ML20058D592
| ML20058D592 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 07/23/1982 |
| From: | Weiss E HARMON & WEISS, PUBLIC INTEREST RESEARCH GROUP, NEW YORK, UNION OF CONCERNED SCIENTISTS |
| To: | CONSOLIDATED EDISON CO. OF NEW YORK, INC., POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| References | |
| ISSUANCES-SP, NUDOCS 8207270226 | |
| Download: ML20058D592 (30) | |
Text
{{#Wiki_filter:, a g('{I-{} (O!ildMOI'W p y y. M N t i fV e UNITED STATES OF AMERIC A ,;c3 NUCLE AR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 33YI. j' 9 In the Matter of ) [+ ) CONSOLIDATED EDISON COMPANY OF NEW YORK ) D o ck e t ' No's'. 50-247 SP (Indian Point Unit 2) ) 50-286 SP ) POWER AUTHORITY OF THE STATE OF NEW YORK ) (Indian Point Unit 3) ) ) UCS/NYPIRG RESPONSE TO LICENSEES' FIRST SET OF INTERROGATORIES UNDER COMMISSION QUESTION ONE UCS/NYPIRG herein files the following responses to Licensees ' first set of interrogatories under Commission Question One (dated 6/16/82). Before responding to specific interrogatories, UCS/NYPIRG presents the following general response to Licensees' large number of interrogatories r egarding probabilities of acciden ts. [ GENERAL RESPONSE TO INTERROG ATORIES REGARDING PROBABILITY I WAS H -14 00, the Reactor Safety Study, was the first major l attempt to quan tify the probability of serious reactor accidents. There has been a great deal of peer criticism of WASH-1400 including strong criticism of its completeness, its methodology, inscrutability, data base, its treatment of common cause f ailure and particularly of its f ailure to properly treat uncer t ain ties. See, e.g. The Risks of Nuclear Power Reactors, A Review of the NRC Reactor Safety Study, Union of /7 N& 8207270226 820723 l PDR ADOCK 05000247 gpy l g PDR -,_m-
s 2 i Concerned Scientists, 1977; NUREG/CR-0400, Risk Assessment Review Group Report to the U.S. Nuclear Regulatory Commission, H.W. Lewis, Ch airm an, 1978. Indeed, subsequent to the publication of the so-called 'tewis Committee " Repo rt, the NRC publicly disavowed its belief in the reliability of WASH-1400 's estimates of the absolute probability of reactor accidents. WASH -14 00 is the direct ancestor of the PRA 's done for Indian Point. Since the issuance of WASH-1400 in 19 75, a number of additional f actors lead to a conclusion that the WASH-1400 estimate of the frequency of core melt (or, at the very least, sufficient core damage to threaten containment and a major release of radioactive materials ) was too low. These factors include but are not limited to: (a) the occurrence of the TMI-2 accident; (b) a re-evaluation of the f ailure rate data used in WASH-14 00 in the light of new data suggests an increase of a f actor of three in core melt frequency due to this f actor alone [See letter from Milton S. Plesset, ACRS, to Hon. Morris K. Udall, dated 2/20/80, Subj e ct : Actual Component Failure Experience]; (c) a large number of occurrences of complete loss of safety function. [See memorandum from C.S. Long to Roger Mattson, dated July 30, 1979, subject : Review of LER 's for Loss of
3 I l i Safety Function due to Personnel Error and Defective Procedures", and NUREG/CR-2497, " Precursors to Potential Severe Core i Damage Accidents, 1969-1979, A Status Report", J. W. Minarick l and C. A. Kukielka, Science Applications, Inc., under subcontract to Oak Ridge National Laboratory, June 1982) ; (d) the occurrence of the Browns Ferry partial scram failure, which raised questions about the application of probabilistic risk assessment to reactor technology. The BWR scram system was probably one of the most studied systems using PRA, and the industry consistently argued that it was an extremely reliable system, but the Browns Ferry incident and subsequent investigations revealed numerous design flaws in the scram discharge volume system of existing BWRs. (John Stampelos, Garry Young, and Dorothy Zukor, ACRS Staff and Fellows, " An Interim Report to the ACRS on a Review of Recent Malfunctions of BWR Scram Systems", Preliminary Draft, August 8, 1980, cover memorandum from Stampelos, Young, and Zukor to Milton S. Plesset; Stuart Rubin and George Lanik, Office i for the Analysis and Evaluation of Operational Data, NRC, " Report on the Browns Ferry 3 Partial Failure to Scram Event on June 29, 1980", July 30, 1980; NUREG-0785, " Safety Concerns Associated with Pipe Breaks in the BWR Scram System", Draft, April 10, 1981; and NUREG-0803, " Generic Safety Evaluation l Report Regarding Integrity of BWR Scram System Piping", June 1981); t
(e) well-documented problems concerning the ability of anyone to perform an adequate systems interaction review (See NUREG/CR-1901, " Review and Evaluation of System Interactions Methods", A. J.
- Buslik, I.
A. Papazoglou, and R. A.
- Bari, i
Brookhaven National Laboratory, January, 1981; NUREG/CR-1896, i " Review of Systems Interaction Methodologies", P. Cybulskis, R. S.
- Denning, R.
- Gallucci, P.
Pelto, A. M. Plummer, and R. D. Widrig, Battelle Columbus Laboratories, January, 1981; NUREG/ f CR-1859, " Systems Interaction: State-of-the-Art Review and Methods Evaluation", J. J.
- Lim, T.
R.
- Rice, R.
K. McCord, and J. E. Kelly, Lawrence Livermore Laboratory, January 1981; and periodic computer printout reports by NRC of the TMI Action Plan Tracking System under item II.C.3, for example, September, 1981 report); (f) lack of consideration of sabotage in the Reactor Safety Studies and Subsequent PRA's and lack of inherent justification for assuming that sabotage will lead to core melt less frequently than other causes; (g) conclusion of WASH-1400 that seismic contribution to risk is negligible has been shown to be in error. (See paper by T. M. Hsieh and David Okrent, "On Design Errors and System Degradation in Seismic Safety", Trans. of 4th Int. Conf., San Francisco, Calif., 1977, paper K 9/4; and recent disclosures of design errcrs at the Diablo Canyon nuclear power plant in California); and
} 5 i (h) inadequate consideration of human error in WASH-1400 ( See NUREG /CR-18 79, " Sensitivity of Risk P arameters to Human Errors in Reactor Safety Study for a PWR", R.E. Hall, P.K. Sam an ta, and A.L. Swoboda, Brookhaven National Laboratory, January 19 81; NUREG /CR -249 7, " Precursors to Potential Severe Core Damage Accidents, 1969-1979, A S tatus Report ", J.W. Minarick and C. A. Kukielk a, Science Applications, Inc., under subcontract to Oak Ridge National Laboratory, June, 1982; and F.H. Fuchs, " Analysis of Nuclear Reactor Events Reported in M arch, 1979", Xyzxz Information Corporation, May, 1979, contained as an enclosure in a 5/18/79 letter from Kay Inaba, Xyzyx, to Hugh Warren, Richland, Washing ton, and also as an enclosure to a Routing and Transmittal Slip from Gary Bennett to W.S. F armer, et al., dated 6/22/79 ). For these reasons, UCS/NYPIRG believes that it is difficult to establish with confidence that the frequency of severe core damage or core melt for the present generation of nuclear reactors is less than the limit which would be justified based upon experience, i.e., approximately 0.001 to 0.0001 per reactor year. The above references are available in the NRC Public Document Room in Washing ton, D.C. See i
- also,
'"The Risk s of Nuclear Power Reactors : A Review of the NRC Reactor Safety Study WASH-1400 (NUREG-75/014 ) ", Edited by Rich ard B. Hubbard and Gregory C. Minor, Union of Concerned Scie nti sts, C am br idge, Massac huse tts, Augus t, 1977, copy I available upon request for inspection and copying. I
. = -... 6 i RESPONSES TO SPECIFIC INTERROGATORIES i 1. UCS/NYPIRG objects to this interrogatory. The licensing board itself is the only proper authority capable of defining terms used in contentions the wording of which is formulated j by the licensing board. 2. We have produced no such documents as of the present j time. We will provide any such documents if and when they are pr oduced. f i 3. Probabilistic risk assessment may be useful for limited purposes insof ar as the assessors are capable of obtaining l accurate estimates of probability and consequences. It is of i great importance that probabilistic risk assessment not be used to exaggerate the degree of certainty in estimates or give misleading impressions that minimize error range. Such use is clearly inappropriate. In general probabilistic risk assessment is more likely to be successful at approximating l j relative risks posed by different events. Probabilistic risk assessment might also be useful at the system level where j sufficient accurate data is available on component failure ( l modes and failure statistics. It should not be used to predic t the absolute probability of particular accidents. 4. See answer to number 3. 5. UCS/NYPIRG is aware of no such self-generated evaluations at the present time. l \\
7 s 6. Risks which are ascertainably insignificant are acceptable to UCS/NYPIRG. Risks are ascertainably insignificant if either consequences are insignificant or probability of occurrence is assuredly very low and consequences are not large. 7. UCS/NYPIRG objects to this interrogatory as being overbroad and calling for the inclusion of a large amount of irrelevan t material. 8. UCS/NYPIRG is aware of no generally accepted exposure level at which there is no risk of harm to the public. 9. UCS/NYPIRG have made no decision as to whether it will prese n t any witness on Commission Question 1, but will inform Licensees promptly when and if such a decision is made. 10. Dr. Vincent Taylor is preparing testimony on the economic costs of shutdown of Indian Point-Commission Question No. 6. Dr. Taylor is a consultant for the Union of Concerned Scien ti s t s. UCS/NYPIRG has been substantially assisted in the preparation of its case by Mr. Steven Sholly, a s taf f mem ber o f UCS. 11. a. Prompt fatalities are the same as early f atalities. See (b), below. b. Deaths which occur within 60 days of their proximate physical cause. c. Noticeable physical impairmen ts, either temporary or permanent, which occur within 60 days of their proximate physical cause. d. Same as answer to c., but which occur af ter 60 days since the date of their proximate physical cause.
O 8 i e. Carcinomas which are the proximate cause of death, f. Carcinomas which are not the proximate cause of ' death, g. Thyroid nodules are detectable physical growths on the thyroid gland. They m ay o r m ay no t become c an cer ous, h. Genetic defects are noticeable physical or mental impairments in succeeding generations caused by alteration of the DNA structure in preceding generations. Latent genetic defects are abnormalities in the DNA structure that could produce physical or mental impairments in future generations. Genetic effects include spontaneous abortions with causes related to alteration of the DNA structure in preceding generations. 12. Consistently with the board 's directives and expressed wishes. Our assumption is that this term will encompass both genetic defects and latent genetic defects as defined in 11 h. above. 13. (parts a-w ) Through Dr. Jan Beyea and Brian Palenik, UCS/NYPIRG has presented evidence of accident consequences l specific to the Indian Point site for a PWR-2 type release, l as defined in WASH-1400. For specific defining characteristics 1 of such a release se e WAS H-14 00. The witnesses found no l i plant-specific features for Indian Point which in their i j judgment would significantly alter the nature of a PWR-2 1 release from that described in WASH-1400. With regard to j probability, UCS/NYPIRG has not, as of the present time, l j calculated any specific probability of a PWR-2 type release 1 I
9 i
- 'at* Indian Point.
UCS/NYPIRG maintains that such probabilities j cannot be calculated with suf ficient certainty or with a l sufficiently limited error range to justify presentation of a central estimate. If any f urther information is generated by UCS/NYPIRG with regard to range of plausible probability j estimates, this answer will be updated. With regard to parts k throug h w, such information is found in the direct testimony of Jan Beyea and Brian Palenik i which has been prefiled and is in the possession of licensees. In general this testimony takes the Parsons-Brinckerhoff evacuation time estimates as given and calculates health and safety consequences on the basis of those estimates. With I regard to sheltering and meteorological assumptions, also see cross examination testimony of Jan Beyea and Brian Palenik. Transcript of hearings for July 8,19 82, and July 9, 1982. At this time, no other specific accidents sequences have been analyzed in detail; however, in researching these issues for preparation of the contention and for the hearing, UCS/NYPTRG has utilized NUREG CR/1244, " Impact of Rainstorm and Runoff I Modeling on Predicted Consequences of Atmosphere Releases from Nuclear Reactor Accidents, Sandia Laboratories, 1980, (available in NRC 's PDR ). This reference indicates that peak accident l consequences may increase by a f actor of approximately 2 when [ t certain weather conditions are considered.
10 Other references considered include SAND 78-0556, "An Investigation of the Adequacy of the Composite Population Distributions Used in The Reactor Safety Study, " Oct., 1978, Sandia Laboratories. This report concludes that for high consequence events, the probability of fatalities in RSS were underestimated by a f actor of 3 due to the use of composite populations. Site-specific calculations for Indian Point and Zion reported in that reference suggest that if all sites had populations as large and unf avorably situated as these 2 sites, the early f atality risk estimates of RSS would be substantially increased, by a factor of up to 10. Total risk of early f atality for Indian Point Unit 3 is indicated to be 12 times that of the RSS. The 10% difference in reactor size between Units 1 and 2 has an insignificant effect on risk. 14. Based upon a review of portions of the public docket files for Docket Nos. 5 0-2 4 7, and 5 0-286, the only descriptions of licensees' EALs of which UCS/NYPIRG is aware are contained in the respective onsite emergency response plans. Altho ug h the plans contain lists of initiating conditions and associated emergency classes (i.e., unusual event, alert, site emergency, or general emergency ), there is no means by which a reviewer can conclude that these lists are exhaustive. Cer tainly, for example, these lists of EALs do not appear to encompass the accidents analyzed in the Licensees ' Indian Point
11 Probabilistic Safety Study. Moreover, a number of events listed as initiating conditions appear to be relatively trivial matters ( from the standpoint of public risk ) which bear no obvious relati;nship to more severe events of which those detailed in the EALs may represent precursors. With the exception of the mention of the FSAR, no source documents are specified for the derivation of the EALs. The FSAR is not sufficient as such a reference since the FSAR for Indian Point is entirely limited to discussion of accidents which are within the design basis of Indian Point. Of cour se, in establishing EALs, a much broader range of accidents must be considered, as even a cursory reading of Appendix 1 to NUREG-0654 will readily demonstrate. In order to establish an adequate set of EALs, it is first necessary to undertake a detailed engineering evaluation of possible accidents. Then it must be determined how the symptoms of these accidents would be manifested on plant instruments and/or alarm systems. Finally, appropriate instrument readings and/or alarm indications must be related to the specific accident and clearly and unambiguously indicated to plant operators in order that the proper emergency class be promptly declared and necessary emergency response procedures set in motion. There is no evidence that this has been done l for Indian Point. Some of the EALs established in the onsite emergency plans for Indian Point appear to be generic in nature, 1 l l l i I
12 1 i.e., aimed at more than one particular accident sequence. Such EALs may be appropriate, but not without a demonstration that the generic EAL adequatel'y covers the intended events and provides early and phompt' notice of the events to the operator, and a demonstration that the operators vill-be able t to promptly discern whicn of the multiple events has caused the particular EAL limit to be exceeded. This has not been done for Indian Point'. 15. UCS/NYPIRG objects to this interrogatory inasmuch as it calls upon UCS/NYPIRG to perform work for which it is not responsible. The responsibility for creating appropriate EALs for Indian Point rests squarely with the Licensees. 16. " Heroic emergency measures" are emergency ' measures that did all that was humanly possible to protect the population-in the event of a major release of radioactivity. t 17. UCS/NYPIRG presumes that this vague re~ quest refers to Buis #1 of UCS/NYPIRG Contention I(B)( 5 ). Assuming that ~ to be the case, our response is as follows. Delay in the first step can be diminished by analyzing the range of possible 1 i accidents, in st alling instrumentation and alarm systems appropriate, to detecting and anne..ca cing such accidents to the operators, and then establi sh. w; >ropriate EALS. Delay in the second step can be diminished by ' improving operator training and by providing operators with' better instrumentation and operator aids such as computer-generated graphic displays of systems status and operating par ameter,s, symptom-oriented emergency i 1 w
,e1 s t u
- '.-gg-1a e
4e'-A-Dzt MJua 2 .s .,1c 5,._ o 13 procedures, and aids to assist the operators in gaining access to the proper emergency procedures. Delay in the third. step can be diminished by providing multiple and redundant means of communications with of fsite emergency response organizations 2 I and by ensuring through continuing contact and training that all organizations likely to be involved in a response to an i emergency at Indian Point have a mutual understanding of applicable terminology. Delay in the fourth step can be diminished l by providing sufficient and continuing training to emergency i i response personnel from offsite organizations who are charged with decision-making responsibilities, and by providing periodic practice in utilizing plant status information, meteorology, i and source term information together with information on the status of factors associated with public response to emergencies to reach decisions on what protective actions to take. Delay in the fifth and sixth steps can be diminished by installing a public alert and notification system which is capable under i~ all circumstances of alerting essentially 100% of the public within 10 miles of the Indian Point site, and by achieving and \\ j maintaining a high degree of public education so that the public understands what is expected of it in a radiological emergency. I 18. See response to Interrogatory No. 17. i 19. Such delay is assumed because the steps detailed in i l response to Interrogatory No.17 have not been implemented by the f b ... -,.. ~. .____,m m.m__,_.,_
~ 14 l Licensees. In addition, UCS/NYPIRG believe that delays-will i result from the fact that personnel relied upon to implement I the plan will see to the safety of their own families first. 1 The impact of such delays has not been fully quantified, but the impact could be bounded by assuming appropriate delay times between the initiation of an accident sequence the tak ing of protective actions by the general public, and the implementation of such protection. In general, UCS/NYPIRG would expect th at taking such delays into account would I increase the magnitude of the consequences of an accident. 20. UCS/NYPIRG has not evaluated the Indian Point designs for unmonitored release pathways. Nonetheless, we are unaware of any nuclear plant in the U.S. which has redundant means of monitoring each containment penetration and other systems which could become contaminated as a result l of a major accident for releases of radioactivity.
- Moreover, leakage pathways created by containment f ailure are going to 4
be inherently unmonitored since their location cannot be i i predicted in advance with a great deal of specificity. There are clearly circumstances in which the releases from unmonitored pathways could constitute a significant fraction of the total release during an accident, even to the point where the unmonitored releases could equal or exceed that from l 4 monitored pathways. Under such circumstances, it is imperative 5
i* 15 4 that means be in place for estimating the magnitude and isotopic content of the release. Reliance on the Reuter-Stokes 1 Sentri-1101. system may be misplaced for a variety of reasons, including masking of readings due to building shine, inability of the system to discriminate releases from monitored versus unmonitored pathways, inability of the system to effectively monitor buoyant plumes, and insufficient e. umbers of monitors E to ensure accurate readings under a variety of diffusion 1 conditions. These problems are detailed in a report to the NRC 1 l prepared by Exxon Nuclear Idaho Company, "An Assessment of Offsite, l Real-Time Dose Measurement Systems for Emergency Situations ", NUREG/CR-2644, ENICO-1110, W.J. Maeck, L.G. Hof f man, B.A. Staples, and J.H. Keller, April, 1982, available from the NRC's Public Document Room. 21. As stated in response to Interrogatory No. 20 above, UCS/NYPIRG has not undertaken a detailed review of unmonitored pathways at Indian Point. Such releases must, however, be accounted for in all does projections and recommendations for implementation of protective actions by the general public as i a result of an accident at Indian Point. In theory, a conservative bounding approach could be taken by assuming a given leak I rate for unmonitored pathways together with a standard or i l measured source term, but such a study should be soundly based l on an engineering evaluation of the ef fective size and l r flow rates associated with possible unmonitored release pathways. i l h i i e
__.~ _. _ 1 O 16 4 ] 22. UCS has not yet performed an analysis of the particular ways in which Indian Point fails to comply with Regulatory Guide 1.97, Revision 2. j 23. It is widely recognized that human exposure to radiation results in genetic defects. See, e.g. The Effects on Populations of Exposure to Low Levels of Ionizing Radiation (BEIR Report), National Academy Press, 1980; Nuclear Power Issues and Choices, Report of the Nuclear Energy Policy Study Group Ford Foundation, Mitre Corporation, Ballinger Press, 1977, Chapter 5 ; WASH-1400, Appendix VI, pp. 9 9 -31 f f. l 24. The accidents are those which result in a release of radioactivity in the range of that assumed for a PWR-2 i release category in WASH-1400. 25. All consequences to human health and safety due to short term exposure to radiation. These include early deaths, latent can cers, genetic effects, early injuries, and other j deleterious health consequences. 26. Thes e include, but are not limited to, a PWR-2 type release of radioactivity. In general these conditions are any accident sequence that results in the release of large amounts of radioactivity into the environment, j 27. See answer to interrogatory 26. UCS/NYPIRG endorses l no specific estimate of the probabilty of such occurrences. i I l t l l
17 UCS/NYPIRG believes that it is not possible to reliably quantify the absolute probability of accidents. i 28. a. A risk of harm to the public health and safety which intervenors are willing to accept, and a substantial segment of the public is or would be willing to accept. b. Harmful effects to either human beings or their i surrounding environment which intervenors are willing to accept, and a substantial segment of the public is or would be unwilling to accept. 4 29. We believe that the consequences of such accidents are not acceptable. 30-40. The Board has ruled that these interrogatories have already been answered. i 41. See answers to interrogatories 26 and 27. 42. Yes, contamination was defined in terms of total dose accrued in a given period of time.. Dose rate is implicit in this definition. 43. With regard to persons, parameters used were whole body rems and thyroid doses. With regard to land contamination, see question 11, Testimony of Jan Beyea and Brian Palenik, 44. See answer to 43. 45. These include areas which are within the path of the plume out to a distance of (depending upon the threshold level adopted for defining contamination) over 100 miles from the i i i s ---'M m aM-e ,,-u--~' 'r- --m --m-+--- + - -
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18 . Indian Point plants. In general the plume pathway is expected to approximate a wedge of 7.5 degrees. 46. The more highly contaminated areas will be contaminated at levels posing a severe risk to human health and safety for I a period of decades. See testimony of Jan Beyea and Brian Palenik, and particularly the graphs showing land contamination at I a distance of 35 miles. 47. We lack sufficient information to answer this interrogatory at the present time. We have not to date made a detailed analysis of the economic effects of an accident at Indian Point. We note that such estimates are generated by NRC 's CRAC code. 48. The number of persons irradiated at levels posing a risk of early deaths and illnesses will in all likelihood exceed the number of f acilities in the New York City region or the U.S. for supportive treatment of radiological victims. Appendix VI of WASH-1400 includes an estimate of capacity for the supportive treatment of such patients which is about 2,500-5000. 49. The assertion in the contention that neither the metropolitan New York region nor the nation could care for persons irradiated and contaminated as a result of a severe accident at Indian Point was based generally upon extrapolation of the consequence estimates of WASH-1400 to the Indian Pont site. W ASH-14 00 used composite population figures far below those for Indian Poin t. In addition, we believe
4 19 tba't the RSS methodology understates the numbers of early ~ illnesses. See The Risks o f Nuclear Power Reactors, supra, Chapter 8. WASH-14 00 contained an estimate of the number of hospital beds available nationwide to deal with the injuries and illnesses. The estimate was that 2,500-500 people could receive supportive treatment and 50-150 could receive " heroic treatment " WASH-14 00 App. VI. p. 9 -3. In the view of UCS /NYP IRG, this is less than the number of people that could require treatment in the event of an accident like PWR-2 at the Indian Poin t site. We also note that, as the Lewis Report pointed out, although WASH-1400 assumed that resources around the nation would be mobilized in the event of a serious nuclear accident, i the ability to carry out the intervention "not only has not been demonstrated, but isn 't even well planned at this time. " NUREG /CR-0400 Risk Assessment Review Group Report to the U.S. Nuclear Regulatory Commission, " 1978. 50. WASH-1400 and NUREG /CR -0400 are in the NRC PDR. The UCS publication can be purchased on request. 51. The basis for the contention did not include a specific radiation dose that would render geographic ar eas l unsuitable for many forms of lif e. This is a question that t c annot be answered with precision, since it involves social i and political judgments about the degree of risk which persons re-inhabiting contaminated areas would be willing to take and I at what cost the radioactivity can be removed from land and 1 buildings. i
20 The following quotation from the Risks of Nuclear Power
- Reactors,
- p. 94, is instructive:
The two most sig nif ic an t questions here are : How much long-term external radiation dose and internal radioactivity are members of the public to be allowed upon returning to a contaminated area, and how ef fectively and at what cost can reactor-released radioactivity be removed from land and buildings? The RSS answers the first by reference to ten-year old guideline of the former Federal Radiation Council (FRC), to recommendation of the International Commission on Radiological Protection, and to those of the Medical Research Council ( MRC ) of Great Britain. The RSS authors chose in the end to ad apt from the FRC and MRC models. (See RSS, Appendix VI, Table VI 11-6) The resulting criteria, when taken toge ther, suggest that those living in rural areas will be allowed 10 rem of whole body dose in a 30-year period, while those living in urban areas will be allowed 25 rems, in the same period. These are permissive standards indeed. Although the doses may be unavoidable following a major acciden t, they are nonetheless in striking contrast to the exposure standards recently promulgated by the EPA for normal operation of the entire nuclear fule cycle. (Feder al Register, 42, No. 9, pp. 2 55 8-25 61, Janu ary 13, 1977: 0.025 rem per year to the whole body including bone marrow, and 0.075 rem per year to the throid which are equivalent to 0.75 rem and 2.25 rem, over 30 years, respectively ). Over a 30-period the RSS criteria are 13 to 33 times less strigent for whole body than the EP A's. Using an estimate from the widely quoted NAS/NRC Biological Effects of Ionizing Radiation Report (1972), a dose of 30 rem in 30 years is most likely to cause approximately 180 additional c ancer deaths per year, or 5,400 total in 30 years, in a population of one million exposed persons. In addition, there would be an inevitable increase in public anxiety. It is also not clear to these authors that the public would reoccupy contaminated land under such circumstances. If the RSS dollar costs for land denial and decontamination were reanalyzed based on the latest EPA standards, it is certain that the costs would be much higher than the present RSS values. h L
21 52. We had no specification in mind of the "m any f o rms of life ". Certainly, the consequences to human life ar e most significant and the testimony of Beyea and Palenik treated the consequences in terms of human life. It is worth noting, however, that virtually all experimentation in the effects of radiation, with the exception of Hiroshima and Nag asaki, has been on animals. 53. In the event that the New York City port area became radioactively contaminated, persons engaging in commerce would be less likely to enter that area voluntarily. Without the voluntary par ticipation of such persons, commerce in the ar e a would decline. See also NUREG/CR-2591, " Estimating the Potential Impact of a Nuclear. Accident, " Regional Economic Analysis Division, Bureau of Economic Analysis, U.S. Dept. of Commerce, April, 1982. 54. No simple answer to this question is possible. The answer obviously depends upon a large number of factors, including size of release and wind direction, which cannot be quantified except for specific accident sequences. The question is therefore hopelessly overbroad. 55. See answer to question 47. 56. See answers to 47 and 54. 57. NUREG-0850 is available at the NRC's Public Document Room in Washington, D.C.
22 58. See response to Interrogatory No. 57, above. 59. See NUREG-0850. 1 - 16 0. See response to Interrogatory No. 59, above. 61. See response to Interrogatory No. 59, above. 62. UCS/NYPIRG is in possession of no such data. 63 through 66. UCS /NYPIRG has not calculated any specific 1 probabilities asked for in these interrogatores as of the present time. If and when such probabilities are calculated, } our answers will be updated. As noted above, we do not believe that reliable estimates of the absolute probability of accidents can be made. 67. UCS/NYP IRG has not yet calculated any such fraction or percentage. See answer to interrogatories 63 through 66. 68. As a general matter UCS/NYPIRG believes that the i nuclear industry should not expand. One basis for this belief is the fact that there is a probability substantially greater than zero of a serious accident at a nucle ar power plant. 69. We have no such quantitive figures. l 70. UCS /NYP IRG has not performed any study of probabilit3 i of accidental release of radiation from specific nuclear power i plan ts. For a UCS critique of the methodology of probability estimates in WASH-1400, see The Risks of Nuclear Power Reactors, s A Review of the NRC Reactor Safety Study WASH-1400, Union of i Concerned Scientists, 1977. If and when such a study is done we will update this answer. i 1 .e-n -wr ._..~..o_, y .r.- ,-- --.,.. -, -,. - -. - ~. _. _ _ _ - - .--c-,-----.
23 71. Quantitive comparisons of this type are dif ficult because of attempting to calculate reliable risk coef ficients for a relatively new technology such as nuclear power plants. Where comparisons can be made without precise quantitive figures (e.g., nuclear power Irsessgreater risks than the use of solar power, windmills, or cogeneration), these difficulties are attenuated. It remains true, however, that no single statement of relative risk can accurately describe the nature of the risks pos ed ( e.g., sudden and catastrophic vs. lo ng term and chronic ). 72. See answer to 71. 73. UCS/NYPIRG does not accept the. initial premise of this question and is aware of no credible support for it. We therefore obj ect to answering such a purely hypothetical question. The question does not even provide sufficient information to enable it to be answered as a hypothetical. Eg. no definitions ar e provided of " aver ag e risk " or "compe ting technologies ". 74. See answer to 73. 75. UCS/NYPIRG believe that the consequence modeling component of risk assessment can and should be used to make comparisons between nuclear plants. That is, it is possible to show that the consequences of postulated accidents are far greater at some sites, Indian Point most prominently, th an at others.
24 76. UCS/NYPIRG object to this question on the ground-that, as stated, it does not appear to be relevant to any issue in this proceeding. In addition, we do not understand what is meant by the phrase "incorpor ate... finding s... in to its regulatory program" and hence object to the question as vague. Insofar as this proceeding is concerned, UCS/NYPIRG believes that probabilistic risk assessment does not yield suf ficiently reliable estimates of the probability of accidents to be a meaningful decision-making tool on that issue. We believe in addition that consequence modeling, such as that don e by Dr. J an Beyea, should be used, for example to support remote siting of nucle ar f acilities and, as in this proceeding to determine whether the consequences associated with major reactor accidents at sites surrounded by densely populated areas are acceptable. 77. See answer to 76. 78. UCS /NYP IRG is currently engaged in reviewing the Indian Point Probabilistic Safety Study and cannot now give any final answer to this interrogatory. The specific probability i estimates in the Indian Point study appear on their face to be l implausible low. As to which study is "more exhaustive, " we t I i I l
- 25 have not formed an opinion. As a general matter we concede j that both studies are voluminous, and that the gross tonnage of the Indian Point study appears to exceed that of WASH-14 00. 79. See answer to 78. If and when additional pertinent information is generated, our answers to interrogatories 78 and 79 will be updated. 80. Probabilistic methodology may be useful for identifying specific weaknesses in plant design that most urgently need to be corrected. If fault trees show a relatively high risk component attributable to f ailure of specific equipment or to a particular type of operator error, a high priority should be placed on correcting pertinent defects or ensuring that such error or component f ailure does not happen. 81. The current standard for backfitting, which appears in i 10 CFR 50.109 requires modification to a plant if it would " provide substantial, additional protection which is required for the public health and safety... " We believe that this is the appropriate standard. 82. In general UCS/NYPIRG finds t his question unintelligible. We are aware of no study showing that plants which have "ag ed " are more safe than new plants. Indeed, as plants age, equipment i ) deteriorates (e.g. steam generators ) and the environmental i e f 4 1 y p.
26 j qualification of equipment is called into question. It would appear intuitively correct that operators with increased experience would, all other things being equal, per form better than new operators, but we have seen no study of the degree of reduction in risk that might thereby result. In addition, factors other than experience would affect operator behavior, including training, intelligence, f amiliarity with equipment, man ag emen t attitude, etc. 83. The answer to this would probably depend upon the specific reactor and the specific site. In gener al, both new and old plants can constitute substantial risks to the I ] public health and safety. In addition, larger plants would pose greater risk than smaller plants. 84. "Never " and " older " are relative terms. We have not attempted to associate precise ages with these terms. UCS/NYPIRG believes that a reactor which has been in operation for less than two years is an example of " newer " reactor, whereas one that has been in operation for more than ten years is an example of an " older " reactor. However, the terms might also be used to differentiate among plants according to the standards which applied to each plant at its licensing. Depending upon the particular standard in question, a relatively recently licensed plant might be "old er. " i t 1 i
27 Submitted by : Ellyn R. Weis s, General Counsel Union of Concerned Scientists HARMON & WEISS 1725 I Street, N.W. Suite 506 Washington, D.C. 20006 (202) 833-9070 DATED: July 23, 1982
UNITUD STATUS OF Af1Mit I CA NUCLEAR El:CUI.ATORY COMMISSION BEFORE THE ATOMIC SAFETY AND 1.fCMNSING HOARD In the Matter of ) ) CONSOLIDATED EDISON COMP ANY OF NEW YORK ) Docket Nos. (Indian Point Unit 2) ) ) 50-247 POWER AUTHORITY OF T!!E STATE OP NEW YORK ) 50-286 (Indian Point Unit 3) ) I hereby certify that copies of the foregoing UCS/NYPIRG Response to Licensees' First Set of Interrogatories Under Commiss. ion Question One, was sent first class postage prepaid on this 26th day of' July, and as specified below to the following: Louis J. Cart er, Esq. .let f r oy M. Blum, Esq. Atomic Safe ty and Licensing floard New York lini versit.y Law School United Stat.cs Nuclear 4 /..! V.md cr bi l t llal l Regulatory Commission 10 Washing ton Square South Washing ton, D.C. 2 0555 New York, New York 10012 Dr. Oscar 11. Paris MS- 'I"" I10 I L Atomic Safety and Licensing Board New York Pubtic In ter es t Researc Uni ted States Nuclear Group Regulatory Commission S neckman Street Washing ton, D.C. 2 0555 New York, New York 10038 Dock e t ing & Service (2) Mr. Frederick J. Shon U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Boar Washing ton, D.C. 2 055 5 Unit ed States Nuclear Regulatory Commission Washing to n, D.C. 2 0555 Drent L. Brandenburg, Esq. Richard P. Remshaw John D. O 'Toole
- Jan i cc Moore, Hsq.
Consolidat ed Edison Company 01iico of t.hu Exocu t. i ve of New York, Inc. I.eg al Director 4 Irving Place Ifnited r,Lat.cs Nuclear Ni w York, New York 1000.1 RnjuIatory Commisuion Washington, D.C. 2 0555 Charles J. Manlkish, Emt. Ms. I'.i t l'osner, Spokesperson General Counsel Parents Concerned About
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_ Ms. Amanda Potterfield, E;quir Renee Schwart z, Esq. P. O. Dox 384 Botein, llay8, Sklar and lit'rzberg Villayo Station A* 200 Park Avenue New York, New York 10014 New York, New York 10166 lionorable Ruth W. Messinger Mr. Donald L. Sapir, Esquire Council Member 60 Eant Mount Airy Road 4th District, Manhattan Croton-on-iludson, N.Y. 10520 City llatl New York, New York 10007 Richard M. Hartzman, Esq. Ms. Lorna Salzman Atomic Safety and Licensing Friends of the Earth Board 200 West 13th Street U.S. Nuc1 car Regulatory New York, New York 10011 Commission Washington, D.C. 20555 Mr. Alfred D. Del Dollo Westchester County Executive westchester County Atomic Safety and Licensing 148 Martine Avenue Appeal Daard New York, New York 10601 U.S. Nuclear Board charles Morgan, Jr. Joan Miles Morgan Associates, Chartered Indian Point Coordinator 1119 9 1.. S t., N.W. New York City Audbon Society Washing ton, D.C. 20036 71 West 23rd Street, suite 18. New York, NY 10010 David B. Duboff thomas R. Prey, Esq. Westchester Peoples' Action Charles M. Pratt, Esq. Coalition Officp of the concral Counsel 255 Grove Street Power' Authority of the State of New YorkWhite Plains, N.Y. 10601 10 Columbus Circle New York, New York 10019 Craig Kaplan, Esq. Ruthanne G. Miller, Esq. National Emergency Civil Atomic Safety and Licensing Committee Board Panel 175 Fifth Avenue, Suite 712 U.S. Nuclear Regulatory New York, N.Y. 10010 Commission Washington, D.C. 20555 Donald Davidoff Director, Radiological JONATHAN D. FEINBERG Emergency NEW YORK STATE PUBLIC SERVICE Preparedness Group COMMISSION Empire State Plaza, Tower THREE, EMPIRE STATE PLAZA
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