ML20069G423
| ML20069G423 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 03/22/1983 |
| From: | Dennis Bley, Paddleford D, Potter T, Richardson D CONSOLIDATED EDISON CO. OF NEW YORK, INC., POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20069G411 | List: |
| References | |
| RTR-NUREG-0880, RTR-NUREG-880 ISSUANCES-SP, NUDOCS 8303250149 | |
| Download: ML20069G423 (43) | |
Text
._.
e flf.((ED UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD J,0 Before Administrative Judges:
James P.
Gleason, Chairman Frederick J.
Shon Dr. Oscar H.
Paris
)
In the Matter of
)
)
CONSOLIDATED EDISON COMPANY OF
)
Docket Nos.
NEW YORK, INC.
)
50-247 SP (Indian Point, Unit No. 2)
)
50-286 SP
)
POWER AUTHORITY OF THE STATE OF
)
)
March 22, 1983 (Indian Point, Unit No. 3)
)
)
LICENSEES' TESTIMONY OF DENNIS C.
BLEY, DONALD F.
PADDLEFORD, THOMAS E.
POTTER, AND DENNIS C.
RICHARDSON ON COMMISSION QUESTION FIVE i
ATTORNEYS FILING THIS DOCUMENT:
1 Brent L.
Brandenburg Charles Morgan, Jr.
Paul F.
Cola rulli Joseph J.
Levin, Jr.
CONSOLIDATED EDISON COMPANY OF NEW YORK, INC.
MORGAN ASSOCIATES, CHARTERED 4 Irving Place 1899 L Street, N.W.
New York, New York 10003 Washington, D.C.
20036 (212) 460-4600 (202) 466-7000 8303250149 830322 PDR ADOCK 05000247 T
pop
o TABLE OF CONTENTS i
Page I.
PRESENTATION AND QUALIFICATIONS OF PANEL MEMBERS............................................
1 II.
INTRODUCTION.......................................
2 III. COMPARISON WITH THE NUCLEAR REGULATORY COMMISSION ' S PRELIMINARY SAFETY GOALS.............. 4 A.
Individual Risk...............................
5 B.
Societal Risk.................................
6 C.
Co re Melt Fr equen cy........................... 6 IV.
PRA COMPARISONS....................................
9 A.
Comparison of Risks Among Nuclear Power Plants........................................
9 1.
Individual Risk...........................lO 2.
Societal Risk............................
14 B.
Comparison With The Commission's Task Force Results......................................
18 V.
SPECIAL DESIGN FEATURES AT INDIAN POINT...........
20 VI.
CONCLUSIONS....................................... 24 i
a I.
PRESENTATION AND QUALIFICATIONS OF PANEL MEMBERS My name is Dennis C. Bley, Ph.D.
I am a consultant at Pickard, Lowe and Garrick, Inc., in reliability, risk, and decision analysis for electrical generating plants.
I was a principal investigator on the Indian Point Probabilistic Safety Study.
A statement of my professional qualifications is attached.
My name is Donald F. Paddleford.
I am an Advisory Engineer in the Risk Assessnent Section of the Nuclear Safety Department of the Nuclear Technology Division of Westinghouse Electric Corporation.
I was a principal investigator on the Indian Point Probabilistic Safety Study.
A statement of my professional qualifications is a ttached.
My name is Thomas E.
Potter.
I am a consultant at Pickard, Lowe and Garrick, Inc., in public health conse-quence analysis of radioactive releases.
I was a principal investigator on the Indian Point Probabilistic Safety Study.
A statement of my professional qualifications is attached.
My name is Dennis C.
Richardson.
I am the Risk Assess-ment Technology Manager in the Nuclear Safety Department of the Nuclear Technology Division of Westinghouse Electric Co rpo ration.
I was a principal investigator on the' Indian Point Probabilistic Safety Study.
A statement of my profes-sional qualifications is attached.
b
. II.
INTRODUCTION A central issue in this hearing is whether the Indian Point nuclear power plants produce risks that significantly exceed the range of risks posed by other nuclear power plants in light of the demographic characteristics of the area surrounding the Indian Point site.
This issue is l
articulated in Question 5 of the Commission's Memorandum and Order of January 8, 1981, which asked:
Based on the foregoing, how do the risks posed by Indian Point Units 2 and 3 com-pare with the range of risks posed by other nuclear power plants licensed to operate by the Commission?
(The Board should limit its inquiry to generic examination of the range of risks and not go into any site specific examina-tion other than for Indian Point itself, except to the extent raised by the Task i
j Force.)
I Risk can be measured by several health and economic indices and from both an individual and a societal stand-point.
Population distribution and plant characteristics i
affect these indices differently.
In selecting which indices are most important, guidance is taken from the i
Nuclear Regulatory Comndssion's (Comnission's) preliminary safety goals, which emphasize early and latent fatality risks (Reference 1).
Similarly, the emphasis here is on the effects of population distribution on early and latent fatality risk.
Three different approaches to addressing Commission Question 5 are taken in this testimony.
First, a comparison p*-
s h
4
.s..
4 1
. is made of the risks from the Indian Point plants to the Commission's preliminary safety goals.
Second,*the risks from the Indian Point plants, as analyzed in the Indian Point Probabilistic Safety Study (IPPSS) (Reference 2), are corpared to the results of site and plant specific proba-bilistic risk assessments (PRAs) of a number of other nuclear power plants.
Third, there is a discussion of the benefits resulting from the special design features at Indian Point which are not present at all nuclear power plants.
Individually and collectively, each of these comparisons supports the conclusion that the Indian Point nuclear power plants are in.the range of risks posed by other nuclear power plants.
Spec.ifically, (1) the risk of latent fatalities at Indian Point is low and information available suggests that latent fatality risks may not vary greatly among nuclear power plants; (2) the absolute risk of early fatalities is even lower than the latent fatality t
risk, thereby reducing the significance of plant-to-plant variability; (3) for both risk indices, the Indian Point plants meet the Commission's preliminary safety goals; and (4) anticipated reductions in source term estimates would reduce both early and latent fatality risk and, in fact, could effectively eliminate the early fatality risk.
See Licensees' Testimony of William R.
Stratton, Walton A.
i Rodger, and Thomas E.
Potter on Question One (Jan. 24, I
t i
. 1983).
In addition, plant features present at Indian Point but not included at other plants are among the important i
factors supporting the conclusion that the Indian Point nuclear power plants are within the range of risks posed by other nuclear power plants.
III.
COMPARISON WITH THE NUCLEAR REGULATORY COMMISSION'S PRELIMINARY SAFETY GOALS On March 14, 1983, the Commission published a Policy Statement on Safety Goals for the Operation of Nuclear Power Plants.
48 Fed. Reg. 10,772 (1983).
The preliminary safety goals and design objectives apply to both individual and societal risk.
Subordinate to these goals is a design objective for risk to the plant, core melt frequency.
The preliminary safety goals represent a national benchmark against which all nuclear power plants can be com-pared.
Therefore, the comparison of the risks from the i
Indian Point plants to the Commission's preliminary safety goals is one method of determining if these plants are within the range of risks posed by other nuclear power plants.
Both Indian Point Units 2 and 3 are among those plants which have health risks smaller than those adopted by the Commission's preliminary safety goals.
Uncertainties in the calculated health risks for Indian i
Point are offset by the large margins between these i
calculated risks and the preliminary safety goals.
Reduced source terms will result in even larger margins.
A.
Individual Risk The Commission's preliminary safety goals state that the early fatality risk to an average individual in the 1
vicinity of a nuclear power plant should not exceed one-tenth of one percent of the sum of early fatality risk to that individual from other accidents.
48 Fed. Reg. 10,774.
To translate this goal into numerical form, we use the United States national average accident risk of 5 fatal accidents per 10,000 people per year (5 x 10-4 per year)
(Reference 2).
For the purpose of assessing the individual risk, the commission defines " vicinity" of the plant as a 1-mile radius.
48 Fed. Reg. 10,774.
Using this definition of vicinity and IPPSS emergency response assumptions, the average individual early fatality risk has been calculated 1.
According to the Commission, the average individual in the vicinity of the plant is defined as the average individual biologically (in terms of age and other risk factors) and locationally who resides within a mile from the plant site boundary.
This means that the average individual is found by accumu-i lating the estimated individual risks and dividing by the number of indiv-iduals residing in the vicinity of the plant.
48 Fed. Reg. 10,77 4.
1 I
l'
. as a fraction of the national average accident risk.
This is then compared with the Commission's preliminary safety goal in Table 1.
The risk of Indian Point is well within this goal, by a factor of approximately 70 for Indian Point Unit 2 and a factor of approximately 75 for Indian Point Unit 3.
B.
Societal Risk For societal risk, the Commission's preliminary goal is that the latent cancer fatality risk to the population in the vicinity of a nuclear power plant should be less than one-tenth of one percent of the cancer fatality risks from other causes.
48 Fed. Reg. 10,774.
For latent fatalities, vicinity is defined as 50 miles.
Idl.
The national average cancer risk for a person in the United States is two deaths per 1,000 people per year (2 x 10-3 per year) (Reference 2).
For this radius from the Indian Point plants, the aver-age latent cancer fatality risk has been calculated as a fraction of the national cancer fatality risk and is com-pared with the Commission's prelimina ry goal in Table 1.
The risk of Indian Point is well within this goal, by a factor of approximately 165 for Indian Point Unit 2 and 710 for Indian Point Unit 3.
C.
Core Melt Frequency Table 1 also shows the comparison of the Indian Point Units 2 and 3 median core melt frequencies with the Commission's preliminary st?ety goal.
Because the Zion PRA l
i
. TABLE 1 COMPARISON OF RISKS FROM INDIAN POINT PLANTS WITH NRC SAFETY GOALS Indian Point 2 Indian Point 3 NRC Goal Average Early Fatality Risk Within 1 Mile as a Fraction of Other 1.4 x. 10-5 1.3 x 10-5 1 x 10-3 Accident Fatality Risk Within 1 Mile Average Latent Cancer Fatality Risk Within 50 Miles as a Fraction of 6.0 x 10-6 1.4 x 10-6 1 x 10-3 Other Cancer Fatality Risks Within 50 Miles Core Melt Frequency (per year of reactor operation) 1.4 x 10-4, 5.0 x 10-5, 1 x 10-4 internal plus external Core Melt Frequency (per year of reactor operation) 5.0 x 10-5, 3.0 x 10-5, no explicit internal only goal stated
(
l
- Median Frequency l
I
. (Reference 3) and the IPPSS are the only risk assessments of which we are aware that give comprehensive treatment to
~
external events, Table 1 also includes the median core melt frequency of internal initiating events only.
Considering internal initiating events only, both Indian Point plants meet the Commission's preliminary safety goal.
Although information on core melt frequency is provided here for completeness in comparing the Indian Point plants against the preliminary goals, the s of this parameter are not of particular use in addressing Commission Question 5.
This is because core melt frequency is a poor indicator of public risk, as discussed in Licensees' Testimony on Commission Question One, Board Question 1.1, and Contention 1.1 (Jan. 24, 1983).
This can be shown in two ways.
- First, approximately 65 percent of the postulated core melt scenarios at Indian Point Unit 2 and almost 95 percent of those at Indian Point Unit 3 do not lead to significant releases of radioactive material to the environment.
Second, approximately 95 percent of the calculated early t
fatality risk at each plant is due to the interfacing systems LOCA, which contributes less than one-half of one percent to the core melt frequency.
On the other hand, core melt frequency is a useful indicator of economic risk to the customers and owners of Indian Point Units 2 and 3, as it is a measure of the likelihood of losing the benefits of these plants.
I
. 4 IV.
PRA COMPARISONS Another way to compare the risks posed by the Indian Point plants with those posed by other nuclear power plants is to compare site and plant specific PRAs for varioue i
plants, all identical in scope and using state-of-the-art methodology.
At the present time, however, such a compar-ison cannot be made due to the limited number of available, comparable studies.
The following comparisons, however, are possible:
1.
A comparison of the IPPSS risk results with those of other plants for which L
reasonably complete PRAs have been pub-j lished.
Only the Indian Point and Zion PRAs include external events; therefore, only comparisons on an intgrnal initi-ator basis have been made.
l 2.
A comparison of the IPPSS results with the range of risks for nuclear plants calculated by the Commission Task Force Report on the Interim Operation of Indian Point (Reference 4).
A.
Comparison Of Risks Among Nuclear Power Plants In connection with the comparison of risks among nuclear power plants, it is important to note that PRA methodology has been evolving rapidly over the last 10 years.
The various published studies, therefore, differ considerably in certain recpects.
Thus, when comparing the l
1 1.
While the Big Rock Point PRA did consider fires, it did not evaluate other external initiating events.
l
l W results of IPPSS with those of other PRA studies, it must be recognized that such comparisons are not only of different plants, but are also of different data bases and, in some i
cases, of different methodologies.
These studies vary in scope and sophistication.
Some did not include external events and/or public health effects, while others focused only on a few systems or on one type of accident initi-a to r.
With these reservations in mind, quantitative com-parisons can be made.
1.
Individual Risk Table 2, which was compiled by the Commission Staff (Reference 5), presents data from a number of plant specific l
PRAs on the frequency of core melt, the frequency of a major release, and the early and latent fatality risks to an indi-i vidual living within one mile of plant boundaries.
The i
values in the "Early fatality" column can be directly com-
[
pared with the Commission's preliminary safety goal for this health index (5 x 10-7).
This table generally reflects the range of risks from internal initiating events at United States nuclear power plants because it includes a represent-ative sampling of PWRs and BWR, high and low population density sites, power levels from 71 to 1250 MWe, and prin-cipal reactor vendors and architect engineers.
Although this table has been reproduced verbatim from Reference 5, additional information is also presented for Indian Point Units 2 and 3, and appears in a box directly below the l
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merin SAft/ POWER F C0tt
~ F MAJOR
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!stP lab /8137tL 81 835 5 a 10*I 2 a 10*I 6 s 10*#
2 a 10*#
all.Pwt 2 M cistis 8 CEttAn til y4G (4 78 1300 4 a 10-5 1 a 10*'
3 a 10**
J a 10*8 Coataamment stroaver end larger taas U.S.
nig ascs l#
W000-LEAftR/ G1/8137th 81 71 1 a 10*I O
O La= pe=er }evel, runete SAI sittag trowns ferry latt Gt/TVA 81 1047 F
10~4 4 a 10 2
10*I 1 a 10*'
ATW5 ans laterseees-
- 4 6
IBwn 4, N Il coacy in rec.etant Pd trains core mit pir. ate Calvert C1tffs RSSMAP C1/8tCMTIL 82 250 F e IS*I 1 s 10*I 9 e 10*0 2 s 10*I More consrenensive lREP*
steer ta progress.
AFV5 rosesign will tenee rest and cag
- melt fregeencs Crystal River IRt?
84W/G!LBERT 80 825 a
10**
2 a 10**
3 e 10-4
? e 10*'
P core seit res.casti' factor of 3 ty auro changes 1proca-Crane G.if 455 MAP GE/8tCMTEL 81 1250 4 a 10*I 4 a 10*I 1 a 10*I 1
10*I Contaiment al ays fails (swa 6. n !!!)
directly to atmos.
PAere, eoes Act assent staf f's analysis 6f ATW5 risa 1.F. s2 $#
PLG W/Ut&C 82 873 a e 10**
3 a 10**
3 a 10*8 1
10*8 lac 1 ses,taggynt eventsd =8 PLA W/veAC 43 073 L a 10" a s 10**
7 a IO*
6 a 60*
ta61eams esserno!
eventa yts wiutac s3 873 5 a 10 '
6 a 10*8 6 a to 3 a 10 tesernet evease enty e.
nie to e seen wa. rie aestas vet.e 14 he -
- towee.
M PG W/Ut&C 82 96 5 g a le 3 a 10-5 I s 10*'
3 a 10*l0 g,g %,
,g l.8 e3 events -
PLM w/utac 03 965
$ a 40*I
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- a a le 3 a 10 Lastases esterna&
event e rLC W/ut&C 83 945 3 a 10*I 4 a 10
- 4 a 16' 3 a 10*
taternet evesse only e.
nse se e seen value. The amatan es&me imm14 he :
- lowes.
L p=rica M 5Al CE/SECHTEL J1 1055 2 a 10*8 3 a 10*'
I i 10-8 I a 10**
rama ulve, ets6tes AT.:
(Su4 e. M !!)
fta stil a s tone let?
CE/tSA5CO 82 652 3 = 10**
I a 10*4 I a 10*I 6 a 10*I Hajor release is na a lease Category 3 e
ot6aes R55ftAP B&W/813TEL 80 860 8 a 10*I 4 a 10*I 2 a 10*I 1 a 10*I 1/4-Put 2; 3/4-#wt 3 Peach 80ttes lJA561400 Gt/IECHTEL 75 1045 3 a 10 5 7 a 10*'
4 a 10*8 3 a 10*8 Staff's analysis of Aib5 (Swa 4. u !)
woule likely resvit in rist escatetet safety gNI tequeya n R55 MAP W-IC/TVA 78 1148 4 a 10 5 4 a 10-5 1
10*3 5 a 10*#
'I M2 *y** to3' ' '"'
b 2 Svery WASM-lac 0 W/5&W 75 775 6 105 1 a 10*I 2 a 10*#
1 e 10*#
2/3-Pua 2: 1/3-PWR 3 lion M 8LG W/5&L 81 1100 4 a 10*I 4 a 10*'
2 a 10*8 1 a 10*O Incluses esternal enats M All aussers are mediaa velves or potat estimates from taternal tattiators unless otherwise spectfisc.
M freo ency of core melt I a 10~4 ts the Safety Goal value for Accident Probaatitty Campartson.
M 8reovency of release witn Potential for Early Fatalities Assuming nominal Evacuatton and Waratag flees (a55).
F 5 a 10*I ts the safety Goal for Early fatality tint canoarison. Same assumption 1 as 3 above valess spectf ted.
I M 2 a 10-4 is the Safety Goal for Cancer Fatality Rist Camearissa. Same assumptions as 3 above salass specif tee.
F Utility-serformes PtAs. All values are rough estimates based vaca tattial interpretattoa af resvits.
M otisisttc energency response assumettons (1-nour celay with at least 8-nour earn 6ag) for destaaet sequence unen seterstaing teetvidual rest.
j C
l M **+etc. tee ess,a is c,ominates by sma.ll LOCAs aae transients.
Sovce tere reenstion esp to rea p
.ca.y oe a. paraitei v.ives at tne aiun.e,,cted.f ine.uce resistes rist to itate guiseltaes.,
i t.en..
430 reie.se ce.ie e e
ora t
.a ter st rege los.,., i.or.,ta l
ur. po.ee reeunaaacy, F to po=ce level (71 Mu l results to low tadtstaval rist. Estenstve aestga emelffcations necessary to reevce core seit freovency, e
$ see ctica af core melt freovency wouls reoutre reeestga of the resteval heat removal systen to ellatnate cammonalities between tratas weien reevce the sigat ticance of multiple reovneancy.
2 FuS recestga is espected te stgnificantly recuce core melt frecuency and indiviaual rist.
I A
1 REP study incleatng isoreved AFv5 sesiga I
well te availante ta Spring 1983. Modtftcation to DC power system and engineered safety systee actuation system any na reevires to lower core cent freovency =ttata gutseltaas. Predicted rtst is dominated by transient event and shouls be significantly reo.as by new source tern esta.
d ore.eelt freovency could be reseced to less tnan guiceltnes levels by taproving written proceaures and imoroving the reltaatitty of the C
stese smonly to the IPJ5 turtsae artven puso. Preatsted rest is dominated by small LOCA events. #ew source term informattoa is espected to res it in a moderate resection in predictes rist.
IM ore melt freovency is seminated by solemic considerations.
C 3 ore melt freeuency could he reo ced to below guideline levels by recesigning the energency AC comer system to reeuce aeoeneency en tne g C
tertine and improving procedures for ressonatag to transtants. Preetcted rist is contNtao by traastant events. raev source tern saformation i
s*=,la resvit in a significaat reovction.
I
~ _ _
. 1 Indian Point results presented by the Staff.
This additional information is drawn from the risk calculations j
in Licensees' Testimony on Commission Question One, Board Question 1.1, and Contention 1.1.
It includes risk results i
from internal initiating events only to avoid an erroneous comparison of Indian Point internal plus external results with internal only results from other plants.
It also a
includes the internal plus external results for Indian i
Point.
Based on the results in this table, the risk to an l
individual living within 1 mile of Indian Point compares favorably with the estimated risk to individuals living within 1 mile of other nuclear power plants.
Additionally, the Indian Point core melt frequency is within the range of other estbnates presented in the table, and the frequency of a major release compares favorably with the estimates for the other plants in the table.
Another valuable comparison is the frequency of the I
interfacing systems LOCA, which is believed to be an important contributor to early fatality risk at PWRs and is the major initiating event contributing to the early fatality risk at Indian Point.
Estimates of the frequency of this event at Indian Point and several other nuclear f
power plants are presented in Table 3.
The differences in these estimated frequencies are due to a combination of design differences among plants, as well as to testing and 1
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. TABLE 3 COMPARISON OF INTERFACING SYSTDiS IXX'A MEDIAN FREQUENCIES Recurrence Interval Median (Nunber of Study PWR Plant Frequency Reactor Years)
Reference RSS Surry 4 x 10-6 250,000 6
IPPSS Indian Ibint 2 4 x 10-8 25,000,000 2
IPPSS Indian Point 3 4 x 10-8 25,000,000 2
ZPSS Zion 3 x 10-C 33,000,000 3
RS94AP Oconee 7 x 10-5 14,000 7
RSSMAP Sequoyah 5 x 10-6 200,000 7
IREP Crystal River-3 2 x 10-9 500,000,000 8
L i
i l
l
. maintenance procedures.
The IPPSS accounted for testing and maintenance, including some procedures which are not in effect at all other plants.
2.
Societal Risk Societal risk comparisons have been compiled for the PRAs listed in Table 4.
Graphical comparisons of results from these studies are presented in Figures 1 and 2.
Ma ny of the studies listed in Table 2 did not calculate risk curves and are, therefore, not included in Figures 1 and 2.
The results from the German Biblis B risk study are included in these figures, as in the Staff table, because the study is recent, reasonably comprehensive, and analyzes a high population site.
Because so few PRA studies have comprehensively exam-ined external initiating events as does IPPSS, the compari-sons in these figures are for internal initiating events only.
(The risk curves presented in the licensees' Question 1 testimony included both internal and external events. )
Figure 1 shows the median risk curves for early fatalities as presented in the various studies, and Figure 2 presents
(
similar results for latent cancer fatalities.
These figures support the conclusion that Indian Point is within the range of societal risks posed by other nuclear power plants.
I I
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s TABLE 4 3
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s PLANTS-USED IN THE GRAPHICAL COMPARISONS s..
4x Plant k
Reference K.
Indian Point 2 2
Indian Point 3 2
I Surry
..- 6 Peach Bot. tom -
-6 s
s Zion 3
N s.
- W Biblis B 9
x s
Big Rock Point 10
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O LIMERICK (D)#
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N 10'O INDIAN POINT 2 OR 3 (E)
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Z ZION (F)
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Q DATA SOURCES E
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CURVE REFERENCE FIGURE 10'II -
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C 6
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11 5-1 k
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F 3
8.6 Sa g
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10-12 I
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2 3
4 5
10 10 10 10 10 10 EARLY FATALITIES Figure 1.
Comparison of PRA Median Risk Curves for Early Fatalities (Internal Risk Only)
. 10-3 BIG ROCK POINT (G) \\
GERMAN RISK STUDY (B)
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10-4 l
PEACH BOTTOM (C)
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e 10-5 O
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- ALSO TABLE 5.6 3g-10 I
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10 10 10 10 10 10 LATENT CANCER FATALITIES Figure 2.
Comparison of PRA Median Risk Curves for Latent Fatalities (Internal Risk Only)
B.
Comparison with the Commission's Task Force Results In 1980, a Commission task force studied the effects on -
risk of:
(1) a typical pressurized water reactor (Reactor Safety Study, Surry) at different sites; (2) different plants at the same site (Indian Point); and (3) different public protection measures (Reference 4).
Because the results of these studies are an indication of the range of risks posed by nuclear plants in general, they are also used for the comparison requested in Commission Question 5.
For this purpose, the median internal risk curves from e
the IPPSS are presented in Figure 3 along with results from Figure 11 of the Task Force Study for early and latent fatality risk.
These curves support the view that the risk from Indian Point is within the range of risks from other nuclear power plants.
As can be seen from Figure 3, the early fatality risk curve calculated in the IPPSS lies more than an order of magnitude below the range of results presented in the Task Force Study.
A large part of this difference results from the Task Force Study's failure to evaluate the strength of the containment, which precludes prompt containment failure.
Thus, the Task Force did not include a release category for late containment failure.
The IPPSS latent fatality risk curve lies within the range of the latent fatality risk calculated by the Task Force for the Indian Point site, and is below the range calculated for the Surry reactor at various sites.
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NOTE 1.
THE R ANCES REPRtStNT t[5T ESTIMATES ON A COMPARATIVE SA$ls. THERE ARE LARGE UNCE RT AINTIES WITH THE ABSOLUTE VALUt3 PRESENTED IN THIS FICURE.
2.
PUBL'C PROTE CTIVE ME ASURES MAO N0 5tGNIFICANT IMPACT ON TOT AL LATENT CANCER
~~--"l ESTiu ATED R ANGE OF CON 5(OVENCt5 FOR V ARIOU5 0tSIGN5 CON $tD(R E D AT INDI AN POtNT SITE.*
[
~----
l l
ESTsuATED R ANCE OF'CONSEQUINCE5 FOR $lX 5BTES CON 510ER(D l
wsTH SURRY DEstCN.*
h[i3
- - --'- l 157tuaTED R ANCE OF CON $toutNCES FOR VARICUS PUSLIC PROTECTlvt Mt ASURES CON 51 CERED AT INDIAN POINT Slit.'
INDIAN POINT 2 OR 3 -INTERNAL ONLY.*
- l l
l
' REFERENCE 1.
l
- REFERENCE 2.
l Figure 3.
Ranges of Risk Variation l
l
V.
SPECIAL DESIGN FEATURES AT INDIAN POINT The comparisons presented in the previous sections indicate that the Indian Point Units are within the range of risks of other nuclear power plants, despite the demographic characteristics of the area surrounding the Indian Point plants.
It is thus appropriate to ask whether this conclusion is supported by information about the engineering and design features of the plants.
Nuclear power plants located at high population sites have received special attention from regulatory agencies.
During the licensing review for the Indian Point Units, additional features were incorporated into the plant designs to supplement the standard safety features.
These features were highlighted in the Director's Order of February 11, 1980.
Among the features that could lead to lower frequencies of major releases from the Indian Point containments than from some other containments ares (1)
The design and construction of these containments, with a pressure limit of 6 141 psia and a large volume of 2.6 x 10 l
cu.
ft.,
gives them the capability to withstand internal pressures well in excess of the design pressure of 62 psia.
Additionally, the containments can withstand without significant structural damage all credible seismic l
events that could occur in this area.
The containment building configuration allows gases to circulate and mix easily t'
to prevent local accumulation of hydro-gen.
This configuration also provides for more effective containment heat l
h
. removal capability.
In addition, the geometry of the reactor cavity promotes dispersion of the core debris, thereby increasing its coolability.
Also, the geometry of the containment floor pro-vides for easy entry of water to the reactor cavity to cool the debris.
(2)
Containment cooling capability is pro-vided by diverse systems.
The design includes five fan cooling units in addi-tion to four purps capable of providing containment spray recirculation.
The availability of any one of the fans or sprays is sufficient to prevent containment overpressure failure.
Two recirculation pumps, located inside containmen t, are unique to Indian Point and are two of the pumps capable of providing containment spray.
(3)
The Indian Point containments have two sumps that provide for recirculation of emergency core cooling water.
The presence of two sumps is also unique to Indian Point.
(4)
The presence of the recirculation pumps inside containment provides the capability of recirculating emergency core cooling water without its leaving the containment building.
(5)
Three gas turbine-generators are avail-able for supplying power to either unit.
This feature is unique to Indian Point and provides an unusual degree of diversity in emergency power sources.
(6)
Confirmatory signals (S signals) are sent upon actuation of emergency saf e-guards to certain power operated isola-tion valves to ensure that, if a valve had been inadvertently placed in an incorrect position, it would be restored to its' correct position.
This feature reduces the likelihood of bypassing the containment.
'7)
The containment weld channel pressur-ization system and the isolation valve
. seal water system help to assure that the containment leaktightness is main-i tained.
(8)
The service water and component cooling water systems are arranged to maximize redundancy of active components.
Any one of six service water pumps can sup-ply any service water load.
Similarly, either of two component cooling water pumps can be connected to any component cooling water load.
The flexibility provided by these and similar intercon-nections within and between systems results in particularly low risk from internal initiating events at Indian Point.
The risk reductions afforded by some of the design fea-tures discussed above have been quantified using information from the IPPSS.
For example, the frequency of late over-pressure containment failure from internal initiating events is reduced by one to two orders of magnitude by the presence of fan coolers, which back up the spray recirculation system.
The gas turbines, an additional source of AC power recovery for the time period of one to three hours following a core melt, provide up to an order of magnitude reduction in the frequency of late overpressure containment failures from internal initiating events.
When external as well as internal initiating events are considered, the fan coolers provide up to an order of magnitude reduction and the gas turbines provide less than a factor of two reduction in the frequency of late overpressure containment failures.
While not specifically quantified, the other design features l
l i
. discussed above would certainly provide further risk i
reduction.
On the strength of these special design features and other specific Indian Point systems, less than 2 percent of the internally initiated core melts lead to containment failure.
As indicated in Table 2 and supported here, the frequency of a major release resulting from internal ini-tiating events is thought to be less at Indian Point than at a number of other nuclear power plants.
In addition, as stated above, the various safety features, particularly the fan coolers, provide significant reductions in overall (internal plus external) frequency of late containment
~ Pailure.
As discussed in Licensees' Testimony of Thomas E.
Potter on Commission Question Five ( Ma r. 2 2, 1983), the range of latent fatality risk among nuclear power plant sites, given a severe release, is relatively narrow.
Based on the information in Table 2, the strength of the Indian Point containments, and the special design features at the plants, the release frequency at Indian Point is lower than the estimated release frequencies at many other plants.
The narrow range of latent fatality risk, in conjunction with a lower than average release frequency, supports the conclu-sion that the Indian Point latent fatality risk is within the range of latent fatality risk of other nuclear power plants.
+
Based on the information in Table 2, the absolute value of the early fatality risk at a number of nuclear power plants is very low.
At Indian Point, this in largely due to the strength of the containments, which essentially precludes prompt containment failure.
The only accident contributing to early fatality risk is the interfacing
]
systems LOCA which, as shown in Table 4, has a very low frequency of occurrence.
Special design features, together with standard nuclear power plant safety systems, result in very low early and latent fatality risk at Indian Point Units 2 and 3.
4 VI.
CONCLUSIONS Each of the several comparisons used in this testimony to address Commission Question Five supports the conclusion that Indian Point Units 2 and 3 are within the range of risks posed by other nuclear power plants.
A comparison of the Indian Point plants to the Commis-sion's preliminary safety goals shows that these plants are within these goals.
As such, they are in the class of plants whose risks are in a range below the limits estab-lished by these goals.
Various comparisons of the results of other PRAs to the results of the IPPSS show that the Indian Point plants are within the range of risks estimated for other nuclear power I
plants.
Table 2 indicates that the early fatality risk for
. a number of nuclear power plants, including Indian Point, is very low.
The Indian Point early fatality risk is low because, based on the strength of the containments, the low f requency interfacing systems LOCA is the only contributor to early fatality risk at Indian Point.
Using the source terms proposed in the previously sub-mitted Question 1 testimony of Dr. William Stratton, Dr.
Walton Rodger, and Thomas Potter, no early fatalities would occur for any Indian Point accident scenario.
When absolute risks are very low, differences between these low numbers are relatively unimportant.
With regard to the latent fatality risk, the Indian 3-.
Point plants are close to the national average of the.mean values of latent fatality consequences, based on the generic work reported in NUREG/CR-2239 (Reference 12).
This report shows that the range of latent fatality consequences, given a specified release, is relatively narrow.
See Licensees' Testimony of Thomas E. Potter on Commission Question Five (Mar. 22, 1983).
Based on the strength of the Indian Point containments l
l and the special features of the Indian Point plan _ts,3 radio-active releases from these plants would be less frequent
?
than at many other plants.
See Table 2.
The narrow range I
of the consequences and the lower frequency of containment failure support the conclusion that the latent fatality risk i
-t,
is within the range of such risks posed by other nuclear power plants.
i The above conclusion on latent fatalities is also relevant to the issue of whether any mitigation devices are warranted for the Indian Point plants.
As discussed under Commission Questions 1 and 2, the principal application of these mitigation devices would be to reduce latent fatal-ities.
The Indian Point plants have latent fatality risks i
which meet the Commission's preliminary safety goals and are within the range calculated for other nuclear power plants.
This range itself will be lower and narrower with reductions in source terms.
There fore, no additional mitigation features are necessary to bring Indian Point within the range of risks posed by other nuclear power plants.
l
. REFERENCES 1.
" Policy Statement on Cafety Goals for the Operation of Nuclear Power Plants," 48 Fed. Reg.
10,773 (1983).
2.
Consolidated Edison Company of New York, Inc., and the Power Authority of the State of New York,
" Indian Point Probabilistic Safety Study" (Mar.
1982 ), and Amendment 1 (Jan. 1983).
3.
Commonwealth Edison Company, " Zion Probabilistic Safety Study" (Sept. 1981).
4.
NUREG-0715, " Report of the Task Force on Interim Operation of Indian Point, " SECY-80-283 (1980).
5.
Memorandum from William J.
Dircks to the Commission (Draft) (Jan. 5, 1983) (Attachment).
6.
WASH-1400 (NUREG/75-014), " Reactor Safety Study:
An Assessment of Accident Risks in U.S. Commercial Nuclear Pswer Plants" (1975).
7.
NUREG/CR-1659, " Reactor Safety Study Mefhodology Applications Program" (1981).
8.
NUREG/CR-2515, " Crystal River-3 Safety Study, Main Report," Vol. I (1981).
9.
EPRI-NP-1804-SR, " German Risk Study - Main Report.
A Study of the Risks Due to Accidents in Nuclear Power Plants" (1981) (English Translation).
10.
Consumers Power Company, "Probabilistic Risk Assessment, Main Report, Big Rock Point Plant" (1981).
11.
Philadelphia Electric Company, " Limerick Probabilistic Safety Study" (1981).
12.
NUREG/CR-2239, " Technical Guidance for Siting Criteria Development" (1982).
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D.F.
Paddleford Kansas State University, B.S. (1960) and M.S. (1962) in Nuclear Engineering Post Graduate Courses in Engineering at UCLA, SMU, MT Mr. Paddleford is an Advisory Engineer in the Risk Assessment Section of the Nuclear Safety Department at the Westinghouse Nuclear Technology Division.
Since joining Westinghouse PWR Systems Division in 1965, Mr. Paddleford has held a variety of positions in areas of increasing responsibility related to PWR plant safety, licensing, reliability, safety standards develop-ment, transient analysis, including core melt behavior, and probabilistic risk assessment.
Most recently he has been engaged in the management and analysis of degraded core related issues, including test programs.
He is currently active on AIF and IEEE Committees on Development of Risk Criteria and Utilization of PRA methods and is one of the principle authors of the Technical Writing Group which prepared the Industry /NRC "PRA Procedures Guides" under sponsorship of ANS and IEEE.
He is a member of the IDCOR Technical Advisory Group and several IDCOR Expert Review Groups.
In the early 1970's his experience and responsibility included lead on research projects to develop a probabilistic approach to safety analysis,, including systems reliability and data, probabilistic fracture mechanics models, core migration assess-ment, and probabilistic modeling of consequences associated with major fission product releases.
Additional pertinent experience has included development of methods for parameter uncertainty propagation through design analysis computer codes and analysis of TMI and alternative scenarios at the request of the Kemeny Commission.
Prior to joining Westinghouse, Mr. Paddleford was at l
Atomics International where he worked in areas of reactor physics and transient analysis in support of the SNAP 2/10 safety development.
Mr. Paddleford is a member of ANS and Sigma Xi and is a registered Professional Engineer.
He is author or co-author of a number of papers on reactor safety and risk assessment.
I i
a HAME THCMAS E PCITER EDUCATION M.S., Environmental Science, University of Michigan,1972.
3.S., Chemistry, University of Pittsburgn,1963.
PRGFESSICHAL EXPERIENCE General Summary Consultant on health and safety' aspects of nuclear power. Performia; environmental dose assessments for nuclear power plant safety analysis, environmental reports and operating reports. Assisting clients in design and imolementation of raciological or environmental mont oring programs and interpretation of results. Providing indepencent review of in-plant radiological protection programs and effluent analysis programs.
Consultant in radiological healtn aspects of nuclear power.
Prepared radiological health section of safety analysis reports and environmental monitoring programs and evaluated data from those programs. Developed a mathematical model to predict radiation doses from nuclear power plant e ffluents.
License acministrator, plutonium fuel facility health and safety supervisor. Proviced radiological safety review of major facility modi fications.
Usec these analyses and nuclear criticality analyses perfomed by others to prepare AEC special nuclear materials and byprocuct license applications. Served as corporate contact with AEC in matters related to licensing. Organi:ed and supervised a radiological protec; ion orogram for a plutonium fuels fabrication facility and not cell f acil ity. Instituted personnel monitoring programs using thermoluminescent dosimetry and breathing-:ene arecsci sampling in 1967.
Served as secretary of a plant safety committee wnich insoected all operations and reviewed detailed written procedures for operators.
Servec as memoer of a corporate safety ecmmittae wnica determined corporate policy regarding health and safety matters.
Chronological Summary 1973-Present Consul tant, Pickard, Lowe and Garrick, Inc.
1972-1973 Consultant to Dr. G. Hoy Whipple, University of Michigan.
1963-137C Nuclear Matarials and Equipment Corporation (NUMEC).
License acministrator, plutonium fuel facility health and safe y supervisor.
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,,* O POTTER - 2 MEMSERSHIPS American Chemical Society.
American Nuclear Society.
Health Physics Society.
Certified by American Board of Health Physics.
REPORTS AND PUBLICATICNS Woodard, X., and T. E. Potter, "Considera:ica of Source Term in Relation to Emergency Planning Requirements," presented to the Workshop of Tecnnical Factors Relating Impacts from Reactor Releases t: Emergency Planning, Be:nesca, Maryland, January 12-13, 1982.
Ga rrick, B. J., S. Xaplan, G. Apostolakis, D. C. Iden, X. Woodard, and T. E. Potter, " Seminar: Probacilistic Risk Assessment of Nuclear P0wer Pl ants," PLG-0141, July 1980.
Garrick, S. J., S. Kaplan, G. E. Apostolakis, D. C. Bley, and T. E. Potter, " Seminar: Probabilistic Risk Assessment as Applied to Nucl ear Power Pl ants," PLG-0124, March 1980.
- Woodard, X., and T. E. Potter, " Modification of the Reactor Safety Study Consequences C0motter Program (CRAC) to Incluce Plume Trajectories,"
presented to the 1979 ANS 25th Winter Meeting, San Francisco, California, Novemoer, 11-15, 1979.
- Woodard, X., and T. E. Potter, " Assessment of Noble Gas Releases frcm tne Three Mile, Island Unit 2 Accicent," present?d to the 1979 ANS 25th Winter Meeting, San Francisco,, California, November, 11-15, 1979.
Garrick, S. J., S. Kaplan, P. P. Bieniar:, X. Woodard, D. C. Iden, H. F. Perl a, W. Dicter, C. L. Cate, T. E. Potter, R. J. Duphily, T. R. Roboins, D. C. Bley, and S. Ahmed, "0PSA, Oyster Creek Probabilistic Safety Analysis," (Executive Summary, Main Repor,
Appendixes), PL3-0100 DRAFT, August 1979.
Woodard, X., ~ and T. E. Potter, "Precabilistic ?rediction of X/0 for R0utine Intermittant Gaseous Releases," Transactions of the American Nuclear Society, Vol. 26, June 1977.
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,, a Cen.d s C. Ricnards:n - Ri st Assass:en: -'echnei:gy Manager Fenn 5: :a University, 5.3. 24 es= ace Emineering est M.5. Cent :1 E.gineeri.y 5.4..=
San Cieg: 5. ara !Jnive.-si y, M.J..w :hea:f:s a
1970 University Of Pittsburgn,.v5A 1950 Mr. ?.icnares n has many yes?t cf pr:fassi:nsT nnd :.anagacen: e=<~4encs f n tne ruclear fisic.
?4 f 1.4c =e ?ressart:ac '4atar Ranc: r Otytsi:n
- ? 'destingneuse in 197E r.en Me =anaged tna React:r Fr::acticn Analy:ts Gnu; f:r per'eming nuciear plan safety analysi s aM, ::st 'recen:Ty, h:s =anaged Oe Rist Assas=en: Techecicqy Cr;ani:stica.
?-ice c this, W. RT:hardson us vf = Csif General A::cic wnere he w-iec :n dasign Of ce=r:1 ar4 safs y sys:acs fer 2e gas-csciec nuclear plant. At Wstinghousa, he has ;artici;4:ad in anc dire::ad a
=cer of ri sk assessmefr: ar4 safety analysis stacies fer a wide variety cf a;;it ariens.
W was a.;rincipai inyestiga::r in be:h :ne II.cn 5ta-tic: anc Incia:: Fofn: Station Rea : r Safety 5::cies. He direc:a4==
P9A s:::tes f: :he %szingncusa Cwrars Gr:u: 2a: addnssac =e
?:st-7MI NUKEG nguirecents en e=ergemy ;r:<acures and :;ert::r disciay e uincen:5. Mr. Rf =ar:s:n was :acnMcal a~d ; :gra: =arager ?:r :he Sritisa (NNC) ?.cferenca 'datar ?.sec r Safety Stu y.
He has ais: Ted the cavei:pmen f ecenccic anc finaxial ef si assass=ent eachnicuas f:r =e use f n new reac::r accei dastgn : mas:s.
Mr. Richards:n is a mecher Of =e IEEI and ANS ar.: has se-va: :n On Artirs ; ;u:s fer :-c sta:Atrds c: cit.aes. Ye is reviewing ta se -
ti:ns f:r : e ?RA r.ar.uai directed by MRC :: :e finisnad in i3Ei.
He f s ac acr er : -a::::: cr of : re =an 15 re::.~.: and pa:ers dealing with rf sk assas=e= and varicus aspects of nuclear pian dasign i
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T T T,'
- w
u UNITED STATES OF AMERICA 00LKETED N
NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD 83 MAR 24 A10:55 Before Administrative Judges:
James P. Gleason, Chairman t.
_ A a.
Frederick J.
Shon In -
. s SE.kVF l
Dr. Oscar H.
Paris MCd
)
In the Matter of
)
)
CONSOLIDATED EDISON COMPANY OF
)
Docket Nos.
NEW YORK, INC.
)
50-247 SP (Indian Point, Unit No. 2)
)
50-286 SP
)
POWER AUTHORITY OF THE STATE OF
)
March 22, 1983 NEW YORK
)
(Indian Point, Unit No. 3)
)
)
CERTIFICATE OF SERVICE I hereby certify that on the 22nd day of March, 1983, I caused a copy of (1) Licensees' Testimony of Dennis C.
- Bley, Donald F.
Paddle ford, Thomas E.
Potter, and Dennis C.
Richardson on Commission Question Five, and (2) Licensees' Testimony of Thomas E.
Potter on Commission Question Five, to be served by first class mail, postage prepaid on the folicwing:
l 4
l
James P. Gleason, Chairman Charles M. Pratt,.Esq.
l Administrative Judge Stephen L.
Baum, Esq.
Atomic Safety and Licensing Board Power Authority of the 513 Gilmoure Drive State of New York 4
Silver Spring, Ma ryland 20901 10 Columbus Circle-New York, New York 10019 Mr. Frederick J.
Shon f
Administrative Judge Janice Moore, Esq.
Atomic Safety and Licensing Board Counsel for NRC Staff U.S.
Nuclear Regulatory Office of the Executive j_
Commission Legal Director Wa shington, D.C.
20555 U.S.
Nuclear Regulatory Commission Washington, D.C.
20555 Dr. Oscar H.
Paris Administrative Judge Brent L.
Brandenburg, Esq.
Atomic Safety and Licensing Board Assistant General Counsel U.S. Nuclear Regulatory Consolidated Edison Company Commission of New York, Inc.
Washington, D.C. 20555 4 Irving Place New York, New York 10003 Docketing and Gervice Branch Office of the Secretary Ellyn R. Weiss, Esq.
U.S.
Nuclear Regulatory Cammission William S. Jordan, III, Esq.
Washington, D.C.
20555 Harmon and Weiss 1725 I Street, N.W.,
Suite 506 Joan Holt, Project Director Wa shington, D.C.
20006 Indian Point Project New York Public Interest Research Charles A.
Scheiner, Co-Chairperson Group Westchester People's Action 9 Murray Street Coalition, Inc.
Box 488 3
White Plains, New York 10602 I
Jeffrey M. Blum, Esq.
New York University Law School Alan Latman, Esq.
423 Vanderbilt Hall 44 Sunset Drive 40 Washington Square South Croton-On-Hudson, New York 10520 New Yo rk, New York 10012 Ezra I.
Bialik, Esq.
Charles J.
Malkish, Esq.
Steve Leipzig, Esq.
Litigation Division Environmental Protection Bureau The Port Authority of New York New York State Attorney and New Jersey General's Office One World Trade Center Two World Trade Center New York, New York 10048 New York, New York 10047
)
Alf r ed B. Del Bello Westchester County Executive Wescchester County 148 Martine Avenue White Plains, New York 10601 l
Andrew S.
Roffe, Esq.
New York State Assembly Alba ny, New York 12248 m.
e Ma rc L.
Pa rris, Esq.
Atomic Safety and Licensing Eric Thorsen, Esq.
Board Panel County Attorney U.S.
Nuclear Regulatory Commission County of Rockland Washington, D.C.
20555 11 New Hempstead Road New City, New York 10956 Atomic Safety and Licensing Appeal Board Panel Phyllis Rodriguez, Spokesperson U.S.
Nuclear Regulatory Commission Parents Concerned About Indian Washington, D.C.
20555 Point P.O. Box 125 Honorable Richard L.
Brodsky Croton-on-Hudson, New York 10520 Member of the County Legislature Westchester County Renee Schwartz, Esq.
County Office Building Paul Chessin, Esq.
White Plains, New York 10601 Laurens R.
Schwartz, Esq.
Margaret Oppel, Esq.
Zipporah S.
Fleisher Botein, Hays, Sklar and Hertzberg West Branch Conservation 200 Park Avenue Association New York, New York 10166 443 Buena Vista Road New City, New York 10956 Honorable Ruth W. Messinger Member of the Council of the Mayor George V.
Begany City of New York Village of Buchanan District #4 236 Tate Avenue City Hall Buchanan, New York 10511 New York, New York 10007 Judith Kessler, Coordinator Greater New York Council Rockland Citizens for-Safe Energy on Energy 300 New Hemstead Road c/o Dean R.
Corren, Director New City, New York 10956 New York University 26 Stuyvesant Street David H.
Pikus, Esq.
New York, New York 10003 Richard F.
Czaja, Esq.
Shea & Gould Joan Miles 330 Madison Avenue Indian Point Coordinato,r New York, New York 10017 New York City Audubon Society 71 West 23rd Street, Suite 1828 Amanda Potterfield, Esq.
New York, New York 10010 New York Public Interest Research Group, Inc.
Richard M. Ha rtzma n, Esq.
9 Murray Street, 3rd Floor Lorna Salzman New York, New York 10007 Mid-Atlantic Representative Friends of the Earth, Inc.
David R.
Lewis, Esq.
208 West 13th Street Atomic Safety and New York, New York 10011 Licensing Board Panel U.S.
Nuclear Regulatory i
Stanley B. Klimberg, Esq.
Commission l
General Counsel Washington, D.C.
20555 l
New York State Energy Office 2 Rockefeller State Plaza l
Albany, New York 12223
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Mr. Donald Davidoff Director, Radiological Emergency Preparedness Group Empire State Plaza Tower Building, Rm. 1750 Albany, New York 12237 Craig Kaplan, Esq.
National Emergency Civil
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Liberties Committee 175 Fifth Avenue, Suite 712 New York, New York 10010 Michael D.
Diederich, Jr.,
Esq.
Fitgerald, Lynch & Diederich 24 Central Drive Stony Point, New York 10980 Steven C.
Sholly Union of Concerned Scientists 1346 Connecticut Avenue, N.W.
Suite 1101 Washington, D.C.
20036 Spence W.
Perry Office of General Counsel Federal Emergency Management Agency 500 C Street, S.W.
Washington, D.C.
'20472 St ewa rt M. Glass Regional Counsel Room 1349 Federal Emergency Management Agency 26 Federal Plaza New York, New York 10278 l
l Melvin Goldberg l
Staff Attorney New York Public Interest Research Group 9 Murray Street New Yo rk, New York 10007 Jonathan L.
Levine, Esq.
P.
O. Box 280 New City, New York 10958 I
(
s i
Paul F.
Colarulli l
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