ML20083M108

Testimony of Sc Sholly on Emergency Planning Contention 11 Re Plume Exposure Pathway Zone.Certificate of Svc Encl. Related Correspondence

ML20083M108
Person / Time
Site:	Catawba
Issue date:	04/16/1984
From:	Sholly S CAROLINA ENVIRONMENTAL STUDY GROUP, PALMETTO ALLIANCE, UNION OF CONCERNED SCIENTISTS
To:
References
NUDOCS 8404170470
Download: ML20083M108 (40)

Text

3

, _HELATED CORRESPONDENCE m

00CKETED UStiRC UNITED STATES OF AMERIC APR 17 40:34 NUCLEAR REGULATORY COMMISSION

' ' O f O' SE9dir.p -

r.;TJYi s Scpyt BEFORE TIIE ATOMIC SAFETY AND LICENSINGY$ARD In the Matter of )

) Docket Nos. 50 -413 DUKE POWER 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. SIIOLLY ON EMERGENCY PLANNING CONTENTION NUMBER ELEVEN 170470 840416 ADOCK 05000413 T PM a

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! UNITED STATES OF AMERICA l NUCLEAR REGULATORY COMMISSION i.

B.E.F.O.R.E..T.i.l E..A.T.O.M..I C..S.A.F.E.T.Y.

~ ..A N.D..L.I C.E.N.S..I N.G..B.O.A.R.D.

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a. In the Matter of )

{ } Docket Nos. 50 -413 DUKE PCWER CCt!PANY, ET AL. ) 50-414  ;

)

(Catawba Nuclear Station, Units ) 16 April 1984 1 and 2) )

i PALMETTO ALLIANCE AND CAROLINA ENVIRONMENTAL i

STUDY GROUP TESTIMONY OF STEVEN C. SIICLLY.ON EMERGENCY PLANNING CONTENTICN NUMBER ELEVEN i 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 4

in technical and policy analysis concerning risk assessment and emergency planning. My business address ist Union of Concerned Scientists, Dupont Circle Building, 1346 Connecticut Avenue, N.W., Suite 110 1, 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.

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Q.0 3 What is 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 its 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 at the Catawba facility would endanger the residents of that area and make their evacuation prudent. The likely flow of evacuees from the present plume EPZ through Charlotte access routes also indicates the need for evacuation planning for southwest Charlotte. There appear to be suitable plume EPZ boundarles inside the city limits, for example, highways 74 and 16 in southwest Charlotte. The boundary of the i northeast quadrant of the plume EPZ should be reconsidered and extended to take account of those demographic, meteorological and access route considerations.

0 04 What is the plume exposure pathway emergency planning zone?

A.0 4 The plume exposure pathway emergency planning -zone

(" plume EPZ") is an area surrounding a nuclear power plant for which emergency response plans are required in order to assure that prompt and ef fective actions can be taken to protect the public in the event of an accident it s

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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-0 396, pp. 27-28; NUREG-0 65 4, 7 ,

I Rev. 1, pp. 8-10 ] .

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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 e

nuclear power reactors is to provide does savings (and in some cases immediate life savings) for a spectrum of i

accidents that could produce offsite doses in excess of f Protective Action GuidesM [NUREG-0 654, Rev. 1, p. 6].

I' O.06 What protective actions for the general public are

! available to avoid or minimize exposures from the dose I

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 j exposures are:

a. Evacuation --

expeditious movement of the i population before plume passage to avoid

! exposure from a radioactive plume and-exposure due to ground ' contamination by deposition from the plume;

b. Relocation --

expeditious movement- of the

, population from contaminated areas:after 1 plume passage to avoid ~further exposure'from j ground contamination;

c. Sheltering --

expeditious movement of the l population indoors before plume passage 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 j conjunction ~with relocation);

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d. Respiratory protection --

use by the population of measures to reduce inhalation exposure during plume passage; and

e. Thyroid blocking --

use by the population (before plume passage) of potassium icdide to block the uptake of radioactive iodine by l the thyroid gland.

The choice of protective actions in any given accident situation depends on a number of f actors, including the magnitude and composition of the release from the plant (i.e., the source term), weather conditions at the time of and subsequent to the release, the amount of time i

available before plume passage, the distance of populated areas from the plant site, the speed with which various protective actions can be implemented, and the level of 1

j protection afforded by various protective actions.

Q.0 7 What is the spectrum of potential accidents at the Catawba Nuclear Station?

s A.0 7 The spectrum of potential accidents at' the Catawba

! Nuclear SF-tion range from relatively trivial plant

! upsets the ugh accidents involving severe core damage and large-scale melting of the core and subsequent breach of l 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) . Cther references describe additional incidents in some detail in both commercial nuclear plants and experimental reactors (ORNL/NSIC-176; i

ORNL/NSIC-217 draft; and NUREG/CR-2497].

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C.0 8 What is the significance of this spectrum of potential accidents for emernency 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.,

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the quantity and chemical form of radioactive materials available for release from containment), performance of engineered safety features, plume dispersion, and 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 I exposure for a two-hour period at the exclusion area l boundary and the entire period of plume passage at the low population zone boundary.W I

In contrast, realistic evaluations of design basis accidents result in exposures significantly lower than j these guideline levels. For example, the NRC Staff's FES-Catawba provides such calculated doses for design basis accidents at Catawba [NUREG-0921, p. 5-79). The largest calculated doses for Catawba design bas'is accidents are 0.0 6 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

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a doses [ 10 CFR 100 . Il(a ) (1) and (a) (2) ] , they are only .

j small fractions.of the Protectiive Action Guide doses (6%

and 1.4%, respectively,' for whole, body and thyroid exposures). \ * -

i Thus, even if these calculated doses are optimistic by a J factor of ten, the estimated doses from a realistic I evaluation of design basis' accidents at Catawba will not exceed the Protective Action Guide doses at the exclusion area bounoary. This observation leads to the conclusion that design, basis acci h'n t ts are not significant with i

respect to offsite emergency response.

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i As a practical matter, should a design basis' accident actually occur, offsit.e officials may decide to implement l' precautionary protect.ives measures such as sheltering or a limited. evacuation of areas near- the plant until condit' ions are stabilized.a.nd the potential for a release ,

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of radioactivity;to the environment han diminished.

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For accidentp beyond the design basis, a range of possible offsitU doses and consequence's is possible. It i

is conceivable- that a severe core damage accident could s i be sur.consfully " bottled up"' by the containment so long

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as containment, hat removal systems function adequately i and exces,sive 'hmounts' of ir.oncondens ible gases arp not i generated. On-yhe other hand, accidents beyond the design basic c6ul'd result in core molting and the,, release l 6 of radioactive materials, to the environment rang i,ng in i- quantity from trivial to vejyglatge. The magnit'de u of the reinase will depend uporf ti degree of cose $amage, the operating history of ' sitWe care,- the perfoEmAnce (or of ~ ehgineer\ad 3 (

lack thereof) 3 safety features, gnd the timihg and(ede of cb tainrent failure. t

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4 0 09 What magnitude of radiation exposure could result from ,

core melt accidents in which the containment fails in the absence of emergenty 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) ( WASil- 1400 , Appendix VI], Sandia calculated bounding doses from such releases. The dose calculations were carried out using the CRAC 2 accident consequence model [NUREG/CR-2326; NUREG/CR-2552; and NUREG/CR- 2901] ,

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 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. 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]:

RELEASE Wi! OLE BODY TilYROID CATEGORY DOSE (REM) DCSE (REM) 0 PWR-7 1 x 10 5 x 10 PWR-6 6x 10 2 x 10 3

PWR-5 1 x 10 8 x 10 4

PWR-4 5 x 10 3 x 10 4 4 PWR-3 2 x 10 2 x 10 4 4 PWR-2 7 x 10 7x 10 i 4 PWR-1A 8 x 10 9 x 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 i Protective Action Guide dose levels of 1-5 Rem whole body 4

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and 5-25-P.em thyroid) in core melt accidents in which the cory ta inm5n t dails. The r e s u l t's also indicate that sheltering -- not an adequate long-term protective action in areas close to , the sito for the more severe release categories -(this is d u e. to 'both the large initial ex posu re_..d u r ing plume' passage and the accumulation of exposure from radioactive materials deposited f rom' the plume on the ground during plume. passage).

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0 10 What are the ireplications of the above for emergency J

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 [ WASil-1400 ] .

and a comparative risk evaluation of accidents within and exceeding the design basis [ NUREG/CR-0 60 3) .

Indeed, NRC regulations -and joint NRC/ FEMA emergency planning guidance ref erence ' NUREG-0 396 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-ll31] . The

. dose versus distance and accident consequence:

calculations presented in NUREG-0 396 ' and ' NUREG/CR-ll31

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are explicitly- based on the characteristics of core melt accident- release categories f rom the - Reactor Safety Study. Thus, - we need _ to look to . analyses of .of f site

, ' doses and consequences.for.corefmelt-accidents at Catawba

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to gain perspective 'on the . size - and ~ configuration Lof the-plume.EPZ.

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0 11 Which reactor served as the model for the calculations in NUREG-0396 and NUREG/CR-ll31?

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 [ WASH- 140 0 ] analysis of a pressurized water reactor. The Surry Unit I reactor served as the surrogate in that analysis for all pressurized water reactors in the U.S.

Q.12 Briefly describe the Surry Unit I 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 reactors are four-loop Westinghouse pressurized water reactors with a thermal power output of 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 of equipment provided in the two plants. These "

differences can be determined by. comparing the. Final Safety Analysis Reports and Safety Evaluation Reports for the facilities.

0 13 How do the differences between Surry Unit. 1.' and .the Catawba Nuclear Station reactors affect their performance

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in severe core damage or core melt accidents?

A.13 The NRC Staff's FES-Catawba ' states _ that the design and operating characteristics of the two plants are similar I- .

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[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 Westinghouse 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., earthquakes, 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 O

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guidance as to the performance characteristics of the Catawba reactors under severe accident conditions.

The differences 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 transients 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 hestinghouse reactors, plant differences are manifested in significantly different dominant accident seque r.ce s [NUREG/CR-1659, Vol. 1, p. 9-12).

Plant systems and design features which are important to risk are -dif ferent 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 containment . cooling following a - small LOCA does not lead to core melt at Sequoyah.s(core melt a t- Surry for such sequences' was predicted to - occur due ' to boiling of sump

water leading to cavitation of emergency core cooling system pumps) [ Ibid.].

• Khile there were only four dominant accident sequences for Surry, there were nine for Sequoyah [ r4UREG/CR-165 9, 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).

Q.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. 6-8]. In three of these sequences, containment f ailure. 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-4 #

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failure are complete within roughly four hour of accident initiation for seven of the thirteen.

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Thus, ten of the sixteen sequences analyzed will be accompanied by containment failure within 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-0654, Rev. 1, p. 17]. For a twenty mile 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.

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 t

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blowing in the direction of Charlotte). In some cases, as with Event V, the time period will be shorter; in I other cases, where the release does not occur until about thirteen hours, the time will be longer.

In many 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 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 provided in the " warning" time values used in accident consequence calculations. For the Sequoyah ice condenser release categories [NUREG-0773, p. 40 ] , the warning time (the time available between notification of 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 distribution of potential time periods required for accident . diagnosis. This pessimism is due :to the adoption since the analyses were performed 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 e ither .. ~ some core . damage indication is annunciated in the control room or there is a clear indication-of the-failure of key. safety functions 4

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(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.

4 0.16 What sources of information are available on accident likelihoods and accident consequences (both doses and 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 pressurized water reactors in the U.S., and documents which provide summaries of such information. The principal sources of information on accident consequences are NUREG-0396, NUREG/CR-ll31, and NUREG-0 921.

Q.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 to date?

A.l? PRA estimates of core. melt and large release likelihoods for U.S. pressurized water .eactors were summarized in a memorandum prepared for the NRC Commissioners in January 1983 [Dircks]. The results for core melt likelihoods range from about 1: 500 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 factor of 250 t

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(there are large uncertainties in the individual estimates).

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 risk in each PRA done to date. Simply accepting the Sequoyah results as completely applicable to Catawba ignores the t

possibility that risk outliers may be present at Catawba.

Further, it should be noted that the range of core melt i and large release likelihoods presented in A.16 above did not include so-called " external events" for many 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

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judgment). It is worth noting that if we assume that all the pressurized water reactors analyzed in PRAs meet NRC regulatory requirements, the range of performance in severe accident conditions implied by the ranges of core melt and large release likelihoods suggests that meeting NRC regulatory requirements does not equate to any I

particular level of risk as estimated in a PRA.

Absent site- and plant-specific analysis, it is not possible to judge whether the influence of external events will affect the comparison between Sequoyah and Catawba, or whether there are risk outliers for the Catawba reactors which render the comparison less robust.

l For emergency planning purposes, however, ' the Sequoyah PRA-results provide the best available guidance.

Q.19 What are the implications of accident consequence analyses for emergency planning at Catawba?

A.19- NUREG-0 396 serves as ~ the explicit technical basis for the r 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.

Figure I-ll from NUREG-0 396 ' (attached to this testimony)

[ NUREG-0 3 96, p. I-38] presents-curves of.the-conditional probability of. whole body dose versus' distance - for core melt accidents. These-curves are_ explicitly-based on the source . terms and . relative _ 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 1- calculated based on' straight line plume trajectory and an

. assumption of'no_ protective-actions,-and were calculated using' the CRAC (" Calculation 'of Reactor Acc'id e n t'-

o Consequences") computer model developed for the Reactor Safety Study [ WA S H- 140 0 , Appendix VI; NUREG-0 3 40 ;

NUREG/CR-3185].

From Figure I-ll of NUREG-0 3 96 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 likelihoods for release categories PWR-1 through PWR-3, however, are not very different between the Surry and Sequoyah analyses (there are large differences for release categories PWR-4 and PWR-5). Another consideration is that the curves will be slightly conservative for Catawba since the WASH-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 n. ile s . 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.

Further, again based on Figure I-ll f rom 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 200 Rem _whole body (which is in the range of early fatality threshold without- medical

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intervention) at 10 miles is about 3% (about 1 chance in 30 ) given a core melt accident.

Additional perspective can be gained, however, by i

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 basemat melt through (i.e., PWR-6 and PWR-7) . This was done in NUREG/CR-ll31 [NUREG/CR-1131, 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 l equalled or exceeded in only one weather _ sequence out of twenty) cases.

Given a core melt accident with a basemat melt through release (examining Figures 5.2 and 5.3 from NUREG/CR-1131, 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. In 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-ll31, using curve A representing Eno protective actions), the average distance to which the-1=and 5 Rem whole body PAG is reached is about 100 miles and ' 80 miles, respectively. Moreover, a _50 E Rem whole . body dose; 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 (510' ~ Rem is the so-called "LD-50 /60 " -dose

'in . WASH-140 0 , that dose: sufficient to result 'in early

• W

l 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 1 and 5 Rem whole body PAG doses do not appear on the graph, but a 10 Rem dose is reached at about 10 0 miles. A dose of 50 Rem is reached at about 50 miles. A

. 20 0 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 / 7 2) would place winds. blowing toward Charlotte from Catawba (compass headings of NNE, NE, and ENE) about 35% (3.5 x 10 ~ ) of the time.

Release categories PWR-1 through PWR-3 dominate the above relationships where the PWR-- I through PWR-5_ releases'are

~

probabilistically weighted. Based on the Sequoyah RSSMAP PRA, the approximate likelihood of a PWR-1 through PWR-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 blowingiin the direction of Charlotte at the time of the release, a.

very approximate overall _ likelihood offa large_ release-occuring with the Lwind. 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 e -- , t-a v ,

l 1

given a core melt with the likelihood of a the wind-blowing toward Charlotte at the time of the release, we obtain a conditional probability (given a core melt) of a large release with the wind blowing toward Charlotte of about I chance in 4 ( 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 10 0 Rem whole body and the dose at 20 miles will be about 50 Rem whole body if no protective actions are taken. In the 95% case (with a likelihood of I chance in 20 , or'5 x 10 - ), the dose at 10 miles will be about 500 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 RSSMAP 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-the2 PAG levels

=9 5 e

1 I C

in the absence of protective actions. Protective actions would also be needed 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 i the plume EPZ) turns -on the relative difficulty of ,

implementing protective actions. In response to 0 15 5 above, I indicated that the time from accident initiation to the transit of the plume through'a distance f rom 10 -17 l

j miles from Catawba vould be roughly . 5-10 hours. I also J.

] indicated that the actual time between when a warning

[ could be given and plume-transit would be -less than the range of 5 - 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, perhaps substantially so depending upon circumstances. Thus, the range of 5 - 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> would 4

l represent an optimistic upper bound case (i.e., :with almost immediate warning to offsite authorities when'the I accident starts, an immediate decision to implement l protective actions, and' prompt -communication o f - ' t h i s ' ~- ^ ~~ ~ ~~ ~

, information to the public).

i i

In the worst case, assuming only minimal (30 minutes)

warning time before. the release occurs, t.be plume will . -

complete its transit of.the Charlotte-area in about--2-6.5-4 hours. Further, the time. available to z 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; o'f Charlotte to take ' protective .

actions is . open . to speculation ' at ; this. time- L(however,

'some fraction' of- thei. population .will . :be. ; watching televisionf or 1. listening to thel radio ; at < any u given? . time..

~

and will' . receive - broadcast 1 warnings;; further, ifire ' and =

civil defense.; sirens' could. be sounded, Tand : police (and: .

'other emergency vehicles -with' ' sirens could be pressed ~

w k!' (

.Q,' -

,, , , ; , . , . . c,

0 into service). Emergency planning and active public education could improve notification times.

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 efforts would need to be concentrated within the existing 10-mile EPZ where the residents 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 ari ~e f f Ective '~ speed'of~10^ mph ^(the NRC Staff's consequence estimates in NUREG-0921 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 emergency reponse 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 shielding factors of 0.75- for cloud exposure and 0.33 for ground exposure). The most favorable shielding factors assumed were 0.5 for cloud exposure and 0.0 8 for ground exposure.

6

• According to NUREG/CR-2239 [NUREG/CR-2239, pp. A-5 and A-7], Catawba was placed into a sheltering . region with 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 S. 9 and 5.10 in NUREG/CR-ll31.

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

- - -- reloca t ion - f rom contaminated-~ areas ~ af ter 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 accesc .to radiation monitoring equipment -is needed to locate contaminated areas from- which prompt relocation .must occur-and to avoid having' persons relocating after plume e?

passage-into contaminated areas (airborne monitoring from a . helicopter would be a good choice if available).

Third, some consideration should be given to possible egress 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 mcst important, not only so that the public will know what may be expected of them, but so i 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 j radiation and minimal protection against infiltration of

radioactive particulates, and it is most undesirable to have people in vehicles in a traffic queue be overtaken by a radioactive plume.

An en.ergency 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'itzin the event of an accident, and a commitment from the city of Charlotte,. the Applicant, and Federal, state, and local': planners to cooperate in-the development of a plan f or Charlotte- and its

, integration into the overall emergency plan.

L t

i

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4

- 9 C.22 What are your conclusions regarding the necessity of extending the plume EPZ to include the city of Charlotte?

A.22 Based on consideratione of the possible performance of the Catawba reactors under core melt accident conditions, the conditional likelihood of a severe release occuring 1 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).

t I

G f '! I

!. 'i _ ,

E.E.E...E.9.LE 8

' 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 (E PA) in 1975 for exposure to airborne materials released in nuclear accidents. For exposure of the general population to whole body gamma radiation, the EPA has established a range of PAGs from 1 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, however, 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 520/1-75-00 1, pp. 2.1-2.8].

! Among the assumptions made are: (a) a source term consisting of 10 0 % of the core inventory of noble gases, 50 % of the core 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.

D e 4 I

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t

S_E_E_E.B E.E C.E.S DIRCKS 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 Mocel", NUREG-0 3 40 , U.S. Nuclear Regulatory Commission, October 1977.

NUREG-0396 Task Force on Emergency Planning, " Planning Basis for the Development of- State and Local Government. Radiological -.

Emergency Response Plans in Support of Light Water Nuclear Power Plants", NUREG-0396, EPA 5 20 /1-78-016, U.S.

Nuclear Regulatory Commission and U.S. Fnvironmental Frotection Agency, December 1978.

NUREG-0654, Rev. 1 FEMA /NRC Steering Committee, " Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness _ in Support of Nuclear Power Plants",

NUREG-0654, FEMA-REP-1, Rev. 1, U.S. Nuclear Regulatory Commission and Federal 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--O 921 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.

i,-

6f

NUREG/CR-0 60 3 f R.E. IIall, 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.

! NUREG/CR-ll31 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, SAND 78-0 45 4, Sandia Laboratories, prepared )

for the U.S. Nuclear Eegulatory Commission, June 1978.

NUREG/CR-1659, Vol. 1 D.D. Carlson, et al., " Reactor Safety Study Methodology Applications Program: Sequoyah fl PWR Power Plant",

NUREG/CR-1659, Vol. 1, SAND 80 -18 97/1 of 4,- prepared by Sandia National Laboratories for the U.S. Nuclear Regulatory Commission, April 1981.

NUREG/CR-2239 D.C. Aldrich, et a l .' , " Technical Guidance for Siting

Criteria Development", NUREG/CR-2239, SAND 81-1549, i

prepared by Sandia National Laboratories for the U.S.

Nuclear Regulatory Commission, December 1982.

NUREG/CR-2272 D.W. Cooper, .W.C. Hinds, and J.M. Price, " Expedient .

Methods of Respiratory Protection", NUREG/CR-2272, SAND 81-7143, prepared by the - Ha rvard 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, "Precurso'rs'to. Potential.

Severe Core Damage Accidents: 1969 - '197.9, A Status Report", NUREG/CR-2497, ORNL/NSIC-182, Oak' Ridge National Laboratory, prepared for .the U.S. Nuclear Regulatory.

Commission, June 1982.

NUREG/CR-2552~ ~

L.T. Ritchie, et 'al.,- "CRAC2 Model- Description",

NUREG/CR-2552, SAND 82-0 3 4 2, prepared by' Sandia Na t ional--

l' Laboratories for the U.S. Nuclear RegulatoryECommission,.

March 1984.

NUREG/CR-2 901 J.D. Johnson and L.T. Ritchie, "CRAC Calculations for Accident Sections of Environmental Statements",

NUREG/CR-2 901, SAND 82-1693, 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 Gptimal Offsite Response to Reactor Accidents", NUREG/CR-2925, SAND 8 2-20 0 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, Maren 1982.

ORNL/NSIC-176 H.W. Bertini, et al., " Descriptions of Selected Accidents That lla ve Occurred at Nuclear Reactor Facilities",

ORNL/NSIC-176, Oak Ridge National Laboratory, Ap r il 19 80 .

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 Sites", TID-14844, U.S. Atomic Energy Commission, second printing, 23 March 1962.

WASH-1400 N.C. Rasmussen, et al. " Reactor Safety Study: An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants", h A SH- 140 0 , NUREG-75/014, U.S. Nuclear Regulatory Commission, October 1975.

WA S H- 1400 , Appendix VI

" Calculation of Reactor Accident Consequences", Appendix VI, WA Sil-140 0 , October 1975.

I e

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6 Table <-5 MAFCH Res its 'uf ICr sor.denset Pwk Accident Sequences

'- t.SSEi. ICE Mt.. .M coNCMETE ErC5 css .n. cRE mit.T

_ .7/ E M START LM;. F A f1IRF. CO'4PiJ '. I;. 6I MELT LTART SEOCENCE STAPT STGP START STOP . . ,

- 5 e 46 1 . 46 AD-O - - 1

' ad 66 4 e 67

- - 1 e+

AD-Y id 66 *i' 361 67 AD-6 - - 1 361 .

104 225 269 214 269 AMF-6 1 12 s 1 129 17 .

177 21t 159 i' 21s 269

. SgMF-5 1 12- 1 12 t-703 4r el 116 161 he N3 182 .

s yD-6 - - 1 I

e 110 1;- 11G 260 i S H-Y0 1  % 1 110 #4 2 716 260 798

• sa m 110 1. i S H-69 1 t' 1  ;

2 18e 324 109 1 109  ! 152 L t* - 206 t SyMF sa 1 ISJ 1e0 193 301 193 S HF-d 1 1rv 1 109 2- .

2 197 114 109 .- ISJ itM 193 i ,

S ,HF = U' 1 IV9 1 I

84 2 na .' 3 2 238 14 3;e 23e m

.B'-f - - - -

les 2 0' ' JJJ 238 .4. 144 384 TMLB'-f - - - -

ld; 2w J .t 2 238 *la 13d 238 TMI.- v 238 - 1 238 '

JOL .>J

< 136 >l" 238 .

i mL-f 238 - 1 660 i.~

- - 3M 57 91 - 91 V s ,

a All t.ames an minut M-b Ic= bed bypassed af ter steam explosion t

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0001 100 1000 1

10 OISTANCE (MILES)

Figure I-11. Conditio.tal Probability of Exceeding Whole Body Dose Versus Distance. Proba are Conditional on a Core Melt Accident (5 x 10 5).

external dose to the whole body due to the Whole body dose calculated includes:

possing cloud, exposure to radionuclides on ground, and the dose to the whole body from inhaled radionuclides.

Dose calculations assumed no protective actions taken, and straight line plume -

trajectory. .

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Figure 5.2 Conditional Mean Projected Whole Body Dose Versus Distance for Sheltering and Evacuation Stratectes. Projected Doses are Conditional on a WR " Melt-Through" Release (ikT. 6 and 7).

Curve A Indavidual located outdoors without protection. SF's (1.0, 0.71 1-day exoosure to radtonuclides on ground.

Curve B Sheltering, SF's (0.75, 0.33), 6-hour exposure to radionocitdes

+ on ground.

Curve C met terang SF's (0.5, 0.C8), 6-hour exposure to radionuclides on ground.

Y Curve D Evacuation, 5 hout delay tim , 10 MPH. .

N.

Chrve E tvacuation, 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> delay tune,10 TH. .*

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F ;are 5.3 Cudtttenal 956 Mvel Whole h>dy Cose Versus Otstr ce for Sheltering and Evacuation Strategtes. Projr<ted Cbnes are Condtttocal on a M %1t-Bretsp* Belease (M 6 and 7) .

S"'s (1.0, 0.7).

Curve A Indtvadual located outdoors without protection.

1-day eroosure to radionucitdes on ground.

Curse 8 S.elterino, SF's (0.75, 0.33), 6-hour exposure to radtenacitdes .

en ground.

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C.:ree . C 9.elttet.c, 3?'s (0.5, 0.09), 6+cor exx:;te te r a f t:tu-i M t * :

gr e :n?.

. .t. Curve D Evacuatten, 5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> delay tam , 10 V M. ,

• Qarve t Evacuation, 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> delay time,10 .wPH.

60 *

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E T qu:e 'i.3 Cond tt tenal .Wan Pro]ectml Wele hi', Ocsc Versas 0;stc .ce for 9 alter .rs; and tvac23 nan 5tra:-M.es. Pcc;ect rJ :oses are C N t::enal en 3 Nr

" At.?ogt.et sc* Felease f R.R t-u .

Curve A . Ind tvid;.21 locate; .>ut&c. - wiercu* 7!. tec? tm. SF'r (1.0, 0.71.

1-Jay equrare to *21.r.rg !:d s & ;;s:: 3.

Curw ?. Shelter t.91. ST's (0.7 5, C. ! 3; , f.ea.;r nrosara to e29:en; 1 ce: : ,

on ytour.d.

Curve C Dacart ton, *. hour delay t t c. 10 .wpu.

N t at . :. 3i's iu. ,.  ;-t. 9-cu* . cc r*

• o :2;;;: .-

Lt .(

m 1r ;. +*. ,

Curve C tvecaation, 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> delay t una, 10 MP't.

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'10 100 10 1-OlSTANCE (MILES)

Figure 5.10 CoMittenal 956 level Pro;ected '. Sole Ebdy tose versus Ctstance for sieltering and Evacuation Str ategtes. Frc)ee:ec' teses are Conditier al on - a N7 l

"Atinospheric* Pelease (NP 1-5).

SF's (1.0, 0.7).

Curve A Individual located outdoors without protection. 'l 1-day exposure to radscauclades on ground. I 6-hour exposure to radionuciades l Curve n Shelter ing, SF's (0.75, 0.33),

on ground.

) Curve C Evacuation, 5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> delay tise, 10 .St!.

s ..

Sheltering, SF's (0.5, 0.08), 6-*me exposure to radsonuc1 Aces Curve D ,

-on ground.

Curve E Evacuation, 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> ' delay time,10 MPH.

' .2 __

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EISIEEEEI.9E erg {ggg{gggg,gggg((((gg{ggg,;;,g{rygg,g;,,gHgggg My name is Steven C. Sholly. I am a Technical Research Associate with the Union of Concerne Scientists (UCS), Dupont Circle Building, 1346 Connecticut Avenue, N.W., Washington, D.C. 200 36. 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 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 'IMI Public Interest-Resource Center (TM IPIRC) in . Harr isburg , ._ Pennsylvania._ TM IP I RC , . w a s created after the Three Mile Island accident by concerned citizens groups in Pennsylvania. At TMIPIRC, 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 speciall committee of the New -York State Assembly on radiological emergency planning matters. I have _ . also testified before a

.__:_- __L__

.l .

Congress on' safety issues associated with steam generators in i

pressurized water reactors.

I~ 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 Thompson), 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 p;obabilistic 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 I

received a B.S. degree in Education (majors in Earth and Space Science and General Science, and minor in Environmental Education). I have also completed graduate coursework in land use planning. I am a resident of Columbia, Maryland.

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• 00CKETED USNPC 14 APR 17 40:34 UNITED STATES OF AMERICA t-F 1E OF SECRUnk <

NUCLEAR REGULATORY COMMISSI6sCKtilHG & SEPvlCf.

BRANCH B. E.F.O.R

. E..T.I.I

. E .A.T.O.M .I C..S.A.F.E.T.Y. .A.N.D.

. . L .I C.E.N.S..I N.G.

. .B.O.A.R.D 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) )

CgggggigArg_gg_gggvigg I hereby certify that copies of PALMETTO- ALLIANCE AND CA RCLINA ENVIRONMENTAL STUDY GROUP TESTIMONY OF STEVEN C.

SilOLLY ON EMERGENCY PLANNING CONTENTION NUMBER ELEVEN in the above captioned matter have been served upon the-following by deposit in the United States mail this 16th' day of April 1984.

James L. Kelly, Chairman George E. Johnson, Esq.

Atomic Safety and Licensing Office of'the Executive Board Panel' Legal Director U.S. Nuclear Regulatory. U.S. Nuclear Ecgulatory Consincion Commission.

- Kashington, D.C. 20555 Washington, D.C. 20555 Dr. Paul-W. Purdom Albert V. Carr, Jr.~,-Esq.

235 Columbia Drive Duke Power Company Decatur, GA .300 30 P.C. Box 33189

-Charlotte, NC 28242 Dr. Richard F. Foster.

P.O. Box 4263 Richard P. Wilson, Esq.

Sunriver, OR 97702 Assistant Attorney. General' State of South-Carolina.

Chairman P.O. . Box 11549 .' ,

Atomic Safety and Licensing: Columbia, SC 29211-Board Panel-U.S. Nuclear-Regulatory Chairman Commission  : Atomic' Safety and. Licensing Washington, D.C. .20555 Appeal Board Panel ~

U.S. Nuclear ~ Regulatory Commi'ssion Washington,.D.C'. 20 555; 1

-5 N., .

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s e

Jesse L. Riley William L. Clements 851.ienley Place Office of the Secretary Charlotte, NC 28207 Docketing and Service Section U.S. Nuclear Regulatory Bradley Jones, Esq. Commission Regional Counsel, Washington, D.C. 20555 Region II U.S. Nuclear Regulatory Don R. Willard Commission Mecklenberg County Department Washington, D.C. 20555 of Environmental Health 1200 Blythe Bouleva J J. Michael McGarry, III, Esq. Charlotte, NC 282C3 Bishop, Liberman, Cook, Purcell & Reynolds John Clewett, Esq.

1200 Seventeenth Street, N.W. 236 Tenth Street, S.E.

Washington, D.C. 200 36 Washington, D.C. 2000 3 Robert Guild, Esq. Palmetto Alliance Attorney-at-Law 2135 1/2 Devine Street P.O. Box 120 97 Columbia, SC 29205 Charleston, SC 29412 Karen E. Long Spence Perry, Esq. Assistant Attorney General Associate General Counsel North Carolina Department-Federal Emergency Management of Justice Agency P.O. Box 629 Room 8 40 Raleigh, NC 2760 2 500 C Street, S.W.

! Washington, D.C. 20472 Dr. Frank F. Hooper University of Michigan Morton B. Margulies, Chairman School of. Natural Resources Atomic Safety and Licensing -Ann Arbor, MI '48109 Board-Panel-U.S. Nuclear Reg ula tor y Dr. Robert F. Lazo Commission Atomic Safety and Licensing Washington, D.C. 20055 ' Board Panel U.S._ Nuclear _ Regulatory Commission Wa sh'i ng ton , D.C. 20555 5tEvEn~C'~5h5ffy Y

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