ML18017A893

Risks & Alternative Options Associated with Spent Fuel Storage at Shearon Harris Nuclear Power Plant.

ML18017A893
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Site:	Harris
Issue date:	02/28/1999
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INSTI'ICE FOR RESOURCE AND SECURITY STUDIES 27 Ellsworth Avenue, Cambridge, Massachusetts 02139, USA Phone: (617) 491-5177 Fax: (617) 491-6904 Electronic mail: irss@igc.apc.org RISKS AND ALTERNATIVEOPTIONS ASSOCIATED WITH SPENT FUEL STORAGE AT THE SHEARON HARRIS NUCLEAR POWER PLANT A report prepared for Orange County North Carolina by Gordon Thompson February 1999 99i0060329 9'F0930 05000400 PDR ADQCK PDR

Acknowledgements This report was prepared as part of a program of work by the Institute for Resource and Security Studies (IRSS) pursuant to a contract between IRSS and Orange County, North Carolina. The report was written by Gordon Thompson, the executive director of IRSS.

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 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,

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'hat 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.

About IRSS 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.

Table of contents Introduction

2. Present status of the Harris nuclear plant Proposed activation of fuel pools C and D

4. Types of potential accident af 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

Risks & alternative options re. spent fuel storage at Harris Page 1

1. 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 discussion in the appendices.

What is spent fuel?

Figure 1 shows a fuel assembly of the type that is used in the Harris reactor.i 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 descripbons 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 presented (Secbon 8), followed by a discussion of regulatory processes (Section 9). Conclusions are presented in Section 10.

i Figure 1 is adapted from: A V Nero, A uidebook t Nucl ar Reactor University of California Press, 1979, page 79.

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

2. Present status of the Harris nudear 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 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 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 necessary 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.

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

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

3. Proposed acbvation of fuel pools C and D CP&L seeks an amendment to its operating license so that it can activate pools 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 willbe 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. 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 racks in pools A and B (9.0 inches instead of 10.5 inches). This highly compact arrangement allows more PWR fuel to be placed in a given pool area but also has adverse implications for safety.

Cooling and electrical supply for pools C and D The water in a spent fuel pool must be cooled and cleaned. Cooling 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 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 1 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 form, 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 1 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

Risks & alternative options re. spent fuel storage at Harris Page 5 address this situation, CP&L proposes an "alternative plan". fo demonstrate that the previously completed piping and other equipmenf 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 planf 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. Af 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 accidents more severe than the design basis are

'redible. 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 nuclear plant.

These plans allow for large releases of radioactive material, of fhe kind that were identified in the Reactor Safety Study. The Chernobyl reactor accident of 1986 further demonstrated the potential 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 nuclear plants has been guided by a hybrid set of assumptions. Many areas of safety regulation rely upon the assumption thaf accidents will remain within the design basis.

Other areas, such as emergency response planning, assume that severe accidents can occur.

Pool-reactor interactions 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 af the Harris reactor could accompany, initiate or exaceibate an accident at the Harris pools, or vice versa. The NRC has been slow to examine fhe 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 QPE). 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 Planf Examination for External Events (IPEEE).

The Harris IPE and IPEEE could be extended to encompass fuel pool accidents as well as reactor accidents. 'uch 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 wiH otherwise apply PRA techniques to potential accidents at the Harris fuel pools.

Risks & alternative options re. spent fuel storage at Harris Page 7

5. Design-basis pool accidents 'I The Harris FSAR considers two types of 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 FSAR estimates that the release of radioactivity would be relatively small. This report does not review the FSAR analysis.

In its license amendment application, CP&L has considered some other potential accidents, induding the dropping of a rack or a fuel pool gate.

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 s 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 in pools A and B at Harris. CP&L s proposed acbvabon of pools C and D will involve an even higher density of storage. Such high-density configurations inhibit heat loss from the fuel if 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. First, 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 mater A variety of events could cause partial or total loss of water from the Harris pools. These events deserve the level of analysis that would be provided by a thorough PRA. Performing a pool accident PRA is beyond the scope of our License amendment'application, Enclosure 7.

A potential accident in this class, which deserves analysis, would involve the placement of a low-burnup or high-enrichment PWR assembly in the racks in pools C or D.

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 CP&L 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 personnel 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 difficultbut not impossible. The probability of this event cannot be predicted by PRA techniques.

Initiation of exothermic reactions, given mater loss Since the late 1970s, the NRC has sponsored and performed a variety of 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 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.

Partial loss of water can be expected in many scenarios, rather than instantaneous, total loss of water. Partial loss 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 probability estimate should be accompanied by a range of uncertainty. Even with the inclusion of uncertainties, PRA-derived estimates represent lower bounds to actual accident probabilities.

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 potenbal 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 from 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 consequences. The approach is described in Appendix E.

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 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) 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

Risks & alternative 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 MCi) corresponds to a pool accident affecting fuel aged up to 9 years after discharge from a reactor.

Land contamination by cesium-137 Accident 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 5 An 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 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 dose-response function were linear, it would follow that continuous, lifetime exposure to 1 rem/year would increase the number of fatal cancers by 25 S Without countermeasures such as interdiction of food supplies, the internal dose could be of a similar 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 &e 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. CP&L'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 systematic, 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 CP&L's nuclear plants. That option, which could be combined with other options for storage of CP&L'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 dass 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 storage 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

Risks & alternative options re. spent fuel storage at Harris Page 12 described in Table 2.6 If willbe noted from 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'ry 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 more resistant to sabotage than others.

All of the 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 proposed 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.

Lotto-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 discharged fuel. The fuel would remain v'ulnerable to consolidation through a cask drop into a pool or a 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, if is unlikely that this reaction would be propagated to other regions of a pool.

Tables1and2areadapted from: USNuclear Regulator Commission, Inf rmati nDi t 1 E i n -1 V lum 1 November 1998.

MGRaddatzandMDWaters,lnf rmati nHandbo k nInde endentS ntFuelS ra In ll n NURE -1 71 December 1996.

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 alteinatives 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 been 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 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 af plant sites. This move could provide an opportunity for a thorough assessment of risks and options, and for the adoption of safer means of fuel storage.

10. Conclusions 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 exothermic reactions of fuel cladding, in any or all of pools A, B, C and D.

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

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.

C3 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 accident at the Harris reactor.

C4 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.

CS 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.

C6 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 Page 16 Storage Design Storage Design Copaaty Appreml Vendor Model mblies Date Generol Nuclear Mdol Cask Systems, Incorporated CASTOR V/21 21 PWR 09/30/1985 08/17/1990 CASl'OR X/28 28 PWR 04/22/1994 CASTOR X/33 33 PWR 11/24/1995 Tronsnudear, Concrete Module West Incorporated NUHOMS.7P 03/28/1986 Westinghouse Eledric Metal Cask 24 PWR 09/30/1987 08/17/1990 MC-10 FW Energy icotions, Concrete Vault 83 PWR or 03/22/1988 I Modular Vault 150 BWR Dry Storage (MVDS)

NAC fntemotionol, Inc. Metal Cosk 26 PWR 03/29/1988 08/17/1990 NACS/T NAC International, Inc. Metal Cask 28 Canisters 09/29/1988 08/17/1990 NAC<28 S/T (fuel rods from 56 PWR assemblies)

Tronsnudear, Meta)Col 11/04/1993 Incorporated TN 24 24 PWR 07/05/1989 TN-32 32 PWR 11/07/1996 NAC International, Inc. Metal Cask 28 PWR 02/01/1990 NAC-128/ST Sierra Nvdeor Ventilated Cosk 24 PWR 03/29/1991 05/03/1993 Corporation VSC-24 Transnudear West, Inc. Concrete Module 04/21/1989 01/18/1995 Standardized NUHOMS-24P 24 PWR NUHOMS-528 52 BWR NAC Intemationol, Inc. NAC-STC 26 PWR 07/17/1995 Note: PWR - Pressurized-Water Reador; BWR - Boiling-Water Reador Table 1 NRC-approved dry spent hxel storage designs

Risks & alternative options re. spent fuel storage at Harris Page 17 Reactor Name Date Storage Issued Vendor Model VIII'urry 1, 2 07/02/1986 Generals Nudear MAC k Virginia Elediic 8 Systems, CASTOR V/21 Power Company tncorporated TN-32 NAC-128 CASTOR X/33 MC-10 H. B. Robinson 2 08/13/1986 Transnvdear West, Concrete Module Carolina Power & Ught Incoporated NUHOMS-7P Company Oconee 1, 2, 3 01/29/1990 Transnvdeor West, Concrete Module Duke Energy Company Incorporated NUHOMS.24P Fort St.

Vrain'ublic 11/04/1991 FW Energy icotions, Modular Vault Service Compony In Diy Store of Colorado Cd ~Cliff 1,2 11/25/1992 Tronsnudear West, Concrete Module 8altimore Gos 8 Incorporated NUHOMS-24P Eledric Company Palisades Under General license Pacific Sierra Nudear Ventilated Cask Consumers Energy Associates VK-24 Proirie bland 1, 2 10/19/1993 Transnvdear West, Metal Cask Northern States Power Inooipo rated 1N.40 Company Point Reoch Under Generol Ucense Sierra Nudear Ventilated Cask Wisconsin Eledric Power Coqxe6on VSC4f Company Davis-Sesse Under General Ucense Transnvdear West, Concrete Modvle Tdedo Edison Company NUHOMS-24P Incorporated Arkonsos Nudear One Under General Ucense Siena Nudear Corporation Ventiloted Cask Enlergy Operations VK-24 North Anna 06/30/98 Transnvdear West, Metal Cask Virginia Eledric 8 Power Incorporated TN.32 Company

'Plant undergoing decommissioning Table 2 NRC dry spent fuel storage licensees

RISKS AND ALTERNATIVEOPTIONS ASSOCIATED WITH SPENT FUEL STORAGE AT THE SHEARON HARRIS NUCLEAR POWER PLANT.

Appendix A Spent fuel management at the Harris plant

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 here is drawn from CP&L's application to amend the Harris license, from CP&L's Final Safety Analysis Report (FSAR) for the Harris plant, or from viewgraphs shown by CP&L personnel during meetings with staff of the Nuclear Regulatory Commission (NRC).>

2. Present and proposed spent fuel storage capacity 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. 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 1 Meetings between NRC staff and CP&L representatives, to discuss the proposed license amendment, were held on 3 March 1998 and 16 July 1998.

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 cleanup systems for pools C and D were never completed.

Pool A now contains six racks (360 fuel assembly 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 offload capacity in the pools at Brunswick and Robinson.

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.

CP&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 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 en'richment and high burnup.>

~ License amendment application, Enclosure 5.

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

3. Support services for pools 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 terbary 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, CPkL 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 I CCW system already provides cooling to pools A and B and serves other, important safety functions. For example, the Unit I 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 I 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 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 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).~ At present, the Unit I CCW system cannot absorb this additional heat load.

Thus, CP&L proposes to include in the Technical Specifications for Hams 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 I 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 I CCW system, so that this 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 fhe present license amendment application, Apparently, CP&L intends to perform the upgrade of the Unit I CCW system concurrent with a power uprate for fhe Unit I reactor. A 4,5 percent power uprate of fhe reactor willbe 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 I CCW heat load, including the reactor power uprate and the ongoing use of pools C and D, will substantiaHy 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 fhe margin in the Unit I CCW sysfem, 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 License amendment application, Enclosure 7, page 5-16.

4 License amendment application, Enclosure 5.

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

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.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 witNn 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 .

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 8 These numbers suggest that accommodating a design-basis LOCA will already exploit the margin of the CCW system, without any additional load 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 s License amendment application, Enclosure 9.

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

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

Harris FSAR, section 92, Amendment No. 40.

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 piping was completed before the second Harris reactor was cancelled 9 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 address this situation, CP&L proposes an Alternative Plan to demonstrate that the previously completed piping and other equipment is adequate for its purpose.>o 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 thaf pools C and D willbegin 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.<> 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 1 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.>> 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, if is necessary to eshmate the potential inventory of the radioisotope cesium-137 in the Harris pools. As a starbng 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.

1 License amendment application, Endosure 8.

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

Harris FSAR, page 9.1~, Amendment No. 48.

13VLSailoretal, vere A 'd n in ntFuel P lsin u rt f n ri f Isu 2 t >>.*

Risks & alternative 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 cesium-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 fuel in pools C and D until it 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 96 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'tock of spent fuel aged 3 years or less willbe 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'tock of spent fuel aged 9 years or less will be 784 PWR assemblies and 1/24 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'tock of spent fuel aged 3 years or less will contain 2.3 x 10" Curies (870,000 TBq) of cesium-137, with a mass of 260 kilograms. 'Also, the Harris pools'tock of spent fuel aged 9 years or less will contain 7.1 x 10" Curies (2,600,000 TBq) of cesium-137, with a mass of 790 kilograms.

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

15 Scarola of CPS', presentation to Orange County Board of Commissioners, 9 February 1999.

J

Risks & alternative options re. spent fuel storage at Harris Appendix A Page A-8 CP&L could provide a more precise projection of the cesium-137 inventory in the Harris pools over coming years. However, our esbmate 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'nventory 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).>6 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.

US Nuclear Regulatory Commission, Final Environmental Statement Related to the ra' f h a nHarri Nu I rP werPlantUnitsland2 NURE 72 October1983.

Risks & alternative options re. spent fuel storage at Harris Appendix A Page A-9 OF UNIT 4 CONThINIIENT FUEL TRANSFER ( NOT CONSTRUCTED)

CANhLS

~ NORTH POOL B 2i I OF UNIT 1 CONTAINKENT OF UNIT 3 CONTAINI(ENT FUEL TRhNSFER (NOT CONSTRUCTED) EQUIPItENT CANALS HATCH N

CASK LOh D INg POOL C ~

p POOL I

76x

( NOT CONSTRUCTED )

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

Risks & alternative options re. spent fuel storage at Harris Appendix A Page A-10 FUEL POOL CLEANUP SYSTEH SPENT FUEL POOL t

DB4IHERALIZEPS FILTERS FUEL POOL COOLING SYSTEH I HTERHEOI ATE HEAT EXCHANGERS REACTOR BUILOIHG CLOSEO COOLENG MATER SYSTEH OR C COHPONENT COOLING HATER SYSTEH HEAT SIHK (LAKE, RIVER, OCEAN. COOLING TOHER, OR COOLING POND)

Source: M;EEG-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 360 363 723 2178 2946 927 2763 3690 co'7 1025 0 1025 Total 3080 5304 Source: License amendment application Table A-I 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 P%R Fuel Assemblies BWR Fuel Assemblies Number of Assys Decay Period Number of Assys Decay Period 172 5 years 456 5 years 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 stoic of fuel assemblies in Harris pools C and D

RISKS AND ALTERNATIVEOPTIONS ASSOCIATED WITH SPENT FUEL STORAGE AT THE SHEARON HARRIS NUCLEAR POWER PLANT Appendix B Potential for severe accidents at the Harris reactor I. Introduction In examining the risks associated with spent fuel storage at Harris, one must corisider 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 assessment The probabilities and consequences of potential accidents at nuclear facilities can be estimated through the techniques of probabilistic risk assessment (PRA). Nuclear facility 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 1 findings, a PRA will estimate the nature of potential radioactive releases from the facility. In

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-initiating 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. 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, willbe a lower bound to the true probability, which willbe impossible to determine.

NUREG-1150 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 externaHy-initiated accidents to build upon revious analysis of internally-initiated accidents.

US Nuclear Regulatory Commission, R r f A H-14 14 October 1975.

US Nuclear Regulatory Commission, v r A 'd nt Risks: An A m n f r Fi u 1 arP w rPIant -11 2 v l December 1990.

4 H Hirsch, T Einfalt, 0 Schumacher and G Thompson, IAEA fe Tar and Pr babilisti Ri k A ment Gesellschaft fur Okologische Forschung und Beratung, Hannover, August 1989.

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 earthquakes (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 the 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 shown in Figure B-2.

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, rather 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, 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 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 threat 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 terr'orism has moved from far-fetched horror to a contingency that could happen next month. Although the United States still takes conventional terrorism seriously, as'demonstrated by the response to the attacks on its embassies in Kenya and Tanzania in terrorism.

G Thompson, National University, October 1996.

.t November/December 1998, page 80.

August, it is not yet prepared for the new threat of catastrophic ar T rr ri m and Nuclear P wer Plants Peace Research Centre, Australian

Risks & alternative options re. spent fuel storage at Harris Appendix B Page B-4 The effectiveness of licensees'rrangements to resist terrorist attacks on nuclear 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'ntiterrorism efforts'.

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

Carolina Power and Light (CP&L) 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). CP&L submitted an IPEEE for the Harris reactor in 1995. 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 CP&L, 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 and B-2. The IPE predicts that accident sequences involving a loss-of-coolant accident (LOCA) will account for 40 percent of Harris'ore 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).

7 8 8Carolina Power &LightCompany, h r nHarri Nu 1 rP rPI n UnitN .1:

Individual Plan Examinati n u mittal August 1993.

Carolina Power 8c Light Company, h ar n Harris Nucl ar P wer Plant nit No. 1:

Individual Plant Examinati n f r Ex mal Ev n ubmittal June 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.io Some of the results are shown in Figures B-3 and B-4.

Figure B-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 &

Wilcox (B&W). 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 recirculabon 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

& 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 the RLE involved housekeeping and minor modifications, and these actions have been taken. The IPEEE did not invesbgate the implications of an earthquake more severe than the RLE.

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 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.

USNuclear Regulatory Commission, In ividu 1Pl n Examinati n Pr am: P r v n r f n Plant P rf rman e -1 v 1 December 1997.

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 CsI 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 OSLOCA). 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 the containment. This loss of water would initiate the accident, and the water's escape pathway would provide a route for the escape of radioactivity 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 the 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< per reactor-year. Note that the overall probability of core damage is estimated to be 7x10-5 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

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

4. Pool-reactor interacbons Neither CP&L nor NRC have performed an analysis to determine 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.>> Appendix C shows how a severe reactor accident could initiate a pool accident by precluding 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 willraise two safety issues that could increase the probability of core damage af 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 categories involving significant containment failure or bypass are, in descending order of estimated probability, RCR, RC-5, RC4, RC-1B, RC-4C and RC-3. Each of these categories involves a 100 percent release of noble gases. The CsI release fraction ranges from

.001 percent (RC4) to 59 percent (RC-5).

1~ As examples of literature relevant to potential safety interactions between fuel pools and reactors, see: D A Lochbaum, Nuclear Wa Di 1 ri is PennWeil Books, Tulsa, OK, 1996; and NSiu etal, L f n Fu I F l lin PRA: M l n R ul INEL-96 Idaho National Engineering Laboratory, September 1996.

Risks & alternative options re. spent fuel storage at Harris Appendix B Page B-8 1.0E" 03 C

0 1.0E" 04 D

A M

A 1.0E-OG F

R E

0 U

N 1.0E-06 C

Y 1.0E-07 INTERNAL SEISMIC SEISMIC FIRE LIVERMORE EPRI 8'ean 8 Median Note: As discussed in Reference 8.7, core damage frequencies below 1E-S per reactor year should be viewed with caution because of the remaining uncertainties in PRA (e.g., events not con-sidered).

Source: NUlKG-1150 Figure B-1 Estimated core damage frequency for the Surry PWRs

Risks & alternative options re. spent fuel storage at Harris Appendix B Page B-9 P

R 0

8 A

8 I

l T

Y 1.0E-08 1.0E-OT 1.0E-06 'I.OE"05 1.0E-04 1.0E-03 1.0E-02 CORE DAMAGE FREQUENCY SEISMIC, LIVERMORE SEISMIC, EPRI FIRE 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-2 Probability density of estimated external-events core damage frequency for the Surry PWRs

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

k kk 1EQ k

I k kk 4 kki 0 k kkg kkkk krak

a. 1EW k k

kkk 0'EW k k

'0 O

O O

1E-7 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 Page B-7.1

~ 4E4 he 5 1EC Le 1EN f SEA E CE7 v0 4E-8 B8W CE W-2 W-3 WQ Source: NUREG-1560 Figure B-4 Core damage frequencies reported in IPEs for types of PWR

RISKS AND ALTERNATIVEOPTIONS ASSOCIATED WITH SPENT FUEL STORAGE AT THE SHEARON HARRIS NUCLEAR POWER PLANT Appendix C Potential for loss of water from the Harris pools Introduction This appendix considers the potential for parbal 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 reactions could occur in the affected pools, as described in Appendix D.

2. Types of event that might cause water loss A variety of events, alone or in combination, might lead to partial or complete uncovering of spent fuel in the Harris pools. Relevant types of event include:

(a) an earthquake, cask drop, aircraft crash, human error, equipment 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 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.

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 QPE). 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. However, there is no current indication that CP&L willextend 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 ¹tional Engineering Laboratory.> 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 willbe noted that the Harris IPE predicts a core damage frequency of 7 x 10-5 per year. (Years and reactor-years are equivalent for Harris.) The similar magnitudes of these probabilities 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 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> The Robinson plant features one pressurized-water reactor (PWR) and a single fuel pool. By exanmrdng'the vulnerability of 1 As examples of literature relevant to potential safety interactions between fuel pools and reactors, see: D A Lochbaum, Nuclear aste Dis I sis PennWell Books, Tulsa, OK, 1996; andNSiuetal,L f n Fu I P 1 lin PRA:M d land R ul INEL- Idaho National Engineering Laboratory,'eptember 1996.

> N Siu et al, op ot.

PGPrassinosetal, i mi F ilur n kDr Anal f ntFu 1P I atTw R r n 'v Nu 1 arP werPlan -517 January 1989.

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 pool in 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 1& 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 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.

The probability of this cask drop was not estimated.

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.

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 drop.

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

~t 4VLSailoretal, vr A dn in n FulF 1 in u r f nri f I u 2

Risks & alternative options re. spent fuel storage at Harris Appendix C Page C-4 probability was estimated to be lower than 2 x 1& per reactor-year. Such a low probability 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 5 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 willbe seen that an unprotected person one mile from the plant willreceive a whole-body dose of about 1,000 rem over one day. Closer 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. Given containment failure or bypass, doses approaching this level could be experienced outside the containment, in locationssuch as the fuel handling building.

Health effects of high dose levels A radiation dose of 500-1,000 rem will normally kill an adult person within a few weeks, due to bone marrow damage. Doses of 1,000-5,000 rem will damage the gastro-intestinal tract, causing extensive internal bleeding and Release categories involving significant containment failure or bypass are, in descending order of estimated probability, RC-4, RC-5, RC-6, RC-1B, ROC and RC-3. Each of these categories involves a 100 percent release of noble gases. 'Ihe CsI release fraction ranges from .001 percent Radi nA 't (RC-6) to 59 percent (RC-5).

Figure C-1 is adapted from Figure 35-10 of: B Shleien, P r n n R US Department of Health and Human Services, August 1983.

Figure C-2 isadapted fromSlide16of: JA Martinetal, Pil Pr am: NR S vere Rea or n in A In n R Tr inin M nu I E -121 February1987,Volume4.

RP Burkeetal, In-Plan C nsiderati ns f r timal ff ite Re onse to Reactor Accident

Risks & alternative options re. spent fuel storage at Harris Appendix C Page C-5 will lead to failure of the

'eath within a few days. Doses above 10,000 rem central nervous system, causing death wiQun a day 9 Prevention of access, and its implications 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 Qus author's knowledge, CP&L 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 CP&L'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 esQmated to be 15.6 million BTU/hour (4.6 MW). CP&L states that the mass of water in these two pools, above the racks, will be 2.9 million pounds (1,320 tonnes). Then, CP&L estimates that the pools will begin to boil, if pool cooling systems become inoperabve, 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+20 tonnes of water, then Qus 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 slighQy longer period willbe 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 Appendix B discusses the potential for 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 recently discharged fuel within a few hours (see Appendix D). Once such a reaction was inibated, access to the fuel handling building would be precluded. Over the subsequent hours, exothermic reactions would be initiated in older fuel.

9BHowersetal,Ro al Commi 'on n Envir nmentalP llution ixth Re rt Cmnd. 1 Her Majesty's Stationery Office, London, September 1976, page 23.

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 group would be difficultbut not impossible. The probability of this scenario cannot be predicted by PRA techniques.

Risks & alternative options re. spent fuel storage af Harris Appendix C Page C-7 103 Bs' ceQ o CL 102 C In ~

4l C 10 Set o~

10 100 Distance (Miles)

Curve A Individual located outdoors without protection. SF's (1.0, 0.7). 1-day exposure to radionuclides on ground.

Curve B Sheltering, SF's (0.75, 0.33), 6-hour exposure to radionuclides on ground.

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

Curve D Sheltering, SF's (0.5, 0.03), 6-hour exposure to radionuclides on ground.

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

Figure C-I Estimated whole-body dose affer a severe PWR accident

Risks & alternative options re. spent fuel storage at Harris Appendix C Page C-8 GENERAL RELATIONSHIP OF DOSE RATE AND DISTANCE FOR AN ATMOSPHERIC RELEASE 1.0 0.8 I-g 0.6 O

O l- 0,4 0.2 1/r1 5 0 0.5 1 2 3 DISTANCE (milesj Figure C-2 Dose-distance relationship for a severe reactor accident

RISKS AND ALTERNATIVEOPTIONS ASSOCIATED WITH SPENT FUEL STORAGE AT THE SHEARON HARMS NUCLEAR POWER PLANT Appendix D Potential for exothermic reactions in the Harris pools X. 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 reaches 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 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 pslg.1 The potential for a partial or total loss of water from the Harris pools is 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 cesium-137 from spent fuel to the atmosphere outside the Harris plant.

1 G'Ihompson, R lat R n t the Fot n 'al f r Reac r Accidents: The Exam le f B ilin - at r R a r Institute for Resource and Security Studies, Cambridge, MA, February 1991.

Risks & alternative 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 provided in Figure A-1 of Appendix 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 in pool C at Harris. It willbe noted that the largest gap between the racks and the pool wall willbe 2.4 inches, while the gap between racks will typically be 0.6 inches.

In other words, the pool willbe tightly packed with racks. Moreover, the racks willbe 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 pathways Heat willbe generated in the fuel assemblies by radioactive decay. Also, heat 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:

Risks & alternative options re. spent fuel storage at Harris Appendix D Page D-3 (a) upward convection of air (for total loss of wafer) or steam (for partial loss 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) willbe 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 approach to heat transfer 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 modelling, must be guided by physical insight. Scoping calculations can help to provide that insight.

Decay heat output The first parameter to be considered 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/MTHMfor fuel aged 1 year from discharge, and about 1 kW/MTHMfor fuel aged 10 years.>

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.> 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 2 For fuel burnups typical of current practice, Q willactually be 10-20 percent higher than the values shown here.

Assuming that a uranium dioxide pellet has a specific heat of 300 J/K per kg of pellet (340 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 264 rods, each containing 1.74 kg of HM. Each rod is 12 ft long, with an outer diameter of 0.374 inches, a cladding thickness of 0.0225 inches, and a pellet diameter of 0.3225 inches.4 Assume that decay heat is generated uniformly along the length of the rod, conducbon along the rod is the only heat transfer mechanism, 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 5 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 thermal inertia 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 that6 T = 100+ (2@60/2,1) x [(I-S)/SJ Note that Q does not enter this. equation. If one-tenth of a fuel assembly is submerged (S = 0.1), ttus equation yields a T of 9+00 degrees C. A temperature of ttus magnitude would not be generated in practice, because of thermal inertia and the operation of other heat transfer mechanisms'owever, 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 FSAR, Section 13, Amendment No. 30.

Assuming that the cladding's thermal conductivity is 17.3 W/mK, the pellets'onductivity is 1.99 W/mK, and pellets are in perfect contact with each other and the cladding.

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

7 The singularity of the T equation at SW reflects the lack of consideration of other heat transfer mechanisms.

Risks & alternative options re. spent fuel storage at Harris Appendix D Page D-5 Cooling by thermal radiation If residual water is present, there 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 continuous sheets of neutron-absorbing material (Boral), and the center-center distance in pool C willbe 9.0 inches. In this configuration, axial heat transfer by thermal radiation will be strongly inhibited. However, calculations more detailed 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 af 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 effecbve 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 racks.

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 & alternative options re. spent fuel storage at Harris Appendix D Page D-6 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-discharged (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 exothermic reactions will require computer modelling. The modelling should consider both 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 storage 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 adjacent 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 theirassumptions 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

~ A S Benjamin etal, March 1979.

S nt Fuel Heatu Foll win Loss of ater Durin Stora e NUREG CR-

Risks & alternative options re. spent fuel storage at Harris Appendix D Page D-8 "bio'eked 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 reacbon. The runaway nature of that reaction is evident.

Note that the analysis underlying Figure D-3 assumed a cylindrical 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 factors will tend to inhibit convective air cooling at Harris.

6. Subsequent studies The 1979 Sandia study could have been the first of a series of studies that 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 9 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. This 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 exothermic reactions.

Brookhaven National Laboratory (BNL) has developed the SHARP code to 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 fuel in a generic, high-density configuration in an instantaneously emptied pool, the SEVQU? 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.>> The relevance of this finding to the Harris pools is unclear.

See,forexample: VLSailoretal,Sev reA in ntFu 1P lsin u f n ri f I u 2 -4 2 July1987;and RJTravisetal, A Safe and R lat A n f n ricBWRandPWRP rman n hutd wnNu learPowerPlants NURE 1 August 1997.

R J Travis et al, page S4.

ll ibid.

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

The MIT-Sandia group conduded 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 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 reached its melting point.>s 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.

7. The potential for an atmospheric release of radioacbve material Spent fuel at Harris which suffers an exothermic reaction will release radioactive material to the fuel handling building. That building is not 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.>4 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 parbcular 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,

>2NA Pisano etal, Th P n ' f r Pr a tion of a Self-Sustainin Zirc nium ida' F ll win f at ring n Fu t ra Pool DraftReport,January1984.

1 ts V L Sailor et al.

~4 ibid.

Risks & alternative options re. spent fuel storage at Harris Appendix D Page D-10 the reaction will be initiated only in younger fuel, perhaps 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

)I

)I (I )I I )I I

)I )I )I I I (I II lj II (I

)I (I

)I

)I

)I (I l

)1 Source: License amendment application Figure D-I Typical rack used in the Harris pools

Risks & alternative options re. spent fuel storage at Harris Appendix D Page D-12 NORTH 597.88 EIH.

1.43" 2.28" III III P11ASE PIIASE PllASE III PTR PER PIR 9X9 9X9 9X9 PHASE III P11ASE I1 I DIR DIR 13 X 13 13 X 13 320.60 MIH.

PHASE 111 DXR 13 X 13 2.44"

.625" EIH.(TYP) 1.44" 1.43" Source: License amendment application Figure D-2 Proposed rack placement in Harris pool C

Risks & alternative options re. spent fuel storage at Harris Appendix D Page D-13 ISDP PWR SPENT FUEL IN CYLINDRICAL BASKETS IGPP 1-YEAR MINIMUM DECAY TIME 140P BLOCKED INLETS

( NO OXIDATION ASSUMED )

OXIDATION IEPP EF'FECT FOR NO WATER,

= 1. 5' lapp Dh I I NO WATER, Dh = 1.5" I

~ Q I g spa Cl SQO 4DD =3.0" NOWATER, Dh hole I

goo NO WATER, Dh 5.0" hole I

p 0 16 24 32 40 48 TIME AFTER POOL DRAINAGE (Hrsj Source: MJREG/CR-0649 Figure D-3 Estimated heatup of PWR spent fuel after water loss

RISKS AND ALTERNATIVEOPTIONS ASSOCIATED WITH SPENT FUEL STORAGE AT THE SHEARON HARRIS NUCLEAR POWER PLANT Appendix E Consequences of a large release of cesium-137 from Harris X. 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. Characterisbcs of postulated releases Two spent fuel release scenarios are postulated here. The first scenario 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'tock of spent fuel aged 3 years or less, as estimated in Appendix A. The second scenario involves a release of 7.1 x 107 Curies (2,600,000 TBq) of cesium-137, with a mass of 790 kilograms. This represents the cesium-137 inventory in Harris'tock of spent fuel aged 9 years or less. Note that all of the cesium-137 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 Curie is equivalent to 3.7 x 10 TBq. 1 TBq of cesium-137 is equivalent to 03 grams.

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

Page E-2 just as it dominates the offsite exposure attributable to the 1986 Chernobyl reactor accident.> 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 106 Curies (155,000 TBq, or 47 kilograms). As summarized in Appendix B, an individual plant examination QPE) study by Carolina Power & Light Company (CP&L) has identified six categories of potential significant release due to severe accidents 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 cesium-137 to the atmosphere, representing 40 percent of the cesium-137 in the reactor 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 than the fallout from the Chernobyl accident, and a larger fraction of it descended on oceans and lightly inhabited areas.

3. Contamination of land A useful indicator 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 US Department of Energy, Health 8e Environmental n u n f th em b l Nu lear P rPI n A ' E ER 32 June1987; ASKrass, n u n f rn

~~in 3 Krass, Institute for Resource 8c Security Studies, Cambridge, MA, December 1991.

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 mixing layer and an initial plume rise of 300 m (although the results are not 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 (decontaminabon 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.

Land contamination from potential Harris releases Three potential Harris releases of cesium-137 are shown in Figure E-1.

Releases of 70 million Curies and 20 million Curies correspond to liberation I Beyea, The Effects of Releases to the Atmosphere of Radioactivity from Hypothetical Large-Scale Accidents at the Proposed Gorleben Waste Treatment Facility", in Chapter 3 of R rt f rl nIn rnati nal R vi w presented(inGerman) to the Governmentof Lower Saxony, March 1979.

Risks & alternative options re. spent fuel storage at Harris Appendix E Page E-4

'I 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 4,000-5,000 square kilometers of land, A release of 20 million Curies would contaminate 50,000-60,000 square kilometers. Finally, a release of 70 million 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.<

Potentially exposed population According to CPSs 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 20207 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, especially cancer, which are manifested years after the initial exposure.

6Th rI Alm n nd k fF 1 1 Pharos Books, New York, 1990.

" Harris FSAR, Section 2.1.3, Amendment No. 2.

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

Page F;5 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 conditions 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 comparabvely 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 CounciL8 In Table E-1, a continuous exposure of 1 mSv/year (0.1 rem/year) is assumed to occur throughout life 9 Such an exposure is estimated to 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 dose-response function were linear, it would follow that continuous, lifetime exposure to 10 mSv/year (1 rem/year) would increase 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 radioactivity 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.>0 A review of these models, findings and cost-National Research Council, Heal h Eff

~BEIR fE ur to w L vel f I nizin Radia ':

National Academy Press, Washington, DC, 1990. Table E 1 is adapted from Table 4-2 of the BEIR V report.

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

See,forexample:EDThrom,R I t Anal i f r h R luti n f n ri I sue 2 n D B i A n in nFuIP 1 1 April 1989; and 7 H 7o et al,

Risks & alternative options re. spent fuel storage 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 dearly articulated and deserve thorough public review.

Previous sections of tNs 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.

lu Im atAnl fA '

Pr n v andMi'v 'n fr n FulP 1 NURE - 2 1 March 1989.

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

- Page E-7 1O9 Lower Limit Typical Meteorology C4 0

8 10 Upper 0 70 MCi 20 MCi 6

10 O

I M

pf 2 MCi O

lO6 IO 10 Sq. km of land contamination Figure E-1 Contaminated land area as a function of cesium-137 release

Risks & alternative options re. spent fuel storage at Harris Appendix E Page E-8 ESTIMATED LIFETIMERISK PER 100,000 PERSONS EXPOSED TO mSv I PER YEAR, CONTINUOUSLYTHROUGHOUT LIFE Males Females

~ Point estimate of excess 520 600 mortality

~ 90 percent confidence limits 410-980 500-930

~ Normal expectation 20+60 17+20

~ Excess as percent of normal 2.5 3.4

~ Average years of life lost per 16 18 excess death Table F 1 Excess cancer mortality from continuous exposure to radiation:

BEIR V estimate