HNP-13-053, Submittal of Path Forward for Resolution of Generic Safety Issue (GSI) - 191

Submittal of Path Forward for Resolution of Generic Safety Issue (GSI) - 191

ML13136A146
Person / Time
Site:	Harris
Issue date:	05/16/2013
From:	Kapopoulos E Duke Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
HNP-13-053, SECY-12-0093, TAC ME1234
Download: ML13136A146 (13)

Text

Ernest J. Kapopoulos, Jr.

Vice President Harris Nuclear Plant 5413 Shearon Harris Road New Hill NC 27562-9300 919.362.2502 May 16, 2013 10 CFR 50.4 Serial: HNP-13-053 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400

Subject:

Path Forward for Resolution of Generic Safety Issue (GSI) - 191

References:

1. Generic Letter (GL) 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors

2. December 23, 2010, Staff Requirements - SECY-10-0113 - Closure Options for Generic Safety Issue - 191, Assessment of Debris Accumulation on Pressurized Water Reactor Sump Performance

3. October 12, 2011, Pressurized Water Reactor Owners Group (PWROG), Topical Report (TR) WCAP-16793-NP, Revision 2, Evaluation of Long-Term Core Cooling Considering Particulate, Fibrous and Chemical Debris in the Recirculating Fluid

4. May 4, 2012, Nuclear Energy Institute (NEI) to the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Reactor Regulation, Director, Division of Safety Systems,

Subject:

GSI-191 - Current Status and Recommended Actions for Closure

5. July 9, 2012, SECY-12-0093 - Closure Options for Generic Safety Issue - 191, Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance

6. November 15, 2012, NEI Letter to NRC, Office of Nuclear Reactor Regulation, Director, Division of Safety Systems,

Subject:

GSI-191 - Revised Schedule for Licensee Submittal of Resolution Path

7. November 21, 2012, NRC Letter to NEI, Nuclear Regulatory Commission Review of Generic Safety Issue-191 Nuclear Energy Institute Revised Schedule for Licensee Submittal of Resolution Path

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

9. April 8, 2013, NRC Letter to PWROG, Final Safety Evaluation for Pressurized Water Reactor Owners Group Topical Report WCAP-16793-NP, Revision 2, Evaluation of Long-Term Cooling Considering Particulate Fibrous and Chemical Debris in the Recirculating Fluid (TAC No. ME1234)

Ladies and Gentlemen:

Duke Energy Progress, formerly known as Carolina Power & Light Company, herein submits the resolution path forward and schedule for resolution of Generic Safety Issue (GSI)-191 for the Shearon Harris Nuclear Power Plant, Unit 1.

Serial: HNP-13-053 Page 2 In Reference 4, the NEI highlighted the current industry status and recommended actions for closure of GSI-191 which were based on licensees providing a docketed submittal to the NRC by December 31, 2012, that would outline a GSI-191 resolution path and schedule pursuant to the Commission direction Reference 2. By Reference 6, NEI recommended to the NRC that licensees delay submittal of GSI-191 resolution path and schedule until January 31 , 2013, or 30 days following placement of both the Commission response to Reference 5 and the NRC staff safety evaluation (SE) for Reference 3. In Reference 8, the Commission approved the staff's recommendation in Reference 5 to allow licensees the flexibility to choose any of the three options discussed in the paper to resolve GSI-191 . Further, the Commission encouraged the staff to remain open to staggering licensee submittals and the associated NRC reviews to accommodate the availability of staff and licensee resources . The SE, Reference 9, for Reference 3 was made publicly available by the NRC on April 16, 2013.

An industry template was developed by NEI for the identification of a resolution path and schedule, and to describe defense-in-depth and mitigation measures to support the proposed resolution schedule.

The NEI template was used for the development of Attachment 1 for Shearon Harris Nuclear Power Plant, Unit 1, which provides a resolution path forward and schedule for resolution ,

summary of actions completed for GL 2004-02, and defense-in-depth and mitigation measures which will be established and maintained throughout the resolution period.

This letter contains new regulatory commitments as identified in Attachment 2.

If you have any questions or require additional information, please contact Dave Corlett, Manager, Regulatory Affairs, at 919-362-3137 .

1 declare under penalty of perjury that the foregoing is true and accurate.

Executed on May 16, 2013.

Sincerely, Attachments

1. Resolution Path Forward

2. List of Regulatory Commitments cc: Mr. J. D. Austin , NRC Sr. Resident Inspector, HNP Ms. A. T. Billoch Colon, NRC Project Manager, HNP (e-mail only)

Mr. V. M. McCree, NRC Regional Administrator, Region II

HNP-13-053 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400 Attachment 1 Resolution Path Forward

Serial: HNP-13-053 Page 1 of 8 Option 2a: Deterministic Resolution Path Introduction Shearon Harris Nuclear Power Plant, Unit 1, (HNP) selected Option 2a and intends to pursue refinements to evaluation methods and acceptance criteria. To support use of this path and continued operation for the period required to complete the necessary analysis and testing, HNP has evaluated the design and procedural capabilities that exist to identify and mitigate in-vessel blockage. A description of these detection and mitigative measures are provided later in this document. Additionally, a summary of the existing margins and conservatisms that exist for HNP are also included in this document.

Characterization of Current Containment Fiber Status From the debris generation and debris transport analysis, HNP has determined that approximately 495 lbs (includes intact blankets) of fibrous debris could be transported to the strainers (limiting fibrous debris case), as documented in HNP calculations SD-0025, and HNP-M/MECH-1205. Based on previously performed strainer bypass testing, the total quantity of fiber assumed to bypass the strainer is 495 lbs x 9.15% = 45.3 lbs. This is based on maximum single sump flow through one of HNPs two recirculation sumps assuming a single failure of one Residual Heat Removal (RHR) pump. If there are no pump failures, which is a more limiting scenario in terms of bypass, the bypass mass is estimated to be twice this amount. Considering there are 157 fuel assemblies (FA) in the HNP core, and taking no credit for the percentage of fiber that would recirculate through the containment spray system, this equates to an approximate value of 262 g/FA [(45.3 lbs (20548 g)/157 assemblies) x 2].

HNP plans to follow the Pressurized Water Reactor Owners Group (PWROG) efforts to establish acceptable limits for in-vessel debris and is providing the previously determined values of in-vessel fiber. The PWROG Comprehensive Analysis and Test Program Generic Safety Issue (GSI)-191 Closure (PA-SEE-1090) is designed to develop success criteria for long-term core cooling in-vessel effects. The program will include analyses and testing to develop a set of debris-related limits to ensure long-term core cooling that may provide less restrictive in-vessel debris limits than WCAP-16793, Revision 2. At the time the PWROG establishes new acceptance limits, HNP will evaluate previously performed bypass testing to determine whether additional testing is required, or if re-analysis of those results can be performed to demonstrate acceptable in-vessel debris limits.

The fibrous debris sources considered in these analyses include Nukon, Thermal Wrap, and latent fiber.

Characterization of Strainer Head Loss Status HNP previously provided the results of strainer head loss testing, including the impact of chemical effects, in letter HNP-11-006 (ADAMS accession no. ML110400028). The results of this testing demonstrate acceptable results with regard to allowable head loss.

Characterization of In-Vessel Effects As noted above, HNP intends to follow the resolution strategy proposed by the PWROG for establishing in-vessel debris limits for the type of plant design that exists at HNP.

Serial: HNP-13-053 Page 2 of 8 Licensing Basis Commitments HNP currently has a commitment to report to the NRC how it has addressed the in-vessel downstream effects issue per the guidance contained in the NRC letter dated September 29, 2008 (ML082540269) which is due within 90 days following the issuance of the final NRC Safety Evaluation on WCAP-16793-NP. This submittal closes that commitment. As a result of the remaining open questions associated with Generic Letter (GL) 2004-02 for HNP and the information contained within this document, the previously established commitments are considered to be closed based on the intended direction to be taken to resolve GSI-191 in-vessel downstream effects as described in this document. New commitments as a result of this closure effort are listed in Attachment 2.

Resolution Schedule HNP will achieve closure of GSI-191 and address GL 2004-02 per the following schedule.

• HNP has completed fibrous insulation measurements on Feedwater piping, Reactor Coolant System (RCS) Loop piping, and the lower portions of the Steam Generators. If additional measurements for insulation replacement are needed based on the results of the PWROG comprehensive test program, they will be completed by the end of the first refueling outage following the issuance of the staffs safety evaluation endorsing the results of the PWROG comprehensive test program. Test results are expected by mid-2014 with development of the staffs safety evaluation to follow. Based on these schedules it is expected that any additional measurements will take place during HNP Refueling Outage 19, which is currently scheduled for the Spring of 2015, to limit dose for taking measurements to as low as reasonably achievable. In addition, HNPs design and procurement milestones for Refueling Outages 18 & 19, the dates estimated for availability of PWROG test results, and the expected schedule for staff review of these test results makes large-scale insulation replacement prior Refueling Outage 20 not feasible.

• Following determination of new acceptance limits for in-vessel debris, which will occur after issuance of the staffs safety evaluation of the results from the PWROG Comprehensive Closure Program, HNP will evaluate previously performed bypass testing to determine whether additional testing is required, or if re-analysis of those results can be performed to demonstrate acceptable in-vessel debris limits.

• HNP will complete the necessary insulation replacements, remediation, or model refinements by the completion of the third refueling outage following the staffs issuance of the safety evaluation for the upcoming PWROG Comprehensive Test Program to satisfy the newly established in-vessel debris limit. This is based on HNPs debris loading and conservatively assumes that all fiber within the Zone of Influence (ZOI) would have to be replaced. If the upcoming PWROG testing produces a fiber limit substantially higher than the current 15 g/FA limit then this may result in requiring a lesser amount of insulation be replaced and earlier completion of the project. The insulation replacement or remediation is expected to begin in the Fall 2016 outage (Refueling Outage 20). This schedule is based upon the expected availability of results from the PWROG test program and completion of the associated staff review. The project will focus on replacing or remediating the greatest quantity achievable within the currently scheduled outage duration, with the remainder to be replaced or remediated within the following outage (Refueling Outage 21).

Serial: HNP-13-053 Page 3 of 8

• Within six months of establishing a final determination of the scope of insulation replacement or remediation, HNP will submit a supplemental response to support closure of GL 2004-02. This determination will be made following the staffs issuance of the safety evaluation for the upcoming PWROG comprehensive test program.

• HNP will update the Final Safety Analysis Report (FSAR) following NRC acceptance of HNPs updated supplemental response and completion of the identified removal or modification of insulation debris sources in containment per plant modification procedures and processes (10 CFR 50.71(e)).

If HNP determines that a proposed testing or analysis resolution path will not be viable, then an alternate resolution path will be discussed with the NRC to gain acceptance of the proposed path and to establish an acceptable completion schedule.

Summary of Actions Completed To Address GL 2004-02 To support closure of GSI-191 and to address GL 2004-02, HNP completed the following actions:

• Replaced original containment sump strainers which had a convoluted screen geometry that had a total effective surface area of 796 ft2 (398 ft2per sump), with nominal 1/8 inch diameter openings with high-performance top hat style assemblies having a total net effective surface area of approximately 6,000 ft2 (3,000 ft2 per sump), with nominal 3/32 inch diameter openings.

• Replaced Min-K insulation that could be damaged by a Loss of Coolant Accident (LOCA) with Reflective Metal Insulation (RMI).

• Performed a latent debris survey to quantify and characterize latent debris in containment. The survey determined the amount of latent debris to be 117.9 lbs. This is less than the 200 lbs assumed in the debris generation analysis.

• Completed debris generation and debris transport analyses. These analyses are documented in HNP calculations SD-0022 (Minimum Post-LOCA Containment Water Level), SD-0023 (Debris Generation) and SD-0025 (Debris Transport).

• Completed ex-vessel downstream effects analysis. This analysis is documented in HNP calculation HNP-M/MECH-1205 and follows the methodology set forth in WCAP-16406-P, Evaluation of Downstream Sump Debris Effects in Support of GSI-191.

• Completed prototypical sump strainer testing to support Emergency Core Cooling System (ECCS) pump Net Positive Suction Head (NPSH) analysis and quantification of strainer structural margin. Testing was completed in the Fall of 2010 and was conducted consistent with the guidance in NRC Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Strainer Head Loss and Vortexing (ADAMS Accession No. ML080230038). The test results are documented in HNP calculation SD-0026 and the NPSH analysis for the RHR and Containment Spray (CT) pumps are documented in HNP calculations SI-0043 and CT-0026 respectively. Test results were also transmitted to the NRC in letter HNP-11-006 (ML110400028).

Serial: HNP-13-053 Page 4 of 8

• Implemented programmatic and procedural changes to maintain acceptable configuration and protect the design and licensing basis. Those changes include:

o HNP limited further installation of fibrous insulation.

o Design and Review Considerations for Mechanical Design Discipline reviews of design change products was revised to include a screening criterion to prompt a reviewer to consider if potential sources of debris are being created or altered which could interfere with ECCS suction from the recirculation sump.

o The amount of loose, buoyant materials that can be brought into containment during online entries is limited to 40 ft2 total cross-sectional area, to stay within strainer headloss calculation assumptions.

o The containment closeout procedure and surveillance procedure for establishing containment integrity were revised to add verifications that no loose debris (rags, trash, welding slag, grinding debris, insulation debris, etc.) are present in containment that could be transported to the containment sumps. The containment integrity procedure also contains steps to assess the quantity of latent debris and also ensure drainage paths to the sumps are not blocked.

o The procedure for pre-job briefs was revised to provide guidance on ensuring the importance of containment cleanliness is discussed and that the plants stringent criterion for containment cleanliness is communicated.

Summary of Margins and Conservatisms for Completed Actions for GL 2004-02 The following provides a summary description of the margins and conservatisms associated with the resolution actions taken to date. These margins and conservatisms provide support for the extension of time required to address in-vessel effects for HNP.

• Debris Generation o The amount of latent debris assumed (200 lbs) is conservative. This value bounds the value of 117.9 lbs determined as a result of a latent debris survey of Containment.

• Debris Transport o The debris generation calculation shows a size distribution for the fiberglass debris made up of fines, small pieces, large pieces, and large pieces with the jacketing intact. In the debris transport calculation it is assumed that the small pieces of Low Density Fiberglass (LDFG) (smaller than 6) can be conservatively treated as 1 clumps, and the large pieces of LDFG (larger than 6) can be conservatively treated as 6 pieces.

o It is assumed that all debris blown upward would be subsequently washed back down by the containment spray flow. Since all of the debris blown to upper containment was determined to be fines and small pieces, it is conservatively assumed that all of this debris would be washed to lower containment. This conservatively disregards any small piece debris held up on the 286 and 261 elevation gratings as it is washed down.

Serial: HNP-13-053 Page 5 of 8 o With the exception of latent debris washed to the sump and inactive cavities during pool fill-up, it is conservatively assumed that all latent debris is in lower containment, and would be uniformly distributed in the containment pool at the beginning of recirculation. This is a conservative assumption since no credit is taken for debris remaining on structures and equipment above the pool water level.

o Small and large pieces of insulation debris (RMI and fiberglass) not blown to upper containment were conservatively assumed to be uniformly distributed between the locations where it would be destroyed and the sump screens. This is a conservative assumption since the blowdown and the majority of the pool fill-up phases are multi-directional flows that would tend to disperse debris around containment (including areas with lower transport potential).

o No credit is taken for settlement of fiber fines. Fiber fines are assumed to transport 100% during recirculation.

• Strainer Bypass Testing o The fibrous debris used in the testing was all shredded into small pieces, but the predicted size distribution of the fibrous debris that is transported to the sump screen includes fines, small pieces, large pieces, and intact blankets.

• Chemical Effects o A value of 7,000 ft2 is assumed for concrete surface area in the Chemical Model.

This is a conservative assumption, as it bounds the sum of the area of epoxy coatings destroyed in a LOCA (1,578 ft2 for a zone of influence of 5L/D) and the concrete area that is covered by unqualified coatings that are assumed to fail (approximately 1,600 ft2).

o The Chemical Model uses the maximum water depth in containment to maximize the mass of precipitates. This is conservative, as the maximum water depth assumes the entire Refueling Water Storage Tank (RWST) volume is injected into containment, and the entire Reactor Coolant System is empty (except for the water in the Reactor Vessel that is below the nozzles).

• In-Vessel o The HNP core barrel is an upflow design. A feature of this design is an alternate flow path through the barrel-baffle region with communication holes in the baffle plate. Flow through this path should provide some amount of core cooling even in the event that the core inlet is completely blocked. It is noted that the Safety Evaluation Report on WCAP-16793, Revision 2, states that some licensees could potentially credit these alternate flow paths via plant specific evaluations however, it must be demonstrated that these flow paths would be effective. HNP realizes that an evaluation of this type would be needed to fully credit this flow path as increasing the HNP debris load limit. HNP has not yet determined whether an evaluation of this type is to be pursued. This item is only listed in this submittal to point out this design feature exists at HNP and was not considered in the testing which established the 15 g/FA fiber limit.

Serial: HNP-13-053 Page 6 of 8 o No credit is taken for the percentage of fiber that would recirculate through the containment spray system. No credit is taken for the settling of fibrous debris in the reactor vessel lower head.

o The current in-vessel fiber limit of 15g/FA is based on testing with other conservatisms as outlined in OG-12-287. The PWROG Comprehensive Test Program will establish more realistic debris limits.

Summary of Defense-In-Depth (DID) Measures The following describes the plant specific design features and procedural capabilities that exist for detecting and mitigating a fuel blockage condition.

Prevention Emergency Operating Procedure (EOP) EOP-ES-1.3, Transfer to Cold Leg Recirculation, contains a set of potential mitigating actions for symptoms of degraded recirculation sump performance. Included in these actions is direction to throttle RHR flow to reduce the flow rate from the recirculation sumps. The procedure states that if flow is throttled, it must be maintained above the value required to remove decay heat and that core exit thermocouples (CETs) and reactor vessel level instrumentation system (RVLIS) values must be monitored to ensure adequate flow is reaching the core. While not an action taken in direct response to symptoms of fuel blockage, this action can have a positive impact on reducing the potential for fuel blockage if large amounts of debris reach the strainer quickly and result in degraded strainer performance prior to any indications of core blockage. Reduced core flow has benefits such as reduced head loss through a debris bed at the core inlet and delayed onset of chemical precipitates. Controlling flow within the bounds of CET and RVLIS indications ensures that adequate fuel coverage and decay heat removal is maintained during the evolution.

Detection Fuel blockage from chemical and non-chemical post-LOCA debris would result in an inadequate core flow condition. If this were to develop, plant operators have the following primary parameters available to them to detect this condition in a timely manner:

1. Increasing core exit thermocouple indication:

CETs are monitored as part of the EOPs to ensure adequate flow is reaching the core to cool the core. There are two redundant channels for incore thermocouple indication.

Primary display for the CETs is provided by the Safety Parameter Display System (SPDS) located on the main control board for each channel. In addition, SPDS displays are located in the Technical Support Center and the Emergency Operations Facility. All of these displays are linked through the Emergency Response Facility Information System (ERFIS) computer from the Inadequate Core Cooling (ICC) Monitor microprocessor, which processes the temperature data.

2. Reactor Vessel Level Instrumentation System reactor water level indication:

RVLIS indication is monitored as part of the EOPs to ensure that the fuel has adequate water coverage. It consists of two redundant independent trains that monitor the reactor vessel water levels. Each train provides three vessel level indications: full range, upper range, and dynamic head. The full range RVLIS reading provides an indication of reactor vessel water level from the bottom of the vessel to the top of the vessel during

Serial: HNP-13-053 Page 7 of 8 natural circulation conditions. The upper range RVLIS reading provides an indication of reactor vessel level from the hot leg pipe to the top of the reactor vessel head during natural circulation conditions. The dynamic head reading provides an indication of reactor core, internals and outlet nozzle pressure drop for any combination of operating reactor coolant pumps. Comparison of the measured pressure drop with the normal, single phase pressure drop provides an approximate indication of the relative void content of the circulating fluid.

In EOP-FR-C.1, Response to Inadequate Core Cooling, and EOP-FR-C.2, Response to Degraded Core Cooling, indications and trends of these parameters are monitored to ensure an inadequate reactor core flow condition is detected, if it occurs, and to direct additional actions to be taken, such as starting a Reactor Coolant Pump (RCP), if limits are exceeded. Also, EOP-ES-1.4, Transfer to Hot Leg Recirculation, allows a transfer to hot leg recirculation based on recommendation from the Plant Engineering Staff that transfer is required.

Mitigation To mitigate an inadequate core flow condition the following methods are currently available to plant operators:

1. Transfer to hot leg recirculation:

The three loop Westinghouse design at HNP consists of one cold leg injection path per loop and one hot leg injection path per loop. Following a LOCA and upon receipt of a RWST low-low level signal in conjunction with a safety injection signal, the switch is made to ECCS cold leg recirculation. Within 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> following the beginning of the event, switchover to hot leg recirculation is made to terminate boiling in the core and to prevent boron precipitation in the core. Following this, a switch between hot leg recirculation and cold leg recirculation should occur approximately every 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the initiation of hot leg recirculation. Transfer to hot leg recirculation has the potential to disturb any debris bed that has formed on the bottom of the fuel, enabling the cold leg injection flow path to once again be a viable flow path. Having a hot leg injection path on each loop ensures flow will reach the core even in the event of a hot leg break on one loop. EOP-ES-1.4 also allows a transfer to hot leg recirculation based on recommendation from the Plant Engineering Staff that transfer is required.

2. Incrementally starting available reactor coolant pumps:

Procedure EOP-FR-C.1 directs the plant operators response to an inadequate core cooling condition. The first steps in the procedure involve verification of Safety Injection (SI) flow in all trains. While this is occurring, Step 3 directs the operators to establish conditions for running an RCP while continuing with the procedure. Additional steps direct operators to reduce the RCS pressure in order for the SI accumulators and low-head SI pumps to inject. If these actions are not successful in restoring adequate core cooling (i.e. CETs are indicating a temperature greater than 1200°F), then step 19 directs starting an available RCP in one idle cooling loop. This action would provide forced two phase flow through the core and should disturb or remove any established blockage to the core to improve core cooling. If starting one RCP does not result in a decrease in CET temperature to less than 1200°F then other available RCPs in idle cooling loops are incrementally started until a decreasing CET temperature response is obtained.

Serial: HNP-13-053 Page 8 of 8

3. Implementation of Severe Accident Management Guidance (SAMG):

EOP-FR-C.1, Response to Inadequate Core Cooling. directs that if CETs are greater than 1200°F and increasing and actions to cool the core, including running RCPs in all available cooling loops, are not successful in reducing the temperature, SAMG procedures are entered. Flooding of containment is a means described in the SAMGs (SAMG-SAG-008) to establish cooling of the core material when other strategies have been ineffective.

PWROG letter OG-13-137 transmits DW-12-013 which includes EOP changes to address the potential for lower core region flow blockage from in-vessel debris during the cold leg recirculation phase of safety injection and includes guidance for detecting lower core region flow blockage and potential mitigating actions. Guidance from DW-12-013 will be incorporated into the site EOPs.

Although these measures are not expected to be required based on the very low probability of an event that would result in significant quantities of debris being transported to the reactor vessel that would inhibit the necessary cooling of the fuel, they do provide additional assurance that the health and safety of the public would be maintained. These measures provide support for the extension of time required to completely address GL 2004-02 for HNP.

Conclusion HNP expects that this GSI-191 resolution path is acceptable, based on the information provided in this document. The execution of the actions identified in this document will result in successful resolution of GSI-191 and closure of GL 2004-02.

HNP-13-053 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400 Attachment 2 List of Regulatory Commitments

Serial: HNP-13-053 Page 1 of 1 Regulatory Commitments The following table identifies the actions committed to in this letter. Statements in this submittal with the exception of those in the table below are provided for information purposes and are not considered commitments.

No. Commitment Expected Completion Dates 1 Complete any necessary additional insulation The completion of the first refueling measurements based on the PWROGs new in- outage following the staffs issuance vessel acceptance criteria. of the safety evaluation for the PWROG Comprehensive Test Program. This is currently expected to be by the end of the Spring 2015 refueling outage for HNP (RFO-19).

2 Complete the necessary insulation The completion of the third refueling replacements, remediation, or model outage following the staffs issuance refinements to satisfy the in-vessel debris limit of the safety evaluation for the established by the PWROG Comprehensive PWROG Comprehensive Test Test Program. Program. This is currently expected to begin in the Fall 2016 refueling outage (RFO-20) and complete by the Spring 2018 refueling outage (RFO-21).

3 HNP will submit a supplemental response to Six months following the staffs support closure of GL 2004-02. issuance of the safety evaluation for the upcoming PWROG comprehensive test program.

4 HNP will update the Final Safety Analysis Following NRC acceptance of Report (FSAR) following NRC acceptance of supplemental response and in HNPs updated supplemental response and accordance with 10 CFR 50.71(e).

completion of the identified removal or modification of insulation debris sources in containment per plant modification procedures and processes (10 CFR 50.71(e)).