ML20065P644
| ML20065P644 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 10/12/1982 |
| From: | Finlayson F SUFFOLK COUNTY, NY |
| To: | |
| Shared Package | |
| ML20065P624 | List: |
| References | |
| ISSUANCES-OL, NUDOCS 8210260246 | |
| Download: ML20065P644 (50) | |
Text
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00g,ypfc UNITED STATES OF AMERICA u
NUCLEAR REGULATORY COMMISSION
'82 OCT 22 P4:31 Before the Atomic Safety and Licensing Board cr SEcatTAR'(
- H3 & SERviCL
>LNCH
)
In the Matter of
)
)
LONG ISLAND LIGHTING COMPANY
)
Docket No. 50-322 (0.L.)
~
)
(Emergency Planning (Shoreham Nuclear Power Station,
)
Proceedings)
Unit 1)
)
)
DIRECT TESTIMONY OF FRED C. FINLAYSON ON BEHALF OF SUFFOLK COUNTY REGARDING CONTENTION EP 14 (ACCIDENT ASSESSMENT AND DOSE ASSESSMENT MODELS) l l
October 12, 1982 i
O ph ADO l
S
i y
Summary Outlins of Dircct Tootimony of Fred C.
Finlayson on Behalf of Suffolk County Regarding Contention EP 14 (Accident Assessment and Dose ~ Assessment Models)
LILCO's dose assessment models neglect several important processes that lead to the accumulation of radiation doses.
Most importantly; the models calculate whole body dose without regard either for the dose resulting from the inhalation of fission products or the dose resulting from ground contamina-tion.
At distances close to the plant, the eight-hour ground dose contributes about 70% of the whole body doser beyond ten miles, the long term effects of inhalation exposure are the domir. ant source of whole body doses.
Thus, the neglect of these exposure processes in LILCO's models.may lead to signifi-cant underestimation of whole body doses as well as underesti-mation of doses to other human organs.
In addition; LILCO's dose assessment models also neglect important fission product source terms; apparently the noble gases and halogens are the only fission products that LILCO plans to use as input in calculating doses.
However; signifi-l cant accidents contributing to public risk will release many fission products in addition to the noble gases and halogens.
A comparative analysis of dose estimates using the very limited set of fissim products incorporated in the LILCO models and estimates i a more extensive realistic set of fission prod-ucts fr~
risk accident scenarios, indicates that the LILCO mr
.l.d substantially underestimate the magnitudes of predieu tes.
Finally, LILCO's estimates of noble gas and halogen re-lease rator. for use in the dose assessment models are based primarily upon arbitrary assumptions concerning the ratio of noble gases and halogens in the mixture.
The estimates are not based on measurements that reflect the actual fission product makeup of escaping radioactive gases and vapors.
Actual halo-gen release rates could be substantially different (either greater or smaller) from the estimated rates.,
Thus, the pro-jected thyroid dose estimates produced by the LILCO models are inherently unrealistic.
l O
S 1. _. _.
t 4
Attachments to'the Direct Testimony of Dr. Fred C. Finlayson on Behalf of Suffolk County Regarding Contention EP 14 (Accident Assessment and Dose Assessment Models)
ATTACHME'NT 1 Resume of Dr. Fred C. Finlayson ATT;.HMENT 2 Table 1 - Typical Fission Product Inventory for a BWR of Shoreham Size ATTACHMENT 3 Table 2 - SAI Categorization of Potential Severe Accident Releases for the Shoreham Facility ATTACHMENT 4 Figure 1 - Projected Probabilities of Exceeding Specified Doses as a Function of Distance From a Severe Nuclear Power Plan Accident ~(SAI Category 1)
ATTACHMENT 5 Figure 2 - Noble Gas Whole Body Doses (DBA) CBA-1 ATTACHMENT 6 Figure 3 - Mean Whole Body Dose for Severe Core Melt Accidents ATTACHMENT 7 Figure 4 - Whole Body Dose Components for Severe Core Mel-Accidents
l L
5 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the' Atomic Safety and-Licensing Board
)
In the Matter of
)
)
LONG ISLAND LIGHTING COMPANY
) Docket No. 50-322 (OL)
) (Emergency Planning (Shoreham Nuclear Power Station,
)
Proceedings)
Unit 1)
)
)
DIRECT TESTIMONY OF FRED C.
FINLAYSON ON BEHALF OF SUFFOLK COUNTY REGARDING CONTENTION EP 14 (ACCIDENT ASSESSMENT AND DOSE ASSESSMENT MODELS)
Q.
Please state your name and your position.
A.
My name is Fred C.
Finlayson.
I am the Principal Associate of F.
C.
Finlayson & Associates, 12844 East Cuesta Street, Cerritos; California. A copy of my professional quali fications is attached to this testimony as Attac'hment 1.
Q.
What is the purpose of this testimony?
A.
The purpose of this testimony is to address certain of the concerns raised in Suffolk County's emergency planning conten-tion EP 14--Accident Assessment and Dose Assessment Models.
I The contention states:
LILCO's plan fails to provide reasonable assurance that adequate methods, systems and equipment for assessing and monitoring actual or potential offsite consequences of a radiological emergency condition are in use, and therefore does not comply with 10 CFR 50.47(b)(9).
i 1
l
.o a
In this testimony; I will address the issues in contention EP 14 associated with the adequacy of LILCO's methods of accident and dose assessment; in particular the methods used for dose assessment.
Q.
Please describe your experience relative to methods for assessing consequences of a radiological emergency.
A.
For the past ten years, I have been working extensively in a variety of projects related to the evaluation of reactor safety.
Over the last six years, I have been responsible for directing, managing, and performing research projects in which reactor risk assessment and public health consequence assess-ment for severe nuclear power plant accidents were significant factors.
For example, I supported the State of California, office of Emergency Services, in the technical management of
^
site unique probabilistic risk assessments and consequence ana,
lyses conducted for each of the nuclear power plant sites in California.
I am currently a member of the review committee for the "PRA Procedures Guide" (NUPEG/CR-2300) that is under development by the NRC, the American Nuclear Society (ANS) and
(
the Institute of Electrical and Electronic Engineers (IEEE).
As a member of the review committee, I have concentrated my efforts primarily upon evaluating the consequence assessment methods recommended in the PRA Procedures Guide.
Y l
s 1
In addition; I have been involved in the assessment of radioactively induced health effects and fallout patterns from nuclear weapons since 1960.
More recently; I managed a program in which the relative consequences of severe nuclear power plant accidents in above-ground and underground facilities were examined.
In al.'. of these studies, state-of-the-art radioac-l tive plume modeling and public health consequence assessment models have been used.
Q.
Are you familiar with the methods, proposed to be used by LILCO, for assessing and monitoring actual or potential conse-quences of a radiological emergency?
A.
I have reviewed Section 6.1 of the Shoreham Emergency Plan, Section 2 of the Shoreham Nuclear Power Station Offsite Dose Assessment Methodology for Emergency Applications prepared by Entech Engineering; Inc. (the "Entech model"), and the Shoreham Emergency Plan Implementing Procedure SP 69.022.01 Revision 0, Determination of Offsite Doses (effective date July 9,
1982).
It is my understanding that these documents reflect the methods which LILCO will use to project offsite doses, prior to receipt of the results of offsite surveys and sam-pling, in the event of a radiological emergancy.
I have also reviewed Emergency Plan Implementing Procedures SP 69.023.01, Revision 0, Thyroid Dose Commitment Using TCS Air Sampler
( effective date July '9, 1982), and SP 69.023.01; Revision 0, T
Downwind Surveys (effective date July 9; 1982), which reflect methods that LILCO proposes to use to determine thyroid doses based upon offsite surveys and sampling conducted during an emergency.
As far as I know; these documents contain all the information provided by LILCO to the County with respect,to the accident and dose assessment methods to be used by LILCO in the event of an accident.
Q.
Please explain what is meant by the term " dose assessment model."
A.
In the broadest sense, a dose assessment model is a mathe-matical tool by which the health impacts of accidentally re-leased radioactive fission products upon humans can be deter-mined as a function of: magnitude of the accidental releases i
(including both the quantities and specific constituents of the released fission products); the wind and weather conditions at i
the time of the accident; and, the distance of individuals from the site of the accident.
Protective actions taken by the individuals, such as evacuation; and taking shelter from direct exposure to the radioactivity; can also have a significant
~
impact on the magnitude of potential radiological doses that they might receive.
The influences of protective actions on potential doses to exposed individuals are also customarily included in dose assessment models.
Thus, the basic elements
~
of a dose assessment model are (1) an accurate description of
4 the characteristics of the radionuclide release; (2) an acceptable mathematical model of the radioactive plume propaga-tion and dispersion that can be used to define the concentra-tion of radioactive fission products in the air and on the ground; (3) an adequate description of weather conditions, in-cluding atmospheric stability, wind speed and-direction, and precipitation; (4) a model of the relationship between health effects and radioactive exposures; and, finally; (5) a method for evaluating the effects of protective actions taken by indi-viduals to reduce radioactive exposures and/or doses.
O.
In your opinion, do the dose assessment models proposed to be used by LILCO prior to obtaining the results of offsite sur-veys provide an adequate method for assessing actual or poten-tial offsite consequences of a radiological emergency condi-I tion?
A.
Procedure SP 69.022.01 refers to two different LILCO dose i
assessment methods for estimating offsite doses.
One is a com-puterized " radiation monitoring system" (of which the Entech model is evidently one part) and the other is a manual method.
Neither the Entech model nor the LILCO Plan or procedures I have reviewed provide any details on the input to the computer-ized method, nor its output.
I understand, however, that the f
manual model incorporates the same mathematical procedures for
(
calculations to be performed as the computerized system does.
l l
1.
s
[ Deposition of H. Mark Blauer, August 23, 1982; at 146-1473 The mathematical procedures in the LILCO model are based upon a straight-line trajectory model using Gaussian dispersion rules that were derived from the NRC Regulatory Guide 1-111 (Ref.;
Entech Model; p.5).
The manual method is based upon utilization of eight pre-calculated nomograms that provide the operators with pictogra-phic tools for scaling measurements of the radioactivity of fission products escaping through the station ventilation sys-tem and the Reactor Building Standby Ventilation System (RBSVS), to estimate offsite whole body and thyroid doses at locations no more than ten miles from the plant site.1!
Many attempts have been made to develop accurate plume propagation and dispersion models.
No thoroughly satisfactory methods for advance prediction of cloud motions exist as.a're-sult of the complexity of realistically modeling the motions of the earth's atmosphere.
However, over the years, many simpli-fying modeling assumptions for assessing radioactive clon.4 motions have been r ade that have been broadly accepted by the technical community.
The straight-line, Gaussian plume disper-sion model adopted by LILCO is relatively unsophisticated, but 1/
The current e?ition of SP 69.022.01 includes only five of the eight nomograms; it indicates that the remaining three will be provided "later."
~..
f one that is widely accepted in principle; within the technical community; for conducting long-term (annualized) risk assess-ments. These types of models are not generally useful for ad-vance projections of real-time meteorological events -- espe-cially those where the predicted motions of the cloud must be accurately portrayed over long distances.
Changes in wind direction over time and distance will frequently cause the plume trajectory to deviate from a straight-line.
- Moreover, under certain ve,ry stable or nonstable atmospheric conditions, the Gaussian assumptions for dispersion begin to be invalid.
When used for the advance projection of doses from real meteor-ological conditions, it is generally' conceded that such models can be accurately applied over distances no greater than ten to twenty kilometers (or about five to ten miles in round num-bers)(Ref: PRA Procedure Guide; NUREG/CR-2300 (Draft), Appendix D,
- p. 4).
l The plume atmospheric dispersion and dose modeling methods that are outlined in the Entech model follow NRC Reg. Guide recommendations (Reg. Guides 1.111 $ 1.145) reasonably well.
However, it should be noted that the Reg. Guides cited were developed primarily for establishing long-term (annualized) l risks from operation of nuclear power plants.
For these cir-cumstances, the use of straight-line trajectory, Gaussian plume models is relatively reasonable since, over the course of a _ _
o' year; clouds would follow meandering paths that would cover essentially all geographic locations around a site.
Under these circumstances, straight-line trajectory models that in-corporated directional probabilities based upon annualized wind-rose frequencies would be expected to provide results that would be similar to more sop 51sticated numerical methods that simulated actual meandering in cloud paths.
For purposes of predicting off-site doses following an accident, however, the straight-line, Gaussian plume models have far greater uncertainty in their results, than they do for the calculation of annualized average values of potential doses.
Therefore, the plume dispersion' aspects of the LILCO mode.ls are not particularly well suited for real time projec-tion of cloud paths or for accurate estimation of the locations for particular radioactive exposure levels.
They do, however, provide a first-order basis for estimation of cloud fission product concentrations that might occur in a straight line, downwind from an accident.
However, the LILCO models evidently include certain as-(
sumptions concerning the number and quantity of the constituent fission products in an accidental release that are not suf-ficiently general for the spectrum of severe accidents that emergency plans must consider.
Therefore, in this respect the models are not currently applied in an adequate fashion.. - - --
o 0
Q.
Please describe in more detail the inadequacies in LILCO's
~
dose assessment models with respect to the assumptions concern-
)
ing fission product releases.
A.
My conclusions are based upon the limited amount of infor-mation available in the documents that I have reviewed relative to the LILCO fission product relenae assessments. These docu-ments suggest that the fission product constituents that are considered in the LILCO analysis are ILmited to noble gases and some fractional releases of halogens (principally iodine iso-topes). In fact, it appears that LILCO intends to use only these two categories of fission products without regard to po-tential releases from the reactor of a much larger spectrum of the fission product inventory under more severe accident condi-
/
tions that would provide a greater challenge to emergency plan-ning.
The accidental releases in these latter categories would l
be expected to be the source of greater public risks than those associated with releases of only noble gases and a small frac-tion of the iodine fission product inventory.
A review of Shoreham Procedure 69.022.01, Section 3.4, indicates that the LILCO models assume a 100% release of noble gases and a 25% release of the halogen inventory within the containment building.
They then apparently assume that inter-nal stack filters will have a 99% efficiency for halogens.
l Thus, only 1% of the original 25% of the halogens released in.
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the containment building would be released to the atmosphere (i.e.,
the fractional release of halogens to the atmosphere would be.0025 times the initial halogen inventory in the reac-tor core).
Q.
What is the difference between LILCO's use of a fission product release inventory consisting of only noble gases and halogens, and the fission product release categories you have described for severe reactor accidents.
A.
Table 1 (Attachment 2 to this testimony) contains a de-scription of the inventory of the more significant fission products for a Boiling Water Reactor (BWR) such as Shoreham.
Listed on Table 1 are the quantities of the more significant isotopes that are contained in the reactor core.
Of the ele-ments listed, those that are bracketed fall within the cate-gories of " noble gas",and " halogen" (or iodines) as noted.'
In the event of a severe accid.ent, fractional releases of some or all'the elements contained in the fuel could occur to the atmosphere.
The precise contents of the release would vary depending upon the events involved in the accident.
Accidental releases are customarily grouped into categories that reflect common accident characteristics.
Such a grouping was performed for the Shoreham plant by Science Applications Inc. (SAI) in the Probabilistic Risk Assessment (PRA) that they conducted for LILCO.
Table 2 (Attachment 3 hereto) has been reproduced from
the SAI-PRA.
It sets forth the release descriptions developed by SAI for five of the more probable classes of severe acci-dents (though all are projected to have low absolute probabili-ties).
The fraction of the total inventory of chemically and phy-sically related groups of fission products that would be re-leased in the event of an accident are shown in Table 2.
The results shown in Tables 1 and 2 indicate that accidents can result in releases.of many other fission products besi, des the noble gas and halogen categories.
Although many of the fission product groups have relatively small release fractions, I will show belo'w that calculations which exclude those fission prod-ucts would lead to significantly lower dose estimates than those that are obtained from analyses that include their effects.
O.
Does the fact that certain fission products are not incor-porated as inputs to the LILCO model have any impact on the whole body doses that would be calculated using that model?
A.
If the LILCO whole body dose predictions are based exclu-sively upon the noble gas fission product releases, estimated doses would generally be relatively low compared with those that would be calculated with the fission product inventories shown in Table 2 -- i.e.,
those that SAI found to represent the releases of the more significant risk inducing accidents at - --
4 Shoreham.
This is the result of two factors:
The first is associated with the physical mechanisms by which individuals accumulate radiological doses; and the second is associated with the importance of fission product constituents besides noble gases and iodines that contribute to whole body doses.
I will address each of these factors in turn.
There are five dominant ways in which people can accumu-late radiation doses after an accidental release of radioactiv-ity to the atmosphere:
1.
Inhalation.
2.
Exposure to external irradiation from the passing cloud (cloudshine).
3.
Exposure to external irradiation from the deposited radionuclides (groundshine).
4.
Ingestion, including contaminated vegetation; miik, milk products; and crops contaminated by root uptake.
5.
Inhalation of resuspended radioactivity.
For estimating early effects such as deaths or injuries resulting from exposure to the radioactive cloud; the most important of these pathways are (1) inhalation from the cloud, (2) cloudshine', and (3) short-term exposure from contaminated ground (over periods lasting from hours to days).
For estimat-ing latent health effects such as cancers, the importanc path-ways include (1) external exposure from contaminated ground
t (both short and long term), (2) inhalation exposure from the passing cloud and from the subsequent resuspension of radionu-clides; and (3) the ingestion of contaminated foods.
In emergency planning for severe nuclear accidents, early effects; including deaths and injuries occuring within 30 to 60 dayA after exposure (or less) are most frequently of concern.
Latent cancer effects from relatively short periods of external exposure to contaminated ground and inhalation exposure from the passing cloud can also be strongly influenced by protective actions associated with the emergency plan.
An effective dose assessment model will include consideration of all such expo-sure processes.
The first and most significant weakness in LILCO's e:xclu-sive use of noble gases to estimate the whole body dose, is that the noble gases contribute almost exclusively to the cloud doser they play an essentially negligible role in inhalation doses and do not contribute at all to ground doses.
As pre-viously noted, many of the chemically active radioactive fis-sion products (as opposed to the chemicals; inert noble gases) that are inhaled into the lungs during the time of the passage of the cloud, as well as those which are on the ground, also contribute significantly to the whole body dose.
Therefore, a realistic analysis o'f the whole body dose must consider the contributions from each of these elements -- materials inhaled -
into the lungs, the radioactive cloud emissions during its passage; and radioactive materials that have fallen out of the cloud and have been deposited on the ground.
If the Whole body dose assessment were limited to the contributions from noble gases; the contributions to health. effects consequences from inhalation dose and ground dose would, for all practical pur-poses, be neglected.
The second weakness associated with the exclusive use of noble gases for predicting whole body doses is that they repre-sent a relatively small fraction of the significant fission product contributors to the total cloud dose that might be de-livered to a'n individual at a given distance from the reactor under the risk-inducing accident scenarios shown in Table 2.
As shown in Table 1 and Table J2; many fission products, in ad-dition to noble gases, would be released in the event of a 'se-vere accident.
A number of these other fission products, car-ried along in the cloud, could be significant contributors to i
the Whole body cloud dose; such as Te-131 & 132;~Sb-129; I-131, 132, 133, 134, & 135; and La-140, for' example.
Thus, if LILCO's assensment of whole body dose were limited to the noble gas contributions to the cloud dose; important health-effects contributions from fission products other than noble gases that might be carried in the cloud 'could be neglected.
14 -
Q.
Can you quantify the effect of the LILCO model's neglect-ing certain fission products in the calculation of whole body doses?
A.
Yes.
Figure 1 (Attachment 4 hereto) shows the Whole body dose; as a function of distance and probability, for typical Shoreham weather conditions given that one of the more severe accidents (SAI Category 1) has occurred.
Figure 2 (Attachment 5 hereto) shows the results that are obtained if you calculate the Whole body dose based exclusively upon a 100% release of noble gases, as the LILCO model assumes.
It can be seen that there is a subst,antial difference in the values of the doses 4
associated with the release of a full range of fission products as compared to the case Where only the effects of noble gases are considered.
For example, Figure 1 indicates that the prob-ability of exceeding 200 rem (a Whole-body dose at which most exposed individuals would show definite signs of early injuries) exceeds one percent out to a distance of about four miles for the SAI Category 1 accident scenario shown.
In Figure 2, where it is assumed that,only noble gases are re-leased, the probability of exceeding a 200 rem dose becomes vanishingly small beyond 1 mile from the reactor.
At 30 rem (a relatively large dose, but one for which ordinary laboratory or clinical methods would generally show no indications of radioactively-induced early injuries), the results shown in 1,
l
~
t Figure 1 indicate that there is a 1 percent probability of exceeding this dose out to about 30 miles; and a 10 percent probability (or greater) of exceeding 30 ren out to about 15 i
miles.
In Figure 2; the probability of exceeding 30 rem be-comes vanishingl'y small beyond 4 miles.
Thus; the results shown in Figures 1 and 2 suggest that for the risk-inducing accidents involving the release of the complete spectrum of fission products; the distance at which there is an equal prob-ability of exceeding a opecified dose can be increased by a multiple of about five to ten.
Q.
Please explain how the curves shown in Figures.1 and 2 were derived.
A.
The curves in Figures 1 and 2 are based upon a series of calculations using the CRAC-2 code.
The CRAC-2 code is a com-puterized risk assessment program that can be used for assess-ing doses on a probabilistic basis; or for direct evaluation of t.he long-term probability of health effects of all types from accidental releases of radioactivity from a nuclear power plant and/or the associated economic consequences.
The operator of the code may define the fission product source term as broadly or as narrowly as he wishes within the limits of the 54 radio-active isotopes shown in Table 1.
As indicated in Table 2, fission product releases may be categorized in terms of their probabilities and results for individual accident scenario
categories evaluated singly; or several categories may be combined into integrated estimates of doses and consequences in accordance with their individual probabilities.
Risks are annualized by using a statistical sampling of a year's worth of hourly weather data for the site as a basis for defining wind-speed, stability; and precipitation characteristics of the local weather at the site.
The annualized risks are generally based on the results of conrequence calculations for a sample of about 100 explicit hourly weather sequences taken over the representative year's worth of data.
The release characteristics for Figure 1 are exactly those shown for the SAI Category 1 release shown in Table 2.
In per-forming the CRAC-2 calculations for the noble gas releases, the same set of atmospheric and meteorological conditions were used f
that were utilized in the assessme'nt of the doses resul. ting from the more extensive set of released fission products used in the calculations for Figure 1.
However, for the noble gas induced consequences shown in Figure 2, we have restricted the o
accidental releases from the plant to an exclusive release of l
100% of the noble gases.
O.
How do the results reflected in Figures 1 and 2 relate to the results that would be obtained using the LILCO dose assessment models?
l l l
l
t A.
Aside from the mathematical models for estimating plume dispersion and fission product concentrations in the radioac--
tive cloud that are described in the Entech model; as discussed above; it is difficult to be sure exactly what is embodied in LILCO's' manual dose estimation procedures or in the computeri-zed method.
This is because the exact makeup of the fission products that are used as inputs by the Entech Jmalel are not presented in the document describing the me, del; nor is the basis for the development of the nomograms specified in the Procedures document.
Assuming that the 100% noble pas fission productreleaseisthebasisforthewhole,kddydoseestimates r.
shown in the LILCO nomograms; then for comparable fission prod-uct release rates the whole body doses predicted by the LILCO nomograms should be couparable to the noble' ggs-on,1y curve
.c i
shown in Figure 2.
Accordingly; the LILCO who n' body dose' pro-jections would apparently be substantially lower than,,phojec tions based upon a more extensive fission product sou'rce term Ns -
description, as shown by comparing Figures 1 and 2.
^J Q.
Would the substantial difference in dose projectio,n that,>
s you just mentioned have an impact on emergency planning,deci-sions made by LILCO?
A.
Yes, it seems very likely that it would.
If the LILCO procedure is limited to assessment of doses which are based'~ eg upon an assumed release ratio of noble gases to halogen fission sq J {s-
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products, then a more severe accident; with a more extensive
'. fission product release; would probably result in doses greater than those predicted by the LILCO models.
If lower' doses were
' predicted by the use of the LILCO model than would have been
~
e' predicted using a more'significant description of released fis-i n
- m sion products,. then the decisions necessary for successful pro-
~.
i tactive actions might not be made in a timely fashion.
The e
operators might not realize that dose-projections exceeding PfcItactive Action Guide (PAG) triggering levels could be rea-s
.s ched :in the early stages of the event.; Even if subsequent
.measuremen$s indicated 'het dose lev.els had exceeded expecta-
'r I
s'"tions and PkG trigger as were in fact exceeded (for instance,.after offsite sampling had been done and reported q
back to the site),, valuable time would have been lost as a re-
,7 sult of initial reliance upon low doce projections which were w
not properly modeled.
If; on-the other hand, a more realistic dose modeling method were utilized, the projee;tions of the action-triggering dose levels might be!provided at an earlier l
[oint in the accident sequence; thus providing more time for M
r, implementing the necessary protectiveg,etions.
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O.
Is there any other impact on the adequacy of LILCO's whole
'J body dose assessment. arising from the consideration of only l
l noble gases?
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A.
Yes.
As I mentioned before, there is an implicit assump-tion in LILCO's use of only noble gases in projecting the whole bcdy doses that the dose is contributed exclusively by the cloud.
In fact, if the breakdown between the contribution from the cloud and the two other major sources of dose in Figure 1 were shown -- those contributions from inhalation through the lungs and from fission products entrained on the ground -- we would find that both the inhalation and ground dose provided major contributions to the total whole body dose.
Figure 3 (Attachmer.t 6 hereto) and Figura 4 (Attachment 7 hereto) show a representative breakdown of contributions from inhalation, cloud and relatively short (8-hour) ground dose exposures.
From Figure 4, it can be seen that at close-in distances the 8-hour ground dose contributes about 70% of the total whole-body dose.
Beyond ten miles, the long-term effects of inhalation exposure to'the cloud become the dominant source of whole-body doses.
Over a period of a few days, contributions from each of the three elements of the whole body dose are quite similar in magnitude. For all practical purposes, both the inhalation and ground dose contributions are neglected if you assume you are only releasing the noble gases, or that only cloud doses are significant -- an explicit feature of the LILCO, dose assessment models (Ref.: Entech Model, p.2).
Q.
So far we've only been discussing whole body dose calcula-1 tions.
Have you reviewed the thyroid dose calculations con-tained in the LILC'O dose assessment model?
A.
Yes, I have reviewed those referenced in the Entech model, in SP 69.022.01 and in SP. 69.023.01.
Q.
According to the LILCO models, how are thyroid doses pro-jected prior to receipt of offsite survey results?
A.
The LILCO procedure SP 69.022.01 indicates that nomograms are u' sed to project thyroid doses in the following manner.
The first step in applying.the nomograms requires the user to esti-mate the iodine release rate.
Under normal conditions, the
~
person responsible for making the thyroid dose predictions de-rives his initial estimate of the radioactive iodine source strength from a reading of the effluent monitor in the Reactor Building Stand-By Ventilation System (RBSVS) stack.
This moni-tor measures the radioactivity being released from the stack in terms of the " counts per minute" of ionizing particles released I
by the fission products as they pass by the monitor.
The Radiation Protection Manager (or an aide) then uses ths nomo-grams to estimate the release rate of iodines.
The use of the comograms depends upon several assumptions that are not very explicitly defined.
One is related to the constituency of the mixture of gaseous and vaporized fission products assumed to be escaping from the containment building.
Radioactivity measurements taken from the station ventilation I
system exhaust monitor are evidently assumed to be derived from only noble gas fission products (Xe and Kr).
The nomograms based upon RBSVS monitor readings appear to incorporate an
.i inherent assumption about the specific fission product ratios of the mixture of noble gases and halogens (principally Iodines) escaping from the stack.
How the operator can tell from the radioactivity measured from these sources whether the ratios of noble gases to halogens is in accord with the assump-tion, or what he would or should do if it were not, is not described in the procedures.
O.
Is the technique used to estimate the iodine release rate in this procedure justified?
A.
Not necessarily.
The ability of an operator to estimate accurately the actual iodine release rate depends upon his.' rec-
.ognizing that there may be various kinds of accidents at the plant.
Some of them may release iodine quantities in propor-1 i
tions that differ from those that have been assumed in the methods incorporated into Procedure SP 69.022.01.
Accordingly, the reasonableness of the iodine release rate estimates depends upon the adequacy and the accuracy of the measurement 7.sthods used.
Unless a more sophisticated procedure is used to estab-lish the fission product makeup of the gases escaping from the reactor containment building, it is not clear to me that the
i operator can accurately determine whether the reading in counts per minute given by the stack monitor was derived from gamma rays from iodine or from some other gamma emitting fission product.
If the operator is to make reasonable estimates of the potential ofir. ite thyroid doses, then it is very important that he be able to make accurate estimates of the actual radio-f active iodine release rates.
In my opinion; a more definitive measurement method would be a desirable improvement over LILCO's apparent reliance on an assumed fission-product make up of escaping gases.
Procedure SP 69.022.01 does specify in Section 8.2.1.11 (p. 6) that if the radiation monitor is either inoperable or its reading is off-scale, a " grab" sample from some source (probably the RBSVS stack) is to be,obtained as a basis for the source estimates for iodine or xenon release rates.
I am not sure what equipment will be used by LILCO'to assess the fission product makeup of the grab sample or how ouickly such assessment could be accomplished.
Under ideal circumstances, a sample of this type could be used to provide a more accurate definition of the constituent fission product makeup of the escaping gases.
However, becauce the LILCO models identify the grab sample method as one to be used only if the less accurate radiation monitors are unavailable, LILCO appears to base its thyroid dose assessment primarily on iodine release rates using assumed rather than measured values. - -
s Q.
In your opinion, does the use of an assumed ratio of no'ble gases to iodines in the stack effluent represent a conservative approach for purposes of thyroid dose assessment?
A.
This again involves the issue of the realism of the opera-tor's projection of the iodine release rate from the contain-It is very impor5 ant for an operator to have accurate ment.
estimates of the quantity of iodine actually released if his estimates of +he thyroid doses are going to be accurate. If the operator makes his estimate of the iodine release rate from the nomograms of Procedure SP 69.022.01 based upon stack monitor count rate from the RBSVS, then the actual ratio of iodines to other gamma emitting fission products is important.
If the actual fractional quantity of iodine in the stack gases is greater than the assumed values upon which the nomograms are based, then the Radiation Protection Manager's estimate of' thy-roid dose would almost certainly be too low.
Hence it does not appear that the approach used can always be considered conser-i vative. If we have reason to believe that the operator's re-I lease estimates may not be accurate, then it'would be more ap-propriate to use a more accurate source of data for defining the iodine release rate.
If this could not be done in a timely fashion, then it would be more conservative to assume higher fractional concentrations of iodine to the releases for pur-poses of dose projections.
In my opinion, those estimated l
l 0
concentrations should make use of the results of a study of severe accidents and their respective probabilities such as a thoroughgoing PRA.
O.
In your opinion; do the LILCO dose assessment models demf onstrate that the methods to be used to assess thyroid doses prior to receipt of data from offsite survey teams are ade-quate?
A.
As indicated above, there is a need for better methods of determining the iodine release rates from an accident.
In ad-dition, all the weaknesses previously described with respect to the size and propagation path of the radioactive cloud in the atmospheric dispersion methods of the Entech model are as ger-mane to thyroid dose estimation as they are to whole-body dose estimates.
The key question is whether the models have demonstated adequacy.
In the LILCO documents I have reviewed, in my opinion, the documentation does not now demonstrate-ade-quacy, although the methods could be the basis for reasonably adequate thyroid dose predictions; if the qualifying weaknesses were eliminated.
i O.
Could the concerns you have identified with respect to LILCO's thyroid dose assessment have an impact on emergency planning decisions?
i l
A.
Yes.
For the reasons I discussed above, the nomograms as l
l they are now constituted could conceivably yield unrealistic l
doses estimates that might not meet protective action guides (PAG) criteria for implementing protective actions; when a more reliable estimate would have called for their implementation.
Or; under other circumstan'es; unreliable dose estimates might trigger decisions to tLke protective actions unnecessarily.
O.
Do you have any other concerns about the adequacy of the LILCO dose assessment models?
A.
Yes.
According to SP #9 022.01; the dose estimates provi-ded by the nomograms are restricted in their applicability to ten miles or less from the plant.
Q.
What is the significance of limiting the LILCO dose assessments to distances of ten miles or less from the plant?
A.
As seen in Figure 1, there is a high probability of exceeding the 1-5 Rem whole body dose protective action guide limits at distances beyond the ten-mile range.
In fact,'it shows a 50% probability of exceeding the 1 Rem PAG level at distances of thirty miles. Although it might be possible to extrapolate from the LILCO models to determine doses at dis-tances beyond ten miles from the plant; such an extrapolation would require the operator to observe that the dose values at the ten-mile limit exceeded the PAG levels and then to use some simplified procedure for estimating how far beyond the arbi-trary 10 mile boundary the dose levels would be exceeded.
If advance preparation for such a procedure is not made, it is not
l j
clear what mechanism might be used in an ad hoc fashion to make the extrapolations.
Consequently; I would recommend that ad-vance preparations be made to provide the Radiation Protection Manager with this capability before he must make the judgment under emergency conditions.
Q.
Dr. Finlayson; please summarize your testimony on EP 14.
A.
In summary; LILCO has the basic mathematical tools for a relatively unsophisticated cloud dose assessment model that is based upon straight-line trajectory, Gaussian dispersion meth-ods.
As presently constituted; however, this model neglects several important processes that lead to the accumulation of radiation doses.
For emergency planning purposes where rapid responses are required, the most important among these neglec-ted irradiation processes are the inhalation process leading to g
internalization of fission products within the body, and the i
ground dose exposure process.
As indicated in Figures 3 and 4, neglect of these exposure processes may lead to significant underestimation of whole body doses as well as underestimation of doses to other human organs.
In addition, the written evidence provided in LILCO's' doc-umentation of the dose assessment methods suggests that impor-tant fission product source terms are being neglected that should be included as input to the mathematical tools of the plume dispersion, radioactivity concentration, and --_
exposure-to-dose conversion elements of the methodology.
Apparently; the noble gases and halogens are the only fission products that LILCO plans to use as input to their dose assessment modeln.
As indicated in Table 2; probabilistic risk assessments of the Snoreham 'l'nt have shown that the more sig-pa nificant accidents contributing to public risks will release many fission products in addition to the noble gases and halo-gens.
The results of comparative analyses of dose estimates for the limited set of fission products incorporated in the current LILCO procedures and those for the more extensive re-leases from high risk accident scenarios indicate (as shown in Figures 1 and 2) that the LILCO procedures could substantially underestimate the magnitudes of predicted doses.
Noble gas and halogen release rates are primary inputs to 8
LILCO's '.nathematical dose assessment tools.
LILCO's principal method of estimating these release r,ates is based upon measure-ment of ionizing radiation from fission products escaping l
through the normal plant ventilation system and/or the reactor i
building standby ventilation system.
The monitors in these ventilation systems do not provide an explicit method for determining the details of the fission product make up of the escaping radioactive gases and vapors.
Thus LILCO's dose esti-mation procedures are based upon arbitrary assumptions concern-ing the ratios of noble gases and halogens in the mixture.
Actual halogen release rates could be substantially different (either greater or smaller) than the estimated rates that are based upon the ventilation system monitor measurements.
Since the projected thyroid dose estimates are directly related to the magnitude of the input halogen release rates, the output doses will be inherently unreliable if the input data is not precisely defined.
l LILCO's projected whole body dose estimates could lead i
decision makers to fail to institute important protective actions since they may be substantially lower than actual doses in the field.
On the other hand; LILCO's thyroid dose esti-mates may.be inherently unreliable, leaving decision makers' without a reasonable basis for instituting protective action decisions.
In both cases; improvements in the matheds for pre-dicting such doses appear to be needed.
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l ATTACHMENT 1 l
l l
1 1
l ATTACHMENT 1 REACTOR SAFETY ASSESSMENT FRED C. FINLAYSON PRINCIPAL ASSOCIATE NUCLEAR POWER PLANT F.C. FINLAYSON & ASSOCIATES PROBABILISTIC RISK 12844 E. CUESTA STREET ASSESSMENT CERRITOS, CALIFORNIA 90701 REACTOR ACCIDENT CONSEQUENCE ASSESSMENT ENERGY SYSTEMS DESIGN AND ANALYSIS BACKGROUND
SUMMARY
Dr. Fiklayson has extensive experience in the field of assessment of the safety and risks of nuclear power reactors. He recently provided technical direction of the probabilistic risk analyses conducted for the State of California's evaluation of Emergency Planning Zone requirements. He was the principal investigator and program manager of the NRC's first investigation of the adequacy of human engineering in nuclear power plant control rooms under severe accident conditions.
He is currently conducting an investigation for the NRC of the feasibility of instituting a specific reporting system for human errors in nuclear power plants. Dr. Finlayson was also the manager of The Aerospace Corporation program that provided systems integration and technical direction of the California Energy Commission's study of underground nuclear power plant designs, costs, and their relative effectiveness in reducing the consequences of extremely severe accidents.
Dr. Finlayson has been a consultant to the NRC, U.S. General Accounting Office, and other federal and state governmental agencies on nuclear safety related issues such as site-specific risk analyses, human engineering, large-scale reactor test program. design and effectiveness, sabotage, waste transport hazards, and a wide variety of other related toples. He is a member of the review committee for the "PRA Procedures Guide" (NUREG/CR-2300) for probabilistic risk assessments for nuclear power plants that is being prepared under a joint NRC-industry-technical society effort. He served as a member of the NRC's 1980/1981 LOFT Special Review Group and was consultant to the NRC's Rogovin Special Inquiry Group in their investigation of human engineering factors associated with the Three Mile Island incident. He has performed several t
l assessments of the design and effectiveness of ECCS for LWRs, including the analysis conducted for the American Physical Society's Review Committee (1975) on Light I
Water Reactor Safety.
EDUCATION BS Mechanical Engineering, Brigham Ycung University,1958 PhD Mechanical Engineering, Northwestern University,1964 EXPERIENCE
[
l The Aerosoace Corooration. Los Angeles, CA (19"2-Present). Dr. Finlayson is currently Manager, Nuclear and Geothermal Systems, Energy and Resources Division. In this l
capacity, he is responsible for nuclear, geothermal, and energy cor" 2rvation projects.
l
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. = _ - -
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He directed the systems engineering and technical management efforts for the recent California study of statewide nuclear power plant risks and associated emergency planning zone requirements; and directed a similar program for a major study of underground nuclear power plant siting. He was also the program manager for an assessment of the impact of plutonium fuel cycle safeguards, and an evaluation of nuclear control room human engineering. He has also managed and performed systems analyses of industrial process heat applications of geothermal power as well as conceptual design and evaluation studies of hybrid solar / geothermal power systems.
Studies of local and national energy consumption patterns and the effectiveness of selected conservation measures have also been performed under his direction.
Physics International Company, San Leandro, CA (1968-1972). Dr. Finlayson directed and conducted research in strategic and tactical weapon systems survivability /
vulnerability, numerical analyses of the propagation of strong shocks.in geologic media and structural materials, and structure-medium interactions.
The Aerospace Corporation, Los Angeles, CA (1964-1968). Dr. Finlayson conducted investigations of ground based system survivability to all relevant effects of nuclear weapons.
The General American Transportation Corporation, Chicago, Ill (1960-1964).
Dr.
Finlayson conducted research on the interactions of strong shocks in air and earth materials with above ground and buried structures.
PROFESSIONAL' ACTIVITIES Dr. Finlayson is a registered Professional Nuclear Engineer in the State of California.
He is a member of the American Nuclear Society.and the Institute of Electrical and Electronics Engineers (Reliability Society),.
PUBLICATIONS
" Closures for Hardened Protective Hangers", AFSWC-TDR-62-77, MRD Division of General American Transport Corporation, Niles, Illinois, August 1962.
" Air Blast Load Reduction on Above Ground Structures", Proceedings of the 32nd Symposium on Shock, Vibration, and Associated Environments, Part II, Bulletin No. 32, Office of the Director of Defense Research and Engineering, November 1963.
" Design Procedures for Shock Isolation Systems of Underground Protective Structures, Volume II, Structure Interior Motions Due to Directly Transmitted Ground Shock",
i AFWL RTD-TRD-63-3096, Vol. II, General American Transportation Corporation, Niles, Illinois, December 1965. (Coauthor).
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" Wave Interaction of a Viscoelastic Medium with an Elastic Cylindrical Shell", Journal of the Acoustical Society of America, Vol. 40, No. 6, pp.1496-1500, December 1966.
(Coauthor).
" System Vulnerabilities to Craters and Ejecta", (U) Proceedings of the Symposium on Nuclear Craters and Ejecta, Vol. I, SAMSO-TR-68-107, November 1967 (S-RD).
Proceedings of the Symposium on Nuclear Craters ani! Ejecta, (U) Vol. I and II, SAMSO-TR-68-107, November 1967 (S-RD). (Coeditor).
"A Theoretical and Experimental Study of Detonations in Connection w!th Decoupling",
DASA 2505, Physics International Company, San Leandro, California, November 1969.
(Coauthor).
" Deep Based Sanguine System Survivability" (U) PIFR-327, Physics International Company, San Leandro, California, August 1971 (S-RD). (Coauthor).
" Estimated SPRINT II Ground Motions" PITR 350-4, Physic International Company, San Leandro, California, March 1972 (S-3). (Coauthor).
l
" Relative Effectiveness of Energy Corservation Measures Taken in the Pacific Northwest", Aerospace Report No. ATR-74(8166)-1, January 1974.
" Emergency Core Coaling Systems for Light Water Reactors", EQL Report No. 9, California Institute of Technology, Environmental Quality Laboratory, May 1975.
Report to the American Physical Society by the Study Group on Light-Water Reactor Safety, Reviews of Modern Physics, Volume 47, Supplement No. 1, Summer 1975.
(Coauthor).
" Integrated Solar / Geothermal Power Systems Conceptual Design and Analysis",
ATR-75(7512)-1, July 1975. (Coauthor).
l
" Nuclear Reactor Safety: A View from the Outside", Bulletin of the Atomic Scientists, September 1975, pp. 20-25.
" Review of the NRC/ERDA Loss-of-Fluid Test Facility",19 November 1975, pp.67-108 of Enclosure A to This Country's Most Expensive Light Water Reactor Safety Facility, GAO document RED-76-68 A, May 26,1976.
i
" Effectiveness of Safeguards Program for the LWR Plutonium Recycling Industry",
ATR-76(6879)-1, April 1976. (Coauthor).
" Poor Management'of a Nuclear Light Water Reactor Safety Project", GAO document EMD-76-4, 25 August 1976. (Coauthor). Documentation of review of ERDA/NRC Plenum Fill Experiment Program.. _ _.
Q
" Transportation Risks for New and Spent Fuels and Radioactive Wastes with Respect to Road Accident Hazards and Purposeful Diversion", Direct Testimony, SDG&E Sundesert NOI Hearings,30 November 1976.
" Technical Brief -Issues of Record Related to Plans for Protection Against Sabotage end Diversion of. Nuclear Materials for the Sundesert Nuclear Power Plant", SDG&E Sundesert NOI Proceedings,29 December 1976.
" Technical Brief - Issues of Record Related to Transportation Risks for New and Spent Reactor Fuels and Radioactive Waste", SDG&E Sundesert NOI Proceedings, 30 December 1976.
" Control Room Human Engineering Influences on Operator Performance", Proceedings of Toolcal Meeting on Thermal Reactor Safety. CONF-770708, Sun Valley, Idaho, 31 July - 4 August 1977.
" Systems Management Support for ERCDC Study of Undergrounding and Berm Containment: Interim Report, Preliminary Program Assessment and Follow-on Program Development", ATR-77(7652-01)-1, August 1977. (Coauthor).
Review and Critique of Draf t " Report to the U.S. Congress on NRC's Plans for Research 4
Directed Toward the Improvement of Light-Water Nuclear Power Plant Safety", Letter l
report, 22 February 1978.
" Evaluation of the Feasibility, Economic Impact, and Effectiveness.of Underground l
Nuclear Power Plants - Final Technical Report", ATR-78(7652-14. 1, May 1978.
(Coauthor).
" Underground Siting of Nuclear Power Reactors - An ' Analysis of the California Energy Commission Study", Transactions of the American Nuclear Society, Vol. 32, 1979 Annual Meeting, Atlanta, GA., June 3-7,1979, pp. 553, 554. (Coauthor).
" Human Engineering Influences on the Performance of Nuclear Power Plant Operators",
l Testimony for the record of the May 22-24 1979 Hearings on Nuclear Power Plant Safety Systems, Committee on Science and Technology, U.S. House of Representatives.
U.S. Government Printing Office,1979, pp. 255-270.
I
" Residential Photovoltaic Systems - A Review and Comparative Evaluation of Four Independent Studies of Potential Concepts", ATR-80(7823)-1, December 1979 (also published as SAND 80-7010, Sandia National Laboratories, October 1980).
I
" Review of Rogovin Special Investigative Group Staff Report on Human Factors Evaluc. tion Related to the Three Mile Island Accident", Letter Report, 30 November 1979.
" Industrial Process Heat Applications of Solar and Geothermal Energy and Human Engineering Influences on the Performance of Nuclear Power Plants", ATR-79(9538)-1, September 1979.
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" Emergency Planning Zones for Serious Nuclear Power Plant Accidents", State of California - Office of Emergency Services, November 1980. (Coauthor).
"The Technical Basis for Emergency Planning Zones for Serious Accidents at Nuclear Power Plants in California", ATR-81(7870)-1, November 1980.
" Report of the LOFT Special Review Group", U.S. Nuclear Regulatory Commission, NUREG-0758, February 1981. (Coauthor).
"The Feasibility and Effectiveness of Underground Nuclear Power Plants - a Review of the California Energy Commission's Study", pp.19-33, Proceedings of the Symposium on Underground Siting of Nuclear Power Plants, Hanover, West Germany (16-20 March 1981), E. Schweizerbart' sche Verlagsbuchhandlung (Nagele u. Obermiller) Stuttgart, 1982.
" Development of Emergency Planning Zones for Nuclear Accidents in California",
American Nuclear Society Transactions,1981 Annual Meeting, Miami, Florida, June 7-11, 1982, TANSAO 381-776 (1981), June 1981, pp.124-126.
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ATTACHMENT 2 I
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ATTACHMENT 2 Tablo 1 Typiccl Flocion Product Inv0ntory for a BWR of Shorcham Sizo NU MB ER N A ME GRCUP PAREAT INITIAL (CURIES)
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SR-90 4.160E+06 2.670E+00 12 Y-51 8
SR-91 8.768E+07 5.880E+01 1 117E+08
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1.171E+08 7.000E-01 14 ZR-97 8
15 NE-95 8
ZR-95 1.055E+08 3.510E+01 16 MO-99
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3 TE-131H 6.553E+07 8 040E+00 31 I-132 3
TE-132 9 6 4 5E.+0 7._ _ _ __.. _ _._ 9. 5 21 E- 0 2 32 I-133 3
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1.513E+08 3.653E-02 34 I-135 3
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I-133 1.381E+08 5.290E+00 36 XE-135 1
I-135 2.850E+07 3 82iE-01 37 CS-134 4
9.458E+06 7.524E+02 38 CS-136 4
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4.903E+06 1.099E+04 40 8A-140 6
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1.388E+09 2.350E+00 48 PU-238 8
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NP-239 1 936E+04 8 912E+06 50 PU-240 8
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F UNITED STATES OF AMERICA j
7 NUCLEAR REGULATORY COMMISSION 1
1 N5yTED BEFORE THE-ATOMIC SAFETY AND LICENSING BOARD 82 0T 22 P4 3l In the Matter of
)
LONG ISLAND LICHTING COMPANY
)
.- M SECRtTAky 50-322'(0.MgCf
)
Docket No.
(Shoreham Nuclear Power Station,
)
(Emergency Planning Unit 1)
)
Proceedings)
)
CERTIFICATE OF SERVICE I hereby certify that copies of the following items were sent on October 12, 1982 by first class mail, except as other-wise noted, to the persons listed below:
1.
Direct Testimony of Andrew C.
Kanen, Dr. James H.
Johnson, Dr. Stephen Cole and Dr. Kai T.
Erikson On Behalf of Suffolk County Regarding Contentions EP 2B and EP SB (Traffic Congestion Issues).
2.
Direct Testimony of Dr. Kai T.
Erikson and Dr.
Stephen Cole On Behalf Of Suffolk County Regarding Contention 5A (Role Conflict).
3.
Direct Testimony of Gregory C. Minor On Behalf Of Suffolk County Regarding Contentions EP 10B and 10C (Radiation Monitoring) and EP 14 (Dose Assessment).
4.
Direct Testimony of Dr. Fred C. Finlayson On Behalf of Suffolk County Regarding Contention EP 14 (Accident Assessment and Dose Assessment Models).
Lawrence Brenner, Esq.*
Ralph Shapiro, Esq.
Administrative Judge Cammer and Shapiro l
Atomic Safety and Licensing Board 9 East 40th Street U.S.
Nuclear Regulatory Commission New York, New York 10016 Washington, D.C.
20555 Howard L.
Blau, Esq.
Dr. James L.
Carpenter
- 217 Newbridge Road Administrative Judge Hicksville, New York 11801 Atomic Safety and Licensing Board U.S.
Nuclear Regulatory Commission W.
Taylor Reveley III, Esq.*
Washington, D.C.
20555 Hunton & Williams P.O.
Box 1535 Mr. Peter A.
Morris
- 707 East Main St.
Administrative Judge Richmond, Virginia 23212 Atomic Safety and Licensing Board U.S.
Nuclear Regulatory Commission Washington, D.C.
20555
- By Hand
Edward M.
B2rrott, E0q.
Mr. J2y Dunkicb0rgsr
~ G2ncrol Councol N;w York Stato Encrgy Offico Long Iolend Lighting Company Agsncy Building 2 250 Old Country Road Empire State Plaza Mineola, New York 11501 Albany, New Ycrk 12223 Mr. Brian McCaffrey Stephen B.
Lathtm, Esq.
l Long Island Lighting Company Twomey, Latham & Shea 175 East Old Country Road P.O.
Box 398 Hicksville, New York 11801
.33 West Second Street i
Riverhead, New York 11901 i
Marc W. Goldsmith Mr. Jeff Smith Energy Research Group, Inc.
Shoreham Nuclear Power Station 400-1 Totten Pond Road P.O.
Box 618 Waltham, Massachusetts 02154 North Country Road Wading River, New York 11792 Joel Blau, Esq.
MHB Technical Associates New York Public Service Commission 1723 Hamilton urenue The Governor Nelson A.
Rockefeller Suite K Building San Jose, California 95125 Empire State Plaza Albany, New York 12223 Hon. Peter Cohalan Suffolk County Executive David H.
Gilmartin, Esq.
County Executive / Legislative Suffolk County Attorney Building County Executive / Legislative Bldg.
Veterans Memorial Highway Veterans Memorial Highway Hauppauge, New York 11788 Hauppauge, New York 11788 Ezra I.
Bialik, Esq.
Atomic Safety and Licensing Assistant Attorney General Board Panel Environmental Protection Bureau U.S.
Nuclear Regulatory Commission New York State Department of Washington, D.C.
20555 Law 2 World Trade Center Docketing and Service Section New York, New York 10047 Office of the Secretary U.S.
Nuclear Regulatory Commission Atomic Safety and Licensing Washington, D.C.
20555 Appeal Board U.S.
Nuclear Regulatory Bernard M.
Bordenick, Esq.*
Commission David A.
Repka, Esq.
Washington, D.C.
20555 U.S.
Nuclear Regulatory Commission Washington, D.C.
20555 Matthew J.
Kelly, Esq.
Staff Counsel, New York Stuart Diamond State Public Service Comm.
Environment / Energy Writer 3 Rockefeller Plaza NEWSDAY Albany, New York 12223 Long Island, New York 11747
- By Hand
..Charif Ssdky, E2q.
Daniel F.
Brown, EIq.
Kirkpatrick, Lockhart, Atomic Safety cnd Licsnsing Board Johnson & Hutchison U.S. Nuclear Regulatory Commissio 1500 Oliver Building Washington; D.C.
20555 Pittsburgh; Pennsylvania 15222 o
k 4
ristopher M.'McMurray KIRKPATRICK;. LCKHART, HILL,"
CHRISTOPHER & PHILLIPS DATE: October 12, 1982
'1900 M Street; N.W.,
8th Floor Washington; D.C.
20036 O
l