 37-DA-C1 the system train plant specific unavailability so that the total no impact to the 5b remains plant specific. The remainder of the unavailability application..

data is plant specific. No further action is required for this F&O.

Very little discussion could be In general, pump starts were taken from OSI PI. Only one Resolution of this found about rules for counting start is counted if multiple starts occurred during the same F&O added demands and failures. test (assumed less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months apart.) clarification on the collection of 38-DA-C5 For valve strokes, only the test stoke that is timed or used demands and to place the component in service is counted. Returning the failures. There is no valve to its normal standby alignment is not counted during impact to the 5b a test. (This detail is assumed based on time between application.

strokes in OSI-PI, typically less frequent than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .)

The correlation of tests that are Discussion was added to correlate collected test data to Resolution of this counted to model failure modes modeled failure modes by defining failures that constitute a F&O added is not clear PSA functional failure, and a non-PSA functional failure. discussion on the Specifically, out-of-spec. stroke times are not PSA failures. correlation of Component trips and subsequent restarts with no collected test data 39-DA-C10 corrective action are not failures. This type of information to the respective was included in Appendix B. failure modes.

There is no impact to the 5b application.

U.S. Nuclear Regulatory Commission Page 79 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact There is no discussion of Coincident maintenance unavailability at HNP is controlled Resolution of this coincident maintenance except by the 12 week rolling schedule. Generally, only coincident F&O updated for the identification of mutually maintenance within the same train is permitted. The coincident exclusive events protected train flags are mutually exclusive and prevent unavailability cross train maintenance in the model results. The potential documentation.

for coincident maintenance in the same train is calculated There is no impact in the fault tree as the product of two or more unavailability to the 5b events. This is the extent of the analysis and no application.

documentation is required.

Coincident maintenance unavailability is identified for swing trains with the A or B train components for CCW pumps and CSIPs. Data was collected for the coincident time 40-DA-C13 periods and subtracted from the independent maintenance unavailability. The coincident events are modeled explicitly and are mutually exclusive with the independent events.

Another collected coincident unavailability is the alignment of the PORV block valves.

The coincident unavailability documentation was updated or added in Appendix B and the respective system notebooks.

The discussion of mutually exclusive events are documented in system notebooks.

U.S. Nuclear Regulatory Commission Page 80 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact No discussion or evaluation HNP meets the requirements of DA-D6. No action was Resolution of this was found regarding the required. F&O included the consistency of CCF data to addition of HNP component boundaries About half of the MGLs come from NUREG/CR-5497 and discussion on the other half from RSC 01-17. The underlying data used in component NUREG/CR-5497 is not available. RSC 01-17 provides for boundaries. There many components a generic composite based on is no impact to the information related to similar component groups. To the 5b application.

degree possible, common cause components were mapped to equivalent HNP failure events. The following 59-DA-D6 improvements were made:

A discussion of component boundaries was added.

Data source references were verified and corrected as necessary.

Component, MGL selection was corrected in some cases.

Updated MGLs data was used from NUREG/CR-5497 if available.

There is limited discussion of The reliability data collection windows is based on available Resolution of this the basis for determined data OSI-PI data and extrapolated back to initial time period that F&O included collection windows. maintenance rule functional failures were recorded. A adding a section on discussion of the reliability data time window was added to the reliability data Appendix B. time window in the 41-DA-D7 documentation.

The availability data windows are based on MR There is no impact unavailability performance monitoring group data and are to the 5b not extrapolated. The goal was to collect as much data application.

between 1996 and the end of 2005. Because the initial

U.S. Nuclear Regulatory Commission Page 81 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact start date of recording unavailability varied from system to system, start windows varied from 1996 to 2000. Some unavailability data is no longer collected so that the window ended prematurely of 2005.

Appendix B includes a discussion of the various unavailability time windows.

SR says: ESTIMATE the Normal uncertainty analysis has been run on revised Resolution of this mean CDF from internal model. F&O included events, accounting for the assessment of the state-of-knowledge Reviewed cutsets to identify cutsets with multiple failures of SOKC and the correlation between event the same type (TCs and CCFs) to determine any uncertainty probabilities when significant uncertainty correlations. There were no cutsets identified analysis.

42-QU-A2b (see NOTE (1)). with the same type failure occurring in multiple systems. Documentation was Therefore, no variations in service conditions were updated. There is No evidence found that this was reviewed to determine potential effects on the failure no impact to the 5b commonly done. Investigate probabilities of similar components in different systems. application.

and redevelop data as required.

No outliers were identified.

Supporting Requirement QU- Documentation of the comparison has been added to the Resolution of this D3 requires: HNP PRA Summary Document. F&O added The quantification results documentation of compared to other similar results. There is no plants and identify causes for impact to the 5b 46-QU-D3 significant differences. For application.

example: Why is LOCA a large contributor for one plant and not for another. The Quantification Notebook and other Notebooks do not have any evidence that

U.S. Nuclear Regulatory Commission Page 82 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact this was done.

A more robust sensitivity Forty-four (44) assumptions associated with data Resolution of this analysis related to key (Appendix B) were identified and evaluated to determine F&O included assumptions is needed. their effect on PRA results. Results of the evaluation are review of included in Appendix U, Uncertainty Analysis. assumptions and 60-QU-E43 documentation of No assumptions were identified for the quantification uncertainties. There process itself (Section 5). is no impact to the 5b application.

1. The development of the The Level 2 analysis evaluated the status of the Low Resolution of this event trees may not be Pressure Injection (LPI) for each Plant Damage State F&O included an sufficient to uniquely identify the (PDS) by a manual review of the dominant cutsets of each update of the conditions required for Level 2 PDS (stated on page 76 of Section 8 of the PRA). Level-2 analysis analysis. This deficiency Therefore, for the baseline Level 2 analysis, the and development of pertains to the ability to discern determination of LPI availability is accurate for all additional operability status of ECCS. The potentially significant LERF PDSs. 49-LE documentation.

Level 1 event trees ask There is no impact functional questions for events To exactly determine the status of the LPI system, another to the 5b such as H, D, X, G. Functional top event would need to be added to the Containment application.

58-LE-A4 failure could be caused by Safeguards Event Tree (CSET), with an additional question operator error, front line system about LPI status added to the end of each branch of the or support system. CSET. This would double the number of CSETs from 18 to 36, thereby doubling the number of PDSs from 344 to 688.

2. The Level 1 sequence is then Because the 344 PDSs are already a very large number to assigned to a single CDB, consider, it is undesirable to double this number to 688.

depending on the set of failures Therefore, some simplification is utilized in the Level 2 assigned to represent the analysis, but the following evaluation demonstrates that sequence. simplification is acceptable.

U.S. Nuclear Regulatory Commission Page 83 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact

3. The 28 CDBs are expanded There are two independent trains of RHR, which provide into 344 PDS, through a the means for late LPI. If support systems are available, the process which was not failure probability of the RHR system is dominated by the documented for this review. common cause failures (CCFs) to start, run or valve failure.

From gate LINJ of HNP_06.CAF, these CCFs have

4. A split fraction for LERF is probabilities of 2.00E-4, 2.50E-5, and 2.20E-5, then calculated for each PDS respectively. The combined CCF probability of RHR is and the split fraction is therefore 2.47E-4. Even if the system failure probability assigned back to the Level 1 were on the order of 1E-3 (e.g., if one train was unavailable sequence. It is not clear how due to maintenance), the impact on the total LERF would the 344 PDS split fractions are be negligible when this probability is multiplied by the assigned to the 28 CDBs. frequency of the PDSs. For example, a PDS with a frequency of 1E-6 would have a maximum possible effect

5. It is also not clear how on LERF of 1E-3

1E-6 = 1E-9. The real impact would be operability of low pressure even lower.

systems is considered when they are not asked on the level Given that the independent failure of RHR is not a 1 event tree. significant issue, the real question is how are supporting systems status considered for a given cutset. To consider

6. It is not clear how the this question, the similarity of supporting systems between dependence of OPER-IV is RHR and Containment Spray injection are examined. Per maintained through the Level 1 gate S007 of HNP_06.CAF, the Containment Spray Pump and Level 2. A power dependencies are 125 VDC bus DP-1A-SA and 480 VAC bus 1A2-SA. Per gate L094, the dependencies for RHR Pump A are the same. Per gates S016 and L096, the Containment Spray Pump B and RHR Pump B also share power supplies (125 VDC bus DP-1B-SB and 480 VAC bus 1B2-SB). Since the Containment Spray injection is considered in the CSET end states, success of Spray injection means that the power supply for Spray injection is available, so power supply for RHR must also be available.

U.S. Nuclear Regulatory Commission Page 84 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact If the Containment Sprays failed to inject, then it is possible that the failure was due to support system failure. For the Harris Level 2 analysis, it will therefore be assumed that CSET end states M, N, O, P, Q and R (each has failure of spray injection) should indicate LPI unavailability. This is partially conservative in that the sprays can fail independently which would not cause LPI failure, but this conservatism is considered small and acceptable.

The Containment Event Tree (CET) in HNP_LEVEL_2-slm.XLS (sheet PDSFLAG), was then examined. As seen in cells T33 to T376, LPI is not credited (FL-LPI = 1) for any PDSs in which Containment Spray Injection is failed (CSET end states M, N, O, P, Q, R), except for 7M, 7P, 12M, 15M, all of which are shown in the spreadsheet in bold red, indicating that they contained no cutsets above the truncation. The only exceptions are B16N and B18R, which are not shown in red, but have FL-LPI set to 0. These two PDSs are SGTR, and the LPI status is not relevant.

Therefore, to ensure that LPI is not credited in any PDS for which Containment Spray injection has failed, HNP_LEVEL_2-slm.XLS (sheet PDSFLAG) cells T153, T156, T243, T297, T340 and T363 are all changed from 0 to 1.

Regarding the question of dependence of OPER-IV with operator actions in the Level 1 analysis, these actions can be considered independent. OPER-IV represents the operator action to open the PORVs late in an accident sequence when core damage is likely (high core exit thermocouple temperature). During an event, operators are

U.S. Nuclear Regulatory Commission Page 85 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact repeatedly guided to check the status of the critical safety functions, and a loss of core cooling directs to EOPFRC-C.1. The procedure repeatedly checks the status of secondary side cooling, which is failed in high pressure sequences. The cue for opening a Pressurizer PORV (core exit thermocouple temperature >1200°F) is independent of other cues for Level 1 operator actions, and the timing of OPER-IV is separate (i.e., later) from the Level 1 actions.

Therefore, OPER-IV is considered independent of the level 1 operator actions.

This evaluation has been added to PRA Section 8.5.2.1.

The requirement is to perform a 1. Per gate #WR of the HNP_06.CAF fault tree, the MSIV Resolution of this realistic evaluation of failure to close contribution to SG isolation failure is 2.4E-3. F&O added a secondary side isolation The contribution from the other component failures is discussion of capability for SGTR accidents. (1.3E-3 + 7.7E-3 =) 9.0E-3. Therefore, the MSIV failure analysis done for There are two observations contributes only 2.4E-3 / (9.0E-3 + 2.4E-3) = 21% of the evaluation of discussed here. isolation failure probability. A detailed modeling fission secondary side product behavior in the TSV and main condenser would be isolation capability

1. SGTR (IE) events with MSIV difficult and contain large uncertainties since the systems for SGTR fail to close are considered a would be exposed to conditions and particles for which they accidents. There is 48-LE-D4 LERF. No credit is given for the were not designed. Some portion of the 21% of isolation no impact to the 5b retention capability of the TSV, failures would still result a large, early release condition (a application.

and main condenser. 50% success of holdup would yield approximately a 10%

Considering the SGTR reduction in SGTR LERF). Therefore, given the difficulty of sequences are 80% of LERF, modeling of fission product holdup and the fact that the this may be a conservative reduction in SGTR LERF would be small, the conservative assumption. The HNP-PRA assumption is considered acceptable.

meets Category 1 in this regard. 2. The Induced SGTR model developed in response to

U.S. Nuclear Regulatory Commission Page 86 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact Observation 49 evaluated the probability of a SGSV

2. For SGTR (IE) events with sticking open during the core damage progression. Event 2 the SGSV cycling, the release of the ISGTR event tree (PRA Section 8.3.2, Figures 8.9 is assumed to be small, based and 8.10) evaluated the probability of a stuck open SG on MAAP analysis (which is relief or safety valve, given the large number of cycles acceptable). However, the (assumed to be 70 cycles) through the core damage probability of the SGSV failing progression. The ISGTR model evaluated the chance of a to reclose is based in the Level stuck valve occurring in any of the three SGs to be 0.15 (all 1 analysis. The duty cycle and SGs intact=0.85). Since there are 3 SGs, the probability environment of the SGSV that the stuck open valve would be in the SG with the during the core melt processes SGTR would be 0.15/3, or 0.05. Therefore, those SGTR is not considered. The sequences in which there was no initial failure of SG probability of .0077 for SGSV isolation (Level 1 event tree top event W) will be assigned a fail to reclose may not be 0.05 conditional probability of SG isolation failure during the appropriate for the Level 2 progression to core damage. This applies to SGTR accident space. This aspect of sequences RPY, RBX, RBH, RUG, and RUP (i.e., CDBs analysis does not meet the SR. B1, B3, B6, and B16). Note that sequence RUB is not included because it is already binned as a large bypass.

Failure of the long term SG isolation will be treated as having the same effect as if the SG isolation had failed prior to core damage as is modeled in the Level 1 event tree. Therefore, in the CET (file HNP_LEVEL_2.XLS sheet CETEXEC), the IFL split fraction is modified to assign a 0.05 conditional probability of a stuck relief valve, given no stuck relief valve from the Level 1 model.

This discussion has been added to PRA Section 8.5.2.5.

U.S. Nuclear Regulatory Commission Page 87 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact The Level II categorically An Induced SGTR (ISGTR) model was developed and Resolution of this dismisses induced SGTR included, based on the guidance in NUREG-1570, Risk F&O included the during an accident, stated in Assessment of Severe Accident-Induced Steam Generator development of an 8.3.4 of HNP Report. A Tube Rupture, USNRC, March 1998. Section 8 of the PRA ISGTR model.

qualitative reason is provided - documentation and the Level 2 results have been updated There is no impact water will remain in the loop to reflect the ISGTR model and results. to the 5b 49-LE-D5 seal, preventing natural application.

circulation. RCP will not be started unless there is water in the secondary side of the SG.

This position is not substantiated for all sequences considered.

There are 2 HFE events in the The OPER_IV probability has been analyzed with a Resolution of this Level 2. They are OPER-IV (fail detailed HRA evaluation, which is added to Section 8.7 of F&O includes an to depressurize and inject LPI the PRA document. The value has been updated in PRA evaluation of and OP_H2REC (failure to Section 8.5.2.1 and Table 8.9. OPER_IV and preclude hydrogen burn development of following recovery. There are 3 Per Section 8.5.2.9 of the PRA, OP_H2REC actually additional other recovery events, which represents the probability of success in preventing a documentation.

represent failure to recover ac- hydrogen burn (0.999 probability of failure, not 0.001 There is no impact power. These are based on the probability of failure as indicated in Observation 50), which to the 5b 50-LE-E1 OSP recovery curve from is essentially 100% failure. The event exists in the model application.

historical experience and are only to allow for sensitivity analysis. Therefore, no further not a concern for this F&O. evaluation of OP_H2REC is needed.

OPER-IV has a probability of .1 and OP_H2REC has a probability of .001. These probabilities are qualitatively assessed with no basis given

U.S. Nuclear Regulatory Commission Page 88 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact for the selection of probability.

This does not meet the SR.

The definition of LERF states Per Section 9 of the PRA, the Early definition for Harris is Resolution of this that any sequence for which not strictly within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of core damage, but state that F&O referenced release occurs within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of alternative definitions can be justified based on event discussion on the core uncovery is early. For sequence characteristics. For example, long-term loss of definition of what timing of Early vs LATE, some decay heat removal sequences in which a long time exists early means as plants use the emergency plan between event initiation and core damage may be related to HNP to identify when a site considered as non-large early release scenarios, if the LERF. There is no emergency would be declared facility emergency plan would allow time for emergency impact to the 5b and evacuation would be protective action prior to radionuclide release. Additionally, application.

started. This is often prior to if time periods of less than four hours (from vessel breach) core uncovery. can be justified based on plant specific procedures and emergency response features, an alternate definition can 43-LE-E3 Using an earlier starting point be used.

for evacuation than core uncovery, may result in the late Applying this definition to SGTR, it is unlikely that the SSO SGTR would declare a General Emergency (GE) preemptively for sequences from being LERF. a SGTR. The Harris EAL flowpaths have a provision to declare a GE if AFW flow of 210 kpph is not available with RVLIS <62%, but other than this, the general approach is not to declare a GE until 3 fission product barriers are breached.

There are provisions for declaring a Site Emergency (SE) for some equipment unavailability, but this would not necessarily lead to any public evacuations. In the event of a GE, the Harris Plant merely provides recommendations

U.S. Nuclear Regulatory Commission Page 89 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact to State and County authorities for sheltering and evacuation. These authorities can choose to act on the recommendations, act early on the recommendations or not act at all. Given the uncertainties surrounding the question of when the SGTR sequences would result in a general emergency, the conservative assumption of Early vs. Late timing is considered appropriate.

The SR requires review of 1) In the new Section 8.8, the LERF contributors are Resolution of this contributors for identified and ranked. F&O included reasonableness. The HNP PRA 2) The Level 2 uncertainties are evaluated through the use various does not meet this SR on of sensitivities in improvements on several counts. Sections 8.1.4 for MAAP and the new Section 8.8.3 for the the level 2 analysis.

rest of the LERF. There is no impact

1) The contributors are not 3) The new Section 8.8.3 varies selected parameters for to the 5b identified or ranked. which assumptions and uncertainty were believed possible application.

2) There are no uncertainties to have a significant impact on LERF associated with the level 2 4) The new Section 8.8.3 provides the results by varying analysis. key assumptions.

44-LE-F1 3) There are no sensitivity 44-LE-F2 studies performed which vary a) It is difficult to defend a SGTR with a stuck open relief the parameters (although valve no being LERF, but in any case, this assumption justification for the selected does not significantly affect the results. Release Category 5 values is given). represents SGTR with a stuck relief valve and some

4) There are no alternate scrubbing; RC 5C is the same but no scrubbing. Per results showing the effect of HNP_

SUMMARY

_2006.XLS, worksheet CNMT FAILURE, alternate assumptions. the frequency of RC 5 is 1.66E-7/yr, compared to 9.07E-7/yr for RC 5C. Therefore, even if some of RC 5 There are some important were justified to not be part of the LERF, it would not assumptions in the Level 2 significantly alter the results or conclusions. Given the which could affect results: difficulty of justifying RC 5 as non-LERF, the current assumptions are appropriate.

U.S. Nuclear Regulatory Commission Page 90 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact a) all SGTR CDF sequences with SO SRV are LERF. b) The statement that all SGTR CDF sequences with SRV b) All SGTR CDF sequences intact are not LERF is incorrect. Per Section 9, all SGTR with SRV intact are not LERF. sequences were binned to LER.

c) Induced SGTR are not possible in any case. c) Induced SGTR has been added to the model in the new d) Scrubbing is ultimately not Section 8.3.2.

credited.

e) Containment overpressure d) Scrubbing is credited in CET top event SRCS.

failure is always underground at Containment overpressure failure location sensitivity to the base mat, resulting in a LERF was examined in sensitivity #2 in the new Section filtered released. 8.8.3. The LERF is not sensitive to the assumption because of the very small contribution to LERF from non-bypass releases.

The SR requires the It is true that the spreadsheet methodology does not allow Resolution of this identification and for a detailed uncertainty analysis of the Level 2 results. F&O provided characterization of contributors However, the majority of uncertainty in Level 2 analyses is justification for the to LERF. The method employed due to the phenomenological uncertainties, for which existing method at HNP-PRA for LERF does not quantitative uncertainty bounds usually have no basis. and references to retain the individual elements of Therefore, the method used is to evaluate the key the appropriate the LERF process. The LERF phenomenological uncertainties through sensitivity documentation.

split fractions are developed analyses. These sensitivities are presented in the new There is no impact 45-LE-F3 through a spreadsheet Section 8.8.3. to the 5b quantification process, which application.

does not retain event names or uncertainty factors. The split fractions are then attached to the appropriate Level 1 sequences. The Level 2 contributing elements are not tracked through the

U.S. Nuclear Regulatory Commission Page 91 of 91 HNP-15-083, Enclosure 1 Table 10. Historical, Closed Internal Events F&Os From Peer Reviews Conducted Prior To The 2007 Peer Review Of Record.

F&O # Finding Resolution 5b Impact quantification process, so the results cannot be manipulated to examine the importance of each one to the overall results.

For example, it is possible to find the contribution of DCH to LERF. It is not possible to find the uncertainties associated with each Level 2 issues.

Document the quantitative The following text is added to Section 8.0 of the PRA: Resolution of this definition used for significant F&O added accident progression The Level 2 assessment is developed to analyze the documentation sequences. If other than the significant accident progression sequences. The ASME about the definition used in Section 2, PRA Standard (Reference 24) defines a significant quantitative Justify the alternative. accident progression sequences as: one of the set of definition for accident sequences contributing to large early release significant accident The HNP PSA needs to frequency that, when rank-ordered by decreasing progression document the quantitative frequency, aggregate to a specified percentage of the large sequences. There definition and provide early release frequency, or that individually contribute more is no impact to the 51-LE-G6 justification if other than Section than a specified percentage of large early release 5b application.

2 definition. frequency. For this version of the standard, the aggregate percentage is 95% and the individual percentage is 1%.

However, since the Level 2 model must be developed before the relative sequence contribution to LERF can be determined, the HNP Level 2 analyzed all of the Level 1 core damage sequences and Plant Damage States (see Section 7) that had cutsets greater than the truncation limit used in the model. No sequences were eliminated from the calculations.

v t e Letter Sequence Supplement
TAC:MF6583, Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5B (Open)
Initiation Request, Request Acceptance Supplement, Supplement Administration Questions Answers 5 February 2016 Results Approval... Other:
MONTHYEAR2015-09-29 [Table View]

HNP-15-083, Supplement to Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program

Text

Subject:

References:

1.0 INTRODUCTION

SUMMARY

Navigation menu

HNP-15-083, Supplement to Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program

Text

Subject:

References:

1.0 INTRODUCTION

SUMMARY

Navigation menu

Search