ML20196K886

Declaration of Gordon Thompson.* Informs of Participation in Preparation of Orange County Contentions Re Proposed License Amend

ML20196K886
Person / Time
Site:	Harris
Issue date:	04/05/1999
From:	Thompson G AFFILIATION NOT ASSIGNED
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ML20196K880	List: ML20196K877 ML20196K886 ML20196K904
References
99-762-02-LA, 99-762-2-LA, LA, NUDOCS 9904080123
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i EXHI3IT 1 April 5,1999 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 (Shearon Harris Nuclear )

Power Plant) )

)

DECLARATION OF DR. GORDON THOMPSON i l

1, Gordon Thompson, declare as follows:

A. Introduction

1. I am the executive director of the Institute for Resource and Security Studies (IRSS), a l nonprofit, tax-exempt corporation based in Massachusetts. Our office is located at 27 I Ellsworth Avenue, Cambridge, MA 02139. IRSS was founded in 1984 to conduct technical and policy analysis and public education, with the objective of promoting peace and intemational security, efficient use of natural resources, and protection of the enviromnent.

2. I received an undergraduate education in science and mechanical engineering at the University of New South Wales,in Australia. Subsequently, I pursued graduate studies at Oxford University and received from that institution a Doctorate of Philosophy in mathematics in 1973, for analyses of plasmas undergoing thennonuclear fusion. During my graduate studies I was associated with the fusion research program of the UK Atomic Energy Authority.

3. During my professional career, I have performed technical and policy analyses on a range ofissues related to intemational security, energy supply, environmental protection, and sustainable use of natural resources. Since 1977, a significant part of l my work has consisted of technical analyses of safety and environmental issues

related to nuclear facilities. These analyses have been sponsored by a variety of nongovernmental organizations and local, state and national govemments, predomimmtly in North America and western Europe. Drawing upon these analyses, I have provided expert testimony in legal and regulatory proceedings, and have served on committees advismg US government agencies. A copy of my resume is attacned as Attachment 2 to the Declaration of Dr. Gordon Thompson (February 12,1999)

(" Thompson No Significant Hazards Declaration"), which is attached as Exhibit 2 to Orange County's Supplemental Petition to Intervene Spril 5,1999).

9904080123 990405 PDR ADOCK 05000400 m

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4. I have reviewed the D~*=har 23,1998, license amandmaat application Sled by Carolina Power and Light (CP&L) for an amendment to Facility Operating License I'

No. NPF-63, which seeks permission to activate spent fuel storage pools C and D at the Shearon Harris nuclear power plant. I have also reviewed the NRC's Federal Register notice for the proposed license amaadmaat, the Final Safety Analysis Report i for the Shearon Harris Nuclear Power Plant, and the Finai Envi Unmental Statement related to the operation of Shearon Hards Nuclear Power Plant, Units 1 and 2 -

(NUREG-0972, October 1983). In addition, I reviewed various corrydence and technical documents relating to the proposed license amendment and to dsks of spent fuel storage, which are identi6ed in Orange County's contentions.

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5. I participated in the preparation of Orange County's contentions regarding the proposed license smaadmar* The factual assertions and expressions of technical

- judgment in tuvi,; ;;;;:ctd:,ns summadze the facts and professiocal opinions to which I would testify if called as a witness in this procaa&ag. and the documents cited in the contentions constitute decaments that I would rely on in my ta=+imany.

eeeeeeeeeeeeeeeeeeeeeeee I declare, under penalty of perjury, that the forego ~mg is true and correct.

Executed on 5 Apdl 1999.

i wy.C m Gorden Thompson

February 12,1999 EXHIBIT 2 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE NRC STAFF In the Matter of )

)

CAROLINA POWER & LIGHT ) Docket No. 50-400 (Shearon Harris Nuclear )

Power Plant) )

)

DECLARATION OF DR. GORDON THOMPSON I, Gordon Thompson, declare as follows:

A. Introduction

1. I am the executive director of the Institute for Resource and Security Studies (IRSS), a nonprofit, tax-exempt corporation based in Massachusetts. Our office is located at 27 Ellsworth Avenue, Cambridge, MA 02139. IRSS was founded in 1984 to conduct technical and policy analysis and public education, with the objective of promoting peace and intemational security, efficient use of natural resources, and protection of the environment.

2. This Declaration penains to an application by Carolina Power and Light (CP&L) for

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an amendment to Facility Operating License No. NPF-63, which covers the Shearon Harris nuclear power plant. The staff of the Nuclear Regulatory Commission (NRC) has reviewed CP&L's application and proposes to determine that the amendment request involves no significant hazards consideration. The NRC has sought public comments on the proposed detennination.1 Through this Declaation, I offer comments on the NRC staffs proposed determination. I have prepared these comments pursuant to an agreement by IRSS to provide technical information and other services to Orange County, North l Carolina.

B. My Professional Background  !

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3. I received an undergraduate education in science and mechanical engineering at the l University of New South Wales,in Australia. Subsequently, I pursued graduate studies i at Oxford University and received from that institution a Doctorate of Philosophy in l mathematics in 1973, for analyses of plasmas undergoing thermonuclear fusion. During l my graduate studies I was associated with the fusion research program of the UK Atomic )

Energy Authority.

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3 Federal Register: January 13,1999 (Volume 64, Number 8), pages 2237-2241.

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4. During my professional career, I have performed technical and policy analyses on a l range ofissues related to intemational securky, energy supply, environmental I protection, and sustainable use of natural resources. Since 1977, a significant part of my work has consisted of technical analyses of safety and environmental issues related to nuclear facilities. These analyses have been sponsored by a variety of nongovernmental organizations and local, state and national govemments,

% rdominantly in North America and western Europe. Drawing upon these analyses, i r> ave provided expert testimony in legal and regulatory proceedings, and have served on committees advising US govemment agencies. My CV is provided here as Attachment A.

C. Scope of My Review l

5. In preparation of this Declaration, I reviewed the NRC's Federal Register notice for the proposed license amendment, the Final Safety Analysis Report for the Shearon Harris j Nuclear Power Plant, the Final Environmental Statement related to the operation of 1 Shearon Harris Nuclear Power Plant, Units 1 and 2 (NUREG-0972, October 1983), and CP&L's application for the proposed license amendment. I also reviewed various correspondence and technical documents relating to the propose license amendment and to risks of spent fuel storage, which are identified below.

6. The information that has been provided by the NRC and CP&L to date does not I contain all of the detail that I would need to provide a complete, final statement about the l hazards associated with the proposed license amendment. I would expect to review the j full body of detailed evidence and present my final evaluation in the context of a hearing. l However, even the limited information provided so far is adequate to permit me to identify serious safety concems which preclude the NRC from making a "no significant hazards" determination. These issues should be addressed through the systematic, public i process that a prior licensing hearing can provide. I D. The "No Significant Hazards" Standard i l

7. The NRC has stated its standard for determining that a license amendment request involves no significant hazards consideration.2 The standard is met if operation of the i facility in accordance with the proposed amendment would not: (1) involve a significant increase in the pro'oability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

8. In my professional opinion, based on the preliminary evidence provided by the NRC and CP&L, operation of the Shearon Harris plant in accordance with the license j amendment proposed by CP&L will violate all three of the conditions set forth in the preceding paragraph. Therefore, the NRC staff should reverse its position and should 2 Ibid.

3 determine that CP&L's license amendment request does act involve no significant hazards consideration.

E. The License Amendment in Context - Spent Fuel Management at Harris

9. Before discussing my concerns about the safety implications of the proposed license amendment, I provide here some background information about spent fuel management at the Harris plant and CP&L's proposal to increase the spent fuel storage capacity at Harris. Unless specified otherwise, the information presented here is drawn from CP&L's license amendment application or from CP&L's Final Safety Analysis Report (FSAR) for the Harris plant.

10. The Harris plant features one pressurized-water reactor (PWR). The core of this reactor contains 157 fuel assemblies, with a center-center distance of about 8.5 inches.

The Harris plant was to have four reactors but only one was built. A fuel handling building was built to serve all four reactors. This building contains four fuel pools (A, B, C, D), a cask loading pool and three fuel transfer cr.nals, all interconnected but separable by gates. Pools A and B contain fuel racks. Pools C and D are flooded but do not contain racks. The cooling and water cleanup systems for pools C and D were never completed.

11. Pool A now contains six PWR racks (360 fuel assembly spaces) and three BWR racks (363 spaces), for a total pool capacity of 723 fuel assemblies. Pool B contains twelve PWR racks (768 spaces) and seventeen BWR racks (2,057 spaces), and is licensed to store one additional BWR rack (121 spaces), for a total pool capacity of 2,946 fuel assemblies. Thus, pools A and B now have a combined capacity of 3,669 fuel assemblies. The center-center distance in pools A and B is 10.5 inches for PWR fuel and 6.25 inches for BWR fuel.

12. Pools A and B store spent fuel from the Harris reactor and from CP&L's Brunswick plant and Robinson plant. The Brunswick plant has two boiling-water reactors (BWRs) while the Robinson plant has one PWR. Shipment of spent fuel from Brunswick and Robinson to Harris is said by CP&L to be necessary to allow core offload capacity in the pools at Brunswick and Robinson.

13. CP&L seeks an amendment to its operating license so that it can activate pools C and D at Harris. By activating these pools, CP&L expects to have sufficie' . spent fuel storage capacity for all four CP&L reactors (Harris, Robinson and the .wo Brunswick reactors) through the end of their current operating licenses.

14. O&L plans to install racks in pool C in three campaigns (approximately in 2000, 2005 and 2014), to create 927 PWR spaces and 2,763 BWR spaces, for a total pool capacity of 3,690 fuel assemblies. Thereafter, CP&L plans to install racks in pool D in two campaigns (approximately in 2016 and at a date to be determined), to create 1,025 PWR spaces. Thus, the ultimate capacity of pools C and D will be 4,715 fuel assemblies.

The center-center distance will be 9.0 inches for PWR fuel and 6.25 inches for BWR fuel.

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15. The PWR racks in pools C and D have a smaller center-center distance than the racks in pools A and B (9.0 inches instead of 10.5 inches). This arrangement allows more PWR fuel to be placed in a given pool area but also means that PWR fuel in pools C and D is more prone to undergo criticality. In response, CP&L proposes to include in the Technical Specifications for Harris a provision that PWR fuel will not be placed in pools C and D unless it has relatively low enrichment and high burnup.)

F. Some Technical Safety Issues Raised By the Proposed License Amendment

16. CP&L's plan for the activation of pools C and D raises a variety of technical safety issues. This section of my Declaration describes some of those issues. Later parts of the Declaration relate these issues to the NRC's standard for a "no significant hazards" ,

determination. l 1

l7. NRC regulations require that spent fuel storage pools must be cooled by safety grade {

,ooling systems. When the Harris plant was designed, the intention was that pools C and D would be cooled by the component cooling water (CCW) system for the second unit of the Harris plant.d That unit was never built, and therefore the Unit 2 CCW system does not exist. In the absence of a second CCW system, CP&L plans to cool pools C and D  ;

by connecting their cooling systems to the CCW system of the first unit. This system )

already provides cooling to pools A and B and serves other, important safety functions.

Attachment B provides supporting information.5 It should be noted that CP&L considered, but has not pursued, the option of cooling pools C and D by a new, independent system that could have had dedicated emergency diesel generators.

Attachment C provides information in support of this point.6 Three significant safety issues are raised by the fact that the spent fuel pool cooling arrangement originally I designed for pools C and D of the Harris plant was not completed. These issues relate to i the heat loading of the existing CCW system, the load on the existing emergency diesel generators, and the loss of some important quality assurance documentation for cooling piping at pools C and D.

18. Heat load. According to CP&L's license amendment application, the bounding heat load from the fuelin pools C and D will be 15.6 million BTU / hour.7 At present, the CCW system cannot absorb this additional heat load. Thus, CP&L proposes to include in 3 License amendment application, Enclosure 5.

8 The Harris pools have their own closed-circuit cooling systems, which can transfer beat to the relevant CCW system through heat exchangers.

5 Attachment B is a portion of a set of viewgraphs (titled " Harris Spent Fuel Pool 'C and 'D' Activation")

shown by CP&L representatives during a meetmg with NRC staff on 16 July 1998.

6 Attachment C is an NRC staff memo about a meeting between CP&L representatives and NRC staff on 3 March 1998, together with a portion of a set of viewgraphs (titled "HNP Spent Fuel Pool 'C and 'D' Activation") shown by CP&L during that meeting.

7 License am ndment application, Enclosure 7, page 5-16.

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5 the Technical Specifications for Harris an interim provision that the heat load in pools C and D will not be allowed to exceed 1.0 million BTU / hour.8 CP&L claims that an additional heat load of 1.0 million BTU / hour can be accommodated by the existing CCW system, and that the fuel to be placed in pools C and D will not create a heat load exceeding 1.0 million BTU / hour through 2001.

19. Apparently, CP&L contemplates a future upgrade of the CCW system, so that the CCW system can accommodate an additional heat load of 15.6 million BTU / hour from pools C and D. This contemplated upgrade is not described in the present license amendment application. Attachment C indicates that CP&L plans to perform the upgrade of the CCW system concurrent with a power uprate for the Harris reactor. Apparently, a 4.5 percent power uprate will be associated with steam generator replacement, and there will be a subsequent further power uprate of 1.5 percent. A chart in Attachment C shows that the projected CCW heat load, including the reactor power uprate and the use of pools C and D, will substantially exceed the capability of the present CCW system.

20. To summarize, CP&L's short-term plan (through 2001) for cooling pools C and D is to exploit the margin in the existing CCW system, so as to accommodate an additional l heat load of 1.0 million BTU / hour. CP&L's longer-term plan is to upgrade the CCW system, in a manner not yet speciDd, so as to accommodate an additional heat load of 15.6 million BTU / hour. The CCW upgrade must also accommodate an increase in the rated power of the Harris reactor. Attachment B indicates CP&L's expectation that the design of the CCW upgrade will commence in mid-1999 and will be completed in early 2001, one year after pool C enters service.

21. In order to avoid exceeding the available margin in the existing CCW system while cooling pools C and D, CP&L may be obliged to require its operators to divert some CCW ilow from the residual heat removal (RHR) heat exchangers during the recirculation phase of a design-basis loss-of-coolant accident (LOCA) event at the Harris reactor.' This raises a safety issue because, during the recirculation phase of a LOCA, operation of the RHR system is essential to keeping the reactor core and containment in a safe condition. Both CP&L and the NRC have identified the proposed additional heat load on the Unit 1 CCW system as an "unreviewed safety question," i.e., a safety question that has not been previously reviewed by the NRC Staff.10 It should be noted in this context that exploitation of the margin in the existing CCW system may involve changes in design assumptions that include fouling factors and tube plugging limits. See Attachment C. The discussion of CCW capability which is provided in Enclosure 9 of CP&L's license amendment application is insufficient to determine the nature and significance of the assumptions made by CP&L. <

22. Backup diesel generators. The cooling systems for pools C and D will draw electrical power from the electrical systems of the existing Harris plant. If electricity i

) 8 License amendment application, Enclosure 5.

l 9 License amendment application, Enclosure 9.

to Ibid; Federal Register notice for this application.

6 supply to the cooling pumps for pools C and D is interrupted, the pools will heat up and eventually boil. CP&L says that pools C and D will begin to boil after a time period "in execss of 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />", assuming a bounding decay heat load of 15.6 million BTU / hour.Il To prevent the onset of pool boiling in the event of a loss of offsite power, the Harris operators may be obliged to provide electrical power to pools C and D from the emergency diesel generators, which also serve pools A and B and the reactor. In the present license amendment application, CP&L does not address the ability of the emergency diesel generators to meet the additional electncal loads associated with pools C and D. CP&L does mention in the Harris FSAR the potential for connecting " portable pumps" to bypass the pool cooling pumps should the latter be inoperable.12 However, the characteristics, capabilities and availability of such portable pumps are not addressed in the present license amendment application. Meeting the electrical load of pools C and D from the systems of the existing Harris plant is a safety issue because it could increase the probability of design-basis or severe accidents at the Harris reactor or at pools A through C.

23. Lack of QA documents. Activation ofpools C and D will require the completion of their cooling and water cleanup systems, and the connection of their cooling systems to the existing CCW system. CP&L states that approximately 80% of the necessary piping was completed before the second Harris reactor was cancelled.o However, some of the quality assurance documentation for the completed piping is no longer available. Much of the completed piping is embedded in concrete and is therefore difficult or impossible to inspect. To address this situation, CP&L proposes an Alternative Plan to demonstrate that the previously completed piping and other equipment is adequate for its purpose.18 4 Nevertheless, the cooling systems for pools C and D will not satisfy ASME code requirements. Attachment D provides supporting information.ts Failure to satisfy ASME code requirements could increase the probability of design-basis or severe accidents at pools C and D.

G. The Degree of Hazard Posed by Spent Fuel Storage at Harris

24. The NRC and CP&L have performed and published site-specific analyses which provide information about potential severe accidents at the Harris reactor. However, to my knowledge neither NRC nor CP&L has performed any site-specific analysis which 11 License amendment application, Enclosure 7, page 5-8.

12 Harris FSAR, page 9.1.3-4, Amendment No. 48.

13 License amendment application, Enclosure 1, page 4 Id License amendment application, Enclosure 8.

15 Attachment D is a portion of a set of viewgraphs (titled "10CFR50.55a Alternative Plan") shown by CP&L representatives during a meeting with NRC staff on 16 July 1998.

t

7 examines potential severe accidents affecting any of the Harris fuel pools, including pools C and D.

25. The NRC examined severe reactor accidents in its Final Environmental Statement for the Harris plant.i6 Site-specific consequence modelling was performed by the NRC for hypothetical accidents that released as much as 82 percent of the inventory of cesium isotopes in the reactor core. CP&L has submitted to the NRC an Individual Plant Examination (IPE) for the Harris plant.17 In addition, CP&L has submitted a similar analysis (an IPEEE) for " external" initiating events.18 The IPE and IPEEE studies examined the potential for severe reactor accidents that could release substantial amounts ofradioactivity.

26. In the absence of similar studies for the Harris pools, one must perform scoping calculations to indicate the degree of hazard posed by spent fuel storage at Harris. The degree of hazard is important when one considers the relevance of a safety issue to a determination of"no significant hazards". Ifpreliminary evidence about a safety issue suggests the potential for accidents with either high probability or large consequences, then the NRC staff should not make a determination of"no significant hazards"

27. The radioisotope cesium-137 is one important indicator of the hazard potential posed by a nuclear facility. This isotope has a half-life of 30 years, emits intense gamma radiation, and is released comparatively readily during severe accidents. The 1986 Chernobyl accident released about 90,000 TBq (27 kg) of cesium-137 to the atmosphere, which accounted for most of the offsite radiation exposure atttibutable to that accident.

Official estimates indicate that this exposure will cause 50-100 thousand extra cancer fatalities worldwide over the next 70 years.19

28. The core of the Harris reactor contains 157 PWR fuel assemblies. At shutdown, this core contains about 155,000 TBq (47 kg) of cesium-137.20 When a spent fuel assembly is ,

discharged from the reactor, it will contain more cesium-137 than the average assembly at shutdown. CP&L plans an eventual, aggregate capacity in the Harris pools of 3,080

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j PWR assemblies and 5,304 BWR assemblies. Note that the cesium-137 content in each l BWR assembly will be about one quarter the cesium-137 content in each PWR assembly, 16NRC, Final Enviromnental Statement related to the ope,ation of Shearon Harris Nuclear Power Plant, Ur.its I and 2, NUREG-0972, October 1983. .

j 17CP&L, Shearon Harris Nuclear Power Plant, Individual Plant Examination Submittal, Final Report,31 August 1993.

Is CP&L, Shearon Harris Nuclear Power Plant Unit No.1, Individual Plant Examination for Extemal Events Submittal, June 1995.

I 19Allan S Krass, Consequences of the Chemobyl Accident (Cambridge, Massachusetts: Institute for l Resource and Security Studies, December 1991).

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20NRC, Final Environmental Statement, page 5-50.

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8 if both assemblies have been discharged for an equal period.21 After discharge, the content of cesium-137 in a fuel assembly will decay exponentially with a half-life of 30 years.

29. As a simplified illustration, assume that all fuel assemblies in the Harris pools have been discharged for an equal period. Further assume that all four pools are full and contain 3,080 PWR assemblies and 5,304 BWR assemblies. The pools will then contain as much cesium-137 as 4,406 PWR assemblies. (3,080 + 5,304 x 1/4 = 4,406) Note that 4,406 PWR assemblies represent 28 cores of the Harris reactor.

30. If an accident can be postulated that releases to the enviromnent a significant fraction of the cesium-137 in the Harris pools, then it is clear that the consequences of this accident would be large. The offsite radiation exposure could be an order of magnitude larger than the exposure from the Chemobyl accident. Activation ofpools C and D could lead to an accident which creates offsite radiation exposure as much as two times higher than the exposure that would arise from a similar accident involving only pools A and B.

H. Loss of Water from Spent Fuel Pools at Harris

31. Loss of water from one or more of the Harris pools could initiate a release to the environment of a significant fraction of the cesium-137 in the pools. This potential exists because the cladding of PWR or BWR fuel is a zirconium alloy which can react exothermically with air or steam. Thus, if the water in a fuel pool is removed and the fuel 4 is partially or totally uncovered, one must be concemed abo"t the possibility of a I runaway air-zirconimn or steam-zirconium reaction. Such a reaction could release l cesium-137 and other radioisotopes from affected fuel into the fuel building. That building was not designed to contain radioisotopes released during a vigorous exothermic l reaction in the pools, and it can be assumed that most of the volatile radioisotopes entering the building from the affected fuel would be released from the building as an atmospheric plume.

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32. Several reports prepared by or for the NRC have examined the conditions under which a runaway zirconium reaction might occur.22 However, these reports have concentrated almost entirely on a postulated condition ofinstantaneous, complete loss of water from a pool. Such a condition is unrealistic in any scenario which preserves the configuration of the spent fuel racks. If water is lost by drainage or evaporation and no makeup occurs, then complete loss of water will always be preceded by partial 21 The ratio of one quaner derives from the parameters shown in the license amendment application, Enclosure 7, page 515.

22 Relevant repons include: V L Sailor et al, Severe Accidents in Spent Fuel Pools in Suppon of Generic Safety Issue 82, NUREG/CR-4982, July 1987; E D Throm, Regulatory Analysis for the Resolution of Generic Issue 82, "Beyond Design Basis Accidents in Spent Fuel Pools", NUREG-1353, April 1989; and R l J Travis et al, A Safety and Regulatory Assessment of Generic BWR and PWR Permanently Shutdown Nuclear Power Plants, NUREG/CR-6451, August 1997. 1

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uncovering of the fuel. If makeup is considered, the water level could fall, rise or remain L static for long periods.

33. Partial uncovering of the fuel will often be a more severe condition than complete loss 'of water because, during partial uncovering, convective heat loss is suppressed by the residual 4er at the base of the fuel assemblies. As a result, longer-discharged fuel with a lower hLt output may undergo a runaway steam-zirconium reaction daing partial ,

l uncovering while it would not undergo a runaway air-zirconium reaction if the pool were instantaneously emptied.

34. I am aware of only one instance in which reports produced by or for the NRC address the hazard posed by panial uncovering, namely in a repon prepared for the NRC by Sandia Laboratories and published in 1979.D Part of this report did address a situation of partial uncoveri' 3, but used a crude heat transfer model and neglected to consider the

. onset of a steam-zirconium reaction. Nevertheless, the report found (page 76) that

......an incomplete drainage can potentially cause a more severe heatup problem than a complete drainage, if the residual water remains near the baseplates". A portion of the 1979 Sandia repon is provided here as Attachment E. An internal NRC memo mentions the consideration of partial uncovering in the 1979 Sandia repon.2% Otherwise,it appears that the NRC has ignored the hazard posed by panial uncovering. This hazard was not reflected in the regulatory analysis whereby the NRC ourportedly resolved Generic Issue 8 2.25

35. In a situation of falling water level, a fuel assembly might first undergo a runaway steam-zirconium reaction, then switch to an air-zirconium reaction as water falls below the base of the rack and convective air flow is established. In this manner, a runaway air-zirconium reaction could occur in a fuel assembly that is too long-discharged (and therefore produces too little heat) to suffer such a reaction in the event ofinstantaneous, complete loss of water, Conversely, a rising water level could precipitate a runaway steam-zirconium reaction in a fuel assembly that had previously been completely uncovered but had not necessarily suffered a runaway air zirconium reaction while in that condition. The latter point is highly significant in the context of emergency measures to recover control of a pool which has experienced water loss. Inappropriate addition of i

water to a pool could exacerbate the accident.

36. The NRC's failure to consider partial uncovering of fuel should be borne in mind when one leviews NRC-sponsored reports that purpon to address the hazard posed by water loss from a fuel pool. This hazard should be re-analyzed through detailed modelling. The modelling should consider both partial and complete uncovering and the U Allan S Benjamin et al, Spent Fuel Heatup Following Loss of Water Laring Storage, NUREG/CR-0649, March 1979.

24 Internal NRC Memorandum from J T Han to M Silberberg, " Response to a NRR request to review SNL studies regarding spent fuel heatup and burning following loss of water in storage pool",21 May 1984.

25 E D Throm, op cit.

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10 transition from one of these states to the other. Also, the modelling should cover: (1) thermal radiation, conduction, and steam or air convection; (2) air-zirconium and steam-zirconium reactions; (3) variations along the fuel rod axis; and (4) radial variations within a mpresentative fuel rod, including effects of the pellet-cladding gap. Experiments will probably be required to suppor't and validate the modelling.  ;

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37. Until the problem of water loss is re-analyzed in this manner, there is no basis for j determining when fuel has been discharged for a sufficiently long period thAt it will not suffer a runaway zirconium reaction in the event of water loss. If the problem were to be properly analyzed through validated models, such a determination could be made within some margin of error, but the determination should consider site-specific factors. For example, the detailed design of a rack might be an important site-specific factor.

38. No determination of this kind has been made for pools C and D at Harris, nor does the methodology now exist to make such a determination. In any case, there is nothing in the license amendment application and its proposed modifications to the Harris Technical Specifications which prohibits the placing of freshly discharged fuel in pools C and D.

Reports previously prepared for the NRC concede that freshly discharged fuel can ,

experience a runaway air-zirconium reaction in the event of complete water loss.

39. A variety of events, alone or in combination, could lead to partial or complete uncovering of spent fuel in the Harris pools. This class of events should be subjected to the kind of systematic analysis that is performed in an IPE and an IPEEE. Relevant i events include: (1) an earthquake, cask drop, aircraft crash, human error, equipment failure or sabotage event that leads to direct leakage from the pools; (2) siphoning of water from the pools through accident or malice; (3) intenuption of pool cooling, leading to pool boiling and loss of water by evaporation; and (4) loss of water from active pools l into adjacent pools or canals that have been gated off and drained. Interactions with the Harris reactor shouid be considered. For exauple, a reactor accident might release radioactivity that precludes personnel access to the plant for purposes of maintaining or restoring pool cooling.

I. Increased Probability or Consequences of Accidents Previously Evaluated

40. The Federal Register notice of this license amendment application claims that the

. probability of a spent fuel assembly drop or a misloaded fuel assembly is not significantly  !

' increased if the license amendment is approved and pools C and D are activated. This claim is false, because activation of pools C and D will roughly double the total number l of fuel handling operations to be conducted at Harris. Assuming that the general nature l of fuel handling operations continues as before, the probability of a fuel assembly drop or misloaded fuel assembly, integrated over the entire period of the Harris operating license, will increase significantly, by a factor of two. This point has been made by David 4 Lochbaum of the Union of Concemed Scientists, in a 22 January 1999 letter to the NRC Commissioners. A copy of his letteris provided here as Attachment F. If probability is integrated over the remaining period of the Harris operating license, rather than over its total duration, then activation of pools C and D will more than double the probability of a fuel assembly drop or a misloaded fuel assembly, i

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41. A spent fuel assembly drop or a misloaded fuel assembly are members of a broader class of accidents that could arise during the movement of fuel from other CP&L stations to Harris, and during fuel movement within Harris. This class of accidents will include design-basis accidents and severe accidents. Assuming that the general nature of fuel movement continues as before, the probability of accidents in this class, integrated over the entire period of the Harris operating license, will double if pools C and D are activated. Ifintegrated over the remaining period of the operating license, the probability will more than double.

42. The PWR racks in pocls C and D will be safe against criticality for a comparatively narrow range of fuel enrichment and burnup. Tnus, assuming that the general nature of l fuel movement continues as before, the probability of a criticality accident will be significantly increased if pools C and D are activated. This probability will increase on a per-movement basis, so it will more than double when integrated over the entire period of the Harris operating license. The consequences of a criticality accident may also be significantly increased.

43. Activation of paols C and D will add to the electrical load and CCW heat load of existing Harris systems. It will also add to the burden of work on the Harris operators.

These effects will increase the probability of two categories of design-basis or severe accidents. First, they will significantly increase the probability of accidents associated with the Harris reactor, because the reactor's CCW and electrical systems and its operators will be under greater stress. Second, they will significantly increase the probability of accidents at the Harris pools that are attributable to interruptions in cooling and electricity supply and to increased operator stress. Also, the inability of cooling piping at pools C and D to meet ASME code requirements could significantly increase the probability of design. basis or severe accidents at these pools.

44. As mentioned in paragraph 24 above, to my knowledge there has been no site-specific analysis of severe accidents affecting eny of the Harris pools. To the extent that such accidents have been previously evaluated, their consequences will be significantly increased by the activation of pools C and D. The fuel storage capacity of these pools will roughly double the storage capacity at Harris, creating the potential for a doubled inventory of radioactivity. Severe accidents could affect some or all of the Harris pools.

As I have discussed in paragraph 30 above, the potential doubling of radioactivity in the pools could significantly increase the consequences of severe accidents.

J. Possibility of New or Different Kinds of Accident from any Accident Previously Evaluated

45. To my knowledge, there has been no site-specific evaluation of the probability or consequences of :: vere accidents at pools A and B at Harris. A variety of severe accidents are possible and should be subjected to the kind of systematic analysis that is performed in an IPE and IPEEE. The NRC has performed evaluations of accidents involving loss of water from fuel pools, generically and for sites other than Harris.

l. .

12 -

However, these evaluations are seriously deficient because they failed to consider partial uncovering of fuel. To summarize, at pools A and B there exists the possibility of new or different kinds of accident from any accident previously evaluated. The same possibility will exist at pools C and D if these are activated.

46. Provision of electrical power, including power from emergency diesel generators, and CCW service ' rom the existing Harris plant to pools C and D could introduce the potential for design-basis or severe accidents that are new or different from any accident previously considered. The IPE and IPEEE studies performed for Harris did not address the provision of electrical power and CCW service to pools C and D. As an example of the potential for new or different accidents, the need to provide cooling to pools C and D will place increased stress on the CCW system, the emergency diesel generators, and the plant operators during a design-basis LOCA.

47. Severe accidents at some or all of the Harns pools could lead to offsite radiation exposure an order of magnitude larger than the exposure from the Chernobyl accident.

Activation of pools C and D ccaid significar.dy increase both the probability and consequences of such acciden.:s. Thus, CP&L's proposed license amendment poses a j "significant hazard" by any reannable definition of that term. j J. Significant Reductions in Margins of Safety.

48. Activation ofpools C and D will create an additional heat load on the existing CCW system. CP&L proposes to meet this load in the short term by exploiting the margin in the CCW system. In my professional opinion, the reduction in the CCW safety margin caused by the increased heat load is significant. Both the NRC and CP&L have also recognized that increasing tne heat load on the CCW system constitutes an unreviewed safety question. The safety margin will be especially reduced if, during a LOCA, the operators must divert water from the RHR to the spent fuel pools. This will increase stress on the operators and create opportunities for human error.

49. As pools C and D become filled and the reactor receives a power uprate, the load on the CCW system will increase further. CP&L offers no assurance that the present margin of safety will be restored by upgrading the CCW system to accommodate these burdens.

50. CP&L proposes to ac&.te pools C and D using cooling systems that will not satisfy ASMd code requirements. This action could potentially cause a significant reduction in margins of safet" for pool cooling. CP&L's Altemative Plan has not been subjected to any public scruuny or rigorous review. It deserves, at the least, thorough consideration at a 1: censing hearing before the 1. 'nse amendment is issued.

n, .

13 .

51. CP&L proposes to provide electrical service to pools C and D from the exisung (Unit

1) electrical system at Harris, having rejected the option of dedicated emergency diesel generators to serve pools C and D. The existing diesel generators already serve the safety systems in Unit I and spent fuel storage pools A and B. By adding ,nools C and D to the lordi carried by the Unit I diesel generators, CP&L would add stress on the diesel generators and on the plant operators. In the event of a loss of offsite power, these effects could significantly reduce the margin of safety at the Harris reactor and the fuel pools.

L Environmental Review

52. As discussed above, the original design of the Shearon Harris plant called for coo'bg of spent fuel pools C and D by the Unit 2 CCW system. The FEIS for the operating license presumably based its conclusions on this design. I have seen no analysis by the NRC Staff, either in the 1983 FEIS or in a subsequent Environmental Impact Statement ct Environmental Assessment, of the environmental impacts of altering the Shearon Harris design to provide for cooling of pools C and D by the Unit 1 CCW-system.

M. Conclusions i

53. From the preliminary evidence presented by the NRC and CP&L, I conclude that operation of the Shearon Harris plant in accordance with the license amendment proposed by CP&L will violate all three of the NRC's conditions for a de'.ermination of"no significant hazards." Therefore, the NRC staff should reverse its position and should determine that CP&L's license amendment request does D9.1 involve no significant hazards consideration.

I 54. The proposed license amendment raises serious safety concerns which deserve prior consideration at a licensing hearing.

I declare, under penalty of perjury, that the foregoing facts provided in my Declaration are true and correct to the best of my knowledge and belief, and that the opinions expressed herein are based on my best professionaljudgment. 1 l

Executed on 12 February 1999.

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, Gordon Thompson i

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I .

ATTACHMENT A INSTITUTE FOR RESOURCE AND SECURITY STUDIES Curriculum Vitae:

GORDON R. THOMPSON December 1996  ;

Professional expertise I

Consulting technical and policy analyst in the fields of energy, environment, l sustainable development, and international security.

Education

• Ph.D. in applied mathematics, Oxford University (Balliol College),1973.

• B.E. in mechanical engineering, University of New South Wales, Sydney, Australia,1967.

• B.Sc. in mathematics & physics, University of New South Wales,1966.

Current appointment

• Executive director, Institute for Resource & Security Studies (IRSS),

Cambridge, MA.

Plphet sponsors and tasks (selected)

• Environmental School, Clark University, Worcester, MA,1996:

session leader at the Summer Institute, " Local Perspectives on a Global Environment".

• Nuclear Free Local Authorities, UK,1996: review of the safety of high level radioactive waste storage at the Sellafield complex.

• Greenpeace Germany, Hamburg,1995-1996: a study on war, terrorism and nuclear power plants.

• HKH Foundation, New York, and Winston Foundation for World Peace, Washington, DC,1994-1996: studies and workshops on preventive action and its role in US national security planning.

• Carnegie Corporation of New York, Winston Foundation for World Peace, Washington, DC, and others,1995: collaboration with the ]

Organization for Security and Cooperation in Europe to facilitate I

improved coordination of activities and exchange of knowledge in the field of conflict management.

• World Bank,1993-1994: a study on management of data describing the performance of projects funded by the Global Environment Facility (joint project of IRSS and Clark University).

• International Physicians for the Prevention of Nuclear War,1993-1994:

a study on the international control of weapons-usable fissile material.

(

(

l

Curriculum Vitae for Gordon R. Tho:npson i December 1996 i

• Government of Lower Saxony, Hannover, Germany,1993: analysis of standards for radioactive waste disposal. ,

a University of Vienna (using funds supplied by the Austrian i government),1992: review of radioactive waste management at the Dukovany nuclear plant, Czech Republic. i

• Sandia National Laboratories,1992-1993: advice to the US Department of Energy's Office of Foreign Intelligence.

• US Department of Energy and Battelle Pacific Northwest Laboratories, 1991-1992: advice for the Intergovernmental Panel on Climate Change regarding the design of an information system on technologies that can limit greenhouse gas emissions (joint project of IRSS, Clark University and the Center for Strategic and International Studies). )

• Winston Foundation for World Peace, Boston, MA, and other funding sources,1992-1993: development and publication of recommendations for strengthening the International Atomic Energy Agency. f

• MacArthur Foundation, Chicago, IL, W. Alton Jones Foundation, Charlottesville, VA, and other funding sources,1984-1993: policy analysis and public education on a " global approach" to arms control and disarmament.

• Energy Research Foundation, Columbia, SC, and Peace Development ,

Fund, Amherst, MA, 1988-1992: review of the US government's {

tritium production (for nuclear weapons) and its implications.

• Coalition of En7ironmental Groups, Toronto, Ontario (using funds supplied by Ontario Hydro under the direction of the Ontario government),1990-1993: coordination and conduct of analysis and preparation of testimony on accident risk of nuclear power plants.

• Greenpeace International, Amsterdam, Netherlands, 1988-1990:

review of probabilistic risk assessment for nuclear power plants.

• Bellerive Foundation, Geneva, Switzerland,1989-1990: planning for a June 1990 colloquium on disarmament and editing of proceedings.

• Der Research Institute, Harrow, Ontario,1989-1990: analysis of regulatory response to boiling-water reactor accident potential.

• Winston Foundation for World Peace, Boston, MA, and other funding sources,1988-1989: analysis of future options for NATO (joint project of IRSS and the Institute for Peace and International Security).

• Nevada Nuclear Waste Project Office, Carson City, NV (via Clark University, Worcester, MA),1989-1990: analyses of risk aspects of radioactive waste management and disposal.

• Ontario Nuclear Safety Review (conducted by the Ontario government), Toronto, Ontario,1987: review of safety aspects of CANDU reactors.

• Washington Department of Emlogy, Olympia, WA,1987: analysis of risk aspects of a proposed radmetive waste repository at Hanford.

Page 2

Curriculum Vitae for Gordon R. Thompson December 1996 Natural Resources Defense Council, Washington, DC,1986-1987:

pri.o ration of testimony on hazards of the Savannah River Plant.

Lakes Environmental Association, Bridgton, ME,1986: analysis of federal regulations for disposal of radioactive waste.

Greenpeace Germany, Hamburg,1986: participation in an international study on the hazards of nudear power plants.

Three Mile Island Public Health Fund, Philadelphia, PA,1983-1989:

studies related to the Three Mile Island nudear plant.

• Attorney General, Commonwealth of Massachusetts, Boston, MA, 1984-1989: analyses of the safety of the Seabrook nuclear plant.

Union of Concerned Scientists, Cambridge, MA,1980-1985: studies on energy demand and supply, nudear arms control, and the safety of nudear installations.

Conservation Law Foundation of New England, Boston, MA,1985:

prepaiatic.n of testimony on cogeneration potential at a Maine papermill.

Town oc Country Planning Association, London, UK,1982-1984:

coordination and conduct of a study on safety and radioactive waste implications of the proposed Sizewell nudear plant.

US Environmental Protection Agency, Washington, DC, 1980-1981:

assessment of the deanup of Three Mile Island Unit 2 nudear plant.

Center for Energy de Environmental Studies, Princeton University, Princeton, NJ, and Solar Energy Research Institute, Golden, CO,1979-1980: studies on the potentials of renewable energy sources. '

• Government of Lower Saxony, Hannover, FRG, 1978-1979:

coordination and conduct of studies on safety aspects of the proposed Gorleben nuclear fuel cyde center. j Other experience (selected)

• Principal investigator, project on " Exploring the Role of ' Sustainable l Cities' in Preventing Climate Disruption", involving IRSS and three other organizations,1990-1991.

• Visiting fellow, Peace Research Centre, Australian National University,1989.

• Principal investigator, Three Mile Island emergency planning study, involving IRSS and Clark University, Worcester, MA, 1987-1989.  ;

• Co-leadership (with Paul Walker) of a study group on nudear weapons prolifen@a, Institute of Politics, Harvard University,1981.  :

• Foundation (with others) of an ecological political movement in Oxford, UK, which contested the 1979 Parliamentary election.

• Conduct of cross-examination and presentation of evidence, on behalf of the Political Ecology Research Group, at the 1977 Public Inquiry into

_ proposed expansion of the reprocessing plant at Windscale, UK.

Page 3  :

Curriculum Vitae for Gordon R. Thompson December 1996

• Conduct of research on plasma theory (while a PhD candidate), as an associate staff member, Culham Laboratory, UK Atomic Energy Authority, 1 % 9-1973.

• Service as a design engineer on coal-fired plants, New South Wales Electricity Commission, Sydney, Australia,1968.

Publications (selected)

• ' Safety of the Storage of Liquid High-Level Waste at Sellafield (with Peter Taylor), Nuclear Free Local Authorities, UK, November 1996.

• Assembling Evidence on the Effectiveness of Preventive Actions, their Benefits, and their Costs: A Guide for Preparation of Evidence, Version 1.0, IRSS, Cambridge, MA, August 1996.

• War, Terrorism and Nuclear Power Plants, Working Paper No.165, Peace Research Centre, Australian National University, Canberra, October 1996.

• "The Potential for Cooperation by the OSCE and Non Governmental Actors on Conflict Management" (with Paula Gutlove), Helsinki Monitor, Volume 6 (1995), Number 3.

• " Potential Characteristics of Severe Reactor Acddents at Nudear Plants", " Monitoring and Modelling Atmospheric Dispersion of Radioactivity Following a Reactor Accident" (with Richard Sclove, Ulrike Fink and Peter Taylor)," Safety Status of Nudear Reactors and Classification of Emergency Action Levels", and "The Use of Probabilistic Risk Assessmerit in Emergency Response Planning for Nudear Power Plant Accidents" (with Robert Goble), in D. Golding, J.

X. Kasperson and R. E. Kasperson (eds), Preparing for Nuclear Power Plant Accidents, Westview Press, Boulder, CO,1995.

• A Data Manager for the Global Environment Facility (with Robert Goble), Environment Department, The World Bank, June 1994.

• Preventive Diplomacy and National Security (with Paula Gutlove),

Winston Foundation for World Peace, Washington, DC, May 1994.

• Opportunities for International Control of Weapons-Usable Fissile Material, ENWE Paper #1, International Physicians for the Prevention of Nudear War, Cambridge, MA, January 1994.

• "Artide III and IAEA Safeguards", in F. Barnaby and P. Ingram (eds),

Strengthening the Non-Proliferation Regime, Oxford Research Group, Oxford, UK, Decemtier 1993.

• Risk !mplications of Potential New Nuclear Plants in Ontario (prepared with the help of eight consultants), a report for the Coalition of Environmental Groups, Toronto, submitted to the Ontario Environmental Assessment Board, November 1992 (3 volumes).

• Strengthening the International Atomic Energy Agency, Working Paper No. 6, IRSS, Cambridge, MA, September 1992.

Page 4

r Curriculum Vitae for Gordon R. Thompson December 1996

• Drdgn of an Information System on Technologies that can Limit Greenhouse Gas Emissions (with Robert Goble and F. Scott Bush),

Center for Strategic and International Studies, Washington, DC, May 1992.

• Managing Nuclear Accidents: A Model Emergency Response Plan for

{

Power Plants and Communities (with six other authors), Westview Press, Boulder, CO,1992.

• "Let's X-out the K" (with Steven C. Sholly), Bulletin of the Atomic Scientists, March 1992, pp 14-15.

• "A Worldwide Programme for Controlling Fissile Material", and "A Global Strategy for Nuclear Arms Control", in F. Barnaby (ed),

Plutonium and Security, Macmillan Press, UK,1992.

• No Restart for K Reactor (with Steven C. Sholly), Working Paper No.

4, IRSS, Cambridge, MA, October 1991.

• Regulatory Response to the Potential for Reactor Accidents: The Example of Boiling-Water Reactors, Working Paper No. 3, IRSS, Cambridge, MA, February 1991. I e Peace by Piece: New Options for International Arms Control and Disarmament, Working Paper No.1, IRSS, Cambridge, MA, January 1991.

• Developing Practical Measures to Preoent Climate Disruption (with Robert Goble), CENTED Research Report No. 6, Clark University, Worcester, MA, August 1990.

• " Treaty a Useful Relic", Bulletin of the Atomic Scientists, July / August 1990, pp 32-33.

• " Practical Steps for the 1990s", in Sadruddin Aga Khan (ed), Non-Proliferation in a Disarming World, Proceedings of the Groupe de Bellerive's 6th International Colloquium, Bellerive Foundation,  !

Geneva, Switzerland,1990.

• A Global Approach to Controlling Nuclear Weapons, Occasional Paper published by the Institute for Resource and Security Studies, October 1989.

• IAEA Safety Targets and Probabilistic Risk Assessment (with three t other authors), Greenpeace International, Amsterdam, August 1989.

• New Directions for NATO (with Paul Walker and Pam Solo), .

published jointly by IRSS and the Institute for Peace and International Security (both of Cambridge, MA), December 1988.

• " Verifying a Halt to the Nuclear Arms Race", in F. Barnaby (ed), A i Handbook of Verification Procedures, Macmillan Press, UK,1990.

• " Verification of a Cutoff in the Production of Fissile Material", in F.

Barnaby (ed), A Handbook of Verification Procedures, Macmillan Press, UK,1990.

1 Page 5

]

Curriculum Vitee for Gordon R. Thompson December 1996

\

" Severe Accident Potential of CANDU Reactors," Consultant's Report in The Safety of Ontario's Nuclear Power Reactors, Ontario Nuclear Safety Review, Toronto, February 1988.

• Nuclear-Free Zones (edited with David Pitt), Croom Helm Ltd, Beckenham, UK,1987.

Risk Assessment Review For the Socioeconomic Impact Assessment of the Proposed High-Level Nuclear Waste Repository at Hanford Site, Washington (edited; written with five other authors), prepared for the Washington Department of Ecology, December 1987.

• The Nuclear Freeze Revisited (written with Andrew Haines), Nuclear Freeze and Arms Control Research Project, Bristol, UK, November 1986. Variants of the same paper have appeared as Working Paper No.18, Peace Research Centre, Australian National University, Canberra, February 1987, and in ADIU Report, University of Sussex, Brighton, UK, Jan/Feb 1987, pp 6-9.

• International Nuclear Reactor Hazard Study (with fifteen other authors), Greenpeace, Hamburg, Federal Republic of Germany (2 volumes), September 1986.

• "What happened at Reactor Four"_ (the Chernobyl reactor accident), Bulletin of the Atomic Scientists, August / September 1986, pp 26-31.

• The Source Term Debate: A Report by the Union of Concerned Scientists (with Steven C. Sholly), Union of Concerned Scientists, Cambridge, MA, January 1986.

• " Checks on the spread" (a review of three books on nuclear proliferation), Nature,14 November 1985, pp 127-128.

• Editing of Perspectives on Proliferation, Volume I, August 1985, published by the Proliferation Reform Project, IRSS.

• "A Turning Point for the NPT ?", ADIU Report, University of Sussex, Brighton, UK, Nov/Dec 1984, pp 1-4.

. " Energy Economics", in J. Dennis (ed), The Nuclear Almanac, Addison-Wesley, Reading, MA,1984.

I

• "The Genesis of Nuclear Power", in J. Tirman (ed), The Militarization of High Technology, Ballinger, Cambridge, MA,1984.

• A Second Chance: New Hampshire's Electricity Future as a Model for the Nation (with Linzee Weld), Union of Concerned Scientists, j Cambridge, MA,1983.

• Safety and Waste Management Implications of the Sizewell PWR (prepared with the help of six consultants), a report to the Town &

Country Planning Association, London, UK,1983.

• Utility-Scale Electrical Storage in the USA: The Prospects of Pumped Hydro, Compressed Air, and Batteries, Princeton University report PU/ CEES #120,1981.

c Page 6

I Curriculum Vitae for Gordon R. Thompsan December 1996

• The Prospects for Wind and Wave Power in North America, Princeton University report PU/ CEES # 117,1981.

=

Hydroelectric Power in the USA: Evolving to Meet New Needs, Princeton University report PU/ CEES # 115,1981.

Editing and part authorship of " Potential Accidents & Their Effects",

Chapter III of Report of the Gorleben International Review, published in German by the Government of Lower Saxony, FRG, 1979--Chapter III available in English from the Political Ecology Research Group, Oxford, UK.

=

A Study of the Consequences to the Public of a Severe Accident at a Commercial FBR located at Kalkar, West Germany, Political Ecology Research Group report RR-1,1978.

Expert presentations and testimony (selected)

Center for Russian Environmental Policy, Moscow,1996: presentation at a forum in parallel with the G-7 Nuclear Safety Summit.

Lacey Township Zoning Board, New Jersey,1995: testimony regarding radioactive waste management.

Ontario Court of Justice, Toronto, Ontario,1993: testimony regarding Canada's Nuclear Liability Act.

Oxford Research Group, seminar on "The Plutonium Legacy", Rhodes House, Oxford, UK,1993: presentation on nuclear safeguards.

Defense Nuclear Facilities Safety Board, Washington, DC,1991:

testimony regarding the proposed restart of K-reactor, Savannah River Site.

Conference to consider amending the Partial Test Ban Treaty, United Nations, New York,1991: presentation on a global approach to arms control and disarmament.

• US Department of Energy, her-Ug on draft EIS for new production reactor capacity, Columbia, SC,1991: presentation on tritium need and implications of tritium production options.

• Society for Risk Analysis,1990 annual meeting, New Orleans, special session on nuclear emergency planning: presentation on real-time techniques for anticipating emergencies.

Parliamentarians' Global Action,11th Annual Parliamentary Forum, United Nations, Geneva,1990: presentation on the potential for multilateral nuclear arms control.

Advisory Committee on Nuclear Facility Safety, public meeting, Washington, DC,1989: submission on public access to information and' on government accountability.

Peace Research Centre, Australian National University, seminar on

- " Australia and the Fourth Nlrr Review Conference", Canberra,1989:

proposal of a universal nuclear weapons non-proliferation regime.

Page 7 l

Curriculum Vitae for Gordon R. Thompscn December 1996

• Carnegie Endowment for International Peace, Conference on "Nudear Non-Proliferation and the Role of Private Organizations",

Washington, DC,1989; options for reform of the non-proliferation regime.

• US Department of Energy, EIS scoping hearing, Columbia, SC,1988:

appropriate scope of an EIS for new production reactor capacity.

• International Physicians for the Prevention of Nudear War,6th and 7th Annual Congresses; Koln, FRG,1986 and Moscow, USSR,1987:

relationships between nuclear power and the threat of nudear war.

• County Council, Richland County, SC,1987: implications of severe reactor accidents at the Savannah River Plant.

• Maine Land Use Regulation Commission,1985: cogeneration potential at facilities of Great Northern Paper Company.

• Interfaith Hearings on Nudear Issues, Toronto, Ontario,1984:

options for Canada's nudear trade and Canada's involvement in nudear arms control.

• Sizewell Public Inquiry, UK,1984: safety and radioactive waste implications of the proposed Sizewell nuclear plant.

• New Hampshire Public Utilities Commission,1983: electricity demand and supply options for New Hampshire.

• Atomic Safety de Licensing Board, US Nuclear Regulatory Commission,1983: use of filtered venting at the Indian Point nuclear plants.

• US National Advisory Committee on Oceans and Atmosphere,1982:

implications of ocean disposal of radioactive waste.

• Environmental de Energy Study Conference, US Congress,1982:

implications of' radioactive waste management.

Miscellaneous

• Married, two children.

• Extensive experience in public speaking before professional and lay ,

audiences, and in interviews with print and broadcast journalists.

1

• Author of numerous newspaper, newsletter, and magazine artides and book reviews.

1 Contact information Institute for Resource and Security Studies 27 Ellsworth Avenue, Cambridge, Massachusetts 02139, USA Phone: (617) 491-5177 Fax: (617) 491-6904 E-mail: irsseige. ape.org j l

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March 11, 1998 4 LICENSEE: CAROLINA POWER AND LIGHT COMPANY .

PLANT: SHEARON HARRIS, UNIT 1

SUMMARY

OF MEETING WITH THE CAROLINA-POWER AND LIGHT COMPANY (CP&L)

On March 3,1998, the staff met with represordatives of Carolina Power & Light Company (CP&L) to discuss the Shearon Harris Nuclear Power Plant (SHNPP) 'C' and 'D' spent fuel pools activation project. Enclosure 1 is a list of meeting attendees. Enclosure 2 is a copy of the handout provided at the meeting. The CP&L presentation included background informstien, a discussion of licensing activities, and the project schedule.

Backoround Originally, SHNPP was intended to be a four unit site with four fuel pools (A.B,C, and D) and two Fuel Pool Cooling and Cleanup systems (FPCCS). Although three of the four units were canceled, the construction of all four pools and one of the FPCCS was completed. Also, a portion of the piping for the other FPCCS was Installed. Currently pools 'A' and 'B' are in service and not only store SHNPP fuel, but also store spent fuel from other CP&L plants (Brunswick Units 1& 2, and Robinson). Pools 'C' and 'D' are not in service.

CP&L has determined that pools 'C' and 'D' will be needed to ensure all four units maintain a prudent operating reserve for com ofiloads. According to CP&L, pool 'C'is needed by early 2000 to support fuel shipments from Brunswick and Robinson. In order to pitce pools 'C' and

'D' in service, the FPCCS and pool racking must be completed for pools 'C' and 'D'.

Licensino Activitiej CP&L identified three licensing activities associated with the completion of pools 'C' and 'D'.

The first is a potential unreviewed safety question (USO) associated with the modification of the Unit 1 Component Cooling Water (CCW) System. Although 15e Unit 1 CCW system was not origina!!y designed to cool the FPCCS for pools 'C' and 'D', CP&L has deterrnined that tne Unit 1 CCW system has sufficient margin to accept the 'C' and 'D' FPCCS lead. The original design was for the Unit 1 CCW to cool the FPCCS for pools 'A' and 'B', and for the Unit 2 CCW system to cool the FPCCS for pools 'C' and 'D'. 9e staff asked several questions about the spent fuel pool, the FPCCS, and CCW system y he staff also inquired about SHNPP fuel handling practices.

The second licensing activity discusseJ involved piping certification for the 'C' and 'D' FPCCS.

A portion of the piping for the 'C' and 'D' FPCCS is already insta!!ed, with some embedded in concrete, making approximately 14 field welds inoceessible. CP&L inadvertently dispesed of the piping certification records for the instafled piping, which makes it unable to demonstrate that the piping satisfies the design requirements of American Society of Mechanical Engineers (ASME) Code Section 111. CP&L stated that it intends to request relief from ASME Code Section Ill. The staff stated that a relief request from the requirements of ASME Code

3014152 02 JUL-10-1??B 12:43 t!5 NRC'DIV RX PROJ 301c1:2102 P.03 4.

-2 Section lit would not be appropriate. The staff recommended that CP&L propose an altomative method, as allowed by 10 CFR 50.55a, that provides an acceptable level of safety and quality.

CP&L agreed with the stars comments and stated that a relief request was not the appropriate terminology for its request. CP&L stated that it intands to propose a piping certification plan, wnich includes tests and inspections, as an alternative method to the requirements of ASME Code Section lit.

CP&L also intends to submit a Tech'nical Spec,L%n (TS) change for high density racks in pools 'C' and 'D'. The TS change would modify SHNPP spent fuel capacity.

Schedule CP&L stated that the TS change and the piping certification plan wPJ be ready for submittal this summer, and the CCW USQ will be ready by fall. Due to the complex nature of this review, the staff recommended that CP&L make one complete submittal that includes all three licensing 3

activities as oppose to three seperate submsttals. The staff also recommended that CP&L may want to meet with the staff again in the summer to discuss, in more detail, the TS change and ",

the piping cartrficabon plan. ,

CP&L agreed with the staffs recommendations, and intends to submit one amendment encompassing all three licensing activrties in the fall. CP&L stated that approval is needed by the end of 1999.

.9 Scott C. Flanders, Project Manager Project Directorate !!-1 Division of Reactor Projects - t/II Office of Nuclear Reactor Regulation Docket No. 50-400 cc w/ enclosures:

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• i l' ATTACHMENT E l

NUREG/CR-0649 SAND 77-1371 R-3 l

• P l

SPENT FUEL HEATUP FOLLOWING LOSS OF WATER DURING STORAGE l

Allan S. Benjamin David J. McCloskey Dana A. Powers Stephen A. Dupree i

Date Published: March 1979

! i Sandia Laboratories Albuquerque, New Mexico 87185 operated by Sandia Corporation for the U.S. Departnent of Energy i

Prepared for UcS. Nuclear Regulatory Com...fssior l washington, DC 20555 Under Interagency Agreement DOE 40-550-75 NRC FIN No. A2050 3

5. OTHER CONSIDERATIONS

'5. 1 Effect of Incomplete Drainage Many spent fuel holder designs provide only a single inlet hole for convective flow through each fuel element, located in the baseplate or near the bottom of the holder. If there is a complete pool drainage, the air must circulate down and under the fuel elements before passing through the baseplate inlet hole into the fuel assembly. An incomplete drainage could block this flow and reduce the effectiveness of natural con-vective cooling. Open frame configurations are, of course, exempt from this possibility because the flow does not have to pass through an inlet hole in order to gain proximity to the fuel element.

A detailed analysis of spent fuel heatup in the event of an incomplete drainage has not been undertaken. However, an approximate analysis has been performed to estimate the amount of aggravation that might occur if the water ceased to drain after exposing all but the bottom portion of the fuel elements.

The analysis is included in Appendix B and is based, among other things, upon upper and lower bound estimates of the thermal radiation absorbed by the water from the hot fuel rods above. The temperature distribution along the rods is pre-scrfoed in this analysis according to estimates made of the likely distribution that would occur just prior to the onset c,f self-sustaining clad oxidation. The amount of heat produced above the water level is then determined together with the amount that could be removed by various mechanisms, including water boiling (latent heat) , convection to the steam produced 73 m

by_ boiling- (sensible heat) , radiation to the building, and convection to the air. If the heat removal rate is determined to be larger than the rate of production, then the configura-tion is coolable; if the heat removal rate is smaller than the rate of production, overheating.resulting in clad rupture or melting will occur. 1 The results for a 1-year decay time are presunted in Table VIII. Consider first the case where the drainage un-covers the upper 80 percent of the fuel rods, leaving the lower 20 percent still covered (third column). The heat

. transferred to the remaining water by decay from the im-mersed portions and by radiation from above is 3.6 - 4.9 KW per assembly (line 2c). This implies that about an hour might be required to raise the water tempere.ture to boiling I (assuming all the assemblies produce the same decay heat) and that the water recession rate following the inception of boiling- will be about 10 cm/h (lines 3 and 4) . Meanwhile, the decay heat produced above the water line is about 4.5 K!1 per assembly (line 5), and the capability for removing heat as the clad temperatures approach the lower limit of self-sustaining oxidation is 5.7 - 8.7 KW per assembly (line 6e).

Since the heat removal capability exceeds the heat' production l (line 7), the geometry is temporarily coolable.

'If, however, . the drainage were to uncover the whole length ;

of the rods but still to constrict the flow, either by blocking l the baseplate holes or by not allowing enough space for un-restricted flow in the base region, then the heat production  ;

.would exceed the heat removal capability (line 7, first column) and the clad would overheat. The same situation would event-

- ually occur if, rather than immediately draining to this posi- l tion, the water were to drain part way down the rods and then boil off down to the baseplates over a period of time. Table

-VIII indicates that there is a good chance of overheating, in 74

Table VIII.

• 1 i Estimates of Heat Removal Capability in an

+

Incompletely Drained Pool, One Year Decay Time * ,

1 0.0 0.1 0.2

1. Normalized water level (zw/L) _

2. ljest transferred. to water, per assembly

@b' ~

0.0 0.2 0.6 q$ py decay heat by thermal radiation 0.3 - 1.3 1.2 - 2.6 3.0 - 4.3 4

from above c '. total 0.3 - 1.3 1.4 - 2.8 3.6 - 4.9 0.9 - 1.8 0.7 - 1.0

3. T$me to start boiling (hours) 1.0 - 4.3 0.7 - 3.2 3.5 - 7.0 9.0 -12.2

4. Water surface recession rate (cp/hr) 5, Decay heat produced by spent 5.1 4.9 4.5 fuel above water level, per assembly (KW)

6. Rsmoval of heat produced by apppt fuel above water level, l per ' assembly (KW) :

A. ' by radiation to water 0.3 - 1.3 1.2 - 2.6 3.0 - 4.3

b. by radiation to building 0.0 - 0.9 0.0 - 0.9 0.0 - 0.9

c. by transfer to water vapor 0.2 - 0.8 0.9 - 1.8 2.3 - 3.1

.d . by transfer to air 0.4 0.4 0.4 e, total 0.9 - 3.4 2.5 - 5.7 5.7 - 8.7

7. Heat removal surplus (deficit) ( 4 . 2 ) - (1.7 ) (2.4)-0.8 1.2 - 4.2 per assembly (KW), line 6e pinus line 5.

t

• PWR spent fuel in cylindrical baskets. One year decay time assumed, (e.g., 0.3 - 1.3) give uniformly throughout pool. Numerical ranges lower ap4 upper-bound estimates. See Appendix B. .

\

N w e . ..

75

' fcct, if the water were to recede below the level where the lower 10% of the rods is still immersed.

A comparison of the peak clad temperature rise versus time for PWR spent fuel with a 1-year minimu decay time in a:well-ventilated room is shown in Figure 26. The temperature rise corresponding to an incomplete drainage down to the bottom

.of the rods, calculated-by utilizing the lower-bound radiation estimate, is compared with previous cases for a complete drain-cge with varying baseplate hole sizes. The clad oxidation offect has not been calculated for the case of incomplete drain-age (blocked inlets), because it is believed to be substantially reduced by the unavailability of oxygen within the assembly.

Clearly, a 1-year minimum decay time is not sufficient to preclude overheating for this case.

The approximate method used for bracketing the thermal radiation downward to the water and upward to the building is not considered to be precise enough to allow prediction of the minimum allowable decay time in the event of an incomplete drainage. This problem could be approached by formulating a detailed thermal radiation model to calculate . shape f actors and include the shadowing of radiating surfaces by fuel rods and tie plates. By incorporating this radiation capability into the overall heat transfer models described in Sections 3.3 and 3.4, a credible prediction of the minimum allowable decay time could be obtained. No attempt to do this, however, l l

has been made. .

It. is clear, however, that an incomplete drcinage can )

potentially cause a more severe heatup problem than a complete drainage, if_the residual water level remains near the base-plates. From a practical point of view, it might be possible to-make provisions for either completing the drainage or re-filling the pool, if this should happen. Hcwever, it would

,-76

1900 . - ' ' '

PWR SPENT FUEL IN CYLINDRICAL BASKETS 1809 1-YEAR MINIMUM DECAV TIME .

I499- BLOCKED INLETS-( NO OXlDATION ASSUMED )

OXIDATION IRQ9 - EFFECT FOR N0 i g WATER, il000, Dhole" 8 -

a: -

a NO WATER, D =1.5" hole

$ 800 -

W ,v.:

3 ',' .'

4 un .

4' E

4Q0 -

NO WATER, D = 3. 0" hole

~

290 - NO WATER, D 2 5.0" hole l

t i e i I

0 8 16 24 32 40 48 ,

,.. TIME AFTER POOL DRAINAGE (Hrs)

Figure 70. Estimated Heatup of PWR Spent Fuel With Residual Water Sufficient to Block Flow ,

Inlets, Well-Ventilated Room

id

. .

77

r-~ .

seem that the special problems associated with an incomplete drainage could best be circumvented by modifying the spent fuel holders to include inlet holes at various elevations along the vertical, rather than just at the baseplate level. According to the predictions, these inlet holes would only be required for the bottom 20 percent of- the fuel rod lengt; if the spent fuel were at least a year old. With these additional inlets, the beneficial effect of natural convection would not be cancelled by an incomplete drainage.

5.2 Effect of Surface Crud Iron oxides are known to deposit upon the outside of the fuel pins during normal operation of the reactor, and these deposits are likely to remain on the fuel pins during storage of the spent fuel. Typically, the iron oxide crud buildup on SWR fuel pins is on the order of 25 to 100 microns and in the for;.t of Fe O2 3, whereas the buildup on PWR pins is on the order of only 1 to 5 microns and in the form of Fe34 0 .16 A calcula-tion was made to determine whether a 100 micron Fe 23 O coating on the BWR fuel pins would affect the heatup of these pins during a pool drainage accident, and it was found that the overall effect on the fuel pin temperature was less than one degree.

The question was also raised as to whether some of the crud, which would be contaminated, could be levitated by the air flows produced by natural convection after a pool drainage and thereby produce a health hazard. An analysis of the weight and drag characteristics of iron oxide particles revealed that a BWR fuel assembly having a decay time of 90 days prior to loss of water can produce upward air currents sufficient to levitate a 200-micron sized particle, whereas an assembly allowed to decay for 250 days can levitate a 175-micron sized particle. Since any spallation of the crud would produce particles of roughly the same size as the thickness of the 78~

ATTACHMENT F UNION OF CONCERNED SCIENTISTS January 22,1999 Chairman Shirley A. Jackson Commissioner Nils J. Diaz Commissioner Greta J. Dieus Commissioner Edward McGaffigan, Jr.

Commissioner Jeffrey S. Merrifield United States Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

CURRENT EXAMPLE OF RISK-DEFORMED REGULATION

Dear Chairman and Commissioners:

During the January 11* Commission briefing on risk-informed regulation and during the January 20*

briefing on the proposed reactor oversight process, I expressed our concem that the NRC and the nuclear industry are making risk decisions using incomplete and inaccurate data. As a current example,I ctl!

your attention to the license amendment application dated December 23,1998, by the Carolina Power &

Light Company involving spent fuel storage at the Harris Nuclear Power Plant and the subsequent proposed no significant hazards consideration determination (Federa/ Register: January 13,1999, Vol.

64, No. 8) prepared by the NRC staff.

He licensee and the NRC staff.have improperly downplayed the risk associated with the proposed activity. Their risk characterization is wrong. He licensee should be required to resubmit a corrected application and another FederalRegister notice issued with a corrected proposed no significant hazards consideration determination.

The error involves the determination made by the licensee and endorsed by the staff regarding the affect cf the proposed activity, namely placing storage racks in Spent Fuel Pools 'C' and 'D' at the Harris plant, on the probability of a fuel handling accident. From the Federal Register notice:

"The probability that any of the accidents in the above list [a spent fuel assembly drop in a spet.t fuel pool / loss of spent fuel pool cooling flow / a seismic event / mistoaded fuel assembly) can occur is not significantly affected by the activity itself. ... He probabilities of accidental fuel assembly drops or misloadings are primarily influenced by the methods used to lift and move these loads. The method of handling loads during normal plant operations is not signficantly changed, since the same equipment (i.e., Spent Fuel Handling Machine and tools) and l procedures as those in current use in pools ' A' and 'B' will be used in pools 'C' and 'D.' Since I the methods used to move loads during ; armal operations remain nearly the same as those used previously, there is no signific nt increase in the probability of an accident."  !

I i

Washington Osr.ce: 1616 P Street NW Suite 310 . Washington DC 20036-1495 202 332 0900 . FAX: 202 332 0905 Cambndge Headquarters: Two Brattle Square . Cambridge MA 02238-9105 617 547 5552 . FAX: 617-864-9405 California Office: 2397 Shattuck Avenue Suite 203 . Berkeley CA 94704-1567 510-843-1872 . FAX: 510-843-3785

• thu 3 &OO k _ Ds'6"-

3

. . :, 1 m

January 22,1999 Page 2 of 2 It is precisely this type of" smoke and mirrors" shenanigans that we decried during the briefings.The logic seems proper at face value, but it does not take much effort to show that it is wrong. In Enclosure 1 to the license amendment submittal, the 1A- reported that the total storage capacity ofpools 'A' and

'B' is 3,669 assemblies a wi that the pmposed activity will add 4,715 storage locations in pools 'C' and

'D.* Thus, if the amendment is granted, CP&L will hamfle -pick up and move - about twice as many irradiated fuel assemblies as they will if the amendment is not granted.

Consider for a moment the old game of Russian roulette using a six. chamber revolver loaded with a single bullet. CP&L and the NRC staff would apparently conclude that the probability oflosing the game are not ired whether one or two turns are taken because, after all, the same method and the same equipment se used each turn. Their logic is simply wrong. "Ihe probability of a fuel handling accident at Harris will nearly double if the license amendment request is granted. 'Ihis material fact contradicts the conclusion of the licensee and the staff that there will be "no significant increase in the probability,"

unless doubling the risk is not significant.

Luckily, there's an opportunity to fix the mistake this time. Unfortunately, it's not the first, and probably won't be the last, time this mistake is made. The NRC staff made this same mistake in April 1998 when it allowed the Paducah facility to continue operating with its risk doubled.

We have no intention at this time of forvrily intervening in this Hanis licensing action. We trust that the NRC staff will take the necessary steps t1 bave the licensee fix the fundamental flaw in the licensing amendment request before granting it.

Sincerely, 0Ad0 Cww David A. Loch aum Nuclear Safety Engineer 0

4

I INSTITUTE FOR RESOURCE AND SECURITY STUDHEG 27 Ellsworth Avenue, Cambridge, Massachusetts 02139, USA Phone:(617) 491-5177 Fax:(617) 491-6904 Electronic mail: irss@igc.apc.org i

EXHIBIT 3 RISKS AND ALTERNATIVE OPTIONS ASSOCIATED WITH I SPENT FUEL STORAGE AT THE SHEARON HARRIS NUCLEAR POWER PLANT l

1 i

l A report l prepared for Orange County North Carolina by Gordon Thompson

+

4 February 1999

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. -i-Acknowledgements This report was prepared as part of a program of work by the Institute for Resource and Fa.. arity Studies (IRSS) pursuant to a contract between IRSS and Orange County, rJorth Carolina. The report was written by Gordon Thompson, the executive director of IRSS. l The author acknowledges help with the acquisition of information and documents, from Diane Curran, David Lochbaum, Mary MacDowell and the staff of the NRC public document room in Washington, DC. Paul Thames, county engineer of Orange County, has provided efficient oversight of the j contract between IRSS and Orange County. Paula Gutlove of IRSS has assisted in the preparation of this report. Gordon Thompson is solely responsible for the content of the report.

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1 About the author Gordon Thompson is the executive director of IRSS. He received an undergraduate education in science and mechanical engineering, in Australia. Subsequently, he studied at Oxford University and received from that institution a doctorate of philosophy in mathematics in 1973.

During his professional career, Dr Thompson has performed technical and policy analyses on a range of issues related to international security, energy supply, environmental protection, and the sustainable use of natural resources. Since 1977, a significant part of his work has consisted of technical analyses of safety and environmental issues related to nuclear facilities.

These analyses have been sponsored by a variety of nongovernmental organizations and local, state and national governments, predominantly in north America and western Europe. Dr Thompson has provided expert testimony in legal and regulatory proceedings, and has served on committees advising US government agencies.

AboutIRSS The Institute for Resource and Security Studies is an independent. non-profit corporation. It was founded in 1984 to conduct technical and policy analysis and public education, with the objective of promoting international security and sustainable use of natural resources. IRSS projects always reflect a concern for practical solutions to resource, environment and security problems, and can range from detailed technical studies to preparing educational materials accessible to the public. IRSS actively seeks collaborative relationships with other organizations as it pursues its goals.

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Abstract Orange County, North Carolina, commissioned this report because the licensee of the Shearon Harris nuclear plant has requested an amendment of its operating license. The amendment would permit the activation of two currently unused spent fuel pools at Harris.

This report examines the risks and alternative options associated with spent fuel storage at Harris. The report identifies a potential for severe accidents at the Harris pools. Such accidents could release to the atmosphere an amount of cesium-137 an order of magnitude larger than the release from the 1986 ,

Chernobyl accident. A severe accident at the Harris PWR, with containment failure or bypass, can be expected to initiate a large release from the fuel pools.

Alternative, safer options for spent fuel management are available. These options include dry storage of spent fuel, which is a well-established practice.

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Table of contents

1. Introduction  !

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2. Present status of the Harris nudear plant

3. Proposed activation of fuel pools C and D

4. Types of potential accident at the Harris plant

5. Design-basis pool accidents

6. Severe pool accidents

7. Consequences of potential pool and reactor accidents

8. Alternative options for spent fuel management

9. Addressing risks and alternatives in the regulatory arena

10. Conclusions Appendix A Spent fuel management at the Harris plant Appendix B Potential for severe accidents at the Harris reactor Appendix C Potential for loss of water from the Harris pools Appendix D Potential for exothermic reactions in the Harris pools Appendix E Consequences of a large release of cesium-137 from Harris

L" Risks & alternative options re. spent fuel stor:ge at Harris Page1 -

L Introduction Carolina Power & Light Company (CP&L) requested, in December 1998, an amendment of its operating license for the Shearon Harris nuclear plant. The amendment, if granted by the Nuclear Regulatory Commission (NRC), would permit the activation of two currently unused spent fuel pools at Harris. In January 1999, Orange County commissioned this report, which examines the risks and alternative options associated with spent fuel storage at Harris.

Structure of this report This report has two major components. One component is a main report which is comparatively brief and is intended for a non-specialist audience.

The second component is a set of five appendices. These appendices contain detailed, technical material and citations to technical literature. Unless otherwise indicated, discussion in the main report rests upon the more detailed di-ion in the appendices.

What is spent fuel?

Figure I shows a fuel assembly of the type that is used in the Harris reactor.1 The fuel rods are 12 feet long, and the assembly is 8.4 inches square. After a fuel assembly is discharged from a reactor, it is " spent" in the sense that it can no longer be used to generate power. However, at this point in its life the assembly is much more dangerous than when it entered the reactor. It emits heat and intense radiation, and contains a large inventory of radioactive material.

Remainder of this report The remainder of this main report begins with descriptions of the Harris plant (Section 2) and CP&L's intentions regarding the fuel pools at Harris (Section 3). Then, categories of potential accident at Harris are identified

. (Section 4), followed by descriptions of potential design-basis (Section 5) and severe (Section 6) accidents at the Harris pools. The offsite consequences of potential pool and reactor accidents are addressed in Section 7. Alternative options for spent fuel management are p. resented (Section 8), followed by a discussion of regulatory processes (Section 9). Conclusions are presented in Section 10.

1 Figure 1 is adapted from: A V Nero, A Guidahook to Nuclame Ran< totI Jniversity of California Press,1979, page 79.

Risks & alternative options re. spent fuel storage at Harris Page 2

2. Present status of the Harris nuclear plant The Harris plant features one pressurized-water reactor (PWR). The core of this reactor contains 157 fuel assemblies, with a center-center distance of about 8.5 inches. The Harris plant was to have four units but only the first unit was built. (A unit consists of a reactor, a turbine-generator and associated equipment.) A fuel handling building was built to serve all four units. This building contains four fuel pools (A, B, C, D), a cask loading pool and three fuel transfer canals, all interconnected but separable by gates.

These pools and transfer canals allow spent fuel to be moved around and stored while remaining under water. The water provides cooling and also shields personnel and equipment from the radiation emitted by the fuel.

Shipping casks can carry spent fuel to or from Harris. Casks are loaded and

• unloaded while submerged in the cask loading pool.

Pools A and B Pools A and B contain fuel racks, and are in regular use. CP&L say s that fresh fuel, and spent fuel recently discharged from the Harris reactor, is stored in pool A. Fuel examination and repair are performed in an open space in pool B. At present, pools C and D are flooded but do not contain racks. The cooling and water cleanup systems for pools C and D were never completed.

Currently, pools A and B store spent fuel from the Harris reactor and from CP&L's Brunswick plant and Robinson plant. The Brunswick plant has two boiling-water reactors (BWRs) while the Robinson plant has one PWR.

Shipment of spent fuel from Brunswick and Robinson to Harris is said by -

CP&L to be n~*===ry to allow sufficient capacity in the pools at Brunswick .

and Robinson so that the entire core can be removed from the reactor.

Pools A and B now have a combined, potential capacity of 3,669 fuel assemblies. The center-center distance in the racks in pools A and B is 10.5 inches for PWR fuel and 6.25 inches for BWR fuel. This is a much more {

compact pool storage configuration than was used when nuclear plants first entered service. The United States has no national storage site or repository

[ for spent fuel, so CP&L is currently obliged to store fuel at its plant sites. '

i Compact storage in the existing pools is a comparatively cheap option for on-site storage.

Risks & alternative options re. spent fuel storage et Harris Page 3 .

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3. Proposed activation of fuel pools C and D CP&L seeks an amendment to its operating license so that it can activate pools

, C and D at Harris. _ By activating these pools, CP&L expects to have sufficient storage capacity at its three nuclear plants to accommodate all the spent fuel discharged by the four CP&L reactors (the Harris and Robinson PWRs and the two Brunswick BWRs) through the ends of their current operating licenses.

Capacity and configuration of pools C and D CP&L plans to install racks in pool C in three campaigns (approximately in 2000,2005 and 2014), to create a total capacity in this pool of 3,690 fuel assemblies. Thereafter, CP&L plans to install racks in pool D in two campaigns (approximately in 2016 and at a date to be determined), to create 1,025 spaces. Thus, the ultimate capacity of pools C and D will be 4,715 fuel amammhlies. The center-center distance in the racks used in these pools will be 9.0 inches for PWR fuel and 6.25 inches far BWR fuel. In pool C, the space between the outermost racks and the pool wall will be 1-2 inches.

The PWR racks in pools C and D will have a smaller center-center distance than the racka in pools A and B (9.0 inches instead of 10.5 inches). This highly compact arrangement allows more .'WR fuel to be placed in a given pool area but also has adverse implications for s.Jety.

Cooling and electrical supply for pools C and D The water in a spent fuel pool must be cooled and cleaned. Cooliag is performed by circulating pool water through heat exchangers, where its heat is transferred to a secondary cooling system. At Harris, the secondary cooling system is the component cooling water (CCW) system. When the Harris plant was designed, the intention was that pools C and D would be cooled by the CCW system for Unit 2. Also, electricity would have been supplied to the circulating pumps at pools C and D from the electrical systems of Unit 2.

However, Unit 2 was never built and its CCW and electrical systems do not exist.

CP&L's current plan is to cool pools C and D by completing their partially built cooling systems and connecting those systems to' the Unit 1 CCW system. Electricity will be supplied to pools C and D from the electrical systems of Unit 1. The Unit 1 CCW system already provides cooling to pools l A and B and serves other, important safety functions. For example, the Unit 1 CCW system provides cooling for the residual heat removal (RHR) system and reactor coolant pumps of the Unit I reactor.

Risks & alternative options re. spent fuel storage at Harris Page 4 '

Independent support systems for pools C and D During CP&L's planning for the activation of pools C and D, the company considered the construction of an independent system to cool these pools.

Within that option, CP&L considered the further possibility of providing dedicated emergency diesel generators to meet the electrical needs of pools C and D if normal electricity supply were unavailable. Construction of an independent cooling system for pools C and D, supported by dedicated emergency diesel generators, could provide the level of safety that was associated with the original design concept for Harris. However, CP&L has not proceeded with this option.

Capacity of the Unit 1 CCW system In its present forns, the Unit 1 CCW system cannot absorb the additional heat load that will ultimately arise from activation of pools C and D. Over the first few years of pool use, while the heat load is comparatively small, CP&L proposes to exploit the margin in the Unit 1 CCW system. Subsequently, CP&L intends to upgrade the Unit 1 CCW system so that it can accommodate the full heat load from pools C and D, and can also accommodate an anticipated power uprate for the Unit I reactor.

Safety implications In order to exploit the margin in the existing CCW system so as to cool pools C and D, CP&L may be obliged to require its operators to divert some CCW flow from the RHR heat exchangers during the recirculation phase of a design-basis loss-of-coolant accident (LOCA) event at the Harris reactor. This is a safety issue because, during the recirculation phase of a LOCA, operation of the RHR system is essential to keeping the reactor core and containment in a safe condition. CP&L's exploitation of the margin in the existing CCW system is deemed by CP&L and NRC to constitute an "unreviewed safety question".

Lack of QA documentation Activation of pools C and D will require the completion of their cooling and water cleanup systems, and the connection of their cooling systems to the Unit 1 CCW system. CP&L states that approximately 80 percent of the necessary piping was completed before the second Harris reactor was cancelled. However, some of the quality assurance (QA) documentation for the completed piping is no longer available. Much of the completed piping is embedded in concrete and is therefore difficult or impossible to inspect. To

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Page 5 address this situation, CPacL proposes an " alternative plan" to demonstrate that the previously completed piping and other equipment is adequate for its purpose. Nevertheless, the cooling systems for pools C and D will not satisfy prevailing code requirements.

4. Types of potential accident at the Harris plant Most of the radioactive material at the Harris plant is either in the reactor or in the spent fuel pools. Thus, these locations are of primary concern when one considers the potential for accidents. This report focusses on the potential for accidents in the reactor or the pools. At present, pools C and D at ,

Harris pose no accident potential, because they are unused.

Some potential accidents could cause injury to plant personnel, without causing any offsite effects. Other potential accidents could release radioactive material beyond the plant boundary, causing offsite effects. The radioactive material could be released as an atmospheric plume, or into ground or surface waters. This report focusses on accidents that release an atmospheric plume which travels beyond the plant boundary. Such a plume will contain radioactive material in the form of gases and small particles. As the plume travels downwind, the small particles will be deposited onto land, bodies of water, structures and vegetation.

Design-basis and severe accidents A. nuclear plant is designed to accommodate the effects of a specified set of accidents, known as " design-basis" accidents. If the plant is properly designed and constructed, if its equipment and operators function in the required manner, and if external influences (e.g., earthquakes) do not exceed specified levels, then the offsite effects of a design-basis accident will be small. Design-basis accidents and their anticipated effects are described in a Final Safety Analysis Report (FSAR) prepared and regularly updated by the licensee.

In the early years of the nuclear industry, some people equated design-basis accidents with " credible accidents. However, research and operating experience soon revealed that acddents more severe than the design basis are credible. The first systematic study of the potential for severe accidents was the Reactor Safety Study, completed and published by the NRC in 1975.

" Severe" accidents are conventionally defined as accidents involving substantial damage to fuel, with or without a substantial release of radioactivity to the environment.

The Three Mile Island (TMI) reactor accident of 1979 was a demonstration of the potential for severe accidents. Soon thereafter, the NRC promulgated

Risks & alternative options re. spent fuel storage at Harris Page 6 -

regulations which require an emergency response plan for each nudear plant.

These plans allow for large releases of radioactive material, of the kind that were identified in the Reactor Safety Study. The Chernobyl reactor accident of 1986 further demonstrated the potenti-! for severe accidents. While the TMI accident released a small fraction of the reactor core's inventory of radioactivity, the release fraction during the Chernobyl accident was large.

Since'the TMI accident, the NRC's safety regulation of nudear plants has been guided by a hybrid set of assumptions. Many areas of safety regulation rely upon the assumption that accidents will remain within the design basis.

Other areas, such as emergency response planning, assume that severe

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accidents can occur.

Pool-reactor hieractions At the Harris plant, the reactor and the fuel pools are adjacent, and they share support systems such as the Unit 1 CCW system and the emergency diesel generators. Thus, it is important to understand if an accident at the Harris reactor could accompany, initiate or exacerbate an accident at the Harris pools, or vice versa. The NRC has been slow to examine the potential for safety interactions between reactors and fuel pools. Neither CP&L nor the NRC has assessed the potential for these interactions at Harris.

PRAs and IPEs A discipline known as probabilistic risk assessment (PRA) has been developed to examine the probabilities and consequences of potential accidents at nuclear facilities. PRA techniques are most highly developed in their application to reactor accidents, but can be applied to fuel pool accidents.

Appendix B describes the characteristics, strengths and limitations of PRA.

CP&L has prepared a Level 2, internal-events PRA for the Harris reactor,in the form of an Individual Plant Examination (IPE). Also, CP&L has

. performed a limited assessment of the vulnerability of the Harris reactor to ,

earthquakes and in-plant fires, in the form of an Individual Plant l Ev=mination for External Events (IPEEE).

The Harris IPE and IPEEE could be extended to encompass fuel pool accidents as well as reactor accidents. Such an extension would be logical, because there are various ways in which a severe accident or a design-basis accident at the Harris reactor might accompany, initiate or exacerbate an accident at the .

Harris fuel pools, or vice versa. However, there is no current indication that CP&L will extend the IPE or IPEEE, or will otherwise apply PRA techniques to -

potential accidents at the Harris fuel pools.

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5. Design-basis pociaccidents The Harris PSAR considers twb :ype c' design-basis accident in the Harris fuel pools. One type of accident involves the dropping of a fuel assembly, while the other type involves the dropping of a shipping cask (but not into a fuel pool). In both cases, the PSAR estimates that the release of radioactivity would be platively small. This report does not review the PSAR analysis.

In its license amendment application, CP&L has considered some other potential accidents, including the dropping of a rack or a fuel pool gate.2 CP&L's analysis of these accident scenarios is limited in scope. Accidents of this type may be in an intermediate class of severity, and that potential class deserves further analysis.3 This report focusses on the potential for severe accidents.

It should be noted that the use of pools C.and D at Harris will involve many additional cask, fuel and rack movements. These additional movements will increase the cumulative probability of accidents associated with such movements.

6. Severe pool accidents Spent' fuel is stored in a compact, high-density configuration M pools A and B at Harris. CP&L's proposed activation of pools C and D willinvolve an even higher density of storage. Such high-density configurations inhibit heat loss from the fuelif water is partially or totally lost from a pool. As a result, partial or total loss of water can lead to an exothermic (heat-producing) reaction of the fuel cladding with air or steam. Such a reaction could liberate a large amount of radioactive material from the fuel.

Thus, two questions become important. Hrst, what circumstances could cause a partial or total loss of water? This question is addressed in Appendix C. Second, will an exothermic reaction be initiated if water is lost? That question is addressed in Appendix D.

Potential for loss of water

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A variety of events could cause partial or total loss of water from the Harris j pools. These events deserve the level of analysis that would be provided by a i thorough PRA. Performing a pool accident PRA is beyond the scope of our 2 License amendment'applation, Enclosure 7.

3 A potential accident in this class, which deserves analysis, would involve the placement of a low tumup or 4. c.&ic.s.t PWR assembly in the racks in pools C or D.

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Risks & alternative options re. spent fuel storage at Harris Page 8 present work for Orange County. Here, the focus is on two types of event - a reactor accident, and a sabotage / terrorism event. Consideration of these events demonstrates clearly that loss of water from the Harris pools is a credible accident.

The Harris IPE - prepared by CPdet - examines the potential for severe accidents at the Harris reactor. It identifies a category of severe accidents that would involve failure or bypass of the reactor containment. The IPE estimates the collective probability of accidents in this category to be 1 per 100,000 reactor-years.4 Occurrence of accidents in this category would contaminate the plant with radioactivity, to the point where peonnel access would almost certainly be precluded. Water would then be evaporated from the fuel pools, and fuel would be uncovered after a delay of perhaps 10 days.

A credible sabotage / terrorism event at Harris would involve a group taking control of the fuel handling building, shutting down the pool cooling systems, and siphoning water from the pools. The group would require military skills and equipment to take control of the fuel handling building.

Siphoning water from the pools would be a comparatively easy task. Escape by the group would be difficult but not impossible. The probability of this event cannot be predicted by PRA techniques.

Initiation of exothermic reactions, given water loss Since the late 1970s, the NRC has sponsored and performed a variety of j studies that have examined the outcomes of a loss of water from a fuel pool.

These studies have focussed almost entirely on the instantaneous, total loss .

j of water from a pool. Computer models have been developed to investigate this situation. For a high-density pool configuration, current models suggest that an exothermic reaction will be initiated in fuel aged up to 1-2 years after discharge from a reactor. These models have not been applied to the specific configuration of the Harris pools. 4 Partial loss of water can be expected in many scenarios, rather than instantaneous, total loss of water. Partialloss of water can be a more severe situation, because convective heat transfer from fuel assemblies is inhibited.

The NRC has neglected this issue. Preliminary analysis suggests that partial water loss could initiate an exothermic reaction in fuel aged 10 years after -

discharge.

4 This probatality estimate should be accompanied by a range of uncertainty. Even with the inclusion of uncertainhes, PRA-derived estimates .cy_..t lower bounds to actual acodent probabilities.

I Risks & alternative options re. spent fuel storage at Harris Page 9 An exothermic reaction could propagate from one set of fuel assemblies to an adjacent set of assemblies that might not otherwise suffer such a reaction.

The NRC's studies of propagation are incomplete, but they acknowledge the potential for propagation.

Exothermic reactions in the Harris pools CP&L representatives have stated that spent fuel assemblies will not be placed in pools C and D at Harris until the assemblies have aged for 5 years after discharge. However, there is nothing in CP&L's license amendment application that prohibits the placement of more recently-discharged fuel in pools C and D. In any case, preliminary analysis suggests that partial water loss could initiate an exothermic reaction in fuel aged 10 years after discharge.

Thus, exothermic reactions could occur in pools C and D.

For the purpose of estimating the potential consequences of a pool accident at Harris, this report considers two scenarios for exothermic reactions. One scenario involves fuel aged up to 3 years after discharge from a reactor, while the second scenario involves fuel aged up to 9 years after discharge fmm a reactor. In both cases, it is assumed that the entire inventory of cesium in the affected fuel assemblies would be released to the atmosphere. This assumption is consistent with NRC studies.

7. Consequences of potential pool and reactor accidents This report focusses on accidents that release an atmospheric plume which travels beyond the plant boundary. The consequences of such a release can be estimated by site-specific computer models. Here, a simpler approach is used, but this approach is adequate to show the nature and scale of expected I consequences. The approach is described in Appendix E. l The role of cesium-137 The consequences of a pool accident can be adequately illustrated by  !

examining a release of only one radioisotope - cesium-137. This isotope has a l half-life of 30 years and is liberally released from damaged fuel. It dominates the offsite radiation exposure from the 1986 Chernobyl accident, and is a major contributor to radiation exposure attributable to fallout from the atmospheric testing of nuclear weapons in the 1950s and 1960s.

Three atmospheric releases of cesium-137 are postulated here for the purpose  ;

of examining consequences. First, a release of about 2 million Curies (2 mci) j corresponds to the most severe reactor accident identified in the Harris IPE. ,

Second, a release of about 20 million Curies (20 mci) corresponds to a pool i

Risks & elternative options re. spent fuel storage at Harris Page 10 -

accident affecting fuel aged up to 3 years after discharge from a reactor. Third, a release of about 70 million Curies (70 MCl) corresponds to a pool accident affecting fuel aged up to 9 years after discharge from a reactor. ,

Land contamination by cesium-137 Acddent consequences are illustrated here by estimating the area of land that would be contaminated by cesium-137 to a level such that inhabitants would suffer an external radiation dose in excess of 10 rem over 30 years.s An exposure of 10 rem over 30 years wouldupcant about a three-fold increase above the typical level of background radiation (which is about 0.1 rem / year).

In its Reactor Safety Study, the NRC used a threshold of 10 rem over 30 years as an exposure level above which populations were assumed to be relocated from rural areas. The same study used a threshold of 25 rem over 30 years as a criterion for relocating people from urban areas, to reflect the assumed greater expense of relocating urban inhabitants.

In an actual case of land contamination in the United States, the steps taken to relocate populations and pursue other countermeasures (decontamination of surfaces, interdiction of food supplies, etc.) would reflect a variety of political, economic, cultural, legal and scientific influences. It is safe to say that few citizens would calmly accept a level of radiation exposure which substantially exceeds background levels.

For typical meteorology, a release of 2 mci would contaminate 4,000-5,000 square kilometers of land, A release of 20 mci would contaminate 50,000-60,000 square kilometers. Finally, a release of 70 MCI would contaminate about 150,000 square kilometers of land. Note that the total area of North Carolina is 136,000 square kilometers and the state's land area is 127,000 square kilometers.

Health effects of radiation There is ongoing debate about the health effects of radiation at comparatively low doses. According to estimates by the National Research Council's BEIR V committee, a continuous exposure throughout life at a rate of 0.1 rem / year (above background) will increase the number of fatal cancers, above the -

normally expected level, by 2.5 percent for males and 3.4 percent for females, with an average of 16-18 years of life lost per excess death. If the dome-response function were linear, it would follow that continuous, lifetime exposure to I rem / year would increase the number of fatal cancers by 25 5Without countermeasures such as interdwtion of food ' supplies, the internal dose could be of a sunilar magnitude to the external dose.

Risks & alternative options re. spent fuel storage at Harris Page 11 percent for males and 34 percent for females. The shape of the dose-response function is a subject of debate.

8. ' Alternative options for spent fuel management The present mode of spent fuel storage in Harris pools A and B poses a major hazard. This hazard will be substantially increased if pools C and D are -

activated. CPacL has not properly characterized the present and potential hazard, nor has the company provided a systematic assessment of alternative options.

A situation like this calls for a systemhtic, comprehensive assessment of alternative options and their impacts. A full range of alternatives should be identified, and their impacts and other characteristics should be assessed.

Performance of such an analysis is beyond the scope of the author's current work for Orange County. An abbreviated discussion is presented here.

Options not reviewed here One option would be to cease operation of CPicL's nuclear plants. That

. option, which could be combined with other options for storage of CPdzL's present stock of spent fuel,is not reviewed here. Another set of options would employ high-density pool storage but would introduce technical measures that sought to increase the reliability of the cooling systems for some or all of the Harris pools, or to decrease the potential for safety interactions between the pools and the reactor. Independent support systems for pools C and D, as mentioned in Section 3, would be in this class of options.

Such options are not reviewed here.

Options reviewed here This report focusses on two classes of options for spent fuel storage. One class involves dry storage of spent fuel, using proven technology. The second class, which could complement dry storage, involves low-density storage in pools. A combination of dry storm and low-density pool storage could offer a practical, proven means of dramatically decreasing the hazard posed by high-density pool storage at Harris.

Dry storage The NRC has approved a variety of designs for the dry storage of spent fuel.

These designs are described in Table 1, and their current use by licensees is

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described in Table 2.' It wiR be noted fmm Table 2 that a dry storage installation is licensed at CP&L's Robinson plant. This installation employs eight NUHOMS-7P modules, each of which can hold 7 fuel assemblies. All eight modules are fully loaded?

Dry storage could be implemented at any of.CP&L's three plant sites. This report does not recommend any particular design, but notes that the designs  !

vary in their level of safety and other features. For example, some designs are l i

more resistant to sabotage than others.

All of tiie approved dry storage designs are safe in the event that access to the plant site is precluded by the release of radioactive material during a reactor accident. None of the designs requires active cooling, electricity or operator attention. A sabotage / terrorism event at a dry storage installation could release only a small fraction of the radioactive material that could be released by a sabotage / terrorism event at the Harris pools in their present and mposed configuration. Overall, dry storage poses a much lower level of hazard than high-density pool storage, for the same quantity of fuel.

At present, the NRC licenses dry storage installations for only 20 years.

However, the technology is capable of storing fuel for much longer periods. If CP&L employs the dry storage option, they should choose a design that has this capability. This choice, properly documented and supported by ongoing testing, would establish the basis for a license extension in the future.

Low-density pool storage Spent fuel can be stored in pools in a low-density, open-rack configuration, as was common practice when nuclear plants were first operated. Given a sufficiently low-density configuration, partial or total uncovering of the fuel will not initiate an exothermic reaction in the fuel cladding, even for recently l discharged fuel. The fuel would remain vulnerable to consolidation through l a cask drop into a pool or L severe earthquake which disrupts the fuel racks. If such consolidation were accompanied by partial or total uncovering, an exothermic reaction could occur in the consolidated region. However, it is unlikely that this reaction would be propagated to other regions of a pool.

6 Tables 1 and 2 are adapted from: US Nuclear Regulatory Canrrussion, k/ormation Digest.

1998 NAi+irm NURNC-118io. Vol-a 10. November 1998.

7M G Raddatz and M D Waters, Inform =+ian Handhaak on Indeoendent Soent Fuel Storane Installations. NUREG-1571. December 1996.

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Risks & alternative options re. spent fuel storage at Harris Page 13 Summary CP&L could employ a spent fuel storage strategy which combines dry storage with low-density pool storage. Some or all of pools A, B, C and D at Harris would be used in a low-density configuration. If appropriately designed and implemented, this strategy could dramatically reduce the hazard posed by present and proposed fuel storage arrangements at Harris.

9. Addressing risks and alternatives in the regulatory arena Orange County has requested the NRC to hold a hearing regarding CP&L's license amendment application, and the NRC has established a Licensing Board for this case. These actions have initiated a regulatory process which has lieen employed many times before. A review of this process is beyond the scope of this report, but some brief observations may be helpful.

The licensing process will typically assume that regulatory decisions taken in the past were correct. Thus, the existing operations at Harris pools A and B might be held to establish a precedent for the proposed operations at pools C and D. However, this report shows that the NRC has not properly analyzed the potential for severe pool accidents at a generic level. This point may or may not influence the NRC's regulatory process, but it deserves continuing emphasis through all available channels.

At Harris, and nationwide, there is a need for a thorough assessment of the hazards associated with high-density pool storage, and of alternative options which could pose a lower hazard. Orange County would provide an important public service if it could persuade the NRC or another body to J conduct such an assessment, perhaps in the form of an environmental impact statement. There has been discussion about the US Department of Energy taking title to the nation's spent fuel, while the fuel remains at plant sites. This move could provide an opportunity for a thorough assessment of 1 risks and options, and for the adoption of safer means of fuel storage.

10. Conclusions l

C1 Given the present and proposed configuration of spent fuel storage in ,

the Harris pools, partial or total loss of wat'er from the pools could initiate l exothermic reactions of fuel cladding,in any or all of pools A, B, C and D. j C2 Partial or total loss of water from the Harris pools could occur through ,

a variety of events including acts of malice, and would be an almost certain  ;

outcome of a severe reactor accident at Harris involving containment failure l l

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Risks & alternative options re. spent fuel storage at Harris Page 14 or bypass; CP&L estimates the probability of the latter event as 1 per 100,000 reactor-years.

O Exothermic reactions in the Harris pools could release to the environment an amount of cesium-137 at least an order of magnitude larger than the amount released by the most severe potential acddent at the Harris reactor.

O A large release of cesium-137, as could occur from exothermic reactions in the Harris pools, could significantly contaminate an area of land equal to the area of North Carolina.

O The probability and magnitude of a potential release from Harris of radioactive material in spent fuel could be dramatically reduced if CP&L adopted a fuel storage strategy which combines dry storage with low-density pool storage; this strategy would employ proven technology.

G Activation of pools C and D at Harris could increase the probability and magnitude of design-basis or severe accidents at the Harris fuel pools or reactor.

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Risks & alternative options re. spent fuel storage at Harris

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EHEHllIEEEE PELLET CUTAWAY VIEW OF OXIDE FUEL "IAI II M 1 FOR COMMERCIAL LWR POWER l PIAIGS.

The basic unit in the core of a light-water reactor is Control rod ,

a fuel rod containing uranium oxide penets in a Zircaloy dadding. The rod is flued will helium gas and welded shut. The cirded portaon exaggerates the annular space between the pellet and the Z dadding. (Figure reproduced from WASH-1250.)

Fuel rod

' ; 11, b

11111)

Ilh l l l

Spring clip - T I l grid assembly Q'"l llll FUEL ASSEMBLY FOR A PRESSURIZESWATER REACTOR.

In a presurized-water reactor, fud rods are assembled into a square array, held together by spring clip assemblics and by nozzles at the top and bottom. The structure is open, permitting flow of i

coolant both verticany and horizontally. AB the anembin ee react r may hm se same Botton nozzle % ~

mechanical design, including prov.sion for pamage d W of a control rod cluster (shown in the figure).

Where there is no cluster, these positions may have neutron sources, burnable poison rods, or plugs.

(Figure reproduced from WASH-1250.)

Bottom view D l l

Figure 1 i Fuel for a pressurized-water reactor

Risks & citernative options re. spent fuel storage at Harris Page 16 .

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om Genwal Nudear MehlCod Syihnu,incorporahd CASTOR V/21 21PWR 09/30/1985 08/17/1990 CASTOR X/28 28 PWR 04/22/1994 CASTOR X/33 33 PWR 11/24/1995 Tramnuclear, Concreen Module 7PWR Wat incorporated NUHOMS-7P 03/28/1986 Weihnghouse Eisdric MehlCod 2JPWR 09/30/1987 08/17/1990 MC-10 FW Enwy.' ' _, Concr* Vauk 83 PWR ar 03/22/1988 incorponned Modulw Vauh 1508WR DryShrage(MvDS)

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Iram 56 PWR anemblies)

Trarunuclear, MehlCod 11/04/1993 incorporated M24 24 PWR 07/05/1989 M 32 32 PWR 11/07/1996 NAC inewnahanol,Inc. MehlCash 28 PWR 02/01/1990 NAC-128/ST Sierra Hudear Vedilated Cod 24 PWR 03/29/1991 05/03/1993 Corporaban VSC 24 Transnudsar Wed,Inc. Carrese Module 04/21/1989 01/18/1995 shndmand NUHOMS-24P 24fWR NUNCW652B 528WR NAC Internahonal,Inc. t%C-STC 26 PWR 07/17/1995 Noir PWR - Pressurized-Wahr Reador; BWR - BoilirwyWater Reador Table 1 NRC-appmved dry spent fuel storage designs

[ .

Risks & alternative options re. spent fuel storage at Harris

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Table 2 NRC dry spent fuel storagelicensees

RISKS AND ALTERNATIVE OPTIONS ASSOCIATED WITH SPENT FUEL STORAGE AT THE SHEARON HARRIS NUCLEAR POWER PLANT Appendix 'A Spent fuel management at the Harris plant l

1. Introduction This appendix summarizes present and proposed arrangements for managing spent fuel at the Shearon Harris plant. Carolina Power & Light Company (CP&L), the licensee for the plant, proposes to introduce new arrangements for spent fuel management. For that purpose, CP&L seeks an amendment to the plant's operating license. Unless specified otherwise, information presented hers is drawn from CP&L's application to amend the Harris license, from CP&L's Final Safety Analysis Report (PSAR) for the Harris plant, or from viewgraphs shown by CP&L personnel during meetings with staff of the Nuclear Regulatory Commission (NRC).1

2. Present and proposed spent fuel storage capacity The Harris plant features one pressurized-water reactor (PWR). The core of l this reactor contains 157 fuel assemblies, with a center-center distance of about  !

8.5 inches. The Harris plant was to have four units but only the first unit was built. (A unit consists of a reactor, a turbine-generator and associated equipment.) A fuel handling building was built to serve all four units. This l building contains four fuel pools (A, B, C, D), a cask loading pool and three  !

fuel transfer canals, all interconnected but separable by gates. Figure A-1 .

shows a plan view of the interior of the fuel handling building.

Pools A and B Pools A and B contain fuel racks, and are in regular use. CP&L says that fresh fuel, and spent fuel recently discharged from the Harris reactor, is stored in pool A. Fuel examination and repair are performed in an open space in pool 4 1 Meetings 1ci_..NRC staff and CP&L . y._..tatives, to / . ss the pmposed hcense amendment, were held on 3 March 1996 and 16 July 1996.

e Risks & alternative options re. spent fuel storage at Harris Appendix A -

Page A-2 B. Pools C and D are flooded but do not contain racks. The cooling and water j cleanup systems for pools C and D were never completed.

Pool A now contains six racks (360 fuel amaambly spaces) for PWR fuel and three racks (363 spaces) for boiling-water reactor (BWR) fuel, for a total pool capacity of 723 fuel assemblies. Pool B contains twelve PWR racks (768 spaces) and seventeen BWR racks (2,057 spaces), and is licensed to store one additional BWR rack (121 spaces), for a total, potential pool capacity of 2,946 fuel assemblies. Thus, pools A and B now have a combined, potential capacity of 3,669 fuel assemblies. The center-center distance in the racks in pools A and B is 10.5 inches for PWR fuel and 6.25 inches for BWR fuel.

Pools A and B store spent fuel from the Harris reactor and from CP&L's Brunswick plant and Robinson plant. The Brunswick plant has two BWRs while the Robinson plant has one PWR. Shipment of spent fuel from Brunswick and Robinson to Harris is said by CP&L to be necessary to allow core cffload capacity in the pools at Brunswick and Robinson.

Pools C and D CP&L seeks an amersdment to its operating license so that it can activate pools C and D at Harris. By activating these pools, CPkL expects to have sufficient storage capacity at its three nuclear plants to accommodce all the spent fuel discharged by the four CP&L reactors (the Harris and Robinson PWRs and the two Brunswick BWRs) through the ends of their current operating licenses.

CP&L plans to install racks in pool C in three campaigns (approximately in 2000,20% and 2014), to create 927 PWR spaces and 2,763 BWR spaces, for a total capacity in this pool of 3,690 fuel assemblies. Thereafter, CP&L plans to install racks in pool D in two campaigns (approximately in 2016 and at a date to be determined), to create 1,025 PWR spaces. Thus, the ultimate capacity of pools C and D will be 4,715 fuel assemblies. The center-center distance in the racks used in these pools will be 9.0 inches for PWR fuel and 6.25 inches for BWR fuel.

The PWR racks in pools C and D have a smaller center-center distance than the racks in pools A and B (9.0 inches instead of 10.5 inches). This arrangement allows more PWR fuel to be placed in a given pool area but also means that PWR fuel in pools C and D is more prone to undergo criticality.

In response, CP&L proposes to include in the Technical Specifications for Harris a provision that PWR fuel will not be placed in pools C and D unless it has relatively low e:arichment and high burnup.2 2 Liense amendment application, Enclosure 5.

i. .

Risks & alternative options re. spent fuel storage at Harris Appendix A Page A-3 Summary Table A-1 summarizes the present and proposed storage capacity in the Harris pools. At present, pools A and B have a coubined, potential capacity of 3,669 assemblies. The proposed, combined capacity of pools C and D will be 4,715 assemblies. Thus, activation of pools C ed D will represent an increase of about 130 percent in the number of fuel assemblies that could be stored at Harris.

3. Support services forpools C and D The water in a spent fuel pool must be cooled and cleaned. Figure A-2 provides a schematic view of typical cooling and cleanup systems. It will be noted that pool water is circulated through heat exchangers, where its heat is transferred to a secondary cooling system. At Harris, the secondary cooling system is the component cooling water (CCW) system. Water in the secondary system is in turn circulated through heat exchangers, where its heat is transferred to a tertiary cooling system. At Harris, the tertiary cooling system is the service water (SW) system.

When the Harris plant was designed, the intention was that pools C and D would be cooled by the CCW system for the second unit. That unit was never built and its CCW system does not exist. Thus, CPLL plans to cool pools C and D by completing their partially built cooling systems and connecting those systems to the CCW system of the first unit. The Unit 1 CCW system already provides cooling to pools A and B and serves other, important safety functions. For example, the Unit 1 CCW system provides cooling for the residual heat removal (RHR) system and reactor coolant pumps of the Unit I reactor.

The original design concept for Harris In the Harris plant's original design concept, pools A and B would have served Units 1 and 4, while pools C and D would have served Units 2 and 3.

There would have been a separate, fully-redundant,100 percent-capacity cooling and water cleanup system for each pair of pools (A+B and C+D).

Cooling of pools C and D would have been provided by the CCW system of Unit 2. Electrical power for the pumps that circulate water from the C and D pools through heat exchangers (see Figure A-2) would have been supplied by the Unit 2 electrical systems. Pools A and B would have been supported by the CCW and electrical systems of Unit 1.

Risks & alternative options re. spent fuel storage at Harris .

Appendix A -

Page A-4 During CP&L's planning for the activation of pools C and D, the company considered the construction of an independent system to cool these pools.

Within that option, CP&L considered the further possibility of providing dedicated emergency diesel generators to meet the cdectrical needs of pools C and D if normal electridty supply were unavailable. Construction of an independent cooling system for pools C and D, supported by dedicated emergency diesel generators, could provide the level of safety that was associated with the original design concept for Harris. However, CP&L has not proceeded with this option.

Capacity of the Unit 1 CCW system According to CP&L's license amendment application, the bounding heat load from the fuel in pools C and D will be 15.6 million BTU / hour (4.6 MW).3 At present, the Unit 1 CCW system cannot absorb this addrional heat load.

Thus, CP&L proposes to include in the Technical Specifications for Harris an interim provision that the heat load in pools C and D will not be allowed to exceed 1.0 million BTU / hour.4 CP&L claims that an additional heat load of 1.0 million BTU / hour can be accommodated by the Unit 1 CCW system, and that the fuel to be placed in pools C and D will not create a heat load exceeding 1.0 million BTU / hour through 2001.

CP&L contemplates a future upgrade of the Unit 1 CCW system, so that this system can accommodate an additional heat Nad of 15.6 million BTU / hour from pools C and D. This contemplated upgrade is not described in the present license amendment application. Apparently, CP&L intends to perform the upgrade of the Unit 1 CCW system concurrent with a power uprate for the Unit I reactor. A 4.5 percant power uprate of the reactor will be associated with steam generator replacement, and will take effect in about 2002. About two years later, there will be a further power uprate of 1.5 percent. CP&L projects that the Unit 1 CCW heat load, including the reactor power uprate and the ongoing use of pools C and D, will substantially exceed the capability of the present CCW system.

To summarize, CP&L's short-term plan (through 2001) for cooling pools C and D is to exploit the margin in the Unit 1 CCW system, so as to accommodate an additional heat load of 1.0 million BTU / hour. CP&L's longer-term plan is to upgrade the CCW system, in a manner not yet specified, so as to accommodate an additional heat load of 15.6 million BTU / hour. The CCW upgrade must also accommodate an increase in the rated power of the Harris reactor. CP&L expects that the design of the CCW 3 License amendment appikation, Enclosure 7, page 5-16.

4 Lkenne amendment application, Enclosure 5.

Risks & citernative options re. spent fuel storage at Harris Appendix A Page A-5 upgrade will commence in mid-1999 and wd oe completed in early 2001, one year after the company expects pool C to enter service.

Safety implications In order to exploit the ma,rgin in the existing CCW system so as to cool pools C and D, CP&L may be obliged to require its operators to divert some CCW flow from the RHR heat exchangers during the recirculation phase of a design-basis loss-of-coolant accident (LOCA) event at the Harris reactor.5 This is a safety issue because, during the recirculation phase of a LOCA, operation of the RHR system is essential to keeping the reactor core and containment in a safe condition. CP&L's exploitation of the margin in the existing CCW system is deemed by CP&L and NRC to constitute an "unreviewed safety question".'

In Enclosure 9 of its license amendment application, CP&L provides a brief description of the analysis that is has performed to demonstrate that an additional load of 1.0 million BTU / hour is within the marginal capacity of the Unit 1 CCW system. That analysis is said by CP&L to take the form of a 10CFR50.59 Safety Evaluation. The description in Enclosure 9 raises more questions than it answers, and does not address the practical issues that affect an analysis of a cooling system's thermal margin. For example, CP&L has mentioned elsewhere that exploitation of the margin in the Unit 1 CCW system could involve changes in design assumptions that include fouling factors and tube plugging limits? These matters are not addressed in Enclosure 9.

As background, note that the Unit 1 CCW system has two heat exchangers, each with a design heat transfer rate of 50 million BTU / hour. During the recirculation phase of a design-basis LOCA, the estimated maximum heat load to be extracted from the CCW system by the SW system is 160 million BTU / hour.s These numbers suggest that accommodating a design-basis LOCA will already exploit the margin of the CCW system, without any additionalload from pools C and D.

Lack of QA documentation Activation of pools C and D will require the completion of their cooling and water cleanup systems, and the connection of their cooling systems to the 5

License amendment application, Enclosum 9. ,

6 Ibid;FederalRegister: January 13,1999 (Volume 64, Number 8), pages 2237-2241.

7 Viewgraphs for presentation by CP&L to the NRC staK,3 March 1998.

8 Harris PSAR, section 9.2, Amendment No. 40.

1

Risks & alternative options re. spent fuel storage at Harris Appendix A -

Page A-6 existing CCW system. CP&L states that approximately 80 percent of the necessary pping was completed before the second Harris reactor was cancelled.- However, some of the quality assurance (QA) documentation for the completed piping is no longer available. Much of the completed piping is embedded in concrete and is therefore difficult or impossiF 3 inspect. To address this situation, CP&L proposes an Alternative 1% i demonstrate that the previously completed piping and other equipment is adequate for its purpose.10 Nevertheless, the cooling systems for pools C and D will not satisfy ASME code requirements.

Electrical power The cooling systems for pools C and D will draw electrical power from the electrical systems of Unit 1. If electricity supply to the cooling pumps for pools C and D is interrupted, the pools will heat up and eventually boil.

CP&L says that pools C and D will begin to boil after a time period "in excess of 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />", assuming a bounding decay heat load of 15.6 million BTU / hour.11 To prevent the onset of pool boiling in the event of a loss of offsite power, the Harris operators may be obliged to provide electrical power to pools C and D from the existing emergency diesel generators, which also serve pools A and B and the Unit I reactor. In its license amendment application, CP&L does not address the ability of the emergency diesel generators to meet the additional electrical loads associated with pools C and D. CP&L does mention in the Harris FSAR the potential for connecting

" portable pumps" to bypass the pool cooling pumps should the latter be inoperable.12 However, the characteristics, capabilities and availability of such portable pumps are not addressed in the license amendment application.

4. Potential cesium-137 inventory of the Harris pools For the purposes of Appendix E of this report,it is necessary to estimate the potential inventory of the radioisotope cesium-137 in the Harris pools. As a starting point, consider the inventory of cesium-137 in a typical PWR spent fuel assembly, represented here by an average assembly in batch 16 from the Ginna plant, discharged in April 1987. At discharge, the Ginna assembly contained 1.4 x 105 Curies of cesium-137 per metric ton of heavy metal (MTHM).13 9 License amendment application, Enclosure 1, page 4.

10 Lianse amendment apphcation, Enclosure 8.

11 Ucense amendment appliation, Enclosure 7, page 5-8.

12 Harris PSAR, page 9.1.3-6, Aws.dws.t No. 48.

13 V L Sailor et al, Severe Arvidents in Soent Fuel Pools in Suonort of Generic Safety Issue 82, NUREG/CR-4982. July 1987, Appendix A.

Risks & citernative options re. spent fuel storage at Harris Appendix A Page A-7 A Harris PWR assembly has a mass of 0.461 MTHM. Thus, one can estimate that a typical Harris assembly contains, at discharge,0.65 x 105 Curies of cesium-137. The assembly's content of resium-137 will decline exponentially, with a half-life of 30 years. At the same age after discharge, a typical BWR assembly in the Harris pools will contain about 1/4 of the amount of cesium-137 in a Harris PWR assembly.14 Potential stock of assemblies in the Harris pools Table A-2 shows CP&L's projection of the stock of assemblies in Harris pools C and D, for the purposes of bounding analysis. A CP&L representative has stated that CP&L will not ship fuel to Harris until it has aged for 3 years, and will not place fuelin pools C and D untilit has aged for 5 years.15 Accepting that fuel aged less than 3 years will not be shipped to Harris, one can assume, to supplement Table A-2, that the Harris pools will contain 456 BWR assemblies aged for 3 years,172 PWR assemblies aged for 3 years, and % PWR assemblies aged for 1 year. Hereafter, these assumptions and Table A-2 are taken to represent the potential stock of fuel assemblies in the Harris pools.

On this basis, the Harris pools' stock of spent fuel aged 3 years or less will be 268 PWR assemblies and 456 BWR assemblies. All of this fuel might be in pools A and B, although there is nothing in CP&L's present or proposed Technical Specifications which prohibits placement of recently discharged fuel in pools C and D. On the same basis, the Harris pools' stock of spent fuel aged 9 years or less will be 784 PWR assemblies and 1,824 BWR assemblies.

Inventory of cesium-137 Now consider the inventory of cesium-137 in the Harris pools. Assume that a ' newly discharged PWR assembly contains 0.65 x 105 Curies of cesium-137, neglect the difference between Harris and Robinson assemblies, allow for radioactive decay, and assume that a BWR assembly contains 1/4 of the amount of cesium-137 in a PWR assembly of the same age. Then, the Harris pools' stock of spent fuel aged 3 years or less will contain 23 x 107 Curies (870,000 TBq) of cesium-137, with a mass of 260 kilograms. Also, the Harris pools' stock of spent fuel aged 9 years or less will contain 7.1 x 107 Curies (2,600,000 TBq) of cesium-137, with a mass of 790 kilograms.

14 'Ihe ratio of 1/4 derives from the parameters shown in the license amendment application, Enclosure 7,page 5-15.

15J Scarola of CP&I, presentation to Orange County Board of CO.TwJssionen,9 February 1999, i

Risks & alternative options re. ayent fuel storage at Harris Appendix A Page A-8 1 CP&L could provide a more precise projection of the cesium-137 inventory in the Harris pools over coming years. However, our estimate will be a reasonable indication of cesium-137 inventory during the next two decades, assuming pools C and D are used as CP&L intends.

For comparison with the pools' inventory of cesium-137, note that the NRC has estimated the inventory of cesium-137 in the Harris reactor core, during normal operation, to be 4.2 x 106 Curies (155,000 TBq, or 47 kilograms).16 This represents an average inventory of 0.27 x 105 Curies in each of the reactor's 157 fuel assemblies. Note that an average assembly in the core will have a lower cesium-137 content than an assembly at discharge, and that the NRC's estimate may have assumed a relatively low fuel burnup.

1 l

1 16 US Nuclear Regulatory Commission, Final Environmental Statement Related to the Operadon of Shearon Harris Nuclear Power Plant Units 1 and 2. NURECr0972. October 1983.

_m

Risks & alternative options re. spent fuel storage at Harris

. Appendix A Page A-9

. 0 0F UNIT 4 CONTAIM MT

\A >(

FUEL TRANSFER CONS M N 1 CANALS x \ _

' + '  !

a d

$ MORTH k*, P0OL B /

.(.

21 7 i ^ 41 0 0F UNIT 1 fCONTAINMENT C 0F HIT 3 CONTAINNDT IQUIPENT j FUEL TRANSFER j (NOT CONSTHCTED) /

CANALS , EATCR l f

l , x r N l POOL C 9lO ' c' P00L D CASK LOABING MOL l l g

- L 45 7 '

N 73 76r 0 0F UNIT 2 y CONTAlHENT (NOT CONSTRUCTED) l 1

Source: License amendment application Figure A-1 Interior of the Harris Fuel Handling Building

Risks & alternative options re. spent fuel storage at Harris )'

Appendix A .

Page A-10 ,

1

>>> ^>> 9 ' - Na

- e FUEL POOL CLEANUP SYSTEM 9 ,.-  ; 9 =m

1. 3

[

x DEMINERALIZERS lllllll//// /f n

FILTERS ae FUEL POOL C00LIIIG SYSTEM iI iL h h INTE2 MEDIATE I! EAT EXCHANGF $

I , , , ,

I = n* -

^

REACTORBUILDINGCLOSEDCOOLINGWAthSYSTEMOR j COMPONENT C00 LING WATER SYSTEM _

l HEAT SINK Y0Ei,Y!iEIl$"f0N0f*"

Source: NUREG-0404 Figure A-2 Typical cooling and cleanup systems for a spent fuel pool

Risks & alternative options re. spent fuel storage at Harris

. Appendix A Page A-11 Pool PWR spaces BWR spaces Total

'A' 360 363 723

'B' 768 2178 2946

'C' 927 2763 3690-

'D' 1025 ~

0 1025 Total 3080 5304 8384 l

I Source: License amendment application Table A-1 Present and proposed storage capacity in the Harris pools

Risks & alternative options re. spent fuel storage at Harris Appendix A .

Page A-12 DECAY PERIODS FOR A BOUNDING POOLS C AND D STORAGE CONFIGURATION PWR Feel Assemblies BWRFuel Assemblies Number of Assys Decay Period Number of Assys I = cay Period 172 5 years 456 5 year:

172 7 years 456 7 years 172 9 years 456 9 years 172 11 years 456 11 years 172 13 years 456 13 years 172 15 years 483 15 years 172 17 years 172 19 years 172 21 years 172 23 years 232 25 years Source: License amendment application Table A-2 Projected stock of fuel assemblies in Harris pools C and D

RISKS AND' ALTERNATIVE OPTIONS ASSOCIATED WITH SPENT FUEL STORAGE AT THE SLEARON HARRIS NUCLEAR POWER PLANT Appendix B Potential for severe accidents at the Harris reactor 1 Introduction In examining the risks associated with spent fuel storage at Harris, one must consider the potential for accidents at the Harris reactor. Such consideration is necessary for two reasons. First, a reactor accident could accompany, initiate or exacerbate a spent fuel pool accident. Second, modification of the Harris plant to increase its spent fuel storage capacity could increase the probability or consequences of accidents at the Harris reactor.

This appendix addresses the potential for severe accidents at the Harris reactor. " Severe" reactor accidents have two major defining characteristics.

First, they involve substantial damage to the reactor core, with a corresponding release of radioactive material from the fuel assemblies.

Second, they extend the envelope of potential accidents beyond the " design basis" accidents that were considered when US reactors were first licensed.'

During a severe reactor accident, radioactive material may be released to the environment, as an atmospheric plume or by entry into ground or surface waters. The release may be large or small. In illustration, the 1979 TMI accident and the 1986 Chernobyly accident were both severe accidents, involving substantial damage to the reactor core. However, the TMI release was comparatively small and the Chernobyl release was comparatively large.

2. Probabilistic risk naeannent The probabilities and consequences of potential accidents at nuclear facilitie.

can be estimated through the techniques of probabilistic risk assessment (PRA). Nuclear fadlity PRAs are performed at three levels. At Level 1, a PRA will estimate the probability of a specified type of accident (e.g., severe core damage at a reactor). At Level 2, which builds upon Level I findings, a PRA will estimate the nature of potential radioactive releases from the facility. In l

I J

Risks & alternative options re. spent fuel storage at Harris Appendix B Page B-2 turn, the Level 2 findings can be used in a Level 3 exercise, which will estimate the offsite consequences (health effects, economic effects, etc.) of radioactive releases. For all three levels, a PRA can be performed for

" internal" accident fnitiating events (equipment failure, operator error, etc.)

and for " external" accident-initiating events (earthquakes, floods, etc.).1 PRA methodology is used for non-reactor nuclear facilities, but is most highly developed in its application to reactors. The first PRA was the Reactor Safety Study (WASH-1400), which was published by the US Nuclear Regulatory Commission (NRC) in 1975.2 The present state of the PRA art is exemplified by a study of five nuclear power plants (NUREG-1150) published by the NRC in 1990.3 Uncertainty and incompleteness of PRA findings An in-depth PRA such as NUREG-1150 can provide useful insights regarding a reactor's accident potential. However the findings of any PRA will inevitably be accompanied by substantial uncertainty and incompleteness.

Uncertainty arises from the intrinsic difficulties of modelling complex systems, and from limited understanding of some of the physical processes that accompany severe accidents. Incompleteness arises from the potential for unanticipated accident sequences, gross human errors, undetected structural flaws, and acts of malice or insanity.4 Thus, a PRA's finding about the probability of an accident should be viewed with two caveats. First, the accident probability, as found in the PRA, will fall within some range of uncertainty. Second, the accident probability, as found in the PRA, will be a lower bound to the true probability, which will be impossible to determine.

NUREG-U50 findings for the Surry PWRs Figures B-1 and B-2 illustrate the findings of NUREG-1150. These figures show the estimated core damage frequency for the Surry nuclear reactors.

These reactors are 3-loop Westinghouse pressurized-water reactors (PWRs), as is the Harris reactor. Core damage frequency is shown per reactor-year of 1 In PRA practice,it is common for analysis of externally-initiated accidents to build upon revious analysis of internally-initiated accidents.

US Nuclear Regulatory Commission, Raetor Safety Study. WASH-1400 (NUREG-75/014).

October 1975.

3 US Nuclear Regulatory Comminninn. Severe Accident Risks: An Assessment for Five US Nuciaar Power Plants. NUREG-1150 (2 vols). Deamber 1990-4 H Hirsch, T Einfalt, O Schumacher and G Thompson,IAEA Safety Tareets and Probabilistic Risk Assessment. Gcaellschaft fur Okologische Forschung und Beratung, Hannover, August 1989.

m

Risks & alternative options re. spent fuel storage at Harris Appendix B Page B-3 operation. Figure B-1 shows core damage frequency for internal events, fires ,

and earthquinken (seismic events). Two estimates are shown for seismic events, one drawing on an estimate of earthquake frequency by Lawrence Livermore National Laboratory, the other on an estimate by tle Electric Power Research Institute (EPRI). The bars in Figure B-1 span an estimated uncertainty range from the 5th to the 95th percentile. An alternative portrayal of estimated uncertainty is provided by the probability densities I shown in Figure B-2. l The authors of NUREG-1150 made a considerable effort to estimate the uncertainty associated with their findings. However, their uncertainty estimates relied heavily on expert opinion, rathat than on a statistical analysis of data. Thus, the uncertainty estimates in NUREG-1150 should be viewed with caution. The reader will observe a cautionary statement attached to Figures B-1 and B-2. Finally, the NUREG-1150 findings of accident probability must be viewed as lower bounds, as explained above. i Acts of malice Nuclear reactor PRAs do not consider malicious acts such as sabotage, terrorism or acts of war. Such acts are less susceptible to probabilistic analysis  ;

than are accident initiators such as human error. Nevertheless, sabotage and l terrorism pose a significant threat to US nuclear plants.5 NRC regulations oblige reactor licensees to take certain precautions against this threat, but ,

these precautions do not preclude the possibility of successful acts of sabotage or terrorism.

The US government is increasing the level of attention and the expenditure that it devotes to the tl:reat of terrorism. Many observers argue that greater effort is required. For example, three authors with high-level government experience have recently written:6 Long part of the Hollywood and Tom Clancy repertory of nightmarish scenarios, catastrophic terrorism has moved from far-fetched horror to a contingency that could happen next month. Although the United j States still takes conventional terrorism seriously, as demonstrated by the response to the attacks on its embassies in Kenya' and Tanzania in August,it is not yet prepared for the new threat of catastrophic  ;

terrorism.

5G ' thompson, War. Terrorism and Nuclear Power Plants. Peace Reseanth Centre, Australian National University, October 1996.

6 A Carter, J Deutch and P Zelikow, Catastrophic Terrorism", Foreien Affairs.

November / December 1998,page 80.

i

Risks & alternative options re. spent fuel storage at Harris Appendix B Page B-4 The effectiveness of licensees' arrangements to resist terrorist attacks on nudear plants has recently been a subject of public debate. According to the head of the NRC's Operational Safeguards Response Evaluation program, plant security arrangements have failed in at least 14 of the 57 mock assaults which the NRC has conducted since 1991. Nevertheless, the NRC intends to weaken its oversight of licensees' antiterrorism efforts.7

3. The HarrisIPEand IPEEE The NRC requires each holder of a reactor license to perform an Individual Plant Evarnination (IPE), to assess the severe accident potential of that reactor.

Carolina Power and Light (CPdcL) submitted an IPE for the Harris reactor in 1993.8 This was a Level 2 PRA for internal events, including in-plant flooding but neglecting in-plant fires.

The NRC also requires each licensee to perform an Individual Plant Examination for External Events (IPEEE). CPdcL submitted an IPEEE for the Harris reactor in 1995.9 This study did not follow PRA practice. Instead,it consisted of a seismic margins analysis and a limited analysis of in-plant fires.

IPE estimate of core damage frequency According to the IPE performed by CPacL, the frequency of severe core damage at Harris is 7 x 10-5 per reactor-year. This must be considered a " point" estimate, because the Harris IPE does not provide an uncertainty band or probability density function of the kind shown in Figures B-1 knd B-2. The IPE predicts that accident sequences involving a loss-of-coolant accident (LOCA) will account for 40 percent of Harris' core damage frequency, while )

sequences involving station blackout (loss of electrical power) will account for 26 percent of the core damage frequency. The 40 percent contribution of LOCAs to core damage frequency is due to LOCAs with injection failure (17 percent) and LOCAs with recirculation failure (23 percent).

7S Allen,"NRC to cut mock raids on atom plants",'Ihe Bostpn Globe,25 February 1999, page A6.

8 Carolina Power & Light Company, Shggpn Harris Nehr Power Plant Unit No.1:

Individual Plant hmmin=+ian 4heniH=L August 1993.

9 Carolina Power & Light Company, Shenron Harris Nehr Power Plant Unit No.1:

Individual Plant Examination for External Events Submittal. Tune 1995.

Risks & alternative options re. spent fuel storage at Harris Appendix B Page B-5 The NRC has compiled and compared IPE findings for all US commercial nuclear reactors.10 Some of the results are shown in Figures B-3 and B-4.

Figure J-3 shows that the reported core damage frequencies tend to be significantly higher for PWRs than for boiling-water reactors (BWRs). Figure B-4 shows that the reported core damage frequencies tend to be higher for 3-loop Westinghouse (W-3) PWRs than for 2-loop and 4-loop Westinghouse PWRs and PWRs made by Combustion Engineering (CE) and Babcock de Wilcox (BdcW). The Harris reactor is a 3-loop Westinghouse PWR.

From its compilation of IPE findings, the NRC concluded that sequences involving LOCAs (especially LOCAs with recirculation failure) and station blackout are major contributors to estimated core damage frequency at 3-loop Westinghouse PWRs. This conclusion is consistent with the Harris IPE findings outlined above. The NRC noted that the 3-loop Westinghouse PWRs exhibit a relatively high dependence of front-line safety systems on service water (SW), component cooling water (CCW) and heating, ventilating oc air conditioning (HVAC) systems.

IPEEE findings The Harris IPEEE consisted of a seismic margins analysis and a limited analysis of in-plant fires. The seismic margins analysis examined the Harris reactor's ability to withstand a review level eathquake (RLE) of 0.3g. Note that the reactor's safe shutdown earthquake (SSE) is 0.15g and its operating basis earthquake is 0.075g. According to the IPEEE, the only actions required to make the Harris reactor safe against tha. RLE involved housekeeping and minor modifications, and these actions have been taken. The IPEEE did not investigate the implications of an earthquake more severe than the RLE. )

i A limited analysis of in-plant fires appears in the IPEEE. This analysis identified four fire scenarios as significant contributors to core damage frequency. One scenario would take place in each of switchgear rooms A and l B, and two scenarios would take place in the control room. The combined core damage frequency, summed over all four scenarios, would be 1 x 10-5 per reactor-year, but the IPEEE argues that a summation of this kind would be inaccurate without further refinement of the analysis.

Figures B.1 and B-2 illustrate the findings that can be generated by the systematic application of PRA techniques to accident sequences initiated by external events. In comparison, the Harris IPEEE is a relatively crude study.

l 10 US Nuclear Regulatory Commission,Indindual Plant Exammation Program. Perspectives on l Rametor Safety and Plant Performarn NUREG-1560 ( 3 volst December 1997. l 1

Risks & alternative options re. spent fuel storage at Harris Appendix B .

Page B-6 Release of radioactive material The Harris IPE analyzes the potential for accident sequences to release radioactive material to the environment. The IPE only considers releases to the atmosphere during accident sequences that are initiated by internal events. Potential releases are described by a set of release categories.

Release category RC-5 represents the largest release identified in.the IPE. This release would include 100 percent of the noble gas inventory in the reactor core,59 percent of the CsI inventory, and 53 percent of the CsOH inventory.

The IPE does not describe how cesium would be distributed between Csl and CsOH. Thus, one can interpret the RC-5 release as including 59 percent of iodine isotopes in the core and 53-59 percent of cesium isotopes.

Accident sequences contributing to release category RC-5 would involve steam generator tube rupture (SGTR) with a stuck-open safety relief valve (SRV), or an inter-system LOCA (ISLOCA). The SGTR could occur as an ,

accident initiating event or through overheating of steam generator tubes during an accident sequence initiated by some other event. A stuck-open SRV, concurrent with a SGTR, would create a direct pathway from the reactor core to the atmosphere, bypassing the containment. In an ISLOCA sequence, ,

reactor cooling water would be lost from a breach in a piping system outside j the containment. This loss of water would initiate the accident, and the  !

water's escape pathway would provide a route for the escape of radimetivity l after core damage began. {

An accident in release category RC-5 would cause substantial offsite exposure to radioactivity. In addition, the Harris plant and its immediate surroundings .

would become radioactively contaminated to Ote point where access by personnel would be precluded. Accidents in other release categories would release smaller amounts of radioactive material, but could also contaminate the Harris plant to the point where access by personnel would be precluded.

This matter is addressed further in Appendix C.

The Harris IPE estimates the probability of release category RC-5 as 3 x 10 4 per reactor-year. Note that the overall probability of core damage is estimated to be 7 x 10-5 per reactor-year. Thus, the IPE predicts that 4 percent of core damage sequences would yield a release in category RC-5. Overall, the IPE predicts that 15 percent of core damage sequences would be accompanied by a

r Risks & alternative options re. spent fuel storage at Harris Appendix B Page B-7 significant degree of containment failure or bypass, with a total probability of about 1 x 10-5 per reactor-year.11

4. Pool-reactor interactions Neither CPdsL nor NRC have performed an analysis to deterinine how a severe accident or a design-basis accident at the Harris reactor might accompany, initiate or exacerbate an accident at the Harris fuel pools, or vice versa.12 Appendix C shows how a severe reactor accident could initiate a pool accident by precl uiing personnel access. From Appendix E it can be inferred that a pool accident could similarly preclude access to the reactor.

The Harris IPE does not analyze the implications that activation of pools C and D at Harris might have for severe accidents at the Harris reactor.

Appendix A points out that activation of pools C and D will raise two safety issues that could increase the probability of core damage at Harris. First, cooling of pools C and D and a planned uprate in reactor power will place an  ;

increased heat load on the component cooling water (CCW) system of Harris Unit 1, thus adding stress to operators and equipment at Harris, potentially increasing the probability of core damage. Second, cooling of pools C and D will create an increased load on the electrical systems at Harris, thereby adding stress to operators and equipment and potentially increasing the probability of core damage. Before activation of pools C and D is permitted, these effects should be examined through a supplement to the Harris IPE.

11 Release meegories involving significant containment failure or bypass are, in descending order of esdmated probability, RC-4, RC-5, RC4, RC-1B, RC-4C and RC-3. Each of these categories involves a 100 percent release of noble gases. 'Ihe Csl release fraction ranges from

.001 percent (RC-6) to 59 percent (RC-5).

12 As examples of literature relevant to potential safety interactions between fuel pools and reactors, see: D A lachbaum, Nuclear Waste Disposal Crisis. PennWell Books, Tulsa, OK,1996; and N Siu et al,i nam of Emt Fuel Pool CanHna PRA: Model and Results. INEL-96/0334. Idaho National Engineering laboratory, Lyta.bs 1996.

Risks & citernative options re. spent fuel stor:ge at Harris Appendix B Page B-8 1.0E- 03 3 g .:

1.0E-04 E s b

D --

~~

~

A ~-

I 1.OE-05 s E

g l

C 1.OE- 06 :

3 1.0E-07 INTERNAL SEISMIC SEISMIC FIRE LIVERMORE EPft1

& Mean 6 Median

- Note: As discussed in Reference 8.7, core damage frequencies below 1E-5 per reactor year should be viewed with caution because of the remaining uncertainties in PRA (e.g., events not con-sidered).

Source: NUREG-1150 Figure B Estimated core damage frequency for the Surry PWRs

F .

l Risks & alternative options re. spent fuel storage at Harris l

. Appendix B l Page B-9 l.

l l

l l

?e i

L o &s ,

E / / '\ 's '

? / / .

I /) ' \.', -

',/ s% ,---

ih ,% n i sin im s a . i i i n. i i ii.

l LCE-08 10E-07 10E-06 10E-05 10E-04 10E-03 10E-02 CORE DAMAGE FREQUENCY

---

SEISMIC. LIVERMORE -- - SEISMIC, EPRI - FIRE i

, Note: As discussed in Reference 8.7, core damage frequencies below 1E-5 per reactor year should be viewed with caution because of the remaining uncenainties in PRA (e.g., events not con-sidered).

Source: NUREG-1150 Figure B-2 Probability density of estimated external-events core damage frequency for the Surry PWRs

(

L

Risks & alternative options re. spent fuel storage at Harris Appendix B Page B-10 1E4

& a,

, itat a

1E4 -

'A' T

• h'$g$

, ^ .^'

I. 'Ig

$I t1 #

1E4 -

A

• &AAA

'A' ' A A A

AA 1E4 -

s 1EJ -

1E4 BWRs PWRs Source: NUREG-1560 Figure B-3 Summary of core damage frequencies as reported in IPEs

Risks & alternative options re. spent fuel storage at Harris

. Appendix B l Page B-11 i

i f

y 1E-3 3 A rt n g4 Diir Er w ae g I 1E-4 3 y' r, g, d Li, ry g J

- 1E4) h 1E4; 1E  !

o 1E a B&W CE W-2 w.3 W-.4 i

Source: NUREG-1560 f Figure B-4 i

re darnage frequencies reported in IPEs for types of PWR

RISKS AND ALTERNATIVE OPTIONS ASSOCIATED WITH SPENT FUEL STORAGE AT THE SHEARON HARRIS NUCLEAR POWER PLANT Appendix C l Potential for loss of water from the Harris pools

1. Introduction This appendix considers the potential for partial or total loss of water from one or more of the Harris fuel pools. The arrangement and use of these pools ,

are described in Appendix A. If a loss of water occurs, then exothermic l reactions could occur in the affected pools, as described in Appendix D.

2. Types of event that might cause water loss i

A variety of events, alone or in combination, might lead to partial or i complete uncovering of spent fuel in the Harris pools. Relevant types of i event include:  !

(a) an earthquake, cask drop, aircraft crash, human error, equipment l failure or sabotage event that leads to direct leakage from the pools; (b) siphoning of water from the pools through accident or malice; (c) interruption of pool cooling, leading to pool boiling and loss of water by evaporation; and (d) loss of water from active pools into adjacent pools or canals that have been gated off and drained.

3. Assessing the potential for water loss: the role of PRA A discipline known as probabilistic risk assessment (PRA) has been developed to examine the probabilities ar i consequences of potential accidents at nuclear facilities. PRA techniques are most highly developed in their application to reactor accidents, but can be applied to fuel pool accidents.

Appendix B describes the characteristics, strengths and limitations of PRA.

Carolina Power & Light Company (CP&L) has prepared a Level 2, internal-events PRA,for the Harris reactor, in the form of an Individual Plant

Risks & alternative options re. spent fuel storage at Harris Appendix C -

Page C-2 Examination (IPE). CP&L has also performed a limited assessment of the vulnerability of the Harris reactor to earthquakes and in-plant fires, in the form of an Individual Plant Examination for External Events (IPEEE). The findings of the IPE and IPEEE are described in Appendix B.

The Harris IPE and IPEEE could be extended to encompass fuel pool accidents as well as reactor accidents. Such an extension would be logical, because there are various ways in which a severe accident or a design-basis accident at the Harris reactor might accompany, initiate or exacerbate an accident at the Harris fuel pools, or vice versa.1 However, there is no current indication that CP&L will extend the IPE or IPEEE, or will otherwise apply PRA techniques to potential accidents at the Harris fuel pools.

As an indication of the need for an extended IPE and IPEEE at Harris, covering fuel pool accidents, consider a study performed for the NRC by analysts at the Idaho National Engineering Laboratory.2 These analysts examined a two-unit boiling-water reactor (BWR) plant based on the Susquehanna plant. They estimated that the plant's probability of spent fuel pool (SFP) boiling events is 5 x 10-5 per year From Appendix B it will be noted that the Harris IPE predicts a core damage frequency of 7 x 1&5 per year. (Years and reactor-years are equivalent for Harris.) The similar magnitudes of these probabi3 ties suggests that pool accidents could be a major contributor to risk at Harris, especially considering the large inventory of long-lived radioisotopes in the Harris pools.

A comprehensive application of PRA techniques to the Harris fuel pools is a task beyond the scope of the author's present work for Orange County. In the remainder of this appendix, selected issues are discussed. These discussions I

illustrate the need for a comprehensive PRA approach.

4. Analyses of earthquake and cask drop at the Robinson plant  ;

Analysts sponsored by the Nuclear Regulatory Commission (NRC) have examined the effects of a severe earthquake and a cask drop on the fuel pool at CP&L's Robinson plant.3 The Robinson plant features one pressurized- ,

water reactor (PWR) and a single fuel pool. By examining the vulnerability of 1 As examples of literature relevant to potential safety interactions between fuel pools and reactors, see: D A lachbaum, Nnche Waste Dimaamt Crisis. PennWell Books, Tulsa, OK,1996; and N Siu et al, i nae of %=nt Fuel Pool Cooline PRA: Model and Results. INEL-96/0334. Idaho National Engineering laboratory, September 1996.

2 N Siu et al, op cit.

3 P G Pransinos et al, hi=mic Failure and r=4 Dron Analyses of the Soent Fuel Pools at Two Recretative Nuche Power Plants. NUREC/CR-5176. january 1989.

l

Risks & alternative options re. spent fuel storage at Harris Appendix C Page C-3 this pool, the NRC sought to obtain knowledge that would be relevant to other PWRs.

Earthquake The NRC's analysis of the Robinson pool showed that there is high confidence (95 percent) of a low probability (5 percent) of structural failure of the poolin the event of an earthquake of 0.65g. A more severe earthquake could cause structural failure and water loss, and the mean probability of such an event was estimated to be 1.8 x 104 per reactor-year.

Cask drop The NRC's analysts examined a four-foot drop of a 68-ton fuel shipping cask ,

onto the wall of the Robinson fuel pool They estimated that the wall would I suffer significant damage. Cracking of the concrete, yield of reinforcing steel, and tearing of the liner could be expected. Loss of pool water could follow. l The probability of this cask drop was not estimated. 1 Relevance of these findings to Harris Each nuclear plant has specific design features. Thus, the findings from Robinson cannot be applied uncritically to Harris. Nevertheless, the Robinson findings suggest that the Harris fuel pools may be vulnerable to water loss in the event of a severe earthquake or a cask drop. I The Harris pools are partly below the site's grade level, and the tops of the fuel racks are at grade level. However, there are rooms and passages below the pools. Also, there are three deep cavities adjacent to the fuel handling building, where the containments for Units 2-4 were to have been constructed. Thus, the pools could drain below the tops of the fuel racks, partially or completely,if damaged by an earthquake or cask dmp.

Administrative and technical measures are employed at Harris to prevent a cask drop onto a pool wall or into a pool. There is some probability that these measures will fail and a cask drop will occur. No PRA estimate of this probability is available. An NRC-sponsored analysis found the probability of structural failure from a cask drop at the Millstone and Ginna plants, prior to improvements, to be 3 x 10-5 per reactor-year.4 After improvements, the 4V L SaGor et al, Severe An idants in Scent Fuel Pools in Sunoort of Generic Safety Issue 82.

NUREG /CR-4982. July 1987, Table 2.10.

Risks & alternative options re. spent fuel storage at Harris Appendix C Page C-4 probability was estimated to be lower than 2 x 10-8 per reactor-year. Such a low prbhility is beyond the range of credibility of PRA techniques.

5. A pool accident induced by a reactor accident The Harris IPE predicts a core damage frequency of 7 x 10-5 per reactor-year. It further predicts that 15 percent of core damage sequences would be accompanied by a significant degree of containment failure or bypass, with a total probability of about 1 x 10-5 per reactor-year.s The resulting releases could initiate a pool accident by precluding personnel access.

Radiation levels close to the plant Figure C-1 shows the estimated whole-body dose to exposed persons following a severe reactor accident.6 The dose shown is averaged over a range of meteorological conditions and a set of potential atmospheric releases (PWR 1-5) from the NRC's 1975 Reactor Safety Study. Those releases involved a cesium release fraction ranging from 1-50 percent. A similar figure could be drawn for the releases predicted by the Harris IPE, with a qualitatively similar result.

From Figure C-1 it will be seen that an unprotected person one mile from the plant will receive a whole-body dose of about 1,000 rem over one day. Closer j to the plant, the dose will be much higher, as shown in Figure C-27 It has been estimated that the dose rate within a reactor containment, following a  ;

severe accident, will be 4 million rem per hour.s Given containment failure )

or bypass, doses approaching this level could be experienced outside the containment, in locations such as the fuel handling building.

Health effects of high dose levels

- A radiation dose of 5001,000 rem will normally kill an adult person within a few weeks, due to bone marrow dam.ge. Doses of 1,000-5,000 rem will damage the gastro-intestinal tract, causing extensive internal bleeding and 5 p,i,.., categories involving significant containment failme or bypass are, in des nding order of estimated probability, RC-4, RC-5, RC4, RC-1B, RC-4C and RC-3. Each of these categones involves a 100 percent release of noble gases. 'Ihe Cs! release fraction ranges from .001 percent (RC4) to 59 perant (RC-5). -

6 Figure C-1 is adapted from Hgure 35-10 of: B Shieien, ns a r and Reennnae in F=dution An idants. US Department of Health and Human Services, August 1983.

7Mgure C-2 is adapted from Slide 16 of: J A Martin et al, Pilot Proeram: NRC Severe Reactor Accidan'Inddent Ra=nanne Trainine Manual NUREC-1210. February 1987, Volume 4.

8 R P Burke et al,In-Plant Considerations for Optunal Offsite Response to Reactor Accidents, NUREG/CR-2925. November 1982, Table B.2.

Risks & alternative options re. spent fuel storage at Harris Appendix C Page C-5 death within a few days. Doses above 10,000 rem willlead to failure of the central nervous system, causing death within a day.9 Prevention of access, and its implications f

It is clear that a severe accident at the Harris reactor, accompanied by containment failure or bypass, would preclude personnel access to the. plant.

To this author's knowledge, CPdcL has made no preparations to maintain pool cooling after such an event. It can be assumed that pool cooling would cease during the accident, and would not resume.

~

In CPacL's application for a license amendment to activate pools C and D at Harris, the bounding decay heat load for pools C and D is estimated to be 15.6 million BTU / hour (4.6 MW). CPdcL states that the mass of water in these two pools, above the racks, will be 2.9 million pounds (1,320 tonnes). Then, CPdcL estimates that the pools will begin to boil, if pool cooling systems become inoperative, after a period "in excess of 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />".10 If we assume that cooling remains inoperative, and that 4.6 MW of heat is solely devoted to boiling off 1,320 tonnes of water, then this water will be entirely evaporated over a period of 180 hours0.00208 days <br />0.05 hours <br />2.97619e-4 weeks <br />6.849e-5 months <br /> (7.5 days). In practice, a slightly longer period will be required, accounting for heat losses.

Thus, a severe reactor accident with containment failure or bypass would lead to uncovering of spent fuel in the Harris pools, after a time delay of perhaps 10 days. Heroic efforts would be needed to restore cooling or to replace evaporated water. If these efforts involved addition of water to the pools after the fuel had been uncovered, they would run the risk of exacerbating the accident by inhibiting convective circulation of air in the pools (see Appendix D).

6. A sabotage / terrorism event involving siphoning j Appendix B discusses the potential fer acts of malice at nuclear plants. A potential act of this kind at Harris would involve a group taking control of the fuel handling building, shutting down the pool cooling systems, and siphoning water from the pools. The consequent uncovering of fuel could initiate an exothermic reaction in recenny discharged fuel within a few hours (see Appendix D). Once such a reaction was initiated, access to the fuel handling building would be precluded. Over the subsequent hours, exothermic reactions would be initiated in older fuel.

9 B Flowers et al, Royal Comrrdanian on Environmental Pollution. Sixth Reoort Cmnd. 6618. Her Majesty's Stationery Office, London, September 1976, page 23.

10 License amendment application, Enclosure 7, page 5-8.

Risks & alternative options re. spent fuel storage at Harris Appendix C Page C-6 The group would require military skills and equipment to take control of the fuel handling building. Siphoning water from the pools would be a comparatively easy task. Escape by the graup would be difficult but not impossible. The probability of this scenario cannot be predicted by PRA techniques.

Risks & alternative options re. spent fuel storage ct Harris Appendix C Page C-7 103 , , , , , , , , ,, , , , , , , , , , _

06' -

102 C A _

_ 8 _

10 _

c _

E -

D _

s

. ,iliini i ti l iiii . . ili,.

3 1 10 100 1000 Distance (MNos)

Curve A Individual located outdoors withcut protection. SPs (1.0, 0.7). 1-day exposure to radionuclides on ground.

Curve B Sheltering, SPs (0.75, 0.33), 6-hour exposwe to radionuclides on ground.

Curve C Evacuation, 5-hour delay time,10 mph.

Curve D Sheltering, SPs (0.5, 0.08), 6-hour exposure to radionuclides on ground.

Curve E Evacuation, 3-hour delay time,10 mph.

Figure C-1 Estimated whole 'sody dose after a severe PWR accident

I .

Risks & citernative options re. spent fuel ster:ge at Harris Appendix C Page C-8 GENERAL RELATIONSHIP OF DOSE RATE AND DISTANCE FOR AN ATMOSPHERIC RELEASE 1.0 g  ; ,  ;

0.8 - -

Y s

@ 0.6 - -

8 54 I 0 - -

u 0.2 - -

1/r t.5 g i I l I i 0 0.5 1 2 3 4 5 .

DISTANCE (miles) i l

l l

Figure C-2 Dose-dietance relationship for a severe reactor accident l

1

\

RISKS AND ALTER 14ATIVE OFTIONS ASSOCIATED WITH SPENT FUEL STORAGE AT THE SHEARON HARRIS NUCLEAR POWER PLANT Appendix D Potential for eWheele rencHa== in the Harris pools 1

1. Introduction If water is totally or partially lost from one or more of the Harris fuel pools, the potential exists for an exothermic reaction between the fuel cladding and air or steam. The cladding is a zirconium alloy that begins to react vigorously with air or steam when its temperature readies 900-1,000 degrees C. Partial or total loss of water could cause the cladding to reach this temperature, because water is no longer available to remove decay heat from the fuel. If the cladding temperature reaches 900-1,000 degrees C and air or steam remain '

available, a runaway reaction can occur. Heat from the exothermic reaction can increase cladding temperature, which will in turn increase the reaction rate, resulting in a runaway reaction.

The steam-zirconium reaction will be familiar to many observers of the 1979 TMI accident. During that accident a steam-zirconium reaction contributed to the partial melting of the reactor core, and generated hydrogen gas.

Accumulation of this gas in the upper part of the reactor pressure vessel was a cause of concern during the accident. Hydrogen entered the containment and i exploded about 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> into the accident, yielding a pressure spike of 28 Psig.1 The potential for a partial or total loss of water from the Harris , .,ols is l addressed in Appendix C. Here, the consequent potential for exothermic  ;

reactions is considered. Also, this appendix considers the potential for exothermic reactions to release radioactive material - especially the radioisotope cesmm-137 - from spent fuel to the atmosphere outside the Harris plant.-

I 1 G Thoinpoon, Rae % = to the Petere l=1 forn=ctor Accidats W Examnie of Boilmg-Water Reactors. Institute for Resource and Security Studies, Cambndge, MA, Fettuary l I

1991.

l l

Risks & citernative options re. spent fuel storage at Harris Appendix D Page D-2

2. Configuration of the Harris pools A plan view of the Harris fuel handling building is prov'ded in Figure A-1 of Apperdix A. Figure D-1 shows a typical rack used in the Harris fuel pools.

Carolina Power & Light Company (CP&L) has not published detailed information about the dimensions and configuration of the Harris racks, claiming that this information is proprietary. The center-center distances in the Harris racks are described in Appendix A.

Figure D-2 shows CP&L's intentions regarding placement of racks b . t 51 C at Harris. It will be noted that the largest gap between the racks and tu s:o) wall will be 2.4 inches, while the gap between racks will typically be 0.6 inches.

In other words, the pool will be tightly packed with racks. Moreover, the racks will be tightly packed with fuel.

Effect of pool configuration on convective heat transfer Examination of Figures D-1 and D-2 shows that convective circulation of air or water through the racks is limited to one pathway. Water (if the pool is full) or air (if the pool is empty) must enter the racks from below and pass upward through the fuel spaces. During Phases I and II of rack placement in pool C, air or water could reach the base of the racks from parts of the pool without racks. After racks are placed in Phase III, air or water must pass downward in the gap (1.4-2.4 inches) between the racks and the pool wall, and then travel horizontally across the bottom of the pool before entering racks from below.

It is further evident that the presence of residual water in the lower part of the pool would prevent convective circulation of air through the racks, in any of the three phases of rack placement. In this case, the only significant source of convective cooling would be from steam rising through the racks.

This steam would be generated by the passage of heat from fuel assemblies to residual water, via conduction or thermal radiation.

Heat transfer pathunys Heat will be generated in the fuel assemblies by radioactive decay. Also, heat l will be generated by exothermic reactions with zirconium, if these reactions are initiated. In the event of partial or total loss of water from a pool, the following pathways will be available to remove heat from the fuel assemblies, assuming that the assemblies remain intact:

1 Risks & alternative options re. spent fuel storege at Harris Appendix D Page D-3 (a) upward convection of air (for total loss of water) or steam (for partialloss of water);

(b) upward or downward conduction along the fuel rods and rack structure; (c) upward or downward thermal radiation along the narrow passages between fuel rods, and between assemblies and rack walls; (d) upward thermal radiation from the top of the racks to the interior of the fuel handling building; (e) downward thermal radiation from the bottom of the racks to the base of the pool or to residual water (if present); and (f) lateral conduction and thermal radiation across the racks to the pool wall.

For a fuel assembly separated from the pool wall by more than a few spaces, pathway (f) will be ineffective. Thus, only pathways (a) through (e) need to be considered. In the event of total loss of water, the effectiveness of pathway (a) will depend upon the extent of ventilation in the fuel handling building.

3. A scoping appmach to heat transfer {

l To assess the effectiveness of the above-mentioned heat transfer pathways, it  ;

is appropriate to begin with a scoping analysis. Detailed calculations, especially if they involve computer nedelling, must be guided by physical  ;

insight. Scoping calculations can help to provide that insight. I Decay heat output l

The first parameter to be conddered - designated here as Q - is the decay heat in a spent fuel assembly. The unit of Q is kW per metric ton of heavy metal (MTHM) in the assembly. For PWR fuel, Q is about 10 kW/MTHM for fuel aged 1 year from discharge, and about I kW/MTHM for fuel aged 10 years.2 i Upper bound of temperature rise  ;

Now consider a fuel pellet which is in complete thermal isolation. Due to  ;

decay heat, this pellet will experience a temperature rise of 11Q degrees C per hour.3 Thus,if Q=10, the temperature rise will be 110 degrees C per hour (2,640 degrees C per day). A temperature rise of 11Q degrees C per hour is the 2For fuel bumups typical of current practice, Q will actually be 10-20 went higher than the values shown here.

3 Assuming that a uranium dbxide pellet has a specific he. t of 300 J/K per kg of pellet 040 J/K per kg of HM).

Risks & alternative options re. spent fuel storage at Harris Appendix D Page D-4 upper bound to the temperature rise that could be experienced by a fuel assembly, absent the initiation of an exothermic reaction of the cladding.

Heat transfer by conduction Next, consider conduction along the fuel rods. A Harris PWR assembly has "4 rods, each containing 1.74 kg of HM. Each rod is 12 ft long, with an outer neter of 0.374 indses, a cladding thickness of 0.0225 inches, and a pellet meter of 0.3225 inches.4 Assume that decay heat is generated uniformly

. .. ag the length of the rod, conduction along the red is the only heat transfer

.aechanism, and the two ends of the rod have the same temperature, Y (degrees C). Then, the temperature at the middle of the rod will be Y+2,000Q degrees C.s This result could be viewed as counter-intuitive, because the decay heat in each rod is only 0.48Q Watts per meter of rod.

Convective cooling by steam Now consider convective cooling of a fuel assembly by upward motion of steam that is generated from residual water at the lower end of the assembly.

Neglect other heat transfer mechanisms, assume that decay heat is generated uniformly along the length of the fuel rods, and assume that the temperature of the residual water is 100 degrees C. Define S as the submerged fraction of the assembly and T (degrees C) as the temperature of steam leaving the top of the fuel assembly. Neglect the thermalinertia of the pellets and cladding.

Then, the amount of steam generated is proportional to S, while the decay heat captured by this steam is proportional to (1-S). It follows that:6 T = 100 + (2,260/2.1)x [(1-S)/S]

Note that Q does not enter this equation. If one-tenth of a fuel assembly is submerged (S = 0.1), this equation yields a T of 9,800 degrees C. A temperature of this magnitude would not be generated in practice, because of thermal inertia and the operation of other heat transfer mechanisms.7 However, the calculation establishes an important point. Convective cooling of fuel assemblies by steam from residual water will be ineffective when the submerged fraction of the assemblies is small.

4 Harris PSAR, Section 13, Amendment No. 30.

s Assuming that the cladding's thermal conductivity is 173 W/mK, the pellets' conductivity is 1.99 W/mK, and pellets are in perfect contact with each other and the cladding.

6 Assuming that the latent heat of evaporation of water is 2,260 kJ/kg and the speofic heat of steam is 2.1 kJ/kgK.

7 The singularity of the T equadon at S=0 reflects the lack of consideradon of other heat transfer mechanisms.

~

Risks & citernative options re. spent fuel storage at Harris Appendix D Page D-5 Cooling by thermal radiation If residual water is present, Nre remains only one potentially effective mechanism of heat transfer from the mid-length of a fuel assembly - thermal radiation along the axis of the assembly. Note that a Harris PWR assembly has an active length of 12 feet, a cross-section 8.4 inches square, and contains 264 fuel rods plus other longitudinal structures. In the Harris fuel pools, the assembly will be surrounded by continuo as sheets of neutron-absorbing material (Boral), and the center-center distance in pool C will be 9.0 inches. In this configuration, axial heat transfer by thermal radiation will be strongly inhibited. However, calculations more deta3ed than those above are required to estimate the amount of heat that can be transferred by this pathway.

Note that downward heat transfer by radiation will increase the generation of steam from residual water, thus improving the effectiveness of convective cooling by steam. A detailed analysis should consider such effects through coupled calculations.

Summary The preceding scoping calculations show that conduction and convective cooling by steam will be relatively ineffective. These cooling mechanisms cannot prevent fuel cladding from reaching a temperature of at least 1,000 degrees C - the initiation point for a runaway exothermic reaction - even for fuel aged in excess of 10 years. An estimate of the effectiveness of axial radiation cooling - the only remaining cooling mechanism if residual water is present - would require more detailed calculations. However, this author does not expect that such calculations would show axial radiation cooling to be more effective than conduction or convective cooling by steam.

If residual water is not present, a fuel assembly can be cooled by convective circulation of air. Estimation of the effectiveness of this mechanism requires an analysis of convective circulation through the pool and the fuel handling building, reflecting practical factors such as constrictions at the base of fuel j 1

racks.

i

4. Specifications for an adequate, practical analysis There has been no site-specific analysis of the potential for exothermic reactions in the Harris pools. Generic analyses have been performed for and by the US Nuclear Regulatory Commission (NRC). Before addressing the findings and adequacy of the NRC's generic analyses,let us consider the

~

Risks & citernative options re. spent fuel storage at Harris Appendix D Page D 4 ingredients that are necessary if an analysis is to provide practical guidance about the potential for exothermic reactions in the Harris spent fuel pools.

Sections 2 and 3 of this appendix provide a basis for specifying those ingredients.

Partial and complete uncovering of fuel First, the analysis should not be limited to instantaneous, complete loss of water from a pool. Such a condition is unrealistic in any accident scenario which preserves the configuration of the spent fuel racks. If water is lost by drainage or evaporation and no makeup occurs, then complete loss of water will always be preceded by partial uncovering of the fuel. If makeup is considered, the water level could fall, rise or remain static for long periods.

Partial uncovering of the fuel will often be a more severe condition than complete loss of water. As shown above, convective heat loss is suppressed by residual water at the base of the fuel assemblies. As a result, longer-discharged fuel with a lower Q may undergo a runaway steam-zirconium reaction during partial uncovering while it would not undergo a runaway air-zirconium reaction if the pool were instantaneously emptied.

In a situation of falling water level, a fuel assembly might first undergo a runaway steam-zirconium reaction, then switch to an air-zirconium reaction as water falls below the base of the rack and convective air flow is established.

In this manner, a runaway air-zirconium reaction could occur in a fuel assembly that is too long diacharged (and therefore has too low a Q) to suffer such a reaction in the event of instantaneous, complete loss of water.

Conversely, a rising water level could precipitate a runaway steam-zirconium reaction in a fuel assembly that had previously been completely uncovered but had not necessarily suffered a runaway air-zirconium reaction while in that condition. The latter point is highly significant in the context of emergency measures to recover control of a pool which has experienced water loss. Inappropriate addition of water to a pool could exacerbate the accident.

Computer modelling An adequate analysis of the potential for exothennic reactions will require computer modelling. The modelling should consider bout partial and complete uncovering and the transition from one of these states to the other.

Also, the modelling should cover: (a) thermal radiation, conduction, and steam or air convection; (b) air-zirconium and steam-zirconium reactions; (c) variations along the fuel rod axis; (d) radial variations within a representative fuel rod, including effects of the pellet-cladding gap; and (e) clad swelling and

Risks & alternative options re. spent fuel storege at Harris Appendix D Page D-7 rupture. Experiments will probably be required to support and validate the modelling.

Site-specific factors The analysis can be strongly influenced by site-specific factors. For convective cooling by air, these factors include the detailed configuration of the racks, the pools and the fuel handling building. All relevant factors should be accounted for. This could be done through site-specific modelling.

Alternatively, generic modelling could be performed across the envelope of site-specific parameters, with sensitivity analyses to show the effects of varying those parameters.

Propagation of exothermic reactions to adfacent assemblies After an exothermic reaction has been initiated in a group of fuel assemblies, this reaction might propagate to adjacent assemblies. Due to their lower Q or to other factors, the adjacent assemblies might not otherwise suffer an exothermic reaction. An analysis of propagation should consider the potential for reactions involving not only the fuel cladding but also material (e.g., Boral) in the fuel racks. The analysis should examine the implications of clad and pellet relocation after a reacting assembly has lost its structural integrity. Those implications include the heating of adjacent assemblies and racks by direct contact, thermal radiation, convection, and the inhibition of air circulation. A bed of relocated material at the base of the pool could have all these effects.

5. The 1979 Sandia study An initial analysis of the potential for exothermic reactions was made for the NRC by Sandia Laboratories in 1979.8 This was a respectable analysis as a first attempt. It considered partial drainage of a pool, although it used a crude heat transfer model to study that problem, and neglected to consider the steam-zirconium reaction. It did not address the potential for propagation of exothermic reactions to adjacent assemblies. The Sandia authors were careful to state their. assumptions and to specify the technical basis for their computer modelling.

Figure D-3 illustrates the findings of the Sandia study. The three lower curves in Figure D-3 show the sensitivity of convective air cooling to the diameter of the hole in the base of the fuel racks. The next higher curve - the 8A S Benjamin et al, Soent Fuel Heatuo Following Imss of Water During Storace. NUREG /CR-Ofd2, March 1979.

Risks & alternative options re, spent fuel storsge at Harris Appendix D Page D-8

~

" blocked inlets" case - shows the suppression of convective air cooling due to the presence of residual water. The dashed curve shows the effect of an air-zirconium reaction. The runaway nature of that reaction is evident.

Note that the analysis underlying Figure D-3 assumed a cylindrial rack arrangement with a center-center distance of about 13 inches. Also, the analysis assumed a gap of 16 inches between the racks and the pool wall. The Harris racks are more compact and are packed more tightly into their pools.

These facturs will tend to inhibit convective air moling at Harris.

6. Subsequentstudies The 1979 Sandia study could have been the first of a series of studies that I moved toward the level of adequacy specified in Section 4. Since 1979 the NRC has sponsored or performed a variety of studies related to the initiation of exothermic reactions in fuel pools.' However, the scope of these studies has narrowed, and their potential for building on the 1979 study has not been realized.

Failure to consider partial uncovering A major weakness of the NRC's studies since 1979 has been their focus on a postulated scenario of total, instantaneous loss of water. 'Ihis appendix shows clearly that partial uncovering of fuel will often be a more severe condition than complete loss of water. Thus, however sophisticated the NRC's modelling of spent fuel heatup might be, the findings have limited relevance to the practical potential for exotherm;c reactions.

Brookhaven National Laboratory (BNL) has developed the SHARP code 50 I

replace the SFUEL code first developed at Sandia. BNL authors have claimed that the SHARP code can more accurately predict spent fuel heatup in realistic spent fuel pool configurations.10 A review of the SHARP code is beyond the scope of this report. Applied to spent fuelin a generic, high-density configuration in an instantaneously emptied pool, the SHARP code finds that the fuel cladding will reach a " critical" temperature (565 degrees C) if aged less than 17 months for PWR fuel or 7 months for BWR fuel.11 The relevance of this finding to the Harris pools is unclear. -

9See, for example: V L Sador et al, Severe Ahk in %nt Fuel Paale in Samnart of FM Safety Issue 82. NUREC/CR-4982. July 1987; and R J Travis et al, A Safety and Regulatory Asaenament of Generic BWR and PWR Permanentiv Shutdown Nudear Power Plants.

August 1997.

R J Travis et al, page 3 4.

11 lbid.

Risks & alternative options re. spent fuel storage at Harris Appendix D .

Page D-9 Propagation of exothermic reactions Pursuant to a Freedom of Information request, the NRC released in 1984 a so-called draft report by MIT and Sandia authors on the propagation of an air-zirconium reaction in a fuel pool.12 This document has been repeatedly cited ,

in subsequent years, although it should properly be regarded as notes toward a l draft report. Those notes were submitted to the NRC after the project ran out of funds;it was never completed.

The MIT-Sandia group concluded from computer modelling and experiments that an air-zirconium reaction in fuel assemblies could propagate to adjacent, lower-Q assemblies. They expressed the view that propagation would be l quenched in regions of a pool where fuel is aged 3 years or more, but noted the presence of "large uncertainties" in their analysis.

BNL analysts subsequently reviewed these experiments and conducted their own modelling using the same code (SFUEL). In their modelling the BNL analysts chose to terminate the air-zirconium reaction when the cladding i reached its melting point.13 Neither the MIT-Sandia group nor the BNL group examined the implications of clad and pellet relocation after a reacting assembly has lost its structural integrity. The author is not aware of other analyses which address this problem. Thus, the specifications set forth in Section 4 for analysis of propagation have not been met.

1'

7. The potential for an atmospheric release of radioactive material Spent fuel at Harris which suffers an exothermic reaction will release  !

rcdioactive material to the fuel handling building. That building is not j designed as a containment structure, and is not likely to be effective in this role, given the occurrence of exothermic reactions in one or more pools. A BNL study has concluded that a reasonable, generic estimate of the release fraction of cesium isotopes, from affected fuel to the atmosphere outside the plant, is 100 percent.14 This release fraction is used in Appendix E. ,

The amount of fuel that will suffer an exothermic reaction, given a loss of water from the Harris pools, will depend upon the particular scenario. For

- scenarios which involve partial uncovering of fuel, the reaction could affect fuel aged 10 or more years. For scenarios which involve total loss of water, 12N A Pisano et al,The Potential for Propagation of a Self-Sustainine Zirconium Ovidation Followina Y- of Wmw in a hmt Puel Stormee Pool. Draft Report, January 1964.

13 V L Sailor et al.

14 mid.

Risks & alternative options re. spent fuel storage at Harris Appendix D Page D-10 the reaction will be initiated only in younger fuel, perhap: aged no more than 1-2 years. However,if clad / pellet relocation is properly factored into a propagation analysis, this analysis may show that a reaction will propagate to much older fuel Appendix E considers two potential releases of cesium-137 from the Harris pools. One release corresponds to an exothermic reaction in fuel aged 9 years or less. The other release corresponds to a reaction in fuel aged 3 years or less.

Risks & alternative options re. spent fuel storage at Harris

. Appendix D Page D-11 N \ '

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Risks & alternative options re. spent fuel storage at Harris

, Appendix D Page D-13 isee , , , , ,

PWR SPENT FUEL IN CYLINDRICAL BASKETS 1808 . 1 YEAR MINIMUM DECAY TIME -

3808 -

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Source: NUREG/CR-0649 Figure D-3 Estimated heatup of PWR spent fuel after water loss

~

RISKS AND ALTERNATIVE OPTIONS ASSOCIATED WITH SPENT FUEL STORAGE AT THE SHEARON HARRIS NUCLEAR POWER PLANT Appendix E Consequences of a large release of cesium-137 from Harris

1. Introduction This appendix outlines some of the potential consequences of postulated large releases of cesium-137 from the Harris plant to the atmosphere. Such consequences can be estimated by site-specific computer models. A simpler approach is used here, but this approach is adequate to show the nature and scale of expected consequences.

2. Characteristics of postulated mienses Two spent fuel release scenarios are postulated here. The first scenatio involves a release of 2.3 x 107 Curies (870,000 TBq) of cesium-137, with a mass of 260 kilograms.1 This represents the cesium-137 inventory in Harris' stock of spent fuel aged 3 years or less, as estimated in Appendix A. The second scenario involves a release of 7.1 x 107Curies (2,600,000 TBq) of cesium-137, with a mass of 790 kilograms. This represents the cesium-137 inventory in Harris' stock of spent fuel aged 9 years or less. Note that all of the cesium-137 I in the affected fuel is assumed to reach the atmosphere, an assumption which is explained in Appendix D.

Releases of the postulated magnitude could occur as a result of exothermic reactions in the Harris fuel pools. Appendix D discusses the potential for such reactions. Cesium-137 would not be the only radioisotope released to the atmosphere if exothermic reactions occurred in the pools. However, cesium-137 is likely to be the dominant cause of offsite radiological exposure, 1 1 Curieis equivalent to 37 x 10-2TBq.1 TBq of cesium-137 is equivalent to 03 grams.

Risks & alternative options re. spent fuel storage at Harris ^

Appendix E Page E-2 just as it dominates the offsite exposure attributable to the 1986 Chernobyl reactor accident.2 Note that cesium-137 has a half-life of 30 years.

A severe accident at the Harris reactor could also release cesium-137 to the

- atmosphere. Appendix A notes that the US Nuclear Regulatory Commission (NRC) has estimated the inventory of cesium-137 in the core of the Harris reactor, during normal operation, to be to be 4.2 x 106Curies (155,000 TBq, or 47 kilograms). As summarized in Appendix B, an individual plant examination (IPE) study by Carolina Power oc Light Company (CPicL) has identified six categories of potential significant release due to severe accidents I

at the Harris reactor. Release category RC-5, the most severe release category,-

would involve a release to the atmosphere of 53-59 percent of the cesium isotopes in the reactor core. Thus, given the NRC's estimate of core inventory, release category RC-5 would involve an atmospheric release of 2.2-2.5 x 106 Curies (82,000-92,000 TBq, or 25-28 kilograms) of cesium-137.

Chernobyl and weapons testing releases For comparison with the above-mentioned potential releases, consider two actual releases - from the Chernobyl accident and from atmospheric testing of nuclear weapons. The 1986 Chernobyl reactor accident released about 90,000 TBq (27 kilograms) of cesiv 737 to the atmosphere, representing 40 percent of the cesium-137 in the r + t core.3 Through 1980, about 740,000 TBq (220 kilograms) of cesium-137 were deposited as fallout in the Northern Hemisphere, as a result of atmospheric testing of nuclear weapons.4 Note that the fallout from weapons testing was distributed over a larger area then the fallout from the Chernobyl accident, and a larger fraction of it descended on oceans and lightly inhabited areas.

3. Contamination of land A usefulindicator of the consequences of a cesium-137 release is the area of contaminated land. Here, contamination is measured by the external (whole-body) radiation dose that people will receive if they live in a contaminated area. When cesium-137 is deposited from an airborne plume, it will adhere to the ground, vegetation and structures. From these locations, it will emit gamma radiation which provides an external radiation dose to an exposed person. Cesium-137 will also enter the food chain and water sources, thereby 2US Department of Energy, Health & Environmental Conseauences of the Chernobyl Nuclear Power Plant Acadent. DOE /ER-0332. June 1987; A S Krass, Consequences of the Chernobyl Accide Institute for Resource & Security Studies, Cambridge, MA, December 1991.

3 Krass, op cit.

4 US Department of Energy, op cit.

Risks & alternative options re. spent fuel storage at Harris Appendix E Page E-3 providing an internal radiation dose to a person living in the contaminated area. Absent any countermeasures, the internal dose could be of a similar magnitude to the external dose.

Figure E-1 shows the relationship between contaminated land area and the size of an atmospheric release of cesium-137. This figure is adapted from a 1979 study by Jan Beyea, then of Princeton University.5 The threshold of contamination is an external dose of 10 rem over 30 years, assuming a shielding factor of 0.25 and accounting for weathering of cesium. The " typical meteorology" case in Figure E-1 assumes a wind speed of 5 m/sec, atmospheric stability in class D, a 0.01 m/sec deposition velocity, a 1,000 m j mixing layer and an initial plume rise of 300 m (although the results are not l sensitive to plume rise). A Gaussian, straight line plume model was used, providing an estimate of contaminated land area that will approximate the area contaminated during a range of actual meteorological conditions. The >

lower and upper limits of land contamination in Figure E-1 represent a range of potential meteorological conditions.

The threshold for land contamination An external exposure of 10 rem over 30 years would represent about a three-fold increase above the typical level of background radiation (which is about 0.1 rem / year). In its 1975 Reactor Safety Study, the NRC used a threshold of 10 rem over 30 years as an exposure level above which populations were assumed to be relocated from rural areas. The same study used a threshold of 25 rem over 30 years as a criterion for relocating people from urban areas, to reflect the assumed greater expense of relocating urban inhabitants.

In an actual case of land contamination in the United States, the steps taken .

to relocate populations and pursue other countermeasures (decontamination )

of surfaces, interdiction of food supplies, etc.) would reflect a variety of l political, economic, cultural, legal and scientific influences. It is safe to say that few citizens would calmly accept a level of radiation exposure which l substantially exceeds background levels. i land contamination from potential Harris releases TLree potential Harris releases of cesium-137 are shown in Figure E-1.  ;

Releases of 70 million Curies and 20 million Curies correspond to liberation  ;

} Beyes,"Ihe Effects of Releases to the Atmosphere of Radioactivity from Hypothetical Large-Scale Axidents at the Proposed Gorieben Waste Treatment Facility", in Chapter 3 of ;

Rennet of the Cortahan inearnational Review. presented (in German) to the Government of Lower Saxony, March 1979.

- Risks & alternative options re, spent fuel storage at Harris Appendix E Page E-4 of cesium-137 from spent fuel aged up to 9 years or up to 3 years, respectively.

A release of 2 million Curies corresponds to the most severe reactor accident '

identified in the Harris IPE.

For typical meteorology, Figure E-1 indicates that a release of 2 million Curies would contaminate 4M5,000 square kilometers of land, A release of 20 million Curies would contaminate 50,000-60,000 square kilometers. Finally,'a release of 70 millica Curies would contaminate about 150,000 square kilometers of land. Note that the total area of North Carolina is 136,000 square kilometers and the state's land area is 127,000 square kilometers.6 Potentially exposed population According to CP&L's Final Safety Analysis Report (FSAR) for the Harris plant, an estimated 1.8 million people will live within 50 miles of the plant in 2000, while 2.2 million people will live within that radius in 2020.7 A 50 mile-radius circle encompasses an area of 20,300 square kilometers.

If a substantial release of cesium-137 occurs at Harris, the shape and size of the resulting contaminated area will depend on the size of the release and the meteorological conditions during the period of the release. If the wind direction is constant during the release and the atmosphere remains stable, the contaminated area will be comparatively narrow and extended downwind. Changing wind direction during the release period and a less stable atmosphere will produce a more " smeared out" pattern of contamination.

A computer modelling exercise could be performed, to predict patterns of contamination under different meteorological conditions. This exercise could ascribe a probability, assuming a postulated release, that a particular population falls within an area contaminated above a specified threshold.

4. Health effects of radiation The health effects of exposure to ionizing radiation can be broadly categorized as early and delayed effects. For our postulated releases of cesium-137, early health effects could be suffered by some people in the immediate vicinity of the plant. However, most of the health effects would be delayed effects, i especially cancer, which are manifested years after the initial exposure.

6'the Wodd Almanac and Book of Facts 1991. Phams Books, New York,1990.

7 Hams PSAR, Section 2.1.3, Amendment No. 2.

, l Risks & alternative options re. spent fuel storage at Harris Appendix E Page E-S Note that a release during a reactor accident (e.g., release category RC-5 at Harris) will contain short-lived radioisotopes as well as cesium-137. Under >

certain mnditions of meteorology and emergency response, the presence of these short-lived radioisotopes in the release could cause many early health effects. Spent fuel contains comparatively small amounts of short-lived radioisotopes. Thus, early health effects are comparatively unlikely if a release occurs from a spent fuel pool.

Table E-1 shows an estimate of the excess cancer mortality attributable to continuous exposure to a relatively low radiation dose rate. This estimate was made by the BEIR V committee of the National Research Council.s In Table E-1, a continuous exposure of 1 mSv/ year (0.1 rem / year) is assumed to occur throughout life.' Such an exposure is estimated to increase the number of fatal cancers, above the normally avp-tad level, by 2.5 percent for males

. and 3.4 percent for females, with an average of 16-18 years of life lost per excess death. If the dose-response function were linear,it would follow that continuous, lifetime exposure to 10 mSv/ year (I rem / year) would increase j the number of fatal cancers by 25 percent for males and 34 percent for females.

The shape of the dose-response function is a subject of ongoing debate.

If people continued to occupy urban areas contaminated with cesium-137 to an external exposure level just below 25 rem over 30 years, as was assumed in the Reactor Safety Study, their average exposure during this 30-year period would be 8 mSv/ year (0.8 rem / year). An additional, internal exposure would arise from contamination of food and water. After 30 years, rates of external and internal exposure would decline, consistent with the decay of cesium-137.

Note that over a period of 300 years (10 half-lives), the activity of cesium-137 will decay to one-thousandth of its initial level.

5. Economic consequences of a release of radiaetivity Computer models have been developed for estimating the economic consequences of large atmospheric releases of radioactive materials. Findings from such models have been used by the NRC to evaluate the cost-benefit ratio of introducing measures to reduce the probabilities or consequences of spent fuel pool accidents.10 A review of these models, findings and cost-8 National Research Coundl, Hamith Effack of F=<ume to Iow Imvels of Ior%e Pa*=dewu BEiE.y., National Academy Pues, Washmgton, DC,1990. TaNe E-1 is adapted from Table 4-2 of the BEIR V report.

9 lhe exposure of 1 mSv/ year is addhional to background radiation, whose effects are accounted for in the normal expectation of cancer mortality.

10See, for example: E D Thmm, Engulatory Analvais for the Resolution of Generic lasue 82.

"Beyond Demian Basis Acddents in Spent Fuel Pools. NUREG-1353. April 1989; and J H Jo et al,

Risks & alternative options re. spent fuel stor:ge at Harris Appendix E Page E-6 benefit analyses is beyond the scope of this report. However, a brief examination of the NRC's literature reveals that findings in this area rest on '

assumptions and value judgements that are not clearly articulated and deserve thorough public review.

Previous sections of this appendix have shown that potential releases of cesium-137 from the Harris spent fuel pools could lead to the relocation of large populations and ongoing radiation exposure to large, unrelocated populations. Relocation implies abandonment of large amounts of land, other natural resources and fixed capital. Political and social effects would be significant, and would have economic implications. Useful analysis of these matters would require a more sophisticated approach than is evident in literature generated by and for the NRC.

Value/I=nact Analvmes of Accidant Preventive and Mitientive Options for Soent Fuel Pools.

NUREG/CR-5281. March 1989.

Risks & titernative options re. spent fuel stor:ge at Harris Appendix E Page E (

109 2,,,, ,

i i g,,iiii i g.,,,,, ,

i ,

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3 Risks & alternative options re. spent fuel storage at Harris Appendix E '

Page E-8 l

ESTIMATED LIFETIME RISK PER 100,000 PERSONS EXPOSED TO 1 mSv PER YEAR, CONTINUOUSLY THROUGHOUT LIFE Males Females -

• Point estimate of excess 520 600 mortality

• 90 percent confidence limits 410480 500-930

• Normal expectation 20,560 17,520

• Excess as percent of normal 2.5 3.4

• Average years of life lost per 16 18 excess death Table E-1 ',

Excess cancer mortality from continuous exposure to radiation: -

BEIR V estimate