ML20099K804

Intervenor Exhibit I-Joint-EP-48,dtd 840416,consisting of Transcript of Sc Sholly Re Emergency Planning Contention 11. Prof Qualifications Encl

ML20099K804
Person / Time
Site:	Catawba
Issue date:	05/24/1984
From:	Sholly S CAROLINA ENVIRONMENTAL STUDY GROUP, PALMETTO ALLIANCE, UNION OF CONCERNED SCIENTISTS
To:
References
I-JOINT-EP-48, OL, NUDOCS 8411290573
Download: ML20099K804 (39)

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{{#Wiki_filter:50'W3t//VOL. f f g. g j, q SD 9' Cf Oh I D, pf s ty d x. w / L'.I ('1 UNITED STATES OF AMERICA -{ ^ 3Q$ .,ddh ~ NUCLEAR REGULATORY COMMISSION 7 ':\\ ~~, c'll50$ El,. Q R, B_E_F O_R_E__T_!_! E__A_T_C_)M_ _I C__S_A_F_ In the Matter of ) ) Docket Nos. 50 -413 DUKE POh'ER COMPANY, ET AL. ) 50-414 ) (Catawba Nuclear Station, Units ) 16 April 1984 1 and 2) ) PALMETTO ALLIANCE AND CAROLINA ENVIRONMENTAL STUDY GROUP TESTIMONY OF STEVEN C. SilOLLY ON EMERGENCY PLANNING CONTENTION NUMBER ELEVEN as mur usar m. g.yV(.7.n cumsun WW -Staff-Intervencr( 3d'difi'dk'ulf f Aeceived Date-ib Rejected ~ hDR Q Reporter: ek#f(Ad ~ O y si

<y a7,- q o UNITED STATES OF-AMERICA NUCLEAR REGULATORY COMMISSION SEEEEE.1EE.k1EE19.EbEE11. BEE.b1EEEE1EE.EEbEE

In the Matter'of

) ) Docket Nos. 50 -413 DUKE PCWER COMPANY, ET AL. ) 50-414 )-

(Catawba. Nuclear Station, Units

) 16 April 1984 1 and 2) ) PALMETTO ALLIANCE AND CAROLINA ENVIRONMENTAL STUDY GROUP TESTIMONY OF STEVEN C. SHOLLY ON EMERGENCY PLANNING CONTENTION NUMBER ELEVEN Q.01 Would you please state your name, position, and business address? A.01 My name is Steven C. Sholly. I am a Technical Research Associate with the Union of Concerned Scientists (UCS) in Washington, D.C. My primary responsibility with UCS is in technical and policy analysis concerning risk assessment and emergency planning. My business address is: Union of. Concerned Scientists, Dupont Circle

Building, 1346 Connecticut

Avenue, N.W.,

Suite

1101, Washington, D.C.

200 36. Q.0 2 lla v e you prepared a statement of professional qualifications? A.0 2 Yes. My statement' of professional qualifications is attached to this testimony. '.t~ e

1 -1 - F '- ' O.03. What1is the purpose of your testimony? A.0 3 This testimony, which is - sponsored jointly by the Palmetto Alliance and the Carolina Environmental Study Group, addresses Emergency Planning Contention 11. That contention, as admitted by the Atomic Safety and Licensing Board. in it's Memorandum and Order of 29 September 1983, is worded as follows: The ' size and configuration of 'the northeast quadrant of the plume exposure pathway emergency planning zone (Plume EPZ) surrounding the Catawba facility has not been properly determined by State and local officials in relation to local emergency response.needs and capabilities, as required by 10 CFR 50. 47 (c) (2). The boundary of that zone reaches, but does not extend past the Charlotte city limit. There is a substantial resident population in the southwest part of Charlotte near the present plume EPZ boundary. Local meteorological conditions are such that a serious accident-. a t the Catawba facility would endanger the residents of that area and make their evacuation _. _ _ _ _ _...prudentT' The 'likely flow of ~ evacuees 1 f rom the' present plume EPZ through Charlotte access-routec also indicates the need for evacuation planning for southwest Charlotte. There appear to be suitable plume EPZ boundaries inside the city limits, for example, highways 74 and 16 in southwest Charlotte. The boundary of - the northeast quadrant of the plume EPZ should be reconsidered and extended to take account of-these demographic, meteorological and access route considerations. 0 04 Khat is the plume exposure pathway emergency planning zone? A.0 4 The plume exposure pathway emergency planning zone' sur' rounding (" plume EPZ") is an area a nuclear power plant for which emergency response plans are required in order -to assure that prompt and effective actions can be taken to protect the public in the event of an accident [ E'

..- L ~ from-two principal pathways: (a) whole ~ body.' external exposure to gamma radiation. from the plume and from deposited materials, and (b) inhalation exposure from the passing radioactive plume.- The plume EPZ should be about 10 miles -in radius (NUREG-0396, pp. 27-28; NUREG-0654, Rev. 1, pp. 8-10 ). 0 05 What is the overall objective of emergency response planning for nuclear power reactors? A.0 5 The overall objective of emergency response planning for nuclear. power reactors is to provide does savings (and in some cases immediate life savings) for a spectrum of accidents that could produce offsite doses in excess of Protective Action Guides 1/ [ NUREG-0 65 4, Rev. 1, p. 6]. Q.0 6 What protective actions for the general public are available to avoid or minimize exposures from the dose pathways of concern for the plume EPZ? A.0 6 The principal protective actions available for the general public to avoid whole body' and inhalation exposures are: expeditious movement of the a. Evacuation population before plume passage to avoid exposure from a radioactive plume and exposure due to ground contamination by deposition from the plume; expeditious movement of the b. Relocation population from contaminated areas after plume passage to avoid further exposure from-ground contamination; c. Sheltering -- expeditious movement of the -population indoors before plume paesage to reduce exposure from a radioactive plume and acute ground contamination by deposition from-the plume, and 'to reduce inhalation exposure during plume passage (used in conjunction with relocation.); t; ~-

gl ,c g/ _.use .by the' q

d. ; Respiratory' ~ protection

~-- L population; of measures. to.. reduce ~ inh'alation ~ -jexposure'during' plume ~ passage;1and .s u e. . Thyroid bloc king. -- use by; th'e _ populati'on - -(before L plume -passage) of t potassium : iodide ~ -to blockjthe uptake of-radioactive iodine.by. the ' thyroid _. gland. - The choice = of protective actions i in any._given acc'ident~ situation ~ depends on ' a. number of f actors,' including the . magnitude. and. composition of tihe. release from the plant' (i.e., the sou'rce term), weather conditions at the time ~of and subsequent-to the r e l e a s e,-- the amount of time available before plume passage,-the distance-of_ populated areas-- f rom the _ pla'nt site, - the speed with'- which various protective actions can be implemented, and ' the level of protection afforded by various protective actions. Q.0 7 What-is the spectrum of -potential accidents at.the Catawba Nuclear Station? A.0 7 The spectrum of potential accidents at the Catawba Nuclear ' Station-range f rom relatively trivial plant upsets through. accidents involving severe core damage and large-scale melting of the core and subsequent' breach of the containment. This spectrum of accidents is sometimes split into two large categories -- accidents within the design basis and accidents exceeded the design basis.. Actual accident experience.to date in nuclear. power plants is briefly reviewed in the NRC Staff's Final Environmental Statement on the Catawba Nuclear Station (FES-Catawba) [NUREG-0 921]. Other-references describe 2 additional incidents in some detail ini both commercial nuclear-: plants Jand experimental reactors [ORNL/NSIC-176; ORNL/NSIC-217 draft; and NUREG/CR-2497]'. ~ u f b o .4g w .L d e, 1 4 Q.0 '8 What is the significance of-this spectrum of potential accidents for emergency planning? A.0 8 Nuclear power plants built in the U.S. are conservatively designed to respond to ~ accidents as severe. as design basis accidents without sustaining severe core damage. The general approach to this design process is based;on the principal of providing multiple barriers to the release of fission products to the environment referred to as the " defense in depth" concept. For the purposes of siting, extremely conservative design basis accident evaluations are mandated.. The - dose calculations for such evaluations are generally governed by the procedures set forth in a 1962 publication of the former U.S. Atomic Energy Commission [ TID-14844). Using a number of assumptions regarding the source term (i.e., the quantity and chemical form of radioactive materials available for release from containment), performance of e ng i nee r ed-sa f e ty~ f~e a t u r e s,- ~ Y l u nie"~d i~s~p e Es'i o n ; ~ T n d ~ ~ ~ ~~~ ~ protective actions, calculated doses from design basis accidents must be demonstrated to be less than 25 Rem whole body and 30 0 Rem to the thyroid from iodine exposure for a two-hour period at the exclusion area boundary and the entire period of plume passage at the low population zone boundary.2/ In contrast, realistic evaluations of design basis accidents result in exposures significantly lower than these guideline levels. For e xa n.ple, the NRC Staff's L FES-Catawba provides such calculated doses for design basis accidents at Catawba [NUREG-0 921, p. 5-79). The largest calculated doses for Catawba design basis accidents are 0.06 Rem whole body and 0.07 Rem to the thyroid at' the exclusion area boundary. Not only are these doses significantly less than the siting guideline t6'

~ [ 10 CFR -- 100.11 (a ) (1) and (a) (2)], they are only doses ~ small f ractions of the' Protective Act' ion Guide doses (6% and 1. 4 %,- respectively, for whole body and thyroid exposures). ~ Thus, even if! these calculated doses are optimistic by'a factor of

ten, the estimated doses from a realistic evaluation of design basis accidents at Catawba will not exceed the Protective Action Guide doses at the exclusion area boundary.

This observation leads to the conclusion that design basis accidents are not significant with respect to offsite emergency response. As a practical matter, should a de' sign basis accident actually occur, offsite officials may decide to implement precautionary protective measures such as sheltering or a limited evacuation of areas.near the plant until conditions are stabilized and the potential for a release of radioactivity to the environment has diminished. For accidents beyond the design basis, a range of possible offsite doses and consequences is possible. It is conceivable that a severe core damage accident could be successfully " bottled up" by the containment so long as containment heat removal systems function adequately and excessive amounts of noncondensible gases are not generated. On the other

hand, accidents beyond the design-basis could result in core melting and the release of ~ radioactive materials to the environment ranging in quantity from trivial to.very large.

The magnitude of the release will depend upon the degree of core damage, the operating history of the core, the performance (or lack thereof) of engineered safety -features, and the timing and mode of containment failure. r

_7_. z. Q.0 9 ' What magnitude of radiation exposure could result from core-melt accidents in which the containment fails in the absence of emergency response? A.0 9 A recent report from Sandia National Laboratories provides one' perspective on accidents involving core melt with containment failure. Using the release -categories for a pressurized water reactor from the Reactor Safety ' Study (RSS) [ WASi!-1400, Appendix VI), Sandia calculated bounding doses from such releases. The dose calculations were carried out using the CRAC2 accident consequence model [NUREG/CR-2326; NUREG/CR-255 2; and NUREG/CR-2 901], and provided estimates of whole body and thyroid doses at a distance of one mile from'the release point assuming no protective actions for 48 hours. The doses presented represent the " peak" or maximum calculated doses based on 10 0 weather sequences. The doses thus calculated were [NUREG/CR-2925, p. 34]: _ _.. _ _ _ _ _ ~ ~ - - - RELEAS F - ' ' - ~ WHOLE' BODY ~~~ ~TilYROID~ - ~~~ '~ ~ CATEGORY DOSE (REM) DCSE (REM) 0 PWR-7 1x 10 5 x 10 PWR-6 6x 10 2x 10 3 PWR-5 1x 10 8x 10 4 PWR-4 5x 10 3 x 10 4 PWR-3 2 x 10 2 x 10 4 4 PWR-2 7 x 10 7 x 10 PWR-1A 8x 10 9x 10 Obviously, these accumulated dose levels would not.be . permitted to accumulate -- protective actions would be implemented to reduce the doses. The results do point out the need for protective actions (compared' with the Protective Action Guide dose levels of 1-5 Rem whole body 9% a 9. b

• n.

_)

.\\ .and 5-25 Rem thyroid) in core melt accidents in which the conta,inment fails. The results also indicate that sheltering is not.an adequate long-term protective action in areas close to the site for the more severe release categories (this is due to both the large initial exposure during plume passage and the accumulation of exposure from radioactive materials deposited from the plume on the ground during plume passage). 0 10 What are the implications of the above for emergency planning for reactor accidents? A.10 It can be concluded from the above information that core, melt accidents dominate public risk considerations, and therefore, to a considerable extent, drive the size and configuration of the emergency planning zone. This'is in accord with prior conclusions of probabilistic risk assessments such as the Reactor Safety Study [ WASH-1400 ) and a comparative risk evaluation of accidents within and '~ ~ ~~ ~ exceeding ~ths design basis [ NUREG/CR-0 60 3 ]. " ~ ' ~~ ~~ ~

Indeed, NRC regulations and joint NRC/ FEMA emergency planning guidance reference NUREG-0396 as providing the technical basis for the size of the plume EPZ.

This report is in turn based to a significant extent on a related Sandia Laboratories report [NUREG/CR-1131]. The dose versus distance and accident consequence calculations presented in NUREG-0396 and NUREG/CR-1131 are explicitly based on the characteristics of core melt accident release categories from tbc Reactor Safety Study. Thus, we need to look to analyses of of fsite doses and consequences for core melt accidents at Catawba to gain perspective on the size and configuration of the plume EPZ. 4 9 e

i .c, 0 11 Which reactor served as-the model for the calculations in .NUREG-0396 and NUREG/CR-1131? A.ll The accident probabilities' and release characteristics used in NUREG-0396 and NUREG/CR-ll31 are based on the results of the. Reactor Safety Study [WASil-1400 } analysis of a pressurized water reactor. The Surry Unit 1 reactor served as the surrogate in that analysis for all pressurized water reactors in the U.S. 0.12 Briefly describe the Surry Unit 1 reactor and contrast it with the Catawba Nuclear Station reactors. A.12 Surry Unit 1 is a three-loop Westinghouse pressurized water reactor with a thermal power output of 2,441 MWt. The plant has a dry subatmospheric containment with a design pressure of 45 psig. The Catawba reactore are four-loop Westinghouse 'eactors with ~a thermal ~ power' output of ' ~ ' - ~ pressurized watbr r 3,412 mkt. The Catawba plants have ice condenser containments with a design pressure of 15 psig. There are differences in design and the number and type 2 of equipment provided in the two plants. These dif ferences can be determined by comparing the Final Safety Analysis Reports and Safety Evaluation Reports for the facilities. 0 13 llow do the differences between Surry Unit 1 and the Catawba Nuclear Station reactors affect their performance in scvere core damage or core melt accidents? A.13.The NRC Staff's FES-Catawoa states that the design and operating characteristics of the two plants are similar

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.~. i .[NUREG-0921, p. 5-36].. This may be accurate for normal operating conditions.. For performance under severe core damage or core melt accidents, however, the performance of the two plants can be expected to be different. Ideally,- a probabilistic risk-assessment (PRA) of-the Catawba reactors would demonstrate this quite well,-but no such analysis of the Catawba reactors has been prepared. The next best choice is a PRA performed on a facility similar to the Catawba reactors. A PRA of the Sequoyah~ Unit 1 reactor was prepared by Sandia. National ~ Laboratories for.the NRC under the Reactor Safety Study Methodology Applications Program (RSSMAP) in 1980 [NUREG/CR-1659, Vol. 1). Sequoyah Unit' 1 is, like the Catawba

reactors, a

3,411 MWt four-loop Wectinghouse pressurized water reactor with an ice condenser containment. It would be reasonable to expect similar performance under severe a6cident conditions for Catawba and Sequoyah. There are two potentially 'important caveats here. The first is that the Sequoyah RSSMAP study did not consider so-called " external events" as accident initiators -- e.g., carthquakes, hurricanes, fires, etc. Because the events classified as " external events" are site-and plant-specific, the effects of such accident initiators are likely to be different for the Catawba and Sequoyah plants, despite their similarities in design. In addition, there may be plant-specific features for Catawba that would result in differences between Sequoyah and Catawba in severe accident performance. Nonetheless, absent a plant-specific PRA for the Catawba reactors, the .RSSMAP PRA for Sequoyah represents the best available 'I S9.

J.' s; _11_ 3 . guidance a s_ to the performance-characteristics of the Catawba reactors under severe accident conditions.

The : dif ferences in _ severe accident performance between Surry Unit 1 and Sequoyah Unit 1 (and, to the-extent that the plants are

similar, Catawba Units 1 and 2) were clearly identified _in the Sequoyah RSSMAP report:

Accident sequences involving trans'ients were found to be important for Surry (indeed, one of the three dominant sequences was TMLB', a station blackout sequence). Only one -transient accident sequence appears in the list of dominant accident sequences for Sequoyah [NUREG/CR-1659, Vol. 1, pp. 7-25 and 9-10 ]. Overpressure failure of the containment for sequences in which containment engineered safety systems operate was found to be far more likely for Sequoyah than for Surry due to the lower containment design pressure and smaller containment volume of Sequoyah (NUREG/CR-1659, Vol. 1, p. S-11). Although both Surry and Sequoyah use Kestinghouse reactors, plant differences are manifested in significantly different dominant accident scquences [NUREG/CR-1659, Vol. 1,

p. 9-12).

Plant systems and design features which are important to risk are different for Surry and Sequoyah [ Ibid.). Unlike the Surry plant, core melt accidents at Sequoyah caused by failure of emergency coolant injection or emergency coolant recirculation can fail the containment due to generation of noncondensible gases (a result similar to the Peach Bottom boiling water reactor, also analyzed in the Reactor Safety Study) [ Ibid.). Unlike the Surry

plant, failure of cor.ta inment cooling following a small LOCA does not lead to core melt at Sequoyah (core melt at Surry for such sequences was predicted to occur due to boiling of. sump.

f ya

4 - ' _12 ~ water leading to cavitation of emergency core cooling system pumps) [ Ibid.]. While there were only four dominant accident sequences for Surry, there were nine for Sequoyah [NUR,EG/CR-1659, Vol. 1, p. 9-13]. Containment base melt through sequences can occur.before above ground containment failure for.Surry,_ whereas for Sequoyah an above ground containment

failure, is predicted to always precede containment basemat melt through.

Containment failure by -overpressurization is predicted to be a certainty for core melt accidents at Sequoyah if other containment failure modes are avoided [NUREG/CR-1659, Vol. 1, pp. 8-2 and 8-12]. 0 14 Which 'results do you recommend using as a ~ basis - for emergency planning for Catawba, Surry or Sequoyah? A.14 Due to the differences in severe accident performance between Surry and Sequoyah, and the similarities between Sequoyah and Catawba, I recommend (in the absence of plant-specific results for Catawba) using' the Sequoyah RSSMAP results as a basis for emergency planning for Catawba. Q.15 What are the implications of using the Sequoyah accident progression analyses for Catawba in the. context of emergency planning? A.15 Accident progression (timing) results for sixteen accident sequences at Sequoyah are found in the RSSMAP analysis [NUREG/CR-1659, Vol. 1, p. 8-8]. In three of these sequences, containment failure occurs in about an hour or less-(including Event V, the interfacing LOCA, in which the containment-is bypassed at the time of accident initiation due to the. nature of the. accident). For the remaining thirteen sequences, core melt and containment l -s.

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?F ' fa'ilure are complete.within roughly four hour of accident ~ ' initiation for seven of the~ thirteen.

Thus, ten of. the sixteen sequences ~ analyzed will be accompanied by containment failure with'in about.four hours or'less.

The remainding. six have times for core melt and containment failure ranging-from about five hours to thirteen-hours. The.-full.results of this analysis.are provided.as an. attachment-to this testimony. Another important consideration is that at least five of the sequences leading.to containment failure within about four hours (and four of the nine ' dominant accident sequences, -for which in some cases no explicit progression calculations were presented) are assigned to release categories involving substantial fractions of the core inventory of the

iodine, cesium-rubidium, tellurium-antimony radionuclide groups.

These radionuclide groups tend to dominate accident ~ ~ consequences. NUREG-0 65 4 provides guidance on plume transit times within ten miles, providing a range of one to four hours [NUREG-0 65 4, Rev. 1, p. 17]. For_a twenty nile distance, these values can be doubled to two to eight hours. The city of Charlotte is in the range of ten to twenty-five miles, with the distance proposed in the contention for the extension of the plume EPZ -of seventeen miles. At seventeen miles, the approximate ' plume. transit times range from one and a half-to six hours, a + - When the core melt accident timing considerations are combined with the plume transit times, we obtain time periods ranging roughly from five and a half to ten hours from the beginning of the accident to the arrival of the plume in the vicinity of Charlotte: (assuming the wind is = 9

ls-;k ..y - blowing' in 'ihe direction of Charlotte). In some cases, the time period will ~ be shorter; in as with' Event .V,. other cases, where the release'does not occur until about thirteen-hours, the time will-be longer. In ' m'any. cases, however, the range of roughly five to ten hours will apply. This time period will be reduced by the time consumed in diagnosing the-accident, and the y time consumed in notifying the public of the need to take protective

actions, and any

delay, time between

notification and the beginning of the implementation of the protective actions by the general public.

? A crude indication of the time consumed in diagnosing the accident is ptovided in the " warning" time values used in accident consequence calculations. For the Sequoyah ice condenser release categories [NUREG-0i73, p. 40 ], the . warning, time (the time available between notification of C) offsite authorities and the time of release) ranges '~' ~ between~ thirty ~ minutes and two hours. ~ ~ ~ ~ ' " ~~ These time periods are probably on the pessimistic side of a distributian of potential time periods required for accident diagnosis. This pessimism is due to-the adoption since the analyses were perforn.ed of the use of " Emergency Action Levels" [NUREG-0 65 4, Rev. 1, Appendix 1) and-symptom-oriented emergency procedures. These

features, if properly

used, should shorten the time required to diagnose an accident and activate emergency plans.

Nonetheless, it must be considered unlikely that plant operators will diagnose an impending severe core damage or core melt accident until either some core damage 9 i indication is annunciated in~the' control room or there is ) a' clear-indication vf the failure of key safety functions D 4' l V

h 5 h'

(e.g.,

emergency core cooling). Thus, the five to ten ' hour period. indicated above _for _ accident progression and plume transit does not~ indicate the amount of time available.for the implementation of protective actions beyond - the present plume EPZ -.the latter time period will be less than five to ten hours, perhaps considerably "so depending upon the. circumstances. 0 16-What' sources of information are available on ' accident likelihoods and ' accident consequences (both doses and 1 health effects) which can aid.in an evaluation of emergency planning for Catawba?- A.16 The principal sources of information of accident. likelihoods. are completed PRAs -for pr'essurized water reactors in the U.S., and documents - which. provide suminaries of such information. The principal sources of information on accident consequences are NUREG-0396, NUREG/CR-ll31, and NUREG-0921. -0 17 What is the range of core melt accident and large release likelihoods for. pressurized water reactors in the U.S. based on PRA results t' date? A.17 PRA estimates of core melt and large release likelihoods for U.S. pressurized water reactors were summarized in a memorandum prepared for the NRC Commissioners ~in January 1983 [Dircks). The results for core melt likelihoods range from about 1: 50 0 to 1:25,000 per reactor year, a-range of roughly a factor of 50 (there are large uncertainties in the individual estimates). The results for.large release likelihoods (i.e., a release with the-potential to cause early fatalities offsite given nominal emergency response assumptions) range from about 1:1,000 to about 1:250,000, a range of roughly a f actor. of 250 i

1

p ~ ~ (there are large. uncertainties in. the individual estimates). i 1 Q'.18 Where do the' Catawba reactors fall within these ranges? A.18 Absent a plant-specific PRA, it is difficult to have substantial confidence in any particular estimate for the Catawba reactors. Given the apparent similarities between Catawba and

Sequoyah, one might have some confidence-that the results would not differ dramatically.

- Such a judgment must be tempered by the recogniti.on that - plant-specific design and operational differences have been found to be important to rick in each PRA done to date. Simply accepting the Sequoyah results as ccmpletely applicable to Catawba ignores the possibility that risk outliers may be present at Catawba. Further, it should be noted that the range of core melt and large release likelihoods presented in A.16 above did ~ not include so-called " external events" for ~mariy reactors. External

events, such as earthquakes, hurricanes, tornadoes,

fires, etc.,

have been analyzed for only a few pressurized water reactors to date (Indian Point Units 2 and 3, Zion Units 1 and 2, and Seabrook Units 1 and 2). In these cases, external events have been found to be risk significant (and sometimes dominate risk), although the results are very site-and plant-specific (for example, the risk posed by Indian Point Units 2 and 3 was different both in magnitude and in the specific accident sequences which dominated risk) [IPPSS). At. most, therefore, one might conclude that the risk posed by the Catawba reactors is reasonably approximated by - the Sequoyah Unit 1 RSSMAP PRA for internal events (there are large uncertainties associated with such a aa + -Q

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3.,

17_. %,? ~, l s worth' noting'that'if-we assume that all i ~

judgment). - It gf 4, '. o n "~ the pressurized' water. reactors analyzed in... PRAs.. meet' NRC ~

' regulatory -re.quirements,. the;-range..of-performance -in severeLaccident conditions.' implied by the ranges of core - melt and large:' release 11ikelihoods suggests that meeting ~ NRC regulatory requirements ' does not equate to.any. particular level of risk.as estimated:in'a PRA. L Absent-site-and. plant-specific

analysis, it. is not possible to ' judge.' whether. the-influence of external events will' af fect.the comparison- 'between Sequoyah <and

~

Catawba, or whether-there-are risk : outliers ~ for - the Catawba reactors which~ render-the comparison-less robust.,

For emergency planning. purposes, however, the? Sequoyah - PRA results provide the'best'available guidance.- 0 19 What are the. implications of accident-consequence analyses for emergency planning at Catawba? ~ A.19 ' NUREG-0 396 servesTas ~ the explicit technical' basi's^f6r~the " ~~ size of the plume EPZ, and therefore represents a logical starting place. In responding to this

question, consideration of consequences will be limited to whole body exposure to gamma radiation.

Fig ure I-ll from NUREG-0396 (attached to this testimony) [NUREG-0 3 96, p. 1-38) presents curves of the conditional probability of whole body dose versus distance-for core melt. accidents. These curves are explicitly: base'd on the

source terms and rela.tive probabilities of the Reactor Safety Study release categories PWR-1 through PWR-7.

The curves result from a probabilistic weighting of separate curves for each release category. The doses ~ were calculated based on' straight line plume trajectory and an assumption of.no' protective actions,-and were calculated' using the CRAC (" Calculation of Reactor Accident

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a - -Consequences") computer model= developed for the. Reactor ' Safety ' Study '[ WASII-140 0, Appendix 'VI; NUREG-0 3 40 ; ^- NUREG/CR-3185). From Figure I-11 of NUREG-0 396 conclusions for Catawba are possible if the' assumption is made that these results' reasonably represent Catawba. This assumption is-somewhat questionable since the results are for release characteristics and relative probabilities for Surry rather than for a reactor with an ice condenser containment. The release like-lihoods for release l categories PWR-1 through PWR-3,

however, are not very

'different between the Surry and Sequoyah analyses (there are large -dif ferences for release categories PWR-4 and PWR-5). Another consideration is that the curves will be slightly conservative for ' Catawba since the WASII-1400 consequence calculations were carried out for a 3,200 MWt core, whereas the Catawba core is somewhat larger at 3412 MWt. This reservation aside, given a core melt accident there is about a 30 % likelihood (about one chance in 3) of exceeding the 1 Rem whole body PAG at 10 miles, and about a 20 % likelihood (about 1 chance in 5) of exceeding the 5 Rem whole body PAG at 10 miles. Another way of stating this is that there is about 1 chance in 5 to 1 chance in 10 of needing to implement protective actions beyond the present 10-mile plume EPZ given a core melt accident. D

Further, again based on Figure I-ll from NUREG-0 3 96, there is about a 10 % likelihood (one chance in 10 ) of exceeding a 50 Rem whole body dose at 10., miles; such a dose is a factor of ten greater than the upper bound whole body PAG dose of 5

Rem. The likelihood of exceeding a dose of 20 0 Rem whole body (which 'is in the range of early fatality threshold without medical II

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69'

intervention) at 10 milen is about 3% (about 1 chance in ~ ' ~ -30 ) given a core melt accident. Additional perspective can be

gained, however, by separating the PWR release categories into those involving direct releases to the atmosphere (i.e.,

PWR-1 through PWR-5) and those involving releases - resulting from base' mat melt through (i.e., PWR-6 and PWR-7). This was done in NUREG/CR-ll31 (NUREG/CR-ll31, Figures 5.2,

5. 3,

5. 9, and 5.10, attached to this testimony] for the mean (average over 91 weather sequences) and 95% (value

-equalled or exceeded in only one weather sequence out of. twenty) cases. Given a core melt accident with a basemat melt through release (cxamining Figures 5.2 and 5.3 from NUREG/CR-ll31, using Curve A representing no protective actions), the average distance to which the 1 and 5 Rem whole body PAG doses will be reached is about 1-2 miles ~ ' ~~ "and 0.' 4

miles, respectively.

I n' ~ the' '95%

case, the~ ~

' ~' distances are about 6 miles and 2 miles, respectively. In addition, in the 95% case (equalled or exceeded only. 5% of the time), the distance to which a 50 Rem whole body dose is exceeded is about 0.2 miles. Given a core melt accident with a release to the atmosphere (examining Figures 5.9 and 5. 10 from NUREG/CR-1131, using Curve A representing no protective actions), the average distance to which the 1 and 5 Rem whole body PAG is reached is about 10 0 miles and 80 miles, respectively. Mo r r.ove r, a 50 Rem whole body done is reached at about 20 miles, and a 20 0 Rem whole body dose is reached at about 8 miles. In addition, a 500 Rem whole body dose (5 10 Rem is the so-called "LD-50 /60 " dose in WAS!!-140 0, that dose sufficient to result in early .A '*}

w fatalities to 50 % of ' those exposed within 60 days) is reached at about 3 miles. In the 95% case - (equalled or exceeded' only 5% of the time),- the 'l and -5 Rem whole body PAG doses do not appear on the graph, but a ~ 10 Rem dose is reached at about ' 100 . miles. A dose of 50 Rem-is reached at about 50 miles. A 200' Rem dose is reached at about 20 miles. A 50 0 Rem dose'is reached at about 10 miles. A very. approximate overall perspective can-be gained as - follows. According to data contained in NUREG/CR-2239 [NUREG/CR-2239, p. A-21], the wind rose-for Catawba (based on data from 6/30 /71 through 6/30 /72) would place winds blowing toward Charlotte from Catawba (compass 10 " ) of headings of NNE, NE, and ENE) about 35% (3.5 x the time. Release categories PWR-1 through PWR-3 dominate the above ~ ~ ~ ~ ~~ relationships where the PWR-1 through PWR-5 releases are probabilistically weighted. Based on the Sequoyah RSSMAP PRA, the approximate likelihood of a PWR-1 through PKR-3 release is about 1 in 25,000 (4 x 10 -5) [Dircks; NUREG/CR-1659, Vol. 1, p. 9-13). The overall core melt probability is about 1 in 17,000 per reactor year (6 x 10 -5).

Thus, the conditional likelihood of a

large release given a core melt is approximately 2 in 3 (6.7 x ~ 10 ). Thus, combining the likelihood of a large release (PWR-1 through PWR-3) with the likelihood of the wind blowing in the direction of Charlotte at the time of the release, a very approximate overall likelihood of a large release occuring with the wind blowing toward Charlotte is about 1 in 72,000 per reactor year (1.4 x 10 -5 ). In addition, combining the conditional likelihood of a large release i - ib

c. 1..

given

a core melt. with the likelihood of a the wind

. blowingr towa rd Charlotte.at the -time '.of the release,- we - obtain 'a conditional probabil'ity- (given -'a core melt) of a large release '- with. the wind ' blowing toward Charlotte of about 1 chance in 4 l(2.3 x 10 -1). On - average (the' mean case), when a-large release occurs with the wind blowing toward Charlotte, the dose at 10 miles-will be about 100 Rem whole body and the cose at 20 ~ miles will be about' 50 Rem whole body if no protective ~ actions are taken. .In the 95% case (with a likelihood of-1 chance in 20, or 5 x 10 ), the dose at 10 miles will ~ be about 'i00 ~ Rem and the dose at 20 miles will'be about 200 Rem. -This case has an approximate overall likelihood (based on calculations above) of about 1 in 1.4.million and a conditional probability (given a core melt accident) of about 1 in 90 (1.1 x 10 -2), The absolute probability values derived above are very ^ ~ ~~ uncertain, and assume that the results from the Sequoyah RSS!1AP PRA are competely applicable to Catawba (which they may not be, but they are certainly more representative than Surry's results). The conditional likelihoods have less uncertainty (being dependent only upon the relative likelihood of a large release given a core melt and the likelihood of the wind blowing toward Charlotte), and are therefore more robust. 0 20 What are the implications of the information provided in response to 0 19 for the configuration of the plume EPZ at Catawba? A.20 Given a large release ~with.the wind blowing toward Charlotte, even in the mean (average) case protective actions will be necessary beyond the existing 10 mile EPZ because whole body doses will be above the PAG levels - 4.- = l. mu

O

. t l

inithe absence'of, protective actions. Protective actions ~ would also be cnceded beyond the existing 10 mile EPZ - if -- the wind was blowing in any other direction from Catawba at the' time of the. release. .The question ~ of. whether' Charlotte - should be included within the plume EPZ (as opposed to other areas outside the plume EPZ) turns on the relative' difficulty or ' implementing protective actions. In response to Q.15 'above, I indicated that the: time from accicent initiation to the transit.of the plume through a distance 'f rom 10-17 miles.from Catawba would be roughly 5-10 hours. I also indicated that the actual time. between when a warning ~ could be given and plume transit would be less than the range of 5 - 10 hours, perhaps substantially.so depending upon circumstances. Thus,-the range of 5 - 10 hours would represent an optimistic upper bound case (i.e., with almost immediate warning to offsite authorities when the accident starts, an immediate decision to implement protective

actions, and prompt communication ~of this

- - ~ ^ information'to the public). In the worst case, assuming only minimal (30 minutes) warning time before the release occurs, the plume will complete it: transit of the Charlotte area in about 2-6.5 hours.

Further, the time available to implement-protective actions will be reduced by the time consumed in notification of the general public of the need to take action.

The length of time required to notify the residents of, the city of Charlotte to take protective actions is open to speculation at this time (however. some fraction of the population will be watching television or listening to the radio at-any given time and will receive broadcast warnings; further, fire and civil defense sirens could be sounded, and police and other emergency vehicles with sirens could be pressed e

? 4 --23-- 4T _y x [ Iint' ~ Lservice).- . Emergency planning. and' active : public o ' education could' improve' notification t'imes. t " Given - these ' considerations, for 'some-accidents (namely, those: in. which containment failure occurs within about~ ' four hours or -less of the -start of-the ' accident). it does-not appear that evacuation would' _be a feasible option. Assuming the population delays-one hour before evacuating [NUREG-0 921, p. F-3], more time.will be lost between.the start of - the accident and plume transit of Charlotte. Evacuation ef forts would E need to. be. concentrated within .the L existing 10-mile EPZ - where the Lresidents of that area.- are at greater risk.(due -to higher-_ ~ exposure levels).

However, as Figures 5.9 and 5. 10 from NUREG/CR-ll31 demonstrate, sheltering, with relocation six hours after plume passage provides ' roughly equivalent protection to evacuation.

Curves B and D represent sheltering with different sheltering

factors, and Curves C

and: E represent' evacuation ~at an effective speed-of 107 mph. (the - NRC Staff's consequence estimates in NUREG-0 921 assume an effective speed of - 6. 7 mph-based on evacuation time estimates for the existing 10 mile EPZ) with' delay-times-of five and three hours, respectively. Even the least favorable of these four,cmergency reponse i sets provides dose ~ reductions of a factor of about 3-5 for the mean case (given an atmospheric release)- and a factor of about 3-for the 95% case in the 10 - 20 mile distance interval. The least favorable set-assumes sheltering with chielding factors of 0.75 for cloud exposure and 0.33 for g round ' e xposut e). The. most favorable shielding f actors assumed were 0.5 for c!oud I -exposure and 0.0 8 for ground exposure. \\ h i4 .e-1 2.

~ .According to NUREG/CR-2239 [NUREG/CR-2239, pp. A-5 and .. _ _3_7j 7 Catawba was placed ' into a sheltering region L with m shielding factors of 0.6 and 0.2 for the Sandia siting study calculations. Thus, the actual sheltering result for Catawba would lie somewhere between curves B and D on Figures 5. 9 and 5.10 in NUREG/CR-1131. Doses might be reduced further -if infiltration of radioactive particulates can be minimized by. shutting-down ventilation systems, moving to basements or the interior areas of. buildings, and blocking cracks in doorways with cloth ~or paper. Inhalation doses could be reduced further with ad hoc respiratory protection [NUREG/CR-2272). These measures should be evaluated in more depth. Implementation of such measures would require an adequate program of public education. These considerations suggest that an emergency plan for Charlotte should consider sheltering with prompt relocation from contaminated areas after plume passage for the relatively fast-moving accidents. For accidents in which the containment is not projected to fail for ten hours or more, evacuation appears to be a more realistic alternative. Q.21 What should be the principal considerations for an emergency plan for Charlotte involving nuclear accidents at Catawba? A.21 Several key considerations emerge from the above discussions. First, redundant communications links with the utility and other offsite emergency response organizations are needed.

Second, prompt access to radiation monitoring equipment is needed to locate contaminated areas Irom which prompt relocation must occur and to avoid having persons relocating after plume 4,

' 'e b4+ 's m

ti; .> passage into contaminated areas- (airborne monitoring f rom m - - a ' helicopter 'would be ~a. good ~ choice if available). ~ Third, some consideration should be given to possible egre'ss routes to facilitate relocation and evacuation. . Fourth, consideration needs to be given to means of-public notification and the content of emergency messages (this requires liason with local media). Public education is most'important, not only_so that the public will'know what may be expected of them, but'so 'that if'the recommended protective action is sheltering, the public will understand the benefits of sheltering and relocation, and understand the reasons why this option has been ' selected. The latter is very important since vehicles provide essentially no shielding against gamma radiation and minimal protection against -infiltration of radioactive particulatcs, and it is most undesirable to have people in _ vehicles ' in a ' traffic queue be overtaken by a radioactive plume. An emergency plan incorporating these features for Charlotte need not be painstakingly detailed or extremely expensive. Existing emergency plans may. already incorporate some of the functions

required, and the remainder could be developed without significant expenditure of resources.

What is required is a recognition of the need for the plan, the benefits which could derive from it in the event of an accident, and a commitment from the city of Charlotte,. the Applicant, and Federal, state, and local planners to cooperate in s the development of a plan for Charlotte and its integration into the overall emergency plan. 4 ? -tt

• t

L' Q.22 What are your conclusions regarding the necessity of-extending the plume EPZ to include the city of Charlotte? .A.22 Based on considerations of, -the possible performance of the Catawba reactors under core melt accident conditions, the conditional likelihood of a severe release occuring with the wind blowing toward Charlotte given a core melt accident, the benefits which can be obtained from the implementation of even - minimal protective actions, and the modest effort involved, I recommend that the plume EPZ be extended as recommended in the contention. As a practical matter, the planning. done for the 10 - 1 7 mile. area of - Charlotte will be applicable to the remainder of the city as well. The preparation of such a plan will have a salutary effect as well -- the planning for sheltering and relocation for radiological emergencies will to a great extent be useful in other emergencies (such as those involving toxic materials - spills).~'~ ~ ~ ~~ ~ ' - ~ k A O I B 'I )

y re' E..N..D....N..O..T..E..S. 'b! Protective ' Action Guides' (PAGs) are projected doses doses that would be received by the population of no protective actions ~are taken -- established _ by the U.S. Environmental Protection Agency (EPA) in 1975 for exposur'e to airborne' materials released in nuclear accidents._ For exposure of the general population to whole body gamma ' radiation,- the EPA has established a .e range of PAGs from'l to 5 Rem.whole body exposure. For thyroid exposure of the. general population, the EPA has established a range of PAGs from 5 to 25 Rem thyroid exposure. According to' EPA guidance, the lower range.of these PAGs should be used when there are no major local constraints in providing protection against exposure, especially to sensitive populations. In no case, F3 wever, should the upper range of these PAGs be exceeded - . In determining the need for protective action. The PAG doses do not include that dose which has unavoidably occurred prior to making dose projections [ EPA 520/1-75-001, pp. 2.1-2.8]. 2/ ?mong the assumptions made are: (a) a source term consisting of 10 0 % of the core inventory of noble gases, ~ 50 % of the ccre inventory of

iodine, and 1% of the remaining core inventory, (b) no consideration of natural attenuation processes' in containment, (c) no consideration of the impact of engineered safeguards features such as containment sprays on ' fission product

behavior, (d) containment isolation and leakage at a constant 0.1%

per

day, (e) time invariant fifth percentile meteorology, and (f) no protective actions for the exposed population.

4 't$ h*

p ?- m E E LE_E_E_E_C_g_g DIRCYS-Memorandum dated 5 January 1983 from William J. Dircks to NRC ' Commissioners :Palladino, Gilinsky,.Ahearne, Roberts, and Asselstine,

Subject:

" Safety Goals", enclosing, " Comparison of Plant' Specific PRAs with Proposed Safety Goals". EPA 520 /1-75-001 Office of Radiation Programs, " Manual of Protective Action Guides and Protective Actions for Nuclear Incidents", EPA 520/1-75-00 1, U.S. Environmental Protection Agency, September-1975, Revised June 1980. IPPSS Power' Authority of the' State of New York and Consolidated Edison Company of New

York, Inc.,

" Indian Point Probabilistic Safety Study", 1982. NUREG-0 340 I.B. Wall, et al., " Overview of the Reactor Safety Study Consequence Model", NUREG-0 3 40, U.S. Nuclear Regulatory Commission, October 1977. NUREG-0396 Task Force on Emergency Planning, " Planning Basis for the i Development of State and Local Government Radiological Emergency Response Plans in Support of Light -Water Nuclear Power Plants", NUREG-0 396, EPA 520/1-78-0 16, U.S. Nuclear Regulatory Commission and U.S. Environmental Protection Agency, December 1978. NUREG-0654, Rev. 1 FEMA /NRC Steering Committee, " Criteria for Preparation and Evaluation of Radiological Emergency Lesponse Plans and Preparedness in Support of Nuclear Power Plants", NUREG-0654, FEMA-REP-1, Rev. 1, U.S. Nuclear Regulatory Commission and rederal Emergency Management Agency,- November 1980. NUREG-0773 R.

Blond, et al.,

"The Development of Severe Reactor ' Accident Source Terms: 1957-1981", NUREG-0 7 7 3, U.S. Nuclear Regulatory Commission, November 1982. NUREG-0921 Office of Nuclear Reactor Regulation, " Final Environmental Statement related to operation of Catawba Nuclear Station, Units 1 and 2", NUREG-0 921, U.S. Nuclear Regulatory Commission, January 1983. A

i o NUREG/CR-0 60 3 R.E. !!a ll, et al., "A Risk Assessment of a Pressurized Water Reactor for Class 3-8 Accidents", NUREG/CR-0 60 3, BNL-NUREG-50 950, prepared by Brookhaven National Laboraotry for the U.S. Nuclear Regulatory Commission, October 1979. 1 NUREG/CR-1131 D.C.

Aldrich, P.

McGrath, and N.C.

Rasmussen, " Examination of Offsite Radiological Emergency Measures for Nuclear Reactor Accidents Involving Core Melt", NUREG/CR-ll31, S AND7 8-0 45 4, Sandia Laboratories, prepared for the U.S. Nuclear Regulatory Commission, June 1978. NUREG/CR-165 9, Vol. 1 D.D. Carlson, et al., " Reactor Safety Study Methodology Applications Program: Sequoyah 11 PWR Power Plant", NUREG/CR-1659, Vol. 1, SAND 60 -1897/1 of 4, prepared by Sandia National Laboratories for the U.S. Nuclear Regulatory Commission, April 1981. NUREG/CR-2239 D.C.

Aldrich, et al.,

" Technical Guidance for Siting Criteria Development", NUREG/CR-2239, SAND 81-1549, prepared by Sandia National Laboratories for the U.S. Nuclear Regulatory Commission, December 1982. NUREG/CR-2272 D.W.

Coopet, W.C.

Hinds, and J.M.

Price,

" Expedient Methods of Respiratory Protection", NUREG/CR-2272, SAND 81-7143, prepared by the Harvard School of Public IIcalth for Sandia National Laboratories under contract to the U.S. Nuclear Regulatory Commission, November 1981. NUREG/CR-2326 L.T.

Ritchie, J.D.

Johnson, and R.M. Blond, " Calculations of Reactor Accident Consequences Version 2, CRAC2: Compupter Code User's Guide", NUREG/CR-2326, SAND 81-1994, prepared by Sandia National Laboratories for the U.S. Nuclear Regulatory Commission, February 1983. NUREG/CR-2497 J.W. Minarick and C.A. Kukielka, " Precursors to Potential

1979, A

Status Severe Core Damage Accidents: 1969 Report", NUREG/CR-2497, ORNL/NSIC-182, Oak Ridge National La bo rii tor y, prepared for the U.S. Nuclear Regulatory Commission, June 1982. NUREG/CR-2552 L.T.

Ritchie, et al.,

"CRAC2 !!od e l Description", NUREG/CR-2552, SAND 82-0 3 4 2, prepared by Sandia National Laboratories for the U.S.. Nuclear Regulatory Commission, March 1984. 4 *-

e \\ NUREG/CR-2901 F J.D. Johnson' and ' L. T. .Ritchie, "CRAC Calculations for . Accident Sections-of. Environmental _ Statements", NUREG/CR-2901, SAND 82-16 93, prepared by Sandia National Laboratories - for the U.S. Nuclear ' Regulatory Commission, March'1983. NUREG/CR-2925 R.P. Burke,. C. D.

Helsing, and D.C.

Aldrich, "In-Plant

-Considerations for Optimal Offsite Response = to Reactor Accidents",- NUREG/CR-2925, SAND 8 2-200 4, prepared by Sandia. National Laboratories for the U.S. Nuclear Regulatory Commission, November 1982.. NUREG/CR-3185 D.W.

Cooper, et al.,

" Critical Review of the Reactor Safety Study Radiological Health Effects Model", NUREG/CR-3185, SAND 8 2-70 81, prepared by the Harvard School of Public Health for Sandia National Laboratories under contract to the U.S. Nuclear Regulatory Commission, March 1982. ORNL/NSIC-176 H.W. Bertini, et al., " Descriptions of Selected Accidents That Have Occurred at Nuclear Reactor Facilities", ORNL/NSIC-176, Oak Ridge National Laboratory, April 1980. ORNL/NSIC-217 draft W.B. Cottrell, et al., " Precursors to Potential. Severe Core Damage Accidents: 1980 - 1981, A Status Report", ORNL/NSIC-217, draft

report, Oak Ridge National Laboratory, prepared for the U.S.

Nuclear Regulatory Commission, July 1983. TID-14844 J.J. DiNunno, et al., " Calculation-of Distance' Factors for Power and Test Reactor Sices", TID-14844, U.S. Atomic Energy Commission, second printing, 23 March 1962. WA Sil-140 0 N.C. Rasmussen, et al. " Reactor Safety Study: An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants", WA S H-140 0, N UR E G-75 /0 14, U ~. S. Nuclear Regulatory Commission, October 1975. WASil-1400, Appendix VI " Calculation of Reactor Accident Consequences", Appendix VI, WASH-140 0, October 1975.

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• Melt-

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r, ?....g .c. *' b ~ ' STATEMENT OF' PROFESSIONAL QUALIFICATIONS -- STEVEN C. SHOLLY My name is Steven C. Sholly. I am a Technical Research Associat'e with the Union of Concerne-Scientists (UCS), Dupont -Circle Building, 1346 Connecticut Avenue, N.W., Washington,- D.C. 20036. I joined the UCS staff in February 1981. My primary responsibilities are UCS are technical and policy ^ analysis concerning probabilistic risk assessment and radiological emergency planning. In addition, I monitor nuclear safety research in'several other ares, including severe i accident research, accident mitigation systems, and alternative reactor designs. I am also a regular contributor to UCS's newsletter, Nucleus. Prior to joining UCS, I served as Research Coordinator and Project Director of the DiI Public Interest Resource Center (TMIPIRC) in Harrisburg, Pennsylvania. TMIPIRC was created after the Three Mile Island accident by concerned citizens t groups in Pennsylvania. At n!IPIRC, I was responsible for directing research and public education activities associated with the proposed restart of TMI Unit 1 and the cleanup of TMI Unit 2. In addition to this experience, I taught secondary school science for two years. I also have two years experience in wastewater treatment, including experience as Chief Process Operator of a 5.0 MGD tertiary treatment facility. In the latter capacity, I obtained state certification to operate activated sludge wastewater treatment plants (Pennsylvania Class B, Type 1 certification). I have provided testimony before Congress and a special committee 'of the New York State Assembly on radiological emergency planning matters.- I have also testified before 6 I ()o 'er y J L

c: ( ;o< o Congress on safety issues associated with steam generators in pressurized water reactors. During the Indian Point Units 2 and 3 Special Investigation in 1983, I provided expert testimony on behalf of UCS and NYPIRG on filtered vented containment systems (jointly with Dr. Gordon Th ompson), severe accident consequences, and comparative risk analysis of nuclear power plants. Most recently, I provided supporting evidence (principal evidence by Dr. Gordon Thompson) on emergency planning and probabilistic risk assessment in the Sizewell B Inquiry in the United Kingdom on behalf of the Town and Country Planning Association. I am a 1975 graduate of Shippensburg State College (now Shippensburg University), Shippensburg, Pennsylvania. I received a B.S. degree in Education (majors in Earth and Space Science and General Science, and minor in Environmental Ed ucation). I have also completed graduate coursework in land I use planning. I am a resident of Columbia, Maryland. l l l l e 9 0 e n e9 l.. ... -}}