Text

Joint Appendix Volume III: Pages 1168 - 1298, Petition to Review a Final Decision of the Nrc. Orange County, Nc V. NRC and Carolina Power & Light Co
	ML021700171
Person / Time
Site:	Harris
Issue date:	06/04/2002
From:	; Harmon, Curran, Harmon, Curran, Spielberg & Eisenberg, LLP, Progress Energy Co, ShawPittman, LLP, US Dept of Justice (DOJ)
To:	; Office of Nuclear Reactor Regulation, US Federal Judiciary, Court of Appeals
References
	01-1073, 01-1246, 50-400-LA, 99-762-02-LA
	Download: ML021700171 (134)
	v • d • e

CASE SCHEDULED FOR ORAL ARGUMENT SEPTEMBER 5,2002 In the United States Court of Appeals For the District of Columbia Circuit Nos. 01-1073 and 01-1246 (Consolidated)

ORANGE COUNTY, NORTH CAROLINA, Petitioner V.

UNITED STATES NUCLEAR REGULATORY COMMISSION And the UNITED STATES OF AMERICA, Respondents CAROLINA POWER & LIGHT COMPANY, Intervenor-Respondent PETITION TO REVIEW A FINAL DECISION OF THE U.S. NUCLEAR REGULATORY COMMISSION JOINT APPENDIX VOLUME III: Pages 1168 throu2b 1785 Diane Curran Karen D. Cyr Anne Spielberg John F. Cordes, Jr.

Harmon, Curran, Spielberg & Eisenberg, L.L.P. E. Leo Slaggie 1726 M Street N.W., Suite 600 Charles E. Mullins Washington, D.C. 20036 U.S. N.R.C.

202/328-3500 Washington, D.C.

301/415-1606 John H. O'Neill David J. Cynamon Thomas L. Samsonetti Douglas J. Rosinski Assistant Attorney General SHAWPITTMAN, L.L.P.

2300 N Street N.W. Ronald Spritzer Washington, D.C. 20037 Attorney, Appellate Section Environment and Natural Resources Steve Carr, of Counsel U.S. Department of Justice Progress Energy Service Co. Washington, D.C. 20530 411 Fayetteville Street Mall 202/514-3977 Raleigh, NC 27602-1551 919/546-4161 Dated: June 4, 2002

- JOINT APPENDIX TABLE OF CONTENTS Vol. III Record Summary of Facts, Data and Arguments on Which Applicant Proposes to Rely at the Subpart K Oral Argument Regarding Contention EC-6 and Volumes 2 and 3 of Exhibits (November 20, 2000).... 1168 Note reearding omissions from the Joint Appendix: Due to their volume, counsel for CP&L has agreed to omit some materials from the Joint Appendix that were included in the record submittal of CP&L's evidentiary presentation of November 20, 2000. Please see Joint Appendix Vol. II for a listing. The Court is respectfully directed to the record to review the omitted materials.

NRC Staff Response to Intervenor's Request for Admission of Late-Filed Environmental Contentions (March 3, 2000) .............................................. 1557 Applicant's Response to BCOC's Late-Filed Environmental Contentions (M arch 3, 2000) ............................................................................................ 1587 Orange County's Comments in Opposition to No Significant Hazards Determination and Conditional Request for Stay of Effectiveness (February 12, 1999) ........................................................................ 1626 Orange County's Request for Admission of Late-Filed Environmental Contentions (January 31, 2000) ............................................................... 1681 Gordon Thompson, Risks and Alternative Options Associated With Spent Fuel Storage at the ShearonHarrisNuclear Power Plant (February 1999) ............................................................................................ 1708

November 20, 2000 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of )

)

CAROLINA POWER & LIGHT ) Docket No. 50-400-LA COMPANY (Shearon Harris Nuclear Power Plant) )) ASLBP No. 99-762-02-LA

SUMMARY

OF FACTS, DATA, AND ARGUMENTS ON WHICH APPLICANT PROPOSES TO RELY AT THE SUBPART K ORAL ARGUMENT REGARDING CONTENTION EC-6 VOLUME 2 EXHIBITS 2-5 001163

EXHIBIT 2 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of )

)

CAROLINA POWER & LIGHT ) Docket No. 50-400-LA COMPANY )

(Shearon Harris Nuclear Power Plant) ) ASLBP No. 99-762-02-LA AFFIDAVIT OF ROBERT K. KUNITA COUNTY OF WAKE )

) ss:

STATE OF NORTH CAROLINA )

I, Robert K. Kunita, being sworn, do on oath depose and say:

1. I am a resident of the State of North Carolina. I am employed by Carolina Power

& Light Company ("CP&L") and work at the Harris Nuclear Plant ("HNP" or "Harris Plant" or "Harris") in the Spent Fuel Management Subunit of the Environmental and Radiation Control Unit. Presently, I am a Principal Engineer, Spent Fuel Management. My business address is 5413 Shearon Harris Road, New Hill, North Carolina 27562-0165.

2. 1 hold a Bachelor of Science degree in Physics from the Illinois Institute of Technology and a Masters of Science degree in Nuclear Science and Engineering from Carnegie Mellon University. Since graduation, I have been employed by the Bettis Atomic Power Laboratory ("Bettis") and CP&L. At Bettis, from 1966 to 001169

1973, 1 was a member of the nuclear core design team for Admiral Rickover's Light Water Breeder Reactor Project, which subsequently ran successfully at the Shippingport Reactor. I performed computerized nuclear design calculations and participated in fuel design changes to optimize breeding while safely generating reactor power.

3. Since joining CP&L in 1973, I have held several positions of increasing engineering responsibility. From 1973 to 1975, I was in the Power Plant Engineering Section responsible for the nuclear fuel related systems for the planned South River Nuclear Power Plant. In 1975, 1 transferred to the Nuclear Fuel Section of the Fuel Department, where my responsibilities included interfacing with the Harris Nuclear Plant project on matters relating to nuclear fuel and I began participating in fuel examinations at the Robinson Nuclear Plant.

4. From 1977 to 1988, I was the Principal Engineer and head of the Surveillance and Accountability Unit. My responsibilities included assuring adequate nuclear fuel mechanical design, monitoring nuclear fuel mechanical performance, providing thermal-mechanical fuel analysis, planning for spent fuel storage and transportation, and providing fuel related support to CP&L's nuclear plants.

During this time my focus on zircaloy clad fuel in-reactor performance, fuel examination efforts, and fabrication of zircaloy clad fuel continued. Fuel performance improved significantly over this period. Around 1983, 1 initiated the Robinson Dry Storage Demonstration project and from then until 1989, 1 was 00117C

involved in the resolution of technical concerns regarding the performance of zircaloy clad fuel in dry storage. I was also a member of the Technical Management Oversight Committee for the dry storage project and participated in a number of technical meetings related to spent fuel issues.

5. In 1989, 1 transferred to the Emergency Preparedness and Spent Fuel Management unit of the Operations and Environmental Support Department where I was responsible for planning and coordinating the implementation of CP&L's spent fuel management program, which included both dry storage and shipment of spent fuel. In approximately 1992, my function and I transferred back to the Nuclear Fuel Section and, in 1998, to the Harris Nuclear Plant.

6. Since 1977, 1 have represented CP&L on numerous Nuclear Energy Institute and Electric Power Research Institute committees dealing with various aspects of nuclear fuel. In my current position as the Principal Engineer, Spent Fuel Management, I continue to be responsible for matters relating to spent nuclear fuel. I am also a Professional Engineer registered in North Carolina. My resume is provided in Attachment A to this affidavit.

7. The purposes of this affidavit are to provide facts, data and my opinion on which CP&L relies in evaluating the postulated occurrence of a zirconium self sustaining exothermic oxidation reaction in Harris spent fuel pools C and D following a postulated loss of most or all pool water through evaporation (i.e.,

"Step 7" in the seven step sequence of events identified on page 13 of the 001171

L Licensing Board's Order dated August 7, 2000 ("Order")) and to address the Board's second question concerning NUREG-1353. First, I describe the principles of a postulated self-sustaining exothermic oxidation reaction of zirconium spent fuel cladding. Second, I discuss the literature survey I conducted to research the likelihood of a self-sustaining exothermic oxidation reaction of zirconium spent fuel cladding occurring in the Harris spent fuel pools. Third, I describe the application of the information obtained in my literature survey to the spent fuel to be stored in Harris spent fuel pools C and D and the analyses I performed to establish that a self-sustaining exothermic oxidation reaction of zirconium spent fuel cladding is vwry unlikely in the Harris pools. Finally, I provide my conclusions on the unlikely occurrence of "Step 7" in the seven step sequence of events identified in the Board's Order.

PRINCIPLES OF THE EXOTHERMIC OXIDATION REACTION OF ZIRCONIUM SPENT FUEL CLADDING

8. Zircaloy, like most metals, undergoes an oxidation reaction in an air environment.

This oxidation reaction is exothermic, meaning that the reaction releases heat.

The oxidation rate, and, therefore, the rate at which heat is released, increases as the temperature of the zircaloy increases. At temperatures less than several hundred degrees Celsius, the exothermic oxidation reaction occurs very slowly.

9. The temperature of spent fuel zircaloy cladding is determined by the balance between the rate at which heat is generated in the fuel and the rate at which heat is transferred from the fuel. If the heat generation rate is greater than the rate at 001172

which heat is transferred from the fi-.l, the temperature rises. If the generation rate is the same as the heat transfer rate, the temperature is constant, and the temperature decreases if the heat generation rate is less than the heat transfer rate.

10. The primary contributor to the heat generation rate in spent fuel is radioactive decay of material in the fuel, referred to as decay heat. The heat input from the spent fuel, also known as spent fuel decay heat, is primarily a function of the '

combination of the reactor power level, the burnup of the spent fuel, in megawatt days per metric ton of fuel (MwD/Mtu), and the age (or "decay time") of the fuel.

The decay heat rate drops drastically with time after the fuel is discharged from the reactor. Approximately five years after discharge from the reactor, the decay heat rate of the old, cold spent fuel is a small fraction of the decay heat rate of the same fuel when it was first stored in the spent fuel pools.

11. It is possible that in some conditions of very high cladding temperatures, the oxidation rate and the corresponding heat generation from the exothermic reaction can become a significant heat source, which, when added to the decay heat from the fuel, can contribute to a further increase in temperature. If the increase in heat generation rate due to the exothermic oxidation reaction exceeds the heat transfer from the fuel, temperatures continue to increase, causing further increase in oxidation reaction rate. This condition is referred to as self-sustaining exothermic oxidation, and is the focus of step 7. The clad temperature at which the self sustaining oxidation reaction occurs is referred to as the critical cladding 001173

oxidation temperature. The result of a self-sustaining oxidation reaction is likely to be loss of clad integrity and loss of fuel pellet containment in the fuel assemblies.

12. Spent fuel cladding below the critical cladding oxidation temperature continues to oxidize, but a protective oxide layer forms on the cladding and slows down the oxidation rate. The oxidation reaction proceeds at such a slow rate that more than sufficient time is available to reestablish cooling of the spent fuel before cladding damage resulting in exposure of spent fuel could occur.

13. The mechanisms for heat transfer from spent fuel are conduction, convection, and radiative heat transfer. Conduction and convection dominate heat transfer until clad temperatures reach several hundred degrees Celsius. Some early analyses assumed spent fuel pools were completely dry when calculating convection heat transfer, but later studies included sufficient detail to model the temperatures with water above the bottom of the fuel racks, which obstruct the free flow of air into the bottom of the fuel assemblies.

RESULTS OF A SURVEY OF THE AVAILABLE LITERATURE

14. 1 performed a survey of the publicly available literature regarding the potential for the initiation and propagation of a self-sustaining exothermic oxidation reaction involving spent fuel cladding following the partial, or complete, loss of water from a spent fuel pool.

15. In conducting my literature survey, I searched a number of sources to identify 001174

documents addressing a self-sustaining exothermic oxidation reaction in spent nuclear fuel. My literiture survey identified as relevant a total of seventeen (17) documents, which are listed in Attachment B to this affidavit.

16. My literature survey did not identify any analysis that reported a zirconium cladding oxidation temperature any lower than 800°C. Numerous studies report that the critical cladding oxidation temperature of zirconium spent fuel cladding is about 900'C. While the Advisory Committee on Reactor Safeguards ("ACRS")

has indicated that the presence of zirconium hydrides on spent fuel may lower the critical cladding oxidation temperature, I did not identify any analysis that indicated zirconium hydrides would lower the critical cladding oxidation temperature below 800°C. NUREG/CR-5597 shows the onset of rapid zircaloy oxidation at 1500°K (1227 0 C). A table of reported critical cladding oxidation temperatures and the associated reports is presented in Attachment C to this affidavit.

17. Actual spent fuel has been heated up in air to a temperature of approximately 800°C under controlled laboratory conditions (see, Attachment B, reference 7).

No zirconium self-sustaining exothermic oxidation occurred even when the spent fuel was heated to approximately 800°C in an air environment. This experimental result is consistent with the analyses reporting 8000 C as a conservative lower bound for the critical cladding oxidation temperature of zircaloy fuel cladding.

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I APPLICATION OF AVAILABLE INFORMATION TO THE HARRIS SPENT FUEL POOLS C AND D

18. The residual or decay heat in nuclear fuel assemblies decreases rapidly after the fuel assembly is discharged from the reactor. Such decay heat is often described on an assembly basis in terms of the number of kilowatts of thermal energy given off (i.e., in units of kilowatts per assembly). It can also be expressed on a per metric ton basis. This is determined by dividing the assembly kilowatts by the metric tons of uranium in the assembly. This kilowatts per metric ton is also referred to as a "specific heat" and is often used to make comparisons between different fuel types which differ in physical size and amounts of uranium.

19. Zircaloy self-sustaining exothermic oxidation does not occur below a fuel clad temperature of 8000 C. This temperature will not be reached unless the specific heat is above some minimum value. For high burnup fuel stored in high density racks, the NRC has conservatively selected, in its "Technical Study on Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants" (Attachment B, reference 5), three (3) kilowatts per metric ton as this minimum value. This value was previously six (6) kilowatts per metric ton for lower burnup fuel stored in more widely spaced storage racks.

20. Harris spent fuel pools C and D will only store spent fuel aged five years or more out of the reactor. It is anticipated :hat all spent fuel shipped in the future from Brunswick and Robinson to Harris will be stored in spent fuel pools C and D.

21. Spent fuel from the Brunswick and Robinson plants that is to be stored in Harris 001176

spent fuel pools C and D is transported to Harris in CP&L's transportation cask.

The existing transportation cask has strict heat load limits on the spent fuel that can be transported in the cask. The transportation cask Certificate of Compliance 9001 limits the maximum heat load of spent fuel shipped in the cask to a peak of 5,725 Btu/hr/PWR and 2,225 Btu/Hr/BWR which is equivalent to 3.7 Kw/Mtu for PWR fuel and 3.6 Kw/Mtu for BWR fuel. As a practical matter, these limits have not been approached and the peak assembly heat load of spent fuel transported to Harris spent fuel pools A and B was 2.5 Kw/Mtu for PWR fuel and 2.9 Kw/Mtu for BWR fuel. The peak assembly specific heats, at the time of shipment, for the shipments made to Harris pools A and B are shown in Figures 1 and 2 for PWR and BWR fuel respectively (Attachments D and E).

22. Figure 3 shows the decreasing specific heat curves for the PWR fuel stored in Harris spent fuel pools A and B (Attachment F). Some of this spent fuel will be moved to pools C or D in the future to make room for spent fuel discharges from the Harris reactor. Present projections show that the Robinson spent fuel will be 15 to 20 years old at the time it is moved to pools C or D. Due to its age, this spent fuel does not have a high enough specific heat to cause a self-sustaining exothermic oxidation reaction in pools C or D.

23. Figure 4 shows the decreasing specific heat curves for the BWR spent fuel stored in Harris spent fuel pools A and B (Attachment G). There are no present plans to move this spent fuel to pools C or D. This fuel has already cooled between 6 and 001177

23 years. Due to its age, this spent fuel will not have a high enough specific heat to cause a self-sustaining exothermic oxidation reaction in pool C or D.

24. Figure 4 also applies to BWR fuel assemblies to be shipped from the Brunswick plant and then unloaded at Harris and stored in pools C or D. Since such fuel will have at least 5 years of cooling prior to shipment, this fuel will also not have a high enough specific heat to cause a self-sustaining exothermic oxidation reaction in pool C or D.

25. Figure 5 shows the decreasing specific heat curves for the high burnup Harris PWR fuel (Attachment H). Some Harris fuel with burnups over 40,000 MwD/Mtu was added to Harris spent fuel pools A and B beginning in 1991. Such spent fuel is not presently projected to be moved to pools C or D until 2011 or later; hence, such spent fuel will have cooled about 20 years. Due to its age, this spent fuel will not have a high enough specific heat to cause a self-sustaining exothermic oxidation reaction in pools C or D.

26. Anecdotal evidence exists that shows that zircaloy self-sustaining exothermic oxidation does not occur for cooled spent nuclear fuel. Between late 1977 and early 198 1, CP&L shipped 290 PWR fuel assemblies from Robinson to Brunswick in over 40 shipments using air coolant in the shipping cask. At the time of shipment, this fuel had cooled between 2.7 and 6.5 years. There is no evidence that there was anything unusual about these assemblies when they were unloaded after receipt at Brunswick.

001178

27. As shown in Figure 6. the amount of Ruthenium- 106 decreases rapidly after the spent fuel assembly is removed from the reactor (Attachment I). This is because Ruthenium- 106 has a half life of approximately one year; hence, half of the Ruthenium decays away every year. Only fuel aged for five years or more will be placed in pools C and D. By the time that pool C is filled, the average age of the fuel will be 23 years for PWR fuel and 10 years for BWR fuel. By the end of the Harris operating license (i.e., 2026), the average age of the fuel in pool D will be 23 years for PWR fuel. No BWR fuel is planned to be stored in pool D. In 2026, the average age of the fuel in pool C will be 30 years for the PWR fuel and 22 years for BWR fuel. Thus, any evaluation of the amount of Ruthenium in pools C and D is not particularly meaningful.

28. Even if a small number of spent fuel assemblies in Harris spent fuel pools C and D could potentially sustain a zirconium self-sustaining exothermic oxidation reaction, it would be unlikely to propagate to adjacent assemblies because they have a heat load far too low to sustain a the reaction. Thus, even if a zirconium self-sustaining exothermic oxidation reaction could occur in Harris spent fuel pools C and D, its extent would likely be extremely limited because of the large quantity of extremely old, cold spent fuel stored in the pools and the small, if any, amount of undecayed Ruthenium remaining. Thus, it would appear that the consequences of a postulated zirconium exothermic oxidation reaction in spent fuel pools C and D would be bound by the consequences on the severe reactor 00-117-9

I accident that initiated the scenario set forth in the Board's Order.

IMPACT OF RECENT DEVELOPMENTS ON NUREG-1353 ESTIMATES

29. I have reviewed the documents listed in Attachment J to this Affidavitto evaluate their impact on two estimates contained in NUREG-1353, "Regulatory Analysis for the Resolution of Generic Issue 82, 'Beyond Design Basis Accidents in Spent Fuel Pools"' (1989).

30. To the extent that any NUREG-1353 estimated value is applicable to the specific seven step scenario contained in the Board's Order, it is my opinion that the data or models that have been reported since the publication ofNUREG-1353 do not suggest any substantive modification of those values. It is also my opinion, however, that, with the exception of the values associated with the probability of a self-sustaining exothermic oxidation reaction occurring in spent fuel following loss of spent fuel pool water level, the estimated values in NUR.EG-1353 do not appear applicable to the scenario postulated in the Order.

31. Regarding the probability of a self-sustaining exothermic oxidation reaction being greater than assumed in NUREG-1353, it is my opinion that this is clearly not the case for PWR spent fuel, as NUREG-1353 assumes a probability of 1.0 for this event. As to BWR spent fuel, I did not find in my literature search any basis for changing the probability value of 0.25.

CONCLUSIONS

32. As long as the heat output of spent fuel is less than the available heat removal 001180

capability, the spent fuel will remain cool, and no self-sustaining zirconium exothermic oxidation reaction will occur.

33. For spent fuel with heat outputs less than the limits identified in the literature, no self-sustaining zirconium exothermnc oxidation reaction will occur even if spent fuel pool water inventory is lost because the available energy is insufficient to initiate and sustain the reaction.

34. For spent fuel with a heat output above the identified limits, it is unclear whether a self-sustaining exothermic oxidation reaction will occur.

35. I conclude, therefore, that because of the low heat load in the old, cold spent fuel to be stored in Harris spent fuel pools C and D, it is highly unlikely that the spent fuel in pools C and D could sustain a zirconium self-sustaining exothermic oxidation reaction, even if most or all of the water in pools C and D is lost through evaporation.

36. I also conclude that in the highly unlikely event that a zirconium self-sustaining exothermic oxidation reaction were to occur in Harris spent fuel pools C and D, its extent would be extremely limited because most of the spent fuel to be stored in pools C and D will be far too old and cold to propagate the reaction. The consequences would certainly be bound by the consequences of the postulated degraded core accident with containment bypass that is the postulated initiator of the seven step sequence in the Board's Order.

001181

I I declare under penalty of perjury that the foregoing is true and correct.

Executed on November F*'2000.

Robert K. Kunita Subscribed and sworn to before me this )* day of November, 2000.

My Commission expires.:" -I )- a 05L 00118?_

V Hams Nuclear 5413 Shear Plant Hams Road New Hill. NC 27562-9217 Work (919) 362-3132 Robert K Kunita Professional 1973 - Present Carolina Power & Light Company Experience Principal Engineer - Spent Fuel Management During my 27 years with Carolina Power & Light, I have worked in the Power Plant Engineering Section, the Nuclear Fuel Section, and the Emergency Preparedness & Spent Fuel Management Sections, all of which were in the Corporate Offices in Raleigh, NC. I have worked for the past two years at the Harris Nuclear Plant located in New Hill, NC in the Spent Fuel Management Subunit of the Environmental and Radiation Control Unit.

My experience covers a broad range of nuclear fuel related items from reactor systems interfaces, fuel design, fuel fabrication, nuclear material accountability, and spent fuel management. I was responsible for and accomplished reviews of system designs and NRC license application submittals, development and implementation of nuclear fuel fabrication surveillance plans, establishment and maintenance of a nuclear material accountability program, development of a dry spent fuel storage demonstration project which was successfully implemented, preparation of implementation of spent fuel shipping emergency exercises, and development of a corporate spent fuel management plan.

I have reviewed documents from the NRC, NEI, EPRI, etc. for technical adequacy and impact on CP&L and I have represented CP&L on numerous NEI and EPRI spent fuel committees.

1966 - 1973 Bettis Atomic Power Laboratory West Mifflin, PA Associate Engineer through Senior Engineer I worked for 7 years at the Bettis Atomic Power Laboratories, which was run by Westinghouse for the Naval Reactors Program. I was a member of the nuclear core design team for Admiral Rickover's Light Water Breeder Reactor Project, which subsequently ran successfully at the Shippingport Reactor. I performed computerized nuclear design calculations and participated in fuel design changes to optimize breeding while safely generating reactor power.

Education Carnegie Mellon University Pittsburgh, PA a 1973 M. S. Nuclear Science and Engineering Illinois Institute of Technology Chicago, IL

1966 B.S. Physics 001183

I Regist-ation Registered Professional Engineer o North Carolina, PE #007015 Awards 1993 CP&L Quality Achievement Award Professional American Nuclear Society Memberships Eastern Carolinas Section of the American Nuclear Society, past membership chairman and treasurer.

00.118 1

Attachment B References I. N. A. Pisano. F. Best. A. S. Benjamin and K. T. Stalker, "'The Potential for Propagation of a Self-Sustaining Zirconium Oxidation Following Loss of Water in a Spent Fuel Storage Pool", Draft Report, January, 1984.

2. A. S. Benjamin, D. J. McCloskey, D. A. Powers. and S. A. Dupree, "Spent Fuel Heatup Following Loss of Water During Storage", NUREG/CR-0649. March, 1979.

3. V. L. Sailor. K. R. Perkins, J. R. Weeks, and H. R. Connell, "Severe Accidents in Spent Fuel Pools in Support of Generic Safety Issue 82-, NUREG/CR-4982, July, 1987.

4. R. J. Travis. R. E. Davis, E. J. Grove, M. A. Azarm, "A Safety and Regulatory Assessment of Generic BWR and PWR Permanently Shutdown Nuclear Power Plants". NUREG/CR-645 1, August, 1997.

5. D. Jackson, T. Eaton, G. Hubbard, "Technical Study of Spent Fuel Pool Accidents for Decommissioning Plants", Draft, February, 2000.

6. A. B. Johnson, Jr.. J. C. Dobbins, F. R. Zaloudek, E. R. Gilbert, and I. S. Levy, "Assessment of the Integrity of Spent Fuel Assemblies Used in Dry Storage Demonstrations at the Nevada Test Site", PNL-6207, July, 1987.

7. D. Stahl, M. P. Landow, R. J. Burian, and V. Pasupathi, "Spent Fuel Behavior Under Abnormal Thermal Transients During Dry Storage", PNL-5456, January, 1986.

8. C. Tinkler, Presentation to the ACRS Reactor Fuels Subcommittee, October 18, 2000.

001185

9. E. D. Throm, "Regulatory Analysis for the Resolution of Generic Issue 82. "Beyond Design Basis Accidents in Spent Fuel Pools-', NUREG-1353.

April. 1989.

10. W. D. Travers memo to The Commissioners, "Improving Decommissioning Regulations for Nuclear Power Plants". SECY-99-168, June 30, 1999.

11. C. F. Boyd, "Predictions of Spent Fuel Heatup After a Complete Loss of Spent Fuel Pool Coolant", NUREG-I1726, July, 2000.

12. L. Matus, L. Maroti, P. Windbert, M. Horvath, I. Nagy, A. Grigger, M. Balasko, "Post-Test Examination of the CODEX-AIT-2 Bundle",

OPSA Final Report, Budapest, June, 1999

13. H. Ocken, "An Improved Evaluation Model for Zircaloy Oxidation", Nuclear Technology, Volume 47, February, 1980.

14. A. W. Cronenberg, "In-Vessel Zircaloy Oxidation/Hydrogen Generation Behavior During Severe Accidents," NUREG/CR-5597, September, 1990.

15. E.T. Bums, D. E. True, V. M. Andersen, "A Review of Draft NRC Staff Report:

DRAFT Technical Study of Spent Fuel Pool Accidents for Decommissioning Plants",

August, 1999

16. D.A. Powers, "Draft Final Technical Study of Spent Fuel Pool Accidents Risk at Decommissioning Nuclear Power Plants", letter to R. A. Meserve (NRC). April 13, 2000.

17. D. A. Powers, "Proposed Resolution of Generic Safety Issue-I 73A, 'Spent Fuel Storage Pool For Operating Facilities"', letter to R.

A. Meserve (NRC), June 20, 2000.

0011861

Attachment C Table of Reported Zircaloy Temperatures Comment temperature escalation-

"The temperature at which clad oxidation becomes self-sustaining is a function of the storage configuration, but tends to occur around 9000 C."

"If the decay heat level is sufficient to heat the rods to about 9000 C (16500 F) the oxidation becomes self-sustaining. That is, the exothermic oxidation reaction provides sufficient energy to match the decay heat contribution and the temperature rises rapidly."

565 0 C 4, page 3-4 'The Workshop on Transportation Accident Incipient clad Scenarios estimated incipient clad failure at failure is not the 5650C with expected failure at 6710 C. onset of self presumably based on expert opinion. Given sustaining zirc that the large seismic event is the dominant reaction; it is the contributor to the configuration 1 initiator, it onset of clad is likely that it would take a prolonged swelling.

period of time to retrieve the fuel, repair the spent fuel pool or establish an alternate means of long-term spent fuel storage.

Therefore, we presume there will be a significant period of time that the fuel will be exposed to air. On this basis, BNL has chosen a temperature of 5650 C as the critical cladding temperature."

800 0 C 5, page Al-1 "'The onset of rapid oxidation may occur as References 3.

low as 8000 C.

0

'-wout YVV k- "9,page tS-I If the decay heat level is high enough to heat the fuel rod cladding to about 9000 C (16500 F) the oxidation becomes self-sustaining, resulting in Zircaloy cladding fire.

I Propagation of the Zircaloy cladding fire to 001181

L older adjacent assemblies is likely if the I

decay heat level in an older adjacent assembly is high enough to heat that assembly to within 100 to 2000 C (200 to 4000 F) of the self-sustaining oxidation temperature. Although propagation of a Zircaloy cladding fire to one or two year old fuel by only thermal radiation can occur, the older fuel would have to be next to the hottest assembly."

8000 C 10, page 4 "The working group is reviewing the temperature criteria used in the spent fuel analysis and the preliminary results indicate that a maximum allowable temperature of 8000 C may be acceptable if certain analysis conditions are met. The conditions for applying this criteria would include demonstrating that the maximum calculated temperature, including uncertainties, remained below the temperature limit, that higher temperature effects are accounted for, and that a release of the radionuclides in the gap between the clad and the fuel is not a concern. The 8000 C temperature limit is based on the lowest temperature for the onset of self-sustaining zirconium oxidation identified by the GSI 82 studies."...

16000 F 13, page 354 "It was concluded that a new evaluation (8710C) This report model for Zircaloy oxidation should be references instanc applicable above 11440K (16000F), which is where zircaloy w the temperature at which the Zircaloy oxidized at oxidation rate becomes appreciable." 927 0 C, 1316 0 C, "Measurements of the oxidation kinetics 10380 C, 1000 to from prefilmed cladding and from specimens 1690'C, 1300 to subjected to anisothermal conditions permit 17500 C. There is one to obtain some measure of the margin mention of associated with different representations of exothermic the Zircaloy oxidation kinetics. The oxidation.

evidence shows that the presence of a prefilm significantly inhibits subsequent oxidation." [Prefilm is preoxidation or anomalous oxidation, see page 350].

001188

Figure 1 - Assembly Maximum Specific Heat (Shipments from Robinson) 4.5 4.0 3.5 3.0 2.5 i PWVR Fuel

2.0 - -" Cask Limit 1.5 1.0 0.5 0.0, 1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 41 43 45 47 49 51 53 Sequential Cask Shipped 0o w

Figure 2 - Assembly Maximum Specific Heat (Shipments from Brunswick) 02.0 Iin8BWR Fuel

" - Cask Limit 1.5 1.0 0.5 0.0 1 9 17 25 33 41 49 57 65 73 81 89 97 105 113 121 129 Sequential Cask Shipped

Figure 3 - Assembly Specific Heat (PWR Fuel in Harris Pools A and B) 24 . . . . . .. . . . . . . . . . . . . . . ....I . .I 3.8 w/o, 45 20 -- GwD/Mtu 18 -- IIANF 15x15, 3.8 w/o, 45 16 GwD/Mtu 14 MOW 17x17, 12 5.0 w/o, 45 CL GwD/Mtu

0. 10

(, *ANF 15x15, ro 8 .I 5.0 w/o, 45 GwD/Mtu 6 I-.

4-4 -- Critical 4

Decay Heat

- - a -

2

.- - a - a a - Earlier Bound 0 "- - - Critical 0.5 2.0 3.5 5.0 6.5 8.0 9.5 11.0 12.5 Decay Heat 0

14.0 15.5 17.0 18.5 20.0 - Latest Cool Down (Years) Bound Bound i-.

C)

C).'

I-

C)

C.Ij

C)

Attachment J

1. E. D. Throm, "Regulatory Analysis for the Resolution of Generic Issue 82.

'Beyond Design Basis Accidents in Spent Fuel Pools"', NUREG-1353. April, 1989.

2. Susquehanna Safety Evaluation Report.

3. D.A. Lochbaum, 'Comments on the Draft Susquehanna Safety Evaluation Regarding Spent Fuel Pool Cooling Issues, letter to Selin (NRC), November 29, 1994.

4. Taylor, "Resolution of Spent Fuel Pool Action Plan Issues", Memorandum to Commissioners, July, 26, 1996.

5. Callan, "Follow-up Activities on Spent Fuel Pool Action Plan", Memorandum to Commissioners, September 30, 1997.

6. J. Ibarra, W. Jones, G. Lanik, H. Ornstein, S. Pullani, "Assessment of Spent Fuel Pool Cooling", AEOD/S96-02, September, 1996.

7. D. Jackson, T. Eaton, G. Hubbard, "Technical Study of Spent Fuel Pool Accidents for Decommissioning Plants", Draft, February, 2000.

8. W. D. Travers memo to The Commissioners, "Improving Decommissioning Regulations for Nuclear Power Plants", SECY-99-168, June 30, 1999.

9. SRM responding to SECY-99-168, December 21, 1999.

10. D. A. Powers, "re Spent Fuel Fires Associated with Decommissioning", letter to Travers (EDO), November 12, 1999.

001195

IL

11. W. D. Travers. "re Spent Fuel Fires Associated with Decommissioning" December 16, 1999.

12. R. J. Travis. R. E. Davis, E. J. Grove, M. A. Azarm.

"A Safety and Regulatory Assessment of Generic BWR and PWR Permanently Shutdown Nuclear Power Plants", NUREG/CR-645 1. August, 1997.

13. N. Siu, S. Khericha, S. Conroy, S. Beck, H. Blackman, "Loss of Spent Fuel Pool Cooling PRA: Model and Results", INEL-96/0334, September, 1996.

14. J. Ibarra, W. Jones, G. Lanik, "Operating Experience Feedback Report:

Assessment of Spent Fuel Cooling", NUREG-1275 Vol. 12, 1997.

15. J.H. Jo, P.F. Rose, S.D. Unwin, V.L. Sailor, K.R.

Perkins, A.G. Tingle, "Value/Impact Analysis of Accident Preventative and Mitigative Options for Spent Fuel Pools", NUREG/CR-5281, 1989.

16. P. Prassinos, C. Kimura, D. McCallen, R. Murray, M. Ravindra, R. Campbell, P.

Hashimoto, A. Nafday, W. Tong "Seismic Failure and Cask Drop Analysis of Spent Fuel Pools at Two Representative Nuclear Power Plants", NUREG/CR 5176, January 1989.

17. V. L. Sailor, K. R. Perkins, J. R. Weeks, and H. R. Connell, "Severe Accidents in Spent Fuel Pools in Support of Generic Safety Issue 82", NUREG/CR-4982, July, 1987.

18. E.T. Bums, D. E. True, V. M. Andersen, "A Review of Draft NRC Staff Report:

DRAFT Technical Study of Spent Fuel Pool Accidents for Decommissioning Plants", August, 1999.

001191

19. D.A. Powers, "Draft Final Technical Study of Spent Fuel Pool Accidents Risk at Decommissioning Nuclear Power Plants", letter to R. A. Meserve (NRC),

April 13, 2000.

001197

EXHIBIT 3 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of )

)

CAROLINA POWER & LIGHT ) Docket No. 50-400-LA COMPANY )

(Shearon Harris Nuclear Power Plant) ) ASLBP No. 99-762-02-LA AFFIDAVIT OF STEVEN A. LAUR COUNTY OF WAKE )

) ss:

STATE OF NORTH CAROLINA )

I, Steven A. Laur, being sworn, do on oath depose and say:

1. I am a resident of the State of North Carolina. I am employed by Carolina Power

& Light Company ("CP&L") and presently serve as Superintendent of the Probabilistic Safety Assessment Unit. This unit is responsible to update and maintain plant-specific risk assessment models in a quality manner; maintain consistency in selection and use of risk-based tools across the Nuclear Generation Group; and effectively use risk-based tools to achieve goals in the safety, production, cost and plant license renewal areas. I have as direct reports engineers located at the General Office and at the Brunswick, Harris and Robinson Nuclear Plants. My business address is 410 South Wilmington Street, Raleigh, NC, 27601.

001198

2. I was graduated from University of Central Florida in 1972 with a Bachelor of Science in Mathematics and from Campbell University in 1985 with a Masters, Business Administration. Since receiving my Bachelor's degree, I have been employed by the United State's Navy, Ford Motor Company, RCA Service Company, Florida Power and Light Company, TENERA L.P., and CP&L. During my tenure at CP&L, specific positions held include Project Engineer in the Nuclear Licensing Section, Corporate Nuclear Safety Section, Risk Assessment Unit, and Probabilistic Safety Assessment Unit. I held a Senior Reactor Operator's license at St. Lucie Nuclear Plant, and have diverse technical and management experience in the areas of project engineering, plant operations, safety analysis, risk management, incident investigation and root cause determination. I am a registered Professional Engineer in the State of Florida.

My resume is provided as Attachment A to this affidavit.

3. The purpose of this affidavit is to describe the information that was developed by CP&L for use by ERIN Engineering and Research, Inc. ("ERIN") for their performance of an analysis of the sequence of events set forth in the Atomic Safety and Licensing Board's Memorandum and Order dated August 7, 2000

("Order"). F I describe the review history of the documents that were used to develop the information. Second, I discuss the specific steps that were taken to ensure that the ERIN analysis was consistent with the plant-specific attributes that were important to the analysis. Finally. I present my conclusions on the quality of the ERIN analyses and the appropriate use of Harris-specific information.

00t199

I

4. My role in the preparing a response to the questions contained in the Board's Order was to provide the updated Harris Individual Plant Examination ("IPE")

PSA model and the Harris Individual Plant Examination of External Events

("IPEEE") analysis to ERIN, which they used as the starting point for their analysis of the sequence of seven events set forth on page 13 of the Board's Order.

Consistent with CP&L standards for configuration control, I ensured that Han-is plant-specific PSA information was provided to ERIN as requested to support the analysis. I was also responsible for ensuring that the resulting ERIN analysis utilized correct and appropriate Harris plant information.

HISTORY OF THE HARRIS PROBABILISTIC SAFETY ASSESSMENT

5. The updated Harris IPE PSA model is a probabilistic safety assessment model that was originally developed for the Harris IPE pursuant to Generic Letter 88-20. The updated Harris IPE PSA model includes: 1) Event trees that model core damage accident sequences and containment response following a core damage event; 2)

Fault trees that represent plant systems and failure modes; 3) Initiating event, component failure, and human reliability data; and 4) Special analyses, such as internal flooding and Interfacing System Loss of Coolant Accident ("ISLOCA").

The updated Harris IPE PSA model considers internal initiating events (except internal fires) and applies when the reactor is critical (Modes I and 2). The results of the updated Harris IPE PSA model include an estimated annualized core damage frequency. The Harris IPEEE analysis was performed pursuant to 001200

Generic Letter 88-20 Supplement 4. The IPEEE considered 1) seismic risk, 2) internal fire risk, and 3) risk from other external events . high winds, tornadoes, and nearby facility accidents).

6. The updated Harris IPE PSA and the Harris IPEEE have been reviewed by organizations outside CP&L to ensure they possess a level of quality commensurate with their intended use. An Independent Peer Review of the model was commissioned. This review utilized industry standard guidance for review of PSA models (NEI 00-02, "Probabilistic Risk Assessment (PRA) Peer Review Process Guide") to determine the acceptability of the PSA model for use in analyzing the sequence of events proposed by the Board. ERIN, which was not involved in the development or update of the Harris updated IPE PSA model, performed the review, which is provided as Attachment B. This review provided high marks for the updated Harris IPE PSA model and concluded that, with one exception, the model formed an acceptable starting point for the sequence of events to be analyzed. This exception was the ISLOCA, which is important to the initiating event analysis of the Board's sequence of events, which ERIN determined to be overly conservative. Because the Board required a best-estimate analysis of the sequence of events, I directed ERIN to update the ISLOCA analysis to more realistically model that initiating event. Attachment C contains a table summarizing these reviews.

7. The CP&L management standard is that the Harris updated IPE PSA model be updated and maintained in a quality manner that is equivalent to applicable 00.1201

portions of 10 C.F.R. Part 50, Appendix B. A CP&L Nuclear Generating Group Common ("NGGC") procedure controls the update and maintenance of the Harris updated IPE PSA model. This procedure requires that: 1) Software used to perform PSA applications and analyses must be qualified and controlled in accordance with NGGC procedures; 2) The PSA models and applications must be described and controlled in accordance with NGGC procedures; 3) PSA model changes, applications and analyses must be performed in accordance with NGGC procedures; and, 4) Errors identified in PSA models, PSA software, PSA methods, or PSA applications must be documented in the CP&L Corrective Action Program.

8. ERIN used the Harris updated IPE PSA and the Harris IPEEE analysis that I supplied as key inputs to their analysis of the Board's sequence of events. Use of these Harris models and analyses ensured that the ERIN analysis was based on the latest available, plant-specific risk information.

HARRIS SPECIFIC INFORMATION

9. To ensure that the ERIN analysis of the Board's sequence of events was consistent with the plant-specific attributes important to the analysis, I coordinated a review of the report by a multi-disciplinary team of individuals within CP&L who have knowledge and expertise in one or more of the areas covered by the report. This review included ensuring that the inputs provided by CP&L were correctly incorporated into the analysis. The review also validated that the methodology 001 202

used by ERIN was appropriate for use at Harris. The review confirmed that assumptions and statements made in the report concerning plant-specific features and characteristics were appropriate. I reviewed portions of the analysis and reviewed the comments provided by 'he other members of the CP&L review team.

10. When the Board's sequence of events was promulgated, I recommended an outside contractor be retained to perform the analysis and selected ERIN. I directed ERIN to provide a best-estimate risk assessment of the sequence of events described in the Board's Order and to report the result in terms of an estimate of the annual frequency of the entire scenario. This analysis was to include not only internal events as modeled in the Harris updated IPE PSA model, but also sensitivity analyses of the scenario frequency to other initiating events, including dominant internal fires and seismic events. The analysis was also to consider the sensitivity of the results to core damage events during shutdown conditions.

11. In order to support the analysis by ERIN, CP&L provided the plant-specific information specified in the table provided as Attachment D to this affidavit.

Cognizant CP&L personnel reviewed each submittal to ERIN to ensure that accurate information was provided.

12. ERIN personnel toured the Harris facility on two occasions to gather plant specific data for the analysis of the Board's sequence of events. ERIN personnel discussed pertinent plant information with cognizant Harris personnel, including 001203

I sources of make-up water to the spent fuel pools and accessibility of the reactor auxiliary building and fuel handling building during accident conditions.

CONCLUSION

13. In conclusion, the ERIN analysis of the sequence of events postulated in the Board's Order used appropriate Harris plant-specific information.

The ERIN analysis utilized the information contained in, and builds upon, the existing Harris updated IPE PSA model and Harris IPEEE analysis, which are plant-specific studies. The Harris updated IPE PSA model was independently peer reviewed and found to be of high quality and appropriate for the analysis of the Board's sequence of events. The CP&L review of the ERIN analysis of the Board's sequence of events confirmed that the assumptions and data used by ERIN reflect the relevant attributes of the Harris plant.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on November L., 2000.

"Steven A. Laur Subscribed and sworn to before me

. C this 1-5 day of November 2000.

S .. o* C

V'.'L

My Commission expires: - -

exie:9-1 Z UN

" 001204

STEVEN A. LAUR, P.E.

Project Engineer Professional Qualifications Mr. Laur has over 20 years of engineering and managerial experience, which includes experience in the operation. training, quality extensive assurance, maintenance, and supervision of power reactors. He held a Senior Reactor nuclear Operator license at St. Lucie Nuclear Plant, diverse technical and management experience and has in the areas of project engineering, plant operations. safety analysis. risk management, incident investigation and root cause determination.

He has provided management consulting to assist nuclear plants in determining the root and developing effective solutions to their causes of performance problems. Additionally. Mr.

conducted performance-based assessments Laur of nuclear utilities, with focus on operational readiness, independent oversight, quality verification, and regulatory compliance issues.

Mr. Laur was Project Engineer assigned to the Nuclear Licensing Section during the process for Shearon Harris Nuclear Power licensing Plant. He supervised four Senior Engineers Specialists and seven Technical Aides, Licensing and Assistants, and Production Assistants. He the Harris Licensing Engineer during the was ASLB hearings, receipt of the Final SER, ACRS sub-committee and full committee and the hearings for the plant. He served on the Robinson tlip reduction team and conducted H. B.

special investigations of plant events at all CP&L nuclear units to determine root causes four and assess the adequacy of corrective actions.

conducted internal assessments of CP&L's He nuclear sites patterned after the NRC SALP Mr. Laur performed independent reviews as program.

a member of the standing nuclear safety organization and provided insight to the organization, charter development, and startup phase of the CP&L Nuclear Assessment Department.

He is currently employed by Carolina Power

& Light Company as the supervisor in charge Probabilistic Safety Assessment Unit, which of the is responsible for Risk analyses for the Brunswick.

Harris. and Robinson nuclear plants.

Education/Special Training MBA Campbell University, 1985 B.S. Mathematics, University of Central Florida, 1972 U.S. Navy Nuclear Power Training. Bainbridge, MD and West Milton, NY, 1975-1976 EG&G Accident Investigation Course, including Management Oversight and Risk Tree (MORT),

1987 (1 week)

TENERA Integrated Risk Management System training, October, 1994 OO4,205

L Steven A. Laur Experience 1998 - Present Superintendent - PSA Unit, Carolina Power & Light Company. First-line supervisory position in charge of Unit responsible to update and maintain plant-specific risk models in a quality manner (equivalent to applicable portions of the IOCFR50 Appendix B program): maintain consistency in selection and use of risk-based tools across NGG; and effectively use risk based tools to achieve goals in the safety. production. cost and plant life extension areas. Position has as direct reports engineers located at the General Office and at Brunswick. Harris and Robinson sites.

1995- 1998 Project Engineer, Carolina Power & Light Company. Currently provides PRA support to the Brunswick. Shearon Harris. and H. B. Robinson nuclear plants. This support includes model development and maintenance as well as applications of the PRA to facilitate managing the risks associated with nuclear power operations. Familiar with current PRA methods and practices, including fault tree and event tree analysis.

data gathering and analysis, human reliability analysis. and quantification of PRA models using the CAFTA computer code. Performed specialized PRA application studies to determine critical components for maintenance, to determine the risk associated with online maintenance activities, and to provide justification for technical specification changes.

Performed special fault tree model of the Harris emergency diesel generator control system to identify critical components for predictive and preventive maintenance to ensure continued high reliability. Greatly enhanced the quantification methodology for the H. B. Robinson PRA.

with the result that the level 1 (core damage) model quantification time was reduced from over 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to around 15 minutes. Provided analyses to evaluate the risk associated with several events at the Brunswick plant.

Provided statistical analysis of increased control rod scram times for Brunswick unit 1.

1992- 1995 Senior Consultant, TENERA. L.P. Provided management consulting to assist nuclear plants in determining the root causes of and developing effective solutions to their performance problems.

From December, 1991 through June. 1993. performed the human reliability analysis (HRA) for the Dresden and Quad Cities Stations as part of the Commonwealth Edison Company's Individual Plant Examination (IPE). Utilized the NUREG/CR-1278 (THERP) methodology. Observed licensed operators perform simulator exercises for Dresden and Quad Cities in support of the HRA. Reviewed plant response (event) trees for Dresden and Quad Cities; developed a fault tree model of the contaminated condensate storage tank inventory control function for Quad Cities. Performed an internal review of the HRAs for the Monticello and Clinton PRA models.

Page 2...

00.1206

Steven A. Laur He assisted the Palisades plant in preparing for an NRC Diagnostic Evaluation Team (DET) inspection. He facilitated the identification of performance improvement initiatives and facilitated validation workshops to ensure all major problems facing the station were identified.

In a similar effort, he assisted the Quad Cities plant in improving their ability to operate the plant efficiently and effectively, including work on the station Management Plan. prioritization. ranking. and management of issues. setting management expectations for site safety culture, and providing guidance regarding how to respond to their NRC Diagnostic Evaluation Team inspection.

In an effort to assist Commonwealth in addressing the placement of the Zion and Dresden plants on the NRC watchlist. assigned the role of technical team leader for the root cause portion of a comprehensive evaluation program (May, 1992). As a follow-on effort, was again the technical team leader for the root cause evaluation of the performance of the Quad Cities station in preparation for and NRC Diagnostic Evaluation Team (DET) inspection (July, 1993).

In 1992. participated in a performance-based assessment of the PSE&G Quality Assurance/Nuclear Safety Review organization.

During this project. interfaced with PSE&G site supervisory and management personnel. Conducted technical information-gathering interviews with site personnel at a variety of levels and reviewed a variety of documentation in support of this project including the products produced by the QA/NSR organization, LERs, incident reports, QA reports, INPO evaluation reports, and NRC correspondence. Primary author of the resultant report to the client.

As a direct result of the 1992 assessment, awarded sole-source project to perform follow-on work for the PSE&G Nuclear Safety Review group.

In February, 1993, taught five one-day classes on the subject of NRC reporting requirements at both the PG&E corporate offices in San Francisco and at the Diablo Canyon nuclear plant. In a separate effort for PG&E. developed a technical basis document to support sampling plans for samples from a finite population without replacement (based on the hypergeometric distribution). Derived algorithm and developed corresponding computer code to aid in the calculation of the cumulative hypergeometric distribution.

1983 - 1991 Carolina Power and Light Company:

Page 3 00.1207

Steven A. Laur I 1991 Project Engineer. Risk Assessment Unit.

Provided analysis and evaluation of nuclear power plant design features utilizing PRA techniques. including work necessary to respond to the NRC IPE requirements and severe accident policy.

Provided human reliability analyses to support plant-specific probabilistic risk fault tree models and core damage sequences assessments. Utilized to provide comparative analyses of options for management decision making regarding modifications, justifications for continued operation, and requests for NRC waiver of compliance issues.

Developed fault trees to support a Brunswick plant application:

Comparison of various proposed changes to the onsite electric supply and distribution system. Quantified the existing Brunswick PRA model using CAFTA computer code to determine the core damage frequency impact for the above application. Reviewed fault trees and event trees for the H.

B. Robinson plant as part of performance of the human reliability analysis (HRA). Utilized the EPRI time reliability correlation, the EPRI decision tree. the NUREG/CR-1278 (THERP), and the NUREG/CR-4772 (ASEP) methodologies. Observed licensed operators perform simulator exercises for Robinson and Brunswick.

Provided statistical expertise in the development Brunswick plant. Wrote a computer program of sampling plans at the and performed Monte-Carlo simulation of emergency diesel generator starts.

1985- 1990 Project Engineer, Corporate Nuclear Safety Section. Provided independent review of plant safety analyses and documents to assure proper maintenance -of nuclear safety; developed recommendations for improvements in nuclear safety; assured proper corrective action taken to prevent recurrence of events involving nuclear safety; and conducted special evaluations of selected safety-related matters.

1983-1984 Project Engineer, Nuclear Licensing Section.

Supervised up to 11 engineers and technicians within Licensing and functioned as first line management for projects and programs under his cognizance. Responsible for development and implementation of a computerized system for centrally tracking the company's regulatory commitments.

1981-1982 Technical Staff Engineer, Florida Power and Light Company St. Lucie Plant. Acted as licensing coordinator between the general office. Responsible for performingthe plant departments and Technical Specification Surveillance Testing and for secondary plant performance testing and evaluation. Successfully completed training programs covering plant systems with emphasis on mitigating core damage in the event of a nuclear accident; qualified to stand watch as needed in the capacity of Shift Technical Advisor. Received Senior Reactor Operator license from the NRC.

Page 4 001208

Steven A. Laur 1979- 1981 Project Engineer/TrialCoordinator.RCA Service Company. Supported suomanne acoustic trials at the Atlantic Undersea Test and Evaluation Center.

1978-1979 Design Engineer, Ford Motor Company. Engineer in the C-4 automatic transmission design unit.

1974- 1978 Nuclear Submarine Service. U.S. Navy. Served in various engineering assignments aboard a nuclear submarine.

Special Qualifications Registered Professional Engineer Senior Reactor Operator License. St. Lucie Nuclear Plant Nuclear Submarine Engineering Officer of the Watch Professional Affiliations Member of the American Nuclear Society (ANS)

Member of the Institute of Electrical and Electronics Engineers (IEEE)

Page 5 001209

PEER REVIEW OF SHEARON HARRIS PSA Prepared for:

Carolina Power and Light Company Prepared by:

&ru1V Engineeringand Research, Inc.

2111 Palomar Airport Road, Suite 180 Carlsbad, CA 92009-1419 (760) 929-0870 November 7, 2000 Final Report 001210

Shearon Hams PSA PeerReview TABLE OF CONTENTS Section

1.0 INTRODUCTION

................................................................................................... 1-1 1.1 Scope ........................................................................................................... 1-1 1.2 Review Approach ......................................................................................... 1-1 1.3 Review Team ................................................................................................ 1-3 2.0 REVIEW RESULTS FOR PRA ELEMENTS ........................................................... 2-1 2.1 IE - Initiating Events ...................................................................................... 2-1 2.2 AS - Accident Sequence Evaluation ............................................................ 2-5 2.3 TH - Thermal Hydraulic Analysis ................................................................. 2-8 2.4 SY - System Analysis .................................................................................. 2-9 2.5 DA - Data Analysis ..................................................................................... 2-10 2.6 HR - Human Reliability Analysis ................................................................ 2-14 2.7 DE - Dependencies .................................................................................... 2-15 2.8 ST - Structural Response .......................................................................... 2-17 2.9 QU - Quantification .................................................................................... 2-18 2.10 L2 Containment Performance ..................................................................... 2-19 2.10.1 Comments on Overall Level 2 Analysis .......................................... 2-19 2.10.2 Comments on ISLOCA Analysis ..................................................... 2-20 2.11 MU - Maintenance And Update Process .................................................... 2-22 3.0

SUMMARY

OF RESULTS AND CONCLUSIONS .................................................. 3-1

4.0 REFERENCES

....................................................................................................... 4-1

. -uiv Engineeringand Research. Inc. S01 1000001-1565-110700 001211

Review HamsPeer I PSA Shearon Shearon Ham~s PSA Peer Review LIST OF TABLES Table 1-1 PSA Elem ents ..............................................................................................

1-2 Table 2-1 Evaluation of Initiating Events Elements .......................................................

2-4 Table 2-2 Evaluation of Accident Sequences Element .................................................

2-7 Table 2-3 Evaluation of Thermal Hydraulics Element ...................................................

2-9 Table 2-4 Evaluation of Systems Analysis Element ....................................................

2-10 Table 2-5 Comparison of Bayes' Updated Failure Rate Distributions ......................... 2-12 Table 2-6 Evaluation of Data Analysis Element ..........................................................

2-13 Table 2-7 Evaluation of Human Reliability Analysis Element

..................................... 2-15 Table 2-8 Evaluation of Dependency Analysis Elements

........................................... 2-16 Table 2-9 Evaluations of Structural Analysis Elements ..............................................

2-17 Table 2-10 Evaluation of Level 1 Quantification Element

........................................... 2-18 Table 2-11 Evaluation of Level 2 Element ..................................................................

2-21 Table 3-1 Summary of Harris PSA Review Findings ....................................................

3-3 5 1W Engineeringand Research, Inc.

i S01 1000000l-1565.1 10700 001212

Shearon HarrisPSA PeerReview SECTION 1 INTRODUCTION 1.1 SCOPE The purpose of this report is to document the results of an independent review of the Shearon Harris PSA ("Harris PSA") that was performed in October 2000 by ERIN Engineering and Research, Inc. under contract with Carolina Power and Light Co. The purpose of the review is to support the use of the Harris PSA in an evaluation of scenarios that could cause or exacerbate a loss of cooling to the spent fuel pool at the Harris plant. The objectives of the review are to assess the quality, scope, and technical adequacy of the existing PSA models to support PSA applications and specifically the evaluation of postulated spent fuel pool scenarios. In addition, the peer review determines what enhancements to the PSA models are desirable to complete the spent fuel pool evaluation.

The scope of the peer review was limited to a documentation review.

No onsite visit or direct interaction with the PSA was performed as part of this peer review.

1.2 REVIEWAPPROACH The approach to conducting the review is based on a PSA peer review process that was originally developed by the Boiling Water Reactor Owners Group and is now being used by all of the existing LWR owners groups, including the WOG (Reference

[1]).

The purpose of this report is to document a peer review of the Harris PSA using the methodology and checklists developed for WOG peer reviews in order to support the application to the spent fuel pool evaluation.

&1V Engineeringand Research, Inc. 1-1 S0110000001-1565-110700 00.r21'3

Shearon HarrisPSA Peer Review The PSA Peer Review Process described in Reference [1] provides a structured process to review a PSA using a checklist that examines the PSA in terms of 11 elements and 209 sub-elements. These PSA elements are listed in Table 1-1.

Table 1-1 PSA Elements Element Code PSA Element IE Initiating Events AS Accident Sequence Evaluation TH Thermal Hydraulic Analysis SY System Analysis DA Data Analysis HR Human Reliability Analysis DE Dependencies ST Structural Response QU Quantification L2 Containment Performance MU Maintenance and Update Process Each of the above elements is further broken down into a total of about 209 sub elements to permit a structured and detailed examination of the PSA and its associated models and documentation.

A summary of the major review findings and recommendations for future updates is provided in Section 3.

-IM 51M Engineering and Research, Inc.

1-2 501 10000001-% 10700 001214

Shearon HarrisPSA PeerReview 1.3 REVIEW TEAM A review team consisting of Karl Fleming, Tom Daniels, Jeff Gabor, Ed Burns, and Grant Tinsley carried out this review. Karl Fleming is a recognized expert in developing and applying PSA technology and in performance of PSA peer reviews.

He served as the leader of the review team and was responsible for the review of about half of the PSA elements and for the preparation of this report. Ed Burns was the principal author of the PSA Peer Review process that was used in this report and has been involved in more than 25 peer reviews that have been completed using this methodology. Mr.

Fleming and Dr. Burns are also principal authors of the ASME PRA standard (Reference [2]) which is currently under development to support the PSA review process. Mr. Daniels has extensive experience as a PSA practitioner and is also playing a lead role in modifying the existing Harris PSA models to support the evaluation requested by the Atomic Safety Licensing Board. Jeff Gabor is a recognized expert in PSA technology and was a principal author of the accident progression analysis methodology that was used in the Harris PSA to support success criteria, sequence timing and source term assessment. Grant Tinsley also has extensive experience as a PSA practitioner and has experience in applying the industry peer review process at several plants. As a team, this group is adequately qualified and experienced to reach technically sound conclusions about the technical quality of the PSA and its capability to support the spent fuel pool licensing evaluation.

As a final note, while a grading system is used to provide a reasonable degree of consistency in the peer review process, the most valuable result of this type of review is a focused set of strengths and weaknesses that the PSA group can use in existing applications. Hence, the grades themselves do not assure PSA quality, but rather the supporting strengths and weaknesses that are identified from the grading process provide a roadmap for steps that should be taken to implement quality PSA applications.

Z". iV Engineeringand Research, Inc. 1-3 S01 10000001-1565.110700 o O.'21 5

k Shearon Hams PSA PeerReview SECTION 2 REVIEW RESULTS FOR PSA ELEMENTS The technical review was organized into the PSA elements listed in Table 1-1 so that the Industry PSA Peer Review Process could be closely followed. The results of the review for each element are provided in Sections 2.1 through 2.11 below.

2.1 IE - INITIATING EVENTS The scope of this element includes identification and grouping of initiating events (IE) and the estimation of initiating event frequencies.

A summary evaluation of the technical quality and adequacy of this element is provided in Table 2-1. The key points from this evaluation are summarized below:

" The extent of updating of the initiating events analysis and the use of formal calculations to document the in-house reviews provides good evidence that the initiating events analysis reflects the as-built and as-operated plant.

" The selection of an initiator to represent loss of normally running and standby charging pumps is unusual, but reflects thoughtful consideration of plant specific and unique features; this event leaves the plant with no high pressure makeup if the scenario should develop into a LOCA.

" The statistical methods used to estimate the initiating event frequencies include both classical (chi-squared) and Bayesian. In future updates all events should be quantified in a Bayesian framework for consistency, however this does not have an appreciable impact on the numerical results.

" The analysis of the frequency and recovery of loss of offsite power reflects the state of the art in the early 90's and would benefit from more recently collected and analyzed data such as those in Reference [3].

An updated analysis would likely show that the current analysis is somewhat conservative.

fJrljV Engineering and Research, Inc. 2-1 00.1216 S01 1OOO01-1565-110700

Shearon Hams PSA PeerReview

" The IE frequency analysis is well documented and is traceable to sources for generic and plant specific evidence, but the sources should be updated and a consistent Bayesian methodology should be used for all events.

" The steam and feed line break IE frequencies are well documented but based on somewhat old references and should be updated. An updated analysis would likely show that the current analysis is somewhat conservative.

"* Some of the IE frequencies developed from failure rates are not weighted by a plant availability factor and are therefore biased on the high side by the inverse of the availability factor. An updated analysis could show that the current analysis is somewhat conservative by 20% to 30%.

" A check of the Bayes update calculations is included as part of the data evaluation in Section 2.5

" The report contains a very good discussion of how initiating events are binned based on functional categorization and binning trees.

" ISLOCA initiating frequency evaluation appears quite conservative compared with other similar plants. This could be reassessed to make the PSA more realistic. Specific comments on the ISLOCA models are included with the review of the Level 2 PSA element in Section 2.10.

The Harris PSA report text says that IE fault trees for certain systemic events such as service water were linked with mitigation fault trees.

These initiating event fault trees were reviewed indicating a number of issues that are listed below.

The fault tree models used for initiating event frequencies include combinations of two or more basic events with an annual exposure time. An example is a cutset at 3.1 E-08 involving a Loss of Instrument Air initiator

(%T1 3) that includes the basic events ACP1ANS%FN and ACP1BN%SFN representing the A and B compressors. Both of these basic events use the compressor failure rate times 8760 hours0.101 days 2.433 hours 0.0145 weeks 0.00333 months . Common cause failures of 2 and 3 compressors are included separately. The fault tree logic should be such that these combinations are not generated or the combinations can be included in the mutually exclusive file.

The exposure time for the first failure is 8760 hours0.101 days 2.433 hours 0.0145 weeks 0.00333 months , but the exposure time for any subsequent failures is the repair time for the first component failed. Similar combinations exist in the Loss of Normal Service Water

(%T9) and Loss of CCW

(%T1 1) initiating event models.

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I Shearon Hams PSA Peer Review

, The fault tree model for the Loss of CCW initiating event does not include combinations of pump failure to run with a failure to reseat of the failed pump's discharge check valve. This is also true for the Loss of Normal Service Water

(%T9) and Loss of Instrument Air (%T13) initiating events.

" The models for loss of a 6.9kV bus initiating event consist of a single basic event representing bus failure during operation multiplied by 8760 hours0.101 days 2.433 hours 0.0145 weeks 0.00333 months .

The model should account for other failure modes, such as circuit breaker transfers open while indicating closed.

"* The model for loss of non-safety DC bus DP-1A (%T15) consists of 2 basic events, one for battery short and one for distribution panel failure during operation. The model should account for other failure modes, such as battery charger failure during operation in combination with battery failure on demand, assuming the battery is sufficient to supply loads without the charger.

" The inclusion of "failure of NSW pump strainer to run" leading directly to Loss of Normal Service Water (%T9) may be overly conservative. It is a dominant contributor to the initiating event frequency.

" The fault tree logic correctly treats the dependencies associated with some of the same components and failure modes appearing in the initiating event frequency fault trees and the mitigation function fault trees by using different basic event names for each. Flags are used to make sure that equipment is not permitted to fail more than once.

" The initiating events quantified using fault trees as well as those quantified using data did not have an availability factor applied to account for the probability that the plant is in operation at the time of the event; this creates a conservative bias on the order of 10% to 20% in the initiating event frequencies.

"* In the loss of charging system initiating event model, it is not clear that the correct times were used in modeling cutsets with valves transferred closed in combination with pump failures.

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Shearon Harmis PSA PeerReview Table 2-1 Evaluation of Initiating Events Elements otWrinuue Assessment Guidance It is not known whether CP&L has any separate guidance doc.ments but the PSA documentation of the initiating event analysis is serves as useful guidance to support PSA updates. very thorough and hencec Grouping The identhcIontr grouping that h of iliti ofittingwas very thorobgh anb events comprehensive. Good use of insights from EPRI ATWS study and PSAs on other Westinghouse PWR plants. Event trees used for binning provide a very clear description of how each event was treated.

Treatment of Support A systematic search for plant specific impotnt inicitors was per formed System/Special system notebooks The final list of initiating events in the Initiators includes a relatively large number of support system initiators which is appropriate only PSA in for PWR plants. The otherat has a larger list of initiating events Calvert Cliffs, which is an industry outlier in this is probably respect. The Harris PSA list appears to be as complete as other PSAs such as South Texas which are regarded as quality PSAs. A very good level of completeness could be brought to excellent if the list was expanded to consider common cause bus failures as initiating events (there have been a few events at other plants where degraded components inside switchgear cabinets have failed energetically, and other events in which fuses inside multiple inverters have blown taking in are failures outmultiple multipl icsistet e support systems samethat time. Another category functionally interact, of events such as failure of a CCW train with a SW train supporting the opposite CCW train (some refer to the occurrence of initiating precursors).

The ISLOCA frequency appears unrealistically high and may bias the results of the PSA in certain applications.

Data One spect otf the initiating events analysis that hassignifi c nt room for improvement The initiating event frequency calculations were very clearly documented and traceable back to the source data, but several problems were noted. One is the inconsistent use of classical and bayesian statistical method. The other is the need to update the generic database as the NRC via INEEL has recently published some more up to date and more realistic sources.

event Such an for frequencies update some is expected to result in somewhat lower initiating events.

DocumerPtation The documentation initiating events analesof s excellent and would likely getquite analysis a grade level 4r There transparent. The documentation is very good use is very clear and of figures and makes the tables and key data and assumptions are justified and the source.

Recomm-ended bases traceable back to the Update the generic dlaia inputs from the INEEL study and perform a consist-ent Enhancements Bayes' update for a initiating dl events using the generic distributions from the INEEL study.Include distributions, Updatethethe lossavailability plant of offsite factor power frequency and timemodels.

Overal Prcess The initiating events analysis is capable ofin all initiating event to repair Assessment supporting risk significance determinations with deterministic input. The current initiating events analysis would be peer reviewed at a solid Grade 3, and if the recommended enhancements were made with the same level of quality of other aspects of Ithe analysis, this element would be at Grade 4.

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L Shearon HarrisPSA Peer Review Recommended Element Grade:

EOGrade 1 - Supports Assessment of Plant Vulnerabilities O Grade 2 - Supports Risk Ranking Applications I Grade 3 - Supports Risk Significance Evaluations w/Deterministic Input O Grade 4 - Provides Primary Basis For Application 2.2 AS - ACCIDENT SEQUENCE EVALUATION The scope of this element includes identification and grouping of functional accident sequence categories, event tree development, core damage and plant damage state characterization and interface with other elements such as initiating event identification and systems dependency analysis. A summary evaluation of the technical quality and adequacy of this element is provided in Table 2-2. The key points from this evaluation are summarized below:

"* Sequence definition is accomplished via event sequence diagrams and event trees that are very clearly documented

"* Very thorough treatment of transient induced LOCAs via PORV and safety valve pressure challenges.

" Used NUREG-1 150 RCP seal LOCA model; according to the Westinghouse Owners' Group, the assumption of a 1.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months time to begin leaking is now regarded as questionable for some of the RCP seal failure modes, especially the so-called popping failure mode. A more up to date model, such as the Rhodes model, should be considered for the next update.

" Good development of event tree logic from safety functions and clear presentation of success criteria.

" Transient ESD/ET does not credit alternate feedwater as indicated in the EOPs/FRGs for sequences in which there is failure of main feed, auxiliary feedwater, and feed and bleed.

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Shearon HarrisPSA Peer Review

" Figure 4-5 has no labels, just graphics. (not critical to the review)

" Table 4-6 presents pump capacities but not the pressures that corresponds to these capacities; it would also be helpful know the shutoff heads of these pumps.

"* No credit taken for operators to terminate sprays results in a conservative treatment of RWST depletion time.

" The LOSP event tree does not discuss how RCP seal LOCAs are treated, nor does it describe how offsite power recovery is factored in. This needed to be determined from other documents. It is confusing that SBO sequences include both Auxiliary Feedwater System (AF) success and AF failed sequences. There must be logic in the recovery file that effectively creates different sequences.

Certain sequences appear to be missing: Loss of offsite power, RCP seal LOCA starts, offsite power successfully restored, core damage due to failure to continue to provide makeup or other LOCA mitigation functions. This would be evidenced by transfers to the Small LOCA event tree if these conditions were met

" Bridge tree is judged to address all relevant plant damage state issues.

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L Shearon HarrisPSA PeerReview Table 2-2 Evaluation of Accident Sequences Element Attribute Assessment Guidance It is not known whether CP&L has any separate PSA documentation of the accident sequence guidance documents but the analysis is hence, serves as useful guidance to support PSA updates. very thorough and Success Criteria and Success criteria tabulated for each event tree and is generally traceable back Bases the source calculation. A number of the success to criteria are acknowledged to be conservative and lead to an overstatement However, the success criteria developed from of risk for some applications.

LOCAs do not initiate for 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months alludes to the assumption that RCP seal information provided by Westinghouse for the an error in NUREG-1550 in how interpreted. Information from the WOG indicates use in the expert elicitation was modes may occur almost immediatey. In addition, that some RCP seal failure if seal injection is lost for 10 minutes it should the WOG has indicated that not be restarted to prevent a thermal shock induced seal failure.

Accident Scenario There was a good use of safety functions and Evaluation (Event document sequence development. The event event sequence diagrams to Tree Structure) relation to many other plants using the same tree structures are simplified in methodology, but otherwise well documented. The LOSP event tree appears to SBO sequences with ,uccessful and unsuccessful be overly simplified; for example collapsed into the same sequence. It does not auxiliary feedwater are with RCP seal LOCAs and successful OSP recovery appear that LOSP sequences have been processed as SLOCA sequences, and this simplification significant impact on the baseline CDF but is optimistic. This does not have a applications in which changes to the ECCS could be a problem for certain system were being evaluated.

Interfacenwith There are numbentrencs of to mitigation EOPsmAOPs the EOPs and FRGs to justify and develop the posi ibilitieslt n that analys re d in of these cases the full mitigation addressed in event trees, however in several some conservatisms. the EOPs is not credited creating Sequence EndJ State IShe Level 1 end states efinition/Treatment interface with the Level are success and core damage; a Bridge Tree is used to 2 Containment Analysis through the definition plant damage rresolve states. The dependencies of a set of plant damage between the Levelstates 1 andtrack sufficient Level 2 eventinformation models. The Level 1 end sequence to state provides a clean interface with the anlysis and is clearly described. Level 2 ocumentation The documentation is excellent making nd/rieniid aam assumptions quite transparent. the event tree analysis and Recommended Enhancements A more up to date RCP seal LOCA model should be used, and for some applications it might conservatisms. be event The LOSP fruitful tree to should eliminate some of and be expanded the the above noted links to the Small LOCA tree for RCP seal LOCAs and successful OSP recovery and/or identified. created Overall Process When the recmme~nded enhancemients Assessment are incorporated this element of the PSA is of sufficient quality to support risk significant evaluations with deterministic input. However even without these enhancements, this aspect of

= ~the PSA is adequate to -neet most applications.

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Shearon Hams PSA PeerReview Recommended Element Grade:

o3 Grade 1 - Supports Assessment of Plant Vulnerabilities O Grade 2 - Supports Risk Ranking Applications

O Grade 4 - Provides Primary Basis For Application 2.3 TH - THERMAL HYDRAULIC ANALYSIS The scope of this element includes the engineering and thermal hydraulic analysis that supports the PSA in several key areas including success criteria for the accident sequence model and systems analysis, time windows for HRA, room heatup analyses, and plant and containment analyses that support the

-Level 2 PSA. A summary evaluation of the technical quality and adequacy of this element is provided in Table 2-3.

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Shearon Hams PSA PeerReview Table 2-3 Evaluation of Thermal Hydraulics Element A#+Ak

I fLu.. W Assessment r."iirlnnr.* It is not known whether CP&L has Assessment any separate guidance documents but the PSA documentation of the thermal hydraulics analysis is very thorough and hence, serves as useful guidance to support PSA updates.

Best Estimate The thermal hydraulics analysis supports the PSA success criteria, time Calculations windows for human operator actions, and analysis to support the Level 2.

These analyses include simple mass and energy balances, plant specific MAAP analyses, and operator training simulator exercises for key accident sequences. In general, these analyses are realistic, technically sound, well documented and traceable to adequately support the assumptions. PSA models and Room heatup calculations were performed and appear to be realistic Room Heatup and yield Calculations reasonable results.

This TH analysis is among the best documented for a PSA that Documentation the review team has seen. There are many figures and charts that display a deep understanding of the results and their limitations.

Recommendations Maintain this quality level as the PSA experiences future updates.

for Enhancements Overall Process This element of the PSA is a key strength and is capable of supporting Assessment the envisioned risk informed applications.

Re-o4,,,*lU; o m n r;IemeInti %raue:

"O Grade 1 - Supports Assessment of Plant Vulnerabilities "O Grade 2 - Supports Risk Ranking Applications "0 Grade 3 - Supports Risk Significance Evaluations w/Deterministic Input j Grade 4 - Provides Primary Basis For Application 2.4 SY - SYSTEM ANALYSIS The scope of this element includes the analysis of systems that support the PSA, including the identification of system functions addressed in the PSA, system success criteria, system dependencies, potential for causing an initiating event, fault tree models that support the accident sequence models, and probability models to support the accident sequence quantification. A summary evaluation of the technical quality and adequacy of this element is provided in Table 2-4.

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Shearon Hams PSA PeerReview Table 2-4 Evaluation of Systems Analysis Element Attribute Assessment Guidance It is not known whether CP&L has any separate guidance documents but the Systems Notebooks are very thorough and hence, serves as useful to support guidance PSA updates.

Systems Modeled 23 original systems notebooks and one additional system notebook (SFP cooling & cleanup) have been developed. All expected SSCs (i.e.,

contributors) modeled. "Extra" SSCs (above usual scope) observed PWR CDF demin water, non-safety related SW and main FW (and SFP cooling including

& cleanup as previously mentioned).

System Model The fault tree models are extensive. They are consistant with Structure (Fault Tree) industry practice in all aspects. Excellent or exceed system-to-system consistency in format, structure, and level of detail.

Success Criteria Plant-specific T-H calcs for CVCS, AFW, RHR, ESW, MFW.

Extensive, appropriate use of plant-specific MAAP model to support systems success critena development.

Recommended More frequent system updates (last complete update was 1995).

Enhancements more flexible, independent update for those SSCs undergoing Consider modifications could be considered. This should be possible given thesignificant either available or imminent (i.e., FORTE solution engine, etc.) and software with the "2000" update to make all notebooks controlled the move engineering calculations.

Overall Process Very difficult to find fault with even the 1995 notebooks; the few Assessment "2000" notebooks reviewed were even better. One of the very bestcompleted examples of systems analysis and documentation this reviewer has seen in the industry, including those studies in which he has managed or participated.

to-system The system consistency is outstanding.

Recommended Element Grade:

"o Grade 1 - Supports Assessment of Plant Vulnerabilities "OGrade 2 - Supports Risk Ranking Applications "O Grade 3 - Supports Risk Significance Evaluations w/Deterministic Input I Grade 4 - Provides Primary Basis For Application 2.5 DA - DATA ANALYSIS The scope of the evaluation for this element of the PSA includes development of component failure rates, maintenance unavailabilities, common cause parameters and miscellaneous data parameters that are needed to calculate the PSA basic event

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IL Shearon Harris PSA PeerReview probabilities. The estimation of initiating event frequencies is covered in the initiating events analysis element.

The following items were identified as part of the data review.

" Plant specific data was only collected and analyzed for selected pumps, the diesel generators, and chillers. No justification was given for why plant specific data was not collected for other key components such as valves, breakers, etc.

It would be preferable to use the risk importance measures to decide where plant specific data should be collected.

" A rather convoluted procedure was used to apply Bayes' updating plant specific experience on the above mentioned components to combine and generic uncertainty distributions. This procedure was used to take advantage of conjugate properties of certain contributions of assumed prior distributions and assumed likelihood functions. ERIN checked the Bayes' update with a more commonly used procedure where the prior distribution is assumed to be lognormal, the binomial likelihood function is used for demand based failure rates, and the Poisson distribution is used for time based failure rates. The ERIN analysis was performed using a proprietary software known as BARTTM (Reference [3]). The results of this comparison are shown in Table 2-5 and show good agreement. CP&L should consider using the recommended procedure as discrepancies in the case of the diesel generator failure rate of approximately 10% could be significant under certain conditions.

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Shearon Harmis PSA Peer Review Table 2-5 Comparison of Bayes' Updated Failure Rate Distributions Componentl Failure Mode Range Method' Mean 50%Tile 95%Tile Motor Driven AFW Pump Fail Factor

!HasPSA 2.2E-3 1.9E-3 4.5E-3 to start (UD-1) 2.4 ERINh 2.OE-3 1.8E-3 4.1E-3 Motor Driven AFW Pump Fail 2.5 Harris PSA 2.6E-5 1.5E-5 8.2E-5 to run (UD-2) 5.4 ERIN 1 2.6E-5 1.5E-5 8.1E-5 Turbine Driven AFW Pump 5.3 Harris Fail to start (UD-3) ERIN 1PSA 1.2E-2 1.4E-2 1.2E-2 1.3E-2 1.5E-2 2.0E-2 1.3 Containment Spray Pump Fail Harris PSA 2.1 2.7E--53 1.6E-5 8.5E-5 5.4 to run (UD-16) ERIN 1 2.7E-5 1.6E-5 8.5E-5 5.4 Diesel Generator Fail to Start Harris PSA 7. 1E-3 6.4E-3 1.3E-2 2.0 (UD-17) "ERIN 1 7.8E-3 7.E-3 ',.3E-2 1.9 Chiller Fails to run (U-D-20) Harris, PSA 7.8E-5 76- 1.E41.5 ERIN 1 7.8E-5 7.5E-5 1.1E-4 "ERIN 2 1.AE-4 1.5 9.9E-5 1.7E-4 1

'METHODS: Harris PSA: As calculated several assumed distributions in Harris PSA using numerous steps converting between ERIN 1: Use of lognormal prior distribution, binomial likelihood for demand, and Poisson likelihood for run; otherwise same generic means, range factors and plant specific evidence as in Harris PSA ERIN 2: Same methodology as ERIN 1 but with increased range factor to correct inconsistency between assumed generic distribution and plant specific evidence.

" in the analysis of the Chiller failure rate for failure to run (UD-20), the plant specific evidence of 9 failures in 76451 hours whose point estimate is 1.18E-4 per hour seems inconsistent with the assumed prior distribution. The probability of observing so many failures in this exposure time is very low under the assumptions of the prior distribution. In Table 2-5 a revised analysis is performed (ERIN 2) in which the prior distribution has been revised to increase the range factor to 5 to be more consistent with the evidence.. As seen in the table the results are sensitive to the assumed prior distribution.

procedure should be revised to check the consistency The Harris PSA of the prior distributions and the evidence and adjustments like this should be made before accepting the results of the Bayes update.

" Maintenance data treatment used same inconsistent methods as initiating events; purely generic data for most components and purely plant specific data IUI Engineeringand Research, Inc. 2-12 S0110000001-1565-110700 001226

I Shearon Hams PSA PeerReview for selected components. A consistent Bayes' update approach could be used as with failure rates.

Very good example of how common cause events should be modeled and how data should be screened and mapped for plant specific application. The common cause data source is somewhat outdated and should be updated using the INEEL database.

A summary evaluation of the technical quality and adequacy of this element is provided in Table 2-6.

Table 2-6 Evaluation of Data Analysis Element Attribute Assessment Guidance/ It is not known whether CP&L has any separate guidance documents but the Documentation Data Analysis Documentation is very thorough and hence, serves as useful guidance to support PSA updates.

Plant Specific Generic database for component failures is well documented. Plant specific Component Data data was only incorporated for selected components including pumps, emergency diesel generators, and chillers. No valid basis for excluding other risk significant components was provided. The Bayes' update methodology employs a sequence of modeling assumptions to take advantage of conjugate properties, but independent verification of these calculations indicates some discrepancies in comparison with a more straightforward Bayes' update procedure. In one case, chiller failure to run, the plant specific evidence was inconsistent with the assumed generic distributions.

System/ Train Plant specific maintenance unavailability data was collected for the same Unavailabilities limited set of components as was used for failure rate determination. Generic maintenance data was used for other components. As with the initiating event frequencies, inconsistent methods were used to quantify the plant specific and generic maintenance unavailabilities: classical statistics used for analyzing plant specific data, but Bayesian method used for those quantified using generic data.

Common Cause The common cause failure analysis including the selection of components for Failure CCF modeling, plant specific screening and mapping of CCF data and Quantification justification and documentation of data interpretation assumptions is excellent.

The only aspect of the common cause analysis that could be improved significantly is the incorporation of the INEEL common cause database.

Unique Unique modeling unavailabilities are documented in the system notebooks and Unavailabilities and throughout the report. Offsite power recovery should be updated to Modeling Issues incorporate the latest industry offsite power recovery data.

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Shearon Hams PSA PeerReview Attribute Assessment Recommended A more straightforward, less convoluted, and consistent Enhancements procedure should be used for initiating event frequencies, Bayes' update failure rates and maintenance unavailabilities. The use of narrow generic inconsistent with the plant specific evidence should distributions that are be avoided. Plant specific data should be applied for all risk significant components.

treatment of CCF data should be updated to incorporate The excellent the INEEL CCF database.

Overall Process This element of the Harris PSA, with the exception Assessment of the excellent treatment of common cause, was not up to the level of quality of most other aspects of the PSA. While the data values used in the PSA are reasonable, the data handling methods for failure rates and maintenance could be improved. The current data treatment should be adequate for risk ranking and most risk informed applicatioris but should be enhanced for risk informed applications in which the plant specific equipment performance is an issue.

Recommended Element Grade:

13 Grade 1 - Supports Assessment of Plant Vulnerabilities M Grade 2 - Supports Risk Ranking Applications O3 Grade 3 - Supports Risk Significance Evaluations w/Deterministic Input O Grade 4 - Provides Primary Basis For Application 2.6 HR - HUMAN RELIABILITY ANALYSIS The scope of this element includes the systems analyses that support the PSA including the identification of system functions addressed in the PSA, system success criteria, system dependencies, potential for causing an initiating event, fault tree models that support the accident sequence models, and probability models to support the accident sequence quantification. A summary evaluation of the technical quality and adequacy of this element is provided in Table 2-7.

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Review Peer PSA L Hams Shearon Shearon Ham~s PSA Peer Review Table 2-7 Evaluation of Human Reliability Analysis Element I Auribute , A1 ntl ;,*l~ll i il

,udncude It is not known whether CP&L has any separate guidance documents but the Human Reliability Analysis Documentation provided in Appendix thorough and hence, serves as useful guidance to support E is very PSA updates.

Proe-initiator Human A reasonably complete treatment of pre-initiator Human Actions Actions was provided in Annex A. A good basic treatment of these events is provided and is well documented.

Both ASEP and THERP methods were use fr post initiator Post Initiator Human hu Tan -actions Actions and these are very well documented in Annex B and C.

A good basic treatment of these events is provided and is well documented.

Treatment of Human actions were placed directly on the fault trees, Which -necessitatedan Dependencies evaluation to see if multiple human errors would impact sequence truncation and quantification. The version of the quantification that was described used a relatively high truncation frequency of lx1O per year only found two cutsets that had been screened out. This procedure should be extended to a lower Docume-ntation truncation no greater than 1x10"°0 .

The documentation of the human reliability analysis in Appendix E Annex's A, Recommended B, C, and D was excellent.

Enhancements Dependencies among multiple HEPs were performed for the 1995 model. The relatively high truncation value (1E-8/yr) indicated only two cases to be addressed. It may be prudent to reevaluate this multiple HEP assessment in the future. These multiple operator actions are believed to impact principally those sequences contributing to late containment failure sequences or to Overall Proes SGTR.

Assessment The Harris PSA includes a treatment of human reliability analysis that is sufficient to treat risk-informed applications.

rnecommenoed Eiement Grade:

O3 Grade 1 - Supports Assessment of Plant Vulnerabilities O3 Grade 2 - Supports Risk Ranking Applica.tions 0 Grade 3 - Supports Risk Significance Evaluations wlDeterministic Input o3 Grade 4 - Provides Primary Basis For Application 2.7 DE - DEPENDENCIES The scope of this element includes the treatment of dependent failures between and among systems and initiating events due to functional, physical, and human sources.

Treatment of common cause failures and spatial dependencies due to internal hazards

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Shearon Hams PSA PeerReview such as flooding are also included in this element. A summary evaluation of the technical quality and adequacy of this element is provided in Table 2-8.

Table 2-8 Evaluation of Dependency Analysis Elements Attribute Assessment Guidance/Document It is not known whether CP&L has any separate guidance documents but ation the Dependency Analysis Documentation is very thorough and hence, serves useful guidance to support PSA updates. as Dependency The Harris PSA did not include dependency matrices, however the system Matrices dependencies are delineated separately on a system by system basis in the system notebooks. A more holistic consolidation of dependencies in a single place would make it easier for system engineers to validate plant to model fidelity.

Common Cause The common cause failure treatment including a comprehensive set Failure Treatment of components and failure modes, and the plant specific screening and mapping of CCF data was excellent.

Spatial The internal flooding analysis is excellent including the screening of flooding Dependencies scenarios, treatment of a range of flood rates, analysis of spatial interactions, and detailed quantification of high-risk scenarios. The documentation of this aspect of the PSA was the best that this reviewer has seen.

HI Dependencies The treatment of HI dependencies was adequate but could have been improved by extending the examination of truncated sequences down to a

truncation frequency of lx1 0"10.

Recommended Add dependency matrices, extend common cause data analysis to include Enhancements the INEEL common cause data base, and extend the HI dependency treatment to 1x10x10 per year.

Overall Process The current dependency treatment is adequate to support risk informed Assessment applications and with the recommended enhancements would provide the primary basis for application.

Recommended Element Grade:

0 Grade 1 - Supports Assessment of Plant Vulnerabilities O3 Grade 2 - Supports Risk Ranking Applications O Grade 3 - Supports Risk Significance Evaluations w/Deterministic Input

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F Shearon Hams PSA PeerReview 2.8 ST - STRUCTURAL RESPONSE The scope of this element includes the structural analyses that support key aspects of the PSA. These structural issues include the success criteria for vessel integrity during severe over pressure and over cooling transients, the capability of the containment in response to severe accident challenges, and the modeling of the response of low pressure piping when exposed to high pressure loads during an interfacing systems LOCA evaluation. A summary evaluation of the technical quality and adequacy of this element is provided in Table 2-9.

Table 2-9 Evaluations of Structural Analysis Elements EM ýa

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Shearon Hams PSA PeerReview 2.9 QU - QUANTIFICATION The scope of this element includes the parts of the PSA not considered in the previous elements to support quantification of CDF and interpretation of the results. A summary evaluation of the technical quality and adequacy of this element is provided in Table 2-10.

Table 2-10 Evaluation of Level 1 Quantification Element dLLFIUbue Assessment Guidance/ Very good guidance (HNP-F/PSA-0001) on quantification and Documentation recovery. This procedure contains a step-by-step process for duplicating the documented results with the controlled model and quantification files. FLag files, mutually exclusive event files and recovery files are explicitly documented.

Minor point, but rearange Table 6-1 in decreasing CDF order instead of alphabetically by sequence name (i.e., same order as Section 6.1.1). The summary of results in Section 1 is weak; at least discuss the top few sequences instead of just showing pie charts and saying, "nothing really dominates - see Section Dominant Accident 6."

Very good guidance (HNP-F/PSA-0001) on quantification and Sequences recovery. This procedure contains a step-by-step process for duplicating the documented results with the controlled model and quantification files. Flag files, mutually exclusive event files and recovery files are explicitly documented.

Minor point, but rearrange Table 6-1 in decreasing CDF order instead of alphabetically by sequence name (i.e., same order as Section 6.1.1). The summary of results in Section 1 is weak; at least discus the top few sequences instead of just showing pie charts and saying, "nothing really dominates - see Section TruncationlRecovery 6."

Truncation in the most recent calculation file (2000 model)

Analysis was 4E-9 in contrast with the IE-8 level in the PSA report. This is 1E-4 less than CDF and is consistent with current industry standards. The documentation cites the largeness of the SHNPP fault trees and the excessively large number of cutsets that would result from lower truncations as justification.

The recovery analysis uses the rule-based EPRI R&R QRECOVER software to apply recoveries to sequences on a detailed, systematic and reproducible basis.

Uncertainty No statistical uncertainty analysis of the results was apparent.

Results Summary Documentation in Section 6.1.1 is strong on what has been done. The documentation does not generally cover the underlying reasons or assumptions regarding why certin choices are made. For example, on top sequence TQUB (30.92% of CDF), why is it that fails seal cooling in the internal flooding sequences? Why does the TDEFW pump fail when the battery depletes? The presentation of key assumptions in the the summary is good but a discussion of the impact of these assumptions on the results is missing.

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IL Shearon Hams PSA PeerReview Attribute Assessment Recommended Document an uncertainty and sensitivity analysis Enhancements and provide engineering insights into the results; update the results summary to account for the more recent update.

Overall Process The quantification element subject to the recommended enhancements is Assessment capable of supporting risk informed applications.

Recommended Element Grade:

o Grade 1 - Supports Assessment of Plant Vulnerabilities

[O Grade 2 - Supports Risk Ranking Applications

M Grade 4 - Provides Primary Basis For Application 2.10 L2 CONTAINMENT PERFORMANCE The scope of this element includes the interface with the Level 1, the containment event tree structure, the phenomena considered, applicable systems and operator actions, and end states. A summary evaluation of the technical quality and adequacy of this element is provided in Table 2-11. The key points from this evaluation are summarized below. Comments relevant to the overall Level 2 analysis are presented in Section 2.10.1, while detailed comments on the interfacing systems LOCA analysis are provided in Section 2.10.2.

2.10.1 Comments on Overall Level 2 Analysis The overall Level 2 Analysis is state of the art and is considered a key strength of the Harris PSA. Specific comments are found in Table 2-11.

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Shearon Hams PSA PeerReview 2.10.2 Comments on ISLOCA Analysis Due to the importance of ISLOCA sequences as a contributor to LERF for the Harris PSA, a special review was performed to identify the need for any updates to provide a more realistic evaluation of these events.

General Comments in ISLOCA

" There are combinations of failure modes included that do not make physical sense. Until the inboard valve is failed (by rupture, leak, or mispositioning, if possible) the outboard valve(s) have no demand and cannot be questioned. The only way an outboard valve can be "holding" is if the inboard valve has already failed in some way.

" Common cause failures of valves in the rupture failure mode should not be modeled. Any conceivable maintenance errors that might be called common cause rupture would be discovered long before reaching full power operation.

"* The rupture and leak failure modes can be combined, since a rupture is just a large sized leak and they are both time-dependent failures.

Frequencies can be determined as done for Seabrook, using a Frequency vs. Leak Rate plot developed from actual data.

" The discussion of low pressure pipe failure due to overpressurization should include at least some discussion of other possible pressure boundary failures, i.e., pump seals, bolted flange connections, valve packing, valve body rupture, etc.

RHR Suction Lines

" The model for these lines does not include the leak failure mode. Combining the leak and rupture failure modes would take care of this.

If leaks were modeled, they would have to exceed the flow capacity (900 gpm) of the relief valve in the low-pressure portion of the line in order to pressurize the low-pressure piping.

"* Rupture of the outboard valve followed by rupture of the inboard valve is not a credible scenario. (See General Comments Above)

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I Shearon HarrisPSA PeerReview

Failure of the inboard valve to hold on demand is not a credible failure mode.

The inboard valve always sees the same pressure during power operation.

LHSI Cold Leg Injection Lines

"* The failure modes 'Fails Stuck Open' and 'Fails to Hold on Demand' are just different ways to describe 'Fails to Reseat' and should not both be included for these check valves.

"* Common cause failures by the rupture failure mode and the leak failure mode should not be included. (See General Comments Above)

Table 2-11 Evaluation of Level 2 Element Attribute .

Assessment Guidance/ It is not known whether CP&L has any separate guidance documents Documentation PSA documentation of the L2 containment analysis is very thorough but the and hence, serves as useful guidance to support PSA updates. The documentation is excellent.

Level llLevel 2 The interface is provided by a bridge event tree that links the Interface Level 1 ET and Containment Event tree. The Bridge Tree supports the definition and assignment of a comprehensive set of plant damage states that are sufficient to capture severe accident issues relevant to PWRs with large dry containments.

This part of the Level 2 analysis is very clearly documented and is technically sound.

Containment Event The CET considers all the severe accident phenomena that Tree Phenomena, are expected for this plant and containment type and account for all the relevant Systems, Human NUREG-1 150 issues. In addition, direct corium attack of the liner was identified Actions, Success and modeled as a result of plant specific evaluation of containment features.

Criteria for in-vessel recovery, arrest of corium attack of basemat Success criteria via debris bed cooling, and time windows for restoring vessel and core cooling are reasonable.

This part of the Level 2 analysis is very clearly documented and is technically sound.

Containment A plant specific probabilistic evaluation of the containment failure modes was Capability performed and used to convolute against the assessed Assessment pressure and temperature loads to calculate the containment failure probability.

This part of the Level 2 analysis is very clearly documented and is technically CET End States sound.

The CET release categories provide an adequate spectrum of possible containment releases to support source term definition and LERF Definition calculations.

The PSA does not calculate a LERF but provides sufficient information on the definition of the release categories to estimate LERF.

li-t lEngineerngand Research, Inc. 2-21 S0110000001-1565-110700 001235

Shearon Harmis PSA PeerReview A Assessment i Recommended d interfating system LOCA frequeng--cies are considered of inappropriate common cause failure modesn to aleve Enhancements app Incato usion uet of conservative data for valve failures. It is expected that an updated evaluation would support up to a one order of magnitude reduction in V-LOCA frequency. Some of the CET Probabilities on Gualipitatve judgemniants and could be questioned are simply assigned based as to ao be reasonable. More could be done to compare their basis, but seem these assessments to Snd strn or NUREG-1 150 results, however no significant changes in the results would be expected, only deeper insights into results. the contributors to the Level 2 vaus, s rressf thq s fLevelanysis Asessn sucrienti to support IS State risk o the art determinations significant for plants in its peer with group and is already deterministic input.

u s etIncorporation of enhancements would bring this PSA element to a level sufficent to Provide the primary basis for decision making.

Recomended Element Grade:

0] Grade 1 - Supports Assessment of Plant Vulnerabilities 13 Grade 2 - Supports Risk Ranking Applications

[] Grade 3 - Supports Risk Significance Evaluations w/Deterministic Input

@ Grade C4 - Provides Primary Basis For Application (Conditions) 2.11 MU - MAINTENANCE AND UPDATE PROCESS As part of a normal WOG PSA Peer Review, there is an evaluation of the maintenance Sand update process. This more complete evaluation involves presentations by the PSA team, discussions regarding applications, review of onsite procedures, etc. Due to the nature of what is this element. Oneinvolved in this type of review, a peer review was not comment that is made is that the PSA documentation, performed for particularly the details provided on the use of the CAFTA and R&R Workstation tools to support the various steps of the quantification process, provides excellent guidance for future PSA updates. However, to review this element requires access to information not available to support this peer review, which is based on the review of the existing PSA documentation and models.

-* Engineeringand Research, Inc. 2-22 S01 10000001-1565.110700 0 0.'2 3()

I Shearon Hams PSA PeerReview SECTION 3

SUMMARY

OF RESULTS AND CONCLUSIONS The results of the evaluation of the Harris PSA elements are summarized in Table 3-1.

This table includes the projected PSA Peer Review Grades that the reviewers believe would occur from a full peer review according to the WOG PSA Peer Review Program (Reference [1]). Also included are the strengths and weaknesses of each PSA element that the formal peer review would normally be documented in the Fact and Observation forms. A more complete delineation of these points is found in Section 2.

On balance this PSA is viewed as one of the best-documented PSAs that the reviewers have seen. The systems analysis, thermal hydraulics analysis, containment performance analysis and the dependency analysis were especially well done and are projected for evaluations at grade Level 4 or close to this grade. With the exception of data analysis, which was assessed at grade Level 2, the remaining elements of the PSA were at or near grade Level 3 with only small numbers of issues to clear up in order to achieve this grade level.

The Harris PSA is viewed as capable of supporting risk-informed applications such as the spent fuel pool PSA evaluation that this review was performed in support of. In each application, the applicability of the strengths and weaknesses identified in this review should be reviewed and addressed to determine whether they impact the conclusions of the application. When such impacts are identified, they should be addressed via PSA updates, sensitivity analyses, and/or supplemental engineering analyses as appropriate to support the decisions or conclusions associated with the application. In the opinion of the reviewers, this PSA is in the upper quartile of PSAs in the nuclear industry today; when ranked in terms of the capability to support risk informed applications.

& -iUVEngineeringand Research, Inc.

3-1 S0110000001.o o 1 C1 fffflI

Shearon Hams PSA PeerReview An important finding of the peer review is that the PSA can be used to assess the CDF, Containment Failure Frequency, and Containment Bypass Frequency.

If all the specific technical issues raised in this review were resolved, and incorporated into a PSA update, it is expected that the estimated CDF values would be comparable to or lower than those reported in the Harris PSA report, however the uncertainties are larger than those quoted in the report due to the issues noted for the data element.

If the issues impacting LERF were addressed in a similar fashion, it is expected that the current LERF results that are supported by the existing PSA be determined to be conservative primarily from conservatisms in the estimation of interfacing systems LOCA frequency.

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X2 SO11000001-15-110700 002 38

Shearon Harris PSA Peer Review Table 3-1 Summary of Harris PSA Review Findings Element Element 1 - =m, ,* 1===

Code PSA Element Grade* Strengths 1 .1 Grade' Potentalro nhancsSemern Rvie I: Initiating Events Excellent treatment of support system initiating events; Mixture of classical statistics for some events 3

Clear functional grouping and binning for sequence and Bayes' treatment of other events; problems model; good use of fault trees for selected initiators with system fault trees for initiators AS Accident Sequence C3 Very clear documentation of event trees, success Loss of offsite power event tree shifts important Evaluation criteria and interface with fault tree quantification; good details of the logic into the linked fault trees that use of ESDs; excellent Level 1/Level 2 interface are more difficult to review.

TH Thermal Hydraulic 4 Excellent and traceable documentation None Analysis SY System Analysis 4 Excellent and traceable system notebooks DA None Data Analysis 2 Good generic database that is traceable to sources; Only limited amount of plant specific data; excellent CCF data treatment inconsistent use of Bayes and statistical HR Human Reliability methods 1 3 1 Clear and tr nsnar nt eru-,,rn#,

Analysis

... .. .... , , . -W. .,U , , IL" LIUI" Treatment of dependencies limited to cutsets >

Ix10 8; need to update offsite power recovery.

DE Dependencies C4 Excellent CCF treatment and internal flooding analysis CCF data source could be updated to latest INEEL CCF; addition of dependency matrices and review by plant staff would ensure PSA reflects as-built plant.

ST Structural 3 IGood documentation None

Response

QU Quantification C3 I Very good description of quantification process Results but summary Fisweak includes basicsummary information on insights; results should be updated L2 Containment C4 Excellent state of the art treatment of all severe ýNo credit for SAMGs; conservative treatmeto Performance accident phenomena relevant to PWRs with large dry interfacing ssystems LOCAs; so that results for containments; clean Level1/Level 2 interface. LERF are conservative 0 MU Ih Maintenance Update N/A This element was not evaluated in this review This element was not evaluated in this review 0 _____________ I NIA I

-A *C indicates the grade is conditional on resolving specific issues noted in the evaluation summaries in Section 2.

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KfINV Engineening and Research, Inc.

Shearon Hams PSA PeerReview SECTION 4 REFERENCES

[1] Nuclear Energy Institute, "Probabilistic Risk Assessment Peer Review Process Guidance", NEI-002, Draft, 2000.

[2] American Society of Mechanical Engineers, "A Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications(Draft)",

Revision 12, May 30, 2000.

[3] INEEL, "Evaluation of Loss of Offsite Power Events at Nuclear Power Plants 1980-1996", NUREG/CR-5496, November 1998.

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I Attachment C

. Summary of Reviews to Shearon Harris Nuclear Plant PSA, IPE and IPEEE The Independent Peer Review team concluded that the Shearon Engineering Harris PRA was one of the best-documented PRAs that the and reviewers had seen. The systems analysis, thermal hydraulics Research, analysis, containment performance analysis and the dependency Inc.

analysis were especially well done and were evaluated as grade Level 4 or close to this grade. Grade Level 4 is acceptable for use as a primary basis for developing licensing positions that may change hardware, procedures, requirements, or methods.

The data analysis was assessed at grade Level 2. Grade 2 corresponds to the attributes needed for risk ranking of systems, structures, and components.

The remaining elements of the PRA were at or near grade Level 3.

Grade 3 means that the PRA is adequate to support regulatory applications, when combined with deterministic insights.

The ISLOCA analysis was considered quite conservative compared with other similar plants. The ISLOCA initiating event frequency could be reassessed to make the PRA more realistic.

Jan-00 NN-RC Staff*s Evaluation of the Shearon Harris Nuclear Plant, Unit 1, NRC Individual Plant Examination of External Events (IPEEE) Submittal (TAC No. M83627)

The NRC staff performed a screening review for completeness and reasonableness considering the design and operation of the plant.

A Senior Review Board provided further review.

Apr-98 Shearon Harris Nuclear Plant Probabilistic Safety Assessment Review of Sequence Solutions, Report RSC 98-06, Revision 0 Ricky Summitt The high-level review of the Harris PSA accident sequence results Consulting (CAFTA-generated cut sets) from the model of record assessed the plant system design and the PSA event trees to determine at a

qualitative level what results could be reasonably expected. The review also identified the most important accident sequence contributors and determined their applicability based on expected plant response and general PSA modeling guidance.

00.1241

Jan-96 NRC Staffs Evaluation of the Shearon Harris Nuclear Plant NRC Individual Plant Examination (IPE Submittal) (Serial HNP-93-835)

(TAC No. M74418)

The evaluation package consisted of the Staff Evaluation Report (SER), the contractor's Technical Evaluation Reports (TERs) and a

summary of the IPE submittal on Internal Events.

Oct-93 Appendix K, Review Comment Resolution:

1. Initiating Events and Event Sequence Development

1. SAROS

2. System Modeling

2. SAROS

3. Component Failure Data

3. CP&L

4. Human Reliability Analysis

4. CP&L

5. Sequence Quantification

5. CP&L

6. Documentation

6. CP&L

7. System Models

7. CP&L

8. Comments from INPO Team member

8. INPO

9. Cutset Review Meeting

9. SAIC, SAROS, Appendix K is part of the supporting documentation prepared for NUS, INPO, the IPE submittal. The review team included CP&L members as CP&L well as members from the organizations listed.

001242

Attachment D Plant-Specific Information Provided to ERIN 001243

25. RMA-028, "Review of WOG Severe Accident Management Guidelines as Harris", Rev0, July 1996. Applied to Shearon

26. RMA-032, "Development of Shearon Harris Severe Accident Calculational 1996. Aids", RevO, July

27. Susquehanna Steam Electric Station, Units 1 and 2, Draft Safety Evaluation Fuel Pool Cooling Issues (TAC. No. M85337), October 25, 1994. Regarding Spent

28. SECY-99-168, Improving Decommissioning Regulations for Nuclear Power

21. 1999. Plants, December

29. AOP-013, Rev11, Fuel Handling Accident

30. OP-1 12, Revl3, Containment Spray System

31. OP-1 16, Rev17, Fuel Pool Cooling and Cleanup

32. OP-143.3, Rev19, Demineralized Water

33. HNP spent fuel pool drawings: CPL-2165-S-0805Rev7 and CPL-2165-S-0807Rev4.

34. HNP IPE Appendix J, Attachment 3 (hardcopy of material not in APP_J.ZIP)

35. HNPsum2000.xls, PSA summary of release categories

36. NRC letter to David Lochbaum, Spent Fuel Pool Cooling Generic Review (TAC including 2 attachments: No. M88094)

1. AEOD/S96-02, Assessment of Spent Fuel Cooling, September 1996.

2. 1INEL-96/0334, Loss of Spent Fuel Pool Cooling PRA: Model and Results, September 1996.

3. The Potential for Propagation of the Self Sustaining Zirconium Oxidation Water in a Spent Fuel Storage Pool, Pisano, et al., Draft - Jan 1984 Following Loss of

37. SF-0040, Rev0, Spent Fuel Pools C and D Activation Project Thermal Hydraulic Analysis, November 10, 1998.

38. NRC Report 7590-01-P, Draft Final Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants (Draft for Comment), February 2000.

39. Shearon Harrs Nuclear Power Plant, Unit 1, Individual Plant Examinatio-n for External Events (IPEEE) Submittal, June 1995.

40. NRC's Staff Evaluation of the Shearon Harris Nuclear Power Plant, Unit 1, Individual Plant Examination for External Events (IPEEE) Submittal (TAC No. M83627),

January 14, 2000.

41. IPE and supporting documentation

42. Drawings: CAR-2165-G-011 through 021; CAR-2165-G-151 through G-156 and CPL-2165 S1308, G-808, S-1324, -G-824, S-1300, -G-800

43. HNP Periodic System Review, Spent Fuel Pool Cooling (7110), June 20, 2000.

44. HNP AMMS printout for System 7110 (Spent Fuel Pool Cooling) for 1993 through 1998.

Various plant drawings - see attached table for complete listing.

46. HNP MAAP runs/card image input decks

47. HNP Procedure OMP-003, Rev12, Outage Shutdown Risk Management

48. HNP Procedure OMP-004, Rev8, Control of Plant Activities During Reduced Inventory Conditions

49. HNP Technical Specification 3/4.9.4 and Bases

50. Assumed duration that containment is not isolated during refueling outage.

51. Response to ERIN request: JPMs; Demin Flow; TSC guidance; Containment assumption; HVAC characteristics isolation

52. Response to request for additional information

53. Fire cutsets Non-fire cutsets

,m, 1u-e pool cooing pump unavailability. Basis for skimmer and purification pump our n'4r values.

0o01 44

55. HNP :Steam Generator and IPEEE assumptions

56. Power Source Locations

57. IN 2000-13, NRC Information Notice, "Review of Refueling Outage Risk", September 27,

58. Dose Calculation

59. Source Term (not provided to ERIN; however, implicit in the dose calculation)

60. Clarification: IPEEE/Tier 2 information Fire Area 1-A COMB

_61. Availability of Fire Pumper Truck following seismic event

62. Assumptions for RAB adverse environment

63. System

Description:

SD-145, Rev5, Component Cooling Water System

64. System

Description:

SD-1156, R8,Pant Electrical Distribution System

65. Operatin Procedure: OP-145, Rev26, C-om-ponent Cooling Water

66. System

Description:

SD-1 39, Rev12, Service Water System

67. CCW pump motors and pumps, EQ classification

68. Power supply for Normal Service Water pumps

69. Power Supply Question - FHB Crane

70. LER 89-002-00 Spent Fuel Pool Draining

71. Control Room Habitability - NEI Summary

72. Extract from October 6, 2000 ACRS-Commissioners meeting

73. FSAR 2.3.2 and wind rose data

74. EOP-PP-01 3, Rev5, LOCA Outside Containment

75. EOP-PP-012, Rev12, Loss of Emergency Coolant Recirculation

76. Fire Brigade entry into FHB in a 19OF environment

77. PLP-201, Rev39, Emergency Plan

78. Use of SFP as makeup source for core not identified in response guidance

79. Dose calculations for 4 cases: Early and Late Containment Failure, Containment Isolation Failure and ISLOCA

80. Reference for "Human Tolerance for Heat"

81. Location of Water Treatment Building

82. SGTR dose cases

83. Revised Time-to-Boil Calculations

84. Procedure FPP-013, Rev28, Fire Protection - Minimum Requirements and Mitigating Actions and recent history of Holly Springs FD backup

85. Location of Water Treatment Building and other site structures

86. Revised access times

87. Revised access times - Chi/Q based on 10-year data summary 001245

L0 001246

EXHIBIT 4 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of )

)

CAROLINA POWER & LIGHT ) Docket No. 50-400-LA

-COMPANY )

(Shearon Harris Nuclear Power Plant) ) ASLBP No. 99-762-02-LA AFFIDAVIT OF IL STEVEN EDWARDS COUNTY OF WAKE )

) ss:

STATE OF NORTH CAROLINA )

I, Robert Steven Edwards, being sworn, do on oath depose and say:

1. I am a resident of the State of North Carolina. I am employed by Carolina Power

& Light Company ("CP&L") and work at the Harris Nuclear Plant ("HNP" or "Harris Plant" or "Harris") in the Nuclear Engineering Department. Presently, I am the Supervisor, Spent Fuel Pool Project, and am responsible for commissioning and placing into service Harris spent fuel pools C and D. My business address is 5413 Shearon Harris Road, New Hill, North Carolina 27562 0165.

2. 1 was graduated from North Carolina State University in 1982 with a B.S. in Industrial Engineering. Since graduation, I have been employed by CP&L, first

-1 oo0024

IL as an Associate Engineer, then Engineer, at the Robinson Nuclear Plant, responsible for planning, scheduling and execution of outages and major projects.

Beginning in 1986, I served in the Technical Support Unit at Robinson as a System Engineer - Mechanical Systems. Promoted to Senior Engineer in July 1988, 1supervised a staff of contract engineers responsible for specific projects at Robinson. In June 1991, I assumed the position of Project Engineer Mechanical Systems at Robinson and managed a staff of four system engineers and two component engineers responsible for the operation, performance, reliability and maintenance of various plant systems. In August 1992, 1 became the Director - Information Architecture (Nuclear) in CP&L's Corporate Management Services and served as the management-level liaison and project manager for nuclear-related information technology projects at CP&L's nuclear plants. In October 1994, I moved to the position of Director - Project Control in the Corporate Nuclear Business Operations Group. In that position, I facilitated the development of long-range planning at each CP&L nuclear plant and provided oversight and administration of project management and economic evaluation processes and activities. In July 1996, I moved to Corporate Nuclear Engineering and became Manager of Projects, responsible for scope, cost, schedule, and quality of various nuclear projects. In April 1998, I was assigned to the Harris Major Projects Section and became responsible for the spent fuel pool C and D activation projects, including the completion of the spent fuel pool

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.001248

cooling and cleanup system ("SFPCCS"), spent fuel storage rack design and installation, and related activities. My resume is provided as Attachment A to this affidavit.

3. The purpose of this affidavit is to set forth the data and calculations on which CP&L relies in establishing the time to heat up the Harris spent fuel pools to boiling, and after boiling has started, the additional time necessary to then boil the coolant level down to the top of the spent fuel racks. Fi I summarize the background of the license amendment request and the information submitted in support of the application. Second, I describe the Harris spent fuel pool physical arrangement and associated equipment. Third, I discuss the types of heatup calculations used and their applicability to the Harris spent fuel pools. Fourth, I discuss the data and assumptions used in calculations. Finally, I describe the results of the calculations.

BACKGROUND

4. CP&L's application for a license amendment to place spent fuel pools C and D in service was submitted on December 23, 1998. As the project manager for the Harris spent fuel pool C and D activation projects, I was responsible for development of the factual information set forth in the license amendment request. The information in the license amendment request, as updated by additional information subsequently submitted, is accurate to the best of my knowledge and belief.

00-1249

IL

5. The license amendment request and the need to expand spent fuel storage at Harris results from the failure of the U.S. Department of Energy ("DOE") to begin taking delivery of spent fuel in 1998, as required by the contract between DOE and CP&L and by the Nuclear Waste Policy Act of 1982, as amended.

CP&L originally requested that the license amendment to allow placement of spent fuel in spent fuel pools C and D be issued no later than December 31, 1999, as CP&L had planned to begin loading spent fuel in pool C starting in 2000.

Further delays threaten to adversely impact CP&L's ability to maintain adequate spent fuel storage capacity and, with the loss of full core discharge capability at one or more of CP&L's nuclear plants, could lead to a forced shutdown condition.

6. The NRC Staff reviewers requested additional information regarding the license amendment request by letters dated March 24, 1999, April 29, 1999, June 16, 1999, August 5, 1999, September 20, 1999, and by conference calls on March 30, 2000 and April 4, 2000. CP&L responded to each request for additional information ("RArF) respectively on April 30, 1999, June 14, 1999, July 23, 1999, September 3, 1999, and October 29, 1999, and April 14, 2000. CP&L also provided additional information to the NRC Staff on October 15, 1999 and July 19, 2000, to supplement previous responses. As the project manager for the Harris spent fuel pool C and D activation projects, I was responsible for development of the factual information set forth in the responses to the NRC 00 25

Staff. Following the Board's Memorandum and Order dated August 7, 2000, 1 was also responsible for development of factual information responsive to the NRC and BCOC discovery requests. The information in each of those responses, as supplemented, remains accurate to the best of my knowledge and belief.

7. 1have previously given an affidavit in this matter on December 30, 1999. The information in that affidavit remains accurate to the best of my knowledge and belief.

8. Harris was originally planned as a four nuclear unit site (Harris 1, 2, 3 and 4).

Harris 3 and 4 were canceled in late 1981. Harris 2 was canceled in late 1983.

Spent fuel pools A, B, C and D and the spent fuel pool cooling and cleanup system ("SFPCCS") for spent fuel pools A and B were completed as part of the Harris fuel handling building, are described in the Harris Final Safety Analysis Report ("FSAR"), and are licensed as part of Harris.

9. Construction on the SFPCCS for spent fuel pools C and D was discontinued after Harris 2 was canceled. By that time, concrete had been poured, all four spent fuel pools had been constructed, and the SFPCCS piping immediately outside and under the spent fuel pools was installed, welded in place and embedded in reinforced concrete. The SFPCCS for spent fuel pools A and B was completed and placed in service. Harris I began commercial operations in 1987. Sometime in late 1988 or 1989, before the first discharge of spent fuel and refueling of Harris 1, spent fuel pool A was filled with borated water. Spent fuel pool B was

-5 00.t251

filled with borated water on or about the time spent fuel was first discharged from the Harris reactor. Because spent fuel pools C and D are connected to spent fuel pools A and B by transfer canals, at some point in or after 1989, spent fuel pools C and D were also filled with borated water to allow the gates in the transfer canal to be opened without a loss of water and preclude an inadvertent partial drain-down of spent fuel pools A and B to spent fuel pools C and D.

HARRIS PLANT SPENT FUEL POOLS AND ASSOCIATED SYSTEMS AND EQUIPMENT

10. As the project manager for the activation of spent fuel pools C and D, my work encompasses analytical design and engineering evaluations, management of the hands-on physical implementation of the modifications to the SFPCCS, and inspection and preparation of the spent fuel pools themselves. As a consequence of my extensive work at Harris and with the Harris spent fuel pools, I am familiar with the physical layout, system configurations, equipment installations, operations, and operating procedures for Harris, as they relate to normal and alternate operation of the fuel handling building, the spent fuel pools, and associated support systems and equipment.

11. As Harris was originally envisioned as a four unit facility with a shared fuel handling building, the fuel handling building was designed and constructed with four separate pools capable of storing spent nuclear fuel. Spent fuel pools A and B were originally intended to support Harris Units 1 and 4. Spent fuel pools C 001252

and D were originally intended to support Harris Units 2 and 3. In addition, the fuel handling building contains a cask unloading pool, which can be connected to any spent fuel pool through transfer canals.

12. The layout of the Harris fuel handling building is illustrated and described in detail in the Harris FSAR, sections 3.8.4.1.3 and 9.1. Each spent fuel pool and the cask unloading pool are interconnected by a main transfer canal, oriented in a north-south direction, and two fuel transfer canals, oriented east-west. The spent fuel pools and transfer canals contain sufficient amounts of water to facilitate safe fuel handling and storage activities. The spent fuel pools, transfer canals, and cask unloading pool contain openings for the underwater movement of fuel assemblies between the pools and transfer canals. These openings also allow the communication of water between the pools and transfer canals. Removable bulkhead gates are installed in the openings when there is a need to isolate a particular pool or canal from the others. The isolation function of the bulkhead gates is provided by stainless steel structural components and inflatable seals, which are installed around the sides of the gates that fit into slots in the pool and canal openings. The seals are normally inflated using instrument air supplied at the installed gate location.

13. The normal configuration of the spent fuel pools (i.e., the configuration present 99% of the time an on annual basis) is with open communication (i.e., the gates removed) between spent fuel pools A and B and the interconnecting south ("Unit 001253

L 1/4") transfer canal. Plant Operating Procedure OP-1 16, section 8.27 requires that a "clear path" be maintained between pools A and B. The cask unloading pool is normally open with the gate removed between it and the north transfer canal. Spent fuel pools C and D are currently normally isolated from the main transfer canal and spent fuel pools A and B. This alignment is illustrated in Attachment B. The expected configuration of the spent fuel pools (ie., the configuration expected to be present 99% of the time on an annual basis) following approval of the pending license amendment request is with open communication between spent fuel pools A and B, the connecting transfer canal and the main transfer canal. The cask unloading pool will normally be connected to spent fuel pools C and D through their interconnecting north ("Unit 2/3") fuel transfer canal. Spent fuel pools C and D, and the cask unloading pool will be isolated from pools A and B by a gate installed at the cask unloading pool end of the main fuel transfer canal. This alignment is illustrated in Attachment C.

14. The original Harris design included a SFPCCS to service spent fuel pools A and B, and a separate SFPCCS to service spent fuel pools C and D. The SFPCCS for spent fuel pools A and B is in service. The SFPCCS for spent fuel pools C and D was not completed, but will be finished and placed in service to support spent fuel operations pursuant to the pending license amendment request.

00125 4

SPENT FUEL POOL HEATUP CALCULATIONS

15. 1 directed that calculations be performed to determine two values for the Harris spent fuel pools: (1) the time to heat up the individual pools (A, B, C, D) to boiling temperature (i.e., 212 degrees Fahrenheit) and (2) the additional time to boil the coolant level down to the top of the spent fuel racks. In addition, I directed calculation of the amount of water required to offset the boiling rate in gallons per minute ("gpm') for each case. In turn, I used these calculations to perform a "best estimate" analysis, meaning that assumed input values are based on normally expected operating conditions based on historical data and plant operating records.

16. As a first step, calculations using standard, commonly used heat transfer equations were performed. The individual calculation steps are described in Attachment D. The heatup and boiloff calculations were performed by Andrew Howe, a degreed nuclear engineer and civil engineer in the Harris Engineering Support Section with 18 years experience performing these types of calculations.

Mr. Howe has been previously licensed as a Senior Reactor Operator and is currently assigned as the Harris spent fuel pool cooling and cleanup system engineer. The methodology and inputs were independently reviewed by Tom Scattergood, a second qualified engineer in the Harris Engineering Support Section. Mr. Scattergood has Bachelors and Masters degrees in mechanical engineering and over eight years experience performing this type of work.

001255

L.

17. 1 then used these heatup and boiloff calculations to prepare an analysis responsive to the Board's request for a "best estimate" analysis by revising the performance assumptions to reflect normally expected operating conditions based on historical data and plant operating records. My analyses resulted in three values for each scenario: (1) time to heat the pools to boiling temperature, (2) additional time to boil the water down to the top of the racks, and (3) makeup flowrate required to offset boiling. The calculations I performed were independently reviewed by Edison Morales, a licensed professional engineer and Harris mechanical engineer with over 29 years of relevant experience.

18. The final calculations were performed using a Microsoft Excel spreadsheet.

Results were independently verified using manual techniques and a hand calculator.

DATA AND ASSUMPTIONS

19. The input values and initial conditions for the heatup calculations were obtained from several sources. A complete list of calculation input values and sources is provided as Attachment E.

20. Initial condition assumptions were based on current knowledge, existing license and administrative controls, and professional judgments on expected future operating conditions. In each case where a future condition could not be definitively established, a best estimate assumption of that condition was used.

Key assumptions in this category include: (a) Beginning of Cycle heat load for 00125G

spent fuel pools A and B is assumed to be a base heat load from HNP Calculation SF-40, Spent Fuel Pools C and D Activation Project Thermal Hydraulic Analysis, dated November 10, 1998, which includes a one-third freshly discharged core after startup from a refueling outage; (b) spent fuel pool temperature at the initiation of the heatup is the temperature expected based on Harris historical operating records; (c) spent fuel pools C and D heat load is either 1 MBTU/hr (the maximum allowed by the pending license amendment request) or 15.6 MBTU/hr (the maximum calculated heat load with both pools C and D completely filled); and (d) the water level in the spent fuel pool and transfer canals is the normal pool level expected during plant operation. In addition, several conservative assumptions are incorporated in the heatup calculations, including: (a) water volume in the cask unloading pool was not considered (ie.,

gate 8 is assumed installed); (b) no credit is taken for heat transfer to the pool liners, concrete structure, or atmosphere; (c) no credit is taken for any makeup water addition after the initiation of the heatup.

21. Initial operating conditions for the spent fuel pools and transfer canals were established as the lineup expected during normal operation of the plant (i.e., the configuration expected 99% of the time on an annual basis). In all scenarios, spent fuel pools A and B are interconnected through their transfer canal and connected to the main transfer canal (i.e., gate 2 is installed and gates 1, 3 and 4 are removed). For the I MBTU/hr scenario involving spent fuel pool C, it is

-11 00125 7

L interconnected with its transfer canal and isolated from the main transfer canal (ie., gate 7 removed and gates 5 and 6 installed). Spent fuel pool D is not considered (ie., gate 9 is installed) in the 1 MBTU/hr scenarios, as all the racks and fuel will be in spent fuel pool C. Gate 8 is also assumed to be installed, as the inventory of the cask unloading pool is conservatively not credited in the calculations. In the 15.6 MBTU/hr scenario, racks and fuel will be installed in spent fuel pools C and D and the pools will be interconnected through their transfer canal, but isolated from the main transfer canal (i.e., gates 7 and 9 are removed and 5 and 6 are installed) and the cask unloading pool (i.e., gate 8 installed to reflect the conservative assumption not to credit this water volume).

CALCULATION RESULTS

22. A table listing each heatup calculation and the results for each analyzed scenario is included as Attachment F. These calculations determined that if all cooling and makeup are lost to spent fuel pools A and B at the beginning of an operating cycle, it would take approximately 20.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months for the pools to heat up to boiling temperature. Once boiling begins, it would take an additional 7.2 days for the water in the pools to boil down to the top of the racks where uncovery of fuel could begin. During this period, approximately 54 gallons per minute of makeup water would needed to maintain water level in the pools constant. If all cooling and makeup is lost near the end of an operating cycle (when the heat load from the most recently discharged fuel has diminished significantly), it would take 001258

approximately 38 hours4.398148e-4 days 0.0106 hours 6.283069e-5 weeks 1.4459e-5 months for spent fuel pools A and B to heat up to boiling. Once boiling begins during this scenario, it would take an additional 13.5 days for the water in spent fuel pools A and B to boil down to the top of the racks.

Approximately 29 gallons per minute of makeup water would need to be added to the pools to maintain water level in the pools constant. In the 1 MBTU/hr scenario (involving only spent fue! pool C), it would take 16 days for water in spent fuel pool C to reach boiling temperature if all cooling and makeup water is lost. Once boiling in spent fuel pool C begins in this scenario, boiling would have to continue uninterrupted for an additional 99 days for the pool C water level to boil down to the top of the racks. During this period only slightly more than two gallons per minute of makeup water would be necessary to offset the boiling rate. For the maximum calculated end of life heat loads in spent fuel pools C and D, the pools will heat up to boiling temperature in approximately 34 hours3.935185e-4 days 0.00944 hours 5.621693e-5 weeks 1.2937e-5 months following loss of all cooling and makeup. Once boiling begins in the pools in this scenario, an additional 8.8 days without cooling or makeup is required for the water in spent fuel pools C and D to boil down to the top of the racks. During this period, approximately 34 gallons per minute of makeup water would needed to maintain water level in the pools constant. These calculations are in Attachment G.

23. The calculations performed to determine the best estimate time to boil, additional time to boil to the top of the racks, and makeup required to offset the boiloff rate 001259

I under different heat load scenarios were all performed using standard heat transfer techniques. These techniques are well understood and straightforward.

These calculations used the same approach that has been historically employed at the Harris Plant to perform similar calculations. These calculations were performed by experienced and qualified engineers knowledgeable in both heat transfer and Harris plant design and operation. In addition, the inputs, methodologies and results were independently reviewed by experienced and qualified engineers.

24. As an additional check, the results of these calculations were compared for consistency with the spent fuel pool heat up rates identified in FSAR 9.1.3.3.

(Amendment No. 50). The information contained in this section of the FSAR discusses expected heat loads and heat up rates through operating cycle 10. The FSAR identifies that under these conditions (which includes an assumed heat load in spent fuel pools A and B of 16.84 MBTU/hr) spent fuel pools A and B would heat up from I 12.7"F to 137*F in 5.56 hours6.481481e-4 days 0.0156 hours 9.259259e-5 weeks 2.1308e-5 months , which equates to a heat up rate of 4.37*/hr. The best estimate time to boil calculations determined that spent fuel pools A and B would heat up from 95¶F to 212*F in 20.57 hours6.597222e-4 days 0.0158 hours 9.424603e-5 weeks 2.16885e-5 months for the 25 MBTU/hr beginning of cycle heat load scenario, which equates to a heat up rate in spent fuel pools A and B of 5.69°/hr. The spent fuel pool A and B end of cycle scenario calculations determined that spent fuel pools A and B would heat up from 95 0F to 212OF in 38.67 hours7.75463e-4 days 0.0186 hours 1.107804e-4 weeks 2.54935e-5 months , which equates to a heat up rate of 3.030 /hr.

-14 0012.60

Based on these results, the minimum and maximum heat up rates calculated as best estimates range from 3.03*/hr to 5.690/hr compared to a value based on the FSAR of 4.37*/hr. These values are consistent and the difference, to the extent they are significant at all, appear primarily because of different heat load assumptions (i.e., 25 MBTU/hr and 13.3 MBTU/hr heat loads used in the best estimate calculations assume spent fuel pools A and B are 'full' at the beginning and end of an operating cycle. The 16.84 MBTU/hr heat load identified in the FSAR is a bounding heat load calculated through operating cycle 10, which is the current Harris fuel cycle). This comparison provides me a high level of confidence that the best estimate analyses produced results that appropriately characterize the expected plant performance under the postulated conditions.

25. The results of these calculations show that in the highly unlikely event that all cooling and makeup to the spent fuel pools is lost, a considerable amount of time is available for Harris operators to re-establish cooling or makeup flow in order to prevent the spent fuel pool water level from boiling down to the point where fuel uncovery could occur. Even considering the worst case scenario where cooling and makeup is lost to 'full' spent fuel pools A and B at the beginning of an operating cycle, then the Harris Plant operators would have over 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months to take actions necessary to establish makeup or cooling flow before spent fuel pools A and B reached boiling temperatures. The plant operators would then have approximately an additional week to re-establish spent fuel pool cooling or make 001-261

I up before the water level boiled down to the top of the fuel racks. Under this scenario, the operators would need to provide less than 54 gallon per minute of make up water in order to offset the boiling rate in spent fuel pools A and B.

Since heat loads in spent fuel pools C and D are less than the corresponding heat loads in pools A and B, an even longer time is available to establish cooling or makeup flow to these pools.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on November 5, 2000.:&z R. Steven Edwards Subscrild and sworn to before me , %,

this day of Novembr o2000. .

JGOIARY b My Commission expires:

e :WBL1 001262

R. Steven Edwards Summary: Eighteen years experience in engineering, project management and outage management.

EXPERIENCE: Carolina Power & Light Company, June 1982 - Present Supervisor, Spent Fuel Pool Project, Harris Plant, Nuclear Engineering (April 1998 - Present)

Project manager for Harris spent fuel pool 'C' and

'D' activation projects including spent fuel pool cooling and cleanup system completion, spent fuel storage rack design and installation, pool cleanup, and related activities. Responsible for all aspects of scope, cost, schedule and quality of projects. Responsible for study, design and implementation activities. Supervise multi-disciplined modification engineering staff that includes mechanical, civil and electrical engineers that develop plant design change modifications, oversee architect/engineer designs, write procedures, perform 10CFR50.59 analyses, perform ANSI N45.2.11 design verification reviews, and perform owner reviews of NE developed modifications and calculations. Manage activities of various NE including Bechtel, Sargent & Lundy, Duke Engineering, engineers performing design activities Raytheon, Protopower and Holtec.

Responsible for development of License Amendment Request for SFP Activation project.

Provide technical support to spent fuel communications evaluations. team. Perform root cause Serve as Emergency Response Organization Company Technical Spokesperson.

Manager of Projects, Nuclear Engineering (July 1996- April 1998)

Project manager responsible for scope, cost, schedule and quality of various nuclear projects.

Responsible for A/E design and analysis. Managed outsource engineering activities (scope development, schedule & cost management, AE negotiations & interface) for preferred and specialty engineering AE's and contractors. Provided group-wide oversight and administration of project management and economic evaluation Responsible for three-phase project authorization processes, procedures and activities.

including value-added technical and financial review of projects requiring executive approval. Delivered economic evaluation module at NGG Business Concepts Course. Taught Project Cost Management module for Project Management Institute (PMI) project manager certification course. Developed and delivered various project management/ project controls presentations to industry groups such as Integrated Scheduling & Planning Utility Group (ISPUG) and Institute for International Research Budgeting and Forecasting Conference.

Director - Project Control, Nuclear Business Operations/

Operations & Environmental Support (October 1994 - July 1996)

Provided group-wide oversight and administration of project management and economic evaluation processes and activities. Lead development of NGG project management procedure. Responsible for three-phase project authorization. Developed and delivered project management and economic analysis training to plant personnel focusing on fundamentals and NGG specifics. Delivered presentations to industry groups and internal various project management related company management. Managed implementation of integrated project cost/schedule reporting system that combined FAIM financial data with Prestige schedule information.

Developed and delivered economic evaluation module of NGG Business Concepts Course. Managed project budgeting team that implemented process to use Prestige schedule and resource data to build budget for plant projects. Facilitated development of Long Range Planning process at each nuclear plant. Project management peer group facilitator.

-0 001263

F L

R. Steven Edwards Director - Information Architecture (Nuclear),

Management Services (August 1992 1994) - October Served as management-level liaison and project manager for nuclear related technology projects. Provided technical and business process perspective for information implemented nuclear I/T projects. Coordinated corporately the development of the nuclear portion Corporate Information Technology (I/T) of the process. Evaluated NGG generated Plan including administration of project prioritization requests evaluation of business justification, development for I/T products and services including of cost/benefit analyses and approval resource allocations. of I/S Project Engineer - Mechanical Systems, Technical Support, Robinson Plant (June 1991 - August 1992)

Managed staff of four system engineers and two component engineers responsible operation, performance, reliability and for maintenance of various plant NSSS, secondary mechanical systems and equipment support and SI/residual heat removal, containment such as high head safety injection, low spray, reactor coolant pumps, liquid & head disposal, steam generator blowdown, HVAC, gaseous waste make up water treatment, condensate etc. Provided extensive coaching and polishing, mentoring to staff with varied experience/education levels in development of their customer focused, performance oriented system component engineering skills. Served and as refueling outage Technical Support responsible for timely and successful completion Shift Manager through coordination of efforts with operations, of all engineering related outage activities site management as well as supervision maintenance, corporate engineering and of engineers assigned to emergent activitiesother planned projects. Served on Emergency and Response Organization as Accident Assessment Team - Mechanical Engineer and Emergency Communicator.

System Engineer - Mechanical Systems, Technical Support, Robinson Plant Senior Engineer (July 1988 - June 1991);

Engineer (November 1986 - July 1988)

Supervised staff of contract engineers responsible for specific projects including performance monitoring, procedure plant rewrite, backlog assessment, engineering program, and work management system training development (1990-1991).

System engineer responsible for operation, various mechanical systems including performance, reliability and maintenance all plant HVAC, containment vessel (civil of systems), LHSI/RHR, containment spray, and support post accident containment venting/H2 primary and post-accident sampling, recombiner, etc. (1986-1990). As system engineer, system/equipment performance; performed monitored evaluations, temporary plant modifications, surveillance tests; developed engineering N45.2.11 design verification reviews, procedures, 10CFR50.59 safety analyses, procurement engineering reviews, etc. ANSI oversight to maintenance staff in troubleshooting Provided root cause analyses. Served on Emergency system/equipment problems. Conducted Response Organization as Accident Assessment Team - Mechanical Engineer and Emergency Communicator.

Outage Planning and Scheduling Engineer, Outage Management, Robinson Plant Engineer (June 1984 - November 1986);

Associate Engineer (June 1982 - June 1984)

Responsible for planning, scheduling and execution of outages and major projects.

-2 001264

R. Steven Edwards Developed detail and summary level schedules for forced outages, refueling outages, generator replacement outage and normal operating periods using steam manual CPM and ARTEMIS project management system. Led plan-of-day meetings.

Served as field "coordinatorin outage management organization for major projects such as S/G eddy current.

PROFESSIONAL DEVELOPMENT: Attended American Management Association Project Management and Financial Analysis training, Reengineering Fundamentals Seminar, Harvard University In-Place Filter Testing Workshop, industry sponsored ANSI Filter Testing Workshop, and NCSU Fundamentals of HVAC Design. Participated N510 Fan and in company sponsored technical, project management and management/supervisory development training. Engineer in Training Certification - State of North Carolina.

EDUCATION: Bachelor of Science in Industrial Engineering, North Carolina State University, May 1982

.3-001265

I

- Attachment B Diagram Illustrating HNP Spent Fuel Storage Pools, Transfer Canals, and Current Bulkhead Gate Configuration.

II Bulkhead Gate Configuration 0- (normally closed) t (normally open) oo126i.

Attachment C Diagram illustrating anticipated bulkhead gate configuration in the HNP spent fuel pools subsequedt to operational use of C and D pools.

Fi Bulkhead Gate Configuration &* (normally closed) * (normally open)

The "normally open" configuration for gate 9 would apply subsequent to placing this pool in service that is scheduled for early the next decade. Otherwise, this gate would remain normally closed.

001267

Attachment D Description of the Key Steps in the Spent Fuel Pool Heatup Calculations The following steps were taken in the calculation of spent fuel heatup and boiling:

Water volume The amount of water available in the pools was determined by calculating the volume of the pools and then subtracting out the volume of the racks and the spent fuel. Since pools A and B are operated such that the pools remain interconnected and connected to the main transfer canal during all normal conditions, then the combined volume of the two pools, the main transfer canal and the Unit 1/4 transfer canal was considered together for these calculations. For the spent fuel pool A and B time to boil calculations, the water volume from the main transfer canal was not included; it was assumed that, due to its length and configuration, the temperature in the main transfer canal would lag behind the temperature in the pools during the heatup. However, the volume of water in the main transfer canal was included during the calculation of the additional time required to boil down to the top of the racks, since this water volume would definitely be available to displace water in the pools during the period when the boiling is actually occurring. Once both pools C and D are placed in service these pools will operate such that they will remain interconnected. Therefore, the combined volume of pools C and D and the Unit 2/3 transfer canal was considered together for these calculations.

Heatinputrequired to raise the temperature of this water volume to boiling.

In the highly unlikely event that all cooling and makeup to the pools is lost, the heat input into the water from the spent fuel will cause the water temperature to increase. The amount of heat input required to increase the temperature of this volume of water from the normal expected temperature to boiling temperature was calculated using standard heat transfer equations.

Time required to reach boiling.

The heat input required to reach boiling divided by the heat load present in the pools yields the time required to reach boiling temperature.

Best estimate heat loads were used to determine the time to boil under the specified scenarios.

Additional heatinputrequired to boil the water to the topof thespent fuel racks.

The volume of the pools that is above the top of the racks was determined. Using this volume, the heat input required to change this volume of water from a saturated liquid to 00..068

a saturated vapor was calculated using standard heat transfer equations.

Additional time to boil down to the top of the racks.

The heat input required to boil the water down to the top of the spent fuel racks was divided by the heat load in the pools to yield the additional time after the start of boiling required to reach the top of the racks. Best estimate heat loads were used to determine the time to boil off the water under the postulated scenarios.

Total time from initiation until the water has boiled down to the top of the racks.

The previously calculated time to boil plus the additional time to reach the top of the racks once boiling begins provides the total time available for the operators to reestablish either cooling or makeup to the pools in order to prevent uncovery of the fuel in the spent fuel pools.

Makeup flow rate required to offset boiling.

The rate (in gallons per minute) that the water is being boiled off was determined by dividing the previously calculated volume of water above the racks by the additional time required to boil the water down to the top of the racks. This boiloff rate is also the amount of makeup required to maintain pool level constant under boiling conditions.

001269

I Attachment E Data Sources for Input Values and Initial Conditions Nominal spent fuel pool level (284.5 ft) CAR-2165-G-024, System Description SD 116 (Fuel Pool Cooling and Cleanup System). FSAR and DBD Elevation of top of racks (260.08 ft) CAR-2165-G -024, System Description SD 116 (Fuel Pool Cooling and Cleanup System), FSAR and DBD Bottom of pools (246 ft) CAR-2165-G-024, System Description SD 116 (Fuel Pool Cooling and Cleanup System), FSAR and DBD Bottom of gates (260 ft) CAR-2165-G-024, FSAR and DBD Best estimate of spent fuel pool Derivred from hitrcal spent fIe pool temperature at initiation (95 0F) temperatures between August 1999 and September 2000 as recorded by the plant Emergency Response Facility Information System ("ERFIS") computer.

Pools A and B beginning of cycle heat load Core shuffle refueling alignment heat load (25.0 MBTU/hr) from calculation SF-0040, 'Spent Fuel Pools C and D Activation Project Thermal Hydraulic Analysis." This heat load is the future expected heat load when pools A and B would be essentially full. This number is expected to be conservative since the number calculated in ESR 00-00046 for cycle 10 is less than 16.84 MBTU/hr.

Pools A and B base heat load at the end of Obtained from calculation SF-0040, --Spent an operating cycle (13.3 MBTU/hr) Fuel Pools C and D Activation Project Thermal-Hydraulic Analysis."

Pools C and D initial licensed limit heat Technical Specification limit in the load (1.0 MBTU/hr) pending License Amendment Request (HNP-98-188).

0012"

Pools C and D maximum end of life heat Action Item 2.

ESR 95-00442 95-00442 Action Item -2.

load (15.6 MBTU/hr)

Formulation Specific Volume of water at 950 F IAPWS 97 IAPWS 97 Formulaionii (0.016115 cu ft/lb)

Specific Volume of water at 212 0 F IAPWS 97 Formulation (0.0 16714554 cu ft/lb)

Enthalpy of water at 95°F (63.0459 IAPWS 97 Formulation BTU/Ib)

Enthalpy of water at 212°F under saturated IAPWS 97 Formulation water conditions (180.1802 BTU/Ib)

Enthalpy of water at 212°F under saturated IAPWS 97 Formulation vapor conditions (1,150.2889 BTU/Ib)

IAPWS 97 Formulation Density of water (7.48 gal/cu ft) standard conversion formula Pool A rack layout Drawing CAR-2168-G-0124 SO I (Fuel Rack Arrangement Harris Fuel Pool "A")

Pool B rack layout Drawing CAR-2168-G-0 116 SO I (Fuel Rack Arrangement Harris Fuel Pool "B")

Pool A rack weights Harris Plant procedure MMM-020 Pool B rack weights Harris Plant procedure MMM-020 Pool C rack layout and weights Holtec drawing 1994 Pool D rack layout and weights Holtec drawing 1993 Specific weight of a spent fuel rack (0.29 Mark's Standard Handbook (stainless steel) lb/cu in)

Volume of a BWR fuel assembly (1.164 cu Engineering Service Request 95-00584 ft) 001271

I Volume of a PWR fuel assembly (2.63 cu fi)

Pool A dimensions - (38 ft x 13 ft)

IEngineerin Engineering Service Request 95-00584 Service Request 95-00584 Drawing CAR-2165-G-022 Pool B dimensions - (50 ft x 27 ft) Drawing CAR-2165-G-022 Pool C dimensions - (50 ft x 27 ft) Drawing CAR-2165-G-022 I

Pool D dimensions - (32 ft x 20 ft) Drawing CAR-2165-G-022 Unit 1/4 transfer canal dimensions (38 ft x Drawing CAR-2165-G-022 9 ft)

Unit 2/3 transfer canal dimensions (38 ft x Drawing CAR-2165-G-022 9 ft)

Main transfer canal (288 ft x 3 ft) Drawing CAR-2 165-G-022 001272

Attachment F Summary Results of Heatup Calculations for Analyzed Scenarios Additional time Time to reach for water level Makeup boiling to reach top of required to s temperature racks Total time offset boiling 20.57 hours6.597222e-4 days 0.0158 hours 9.424603e-5 weeks 2.16885e-5 months 7.21 days 8.07 days 53.70 gpm ofZ

.nd of 38.67 hours7.75463e-4 days 0.0186 hours 1.107804e-4 weeks 2.54935e-5 months 13.56 days 15.17 days 28

.57 gpm 384.66 hours7.638889e-4 days 0.0183 hours 1.09127e-4 weeks 2.5113e-5 months 99.99 days 116.02 days 2.15 gpm heat 5.6 34.42 hours4.861111e-4 days 0.0117 hours 6.944444e-5 weeks 1.5981e-5 months 8.80 days 10.23 days 33.64 gpm ieat 001273

A B C I

D I E F iSupportinq Analysis for NRC Specific Interrogatory #7 Request Calculate the best estimate timeto reach boiing temperature and then additional time for water level to b ole dow ton the top of the racs 4 AJso Calculated is the amount of makeup water required to offset the boeloff rate (in gpm) 6Assumptions:

7 8 1 1 Gates 1.3 & 4 are removed. Gate 2 is installed.

9 Thus SFP A. SFP B. 1/4 transfer canal and main tunrsfer canal we interconnected This is normal lineup.

10 2 Gate 7 is removed and gates 5 & 6 are installed for SFP C 1 MBTUisr scenano. Thus SFP C and 2/3 trasfer canal a interconnected for cooling 11 SFP D is not onsidered for 1MSTU/br scenari since Oilracks and fuelwill be in SFP C 12 3 Gates 7 & 9 are removed and gates 5 & 6 are installed for SFP C&D 15.6MBTU/fr scenano 13 _ Thus SFP C. SFP D and 2/3 transfer canal will be interconnected for cooling. This will be thle nomtal hneup when SFP C&D are acivated 14 4 Water volume in main transfer canal and cmlk unloading pool a *ot considered in this analyss. is...

1s - 5 No credit is taken for any SFP cooling or makeup after initiation of ale loss of cooling 16 occurs.

6 Beginning Cycle heat load for SFP A&B assumes base heat load from SF-40 plus 1/3 freshly dislchaged core after st*tupifron a re-fueing outage 17 7 No credit is taken for heat transfer to liner. concrete structure or atmosphere.

20 Inputs:

22 Nominal water level 284.50 ft.

23 1 Top of Racks 2W.08 -ft.

24 Bottom of Pools 246.00 it.

25 BBottomof Gates 260.00 ft.

26 Best estimate - SFP temp at initiation 95 deg. F

- BP o ol A& B he at loa d 25 ,000.0 0 0 B T U /hr " ,,* * , g f y d

_9Pool A&B heat load 13.300.000 -BTU/hr Base heat load (end of Zy e) 30Pool C&D heat load 1.000.000 .BTUfhr Initial Licensed Limit

_Lý Pool C&D heat load 15 661.901 BTU/hr End of Lileheat oad 33Water Poebs 37Specific Volume__ 95F (initaton temertre) 0.0116115214, cu ff~b 35Specific Volume _.2_12F (saturatied liquid) 0.016714554 cu ft/I

_MEuuyR9F(ntaintrieg" 63.0459 BTU/tb 37 tnhaPY42 212F (saturated liquidl) 180.1802 :BTUbo 3 __ Enthalpy @ 212F (saturated vapor) 1.150.2889 BTUIIb 397.48 ,a/au f I

141 Rack & fuel volume:

42POOA 1.675.85 :Cuff 43PoolB 5.627.63 c'ulf Pool C 6.449.04 ,cuft 45 Pool D 3.007.77 cu ft 46 47 Pool A water volume - total 17.343.15 cuIt 38ft x 3Nt x (284.5 - 246)ft minus reckifuel volume Pool A water volume - to top of racks 12.063,48 cu ft 38ft x 3f a (284.5- 260.06)ft 49 Pool B water voluime-total 46.347.37au ft 50ft x 27ft x (284.5 - 240)f menus ra:jck/fuel volume 51 Pool B water volume - to top of nais 52 - C45.525.96 32.967.00 cu ft 5Oft x 27ft x (284.5 - 260.08)ft

cu ft 50ft x 27 x (284.5 - 246)ft minusr raudh volume 5 Pool C water volume - to top of raes 32.967.00 ;cuft 50ft x 27ft x (284.5 - 260.08)fA 53 Pool D water volume - total 21.632.23 !cu 321f x 20/t x (284.5 - 246)fl minus rack/fuel volume 54 Pool D water volume - to top of racks 15.628.0 cu/f 32f x20ftx (284.5.260.08)ft 55 1/4 transfer water volume to bottom of gates U.379.00ccui 38Mx 911x (284.5- 260)f 56523 tansfr water volume to bottom of gates 8.379.00 cu f 5n_

__ , ,transfer.canal, 38ftxgfx*2-k.5- MM01/ ..

w ow ,vnolue 21.168. o cu ft8/faft(284 58 .-2 ) -

- 288_ x--1-x (8-.

59 60 6 &- a __M of te cycle__., . ..... .. .

3 _ Time to boil for A&B - beginning of cycle:

Volume avalalefor__ s 72.069.52 cu ft total volume of SFP A. SFP B & 114 canal 65 sass M -~_

l 66 'Heat .nput rquired

_4.390.93.25 _total volume)Ipcii volume -

514.277.884.92 BTU mass x (enthalpy initiation- enthalpy Q saturated liquid) 67 _ Time required to boiling 20.87 hr heat input I heat load at beginning of cycle . ..

S Additional time to boll down to top of ac s-begn n 'ing -cle: -. .cy.. - ... -. - - - - - - - - - - . .

__ Volume available to top of ,548 cu ft.__ volume to top of racksr o- SFP-A. SFP*B. main canal .i,4 cana 7Mass 4.481.828.86 volumetotopoFrecsIspecicvolume 72.-lHeat inprut _.m(i~ld to r _ach .opracks

_o 4.328.459.070.68 BTU mass x (ent*fapy@ sra ted vapor - enthalpy saturated lIqll )

__-73 Time

. to reach top of racks 173.14 hr heat input Iheal load 75 74 - ~~or7.1dy 76 Total ties from initiation until top of rackM: , -.

77 Tim to bol* + t.me to top of racks 193.71 'hr 78 A&Y -.-

80 __Malceup requiredto oftset bolting: - -- - _ _--

81 Voluw met___volum to top ofracks for SFP A. SFPB. maincana& 1/4canal 2 557839.55 galons convert ci 1110gallons (x by-7-.48) 83 _ Time tobad ltotop ofnicks -

173.14 hr fromabo ,e . . .. . .

4 5ortrue83.70_galmin , volumeto top ofracks/time to ofracks 0fterboilg 001274

A B C D IE F 86 87 88 Pools" -atthe-nd ofte cycle 89 __

90 Tmie to Boil for A&B. lEnd of cycle:

91 - Heat input required to reach boiling 514.277.884.92 BTU calculated above 92 - Time reqirad to boiling 38.67 hr heat input / heat load at end of cycle 94 Additional time to boil down, to top of racks - End of cycle:

95 1Heat input required to reach top of racks 4.328.459.070.68 BTU calculated above 96 - ITime to reach top of racks 325.45 hr heat input heal load at end of cyle 97 or 13.16 days OR wi*TUIhr I

Total time from initiation until too of rcu ofks:

'T;mi [*^ km;I * (,;m. *^**-- ^ll -- *1 load i or 15.17 days heat 103 ._ _Makeup required to offset boiling:

[104 Volume to top licensed of racks 74.577.48 cu f volume to top of racks for SFP A. SFP B & 11 105 V-Volume to lop of racks initial 557.839.55 gallons conven cu ft to gallons (x by 7.48) O4 canal 106 Time theto boil to top of racks 325.45 hr from above 107 It CMake up flow rate required 28.57 oia/rmin volume to top of racks. I t.., ...... f. 1 1,01 10 bderbling begins Pool

t

[j ýlC at the initial licensed hest load of 1 MBTU~hr PIIII 1113f 114 Time to Boil for C - Initial Licensed Heat Load:

115 Volume available for saturation 53.904.96 cu ft total volume of SFP C & 2/3 canal 116 Mass 3.283.907.43 lb total volurmeispecalc volume 117Heat input required to reac boling 384.658.158.06 BTU mass x (enthalpy Q initiation - enthalpy a saturated liquid) 1Time required to boiling 384.66 hr heat input I heat toad at intal licensed hmit 119 or 16.03 days 121 Additional time to boildown to top of racks -Initial Licensed Heat Load:

122 Volume available to top of racks "13Mass 41.346.00 2.473.652.58 !lb fcu it volume to top of racks for SFP C & 2/3 canal volume to top Ofracks IFspecifi volume 124 Heat input required to rea,,; top of racks 2.399.711.933.64 ! BTU mass x (enthalpy

saturated vapor - enlhalpy S saturated liquid) 12Time to rach top of racks 2,390.71 'hr heat input / heat load 17or ").9l -,days 128 - Total time from initiation until top of racks:

129 Time to boil + time to top of racks 2,764.37 hr or 116.02 .days 132 IMakeup required tooffset boiling:

133 Volume to top of racks 41.348.00 rcuaft vume to top of racks for SFPC & 2/3 canal 134 Volume to top of racks 309.268.08 gallons convert cu ft to gallons (x by 7.48)

Time to boil to top of racks 2.399.71 'hr from above 136 'Make up flow rate required 2.15 :gal/mmn volume to top of rackstime totoposfriling s 144t Pool C&D at maximum end of life heat load of 15.6 MSTU/hr

'142 ___Time to Boll for C&D - End of Life Heat Load: ___cn 143 Volu va blefr 75,537,cuft toralvolun e S--F

.- PD--&-2J3canl- .......

144 Mass 4.601.740.81 lb total volume/specific volume 145 _ Heat input required to reahbilig pnhlyiiito.enhly _ 5902667 T as 46

_ Time required to boiling auae iud

342 hr heat input I heat load at masamum end of life heat load 147 _______ or 1.43 days .. .

149 - Addional time to boil down to top of racks. End of Life Heat Load:

_ Volume available to top of racks 56I974.80 r of racks for SFP C. SFPD & 2/3 canal 1Mass 3.408.693.98 lb volume to top of racks I specific volume 153Hea input requred to reach top of k 3.306.803.741.03 BTU mass ax(enthalpy @ saturated vapor -enqialpy 0 saturated liquid) 13 Time to reach top of racks 211.14 hr heat input I heat load 154 or 6.50 days

ý156 Total time from initiation until top of racks:

57 Time to boll + time to top of racks 246.65 hr 158 or 10.23 'days Makeup required to offset boiling:

161 Volumetotopofracks 56.97480 i volume-toop ofrak-s-rSFPCSFPD&2/3canal. w 1 'Volume o toP of racks 426.171*50 'gallons convenrcurfto-gallons-( by7.48 ........

163 Time to boil to top ofnacks 211.14 he fm above .

14Makeup fow rate required 33.64 'gal/mb volume to top-of racks-/ time to top Of'racks after boiling begins 165 _________________________________

001275

I L

EXHIBIT 5 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of )

)

CAROLINA POWER & LIGHT ) Docket No. 50-400-LA COMPANY )

(Shearon Harris Nuclear Power Plant) ) ASLBP No. 99-762-02-LA AFFIDAVIT OF ERIC A. MCCARTNEy COUNTY OF WAKE )

) ss:

STATE OF NORTH CAROLINA )

I, Eric A. McCartney, being sworn, do on oath depose and say:

1. I am a resident of the State of North Carolina. I am employed by Carolina Power

& Light Company ("CP&L") and work at the Harris Nuclear Plant ("HNP" or "Harris Plant" or "Harris") in the Licensing Department. Presently, I am the Supervisor, Licensing/Regulatory Programs, and am responsible for managing regulatory interfaces for Harris. My business address is 5413 Shearon Harris Road, New Hill, North Carolina 27562-0165.

2. I was graduated from the University of Maryland in 1995 with a B.S.

in Nuclear Science. Between 1974 and 1995, 1 have held several positions of increasing responsibility in nuclear power operations, first in the United States Navy, then

-I-001 27r,

with CP&L. I obtained my Reactor Operator license in 1988 and have been a licensed Senior Reactor Operator since 1991. Since graduation from the University of Maryland, I have been employed by CP&L in supervisory positions, first as Superintendent - Shift Operations, responsible for shift activities including radioactive waste and makeup water systems at Harris. In 1997, 1was appointed Harris Superintendent - Work Control, responsible for directing day-to-day licensed activities as the senior licensed operator onsite. Between January 1999 and April 2000, 1 was an Operations Evaluator for the Institute of Nuclear Power Operations ("INPO") performing evaluations of nuclear power plant operations. I also performed World Association of Nuclear Operators ("WANO') peer reviews and assistance visits at Donald C. Cook Nuclear Power Station and Fort Calhoun Nuclear Station. In April 2000, 1 was appointed to my current position of Supervisor - HNP Licensing/Regulatory Programs, where I am responsible for preparing regulatory reports, supporting plant operations, and participating on the Plant Nuclear Safety Committee. I have also served on the Technical Support Center ("TSC") staff as the Site Emergency Coordinator, responsible for operating decisions during emergency conditions. My resume is included as Attachment A to this affidavit

3. The purpose of this affidavit is to set forth facts on which CP&L relies in establishing that numerous, diverse sources of water and methods of delivery exist for establishing makeup to the Harris spent fuel pools. First, I describe the Harris

-2 001277

spent fuel pool physical arrangement, systems configurations, and plant equipment associated with normal and alternate makeup to the spent fuel pools.

Second, I discuss the methods available for supplying makeup water to the Harris spent fuel pools and identify the Harris procedures, controls, conditions, and equipment that establish the viability of each method. Third, I describe the TSC, its functions and personnel, and how the Severe Accident Management Guidelines

("SAMGs") are used to assist the operating staff in responding to emergency conditions outside of existing procedures. Finally, I provide my conclusions on the ability of Harris operators to restore makeup water to the spent fuel pools under emergency conditions.

HARRIS PLANT SPENT FUEL POOLS, ASSOCIATED SYSTEMS AND AVAILABLE EQUIPMENT

4. As a consequence of my extensive experience as a licensed operator and manager at Harris, I am familiar with the physical layout, system configurations, equipment installations, operations, and operating procedures for Harris, as they relate to normal and alternate operation of the fuel handling building, the spent fuel pools, and associated water sources, support systems and equipment.

5. As Harris was originally envisioned as a four unit facility with a shared fuel handling building, the fuel handling building was designed and constructed with four separate pools capable of storing spent nuclear fuel. Spent fuel pools A and B were originally intended to support Harris Units 1 and 4. Spent fuel pools C

" 001278

and D were originally intended to support Harris Units 2 and 3. In addition, the fuel handling building contains a cask unloading pool, which can be connected to any spent fuel pool through transfer canals.

6. The layout of the Harris fuel handling building is illustrated and described in detail in HNP FSAR, section 9.1. Each spent fuel pool and the cask unloading pool are interconnected by a main transfer canal, oriented in a north-south direction, and two fuel transfer canals, oriented east-west. The spent fuel pools and transfer canals contain sufficient amounts of water to facilitate safe fuel handling and storage activities. The spent fuel pools, transfer canals, and cask unloading pool contain openings for the underwater movement of fuel assemblies between the pools and transfer canals. These openings also allow the communication of water between the pools and transfer canals. Removable bulkhead gates are installed in the openings when there is a need to isolate a particular pool or canal from the others. The isolation function of the bulkhead gates is provided by stainless steel structural components and inflatable seals, which are installed around the sides of the gates that fit into slots in the pool and canal openings. The seals are normally inflated using instrument air supplied at the installed gate location.

7. The normal configuration of the spent fuel pools (i.e., the configuration expected 99% of the time an on annual basis) is with open communication (i.e., the gates removed) between spent fuel pools A and B and the connecting transfer canal.

001279

The gate between the transfer canal and the main fuel transfer canal is also normally removed. The cask unloading pool is normally connected to the main fuel transfer canal. Spent fuel pools C and D are currently normally isolated from spent fuel pools A and B by a gate installed between the main fuel transfer canal and pools C and D. This alignment is illustrated in Attachment B. Once placed in operation, spent fuel pools C and D and the associated fuel transfer canal will normally be connected to the cask handling pool. Spent fuel pools A and B and their shared transfer canal will be 2onnected to the main fuel transfer canal. This alignment is illustrated in Attachment C.

8. The original Harris design included a spent fuel pool cooling and cleanup system

("SFPCCS") to service spent fuel pools A and B, and a separate SFPCCS to service spent fuel pools C and D. The SFPCCS for spent fuel pools A and B is in service. The SFPCCS for spent fuel pools C and D was not completed, but will be finished and placed in service to support spent fuel operations pursuant to the pending license amendment request.

9. The purpose of the SFPCCS is to maintain water quality in the spent fuel pools, transfer canals, cask loading pool and the reactor cavity, and remove residual heat generated in the stored spent fuel. The SFPCCS consists of a cooling system and a cleanup system. The major system components are the fuel pool heat exchangers, fuel pool demineralizer, fuel pool cooling pumps, filters, skimmers, water purification pumps, valves, piping, fuel pool gates, strainers, 001280

instrumentation and system controls. The SFCCS is comprised of two separate cooling loops, each with 100% capacity and independence. The fuel pool cooling pumps are powered from train separated power sources with the capability of being connected to the emergency diesel generator should a loss of offsite power occur. The normal source of water for the system is the Refueling Water Storage Tank ("RWST"), which has a capacity of 469,000 gallons and is maintained above 436,000 gallons at all times during reactor operation. Attachment D is the system description, SD-I 16, from Volume 6 of the Harris Plant Operating Manual, which provides a more detailed description of the SFPCCS. Attachment E is a simplified schematic of the system.

10. The demineralized water system ("DWS") is designed to process filtered water from the filtered water makeup system to produce demineralized water sufficient for the expected demands during startup and operation of various Harris plant systems, including the reactor coolant system. The DWS is capable of supplying normal makeup needs with additional capacity for maintaining level in the condensate storage tank, refueling water storage tank, and reactor makeup water storage tank, which provide makeup water capacity to bring the plant to a shutdown condition during accident conditions. The major system components include: the demineralized raw water feed pumps, carbon filters, cation, anion and mixed bed demineralizers, a vacuum degassifier, degassified water transfer pumps, a demineralized water storage tank with a capacity of 500,000 gallons, 001281

I and demineralized water storage tank transfer pumps. Attachment F is the system description, SD-143, from Volume 6 of the Harris Plant Operating Manual that provides a more detailed description of the DWS. Attachment G is a simplified schematic of the system.

11. The RWST has two functions. First, it provides a borated water source for the charging safety injection pumps, containment spray pumps and residual heat removal pumps for injection into the reactor vessel during accident conditions that threaten core uncovery. The RWST is also a source of water to fill the fuel transfer canals for refueling operations. The RWST has a capacity of 469,000 gallons and is maintained greater than 92% at all times while the plant is operating. Technical Specifications require the engineered safety features actuation system to swap safety injection pump suction from the RWST to the containment sump at 23.4% RWST level to prevent possible safety injection pump damage. This leaves approximately 100,000 gallons of water available for use by other systems following safety injection. Attachment H is the system description, SD-1 12, from Volume 6 of the Harris Plant Operating Manual that provides a more detailed description of the RWST. Attachment I is a simplified schematic of the RWST and connecting systems.

12. The normal service water system ("NSW") circulates water from the cooling tower and cooling tower makeup system through plant auxiliary components and back to the cooling tower. The NSW provides cooling water to the ESW headers 001282

during normal operations. The key components of the system include two (2)

NSW pumps and interconnecting piping. The pumps are powered from non safety auxiliary busses. The sources of water for NSW are the main and auxiliary reservoirs. Attachment J is the system description, SD-139 from Volume 6 of the Harris Plant Operating Manual that provides a more detailed description of the NSW System. Attachment K is a simplified schematic of the NSW System.

13. The emergency service water ("ESW") system circulates water from the ultimate heat sink ("UHS") through the plant components required for safe shutdown of the reactor following an accident and returns the water to the UHS. The ESW system provides an emergency source of water for the auxiliary feedwater system, essential services chilled water system and fire protection system. Key system components include two (2) ESW pumps and two (2) ESW booster pumps, are powered from safety-related buses. The sources of water for the ESW system are the main and auxiliary reservoirs. Attachment J is the system description, SD 139, from Volume 6 of the Harris Plant Operating Manual that provides a more detailed description of the ESW system. Attachment L is a simplified schematic of the system.

14. The reactor makeup water storage tank ("RMWST") serves as a storage volume for makeup water to nuclear steam supply system, specifically the component cooling water ("CCW") system and reactor coolant system ("RCS"). The RMWST has a capacity of 85,000 gallons. Attachment M is the system

-8o 00i1283

L description, SD- 102, from Volume 6 of the Harris Plant Operating Manual that provides a more detailed description of the RMWST. Attachment N is a simplified schematic of the RMWST and connecting systems.

15. The purpose of the fire protection system is to ensure the capability to shut down the reactor safely, maintain it in a safe shutdown condition, and limit the radioactive release to the environment in the event of a fire. It also serves to minimize both the probability and the consequences of fires, thereby protecting plant personnel and plant related equipment and property. Two fire pumps are provided, one motor driven pump powered from a non-safety power supply and one diesel engine driven pump, each with a capacity of 2500 gallons per minute.

Both fire pumps draw suction from the Auxiliary Reservoir. Attachment 0 is the system description, SD-149, from Volume 6 of the Harris Plant Operating Manual that provides a more detailed description of the Fire Protection System.

Attachment P is a simplified schematic of the system.

16. Harris has agreements with the Holly Springs and Apex Fire Departments to provide assistance in emergency situations. The Apex fire station is approximately three miles from the Harris site. Generally, Holly Springs and Apex respond to the site in 15 to 30 minutes from the phone call requesting assistance. Additional emergency resources are available from Raleigh and Sanford fire departments, with which Harris has also established working relationships. These fire departments are familiar with Harris because they 001284

participate in annualdrills requiring off-site fire department response as part of the Harris Emergency Plan.

17. Harris Fire Brigade members are provided with a full set of structural fire fighting turnout gear and Self-Contained Breathing Apparatus

("SCBA') suitable for performing actions in high temperature and high humidity conditions. This gear includes boots, bib overalls, coat, hood and helmet.

Harris turnout gear is generally manufactured from NomexTM and GortexTM materials, which are non combustible and vent moisture during use.

18. Plant Fire Brigade members undergo initial and annual re-qualification LIVE Fire Training using turnout gear and SCBA. This training includes structural fire training (entry and extinguishing a actual structural fire) in a building. Fires during these training evolutions typically involve temperatures in the fire room well above the 1950 F that is anticipated in the Harris fuel handling building during postulated spent fuel pool boiling. Temperatures in the range of 3000 F would not be unusual during Fire Brigade training exercises. Fire fighting with a fire hose normally produces high humidity conditions and Fire Brigade members are trained to perform under such conditions. Additionally, Fire Brigade members conduct periodic drill in various plant locations to exercise their ability to use fire fighting equipment and techniques.

19. Emergency lighting is installed and regularly maintained in locations where operator actions may be required following a loss of normal lighting. In 00.1. .8

I particular, the plant areas containing valves and equipment necessary to provide normal and alternate fuel pool cooling and make up contain normal and emergency lighting, either of which is adequate to enable an operator to perform the actions required to establish flow to the fuel pools. In addition, portable, battery-powered flashlights and lanterns are available in each emergency facility and the control room. These storage locations are periodically inventoried to ensure the equipment is available and working. Normal and emergency lighting is more fully described in system de.;cription SD-I 58, Plant Lighting (Attachment Q).

20. Ladders are strategically located in designated ladder storage areas throughout the fuel handling and reactor auxiliary buildings. Ladder locations are marked with placards and a log is maintained in the operator's work area showing the location of ladders in the plant. These ladders are staged for the specific purpose of providing operators access to elevated valves and equipment, and operators are familiar with their locations.

21. Auxiliary operators are trained to operate locked doors and valve locking devices under normal and emergency conditions. The doors in the fuel handling building and reactor auxiliary building do not require electrical power to be operated manually. Power provides only alarm and indication for the security system.

Each operator carries keys that operate plant doors, as well as keys for locked equipment, such as valves and tool boxes. Additional keys of both types are

-U1-001286

maintained in the main control room. In addition, following the activation of the Emergency Plan, or on a loss of site power, security personnel are tasked to assist operating staff with access to locked areas. Finally, bolt cutters and torches can be obtained from the site tool room located in the service building. Operators and security personnel are trained and familiar with operating plant doors and equipment under normal and emergency conditions.

METHODS FOR SUPPLYING MAKEUP WATER TO THE HARRIS SPENT FUEL POOLS

22. There are numerous normal and alternate methods for supplying makeup water to the Harris spent fuel pools. To the best of my knowledge and belief, each of the ten (10) methods described below is individually capable of delivering makeup water to the spent fuel pools at a rate in excess of the highest evaporation or boil off rates calculated for the beyond-design-basis accident with containment bypass scenario described in Contention EC-6.

23. In each method described below, makeup water will be available to each spent fuel pool regardless of the makeup water discharge location or the bulkhead gate configuration. Makeup is available to spent fuel pool A and B (and spent fuel pools C and D when they are placed in service) in the normal bulkhead gate configuration by open communication through the transfer canals. If makeup is required when a spent fuel pool is isolated by one or more installed bulkhead gates, makeup water can be made available simply by depressurization of the inflated gate seals. With the seals depressurized, makeup water will communicate 00128'

L into, and equalize the levels of, the spent fuel pools, transfer canals, and cask unloading pool. The seals can be deflated locally by bleeding the pressure off through a vent valve. In the event the gate seals cannot be depressurized, makeup water from a filling pool or transfer canal will overflow the installed gates and fill the other pools and canals.

24. In addition to the ten (10) methods described below, following installation of the plant modifications associated with spent fuel pools C and D, a completely redundant spent fuel pool cooling system, purification system, and skimmer system will be installed in the north end of the fuel handling building. This will provide four (4) additional redundant delivery locations for operators to align existing makeup water sources to the spent fuel pools, transfer canals, and cask loading pool. Procedure OP-1 16 will be revised to reflect the additional redundant makeup water pathways before adding spent fuel to spent fuel pool C.

Normal Makeup

25. Normal spent fuel pool makeup is accomplished by OP-1 16, Revision 17, section 8.4, "Makeup to SFP B with Demineralized Water with the Purification System in Service" (Attachment R). To initiate normal makeup, plant operators have to open only a single manual valve, I SF-20 1, located on the 216-foot elevation at the south end of the fuel handling building. This action aligns the demineralized water system to the spent fuel purification system and delivers makeup water directly to spent fuel pool B. As described above, makeup is available to spent

-13 0 0128S

fuel pool A (and spent fuel pools C and D when they are placed in service) in the normal bulkhead gate alignment by open communication through the transfer canals and to other pools by either deflating the gate seals or overflowing the gates. The source of water for this mode of makeup is the dernineralized water storage tank, which has a capacity of 500,000 gallons. The flow rate is approximately 100 gallons per minute. An operator can reach the valve needed to align this flow path within approximately 15 minutes from any location in the reactor auxiliary building ("RAB") or fuel handling building ("FHB") and can initiate flow within another 5 minutes. The procedural reference for this method is OP- 116, Revision 17. Once the SFPCCS for spent fuel pools C and D is placed in service, normal makeup to pools C and D will be initiated by opening a single manual valve, 2SF-201, located on the 216-foot elevation at the north end of the fuel handling building.

Alternate Makeup Method No. I - Demineralized Water Tank to Spent Fuel Pool Purification System

26. This method aligns the demineralized water tank to the spent fuel pool purification system and delivers water to spent fuel pool A, spent fuel pool B, the interconnecting transfer canal, the cask loading pool, or all of these locations. To provide makeup water to spent fuel pools A and B, the operators manually opens ISF-201 and any one of the following valves 1SF-26, -27, -28, -29, or -192. One of these valves is located on the 216 foot elevation of the FHB, two are on the 236 foot elevation of the FHB, and one is on the 261 foot elevation of the FHB. To 001289

I align makeup water to the cask loading pool an additional five (5) valves, 1SF 206, 2SF-205, 2SF-141, 188, and 203, are opened. All of these valves are located in the FHB on the 216 foot elevation. After establishing the desired valve lineup, the operator closes power supply breakers 1-4A 1021 -I D and 1-4B 1021-5E for the spent fuel pool purification pumps on the 261-foot elevation of the FHB. Once energized, the operator starts the purification pump from one of two locations (i.e., the FHB operating deck or the FHB 236-foot elevation). The source of water is the demineralized water storage tank with a capacity of 500,000 gallons with a flow rate of approximately 100 gallons per minute. An operator can access these valves within 15 minutes and initiate flow in approximately 30 minutes. The procedural reference for this method is OP-1 16, Revision 17, section 8.5. When spent fuel pools C and D and the associated SPFCCS are placed in service, a redundant alignment to implement this alternate method will be available by opening 2SF-201 and any one of the following valves, 2SF-26, -27, -28, -29 or -192.

Alternate Makeup Method No. 2 - Refueling Water Storage Tank (RWST) to SDent Fuel Pool Purification System

27. This method aligns the RWST to spent fuel pools A and B, the interconnecting transfer canal, the cask loading pool, or all of these locations simultaneously. To align this flow path, an operator manually shuts 1SF-202 on FHB 216 and opens one or all of the following valves: ISF-26 or -27 to spent fuel pool A; ISF-28 or

-29 to spent fuel pool B; or 1SF-192 to the south transfer canal. One of these 00.12 0

valves is on the FHB 216 foot elevation and the remainder are on the FHB 236 foot elevation. The operator then opens 1CT-23 on the RAB 236 foot elevation.

After establishing the desired valve lineup, the operator closes power supply breakers 1-4A 102 1-ID and 1-4B 1021-SE for the spent fuel pool purification pumps on the FHB 261-foot elevation. Once energized, the operator starts the purification pump from one of two locations (i.e., the FHB operating deck or the FHB 236-foot elevation). The source of this flow path is the RWST with a capacity of 469,000 gallons. If the RWST is approximately 50% or more full, as required during plant operations, this flow path will result in gravity flow to the spent fuel pools, transfer canal, or cask loading pool with an expected flow rate of up to 100 gallons per minute. When spent fuel pools C and D and the associated SFPCCS are placed in service, a redundant method to implement this flow path will be available by aligning 2SF-202, -26, -27, -28, -29, and 192. The Unit 1 RWST will be the source of water through 1CT-23. An operator can access these valves within approximately 15 minutes and initiate flow in approximately 30 minutes. The procedural reference for this method is OP-1 16, Revision 17, section 8.5.

Alternate Makeup Method No. 3 - Demineralized Water System to Spent Fuel Pool Skimmer System

28. This method aligns the demineralized water system to the spent fuel pool skimmer system and delivers makeup water to spent fuel pool A, spent fuel pool B, or the 001291

I transfer canal. The method requires the skimmer system to be in service, which is normally the case. To align this flow path, the operator simply opens valve IDW 527 on the 236-foot elevation of the fuel handling building. The source of water is the demineralized water storage tank with a capacity of 500,000 gallons. The flow rate is approximately 100 gallons per minute. An operator can access this valve and initiate flow in approximately 5 minutes. The procedural reference for this method is OP-1 16, Revision 17, section 8.6.

Alternate Makeup Method No. 4 - Refueling Water Storage Tank (RWST) to Spent Fuel Pool Cooling System

29. This method aligns the RWST to the suction of the spent fuel pool cooling pumps and delivers water to the spent fuel pool A/B transfer canal, the main fuel transfer canal, or the cask loading pool. To align this flow path the operator manually aligns eleven (11) valves. Eight (8) of the valves, 1SF-1, -5, -9, -10, -21, -23, -24, and -25, are located on the FHB 236-foot elevation; 1SF-193 is located on the FHB 216-foot elevation, and two (2) valves, 3BR-378 and 1CT-23, are located on the RAB 236-foot elevation. The source of water is the RWST with a capacity of 469,000 gallons. If the RWST level is above approximately 50% full (as is required during reactor operation) then the transfer canal or cask loading pool will fill due to gravity with a flow rate of up to approximately 500 gallons per minute.

The spent fuel pool cooling pump is then started from the main control room to distribute water to all the pools. Ai operator can access these valves within approximately 15 minutes and initiate flow in approximately 30 minutes. The

-17.

001292

procedural reference for this method is OP- 116, Revision 17, section 8.12. When spent fuel pools C and D and the associated SFPCCS are placed in service, a redundant flow path will be available by aligning 2SF-i, -5, -9, -10, -21, -23, -24,

-25, and -193. The Unit I RWST is the source through valve ICT-23.

Alternate Makeup Method No. 5 - Emergency Service Water (ESW) System to Spent Fuel Pool Cooling System

30. This method aligns the ESW system to the spent fuel pool cooling system. To align this flow path, the operator connects a jumper between spent fuel pool cooling system connection valves, 1SW-1239(269) and 1SF-76, on the RAB 236 foot elevation. A toolbox is staged locally which contains all the necessary hose and fittings to install the jumper. The operator then opens the two valves to initiate flow. The source of water is Harris Lake, which provides a virtually unlimited supply of makeup water at a flow rate is approximately 50 to 75 gallons per minute. The operator can align this flow path within 30 minutes, as all the tools and equipment necessary to align this path are with 50 feet of the valves.

The procedural reference for this method is OP-1 16, Revision 17, section 8.13.

Alternate Makeup Method No. 6 - Reactor Makeup Water Storage Tank (RMWST) to Spent Fuel Pool Purification System

31. This method aligns the RMWST to the spent fuel pools. To connect the RMWST to the spent fuel pool purification system, an operator must manually align four (4) valves: 1SF-194 on FHB elevation 216, and 3PM-83, IPM-81, and IPM-150 on RAB elevation 261. The source of water is the RMWST with a capacity of 00 .1293

80,000 gallons. The flow rate is 75 to 100 gallons per minute. An operator can access these valves within approximately 15 minutes and initiate flow in approximately 30 minutes. The procedural reference for this method is OP- 116, Revision 17, section 8.26.

Alternate Makeup Method No. 7 - Fire Protection System to Spent Fuel Pool or Transfer Canals

32. The plant fire protection system draws water from the Harris Lake via a motor driven fire pump or a redundant diesel driven fire pump. There are seven (7) fire hose stations on the operating deck of the fuel handling building, each equipped with a 1 1/2 inch fire hose and connected to a normally pressurized fire water header. Several of these stations, at least one on each level of the building, are seismically qualified. The fire header is maintained pressurized.by the fire pump.

If the motor driven pump fails, a diesel driven fire pump will automatically start as a backup. To begin filling the spent fuel pools or transfer canal, an operator simply aims a fire hose at the spent fuel pool and opens the hose station valve.

Each fire hose can deliver approximately 125 gallons per minute. The fire hoses can be attached to a railing in the fuel handling, precluding the need to station an operator on the fuel handling deck. An operator can access these valves and equipment and initiate flow in approximately 5 minutes.

Alternate Makeup Method No. 8 - Normal Water Service ("NSW") System to Spent Fuel Pool or Transfer Canals

33. This method uses water from the NSW system to fill an accessible spent fuel pool 00 1294

or transfer canal through a hose. The NSW system extends into the waste processing building, where an operator connects a 1-inch drain hose to a NSW system drain valve, 1SW-817, at the FHB 261-foot elevation stair well. Tools and hose can be obtained from the tool room in the service building. The operator then runs approximately 300 feet of hose from the valve to deliver makeup water to spent fuel pool A. A one inch drain hose is expected to provide a flow rate of approximately 50 to 75 gallons per minute. An operator can access these valves and equipment within approximately 45 minutes and initiate flow in approximately another 25 minutes.

Alternate Makeup Method No. 9 - Demineralized Water System to Spent Fuel Pools or Transfer Canals

34. This method provides water from the demineralized water system directly to the spent fuel pools or transfer canals. The west wall of the fuel handling building operating deck contains several service connections for demineralized water designed to connect hoses for use in filling or wash down activities on the fuel handling building operating deck. The demineralized water header is normally pressurized and, by connecting hoses located on the fuel handling building operating deck and opening one (1) header isolation valve, an operator can fill an accessible spent fuel pool or truansfer canal at approximately 100 gallons per minute. An operator can access these valves and initiate flow in approximately 5 minutes.

00.1295

TECHNICAL SUPPORT CENTER (TSC) EMERGENCY MANAGEMENT

35. The TSC is an Emergency Plan facility that is staffed upon declaration of an Alert or higher classification. The Site Emergency Coordinator ("SEC") is the lead position, in charge of coordinating TSC support of main control room response to plant emergencies. The TSC is staffed with representatives from key plant support groups including: operations, radiological control, security, offsite communications, and technical and accident assessment. The TSC is responsible for all on-site emergency response actions. The SEC is trained in the Severe Accident Mitigation Guidelines ("SAMGs"), which provide guidance for beyond design basis events that are not specifically covered in Emergency Operating Procedures. The SEC is in constant communication with the main control room, the operational support center ("OSC") (damage control missions are dispatched from this facility), and the emergency operations facility ("EOF") (coordinates off-site protective actions). The SEC receives important information from the main control room, such as notification of the loss of spent fuel cooling, and coordinates recovery missions with the OSC and control room. The TSC sets the priority for all site missions in response to accident conditions. A significant resource available to the SEC is the Accident Assessment Team, which consists of multi-disciplined, experienced engineers tasked with developing solutions to problems and conditions not specifically addressed in plant procedures.

36. The SAMGs are implemented when core temperatures exceed a pre-determined 00129 6

value and efforts to restore cooling have been unsuccessful. The SAMGs are designed primarily for the Emergency Response Organization staff to fill the void between the EOPs and Emergency Plan regarding formalized guidance for severe accident situations. The primary goal of the SAMGs is to protect fission product barriers and mitigate any ongoing fission product releases, with a secondary goal to mitigate severe accident phenomena and return the plant to a controlled stable condition.

37. In evaluating an emergency, the TSC staff begins by monitoring a Diagnostic Flow Chart ("DFC") and the Severe Challenge Status Trees which provide guidance to diagnose challenges to fission product barriers. Predetermined values are used to direct implementation of specific SAMGs, which provide a number of solutions to potential challenges to a given barrier. The goals of the SAMGs are to 1) prevent core damage, 2) terminate the progress of core damage if it begins and retain the core in the reactor vessel, 3) maintain containment integrity as long as possible, and 4) minimize offsite releases. Work sheets that provide guidance on a number of mitigation options are included to aid in the development of an implementing strategy. The TSC then evaluates the current equipment capabilities and develops a course of action. The SEC also directs the Accident Assessment Team to develop strategies for the given situation. An analysis of the impacts of the strategies is conducted with the SEC and an strategy is selected and implemented.

.22-001297

IL CONCLUSIONS

38. There are numerous, diverse methods for providing cooling and makeup water to the Harris spent fuel pools following a loss of normal cooling.

39. Harris operators are trained and capable of performing the actions necessary to initiate one or more of these methods under emergency conditions, although the exact method or methods employed may depend on the specific plant conditions existing at the time. The necessary tools and equipment are available to perform the required actions.

40. Personnel assigned to the Harris TSC are familiar with the SAMGs and are trained to assist plant operators responding to emergency conditions outside of existing emergency response procedures.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on November /s7 , 2000.

-rnc A. Mk/Cartm~yo Subscribed and sworn to before me this /5 day of November 2000. S Y4.9,S%.

My Commission expires: -* -- /-2-05 K >=

-23.-

001298

ML021700171

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