ML20066J041

Responses to First Set of Interrogatories & Document Requests on ASLB Questions 1,2 & 5.Certificate of Svc Encl

ML20066J041
Person / Time
Site:	Indian Point
Issue date:	11/19/1982
From:	Brandenburg B, Colarulli P CONSOLIDATED EDISON CO. OF NEW YORK, INC., MORGAN ASSOCIATES, POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	PUBLIC INTEREST RESEARCH GROUP, NEW YORK, UNION OF CONCERNED SCIENTISTS
References
ISSUANCES-SP, NUDOCS 8211230429
Download: ML20066J041 (135)

Text

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'82 MV 22 #059 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION 'f 2'C-DT r: r;Rgfy, U"N' e. EC't,r r' ATOMIC SAFETY AND LICENSING BOARD lNCM Before Administrative Judges:

James P. Gleason, Chairman Frederick J. Shon Dr. Oscar H. Paris

)

In the Matter of )

)

CONSOLIDATED EDISON COMPANY OF NEW YORK, INC. ) Docket Nos.

(Indian Point, Unit No. 2) ) 50-247 SP

) 50-286 SP POWER AUTHORITY OF THE STATE OF NEW YORK )

(Indian Point, bnit No. 3) ) Nov. 19, 1982

)

LICENS8ES' RESPONSES TO UCS/NYPIRG FIRST SET OF INTERROGATORIES AND DOCUMENT REQUESTS ON BOARD QUESTIONS ONE, TWO, AND FIVE Preface Pursuant to 10 C.F.R. $$ 2.740b ana 2.741 (1982), the Consolidated Edison Company of New York, Inc. and the Power Authority of the State of New York, licensees, hereby submit these responses to the Union of Concerned Scientists /New York Public Interest Research Group, Inc.'s first set of interrogatories and document requests concerning the Nuclear Regulatory Commission's Questions 1, 2 and 5.

Brent L. Brandenburg Charles Morgan, Jr.

Paul F. Colarulli 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 i212) 460-4600 (202) 466-7000 i

8211230429 821119

{DRADOCK 05000247 PDR psos

TABLE OF CONTENTS I. PRELIMINARY MATTERS........................... 1 II. RESPONSES TO INTERROGATORIES.................. 2 i

, e PRELIMINARY MATTERS In several instances the Union of Concerned Scientists (UCS) and the New York Public Interest Research Group, Inc.

(NYPIRG) have made requests for documents. In accordance with 10 C.F.R. Part 2, the licensees are prepared, while reserving any claims of privilege or other objections to such production, to produce for inspection and copying documents requested under Commission Questions 1, 2, and 5 at a time mutually convenient to the licensees, UCS and NYPIRG.

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, , RESPONSES TO INTERROGATORIES UCS/NYPIRG Interrogatory No. 1:

With respect to each person whom the licensees intend to call as a witness regarding Board [ sic] Questions 1 and 2 in this proceeding:

a. identify the name, address, and organizational affiliation (s) of each such person;

b. state in full the educational and professional background of each such person, including occupation, institutional and professional affiliations, profes-sional certifications and/or licenses, publications, papers, books ana/or other published writings of each such person;

c. provide copies of all such publications, papers, books, and/or other published writings of each such person pertaining to probabilistic risk assess-ment, consequences of nuclear reactor accidents, proba-bilities of nuclear reactor accidents, and risk compar-isons between different nuclear powerplants and/or ,

between nuclear powerplants and non-nuclear electrical generating stations, whether such publications, papers, books, and/or other published writings are generic in nature or deal with Indian Point;

. . d. identify the contention as to which each such person will testify;

e. describe the nature and subject matter (s) of the testimony which will be presented by each such per-son, including an identification of all documents which the person will rely upon in the testimony; and

f. identify by court, agency, or other body, and by proceeding, date, subject matter, and transcript pages all prior testimony by and direct and cross-examination of all prior testimony by each such person.

Response to Interrogatory No. 1 (a) - (c), (f)

See the enclosed resumes.1

d. The following witnesses are presently intended to be offered.2 Contention 1.1 The licensees presently intend to offer as witnesses B. J. Garrick, D. Bley, D. Goeser, S.

Kaplan, D. Richardson, N. Liparulo, H. Perla, R.

Henry, T. Potter, D. Walker, W. Stratton, W.

1. The resumes of D. Goeser, R. Sero, and the Power Authority witnesses will be supplied shortly.

2. The licensees reserve the right with respect to Question 2 contentions to name additional witnesses after the intervenors file their testimony, because the inter-venors to date have failed to sufficiently specify the nature and extent of the safety measures proposed under Question 2. For example, the proper probabilistic evaluation of a safety measure requires that intervenors supply the parameters of the proposed safety measure.

Rodger, and B. Cohen. The Power Authority also intends to of fer as a witness R. DuPont.

Contention 2.l(a) and (d)

The licensees presently intend to offer as witnesses D. Richardson and D. Bley. The Power Authority also intends to offer as witnesses J.

Brons, J. Kelley, and J. P. Bayne.

Contention 2.2(a)

The licensees presently intend to offer as a witness D. Bley. The Power Authority also intends to offer as witnesses J. Brons, W.

Spataro, J. Leona rd , and K. Chapple. Con' Edison also intends to offer as witnesses S. Rothstein and A. Tuthill.

Contention 2.2(b)

The licensees presently intend to offer as witnesses R. Sero and D. Bley. The Power Authority also intends to offer as witnesses K.

Chapple, J. De Roy , J . Brunetti, and D. Speyer.

e. It is presently planned that John Garrick, Stan Kaplan, and Dennis Bley will testify as to the principles of probabilistic risk assessment and their application to the probability and consequences of postulated serious accidents at the Indian Point units in their present and presumed alternative configura-tions; the methodology, input parameters and results of

i l the Indian Point Probabilistic Safety Study (IPPSS), l including the analysis of plant ' systems and the development and use of fault and event' trees in a probabilistic framework; the implications of NUREG/CR-2497 to the IPPSS; the risN of the Indian 1

Point units compared to other units and other sites; and the implications of NUREG/CR-2239 (" Technical Guidance for Siting Criteria Development") to the comparative risk of nuclear power plants.

It is presently planned tha't' David'Goeser will testify as to the principles of probabilistic risk assessment and their application to the probability and consequences of postulated serious accidents at the Indian Point units in their present and presumed alternative configurations; the methodology, input parameters and results of the IPPES; the phenomenology occurring within the containment structure; and the analysis of containment response during postulated serious accidents at nuclear power plants.

It is presently planned that Dennis Richardson I will testify as to the probability and consequences of l

postulated se*-nts accidents at the Indian Point units in their . D ae , and presumed alternative configura-tions employing probabilistic risk assessment tech-niques, and the input parameters and results of the IPPSS, including the analysis of plant systems, the

analysis of containment response during postulated serious accidents at nuclear power plants, and the risk of the Indian Point units compared to other units and other sites.

It is presently planned that Nick Liparulo will testify as to the probability and consequences of postulated serious accidents at the Indian Point units in their present and presumed alternative configura-tions, and the methodology, input parameters and results of the IPPSS, including the analysis of plant systems and the analysis of containment response during postulated serious accidents at nuclear power plants.

It is presently planned that Hal Perla will testify as to the probability and consequences of postulated serious accidents at the Indian Point units, and the methodology, input parameters and results of the IPPSS, including the analysis and modeling of the probability of postulated serious accidents arising from externally initiated events such as earthquakes and hurricanes.

It is presently planned that Robert Henry will I

testify as to the effects of postulated serious i

accidents at the Indian Point units upon plant systems and the phenomenology occurring within the containment l

l structure during postulated serious accidents at l

nuclear power plants.

l

It is presently planned that Thomas Potter will testify as to the probability and consequences of postulated serious accidents at the Indian Point units and the modeling of public protective response actions following such postulated accidents employing prob-abilistic risk assessment techniques.

It is presently planned that D. Walker will testify as to the phenomenology occurring within the containment structure and plant systems during postu-lated serious accidents at nuclear power plants.

It is presently planned that William Stratton will testify as to the phenomenology occurring within the containment structure and plant systems during postu-lated serious accidents at nuclear power plants.

It is presently planned that Walton Rodger will testify as to the phenomenology occurring within the containment structure and plant systems during postu-lated serious accidents at nuclear power plants.

It is presently planned that Sam Rothstein will testify as to the phenomena of postulated steam gener-ator tube rupture and corrosive-initiated events at the Indian Point units, and techniques for preventing such events.

It is presently planned that Bernard Cohen will testify as to the probability and consequences of postulated serious accidents at nuclear power plants

, . e.ider alternative configurations, and the techniques for quantifying the frequency of occurrence of public health effects arising trom initiating events including accidents at nuclear power plants.

It is presently planned that Raymond Sero will testify as to the phenomena of postulated pressurized thermal shock events and preventive techniques at nuclear power plants.

It is presently planned that Arthur Tuthill will testify as to the phenomena of corrosive-initiated events at power plants.

It is presently planned that John Brons will testify as to a filtered vented containment system, a separate containment' structure, and as to the cooling system at the plant.

It is presently planned that John Kelley will testify as to a filtered vented containment system and a separate containment structure.

It is presently planned that J. P. Bayne will testify as to a filtered vented containment system.

It is presently planned that Kenneth Chapple will testify as to the issue of pressurized thermal shock and as to the cooling system at the plant.

It is presently planned that J. De Roy , J . Brunetti and D. Speyer will testify as to the issue of pressurized thermal shock.

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It is presently planned that Robert DuPont will testify on perceptions of risk of nuclear power.

It is presently planned that William Spataro will testify as to the cooling system at the plant.

UCS/NYPIRG Interrogatory No. 2:

Provide all documents which contain and/or pertain to evaluations, assessments, critiques, and/or criticisms of the licensees' " Indian Point Probabilistic Safety Study",

including appendices and attachments thereto. Such docu-ments include but are not limited to such evaluations, assessments, critiques, and/or criticisms whether or not done at the request of or under contract to the licensees, ano specifically include all such evaluations, assessments, critiques, and/or criticisms performed by Norman Rasmussen, Ian Wall, and Saul Levine, either singly or in combination.

Response to Interrogatory No. 2:

See Preliminary Matters.

UCS/NYPIRG Interrogatory No. 3:

Regarding all accidents at Indian Point discussed, analyzed, or evaluated in the Indian Point Probabilistic Safety Study or otherwise considered or evaluated by the licensees in preparing testimony for this proceeding:

a. specify the types of nuclear accidents or other accident scenarios considered, quantify the prob-ability of occurrence of each accident, and if such quantification is generic or for another plant, state how the quantification would vary for Indian Point, j

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which vary the quantification, state the basis for this answer, anc provide all documents'which contain and/or pertain to such quantification;

b. describe the fission product inventory within the containment, including type, chemical forms and quantities of isotopes, for each accident considered;

c. specify the source reduction factors used to calculated [ sic] plateout, washout, filtering, and all other fission product ren. oval processes, whether dependent upon natural processes or initiated as a con-sequence of the operation, misoperation, or malfunction of plant equipment and/or systems for each accident considered;

d. specify the release categories assumed, incluaing the percent released for all classes of radionuclides released, and provide all documents which contain and/or pertain to these release categories and the percent release [ sic] for each radionuclide class assumed;

e. quantity lsic] the probability of occurrence of each such release category, and provide all docu-ments which contain and/or pertain to these release categories and their probabilities;

f. specify all mechanical and structural contain-ment failure modes assumed and the timing of such fail-ures relative to the initiation of each accident con-l

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. . sidered, and idehtify precisely how such failure modes would be identified by licensed operators at Indian Point with specific reference to instrumentation (whether or not safety grade and specify which) and the associated indication on that instrumentation corres-ponding to the failure mode assumed, and provide all documents which contain and/or pertain to these analyses;

g. quantify the probability of the occurrence of each failure mode described in Interrogatory No. 3e;

h. specify all assumptions made in the analysis concerning containment safeguard features and calcula-tions of the availability of each feature;

i. specify all action (s) a licensed operator can take or direct to be taken to terminate a degraded core accident before reactor pressure vessel failure occurs, specify precisely how such an operator would become aware of the existence of such a degraded core acci-dent, specify what procedures would be utilized by such an operator in responding to such a degraded core acci-dent and provide copies of the same, and quantify the probability of the operator correctly identifying the existence of a degraded core accident and correctly initiating and completing the proper corrective action in sufficient time to prevent reactor pressure vessel failure;

. . j. fully describe all training received by licensed operators related to degraded core accident mitigation as described in Interrogatory No. 3i;

k. specify the pressures, temperatures, humidity, hydrogen gas concentrations, and oxygen gas concentra-tions assumed and/or calculated for the containment during the accidents discussed, analyzed, or evaluated, specify all other assumptions, both conservative, realistic, and non-conservative, and specify the effects of both the conservative and non-conservative assumptions upon the release of radioactivity to the environment;

1. specify the accident scenarios and/or accident progressions which form the basis of the calculations and assumptions described in Interrogatory No. 3k, and provide all documents which contain and/or pertain to such calculations;

m. specify the containment leakage rate which forms the basis of the calculations and assumptions described in Interrogatory No. 3k;

n. specify the degree of radiation exposure assumed to produce any and all health effects discussed in the Indian Point Probabilistic Safety Study, and provide all documents which contain and/or pertain to the doses and effects and the relationship between the doses and effects assumed;

, , o. specify how public response to all accidents -

discussed, analyzed, or evaluated was modelled, includ-ing each and every response option modelled, the prob-ability of each response option modelled (including the methods by which the probability was calculated, the uncertainty of the probability, and the assumptions used in calculating the probability and the uncer-tainty), the distance to which sheltering, evacuation, thyroid prophylaxis, relocation, and any and all other protective response options modelled were assumed to be utilized and the degree of effectiveness of each such protective response option, and specify how special population groups, including but not limited to persons who are deaf, blind, too young to understand instruc-i tions, who do not speak English, who are immobile, or who suffer from or are affected by any condition which could limit the extent to which such persons could understand and/or comply with protective action instructions, were treated in the analyses;

p. specify the demographic and population dis-tributions which wre [ sic] utilized, including the source (s) of all such data;

q. specify the meteorological dispersion model(s) and measurements which were utilized, including the source of all measurements, the location of all mea-surement stations, the instrumentation at each such

, . measurement stations [ sic], and the accuracy of each such instrument expressed as a percent of the value being reported by the inetrument or other appropriate expression, and provide all accuments which contain and/or pertain to these analyses;

r. specify the geographic area (s) considered, and provide legible, clearly delineated maps which set forth these geographic areas with reasonable speci-ficity;

s. specify the size (s) of the assumed plume expo-sure and ingestion exposure pathways;

t. specify the population (s) assumed to be evacu-ated, the assumed rates and paths of evacuation, and delay time between any Emergency Broadcast System announcements and tae beginning of the evacuation, the time assumed to be required to complete public notifi-cation of the need for an evacuation, and the time assumed to be required to complete the evacuation;

u. state whether different age groups were treated differently with respect to evacuation;

v. specify the number of persons assumed to be sheltered within the plume and ingestion exposure path-ways, and specify the degree of sheltering assumed to exist for these persons (in terms of protection fac-tors, dose reduction factors, or other appropriate cri-terion);

w. provide all documents which contain and/or pertain to the sheltering assumed to exist for all per-sons in the plume and ingest' ion exposure pathways as discussed in Interrogatory No. 3v;

x. specify the assumed time estimates for com-mencing sheltering and the assumed duration of shelter-ing as discussed in Interrogatory No. 3v; and

y. specify the degree to which probabilistic analysis was considered and/or utilized with respect to meteorological conantions [ sic], containment failure modes, accident scenarios, and release categories, and provide all oocuments which contain and/or pertain to these analyses.

Response to Interrogatory No. 3:

a. Section 1 of the IPPSS addresses the quanti-fication of initiating events and plant systems response. The quantifications of the probabilities of occurrence for all accident scenarios in the IPPSS are specific to Indian Point Units 2 and 3.

The specific initiating event categories or types (internal and external) which have the potential to lead to core damage are identified in Section 1.1. The probability distributions for the frequency of occur-rence of internal initiating events are shown in Sections 1.0 and 1.4, while the external initiating event frequency distributions are shown in Section 7.

. . The IPPSS containment analyses were performed for each of the six core damage states delineated in Sec-tion 2.3, including Table 2.3-2. A further discussion of containment responses to core damage states is included in Section 4.2.

The licensees object to the request for documents as excessive, burdensome, overly broad, and insufficiently specific to permit the licensees to determine which documents have been requested.

b. The fission product inventory within the con '

tainment for each accident sequance is calculated by the COkRAL Code as releases from the containment to the environment as a function of time. The CORRAL results in the IPPSS were for fission product releases to the environment. Some values for quantities of radio-activity present in the containment atmosphere as a function of time are available in the detailed CORRAL printouts, but this level of detail was not included in f the IPPSS. The types and quantities of fission product species considered are listed by release category definition in Section 5.2 of the IPPSS. Forms of fission product species considered were noble gases, vapor phase molecular iodine, and vapor phase organic iodide. All other species were considered to be present in the containment as particulate material.

The size of particulate material was assumed to be the

same as that utilized in the Reactor Safety Study (RSS)

} (Appendix VII, Section J2.4 of the RSS) .

c. Data for particulate and iodine removal rates i

t by filtering are given in Table 5.8.3.2-1 of the IPPSS. This table includes other plant parameters uti-lized in calculating filter removal rates. The CORRAL Code was utilized for the calculation of radioactivity source reduction in the containment for plateout, washout and settling processes. The basic physical parameters utilized in CORRAL calculations are set forth in Appendix J to Appendix VII of the RSS. Plant-specitic parameters necessary for the Indian Point CORRAL calculations are listed in Table 5.8.3.2-1 of the IPPSS.

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d. The release categories utilized are specified in Table 5.4-1 of the IPPSS. Their application is discussed in Section 5.3 and is summarized in Figure 5.3-1. Fractional releases for the release categories are listed in Table 5.4-2. Release values for some of the release categories utilized in the IPPSS were taken i

f rom the RSS as specified in Table 5.4-1. Such release categories are discussed in Appendix V of the RSS.

e. The annual frequencies of occurrence of release categories due to all initiators are found in Figure 8.1-2 of the IPPSS for Indian Point Unit 2, and

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in Figure 8.1-5 for Indian Point Unit 3. These results are discussed in Section 8.1.2. I

f. The types of containment failures identified

_re containment breach due to a large seismic event or averpressurization from steam, hydrogen burn or noncondensable gas generation, penetration of the con-tainment basemat due to core / concrete interaction, and containment bypass.

For accidents involving a loss of all active containment safeguards, containment overpressurization failure is predicted no sooner than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> into the accident for the probable accident progression as indicated in Section 4.3.5 of the IPPSS. Consideration was given to some extremely improbable paths in which overpressurization could occur one to three hours after the initiation of the accident as discussed in Sections 2.2 and 5.3.2.

Containment breach for a large seismic event would occur at the initiation of the accident.

As a result of a containment failure due to overpressurization, the containment would undergo a rapid depressurization. High-range pressure recorders, located in the control room, would identify this rapid depressurization to the operator. These pressure recorders are safety grade and were installed in response to NUREG-0737, Item II.F.1.4.

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, , " Temperature detectors located within the contain-ment measure the environmental temperature. This temperature is indicated in the control room. The +

equipment indicates temperature up to 150*F. The equipment is not safety grade.

Basemat penetration is likely to occur only for those situations in which containment safeguards are not operational; this is discussed in Sections 2.2.4 and 3.2.15.

Containment bypass may be the result of failure of the interfacing systems yielding an early direct release (one-half to three hours after accident initiation) to the auxiliary building, or from a failure to isolate the containment, or from a steam generator tube rupture accompanied by a failed secondary relief valve as described in Section 1.3.4.

A failure of one of these systems would be indicated by an increased steam (humidity) concentration in the l auxiliary building coupled with primary system indicators.

Instrumentation which could be used to indicate containment failure modes consists of primary system l

pressure, temperature and pressurizer level monitoring l

systems; secondary system flows, pressure and level monitoring systems; containment pressure and temper-

, , ature monitoring systems; and plant radiation monitor-ing systems.

g. See Response to UCS/NYPIRG Interrogatory No.

3e.

h. The availability of Indian Point Units 2 and 3 containment safeguards features was quantified and discussed in Sections 1.5 and 1.6 of the IPPSS. For the specified accident sequences with available containment sprays or containment fan coolers, the sprays or fan coolers were assumed to remain available during the progression of the accident sequence. The features used are summarized in Table I, Appendix 4.4.8; minimum safeguards were used to calculate the containment pressure response for each accident scenario analyzed.

A detailed assessment verifying operability of containment safeguard systems during these events was performed and is discussed in Response to Staff Interrogatory Nos. 10, 11, and 12, Licensees' Responses to NRC Staff First Set of Interrogatories Concerning the Commission's Questions 1 and 2 (June 25, 1982).

i. The licensees object to this interrogatory as I

l vague and overbroad. Any accident or unusual event is l

a " potential" " degraded core accident," but hardly any accidents proceed to the " degraded core" state. Thus, i

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. all operator actions are intended, in a wide variety of ways, to " terminate" a " degraded core accident."

Notwithstanding this objection, the licensees provide the following response.

All actions which the operators can take to

" terminate" a " degraded core accident" are described in the Operating and Emergency Procedures. The procedures are related to specific circumstances which may exist at any time.

In the course of performing the 1PPSS, specific operator actions were accounted for in the development of the event scenarios. These actions are described throughout Section 1. For most of the operator actions, see Sections 1.3.2.2, 1.3.2.3, 1.3.3.9, and the individual system analyses in Sections 1.5 and 1.6.

Given that significant core damage has occurred, the probability of the operator failing to take action I

to prevent reactor pressure vessel failure was con-servatively estimated to be one in the IPPSS, except for one accident scenario. For that scenario (Class IV event) the probability of vessel failure was taken to be 0.9 af ter significant core damage has occurred. The risk results in this case are insensitive to the value assigned because containment failure is unlikely.

j. All licensed operators are trained in miti-gation of core damage in compliance with NUREG-0737.

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Training programs are contained in the licensees' cubmittal to the Nuclear Regulatory Commission (Commission) concerning NUREG-0737, dated February 26, 1981, and July 1, 1982, for Indian Point Units 2 and 3, respectively.

h. The results of the calculations for contain-ment conditions during postulated severe accident sequences are given in Vol. 8, Section 4.4.9 of the IPPSS. These results are given for the various acci-dent sequences, and include best estimate and para-metric calculations to illustrate the sensitivity to various assumptions and/or models. Assamptions for each class of accidents analyzed may be found in Section 4.2. The results of the analyses performed for each class of accidents are found in Section 4.3 and are summarized by class in Tables 4.3.1 to 4.3.6, on pages 4.3-12 to 4.3-24.

Even in the best estimate analyses, several con-servative assumptions were employed and are delineated in Vol. 7, Section 3. A detailed uncertainty treatment of the containment analysis is displayed in Section 8.5.7.

1. The various accident scenarios and/or accident progressions which form the basis of the calculations and assumptions described in the Response to UCS/NYPIRG Interrogatory No. 3k are discussed in Section 2.3 of

the IPPSS in the tables at the end of that section, in Section 4.2, and in Section 4.3.

m. For the cases studied in which the Indian Point Unit 2 and Unit 3 containment pressures were below the assumed failure limit of 126 psig as discussed in Section 5.3.6 of the IPPSS, a leak rate of 0.14 per day was employed in order to define the release rate of the radionuclides. This leak rate and associated release categories were used to define the risk associated with two general situations:

1) Containment leakage for a core melt event in which the containment does not fail; and

2) Containment leakage prior to the estimated failure time for the containment.

n. The relationships between dose (rem) and probability of health effect probability for acute fatality in the IPPSS are piece-wise linear relationships fit to the following points:

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. . Lower Large Intestines Marrow Wall Lung Proba- Proba- Proba-bility bility bility of of of Dose Effect Dose Effect Dose Effect

, <320 0. <2000 0. <5000 0.

400 0.03 >5000 1.00 15,000 0.24 510 0.50 22,400 0.73

>615 1.00 >24,000 1.00 The corresponaing relationships for acute injury are:

Lower Large Intestines

! Whole Body Lung Wall Proba- Proba- Proba-bility bility bility of of of Dose Effect Dose Effect Dose Effect

<55 O. <3000 0. <1000 0.

150 0.30 >6000 1.00 >2500 1.0 280 0.80 l >370 1.0 1

The relationships between dose and thyroid cancer cases, and between dose and cancer fatalities from

{

, cancers other than thyroid cancers, are shown in the l

attached table.

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[ Document attached.]

o. The emergency response action assumptions described in Section 6.2.1.2 of the IPPSS were used for all release categories. Distances, delays, shielding factors, and the like are discussed in that section.

People beyond 10 miles were assumed to be exposed for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to radiation from material deposited on the ground. Thyroid prophylaxis was not assumed. Explicit modeling of the emergency response for special population groups is described in Section 6.2.1.2.

Population delay times were distributed to account for slower response of some groups,

p. Population data and their bases are given in IPPSS Section 6.2.1.2, and are included in Tables 6.2-5 and 6.2-6.

q. A complete discussion of the atmospheric dispersion model is provided in Section 6.1 of the IPPSS. The locations and characteristics of the regional meteorological stations are given in Tablo 6.2-1. Regional weather data were obtained from the National Climatic Center (NCC) in Asheville, N.C. The information contained on the data tapes received from the NCC is described in " Surface Observations," TDF14, National Oceanographic and Atmospheric Administration, Oct. 1975.

. . The accuracy of the instrumentation was considered to be a small contributor to the overall uncertainty of 4

the analysis, and therefore was not specified.

r. Statistics for an area within 2000 miles were l included in the analysis for mass balance purposes.

(Health effects beyond 500 miles are predicted to occur i

with low frequency.) A map of the EPZ showing evacuation routes is shown in Figure 6.2-6 of the IPPSS. A map of a larger area is shown in Figure 6.2-1.

s. As required by the Commission's regulations, the plume exposure EPZ is approximately 10 miles and the ingestion exposure EPZ is approximately 50 miles.

t. All evacuation assumptions are discussed in Section 6.2.1.2 of the IPPSS.

u. Age groups were not treated differently with respect to evacuation.

v. Sheltering assumptions are described in Section 6.2.1.2 of the IPPSS.

W. Documents used in producing sheltering assumptions were SAND 77-1725, SAND 77-1555, and SAND 79-2079.

x. It was assumed that persons who take shelter would do so before arrival of the plume, and that they I

would be relocated 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after arrival of the plume.

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y. Probabilistic analysis was utilized throughout the IPPSS. The probabilistic risk assessment method-ology for performing the specific portions of the IPPSS is described in Section 0.

The licensees object to t.he request for documents as excessive, burdensome, overly broad, and insuffic-iently specific to permit the licensees to determine which documents have been requested.

UCS/NYPIRG Interrogatory No. 4:

Provide all documents which contain and/or pertain to the Emergency Action Levels established for Indian Point, including but not limited to, any and all analyses, evalua-tions, and/or discussions of engineering or other analysis upon which these Emergency Action Levels dre based.

Response to Interrogatory No. 4:

See Preliminary Matters.

UCb/NYPIRG Interrogatory No. 5:

State whether the possibility and consequences of sabo-tage were considered in the Indian Point Probabilistic Safety Study (IPPSS) as a possible initiating condition or aggravating condition for the accidents evaluated therein, and:

a. it so, specify how the possibility and conse-quences of sabotage were treated and/or modelled in the IPPSS, and provide all documents which contain and/or pertain to the consideration of the possibility and

consequences of sabotage, whether specific to Indian Point or generic in nature, which were relied upon in performing the analyses contained in the IPPSS;

b. if not, specify precisely the bases for not considering the possibility and consequences of sabo-tage as a possible initiating condition or aggravating condition for accidents evaluated in the IPPSS, and provide all dodcuments [ sic] which contain and/or per-tain to this position, including but not limited to those documents which set forth in detail the basis for not so including sabotage.

Response to Interrogatory No. 5:

a. See Response to UCS/NYPIRG Interrogatory No. 5b.

b. The evaluation of sabotage was not within the scope of the IPPSS. The licensees believe that to include such an analysis in any safety study has the potential to jeopardize plant safety because it identifies potential sabotage scenarios.

UCS/NYPIRG Interrogatory No. 6:

State whether and specify how common-mode failures, common-cause failures, common-environment failures, and sys-tems interaction were treated and/or modelled in the IPPSS, and provide all documents which contain and/or pertain to these matters and how they were treated and/or modelled in the IPPSS.

Response to Interrogatory No. 6:

For a discussion of how these were treated, see Section 0.20 of the IPPSS.

The licensees object to the request for documents as excessive, burdensome, overly broad, and insufficiently specific to permit the licensees to determine which documents have been requested.

UCS/NYPIRG Interrogatory No. 7:

Specify any and all factors which could either initi-ate, aggravate, or mitigate accidents discussed, analyzed, or evaluated in the IPPSS which were not treated or modelled therein, discuss the basis for decisions not to treat or model each such factor, and provide all documents which con-tain and/or pertain to the decision to include and/or exclude each such factor.

Response to Interrogatory No. 7:

The licensees are not aware of any significant

" factors" which could initiate, aggravate, or mitigate accidents discussed, analyzed or evaluated in the IPPSS which were not treated or modeled therein.

During the course of the study, analysts routinely l applied their judgment to exclude events of insignificant contribution to the scenarios being considered on the basis of:

1) frequencies or probabilities too low to contribute; and 1

1 i

2) consequences of failure too insignificant to contribute.

For example, there are perhaps hundreds of " amplifiers" in the plant instrument and controls systems. Each ampli-fier may have four or five failure modes. Each mode could be caused by single or combined failures of hundreds of wires, transistors, and the like, each of which may have several failure modes. No such detailed modeling was done for all components, nor was it necessary, nor was the exclusion of such events " recorded," for the reason given above. The IPPSS records all such important eliminations.

In fact, probabilistic risk analysis methods provide the most rational framework for avoiding such unproductive detailed modeling.

UCS/NYPIRG Interrogatory No. 8:

Discuss and specify precisely how all uncertainties in meteorological conditions, populations, release categories, radionuclide inventories, containment failure mode probabil-ities, accident probabilities, and any other input factor to the IPPSS'were propogated [ sic) through the analyses and expressed in the final results of probabilities and conse-quences.

Response to Interrogatory No. 8:

The probabilistic risk assessment methodology by which uncertainties were propagated throughout the analyses in the

IPPSS are described in Section 0. For specific questiona, see applicable sections of the IPPSS.

UCS/NYPIRG Interrogatory No. 9:

Discuss and specify precisely now [ sic] the existence of, magnitude of, and relevant characteristics of unmoni-tored release pathways are determined for Indian Point under accident conditions, how this information is incorporated into dose projections and protective action decision-making, and provide all documents which contain and/or pertain to unmonitored release pathways at Indian Point.

Response to Interrogatory No. 9:

The accidents which are significant contributors to i

risk involve core melt followed by containment failure. The existence of unmonitored release pathways, if any, would be detected by information f rom the of fsite radiological monitoring stations and verified by field samples taken by survey teams. Plant conditions, meteorological data, and radiological data from the offsite monitoring stations and field samples would be used to determine the path and the magnitude of the release. This information would be conveyed to the appropriate state and local governmental agencies for protective actions.

UCS/NYPIRG In terrogatory No. 10:

State whether it is licensees' position that exposure to ionizing radiation cannot cause genetic effects in subse-quent generations, and provide all documents which contain

and/or pertain to the relationship assumed by the licensees to exist (or lack thereof) between exposure to radiation and the occurrence of genetic effects in subsequent generations.

Response to Interrogatory No. 10:

The licensees do not have sufficient information to answer this interrogatory because of the ongoing debate in the scientific community regarding this question.

UCS/NYPIRG Interrogatory No. 11:

Define, as used by the licensees in the IPPSS and licensees' testimony in this proceeding:

a. risk;

b. probability;

c. acute fatality;

d. acute injury (as this term relates to exposure to ionizing radiation);

e. thyroid nodule;

f. thyroid cancer;

g. cancer fatality;

h. morbidity;

i. man-rem; v

j. initial leukemia;

k. total whole body;

1. total leukemia;

m. evacuation;

n. relocation;

o. sheltering; i

/

33 -

p. interdiction;

q. impoundment;

r. decontamination; I

s. plume;

t. latent effect; ,

u. risk curve; and -

v. sensitivity study.

Response to Interrogatory No. Ils

a. See Sections 0.2, 0.3, and 0.5 of the IPPSS.

b. See,Section 0.4 of the IPPSS.

c.-h. See RSS Appendix'VI, Section 9.

i. Sum over entire population of dose received by each individual.

j. Leuxemia from early phase of exposure. See

+

Section 6.1.1.3 of the IPPSS.

k. Term not commonly used. " Total body" and "whole body" may be considered synonymous terms refer-ring to radiation dose that is approximately uniform over all organs and tissues.

1. Leukemia from both early and chronic phases of exposure. See Response to UCS/NYPIRG Interrogatory No.

Ilj.

m. Removal of persons from defined area in which a radiation hazard exists or is anticipated.

n. Removal of persons from defined area in which a radiation hazard exists.

o. Removal of persons to an area in which structural characteristics result in reduction of radiation doses below those which might be expected from normal activities given a radiation source.

p. Action which cuts off or precludes radiation dose accumulation.

q. Prevention of distribution of materials cont'aning radioactive material.

r. Removal of radioactive material.

s. Volume of air containing radioactive material af ter release to atmosphere.

t. Effect which may be caused by radiation exposure but which develops only after the passage of some time fol' awing the exposure.

u. Plot of frequency versus consequence.

v. Study in which selected parameters are varied to determine impact upon results.

UCS/NYPIRG Interrogatory No. 12:

Specify the dose levels and/or man-rem totals at which the following health effects consequences are assumed by the

, CRACIT model in the IPPSS to occur:

i

a. acute injury;

b. acute fatality;

c. latent effect;

d. leukemia;

, , e. thyroid nodule; and

f. thyroid cancer.

Response to Interrogatory No. 12:

The relationships used between dose level and health ef fects are given in the Response to UCS/NYPIRG Interroga-tory No. 3n.

UCS/NYPIkG Interrogatory No. 13:

Specify the criteria adopted in the IPPSS for decon-tamination (i.e., how it is determined what areas require decontamination in order to permit restricted and/or unre-stricted access by the general public following an acciden-tal release of radioactivity to the environment from Indian Point).

Response to Interrogatory No. 13:

The criteria for decontamination are determined by evaluating the chronic whole body dose in affected areas and comparing them with whole body dose limits. The whole body dose is computed at zero, one, two, and 10 years, and divided by the dose limit. An interpolation is performed, if necessary, to determine the time at which the area would be habitable without the use of decontamination. A maximum decontamination factor of 20 is used, based upon the results in Appendix VI to the RSS, Appendix K. The time at which the area will become habitable with decontamination is determined as above. If the time at which decontamination would be effective is determined to be greater than 10

. . . _ _ . ___~ _

. . years, the area is considered permanently interdicted; otherwise, it is assumed that the area will be decontaminated and will be considered habitable at the computed time.

UCS/NYPIRG Interrogatory No. 14:

State whether the CRACIT code has the capability to calculate the maximum distance to which the following health i

effects consequences will occur as a result of a specified release of radioactivity:

a. acute injury;

b. acute fatality;

c. latent effect;

d. cancer;

e. thyroid cancer;

f. thyroid nodule; and 9 leukemia.

Response to Interrogatory No. 14:

In the CRACIT runs used for the IPPSS, the maximum distance was determined for acute injury (a) and acute -

i fatality (b), but not for the other consequences. The version of the CRACIT Code used in the IPPSS had no capability for determining maximum distance for the other

( 'onsequences.

l l

UCS/NYPIRG Interrogatory No. 15:

If the response to any portion of Interrogatory No. 14 is yes, state whether and which CRACIT runs which formed the I

l

basis for the results in the IPPSS calculated such results, provide all such results together with an identification of the release category, meteorological assumptions, and public response assumptions associated with each such result.

Response to Interrogatory No. 15:

The CRACIT program does not print the results requested in UCS/NYPIRG Interrogatory Nos. 14a and 14b, but produces a file which contains them. This file is processed in a post-processor program called FINSUM to produce conditional risk c.rves for those results.

I See Preliminary Matters.

UCS/NYPIRG Interrogatory No. 16:

If the response to any portion of Interrogatory No, 14 is yes, provide all cumulative complementary distribution functions, risk curves, and/or other probabilistic expres-sions of the probability of causing the specified health eftectors [ sic] consequences versus distance.

Response to Interrogatory No. 16:

l See Response to UCS/NYPIRG Interrogatory No. 15.

UCb/NYPIRG Interrogatory No. 17:

! State what the licensees believe to be the relationship (if any) between the population density in the region sur-rounding Indian Point and the magnitude of the consequences resulting from accidental releases of radioactivity from Indian Point, specify the basis for this relationship, and provide all documents which contain and/or pertain to this i

l

relationship or the lack thereof of any such relationship.

Response to Interrogatory No. 17:

Given the Indian Point plant, or any other nuclear plant, the higher the surrounding population density, the higher will be the magnitude of the consequences in the event of an assumed accidental release of radioactivity.

UCS/NYPIRG Interrogatory No. 18:

State whether the licensees have compared the risk of continued operation of Indian Point to the risk of continued operation of any other nuclear powerplant in the United States, and for each such plant so compared, specify the name of the plant, the method by which the risk posed by that plant was calculated, the uncertainty in the risk and the method by which it was calculated, how the method by which the risk and uncertainty in the risk were calculated for these plants differs from the method by which these parameters were calculated for Indian Point, the population surrounding each such plant, the types of accidents and release categories considered in the assessment of the risk posed by continued operation of each such plant, and state the basis for comparison of the risk posed by continued operation of each such plant with the risk posed by the con-tinued operation of Indian Point.

Response to Interrogatory No. 18:

Comparisons of the risk of continued operation of Indian Point with the results of probabilistic risk analysis

studies for various other plants are being made. The plants presently being compared to Indian Point are Surry, Biblis, Oconee, Sequoyah, Grand Gulf, Peach Bottom, Big Rock, and Limerick. Information regarding,the methods, assumptions, and data used for each plant is being presented in each plant's report, which is public information.

UCS/NYPIRG Interrogatory No. l_9 :

Provide all documents which contain and/or pertain to the analyses described in Interrogatory No. 18.

Response to Interrogatory No. 19:

See Response to UCb/NYPIRG Interrogatory No. 18.

UCS/NYPIMG Interrogatory No. 20:

State whether the licensees have evaluated the risk posed by accidents at non-huclear electrical generating facilities, and for each such evaluation, specify the type of facility, the size of the facility (in megawatts elec-trical generating capacity), the location of any such facility, the method by which the risk posed by accidents at 1 each such facility was calculated, the uncertainty in the results of the risk calculations for each such facility, the ,

j types of accidents occurring at such facilities and their

[

probability of occurrence, the types of consequences result-ing from accidents at such facilities, the magnitude of con-sequences resulting from accidents at such facilities, and the distributica of such consequences with distance from such facilities.

t __

Response to Interrogatory No. 20:

No such evaluations have been made.

UCS/HYPIRG Interrogatory No. 21:

Provide all documents which contain and/or pertain to the evaluations and analyses described in Interrogatory No.

20.

Response to Interrogatory No. 21:

See Response to UCS/NYPIRG Interrogatory No. 20.

UCS/NYPIRG Interrogatory No. 22:

Figure 6.2-1 in the IPPSS depicts graphically the

" Indian Point Meteorological Regions" surrounding Indian Point. Specify the sources of all meteorological data from Indian Point Meteorological Regions 2 through and including (in sequence) 14.

Response to Interrogatory No. 22:

See Response to UCS/NYPIRG Interrogatory No. 3q.

UCS/NYPIRG Interrogatory No. 23:

l Identify the sources of the population totals and evac-uation vectors contained in Tables 6.2-5, 6.2-6, 6.2-7, 6.2-8, and 6.2-9 in the IPPSS, and provide all documents which contain and/or pertain to these data and the means by which they were derived or generated.

Response to Interrogatory No. 23:

The information was provided by Parsons Brinckerhoff Quade & Doug las , Inc. The derivation of the data is discussed in Section 6.2 of the IPPSS.

1 l

l l

UCS/NYPIRG Interrogatory No. 24:

Identify the sources of the " Evacuation Data" contained in Tables 6.2-13A, 6.2-13B, 6.2-13C, 6.2-13D, and 6.2-13E in the IPPSS, and provide all documents which contain and/or pertain to these data and the means by which they were derived or generated.

Response to Interrogatory No. 24:

See Response to UCS/NYPIRG Interrogatory No. 23.

UCS/NYPIRG Interrogatory No. 25:

Identify the sources of the " Time of Release," " Dura-tion of Release," " Warning Time," and " Energy Release" data contained in Table 6.2-16 in the IPPSS, and provide all documents which contain and/or pertain to these data and the means by which they were derived or generated.

Response to Interrogatory No. 25:

Values for the " Time of Release," " Duration of Release," and " Associated Energy of Release" are contained in Tables 5.4-3 and 6.2-16 of the IPPSS. The ration :le for the values of the parameters is discussed in Sections 5.4.2-1 and 5.4.2-2. In general, the values were derived from the

, values contained in the RSS for the same or similar release categories. With respect to " Warning Time," the parameter is defined as the interval between public awareness of the impending core melt and release of radioactivity to the a tmosphe re . See Section 6.2.2-1. With two exceptions, it was assumed that awareness of impending core melt would

occur within one hour of accident initiation. Warning time is calculated by subtracting the period for awareness (one hour) from the " Time of Release." One exception is release category Z-lu. This category results from a seismic event which causes both a loss of coolant accident and containment failure. For such an event, awareness is assumed to occur in 1/2 hour, and warning time of 1/2 hour is calculated.

The second exception is sequences for which " Time of Release" is 2-1/2 hours. For such sequences a longer awareness time of 1-1/2 hours was taken.

UCS/NYPIRG Interrogatory No. 26:

Specify how the wind roses on pages 6.2-57 and 6.2-58 of the IPPSS are interpreted, i.e., state whether the lines represent the direction from which the wind is blowing or whether the lines represent the direction in which the wind is blowing.

Response to Interrogatory No. 26:

The lines on the wind rose refer to the direction from which the wind was blowing in conformance with standard meteorological practice.

UCS/NYPIRG, Interrogatory No. 27:

Provide copies of the wind roses on pages 6.2-57 and 6.2-58 with the compass headings noted on them, or specify precisely where the compass heading of North is located on these wind roses.

Response to Interrogatory No. 27:

The compass heading for North is at the top of the page.

UCS/NYPIRG Interrogatory No. 28:

State whether the CRACIT model incorporates an assump-tion that releases of radioactivity are distributed evenly between day and night and evenly between the 12 months of the year, and, if not, state what assumptions are incor-porated into the CRACIT model regarding the start times of releases and day and night and the months of the year, and, in either case, provide all documents which contain and/or pertain to these relationships and the means by which they were derived or generated.

Response to Interrogatory No. 28:

A list of accident start times was distributed evenly between day and night and between the 12 months of the year. The list was randomly generated using TIMES, a subroutine in the CRAC code written for that purpose.

UCS/NYPIRG Interrogatory No. 29:

State whether the CRACIT model can compute results for up to four " multi-phased" [ sic) releases, define " multi-phase" releases, and specify how the time of release, l duration of release, height of release, magnitude and l

isotopic makeup of the release, heat content of the release, and warning time for the release is determined for each such phase.

- 44 _

Response to Interrogatory No. 29:

The CRACIT Code can compute those results and was used in that way for release category 2RW for the IPPSS. The computer program used to calculate release fractions and associated parameters, CORRAL, yields cumulative release fractions versus time following the accident. These data were grouped into four phases for release category 2RW.

The CRACIT Code can compute consequences for up to four fission product release periods. The term " Multi-Phase Release" is generally defined in Appendix 5.8.4. The time of release, duration, magnitude, and isotopic makeup for each of the multi-phase releases were derived from CORRAL calculations as described in Section 5.8.4-1. Energy content for each phase of the release is identified in Table 5.8.4-3. Height of release of 10 meters was used for all phases as specified in Section 5.4.2. The approach used to obtain the time of release and warning time is discussed in the Response to UCS/NYPIhG Interrogatory No. 26.

I UCS/NYPIRG Interrogatory No. 30:

Provide copies of all documents which contain and/or pertain to a description of the CRACIT model, the, user's manual or guide for the CRACIT model, the computer pro-gram (s) used in the CRACIT model, evaluations, criticisms, and assessments of the CRACIT model, and descriptions of the differences between CRACTIT [ sic] and the CRAC and CRAC2 models.

Response to Interrogatory No. 30:

The CRACIT program is described in detail in Section 6 of the IPPSS. No user's manual or guide has been issued.

The licensees are not aware of any criticisms, reports, assessments or evaluations of the CRACIT model. It was used in the internal benchmark exercise sponsored by the Organization for Economic Cooperation and Development (OECD) for which the final report has not been issued. A detailed discussion of the differences between CRAC and CRACIT is provided in Section 6.1.2 of the IPPSS.

Detailed comparisons with CRAC 2 have not been made; however, it is the licensees' understanding that there are major differences in the technique used to sample meteorological scenarios.

UCS/NYPIRG Interrogatory No. 31:

Provide all final results output from CRACIT model runs which were used in the IPPSS, including total consequences, i distance versus magnitude data, magnitude versus probability l

data, release categories and leakage fractions assumed, I

source term assumptions, public response assumptions, release category probability assumptions, and an identifica- ,

j tion of which Indian Point reactor is analyzed for each such run.

Response to Interrogatory No. 31:

The final results output from the CRACIT runs used in the IPPSS are included in Tables 8.5.2-4a to e, 8.5.8-4a to I

e, 8.5.8-Sa to e, and 8.5.8-6a to e of the IPPSS. These are conditional risk curves for consequences (magnitude versus probability). No distance versus magnitude results were computed in the IPPSS runs. The input assumptions are described in Section 6. The higher power level of the Indian Point Unit 3 reactor was used for all consequence runs.

UCS/NYPIRG Interrogatory No. 32:

Specify the basis for the assumption as expressed on page 6.1-11 of the IPPSS that an individual's exposure to a cloud of radioactivity released from Indian Point would last only about an hour, and provide all documents which contain and/or pertain to this assumption and the technical basis therefore [ sic].

Response to Interrogatory No. 32:

The duration of exposure to the cloud or plume is the length ot time it takes for the plume to pass overhead.

Exposure to the plume stops after the plume has passed. The plume exposure period is approximately equal to the duration of release. For the most severe release categories, the release durations are 1/2 hour to one hour. Longer durations would provide more time for response. See Table 6.2-16 of the IPPSS, which shows release durations.

UCS/NYPIRG Interrogatory No. 33:

State whether the CRACIT model assumes that chronic radiation exposure for the general public will occur for

only one season for crops, and if so, specify the basis for that assumption, and provide all documents which contain and/or pertain to that assumptions [ sic] and the basis therefore [ sic).

Response to Interrogatory No. 33:

The CRACIT model for chronic radiation exposure is identical to the CRAC model, and is described in Appendix 6, Section 8 of the RSS. Chronic exposure from crops is not computed for seasons beyond the first season unless interdiction criteria are exceeded.

UCS/NYPIRG Interrogatory No. 34:

State whether the ORIGEN computer code was used to cal-culate the radionuclide inventory at the time of the accidents discussed, evaluated, or assessed in the IPPSS, and if so, specify all uncertainties in the results of the use of the ORIGEN code, whether the uncertainty will result in a conservative or non-conservative result and the mag-nitude of that result, and the impact such uncertainties have on the health effects consequences estimated using the CRACIT model for Indian Point.

l l Response to Interrogatory No. 34:

1 The ORIGEN code was used to produce inventories for a 3200 MWth reactor in WASH-1400. See WASH-1400, Appendix VI, Section 3.2. These values were scaled to the power level of

( Indian Point Unit 3 -- 3025 MWth. The uncertainties in the inventory estimate or curie content contribute negligibly to i

uncertainties in the consequence analysis. However, the ORIGEN calculation is based upon an end-of-cycle inventory which results in an overestimate of the Cs-137 inventory by a factor of approximately 2 (relative to the average over the cycle). This bias was accounted for in the assessment of uncertainties ir, the estimate of cancer fatalities and man-rem for which Cs-137 is a dominant contributor.

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As to Answers:

CONSOLI T' ISON CO N NEW Y K, INC.

By i Pl

• CA-7ts4 Richara P. Remshaw '

Project Manager - Indian Point Hearings POWER AUTHORITY OF THE STATE OF NEW YORK By A (fie'r'schel Specter /

Project Manager - Indian Point Hearings As to Objections:

CONSOLIDAT' EDISO OMPANY F NEW YORK, INC.

By /X?ni<J w Brent L. brandenburg Assistant General Counse f

POWER AUTHORITY OF ThE STATE OF NEW YORK MORG 4 ASSO ,C ' E D 1

l By N Plttri F. Colarulli X ,

Attorney for Power Authority of the State of New York l

l l

l 1

f

VERIFICATION STATE OF NEW YORK )

SS.:

l COUNTY OF NEW YORK )

I, RICHARD P. REMSHAW, being duly sworn, depose and say:

That I am the Manager, Indian Point Hearings for Consolidated Edison Company of New York, Inc., licensee of the Indian Point Nuclear Generating Station, Unit No. 2; that I am authorized to make this verification on behalf of said corporation; and that the foregoing answers to interrogatories were prepared under my direction and supervision and are true and correct to the best of my knowledge, information and belief.

l V/WA.

• h1 CHARD P. hEMShAW '

Sworn to before me this Q th cay of November, 1962

)

hotary Public v

,\!y Comminion Expires June 14, 1937 l l

l 1

VERIFICATION STATE OF NEW YORK )

SS.:

COUNTY OF NEW YORK )

I, HERSCHEL SPECTER, being duly sworn, depose and say:

That I am the Manager, Indian Point 3 Hearings, Technical Support for Power Authority of the State of New York, licensee of the Indian Point 3 Nuclear Power Plant; that I am authorized to make this verification on behalf of said corporation; and that the foregoing answers to interrogatories were prepared under my direction and supervision and are true and correct to the best of my knowledge, information and belief.

Abk- A wY

/ HERSCHEL SPECTER /

Sworn to before me this th day of November, 1982 Notary Public V My Comminion Expires Jue 1+,1987 l

s ... - -_ r -- _ _

. . o Respectfully submitted, ll - a Brent L. Brandenburg A-

~ Charles Morgan, Jr.

Paul F. Colarulli Joseph J. Levin, Jr.

! CONSOLIDATED EDISON C PANY MORGAN ASSOCIATES, CHARTERED OF NEW YORK, INC. 1899 L Street, N.W.

Licensee of Indian Point Washington, D.C. 20036 Unit 2 (202) 466-7000 4 Irving Place New York, New York 10003 Stephen L. Baum (212) 460-4600 General Counsel Charles M. Pratt Assistant General Counsel POWER AUTHORITY OF THE STATE OF NEW YORK Licensee of Indian Point Unit 3 10 Columbus Circle New York, New York 10019 (212) 397-6200 Bernard D. Fischman Michael Curley l

Richard F. Czaja David H. Pikus l SHEA & GOULD 330 Madison Avenue New York, New York 10017 (212) 370-8000 l

Dated: November 19, 1982 l

l l

, . .o UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD Before Administrative Judges:

James P. Gleason, Chairman Frederick J. Shon Dr. Oscar H. Paris

)

In the Matter of )

)

CONSOLIDATED EDISON COMPANY OF NEW YORK, INC. ) Docket Nos.

(Indian Point, Unit No. 2) ) 50-247 SP

) 50-286 SP POWER AUTHORITY OF THE STATE OF NEW YORK )

(Indian Point, Unit No. 3) ) Nov. 19, 1982

)

CERTIFICATE OF SERVICE I hereby certify that on the 19th day of November, 1982, I caused a copy of the Licensees' Responses to UCS/NYPIRG First Set of Interrogatories and Document Requests on Board Questions One, Two, and Five to be served by first class mail, postage prepaid on the following:

i l

. . . , _ _ y,_, , - . _ _ _ .- ,. ,

James P. Gleason, Chairman Charles M. Pratt, Esq.

Administrative Judge Stephe'n L. Baum, Esq.

Atomic Safety and Licensing Board Power Authority of the 513 Gilmoure Drive State of New York Silver Spring, Maryland 20901 10 Columbus Circle New York, New York 10019 Mr. Frederick J. Shon Administrative Judge Janice Moore, Esq.

Atomic Safety and Licensing Board Counsel for NRC Staff U.S. Nuclear Regulatory Office of the Executive Commission Legal Director Washington, 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 Service Branch Office of the Secretary Ellyn R. Weiss, Esq.

U.S. Nuclear Regulatory Commission William S. Jordan, III, Esq.

Washington, D.C. 20555 Harmon and Weiss 1725 I Street, N.W., Suite 506 Joan Holt, Project Director Washington, 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.

New York, New York 10007 P.O. Box 488 White Plains, New York 10602 Jeffrey M. Blum, Esq.

New York University Law School Alan La tman, Esq.

423 Vanderbilt Hall 44 Sunset Drive 40 Washington Square South Croton-On-Hudson, New York 10520 New York, New York 10012 Ezra I. Bialik, Esq.

Charles J. Maikish, 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 Alfred B. Del Bello Westchester County Executive Westchester County 148 Martine Avenue White Plains, New York 10601 Andrew S. Roffe, Esq.

New York State Assembly Albany, New York 12248

Marc L. Parris, 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 Pat Posner, 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 i 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.

330 Madison Avenue Geoffrey Cobb Ryan New York, New York 10017 Conservation Committee Chairman Director, New York City Amanda Potterfield, Esq.

Audubon Society Johnson & George 71 West 23rd Street, Suite 1828 528 Iowa Avenue New York, New York 10010 Iowa City, Iowa 52240 Lorna Salzman Ruthanne G. Miller, Esq.

Mid-Atlantic Representative Atomic Safety and Friends of the Earth, Inc. Licensing Board Panel 208 West 13th Street U.S. Nuclear Regulatory l New York, New York 10011 Commission Washington, D.C. 20555 l Stanley B. Klimberg, Esq.

General Counsel New York State Energy Office 2 Rockefeller State Plaza Albany, New York 12223 t

. e. o 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 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 Stewart M. Glass Regional Counsel Room 1349 Federal Emergency Management Agency 26 Federal Plaza New York, New York 10278 i

Melvin Goldberg Statt Attorney New York Public Interest Research Group 9 Murray Street New York, New York 10007 Jonathan L. Levine, Esq.

P. O. Box 280 New City, New York 10958 l 2\ '

aa Paul F. Cdisrulli

NAME DENNIS C. BLEY EDUCATION Ph.D., Nuclear Reactor Engineering, Massachusetts Institute of Technology ,1979.

Courses 1972-1974.

in nuclear engineering and computer science, Cornell University, U.S. Navy Nuclear Power School,1968.

University of Cincinnati, B.S.E.E. ,1967.

Courses in Mathematics and Physics, Centre College of Kentucky, 1961-1963.

PROFESSIONAL EXPERIENCE General Summary A consultant at Pickard, Lowe & Garrick, Inc. ,1979-present. Technical analysis of power plant availability and risk. Cost-benefit analysis of power plant system changes. Preparation of technical reports, expert testimony, and proposals. Supervision of the technical quality of PLG reports and direction of some PLG projects. Instructor at availability, risk, and decision analysis courses offered by PLG. Oyster Creek Probabilistic Risk Assessnent (OPSA). Assisted in the completion and review of this complete risk assessment of an operating BWR performed for Jersey Central Power & Light. Work Order Scheduling System (WOSS).

Assisted in developing the San Onofre 2 and 3 plant-model for a computer based work arder prioritizing, scheduling, and record keeping system for Southern California Edision Company. Steam Turbine Diagnostics Cost-Benefit Analysis. Developed and applied a procedure for evaluating diagnostic alternatives for EPRI. Reliability Analysis of Diablo Canyon Auxiliary Feedwater System for Pacific Gas & Electric. Midland Plant Auxiliary Feedwater System Reliability Analysis for Consumers Power.

Technical Review of the " Office of Emergency Services Recommended Emergency Planning Zone Considerations..." for Southern California Edison. Prioriti:ation of NRC Action Plan for NSAC. Development of a methodology and participation in an AIF workshop to apply it for EPRI/NSAC. Zion and Indian Point Probabilistic Safety Studies.

Methods development, systems analysis, and plant modeling. Other PRAs--LaSalle, Browns Ferry, Midland, Pilgrim 1, and Oconee.

On USS Enterprise, Reactor Training Assistant, 5 months, 1971.

Responsible for technical training of approximately 400 nuclear trained officers and men prior to annual safeguards examination. Propulsion Plant Station Officer, 9 months, 1970-1971. Responsible for maintenance and operation of one propulsion plant (two reactors, eight steam generators, and associated equipment) during power range testing of new reactors and during deployment. Approximately 50 enlisted personnel were assigned to the plant. Shif t Propulsion Plant Watch Officer,15 months, 1969-1970. Supervised a crew of about 20 navy enlisted operators and many shipyard workers on 8-hour shift rotation conducting maintenance 24 0457P072082

BLEY - 2 and testing in one propulsion plant during refueling-overhaul. Shipboard qualifications: Propulsion Duty Officer, responsible for all propulsion equipnent during absence of Reactor Officer and Engineer Officer.

Engineering Officer of the Watch, operational watch in Central Control, responsible for all propulsion and engineering equipment and watch standers. Propulsion Plant Watch Officer, operational watch in one propulsion plant, directed and responsible for all operations in the plant.

At Cincinnati Bell, Plant staff assistant, 4 months,1967. Worked in central office and transmission group supplying technical assistance to the line organization. Cocoerative trainee, 3 years,1964-1967, work-study program with alternate three month periods at the University of Cincinnati.

Chronological Summary

• 1979-Present Consultant, Pickard, Lowe and Garrick, Inc.

1974-1979 Massachusetts Institute of Technology.

Research assistant for Department of Energy LWR Assessment Project. Teaching assistant in engineering of nuclear reactors.

Summer 1976 Northeast Utilities.

Engineer: economy studies, plant startup, analysis of physics tests.

1967-1974 U.S. Naval Reserve, active duty.

Instructor of naval science, Cornell University, 1971-1974; Reactor Department of USS Enterprise, deployment and refueling-overhaul, 1969-1971; Nuclear Power training program and Officer Candidate School, 1967-1969.

1964-1967 ' Cincinnati Bell .

Plant staff assistant and work-study program trainee.

MEMBERSHIPS, LICENSES, AND HONORS The Society for Risk Assessment.

Institute of Electrical and Electronics Engineers.

American Nuclear Society.

Anerican Association for the Advancement of Science.

The New York Academy of Sciences.

U.S. Naval Reserve, Commander.

l Registered Nuclear Engineer, State of California.

25 l Oa87P072na?

9 BLEY - 3 Sigma Xi (national science honors society),1976.

Sherman R. Knapp Fellowship (Northeast Utilities), 1975-1976.

51oan Research Traineeship, 1974-1975.

Eta Kappa Nu (national electrical engineering honors society),1967.

REPORTS AND PUBLICATIONS "Seabrook Probabilistic Safety Assessment," Public Service Company of New Hampshire, to be published in 1983.

Pickard, Lowe and Garrick, Inc., " Midland Probabilistic Risk Assessment,"

Consumers Power Company, to be published in 1982.

Oconee Probabilistic Risk Assessment," a joint effort of the Nuclear Safety be Analysis published Center, Duke Power, and other participating utilities, to in 1982. .

Tennessee Valley Authority and Pickard, Lowe and Garrick, Inc., " Browns Ferry Probabilist'c Risk Assessment," to be published in 1982.

4 Apostolakis, G. , M. Kazarians, and D. C. Bley, "A Methodology for Assessing the Risk from Cable Fires," accepted for publication in Nuclear Sa fe ty, 1982.

Kaplan, S., H. F. Perla, and D. C. Bley, "A Methodology for Seismic Safety Analysis of Nuclear Power Plants," proposed presentation at the International Meeting on Thermal Nuclear Reactor Safety, Chicago, Illinois, August 29-September 2,1982.

Bley, D. C. , S. Kaplan, and B. J. Garrick, " Assembling and Decomposing l

' PRA Results: A Matrix FormaTism," proposed presentation at the -

International Meeting on Thermal Nuclear Reactor Safety, Chicago, Illinois, August 29-September 2,1982.

Garrick, B. J. , S. Kaplan, and D. C. Bley, "Recent Advances in Probabilistic Risk Assessment," prepared for the MIL Nuclear Power Reactor Safety Course, Cambridge, Massachusetts, July 19, 1982.

l Fleming, K. N., S. Kaplan, and B. J. Garrick, "Seabrook Probabilistic Safety Assessment Management Plan,"PLG-0239, June 1982.

Garrick, B. J. , " Lessons Learned From First Generation Nuclear Plant Probabilistic Risk Assessments," to be presented at the Workshop on Low-Probability /High-Consequence Risk Analysis, Arlington, Virginia, June 15-17, 1982.

26 Oc87P072082

[ _ _ _ __ , - - - - - -

BLEY - 4 Garrick, B. J., S. Kaplan, D. C. Iden, E. B. Cleveland, H. F. Perla, j

D. C. Bley, D. W. Stillwell, H. Y. Schneider, and 3. Apostolakis, " Power /

Plant Availability. Engineering: Methods of Analysis, Program Planning, and Applications," EPRI NP-2168, PLG-0163, May 1982. ,

Bley, D. C., and R. J. Mulvihill, " Comments on Evaluation of Availa'bility Improvement Options for Mass Landing Units 6 and 7," PLG-0226, March 1982. ,

Stillwell, D. W. , G. Apostolakis, D. C. Bl ey,- P. H. Raabe, R. J. Mulvihill, S. Kaplan, and B. J. GarricK, "EEI Availability Handbook," PLG-0218, January 1982. - 0 Bley, D. C. , L. G. H. Sarmanian, and D. W. Stillwell, " Reliability Analysis of Safety Injection System Modification, San Onofre Nuclear Generating Station - Unit 1," PLG-0206, October 1931.

" Zion Probabilistic Safety Study," Comonwealth Edison Company, September 1981. 'f Buttemer, D.' R.', " Analysis of Postulated Accidents During Low Power Testing at,the San Onofre Nuclear Generating Station--Unit 2," PLG-0199

September 1981. ,

Bley, D. C. , D. W. Stillwell, and R. R. Fray, '" Reliability Analysis of Diablo Canyon Auxiliary Feedwater System," pqesented at the Tenth Biennial Topical Conference on Reactor Operating Experience, Cleveland, Ohio, August 17-19, 1981. ,

! Garrick,'B'. J. , and D. C. Bley, "Lessans Learned from Current PRAs,"

presented to the ACRS Subcommittee on Reliability and Probabilistic Risk

! Assessment, Los Angeles, California, July 28, 1981.

1

( Kaplan, S. , G. Aposto1akis, B. J. Garrick, D. C. Brey, and K. Woodard, l " Methodology for Probabilistic Risk Assessment of Nuclear Power Plants,"

j draf t version of a book in preparation, PLG-0209, June 1981.

Perla, H. F. , " Project Plan: Probabilistic Risk Assessment, Midland Nuclear Power Plant," PLG-0150, May 1981. ',

l Bl ey, D. C. , C. L. Cate, D. W. Stillwell, and B. J. Garrick, " Midland Plant Auxiliary Feedwater System Reliability Analysis' Synopsis," I l PLG-0166, March 1981.

Pickard, Lowe and Garrick, Inc., "A Metnodology to Quantify Uncertainty of Cost of Electricity for Alternate Designs of (Combustion) Turbine

Combined Cycle Plants," PLG-0162, March 1981.

, t 1 i

! p

.  ; 27

0487P072082 ,

i  ? i

BLEY - 5 I

's Garrick, B.. J. , S. Ahmed, and D. C. Bley, "A Methodology for Evaluating

, the Costs and Benefits of Power Plant Diagnostic Techniques," submitted for presentation at the Ninth Turoomachinery Symposium, Houston, Texas, December 9-11, 1980.

Pickard, Lowe and Garrick, Inc., " Seminar: Probabilistic Risk Assessment of Nuclear Power Plants," PLG-0154, November 1980.

Pickard, Lowe and Garrick, Inc., " Project Plan: Probabilistic Risk Assessment, Browns Ferry Nuclear Plant Unit 1," PLG-0149, October 1980.

Garrick, B. J. , S. Kaplan, D. C. Iden, E. B. Cl eveland, H. F. Perla, D. C. Bley, and D. W. Stillwell, " Power Plant Availability Engineering, Methods of Analysis - Program Planning - Applications," 2 Vols.,

PLG-0148. October 1980.

Bl ey, D. C. , C. L. Cate, D. W. Stillwell, and B. J. Garr1ck, " Midland Plant Auxiliary Feedwater System Reliability Analysis," PLG-0147, October 1980.

Bl ey, D. C. , D. M. Wheeler, C. L. Cate, D. W. Stillwell, and B. J. Garrick, " Reliability Analysis of Diablo Canyon Auxiliary Feedwater Sy stem," PLG-0140, September 1980.

Garrick, B. J. , et al, " Project Plan: Oconee Probabilistic Risk Assessment," PLG-0138, August 1980. '

Garrick, B. J., D. M. Wheeler, E. B. Cleveland, D. C. Bley, L. H. Reichers, and C. B. Morrison, " Operating Experience of Large U.S. Steam Turbine-Generators; Volume 1 - Data, Volume 2 - Utility Directory," PLG-0134, June 1980.

Garrick, B. J. , S. Kapl an, and D. C. Bl ey, " Seminar: Power Plant Probabilistic Risk Assessment and Reliability," PLG-0127, May 1980.

Garrick, B. J., and S. Kaplan, "0yster Creek Probabilistic Safety Anal-ysis (OPSA)," presented at the ANS-ENS Topical Meeting on Thermal Reactor Safety, Knoxville, Tennessee, April 8-11, 1980.

Garrick, B. J. , S. Kaplan, G. E. Apostolakis, D. C. Bley, and T. E. Potter, " Seminar: Probabilistic Risk Assessment as Applied to Nuclear Power Plants," PLG-0124, March 1980.

Garrick, B. J. , S. Ahmed, and D. C. Bley, "A Methodology for Evaluating the Costs and Benefits of Power Plant Diagnostic Techniques," PLG-0118, January 1980.

28 0487P0720S2

BLEY - 6 Kaplan, S. , B. J. Garrick, and D. C. Bl ey, " Notes on Risk, Probability, and Decision," PLG-0113, November 1979.

Bl ey, D. C. , C. L. Cate, D. C. Iden, B. J. Garrick, and J. M. Hudson,

" Seismic Safety Margins Research Program (Phase I), Project VII - Systems Analysis," PLG-0110, September 1979.

Cate, C. L. , and B. J. Garrick, "W-501 Combustion Turbine Starting Reliability Analysis," PLG-0103, June 1979.

Pickard, Lowe and Garrick, Inc., " Plant Availability Program Specifica-tion, San Onofre Nuclear Generating Station," March 1979.

Pickard, Lowe and Garrick, Inc., " Work Order Scheduling System, Design Specification," March 1979.

e l

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29 0487P072032 -

L

. o Resume - Bernard L. Cohen Birth: Pittsburgh, PA June 14,1924 Education: B.S. , Case-Western Reserve Univ. , - 1944 M.S., University of Pittsburgh, - 1947

D.Sc. , Carnegie-Mellon Univ. , - 1950 Employment:

Permanent positions 1950-58: Oak Ridge National Labora, tory

- Group Leader for Cyclotron Research 1958 - present: University.of Pittsburgh

- Professor of Physics

- Adjunct Prof. of Chemical and Petroleum Engineering

- Director, Scaife Nuclear Laboratory (1965-78)

Temporary positions (1-9 months) 1959-60 General Atomic Co., La Jolla, CA .

1962 Institute for Defense Analysis, Washington, DC 1965 Brookhaven National Laboratory 1969 Los Alamos Scientific Laboratory 1971 Stanford University .

1974-75 Institute for Energy Analysis, Oak Ridge, Tennessee 1975 Electric Power Research Institt te, Palo Alto, CA 1978-79 Argonne National Laboratory .

Offices - Awards:

Chairman, Am. Physical Society Div. of Nuclear Physics, 1974-75 Chairman, Am. Nuclear Society Div. of Environmental Sciences, 1980-81 Landauer Award (Health Physics) - 1980 Am. Physical Society Bonner Prize - 1981 Books authored:

Heart of the Atom, Doubleday (1967)

Concepts of Nuclear Physics, McGrcw-Hill (1971)

Nuclear Science and Society, Doubleday (1974)

Publications: ,

about 200 articles in Scientific journals about 30 articles in other journals about 120 invited papers at Scientific meetings about 300 invited visiting lectures at Universities, laboratories, etc. -

Research Areas:

Nuclear physics Health effects of radiation Environmental impacts of energy production Societal risks and risk aversion

Recent Publications - B. L. Cohen Impacts of the Nuclear Energy Industry on Human Health and Safety, B.L. ,

Cohen, The American Scientist. Sept.-Oct.1976, p. 550.

Conclusions of the BEIR and UNSCEAR Reports on Radiation Effects per Man-rem, B.L. Cohen, Health Physics 3_0,,

0 351 (1976).

Search for the Double-Direct (p,2p) Reaction at 17 MeV, L.Shabason, B.L.

Cohen, and T. Congedo, Phys. Rev. C15, 260 (1977).

Compound Nucleus (a,p) and (p p') Reactions on Odd-A Nuclei in the Ni Region, K.C. Chan, L. Shabason, B.L. Cohen, J. Alzona and T. Congedo, Phys. Rev. CIS, 1698 (1977).

High Level Waste from Light Water Reactors, B.L. Cohen, Rev. Mod. Phys.

49, 1 (1977).

Hazards from Plutonium Toxicity, B.L. Cohen, Health Physics 32, 359 (1977).

Body Weight as an Application of Energy Conservation, B.L. Cohen, Am.

Jour. Physics 45, 867 (1977).

Methods for Calculating Population Dose from Atmospheric Dispersion of Radioactivity, B.L. Cohen, H. N. Jow, and I.S. Lee, Health Physics 34, 569 (1978).

Profiling Hydrogen in Materials Using Ion Beams, J.F. Ziegier + 19 authors including B.L. Cohen, Nucl. Instr. Meth. 149, 19 (1978).

A Generic Hazard Evaluation of Low Level Waste Burial Grounds, B.L. Cohen and H.N. Jow, Nuclear Technology, 41, 381 (1978).

The Relative Risks of Saccharin and Calorie Ingestion, B.L. Cohen, Science 199, 983 (1978); Nature 271, 492 (1978). ,

Cross Sections, B.L. Cohen in " Handbook of Radiation Measurement and Pro-tection", A. Brodsky (Ed.), Chem. Rubber Co. (1978).

Health Risks of Nuclear Power, B.L. Cohen, The Physics Teacher, Nov.1978,

p. 526.

Laws of Statistics Ignored by Statisticians, B.L. Cohen, Health Physics 35, 582 (1978).

Indoor-Outdoor Relationship for Air Particulate of Outdoor Origin, J.

Alzona, B.L. Cohen, H. Rudolph, H.N. Jow, and J.0. Frohliger, Atmosph.

Env.13,1, 55 (1979).

The BEIR Rep 5rt Relative Risk and Absolute Risk Models for Estimating Effects of Low Level Radiation, B.L. Cohen, Health Physics 37, 509 (1979).

i

hcent publications - B.L. Cohen (cont'd) and I.S.-Lee, A Catalog of Risks, B.L. CohengHealth Physics 36, 707 (1979).

Radon: Characteristics, Natural Occurrence, Technological Enhancement, and Health Effects, Progress in Nuclear Energy 4, 1 (1979).

Methods for Predicting the Effectiveness of Uranium Mill Tailings Covers, B.L. Cohen, Nucl. Instr. Meth. 164, 595 (1979).

Radioactive Waste Disposal, B.L. Cohen, in " Handbook of Radiation Measure-ment and Protection", A. Brodsky (Ed.), Chem Rubber Co. (in print) 1982.

Tests of the Linearity Assumption in the Dose-Effect Relationship for Radiation-Induced Cancer, A.F. Cohen and B.L. Cohen, Health Physics 38, 53 (1980).

Protection from Being Indoors Against Inhalation of Suspended Particulate of Outdoor Origin, Atmosph. Environment 14, 1,83 (1980).

The Low Level Radiation Link to Cancer of the Pancreas, Health Physics 38, 712 (1980).

Ocean Dumping of Radioactive Waste, Nuclear Technology 47, 163 (1980).

Society's Valuation of Life Saving in Radiation Protection and other Contexts, Health Physics 38, 33 (1980).

Occupational Risks of Radiation Workers, B.L. Cohen, Health Physics (L),

39, 121 (1980).

Final State Interaction in (3He,2He) Reactions, T.V. Congedo, I.S. Lee-Fan, and B.L. Cohen, Phys. Rev. C2_2, 985 (1980).

Compound Nucleus Contribution and Even-Odd Effect for (3He,p) Reactions in the Nickel Region, I.S. Lee, B.L. Cohen, and T. Congedo, Nuclear Physics A344, 409 (1980).

Analysis, Critique, and Re-evaluation of High Level Waste Water Intrusion Scenario Studies, B.L. Cohen, Nuclear Technology 48, 63 (1980).

The Cancer Risk from Low Level Radiation, B.L. Cohen, Health Physics 39, 659 (1980).

Health Effects of Radon from Ir.sulation of Buildings, B.L. Cohen, Health Physics 39, 937 (1980).

The Role of Radon in Comparisons of Environmental Effects of Nuclear Energy, Coal Burning, and Phosphate Mining, B.L. Cohen, Health Physics 40, 19 (1981).

Perspective on Occupational Mortality Risks, B.L. Cohen, Health Physics

'40, 703 (1981).

Recent publications - B.L. Cohen (cont'd)

Radon Daughter Exposure to Uranium Miners, B.L. Cohen, Health Physics 42, 449 (1982).

Plutonium Containment, B.L. Cohen, Health Physics 40, 76 (1981).

Proposals on Use of the BEIR-III Report in Environmental Assessments, B.L. Cohen, Health Physics 41, 769 (1981).

Effects of ICRP-30 and BEIR-III on Hazard Estimates for High Level Radio-active Waste, B.L. Cohen, Health Physics 42, 133 (1982).

Long Term Waste Problems from Electricity Production, Science (submitted).

Health Effects of Radon from Coal Burning, Health Physics 42, 725 (1982).

Failures and Critique of the BEIR-III Lung Cancer Risk Estimates, B.L. Cohen, Health Physics 42, 267 (1982).

Breeder Reactors - A Renewable Energy Source, B.L. Cohen, Am. Jour. Phys.

(in print).

Physics of the Nuclear Reactor Meltdown Accident, B.L. Cohen, Nuclear Science

& Eng. 80, 47 (1982).

Health Effects of Radon Emissions from Uranium Mill Tailings, B.L. Cohen, Health Physics 42, 695 (1982).

Limitations and Problems in Deriving Risk Estimates for' Low-level Radiation Exposure, Yale Jour. Biol. & Med. $4, 329 (1981).

Long Term Consequences of the Linear-No Threshold Dose-Response Relationship for Chemical Carcinogens, Risk Ar.alysis, Vol.1, No. 4 (1981).

Applications of ICRP-30, IrRP-23, and Radioactive Waste Risk Assessment Tech-niques to Chemical Carcinogens, Health Phys. 42, 753 (1982).

Theory and Practice of Radon Monitoring by Adsorption in Charcoal, Health Physics (submitted).

Large Scintillation Cells for High Sensitivity Radon Monitoring, fiealth Physics (submitted).

Per pective on Genetic Effects of Radiation, Health Phys. (submitted).

Discounting in Assessment of future Radiation Effects, Health Physics (submitted).

Health Risks from Electricity Generation, Coments on Molec. Cellular Biophys.(inprint).

Articles in Popular Journals - B. L. Cohen Learning to Live with Radiation, Science Digest, April 1975, p. 61.

The Potentialities of Terrorism, Bul. Atomic Scient. , June 1976, p. 34.

Environmental Impacts of Nuclear Power Due to Radon Emissions, Bul. Atom.

Scient. , Feb.1976, p. 61.

Some Issues in the Nuclear Power Controversy, Public Util. Fortnightly 98, 4, 31 (Aug. 31,1976).

Storing Radioactive Waste Need Not be a Problem, Nucl. Eng. Int. , Sept.1976.

Plutonium Toxicity, Nucl . Eng. Int. , Nov.1976, p. 35.

Health Risks from Nuclear Waste, AFL-CIO, Viewpoint, 3rd Quarter,1976, p.10.

The Terrorist Threat, Nucl . Eng. Int. , Feb.1977.

Are Nuclear Side Effects Hazardous to your Health, Family Health, Jan.1977, p.52.

The Case for the Breeder Reactor, National Review, Sept. 16, 1977, p.1044.

Disposal of High Level Fadioactive Waste from Nuclear Reactors, Scientific American, June 1977, p.21.

A Tale of Two Wastes,B.L. Cohen, Comentary, Nov.1978, p.63.

Weighing the Risks of Life Today, B.L. Cohen, Los Angeles Times Syndicate, (June,1978).

Cancer and Low Level Radiation, B.L. Cohen, Bul. of Atomic Scientists, Feb. ,1979, p.53; follow-up letter Dec.1979, p.56.

Understanding a Trillion Dollar Question, B.L. Cohen, and R. Brookhiser, National Review, P.142 (Feb. ,1979).

Living Can Be Hazardous to Your Health, B.L. Cohen, Catholic Digest, p.110 (March 1979).

The Situation at West Valley, B.L. Cohen, Public Utilities Fortnightly,

, Sept. 27, 1979, p.26.

Far Greater Dangers than Nuclear, Jour. Am. Scient. Affil . , June 1980, p.89.

Radiation Fan.tasies, Reason, March 1980, p.24.

l

Articles in Popular Journals - B.L. Cohen (cont'd)

Nuclear Energy, B.L. Cohen, Chapter in book by W.T. Hyde (in print).

How Dangerous is a Nuclear Reactor Meltdown?, B.L. Cohen, National Review, Vol. 33, p.667 (June 12, 1981).

How Much Are We Willing to Spend to Save a Life, Public Utilities Fortnightly, Nov. 1981, p.22.

The Storage of Radioactive Waste, The Military Engineer, March 1981, p. 96.

Health Effects of Radiation, Chapter in book by Jennifer Trainer, Norton (New York) 1982.

The Risks You Run, Consumer's Research, May 1981, p.16.

Q and A on Waste, Nuclear Industry 28, p. 28 (1981).

How Dangerous is Radiation?, Ascent, Vol. 2, No. 4 (1981).

High Level Radioactive Waste, Natural Resources Journal, Oct.1981, p. 703.

Radiation Pollution and Cancer: Comparative Risks and Proof, Cato Journal (in print).

Is Nuclear Power Too Risky?, The American Legion, Jan.1982, p.16.

i Genetic Effects of Radiation, Ascent, Vol. 3, No. 3, p. 8 (1982).

Journalism and Nuclear Power, Commentary (submitted).

l

November, 1980

. s.

ROBERT L. DUPONT, M. D.

BIOGRAPHICAL SKETCH Robert L. DuPont, M. O. , is a practicing osychiatrist and President of the ncn-profit Institute for Behavior and Health, Inc. (IBH). As part of his practice of psychiatry, he directs Washington's first phobia treat-ment program. In addition, he is President of the American Council on Marijuana and contributes to local and national TV, radio, magazines, and

'wspapers en a variety of health topics.

The American Council on Marijuana is the nation's leading private ,

organization linking scientists to ccmmunity acticn programs. It interprets the latest scientific research for the public and offers leadership on mari-juana policy. The American Council on Marijuana and Other Psychoactive Drugs was founded in 1977. ACM has offices in New York City and Rockville, Maryland.

Dr. DuPont has a special interest in substance abuse prevention pro-grams in the schools and in the workplace. The Institute for Behavior and Health conducts research and demonstration programs aimed at preventing drug and alcohol abuse, as well as mere broadly targeted health promotion efforts.

The Phobia Program of Washington is a structured 20-week program which helps phobic people overccme agoraphobia, fear of flying, fear of driving on major highways, claustrephobia, and other phobias. The program, which was founded in 1977, also includes self-help meetings for former program members, and outreach services for housebound agoraphobics. In May, 1978, Dr. DuPont chaired a Special Session at the American Psychiatric Association's annual meeting in Atlanta on the " Treatment of Phobias." He chaired a similar Spe-cial Session at the 1979 APA meeting in Chicago and the 1980 APA meeting in San Francisco. In 1980 he led the second annual National Phobia Conference held in Washington, D. C.

In addition to his work as a health ccmmentator en ABC-TV's Good Morning, America, Dr. DuPont has appeared en many network TV shows, including -

The Phil Donahue Shcw, The David Suskind Show, and The Dick Cavett Show. He is a frequent guest talk-shcw host en WRC-NBC radio in Washington, D. C. He has been quoted in U. S. News and World Recort, Tice and Newsweek, and has appeared on the evening networx news, the Tocay Shcw, and many TV documentaries.

Dr. DuPont was the Director of the National Institute on Drug Abuse from its creation in September,1973, until July,1978. In June, 1973, he was appointed by the President and confirmed by the Senate to direct the White House Special Action Office for Drug Abuse Prevention, a position he held until the office terminated in June,1975. As SACDAP Director with a staff of more than 100, he designed and coordinated the entire 51 billion a year federal drug abuse preventien program.

In his role as Director of the National Institute on Drug Abuse, he directed the Federal Government's major drug abuse treatment, research and preventicn effort with a staff of 400 and a budget of $230 millicn a year.

From 1970 to 1973, Dr. CuFont served as Administratcr of the Narcotics Treatment Administratien (NTA) of the District of Columbia Department of M

. 4 Human Resources. NTA was a comprehensive city-wide multimodality heroin addiction treatment program which treated 15,000 pecple with a staff of 500 working in 20 facilities curing these years.

As a research psychiatrist and Acting Associate Director for Cc=r. unity Services of the District of Columbia Department of Corrections from 1968 to 1970, he directed the city's parole and halfway house programs and developed a pilot narcotics treatment program.

Dr. DuPont has written more than 100 professional articles and one book on a variety of topics in the fields of health promotion, drug abuse prevention, and criminal justice. He holds the faculty positions of Clinical Professor of Psychiatry at Georgetcwn University Medical School, and Visiting Associate Clinical Professor of Psychiatry at Harvard Medical School.

He is a diplomate of the American Board of Psychiatry and Neurology, a fellow of the American Psychiatric Association, and a member of many pro-fessional organizations, including the Academy of Behavioral Medicine Research, the Behhvioral Medicine Special Interest Group, the American Public Health Association, the World Psychiatric Association, the Pan American Medical Asso- '

ciation, the Medical and Chirurgical Faculty of the State of Maryland, and the Montgomery County Medical Society. Dr. DuPont was Chairman of the Drug Dependence Section of the World Psychiatric Association, from 1974 thrcugh 1979.

Dr. DuPont is listed in Who's Who in America and has received honors, including being chosen the Outstanding Young Man in the District of Columbia Government in 1972 by the Dcwntcwn Jaycees. In 1973, he was given the Meri-torious Service Award by the Mayor of the District of Columbia. He was

. awarded the highest honor in the U. S. Public Health ~3ervice, the Superior Service Award, by the Surgeon General in 1978. He has also beer honored by several local and naticnal drug abuse and alcohol abuse prevention organiza-ti ons.

Born on March 25, 1936, in Toledo, Ohio, he attended public high school in Denver, Colorado; received a B. A. from Emory University in Atlanta, Georgia, in 1958; and an M. D. from Harvard Medical School in Boston, Massa-chusetts, in 1963. His postgraduate training includes: Medical Intern, Cleveland Metropolitan General Hospital, Western Reserve Medical School (1963-1964); psychiatric resident and teaching fellcw, Massachusetts Mental Health Center, Harvard Medical School (1964-1966); and clinical associate, Naticnal Institutes of Health (1966-1968).

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. o ROBERT L. DUPONT, M. D.

SUPPLEMENTAL BIOGRAPHICAL SKETCH Ma;erial Pertaining to Fears of Nuclear Power:

Tc 1979, I was asked by the non-profit Media Institute to review network TV news coverage of nuclear power between 1968 and 1979. This led to publica-tion of " Nuclear Phobia -- Phobic Thinking About Nuclear Power."

Subsequently, I have visited nuclear power plants at Three Mile Island, Peach Bottom in Pennsylvania, Diablo Canyon in California, and in France. In addi-tion to talking to employees at each of these nuclear power plants, I spoke with neighbors and county and regional leaders including physicians and poli-ticians.

In October of 1981, I participated in an international Conference at Ditchley Park in England about the media coverage of nuclear power in five countries:

England, France, Germany, the United States, and Japan.

My publications on the psychology of nuclear power include the following: -

DuPont, R. L.: " Phobic Fear as a Nuclear Health Hazard." The Washingt~i ~

Star, July 20, 1980.

DuPont, R. L.: Nuclear Phobia -- Phobic Thinking About Nuclear Power. The Media Institute, March, 1980.

DuPont, R. L.: " Nuclear Phobia: Phobic Thinking About Nuclear Power," in Nuclear Power in American Thought, Decisionmakers, Vol. 8, Edison Electric Institute, 23-41, 1981.

DuPont, R. L.: " Fifty Million Frenchmen Have Few Nuclear Fears." Electric Perspectives, Edison Electric Institute, 33-36, Fall,1981.

DuPont, R. L.: "The Nuclear Power Phobia." Eusiness Week, 14-16, September 7, 1981. (Reprinted, Congressional Record, September 15, 1981.)

DuPont, R. L.: "The Psychology of Phobic Fear of Nuclear Energy," in Phobia:

A Comprehensive Summary of Modern Treatments. Edited by Robert L.

De'ont, M. D. , in Press, to be published in ~the Spring of 1982, Brunner/

Ma zel .

Congressional Hearing:

Statement of Robert L. DuPont, M. D. , before the Committee on Science and Technology, Subcommittee on Energy Research and Production, U. S. House of Representatives, Washington, D. C., December 15, 1981.

Unpublished Manuscripts:

DuPont, R. L.: " Understanding Fear of Nuclear Power." Presented at the International Conference of the Atomic Industrial Forum, Inc. , November 18, 1980, Washington, D. C.

. t - Uncublished Manuscriots (continued):

DuPont, R. L.: " Lessons from France: Fears of Nuclear Power." May 4, 1981.

DuPont, R. L.: " Perspectives of Nuclear Risk: The Role of the Media." Pre-sented at the Annual Meeting of the Canadian Nuclear Association, June 9, 1981, Otta'wa, Canada.

DuPont, R. L.: " Phobia Fear of Nuclear Energy -- Why Don't the French Have It?"

June 12,1981.

DuPont, R. L.: "The Press Isn't to Blame for Nuclear's Problems!" December 20, 1981.

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m_- . . , , __ , . , . - . . ., ,, . , _ _ _ , , . . . _ . _ - . , . . , . . _ . . -, . _ - . _ - _ _ . , , _ _

NAME B. JOHN GARRICK EDUCATION Ph.D., Engineering, University of California, Los Angeles,1968.

M.S., Engineering, University of California, Los Angeles,1962.

B.S., Physics, Brigham Young University,1950.

U.S. Atomic Energy Commission Grant-in-Aid, Oak Ridge School of Reactor Technology , 1954-1955.

PROFESSIONAL EXPERIENCE General Summary A principal of Pickard, Lowe and Garrick, Inc. Consultant in reliability and availability, risk analysis, licensing and safety, management systems, and engineering. Pioneered early use of reliability and risk analysis technology in nuclear and fossil power plants. Served on several design review and safety committees and other task forces related to power plant design and operations. -Study director of numerous major risk studies of nuclear power plants including Oyster Creek, Zion, Indian Point, LaSalle, Pilgrim 1, Midland, and Browns Ferry. Extensive experience with hearings and the general nuclear licensing process.

Coordinator and principal lecturer for the annual UCLA short course on power plant reliability. Presented numerous seminars on risk and safety analysis at such institutions as MIT, the University -of California, and the United Kingdom's National Centre of Systems Reliability. Served on several accreditation teams evaluating engineering curriculum at different universities. Organized and conducted nwnerous workshops and training programs on maintenance, reliability, and availabilty for EPRI, DOE, and many utilities.

Adjunct Professor, University of California, Los Angeles; member of several institutional committees including the UCLA Radiation Committee, the Select Review Comnittee for the Clinch River Breeder Reactor, Design Review Board for the Midland Nuclear Power Plant, Direction and Control System Advisory Connittee of the Governor's Emergency Task Force on Earthquake Preparedness, and Boston Edison's Audit and Nuclear Review Committee.

Chronological Summary 1975-Present Principal, Pickard, Lowe and Garrick, Inc.

1957-1975 Holmes & Narver, Inc.

Key Positions: Member of Board of Directors; President, Nuclear & Systems Sciences Group; Sr. Vice President; Vice President, Science & Technology, The Resource Sciences Corporation, Tulsa, Oklahoma (parent company).

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GARRICK - 2 e

1955-1957 Physicist, Hazards Evaluation Branch, U.S. Atomic Energy Commission, Washington, D.C.

1952-1954 Physicist, Phillips Petroleum Company, National Reactor Testing Station, Idaho.

MEMBERSHIPS, LICENSES, AND HONORS American Nuclear Society.

Fellow, Institute for the Advancement of Engineering.

New York Academy of Sciences.

Registered Professinal Engineer, State of California.

Leaders in American Science (Eighth Edition).

REPORTS AND PUBLICATIONS "Seabrook Probabilistic Safety Assessment," Public Service Company of New Hampshire, to be published in 1983.

Pickard, Lowe and Garrick, Inc. , " Midland Probabilistic Risk Assessnent,"

Consumers Power Company, to be published in 1932.

Oconee Probabilistic Risk Assessment," a joint effor.t of the Nuclear Safety Analysis Center, Duke Power, and other participating utilities, to be published in 1982.

Tennessee Valley Authority and Pickard, Lowe and Garrick, Inc., " Browns Ferry Probabilistic Risk Assessment," to be published in 1982.

Garrick, B. J. , and H. F. Per.la, " Quantitative Risk Management - A New Tool for the Engineering of Facilities," to be presented at the ASCE '

National Convention, New Orleans, Louisiana, October 26, 1982.

Garrick, B. J., " Experiences and Practices in the Applications of Risk Analysis," to be presented at the ORSA/TIMS,1982, Joint National Meeting, San Diego, California, October 25-27, 1982.

Heising, C. D. , A. W. Barsell, K. N. Fleming, S. Kaplan, and B. J.

Garrick, "A comparison of Recent Nuclear' Plant Risk Assessments,"

proposed presentation at the Congres Annuel 1982, SFRP, La Comparaison des Risques Associes aux Grandes Activites Humaines, Avignon, France, October 18-22, 1982.

Garrick, B. J., "Probabilistic Risk Assessments and Quality Assurance,"

to be presented at the Ninth Annual National Energy Division Conference, ASQC, Orlando, Florida, October 12-15, 1932.

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GARRICK - 3 Bley, D. C., S. Kaplan, and B. J. Garrick, " Assembling and Decomposing PRA Results: A Matrix Formalism," proposed presentation at the International Meeting on Thermal Nuclear Reactor Safety, Chicago, Illinois, August 29-September 2,1982.

Kaplan, S., "On a 'Two-Stage' Bayesian Procedure for Determining Failure Rates from Experiential Data," PLG-0191, preprint of a paper to appear in the IEEE Transactions on Power Apparatus and Systems, August 1982.

Garrick, B. J., "Probabilistic Risk Assessment as an Aid to Risk Management," presented to the 17th Intersociety Energy Conversion Engineering Conference, Los Angeles, California, August 8-13, 1982.

Garrick, B. J., S. Kaplan, and D. C. Bley, "Recent Advances in Probabilistic Risk Assessment," prepared for the MIT Nuclear Power Reactor Safety Course, Cambridge, Massachusetts, July 19, 1982.

Fleming, K. N., S. Kaplan, and B. J. Garrick, "Seabrook Prvbabilistic Safety Assessment Management Plan,"PLG-0239, June 1982.

Garrick, B. J., " Lessons Learned From First Generation Nuclear Plant Probabilistic Risk Assessments," to be presented at the Workshop on Low-Probability /High-Consequence Risk Analysis, Arlington, Virginia, June 15-17,1982.

Garrick, B. J. , S. Kaplan, D. C. Iden, E. B. Cleveland, H. F. Perla, D.

C. Bley, D. W. Stillwell, H. V. Schneider, and G. Apostolakis, " Power Plant Availability Engineering: Methods of Analysis, Program Planning, and Applications," EPRI NP-2168, PLG-0165, May 1982.

Garrick, B. J. , and H. F. Perla, " Management of PRA Projects," presented at the ANS Executive Conference, Arlington, Virginia, April 4-7, 1982.

Garrick, B. J., "Results From Recently Performed PRA Studies (Zion / Indian Point) and the Proposed Use of Such Studies for Safety Decision-Making," presented at the ANS Northeastern Section and M.I.T.

Depart:nent of Nuclear Engineering Meeting, Cambridge, Massachusetts, March 16,1982.

" Indian Point 2 and 3 Probabilistic Safety Study," Power Authority of the State of New York and Consolidated Edison Company of New York, Inc.,

March 1982.

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4 GARRICK - 4 Stillwell, D. W., G. Apostolakis, D. C. Bley, P. H. Raabe, R. J. Mulvibill, S. Kaplan, and B. J. Garrick, "EEI Availability Handbook," PLG-0218, January 1982.

" Zion Probabilistic Safety Study," Commonwealth Edison Company, September 1981.

Garrick, B. J. , and D. C. Iden, " Project Plan, Probabilistic Risk Assessment, LaSalle County 1 and 2," PLG-0208, September 1981.

Siu, N. 0., and B. J. Garrick, " Analysis of Severe Transportation Accident Environment Frequencies," PLG-0207, September 1981.

Mulvibill, R. J. , B. J. Garrick, Y. G. Mody, and D. A. Reny, " Comparative Evaluation of Air Quality Control Equipment Availability for Intermountain Power Project," PLG-0200, September 1981.

Iden, D. C., B. J. Garrick, and Loren Reichers, "0yster Creek Feed and Condensate System Reliability Study," presented at the Tenth Biennial Topical Conference on Reactor Operating Experience, Cleveland, Ohio, August 17-19, 1981.

Stillwell, D. W., B. J. Garrick, D. R. Buttemer, G. Apostolakis, J. C. Lin, and S. Kaplan, " Analysis of the Pilgrim Nuclear Power Station Reactor Protection System," PLG-0195, July 1981.

Garrick, B. J., and D. C. Bley, " Lessons Learned from Current PRAs,"

presented to the ACRS Subcommittee on Reliability and Probabilistic Risk Assessment, Los Angeles, California, July 28, 1981.

Kaplan, S., G. Apostolakis, B'. J. Garrick, D. C. Bley, and K. Woodard, '

" Methodology for Probabilistic Risk Assessment of Nuclear Power Plants,"

draft version of a book in preparation, FLG-0209, June 1981.

Mulvihill, R. J. , B. J. Garrick, and Y. G. Mody, " Reliability, Availability, and Maintainability (RAM) Program Plan," PLG-0182, June 1981.

Mulvibill, R. J. , and B. J. Garrick, R. S. Hanson, S. Kaplan, Y. G. Mody, D. A. Reny, L. H. Riechers, and H. V. Schneider, " Comparative Evaluation of Boiler Availability for Intermountain Power Project," PLG-0169, April 1981.

Mulvibill, R. J. , and B. J. Garrick, " Tutorial: Reliability Analysis of Plant Systems and Components; A Boiler Procurement Case History,"

PLG-0171, presented to the Eighth Annual Reliability Engineering Conference for the Electric Power Industry, Portland, Oregon, April 1981.

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GARRICK - 5 Kaplan, S.,

and B. J. Garrick, "On the Quantitative Definition of Risk,"

PLG-0196, Risk Analysis, Vol.1, No.1, March 1981.

Bley, D. C., C. L. Cate, D. W. Stillwell, and B. J. Garrick, " Midland Plant Auxiliary Feedwater System Reliability Analysis Synopsis,"

PLG-0166, March 1981.

Garrick, B. J., S. Kaplan, and N. O. Siu, " Definition of Bounding Physical Tests Representative of Transport Accidents - Rail and Truck,"

PLG-0164, March 1981.

Kaplan, S., B. J. Garrick, and G. Apostolakis, " Advances in Quantitative Risk Assessment - The Maturing of a Disciplinn," IEEE Transactions on Nuclear Science, NS-28, No.1, February 1981.

Garrick, B. J. , and D. C. Ideo, " Project Plan: Probabilistic Risk Assessment, LaSalle County 1 and 2," PLG-0158, January 1981.

Apostolakis, G. , S. Kaplan, B. J. Garrick, and R. J. Duphily, " Data Specialization for Plant Specific Risk Studies," Nuclear Engineering and Design 56, 1980.

Kaplan, S., L. H. Reichers, and B. J. Garrick, " Histogram Convolution Program (HICOP)," PLG-0157, December 1980.

Garrick, B.. J. , S. Ahmed, and D. C. Bley, "A Methodology for Evaluating the Costs and Benefits of Power Plant Diagnostic Techniques," submittad for presentation at the Ninth Turbomachinery Symposium, Houston, Texas, December 9-11, 1980.

Hanson. R. S. , J. C. Lin, D. 'M. Wheeler, S. Kaplan, B. J. Garrick, D. C. , Iden, W. B. Holder, and L. G. H. Sarmanian, "An Assessment of the Reliability of Turbine-Generators," PLG-0155, November 1980.

Garrick, B. J. , S. Kaplan, D. C. Iden, E. B. Cleveland, H. F. Perl a, D. C. Bley, and D. W. Stillwell, " Power Plant Availability Engineering, Methods of Analysis - Program Planning - Applications," 2 Vols. ,

PLG-0148. October 1980.

Bley, D. C. , C. L. Cate, D. W. Stillwell, and B. J. Garrick, " Midland Plant Auxiliary Feedwater System Reliability Analysis," PLG-0147, October 1980.

Bley, D. C. , D. M. Wheeler, C. L. Cate, D. W. Stillwell, and B. J. Garrick, " Reliability Analysis of Diablo Canyon Auxiliary Feedwater System," PLG-0140, September 1980.

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GARRICK - 6 Garrick, B. J. , et al, " Project Plan: Oconee Probabilistic Risk Assessment," PLG-0138, August 1980.

Garrick, B. J. , D. M. Wheeler, E. B. Cleveland, D. C. Bley, L. H. Reichers, and C. B. Morrison, " Operating Experience of Large U.S. Steam Turbine-Generators; Volume 1 - Data, Volume 2 - Utility Directory," PLG-0134, June 1980.

Garrick, B. J. , and D. C. Iden, " Study Plan: Probabilistic Risk Assessment, LaSalle County 1 and 2," PLG-0128, June 1980.

Kaplan, S.,

R. S. Hanson, B. J. Garrick, and J. W. Stetkar, "A Strategic Plan1980.

July for a National Data System for Electric Power Plants," PLG-0144, Garrick, B. J. , S. Kaplan, G. 'Apostolakis, D. C. Iden, K. Woodard, and T. E. Potter, " Seminar: Probabilistic Risk Assessment of Nuclear Power Pl ants," PLG-0141, July 1980.

Garrick, B. J. , S. Kaplan, and D. C. Bley, " Seminar: Power Plant Probabilistic Risk Assessment and Reliability," PLG-0127, May 1980.

Garrick, B. J., " Power Plant Availability Decision-Making Under Uncer-tainty'," presented at the 34th Annual Technical Conference, American Society for Quality Control, Atlanta, Georgia, May 20-22, 1980.

Garrick, B. J. , and S. Kaplan, "A Conceptual Plan for a National Data System for Electric Power Plants," PLG-0131, April 1980.

Garrick, B. J., and S. Kaplan', "0yster Creek Probabilistic Safety Anal-ysis (OPSA)," presented at the ANS-ENS Topical Meeting on Thermal Reactor Safety, Knoxville, Tennessee, April 8-11, 1980.

Kaplan, S., and B. J. Garrick, "A Strategic Plan for a National Reliability Data System," PLG-0125, March 1980.

Garrick, B. J. , S. Kaplan, G. E. Apostolakis, D. C. Bley, and T. E. Potter, " Seminar: Probabilistic Risk Assessment as Applied to Nuclear Power Plants," PLG-0124, March 1980.

Kaplan, S. , and B. J. Garrick, "Try Probabilistic Thinking to Improve Power Plant Reliability," Power, March 1980.

Garrick, B. J. , " Progress Report: Availability Engineering Guide, Electric Power Plants," presented to the EEI Availability Engineering Task Force, Atlanta, Georgia, March 26, 1980.

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GARRICK - 7 Garrick, B. J. , "Probabilitistic Risk Assessment and Regulation,"

presented at the AIF Workshop on Licensing and Technical Issues, - Post TMI, Washington, D.C. , March 9-12, 1980.

Garrick, B. J. , S. Ahmed, and D. C. Bley, "A Methodology for Evaluating the Costs and Benefits of Power Plant Diagnostic Techniques," PLG-0118, January 1980.

Garrick, B. J. , " Power Plant Reliabilty: Practices and Trends," presented to the IEEE Power Engineering Society, San Francisco Section, January 15, 1980.

Apostolakis, G. , S. Kaplan, B. J. Garrick, and W. Dickter, "Assessmeat of the Frequency of Failure To Scram in Light Water Reactors," Nuclear Safety, Vol. 20, No. 6, November-December 1979.

Garrick, B. J. , and H. F. Perla, " Maintenance Management in the Electric utility Industry," presented to the International Conference on Energy Use Management, Los Angeles, Calfornia, October 22-26, 1979.

Bl ey, D. C. , C. L. Cate, D. C. Iden, B. J. Garrick , and J. M. Hudson,

" Seismic Safety Margins Research Program (Phase I), Project VII - Systems Analysis," PLG-0110, September 1979.

Garrick, B. J., S. Kaplan, and S. Ahmed, "A Reliability Prediction Technique for Selected Thermomechanical Components of Gas Turbine Combined Cycl e Plants," PLG-0109, September 1979.

Iden, D. C., S. Ahmed, and B. J. Garrick, " Impact of 18-Month Refueling Cycle on BWR Plant Availability," PLG-0107, Septe'aber 1979.

Garrick, B. J. , S. Kaplan, P. P. Bieniarz, K. Woodard, D. C. Iden, H. F. Perl a, W. Dicter, C. L. Cate, T. E. Potter, R. J. Duphily, T. R. Robbins, D. C. Bley, and S. Ahmed, "0PSA, Oyster Creek l

Probabilistic Safety Analysis," (Executive Summary, Main Report, Appendixes), PLG-0100 DRAFT, August 1979.

l Garrick, B. J. , and S. Ahmed, " Technical Note: Prediction of Reliability Growth for Alternate Designs of Combustion Turbines," PLG-0105, July 1979.

Kaplan, S., and B. J. Garrick, "On the Use of a Bayesian Reasoning in Safety and Reliability Decisions--Three Examples," Nuclear Technology, Vol . 44, July 1979.

Kaplan, S. , and B. J. Garrick, " Notes on Prediction of Reliability,"

PLG-0117, June 1979.

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GARRICK - 8 Garrick, B. J., and R. W. Pack, " Workshop Proceedings: Outage Planning and Maintenance Management," 2 Vols., June 1979.

Kaplan, S., and B. J. Garrick, " Notes for a Workshop on Risk, Reliability, and Decision Under Uncertainty," presented at Battelle Northwest Laboratory, June 1979.

Cate, C. L., and B. J. Garrick, "W-501 Combustion Turbine Starting Reliability Analysis," PLG-0103, June 1979.

Garrick, B. J., P. P. Bieniarz, and S. Kaplan, " Risk Analysis of Transporting Oconee Spent Nuclear Fuel to the McGuire Nuclear Station,"

PLG-0102, June 1979.

Iden, D. C., S. Ahmed, B. H. Garrick, and J. K. Pickard, " Full Core Removal Study," PLG-0101, May 1979.

Garrick, B. J., and S. Kaplan, " Training Engineers to be Reliability Practitioners," presented to the Sixth Annual Reliability Engineering Conference for the Electric Power Industry," Miami Beach, Florida, April, 19-20, 1979.

Dickter, W., B. J. Garrick, and S. Ahmed, " Selecting Corrective Actions,"

March 1979.

Apostolakis, G. , S. Kaplan, B. J. Garrick, and W. Dick'ter, "An Assessment of the Frequency of Failure to Scram," Vol. 20, No. 6, Nuclear Safety, November-December 1979.

Garrick, B. J., " Establishing Availability Goals in Procurement,"

presented to the ANS Topical Meeting on Reliable Nuclear Power Today, Charlotte, North Carolina, April 9-13, 1978.

Garrick, B. J., "0yster Creek Probabilistic Safety Analysis, Phase 1 Report," February 1978.

Garrick, B. J. , " Energy and Property Values," presented to the 1978 Southern Region Conference, California Assessors' Association, Anaheim, February 8-10, 1978.

Garrick, B. J. , S. Kaplan, and P. P. Bieniarz, " Input Material for Reliability Section of Westinghouse Turbine-Generator Proposal to Middle South Utilities," December 1977.

Garrick, B. J. , and W. Dickter, "How Do We Improve the Productivity of Large Power Plants?" presented at the Pacific Coast Electrical Association Engineering and Operating Conference, Los Angeles, California, March 1977.

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GARRIDC - 9 Garrick, B. J., " Reliability Engineering Guide, Sundesert Nuclear Plant,"

October 1976.

Garrick, B. J., "Where Do We Go from Here in Inproving Reliability? The Role of the Consultant," presented to the ANS Executive Conference, Hot Springs, Virginia, September 27-29, 1976. ,

Garrick, B. J., and S. Kaplan, " Reliability Technology and Nuclear Power," IEEE Transactions on Reliability, Vol. R-25, No. 3, August 1976.

Garrick, B. J., " Technical Note: Cost-Benefit Estimate of Transporting

, Spent Nuclear Fuel by Special Trains," August 1976.

Garrick, B. J., " Broadening the Application of Risk and Reliability Technology (Summary)," presented to the ANS Annual Meeting, Toronto, Canada, June 13-18, 1976. -

Garrick, B. J., and W. C. Gekler, "The Impact of Reliability on Power l

Plant Economics," presented to the Pacific Coast Electrical Association Engineering and Operating Conference, San Francisco, California, March 18-19,1976.

Garrick, B. J., " Nuclear Power - Issues with Answers," Engineering Forum, California State University, Long Beach, February 27, 1976.

Garrick, B. J., and S. Kaplan, " Reliability Technology' and Nuclear Power," 197 5.

Garrick, B. J., " Nuclear Power Plant Availability, Its Costs and Benefits (Outline Only)," presented to the Atomic Industrial Forum, Workshop on Reactor Licensing & Safety, San Diego, California, December 10-13, 1974.

I Garrick, B. J. , A. A. Jarrett, E. B. Cleveland, and W. C. Gekler,

" Availability Analysis: Initial Allocation of the Clinch River Breeder Reactor Plant," NSS-8212, Projact Management Corporation, Chicago, Illinois, November 1974.

Garrick, B. J. , W. C. Gekler, E. B. Cleveland, and L. T. Allard,

" Reliability Assurance Program Plan for the Clinch River Breeder Reactor Pl ant," NSS-8212, November 1974.

Garrick, B. J. , C. V. Hodge, and A. A. Jarrett, " Transportation Accident Risks in the Nuclear Power Industry 1975-2020," NSS 8191, November 1974.

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GARRICK - 10 Garrick, B. J., " Problems and Methodology in Power Plant Siting Based on the California Experience," presented to the Power Systems Engineering Program, The University of Arizona, Tucson, March 1974.

Garrick, B. J., " Site Evaluation Decision Methodology," 1973.

Garrick, B. J. , " California Power Plant Siting," Nuclear Safety,1973.

Garrick, B. J. , " Risk and Systems Analysis Projects," Holmes & Narver, Inc., December 18, 1972.

Garrick, B. J., and R. J. Mulvibill, " Maintainability in Automated Postal Systems," presented to the Los Angeles Maintainability Association, SYMP0-72, May 20,1972.

Garrick, B. J., S. Kaplan, "A Method for Evaluating Nuclear Plant Siting Concepts," presenteed before the Joint Committee on Atomic Development and Space, California Legislature, Sacramento, May 19, 1972.

Garrick, B. J. , S. Kaplan, and 0. C. Baldonado, "On a Decison Theory Formalism for Nuclear Pewar Plant Siting," presented to the Conference on Unique Siting Concepts for Nuclear Power Plants, Joint Committee on Atomic Development and Space, Sacramento, California, May 9,1972.

Garrick, B. J., " Reliability Analysis of Bulk Mail Handling Facilities, Holmes & Narver, Inc. , Los Angeles, California,1971.

Garrick, B. J. , R. J. Mulvihill, et al, " Failure Modes Analysis for the New York Bulk Mail Center," HN-8101.20, August 1971.

Garrick, B. J. , R. J. Mulvibill, et al, "RMA Studies and Recommendations for Chicago Bulk Mail Center," HN-8114.1 (Rev.), August 1971.

Garrick, B. J. , R. J. Mulvihill, et al, " Failure Modes Analysis for the i

NBMS Modular BMC," HN-8106.6, n.d.i.

Garrick, B. J., " Reliability Analysis of Bulk Mail Handling Facilities,"

Tenth Annual Reliability and Maintainability Conference, Anaheim, California, June 28-30, 1971.

l Garrick, B. J. , O. C. Baldonado, and W. C. Gekler, " Estimating the Risk Involved in Transport of Hazardous Materials," presented at CREST, i

l Specialist Meeting on Applicability of Quantitative Riiability Analysis of Complex Systems and Nuclear Plants in Its Relation to Safety, Munich, Ge rmany, May 1971.

i l

1 10

, 0487P072082 l

GARRICK - 11 Garrick, B. J. , O. C. Baldonado, and W. C. Gekler, " Estimating the Risk Involved in Transport of Hazardous Materials," presented to the Armed Services Explosives Safety Board Seminar, August 25-27, 1970.

Garrick, B. J. , R. J. Mulvibill, et al, " Reliability A'nalysis of Substation Unit Configurations," HN-216, January 1970.

Gekler, W. C., et al, and B. J. Garrick, "A Risk Model for the Transport of Hazardous Matef als," HN-204, August 31, 1969.

Gekler, W. C. , O. C. Baldonado, H. K. Elder, J. E. Shapley, and B. J. Garrick, "A Risk Model for the Transport of Hazardous Materials,"

HN-204, Augus t 1969.

Garrick, B. J. , C. V. Chelapati, R. A. Williamson, D. V. Dominic, and O. C. Baldonado, "Probabiiistic Analysis of Nuclear Reactor Containment Structures," Part I - Seismic Analysis of Reactor Building, Volume I, HN-203, June 1969.

Chel apati, C.- V. , R. A. Williamson, C. V. Dominic, 0. C. Baldonado, B. J. Garrick, "Probabilistic Analysis of Nuclear Reactor Containment

• Structures," Part I - Seismic Analysis of Reactor Building and Part 2 -

Finite Element Analysis of Drywell Due to Loss-of-Coolant Accident, HN-203, June 1969.

Garrick, B. J. , B. Shimizu, W. C. Gekler, and J. H. Wilson, " Collection of Reliability Data at Nuclear Power Plants," HN-199, December 1968.

Garrick, B. J. , " Unified Systems, Safety Analysis for Nuclear Power Plants," Ph.D. Thesis, University lof California, Los Angeles, California, 1968.

Garrick, B. J. , B. Shimizu, M. T. Campora, R. C. Erdmann, W. C. Gekler, and J. E. Shapley, " Reliability Monitoring Pilot Test Program -

Connecticut Yankee Project, Part I - Program Initiation," (preliminary draf t), HN-lS8, September 1968.

Jackson, M. W., B. J. Garrick, R. A. Williamson, R. P. Kennedy, i

D. L. Platus, Y. A. Smoots, " Report of Study of Hanford Waste Tank Structure," HN-197, May 1968.

Garrick, B. J. , W. C. Gekler, 0. C. Baldonado, E. H. Behrens, and B. Shimizu, " Classification and Processing of Reliability Data from Nuclear Power Pl ants," HN-193, February 19CS.

Garrick, B. J. , and S. Kaplan, " Notes on Fault Trees and Reliability Calculations," (unpublished work), HN 342.32.08, September 1967.

11 04S7P072082

GARRICK - 12 Garrick, B. J. , R. C. Erdmann, and S. Kaplan, "Another Approach to Reliability Calculations," (unpublished work), HN 342.32.07, August 1967.

Garrick, B. J. , B. Shimizu, E. H. Behrens, W. C. Gekler, L. Goldfisher, and J. H. Wilson, " Reliability Analysis of Carolinas Virginia Tube Reactor Engineered Safety Systems," (preliminary draft), HN-191, August 1967.

Garrick, B. J., and W. C. Gekler, " Reliability Analysis of Engineered Safeguards," Nuclear Safety, Vol. 8, No. 5, November 1967.

Garrick, B. J. , W. C. Gekl er, L. Goldfisher, R. H. Karcher, B. Shimizu, and J. H. Wilson, " Reliability Analysis of Nuclear Power Plant Protective Systems," HN-190, Phillips Petroleum Company, May 1967.

Garrick, B. J., "An Approach to Systems Safety Analysis for Nuclear Power Plants," paper presented at the meeting of the Comnittee on Reactor Safety Technology, Brussels, Belgium, April 11-12, 1967.

Garrick, B. J. , W. C. Gekl er, H. P. Pomrehn, "Are Plant Engineered Safeguards Reliable?" Nucleonics, January 1967.

Garrick, B. J. , W. C. Gekler, H. P. Pomrehn, "An Analysis of Nuclear Power Plant Operating and Safety Experience," HN-185, December 1966.

Garrick, B. J., J. H. Wilson, R. H. Karcher, "Shieldin'g Effectiveness of Soils Against Initial Radiation," HN-186, August 1966.

t Garrick, B. J., " Engineering Aspects of Nuclear Plant Safety," American Society of Civil Engineers, January 1966.

l Garrick, B. J., " Siting Aspects of a Nuclear Power Reactor in the Republic of Korea," performed for the International Atomic Energy Agency under the auspices of the U. S. Atomic Energy Comnission, June 1965.

Garrick, B. J. , R. H. Karcher, J. H. Wilson, " Structural and Shielding 1

Considerations in the Design of Hardened Facilities," HN-183, June 1965.

Garrick, B. J. , W. C. Gekler, H. P. Pomrehn, " Preliminary Site Analysis I

and Nuclear Design Evaluation of a Proposed large Central Station Nuclear Power Plant," May 1965.

l l Garrick, B. J. , W. C. , Gekler, H. P. Pomrehn, " Preliminary Site Analysis and Nuclear Design Evaluation of a Proposed Large Central Station Nuclear Power Plant," May 1965.

l l

l l

12 0487P072032

I GARRICK - 13 Garrick, B. J. , W. C. Gekler, H. P. Pomrehn, "Some Aspects of Protective Systems in Nuclear Power Plants," 1965 IEEE International Convention, March 1965.

Garrick, B. J. , W. C. Gekler, " Monitoring the Reliability of Engineered Safeguards," American Nuclear Society National Topical Meeting, February 1965.

Garrick, B. J. , "A Numerical Formalism for Reactor Site Selections," UCLA Research Project, 1964.

Garrick, B. J. , W. C. Gekler, J. M. Duncan, R. H. Karcher, B. Shimizu, "A Study of Research Reactor Operating and Safety Experience," HN-180, June 1964.

Garrick, B. J. , R. H. Karcher; " Design Information for Protection Against Initial Radiation (Secret-Restricted)," HN-179, May 1964.

Garrick, B. J. , "The Meaning of Test Reactor Operating and Safety Experience to Power Reactor Operations," presented to the 1963 Annual Conference of the Atomic Industrial Forum, New York, New York, -

November 18-21, 1963.

Garrick, B, H., P. W. van der Pas, "Least Squares Analysis of Parameters of U-233, U-235, and Pu-239," UCLA Research Project, June 1963.

Garrick, B. J. , W. C. Gekler, " Nuclear Parameter Study of a Pu-239 Gaseous Core Reactor," UCLA Research Project, June 1963.

Garrick, B. J. , W. J. Costley, W. .C. Gekler, "A Study of Test Reactor Operating and Safety Experience," HN-172 (Volumes I and II), May 1963.

Garrick, B. J. , W. C. Gekler, " Temperature Coefficient Measurement by Cold Water Injection," UCLA Research Project,1962.

Garrick, B. J., " Probability Theory as a Couple Between Atomospheric Diffusion and Population," UCLA Research Project,1962.

Garrick, B. J. , et al, "By-Product and Special Nuclear Material Safety Report," September 1960.

Garrick, B. J., W. C. Gekler, " Safety and Engineering Site Study for a Nuclear Fuel Reprocessing Plant," June 1960.

Garrick, B. J., " Nuclear Fuel Reprocessing Plant Accident Analysis,"

published by the Florida Development Commission,1960.

13 0*S7P072082

r GARRICK - 14 Garrick, B. J., "Some Hydrologic and Geologic Considerations of Reactor Site Selection," Second Annual Meeting of California Association of Engineering Geologists, University of Southern California, October 1959.

Garrick, B. J. , "Research and Development in Reactor Safety,"

February 1959.

Garrick, B. J., " Radiation Barriers in a Reactor Plant," Civil Engineering, September 1958.

Garrick, B. J. , et al, " Thermal and Mechanical Design of the Westinghouse Testing Reactor Core Support Structure," September 1958.

Garrick, B. J. , "Multigroup-Multiregion Theory for the Consistent Pi Approximation to the Boltzmann Equation," ORNL 55-8-189, 1955.

t 14 0487P072082

Robert E. Henry - Consultant - Phencoenological Task Univ'ersit3 of Notre Dane Ph.D. Mechanical Engineering 1967 Rchert E. Henry is the Vice President and co-founder of Fauske &

Associates, Inc. He spent thriteen years at the AnJonne National Laboratory achieving the position of Associate Director of the Reactor Analysis and Safety Division. In this position he was responsible for all cut-of-reactor experimental work pertaining to 11guid-cetal, light-water, and gas-cooled reactor safet;y. He served on the progra coanittees for the Marviken III Large Scale criticial Flow Tests (Chaiman) and the Marviken IV Jet Irapingement Tests. In addition he was a member of the team fomed by the Nuclear Safety Analysis Center to investigate the accident at Three Mile Island.

At Fauske & Associates, Dr. Henry is now a consultant to electric power utilities and nuclear reactor equipment manufacturers in the Unt'd He has rade major contributions to the Zion States as well as overseas.

and Indian Point Near Site Studies, the Limerick Probabilistic Study, and the Swedish FILTRA Project.

Fomerly the Chaiman of the Mechanical Engineering Depament and Dean of the Graduate School at the Midwest College of Engineerf ng and author or more than 75 publications and reports on boiling and two-phase flow, Dr. Henry has given numemus lectures at universities and has participated in seminars throughout the world. He also has acted as advisor to several graduate studients in mechanical and cher.lical enginering, and is a met:bar of the American Nuclear Society.

0941Q:1

Fobert E. Henry, Vice President

/

Lc. Henry received his B.S. (1962), M.S. (1964), and Ph.D. (1967) in

- Mechanical Engineering from the University of Notre Daze where his thesis l

work, which was carried out at Argonne liational Laboratory, involved an expericental and analytical study of' one-component, two-phase critical flow. Following cocpletion of his thesis, he spent two years in the field artillery, leaving the service with the rank of captain. During this time

! he was attached to the Lewis Research Cc.cer of the National Aeronautics j and Space Administration where he was responsible for initiating experi-h, cental f acilities for the study of too-component, two-phase critical flow and one-co=ponent critical flows near the thermodynar.ic critical point.

Upon completion of his military obligation, Dr. Henry returned to h Aigonne National Laboratory to work in liquid metal 'f ast breeder reactor safety research. In particular he was responsible for out-of-ree.ctor I experiments on liquid metal superheat and sodium voiding as they relate to t-These ' experimental studies played a

^

.i hypothetical accident conditions.

nijor role in resolving the uncertainties in these phenonena. In addi-i tion, he was also responsible for the vapor explosion studies, both analytical and expericental, carried out at Argonne. It was these experi-h

=ents which eave the first clear deconstration of the physical threshold

[

for such events es " spontaneous nucleation upon contact".

In 1974, Dr. Hecry was precoted to Patnager of the Expericental Modeling Section in the Reactor Analysis and Safety Division. At this i time, the Areas Cf re Spor.sibilit y for the section Vere broadened to I include both l}T3R and LL*R safety research 'as well as scme 11:1ted I t involvement with the CCFR. As part of the LVR responsibilities he was h -

I cade Chairman of the Progra: Corr.ittee for the Marviken 111 Large Scale n .

t i Critical Flow Tests. This coc=.ittee established the origistil test catrix f for these experiments as well as the basic test configuration and f instrumentation requirements. As another part of the LVR safety studies, small scale simulant fluid experiments at ANL revealed that the "early CHF" observed in large LOCA simulation in the Seciscale facility could be

/ governed by spontaneous nucleation. This resulted in two special tests for the Seeiscale prograr., S-29-2 &nd S-29-3, for which pretest f predictions were presented to the Advisory Committee en Reactor f SafeSuards. The test results were in excellent agreement with the pretest

$ predictions, thereby resolving the issue.

h Small scale experiments on vapor explosions, under Dr. Henry's

) direction, revealed the strong effect of systec pressure en the ability to k initiate such esents. This lead to the large scale, molten salt-water f experiments t.' Ispra, which were funded by the German BMFT and the USNRC.

[

Dr. Henry was responsible for planning and supervising these experictnts

$ as well as providing the pretest predictions for the pressure level at i Tne results were in excellent which staae explosions vould be tereinated.

i agreement with the pretest predictions and this had a profound positive 5 effect en LkT. safety analyses.

t Dr. Henry was promoted to Associate Division Director of the Reactor i -

Analysis and Safety Division in March, 1979. Shortly n deer the

/

Fuelear Safety Analysis Center (NSAC) forted a team to study the TMI-2 k' accident, i.e. cause, extent of d a:. age , and "what-if" scenarios. In particular, Dr. Henry was responsible for the evaluations in in-vessel

/ secae explosion potential (negligible), in-vessel cooling of the degraded I f

core, and the potential for ex-vessel cooling if core caterial had been released from the pric.ary system.

$ in March 1980, Dr. Henry lef t Argonne National Laboratory to fore the fir: of Fauske and Associates, Inc. with Dra. Fauske and Crolmes. In this

' capacity, he was inticately involved in the Zion / Indian Point Near Site

- Study and the Limerick Probabilistic Risk Assessment. He has also been involved in the planning and interpretation of the Marviken IV Jet Ir.pingesents Tests which are currently underway.

In addition, his acadeutic background includes: Chairman of Mechani-cal Engineering Department (6-years) and Dean of the Graduate School (2-years) at Midwest College of Engineering, as well as the supervision of several Masters and Ph.D. thesis and runerous lectures at major universi-ties. A complete listing of all publications is given below.

1. "A Study of One- and Two-Component, Two-Phase Critical Flows at Lev Qualities," AK-7430,1968.

2. "Frcpagation velocity of Pressur. Waves in Gas-Liquid Mixtures,"

Coeurrent Gas-Liquid Flow, Plenum Press, 19 6.

3. "Two-Phase Critical Flow at Lev Qualities, Part 1: Experimental,"

Nucl. Sci. Eng., 4_1_, 79-91, 1970.

i

4. "Tvo-Phase Critical Flow at Lov Qualities, Part 2: Analysis," Nucl.

Sci. Eng., 4_1,, 92-98, 1970.

5. " Pressure Wave Propagation in Two-Phase Mixtures," Chem. Eng. Symp, Series, & 1-10, 1970.

t

6. "ne Two-Phase Critical Discharge of Initially Saturated or Subcooled Liquid," Nucl. Sci. Eng., H , 336-342, 1970.

7. "The Ivo-Fbase Critical Flov of One-Compenent Mixtures in Nozzics, Orifices, and Short Tubes," Trans. of ASME, J. Heat Transfer, 179-187, May 1971.

S. " Role of Surf ace in the Meacuter,ent of the Leidenfroct Tempers.ture,"

ASME Sy=pesiu: Series on Augeentation of Convective Beat and Mass Transfer, 1970.

9. "Precrure Wave Propagation to Annular Mist Flows , AIChE Sytp.

Series, 113, 67, 3S-I.7, 1971.

~

10. " Pressure-Pulse Propagation in Two-Phase One- and Two-Component Mixtures," ANL-7792,1971.

11. "Two-Phase Critical Discharge of High Pressure Liquid Nitrogen."

Proc. Third incl. Congress of Refrigeration, Washington, D.C.,1971.

j 12. "Mccentum Flux in Two-Fhase, Tvo-Component Lov Quality Flow," AIChE Sy:p. Series, M, 69, 46-54, 1972.

I 13. "A Cc,rrelation for the Minicu= Film Boiling Te:perature," AIChE Symp.

i Series, 138, 79, 81-90, 1974.

t 14. "Large Scale vapor Explosions " Proc. Tast P.eactor Safety Mtg.,

3 CONT-740401, 1974.

,I 15. " Sodium Expulsion Test for the 7-Pin Ceometry," Proc. Fast Peactor Safety Mtg., CONT-740401, 1974.

t

16. " Transient Sodium toiling Experitr.ents for Tast Reactor Safety," Sy p.

5 on Fast Breeder Reactor Development, 77th Nat. A1ChE Mt g. , Pitts-

.i burgh, FA, 1974.

l 17. " Pressure Drop and Compressible Flov of Cryogenic Liquid-Vapor

! Mixtures," Chpt. 11, Heat Transfer at Lov Temperatures, Plenus Press, 1975.

I 18. " Experimental Study of the Minimum Film Boiling Peint for Liquid-i Liquid Systems," Froc. 711th Incl. Heat Transfer Conf. , Tokyo, Japan, Septe:ber 1974.

. 19. " Incipient Superheat in Convective Sodiu= Systez," Proc Fif th Intl.

' Heat Trcnsfer Conf. , Tokyo, Japan, Septecher 1974.

20. "0FERA Single-Pin Pump-Constdown Expulsion and Reentry Test," ANL '

17, 1975.

! 21. "Ene rge tic s of Vapor Explosiens," ASME Paper 75-HT-66, presented at Nat. Heat Traasfer Conf., San Francisco, CA, 1975.

i

" Vapor Explosions of Subcooled Freon," Proc. Third Specialists Mtg.

22.

on Sodium / Fuel Interactions in Fast Reactors, Tokyo, Japan, March

. 1976.

i

23. " Nucleation Characteristics in Physical Explosions," ' roc. Third Specialists Mtg. on Sodiuz/Tuel Interactions in Fast Reactors, Tokyo, y

Japan, March 1976.

24. "SIKPLE-2: Co:puter Code for Calculation of Steady-St te Tner a1 Echavior of Fmd Sundles with Fuel Sveeping," AS. I Faper 76-HT-8, presented at Kat. Heat Tracsfer Conf., St. Louis, MO, 1976.

I 25. " Fuel Dispersal Experiments with Siculant Fluids," Proc. Mtg. Fast Reacter Safety and Tselattd Physics, Chicago, IL, October 1976.

J

26. " Cladding Relocation Experisents," Proc. Mcg. Fast P.eactor Safety and Related Physics, Chicago, IL, October 1976.

27. " Vapor Dispersal Experiments with Situlant Fluids," Proc. Meg. Yast

, Reactor Safety and Related Physics, Chicago, IL, October 1976.

28. " Analytical and Experisa_ntal Studies of Transient Tuel Freeting,"

Proc. Mtg. Fast Reactor Safety and F41ated Physics, Chicago, II. ,

1976.

29. " Experiments on Pressure-Driven Fuel Co:paction with Reactor Materi-als," Proc. Mtg. Fast Keactor Safety and Related Physics, Chicago, IL, October 1976.

i

30. " Hydrodynamic Instability Induced Liquid-Solid Contacts in Film Eoiling," ASME Paper 76-WA-25, presented at Winter Annual ASME Mtg. ,

, New York, NY, 1976.

31. " Transient Freeting of a Floving Ceramic Fuel in a Steel Channel,"

Nucl. Sci. Eng., 6_1, 310-323, 1976.

[ 32. "Non-Equilibriu Critical Discharge of Saturated Subcooled Liquid Freon-11," Nucl. Sci. Eng., H , 365, 1977.

33. "A Mechanis for Transient ' Critical Heat Flux," Proc. Mtg. Thernal Reactor Safety, Sun Valley, 1D, 1977.

34. " Vapor Explosion Potentials Under LVR Hypothetical Accident Ccndi-tions," Proc. Mtg. Ther:a1 Peactor Sa.f ety, Sun Valley, ID,1977.

35. "An Evaluation of the Fotential for Energetic Fuel-Coolant Interac-cions in Hypothetical IEFER Accidents," Thermal and Hydraulie Aspects of Nuclear Reactor Safety, 2, Licuid_ Metal _ Fast Ereeder Reactors, American Society of Mechanical Engineers, New York, h7, ed. O. C.

l ,

Jones and S. G. Bankoff, p. 223, 1977.

36. " Experiments of Quenching Under Pressures," presented at the 1978

, Intl. Heat Transfer conf., Toronto, 1

37. "An Investigation of the Minimus Film Eoiling Tecperature on Hori-

, zental Surfaces," 1rans. ASME, J. Heat Transfer, 100, May 1978, p.

l 260-267.

38. " Vapor Explosion Experiments with Subcooled Freen," AL77-43,1967.

39. " Occurrence of Critical Heat Flui. During Blowdown with Flow Rever-sal," accepted for publication in Nucl. Eng. Design.

40. " Nucleation Processes in Large-Scale Vapor Explosions," accepted for publication in Trans. ASPE, J. Heat Transfer.

t l . 41. "Ef fects of System Pressure on the Bubble Growth f rc: Highly Super-I heated Water Droplets," ASME Sy=p. Volume on Topics in Tve-Phase Eest i e Transfer and Flow, Dece:ber 1978, p. 1-10.

'. 42. "Ther=al and Hydrodynamic Interactions in a Shock Tube Configura-k tion," ASME Symp. olume on Topics in Two-Phase Heat Transfer and Tiow, Decenber 1978, p. 37-49.

43. " Effects of Dissolved Gas and Downstream Geosetry During Blowdown of a Subcooled Liquid," ASME Symp. Volume on Fluid Transients and Acoustics in the Power Industry, Dece:bar 1978, p.95-104 1

44, "Generi: Considerations of IXFER Hypothetical Accident Energetics,"

presented at the ENS /ANS Intl. Topical Meg, on Nuclear Reactor Safety, October 1978, Brussels, Belgium.

l 45. "IRFER Safety and Sodium Boiling," presented at the ENS /ANS Intl.

• Topical Meg. on Nuclear Reactor Safety, Detober 1978, Brussels,

Belgium.

46. " Fuel-Coolant Interactions in a Shock-Tube Geocatry," presented at the D;S/ANS Intl. Topical Mtg. on Nuclear Reactor Safety, October i

1978, Brussels, Belgius.

47. " Fuel-Sodiuc: Ther=al Interactions in the CAMIL TOP Safety Tests,"

f' accepted for publication at the Tourth CSNI Specialist Mtg. on Fuel-Coolant Interactions, April 1979.

l

48. "A Comparison of the Sodium-UO., Results with the Spontaneous Nuclea-j 2

tion Theor'/," accepted for publication at the Fourth CSNI Specialist i Htg. on Tuel-Coolant Interactions April 1979.

s'

49. "The Ef fect of Pressure on NACL-H 9 0 Explosions," accepted for publi-cation at the Tourth CSNI Specialist Mtg. on Fuel-Coolant Interac-tions, April 1979.

50. " Bubble Crowth During Decompression of a Liquid," sube.it ted for f presentation and publication at the Nat. Heat Transfer Conf., San p

Diego, CA, August 1979.

l

51. "Depressurization of Internally Heated Boiling Pools," subcitted for

[i presentation and publication at the Nat. Heat Transf er Conf. , San Diego, CA, August 1979.

Li=ited Distribution Report _a_

NA. _SA TMX Reports

?

1. "The Ivo-Phase Critical Discharge of initially Saturated or Subcooled Liquid," NASA TMX-52726.

in

2. "The Cocpressible Tiow of Two-Cc:ponent, Two-Phase Mixtures No::les and Orifices," SASA n'.X-52729.

! of the Leidenfrost Te: p er a-

, 3. " Role of the Surf ace in the Measure:ent ture," NAS A TMX- 5286 6.

1 1

, 4. " Pressure Wave Propagation Through A:tnular and Hist Flows," NASA

, M7.-5 2 6 52.

. S. "Two-Phase Critical Discharge of High Pressure Liquid Nitroger,," NASA TKZ-67863.

6. "Mocentu: T1ux in Two-Phase, Two-Corponent, Lew Quality T1ov," NASA TMT.-6 S03 8.

ANL/ RAS R $orta

7. "0FEFA Sin 61 e-Pin Pump Coastdown Expulsion and Esentry Test," ANL/ RAS

, 73-28.

i 6. "An Assessment of Voiding Dynetics in Sodium-Cooled Tast Reactors,"

, ANL/ RAS 74-20.

9. " Energetics of Vapor Explosions," ANL/ RAS 75-22.

i i 10. " Transient Freezing of a Floving Ceramic Fuel in a Steel Channel,"

ANL/ RAS 76-3.

11. " Upper Plenue Injection Tests No. 1 and No. 2," ANL/ RAS 76-4

12. " Vapor Explosion Expericents with Subcooled Freon," ANL/FAS 77-3.

4

'; s. An Investi3ation of the Minimum Film Boiling Te:perature on Horizon-tal Surf aces," ANL/ PAS 77-14.

, 14. "CAhEL TOP / Fuel Sweepout Sin 51e-Pin Test C2," ANL/ RAS 77-22.

15. " Transient critical Heat Flux in a 0.91 c Leng Unifor:17 Heated Test Section DuriCE Blowdown of High Pressure Treon," ANL/ RAS /LVR 77-1.

16. "Wsod's Metal Cladding Relocation Experiments," ANL/ PAS 77-37.

17. "Experi ental Sirsulation sf Boiled-Up Fuel Fools," ANL/FAS 78-24.

! 18. " Test Plan Large Scale Holten Salt-Vater Vapor Explosion Studies to l be Conducted at Ispra, Italy," ANL/ PAS /LVR 78-1.

19. "High Pressure Freon Blosdovn Studies in a 2.75 = Long Non-Uniformly Heated Round Tube," ANL/ RAS /LVR 75-3.

, 20. "Bigh Pressure Freon Elowdove Studies in a 2.75 m long Uniformly f Heated Round Tube," ANL/ RAS /LVR 76-4.

Scientific Sound Movies Produced

1. "Large Scale Vapor Explosions."

2. " Vapor Explosions in a Well Vetted System."

l _. _ __. -

3. "28-Pin Clad Relocation Expericent."

4 "The Consequences At 'INI-2 if Core Melting Occurred."

9 0

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{

NAME STANLEY KAPLAN EDUCATION Senior Post-Doctoral Fellowship, University of Southern California, 1967-1969.

Ph.D., Mechanical Engineering and Applied Mathematics, University of Pittsburgh, 1960. Post-doctoral courses in mathematics at the University of Pittsburgh and Carnegie Institute of Technology, 1960-1965.

M.S. , Mechanical Engineering, University of Pittsburgh,1958.

Graduate of the Oak Ridge School of Reactor Technology,1955.

B.S. , Civil Engineering, City College of New York,1954.

PROFESSIONAL EXPERIENCE General Summary Mathematician and engineer well know for contributions to risk analysis and reliability theory, reactor physics, kinetics, and computational technique. Specializes in probabilistic methodology; decision theory; risk analysis; and, particularly, applications of Bayes' theorem. In this connection has worked specifically and recently on developing probabilistic and decision theoretic treatments of various phases of the energy business. Included here are PRA anlayses of several existing nuclear plants, hazardous material transportation and storage, spent fuel pools, aircraft impact, offshore oil drilling (environmental risk),

underground oil storage, pipelines, and tarsands projects (business and construction risk). Developer of the DPD method for probabilistic calcula*. ions, the two-stage Bayesian technique for data analysis, the

" set of triplets," " probability of frequency," "cause table," and

" environmental table" concepts in risk analysis. Originator of the Matrix Theory of Event Trees and DPD approach to seismic risk analysis.

Chronological Summary 1977-Present President, Kaplan & Associates, Inc. , a consul ting firm specializing in risk analysis and applied decision theory.

Concurrently Adjunct Professor, Department of Chemical, Nuclear and Thermal Engineering, University of California, Los Angeles, and Associate Consultant, Pickard, Lowe and Garrick, Inc.

1975-1977 Private consultant specializing in risk analysis and decision theory.

1972-1975 Holmes & Narver, Inc. , Anaheim, California.

Director, Advanced Technology Division; Director, Systems Sciences Division; Technical Director, Nuclear & Systems Sciencer Group.

30 0487P072082

KAPLAN - 2 1971-1972 Director of Software Development, COMARC Design Systems, Inc., San Francisco, California.

1969-1971 Product Manager and Senior Staff Member, Computer Sciences Corporation, Los Angeles, California.

1967-1969 Special Research Fellow, U.S. Public Health Service at University of Southern California, Los Angeles.

1955-1967 Westinghouse Bettis Atomic Power Laboratory, West Mifflin, Pennsylvania.

Experimentalist, Experimentalist in Charge, Scientist, Senior Scientist, Fellow Scientist, Advisory Scientist.

1954 Lecturer, Department of Civil Engineering, City College of New York. .

1962-1967 Concurrently Adjunct Professor of Mechanical Engineering, University of Pittsburgh; Lecturer, Department of Mathematics, Carnegie Institute of Technology.

MEMBERSHIPS American Society of Civil Engineers.

American Nuclear Society.

Society of Industrial and Applied Mathematics.

New York Academy of Sciences.

REPORTS AND PUBLICATIONS "Seabrook Probabilistic Safedy Assessment," Public Service Company of New Hampshire, to be published in 1983.

Kaplan, S., "A Matrix Theory Formalism for Event Tree Analysis--Application to Nuclear Risk Analysis," Risk Analysis, Vol. 2, No. 1, 1982.

Kaplan, S., "On Safety Goals and Related Questions," KAI-19, Reliability Engineering, Vol. 3, 1982 Kaplan, S., " Methodology for the Zion and Indian Point Probabilistic~ Risk Assessments," proposed presentation at the Congres Annuel 1982, SFRP 1

La Comparison des Risques Associes aux Grandes Activites Humanines, i

Avignon, France, October 18-22, 1982.

31 L 7P072082 I

KAPLAN - 3 Heising, C. D. , A. W. Barsell, K. N. Fleming, S. Kaplan, and B. J.

Garrick, "A comparison of Recent Nuclear Plant Risk Assessments,"

proposed presentation at the Congres Annuel 1982, SFRP, La Comparaison des Risques Associes aux Grandes Activites Humaines, Avignon, France, Oc tober 18-22, 1982.

Kaplan, S., H. F. Perla, and D. C. Bley, "A Methodology for Seisnic Safety Analysis of Nuclear Power Plants," proposed presentation at the International Meeting on Thermal Nuclear Reactor Safety, Chicago, Illinois, August 29-September 2,1982.

Bl ey, D. C., S. Kaplan, and B. J. Garrick, " Assembling and Decomposing PRA Results: A Matrix Formalism," proposed presentation at the International Meeting on Thermal Nuclear Reactor Safety, Chicago, Illinois, August 29-September 2,1982.

Kaplan, S., "On a 'Two-Stage' Bayesian Procedure for Determining Failure Rates from Experiential Data," PLG-0191, preprint of a paper to appear in the IEEE Transactions on Power Apparatus and Systems, August 1982.

Garrick, B. J. , S. Kaplan, and D. C. Bley, "Recent Advances in Probabilistic Risk Assessnent," prepared for the MIT Nuclear Power Reactor Safety Course, Cambridge, Massachusetts, July 19, 1982.

Fleming, K. N., S. Kaplan, and B. J. Garrick, "Seabrook Probabilistic Safety Assessment Management Plan,"PLG-0239, June 1982.

Garrick, B. J. , S. Kaplan, D. C. Iden, E. B. Cleveland, H. F. Perla, D.

C. Bley, D. W. Stillwell, H. V. Schneider, and G. Apostolakis, " Power Plant Availability Engineering: Methods of Analysis, Program Planning, l and Applications," EPRI NP-2168, PLG-0165, May 1982.

" Indian Point 2 and 3 Probabilistic Safety Study," Power Authority of the State of New York and Consolidated Edison Company of New York, Inc.,

March 1982.

Lin, J. C. , and S. Kaplan, "SEIS3: A Computer Program for Seismic and Wind Risk Assessment," PLG-0222, March 1982.

Kaplan, S., "On the Method of Discrete Probability Distributions in Risk i

and Reliability Calculations--Application to Seismic Risk Assessment,"

Risk Analysis, Vol.1, No. 3,1981.

Kaplan, S. , B. J. Garrick, and P. P. Bieniarz, "On the Use of Bayes' Theorem in Assessing the Frequency of Anticipated Transients." Nuclear Engineering and Design 64, 1981.

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KAPLAN - 4 Kaplan, S., "On The Method of Discrete Probability Distributions in Risk and Reliability Calculations--Application to Seismic Risk Assessment,"

Ri sk Analysis, Vol .1, No. 3,1981.

" Zion Probabilistic Safety Study," Commonwealth Edison Company, September 1981.

Risk Analysis," PLG-0207 prepared as input to the NRC/ANS Probabilistic Risk Assessment Procedures Guide, September 1981.

Kaplait, S. , "A Matrix Theory Formalism for Event Tree Analysis--Application to Nuclear Risk Analysis," preprint of a paper to appear in Risk Analysis, Vol. 2, No.1,1983; PLG-0198, August 1981.

Stillwell, D. W., B. J. Garrick, D. R. Buttemer, G. Apostolakis, J. C. Lin, and S. Kaplan, " Analysis of the Pilgrim Nuclear Power Station Reactor Protection System," PLG-0195, July 1981.

Kaplan, S. , " Matrix Format for PRA and Its Possible Usefulness in Licensing," presented to the ACRS Subcommittee on Reliability and Probabilistic Risk Assessment, Los Angeles, California, July 28, 1981.

Kaplan, S., " Scarce Data Analysis Techniques," presented at the ANS 1981 Annual Meeting, Miami Beach, Florida, June 7-12, 1981.

Kaplan, S., G. Apostolakis, B. J. Garrick, D. C. Bley, and K. Woodard,

" Methodology for Probabilistic Risk Assessment of Nuclear Power Plants,"

draf t version of a book in preparation, PLG-0209, June 1981.

Mulvihill, R. J. , and B. J. Garrick, R. S. Hanson, S. Kaplan, Y. G. Mody, D. A. Reny, L. H. Riechers, and H. V. Schneider, " Comparative Evaluation of Boiler Availability for Intermountain Power Project," PLG-0169, April 1981.

! Apostolakis, G., and S. Kaplan, " Pitfalls in Risk Calculations,"

l Reliability Engineering, Vol. 2, No. 2, April 1981.

Kaplan, S., and B. J. Garrick, "On the Quantitative Definition of Risk,"

PLG-0196, Risk Analysis, Vol.1, No.1, March 1981.

Garrick, B. J. , S. Kaplan, and N. O. Siu, " Definition of Bounding

' Physical Tests Representative of Transport Accidents - Rail and Truck,"

PLG-0164, March 1981.

l l

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KAPL AN - 5 Kaplan, S., B. J. Garrick, and G. Apostolakis, " Advances in Quantitative Risk Assessment - The Maturing of a Discipline," IEEE Transactions on Nuclear Science, NS-28, No.1, Feoruary 1981.

Apostolakis, G. , S. Kaplan, B. J. Garrick, and R. J. Duphily, " Data Specialization Design 56, 1930.for Plant Specific Risk Studies," Nuclear Engineering and Kaplan, S., L. H. Reichers, and B. J. Garrick, " Histogram Convolution Program (HICOP)," PLG-0157, December 1980.

Hanson. R. S. , J. C. Lin, D. M. Wheeler, S. Kaplan, B. J. Garrick, D. C. , Iden, W. B. Holder, and L. G. H. Sarmanian, "An Assessment of the Reliability of Turbine-Generators," PLG-0155, November 1980.

Garrick, B. J. , S. Kaplan, D..C. Iden, E. B. Cleveland, H. F. Perla, D. C. Bley, and D. W. Stillwell, " Power Plant Availability Engineering, Methods of Analysis - Program Planning - Applications," 2 Vols. ,

PLG-0148. October 1980.

Kennedy, R. P. , A. C. Cornell, R. D. Campbell, S. Kaplan, and H.

F. Perla, "Probabilistic Seismic Safety Study of an Existing Nuclear Power Plant," Nuclear Engineering and Design, Vol. 59, No. 2, August 1980.

Kaplan, S.,

R. S. Hanson, B. J. Garrick, and J. W. Stetkar, "A Strategic Plan for July 1980.a National Data System for Electric Power Plants," PLG-0144, Garrick, T. B. J." ,Seminar:

E. Potter, S. Kaplan, G. Apostolakis, D. C. Iden, K. Woodard, and Pl ants," PLG-0141, July 1980.Probabilistic Risk Assessment of Nuclear Power Garrick, B. J., S. Kaplan, and D. C. Bley, " Seminar: Power Pl ant Probabilistic Risk Assessment and Reliability," PLG-0127, May 1980.

Garrick, B. J., and S. Kaplan, "A Conceptual Plan for a National Data System for Electric Power Plants," PLG-0131, April 1980.

Garrick, B. J., and S. Kaplan, "0yster Creek Probabilistic Safety Anal-ysis (0PSA)," presented at the ANS-ENS Topical Meeting on Thermal Reactor Safety, Knoxville, Tennessee, April 8-11, 1980.

Kaplan, S. , and B. J. Garrick, "A Strategic Plan for a National Reliability Data System," PLG-0125, March 1980.

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KAPLAN - 6 Garrick, B. J. , S. Kaplan, G. E. Apostolakis, D. C. Bley, and T. E. Potter, " Seminar: Probabilistic Risk Assessment as Applied to Nuclear Power Plants," PLG-0124, March 1980.

Kaplan, S., and B. J. Garrick, "Try Probabflistic Thinking to Improve Power P1 ant Reliability," Power, March 1980.

Apostolakis, G. , S. Kaplan, B. J. Garrick, and W. Dickter, " Assessment of the Frequency of Failure To Scram in Light Water Reactors," Nuclear Safety, Vol. 20, No. 6, November-December 1979.

Kaplan, S., B. J. Garrick, and D. C. Bley, " Notes on Risk, Probability, and Decision," PLG-0113, November 1979.

Garrick, B. J., S. Kaplan, and S. Ahmed, "A Reliability Prediction Technique for Selected Thermomechanical Components of Gas Turbine Combined Cycle Plants," PLG-0109, September 1979.

Garrick, B. J. , S. Kaplan, P. P. Bienf arz, K. Woodarr', D. C. Iden, H. F. Perla, W. Dicter, C. L. Cate, T. E. Potter, ' J. Duchily, T. R. Robbins, D. C. Bley, and S. Ahmed, "0PSA, t.,;ter Creek

• Probabilistic Safety Analysis," (Executive Summary, Main Report, Appendixes), PLG-0100 DRAFT, August 1979.

Kaplan, S., "A Numerical Method for Obtaining Probabil,ity versus Frequency Distributions from EDAC's Data for Failure of a Structural Component Under Specified Accelerations," presented to the International Conference on Structural Mechanics in Reactor Technology, Berlin (West),

Gemany, August 20-21, 1979.

Kaplan, S., and B. J. Garrick, "On the Use of a Bayesian Reasoning in Safety and Reliability Decisions--Three Examples," Nuclear Technology, Vol. 44, July 1979.

Kaplan, S., and B. J. Garrick, " Notes on Prediction of Reliability,"

PLG-0117, June 1979.

Kaplan, S., and B. J. Garrick, " Notes for a Workshop on Risk, Reliability, and Decision Under Uncertainty," presented at Battelle Northwest Laboratory, June 1979.

Garrick, B. J. , P. P. Bientarz, and S. Kaplan, " Risk Analysis of Transporting Oconee Spent Nuclear Fuel to the McGuire Nuclear Station,"

PLG-0102, June 1979.

Garrick, B. J., and S. Kaplan, " Training Engineers to be Reliability Practitioners," presented to the Sixth Annual Reliability Engineering Conference for the Electric Power Industry," Miami Beach, Florida, Ap ril , 19-20, 197 9.

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KAPLAN - 7 Kaplan, S., J. M. Vallance, and C. L. Cate, " Prediction of Frequency of Aircraf t Crashes at the Three Mile Island Site," Nuclear Safety, October 1978.

Garrick, B. J., S. Kaplan, and P. P. Bieniarz, " Input Material for Reliability Section of Westinghouse Turbine-Generator Proposal to Middle South Utilities," Decemoer 1977.

Kaplan, S., " Description of OPTSWU-1, a Program for Computing the Optimum Amounts of Separative Work to be Contracted for," November 1976.

Kaplan, S., " Notes on Pooling, Meaning of, Types of, and Advantages of Also Notes on a Bookkeeping Concept for Equitable and Visible Management of a Nuclear Pool," September 1976.

Kaplan, S., " Notes on the Concept of Inventory as it Relates to Uranium Procurement Planning," September 1976.

Garrick, B. J., and S. Kaplan, " Reliability Technology and Nuclear Power," IEEE Transactions on Reliability, Vol. R-25, No. 3, August 1976.

Kaplan, S., "UPLAN, A Decision Theoretic Tool for Uranium Procurement Planning," May 1976.

Kaplan, S., "On a Probabilistic Approach to Project Co.st Estimating,"

Consulting Engineer, February 1976.

Kaplan, S., "On a Bayesian Type Methodology for Making Accept / Reject Decisions on Offshore Lease Bids," Journal of Petroleum Technology, March 1976.

l

) Kaplan, S., and D. Trujillo, " Numerical Studies of the Partial l

Differential Equations Governing Nerve Impulse Conduction-I, the Significance of Lieberstein's Inductance Term," Journal of Mathematical Biosciences, Vol. 7, pp. 379-404, 1970.

Kaplan, S., and J. M. Vallance, " Notes on a Model for Evaluation and Optimization of Uranium Procurement Strategies for the CAPC0 Companies,"

l January 1976.

l

Garrick, B. J., and S. Kaplan, " Reliability Technology and Nuclear

Power," 1975.

i l Garrick, B. J., S. Kaplan, "A Method for Evaluating Nuclear Plant S~iting Concepts," presenteed before the Joint Committee on Atomic Development l

and Space, California Legislature, Sacramento, May 19, 1972.

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KAPLAN - 8 Garrick, B. J., S. Kaplan, and O. C. Baldonado, "On a Decison Theory Formalism for Nuclear Power Plant Siting," presented to the Conference on Unique Siting Concepts for Nuclear Power Plants, Joint Committee on Atomic Development and Space, Sacramento, California, May 9,1972.

Kaplan, S., " Variational Methods in Nuclear Engineering," Advances in Nuclear Science and Technology, Vol. Y, P. R. Greebler, editor, Acacemic Press, 1909.

Kaplan, S., "A New Derivation of Discrete Ordinate Approximations,"

Nuclear Science and Engineering, Vol. 34, No. 1, 1968.

Kaplan, S., A. J. McNabb, and M. B. Wolf, " Input-Output Relations for a Counter Current Dialyzer by the Method of Invariant Imbedding," Journal of Mathematical Biosciences, Vol. 3, No. 3,1968. ~

Kaplan, S. , A. J. McNabb, J. K. Siemsen, and D. Trujillo, "The Inverse Problem of Radiosotype Diagnosis - A Computational Model for Determining the Size and Location of Tumors," Journal of Mathematical Biosciences, Vol. 5, pp. 29-35, 1969.

Yasinsky, J. B., and S. Kaplan, " Anomalies Arising from the Use of Adjoint Weighting in a Collapsed Group Space Synthesis Model," Nuclear Science and Engineering, Vol. 31, No. 2,1968.

Kaplan, S. , " Canonical and Involutory Transformations 'of Variational Problems Involving Higher Derivatives," Journal of Mathematical Analysis ~

and Applications, Vol . 22, No.1,1968.

, Yasinsky, J. B., and S. Kaplan, "On the Use of Dual Variational Principles for the Estimation of Error in Approximate Solutions of l Diffusion of Problems," Nuclear Science and Engineering, Vol. 31, i pp. 80-90, 1968.

l Kaplan, S., " Properties of the Relaxation Lengths in Pl-Double-P1 and l Angle-Space Synthesis Type Approximations," Nuclear Science and Engineering, Vol. 28, pp. 450-463,1967.

Kaplan, S. , J. A. Davis, and M. Natel son, " Space-Angle Synthesis - An l Approach to Transport Approximations," Nuclear Science and Engineering, Vol. 28, pp. 364-375, 1967.

Kaplan, S., and J. A. Davis, " Canonical and Involutory Transformations of the Variational Principles of Transport Theory," Nuclear Science and Engineering, Vol. 28, No. 2,1967.

l i 37 j 04S7P072032

KAPLAN - 9 Yasinsky, J. B., and S. Kaplan, " Synthesis of Three-Dimensional Flux Shapes Using Discontinuous Sets of Trail Functions," Nuclear Science and Engineering, Vol. 28, pp. 426-437,1967.

Gelbard, E. M., and S. Kaplan, " Reality of Relaxation Lengths in Various Approximate Forms of the Slab Transport Equation," Nuclear Science and Engineering, Vol. 26, No. 4, 1966.

Kaplan, S., and J. B. Yasinsky, " Natural Modes of the Xenon Problem with Flow Feedback - An Example," Nuclear Science and Engineering, Vol. 25, pp. 430-438,1966.

Kaplan, S., "An Analogy Between the Variational Principles of Reactor Theory and Those of Classical Mechanics," Nuclear Science and Engineering, Vol. 23, No. 3, 1965.

Henry, A. F. , and S. Kaplan, "Some Applications of a Multimode Generalization of the Inhour Formula," Nuclear Science and Engineering, Vol. 22, No. 4, 1965.

Kaplan, S., " Synthesis Methods in Reactor Theory," Advances in Nuclear Science, Vol.111, Academic Press, June 1966.

Kaplan, S., E. M. Gelbard, " Invariant Imbedding and the Integration Techniques of Reactor Theory," Journal of Mathematical Analysis and Aoplications, Vol. 11, No. 1-3, 190s.

Kaplan, S. , A. F. Henry, S. G. Margolis, and J. J. Taylor, " Space-Time Reactor Dynamics," Proceedings Third United Nations International Conference on the Peaceful Uses of Atomic Energy, Geneva, Switzerland, 1964.

Kaplan, S., Editor and Contributor, Section 6.5, " Space-Time Kinetics,"

Naval Reactors Handbook, Vol .1.

Kaplan, S., "The Use of the Rayleight-Ritz Method in Non-Self Adjoint Problems," IEEE Transactions, Vol. MIT-12, No. 2,1964.

Kaplan, S., O. J. Marlowe, and J. A. Bewick, " Application of Synthesis Techniques to Problems Involving Time Depencence," Nuclear Science and Engineering, Vol. 18, pp. 163-176, 1964.

Kaplan, S., "Some New Methods of Flux Synthesis, Nuclear Science and

)

4 Engineering, Vol. 13, No. 1, 1962.

Kaplan, S., and G. Sonnemann, "The Methods of Finite Integral Transforms in Heat Transfer Problems," Proceedings of the International Heat Transfer Conference, Boulder, Coioraco,19o1.

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KAPLAN - 10 Kaplan, S., "The Property of Finality and the Analysis of Problems in Reactor Space-Time Kinetics by Various Model Expansions," Nuclear Science and Engineering, Vol. 9, No. 3,1961.

Kaplan, S., and S. G. Margolis, " Delayed Neutron Effects During Flux Tilt Transients," Nuclear Science and Engineering, Vol. 7, No. 3,1960.

Kaplan, S. , and G. Sonnemann, "A Generalization of the Finite Integral Transform Technique and Tables of Special Cases," Proceedings of Mid-Western Conference on Solid and Fluid Mechanics, Austin, Texas,1959.

Goldsmith, M. , T. T. Jones, T. M. Ryan, S. Kaplan, and A. D. Vorhis,

" Theoretical Analysis of Highly Enriched Light Water Moderated Critical Assemblies," Proceedings, Second United Nations International Conference on the Peaceful Uses of Atomic energy, Pager P/2376,1958.

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39 0487P072082

Nicholas J. Liparulo -

Core, Containment and Consequence Analysis Group Manager University of Pittsburgh, B.S. Chemical Engineering 1971 University of Pittsburgh, M.S. Chemical Engineering 1974 Registered Professional Engineer Mr. Liparulo is Manager of the Core, Containment and Consequence Analysis Group. He came to this position from the Safeguards Engineering Group in the Nuclear Safety Department of Westinghouse's Water Reactor Division where he was responsible for many safety related analyses.

Since joining Westinghouse PWR System Division in 1972, Mr. Liparulo has held positions of increasing responsibility. In his present assignment he is responsible for managing all aspects of the analysis of degraded core accidents including the development of analytical models, specification and direction of test programs associated with developing information on degraded core progression, application of models and experimental results to meet customer needs, programmatic and technical direction of degraded core analyses performed by Westinghouse, and interf acing with regulatory and customer personnel. His previous assignments have included the development and application of models for the Zion Probablistic Safety Studies, the Indian Point Probablistic Safety Studies, and other probablistic safety studies, as well as Ice Condenser Hydrogen Studies.

Prior to his present position, Mr. Liparulo was involved in the development and application of analytical techniques for the study of postulated accidents in nuclear power plants.

Assignments included the performance of analysen related to high energy line ruptures, preparation of plant safety reports, modeling of and performing analyses for safety related tests (e.g., LOFT, semi-scale), development of analytical models, core meltdown reactor coolant and containment thermal-hydraulic analysis, and equipment qualification modelling. He became

heavily involved in degraded core accidents during the TMI l incident during which he provided technical support. Since then, he has participated in and lead studies into the effect of non-condensible gases (hydregen and nitrogen) on condensation in the primary system, hydrogen burn flammability limits and testing, debris bed cooling limits and testing, and general code developmeet for degraded core events. Mr. Liparulo has also been the responsible individual for many safety related thermohydraulic computer codes.

Mr. Liparulo is the author or co-author of many nuclear safety related reports, and has many times presented results to the NRC, ACRS and other organizations.

NAME HAROLD F. PERLA EDUCATION B.S., Engineering, University of California, Los Angeles,1951.

Advanced course in Management and Systems Approach, University of California, Los Angeles.

PROFESSIONAL EXPERIENCE General Summary Consulting: Conducted surveys and developed programs for electric utility workshops in maintenance and outage planning. Performed cost studies and productivity improvement studies and developed engineering-construction job controls, cost controls, and management information systems.

Site Planning and Engineering: Expert witness on power plant siting alternatives with considerable experience in plant siting and project pl anning. Managed probabilistic risk assessment projects. Performed probabilistic seismic analysis of nuclear power plants, integrating seismicity and structure and equipment capacities with plant event sequences. Principal investigator for State of California study of aboveground, underground, and offshore siting concepts for nuclear power plants to be located in California. Directed seismic and geologic investigations in site selection studies for nuclear power plant in 4

Florida to serve Jacksonville Electric Authority. Directed master plan for 10,000-acre citrus processing plant in Arizona and for industrial park in New York. Was activation manager for startup of the U.S. Postal Service's bulk mail center in Chicage. Was responsible for the planning and design of research and development and support facilities, utility systems, and site developments related to sereral U.S. Atomic Energy .

Commission programs. Directed the design and structurally engineered '

nunerous industrial and commercial facilities including aircraft assembly plant, shopping centers, bridges, telephone exchanges, and offices.

Systems Analysis: Was principal investigator for unique systems studies including deployment and construction of an exploratory lunar base in

' NASA's Apollo program, an interoceanic canal using nuclear explosives and hydraulic excavation, and a major fallout shelter and parking garage

, under Manhattan. Also developed company job control and management

{ information system.

i Administration: As Vice President, Technical Services, managed business development support and long range planning activities. Directed business development in one broad segment of the firm's activities.

Directed continuing profit improvement program and was key in the i

development of project and overhead cost control and reporting systems.

As Vice President, Nuclear & System Sciences Group, controlled nuclear i

45 0437P072082

i .

PERL A - 2 f acilities design and study activities, and systems analysis projects from proposal to project completion. As Engineering Manager, directed engineering and construction project managers and multidisciplined design departments including architects and civil, electrical, mechanical, and structural engineers. Managed engineering projects and structural engineering department.

Chronological Summary 1976-Present Associate Consultant, Pickard, Lowe and Garrick, Inc.

1951-1976 Holmes & Narver, Inc.

Positions of increasing responsibilities from structural engineer to Vice President.

LICENSES -

Registered Civil and Structural Engineer, States of California and Hawaii.

REPORTS AND PUBLICATIONS "Seabrook Probabilistic Safety Assessmen't," Public Service Company of New Hampshire, to be published in 1983.

Perla, H. F., " Role of External Events in PRA Studies," to be presented at the Proposed Technical Session, ASCE, Philadelphia, Pennsylvania, May 1983.

Oconee Probabilistic Risk Assessment," a joint effort of the Nuclear Safety Analysis Center, Duke . Power, and other participating utilities, to be published in 1932.

Tennessee Valley Authority and Pickard, Lowe and Garrick, Inc., " Browns Ferry Probabilistic Risk Assessment," to be published in 1982.

Garrick, B. J., and H. F. Perla, " Quantitative Risk Management - A New Tool for the Engineering of Facilities," to be presented at the ASCE National Convention, New Orleans, Louisiana, October 26, 1982.

Kaplan, S. , H. F. Perla, and D. C. Bley, "A Methodology for Seismic Safety Analysis of Nuclear Power Plants," proposed presentation at the International Meeting on Thermal Nuclear Reactor Safety, Chicago, Illinois, August 29-September 2,1982.

Garrick, B. J. , S. Kaplan, D. C. Iden, E. B. Cleveland, H. F. Perla, D.

C. Bley, D. W. Stillwell, H. V. Schneider, and G. Apostolakis, " Power Plant Availability Engineering: Methods of Analysis, Program Planning, and Appi f cations," EPRI NP-2168, PLG-0165, May 1982.

46 0487P072082

a .

PERLA - 3 Garrick, B. J., and H. F. Perla, " Management of PRA Projects," presented at the ANS Executive Conference, Arlington, Virginia, April 4-7, 1982.

" Indian Point 2 and 3 Probabilistic Safety Study," Power Authority of the State of New York and Consolidated Edison Company of New York, Inc.,

b March 1982.

Perla, H. F., W. T. Hussey, D. H. Lougeay, and Y. G. Mody, " Cost and Controls Study of San Onofre Units 2 and 3," PLG-0227, December 1981.

" Zion Probabilistic Safety Study," Commonwealth Edison Company, September 1981.

Perla, H. F., "A Perspective of the Seismic-Initiated Hazard from the Indian Point Nuclear Generating Station," PLG-0201, September 1981.

Perla, H. F., " Outage Planning Systems--Status Overview," presented at the Tenth Biennial Topical Conference on Reactor Operating Experience, Cleveland, Ohio, August 17-19, 1981.

Perla, H. F. , " Project Plan: Probabilistic Risk Assessment, Midland Nuclear Power Plant," PLG-0150, May 1981.

Pickard, Lowe and Garrick, Inc., " Project Plan: Probabilistic Risk Assessment, Browns Ferry Nuclear Plant Unit 1," PLG-0149, October 1980.

Garrick, B. J. , S. Kaplan, D. C. Iden, E. B. Cleveland, H. F. Perla, D. C. Bley, and D. W. Stillwell, " Power Plant Availability Engineering, i

' Methods of Analysis - Program Planning - Applications," 2 Vols.,

PLG-0148. October 1980. -

l  :

Kennedy, R. P. , A. C. Cornell, R. D. Campbell, S. Kaplan, and H. F. Perla, "Probabilistic Seismic Safety Stuidy of an Existing Nuclear Power Plant," Nuclear Engineering and Design, Vol. 59, No. 2, August 1930.

Chavez, G., and H. F. Perla, " Managing and Controlling Maintenance,"

Power, June 1980.

Garrick, B. J., and H. F. Perla, " Maintenance Management in the Electric Utility Industry," presented to the International Conference on Energy Use Management, Los Angeles, Calfornia, October 22-26, 1979, i

l 47 0487P072082

. _ - . - - -- -- . .- _- . _. . = _ _ . . - _

NAME THOMAS E. POTTER EDUCATION M.S., Environmental Science, University of Michigan,1972.

B.S. , Chemistry, University of Pittsburgh,1963.

PROFESSIONAL EXPERIENCE General Summary Consultant on health and safety aspects of nuclear power. Perfoming environmental dose assessments for nuclear power plant safety analysis, environmental reports and operating reports. Assisting clients in design and implementation of radiological or environmental monitoring programs and interpretation of results. Providing independent review of in-plant radiological protection programs and effluent analysis programs.

Consultant in radiological health 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.

• 1 License administrator, plutonium fuel facility health and safety supervisor. Provided radiological safety review of major facility modifications. Used these analyses and nuclear criticality analyses perfomed by others to prepare AEC special nuclear materials and byproduct license applications. Served as corporate contact with AEC in matters related to licensing. Organized and supervised a radiological protection program for a plutonium fuels fabrication facility and hot cell facil ity. Instituted personnel monitoring programs using themoluminescent dosimetry and breathing-zone areosol sampling in 1967.

Served as secretary of a plant safety comittee which inspected all operations and reviewed detailed written procedures for operators.

Served as member of a corporate safety committee which determined corporate policy regarding health and safety matters.

Chronological Summary 1973-Present Consultant, Pickard, Lowe and Garrick, Inc.

1972-1973 Consultant to Dr. G. Hoyt Whipple, University of Michigan.

1963-1970 Nuclear Materials and Equipment Corporation (NUMEC).

License administrator, plutonium fuel facility health and safety supervisor.

48 0487P072082

( _ _ _ _ _ _ _ - __ _

's POTTER - 2 MEMBERSHIPS American Chemical Society.

American Nuclear Society.

Health Physics Society.

Certified by American Board of Herlth Physics. .

REPORTS AND PUBLICATIONS Woodard, K. , and T. E. Potter, "Co.isideration of Source Tenn in Relation to Emergency Planning Requirements," presented to the Workshop of Technical Factors Relating Impacts from Reactor Releases to Emergency Planning, Bethesda, Maryland, January 12-13, 1982.

Garrick, B. J., S. Kaplan, G. Apostolakis, D. C. Iden, K. Woodard, and T. E. Potter, " Seminar: Probabilistic Risk Assessment of Nuclear Power P1 ants," PLG-0141, July 1980.

Garrick, B. J. , S. Kaplan, G. E. Apostolakis, D. C. Bley, and T. E. Potter, " Seminar: Probabilistic Risk Assessment as Applied to Nuclear Power Plants," PLG-0124, March 1980.

Woodard, K., and T. E. Potter, " Modification of the Reactor Safety Study Consequences Computer Program (CRAC) to Include Plume Trajectories,"

prese'1ted to the 1979 ANS 25th Winter Meeting, San Francisco, California, November, 11-15, 1979.

Woodard, K., and T. E. Potter, " Assessment of Noble Gas Releases from the Three Mile, Island Unit 2 Accident," presented to the 1979 ANS 25th Winter Meeting, San Francisco, California, November, 11-15, 1979.

Garrick, B. J. , S. Kaplan, P. P. Bientarz, K. Woodard, D. C. Iden, H. F. Perla, W. Dicter, C. L. Cate, T. E. Potter, R. J. Cuphily, T. R. Robbins, D. C. Bley, and S. Ahmed, "0PSA, Oyster Cre:ek Probabilistic Safety Analysis," (Executive Summary, Main Report, Appendixes), PLG-0100 DRAFT, August 1979.

Woodard, K. , ~and T. E. Potter, "Probabilistic Prediction of X/Q for Routine Intermittant Gaseous Releases," Transactions of the American Nuclear Society, Vol. 26, June 1977.

l l

49 0487P072082 l

= ___ = - _

Dennis C. Richardson - Risk Assessment Technology Manager Penn State University B.S. Aerospace Engineering 1963 M.S. Contro1 Engineering 1955 San Diego State University, M.S. Mathesatics 1970 University of Pittsburgh, M3A 1980 Mr. Richardson has many years of professional and manageocnt experience in the nuclear field. He joined the Pressurized Water Reactor Division of Westinghouse in 1972 where he managed the Reactor Protection Analysis Group for perfoming nuclear plant safety analysis and, most recently, has mar. aged the Risk Assessment Technology Organization.

Prior to this, Mr. Richardson was with Gulf General Atomic where he worked on design of control and safety systems for the gas-cooled nucicar plant. At Westinghouse, he has participated in and directed a number of risk assessment and safety analysis studies for a wide variet;y of applications. He was a . principal investigator in both He the Zion Sta-directed the tion and Indian Point Station Reactor Safety Studies.

PRA studies for the Westinghouse Owners Group that addressed the 1 Post-TMI NUREG requirecents on emergency procedures and operator display requirements. Mr. Richardson was technical and progran He manager for the has also led the British (NWC) Reference Water Reactor Safety Study.

development of econoolc and financial rf sk assessment techniques for the use in new reacter ocdel design concepts.

Mr. Richardson is a mecber of the IEEE and ANS He isand has served reviewing on the the sec-working groups for two standards corr.ittees.

tions for the PRA manual directed by NRC to be finished in 1981. Heis author or co-author of more than 15 reports and papers dealing with ri sk assessment and various aspects of nuclear plant design 1

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@ nuclear-safety associates WALEN A PCCGEP.

EDlEATICN Dr Pcdger holds ES degrees in both Chemical and Fetallurgical Engineering frcm the University of Michigan (1939) and an MS degree in Che:nical Engineering, also fecm the University of Michigan (1940). His doctorate in Chemical Engineering was 1 awarded in 1956 by the Illinois Institute of Technology. ,

PFCFESSIONAL Dr Podger has spent seven years in private industry EXPERIENCE (pharmaceutical and nuclear) in production and plant construction,17 years in research and develep.ent at AT National Iaboratories, and 17 years in private consulting. Much of his working lifetime has been spent in the field of waste control. He is recogniz'ed internationally as an expert in the control of radicactive waste and has given papers on the subject at ntrrerous national and internaticnal :reetings in the US and half a dozen foreign countries. He has written extensively, is the author of about three dozen papers, and is the author of sections of several nuclear engineering hanc2:coks.

In the course of his career Dr Rodger was responsible for the design of the licuid waste handling facilities, both nuclear and industrial, for Argenne National Laboratory, Argonne, Illinois, and for the total waste handling facilities for the West Valley, NY, plant of Nuclear Fuel Services, Inc--the world's first ccm:ercial nuclear fuel reprocessing plant.

Dr Fodger has done extensive werk throughcut the entire nuclear fuel cycle, including research and develep: ent, design, construction, cperatien, arx!

censulting. He has also had a large amount of experience in dealing with the licensing of nuclear j facilities.

@ nuclect-safety associates WAL' ION A PCDGER po9e 2 PRCFESSIONAL For several years Dr Rodger was involved in preparing EXPERIEtCE and presenting testimony for the utility industry at (continued) the tric's Appendix I Rulemaking Hearings aimed at determining what releases are "as low as practicable" for light water reactors. This involved evaluating all possible release paths from the reactor emplex, the effective doses therefrom, the costs associated with reducing e:::nissions, and a cost-tenefit analysis of such reductions. This work was done for a group of about thirty utilities. Iater that group was recenstituted to co. ment en EPA's efforts to further limit effluents from the nuclear fuel cycle. Dr Rodger has also appeared as a technical witness in several individual licensing actions testifying on the AIAFA implications of radwaste systems.

Following his post 'IMI service at ':MI-2, which consisted largely of following the behavior of iodine in the system and assuring that no action was taken which would significantly increase the environmental release of iodine, Dr Rodger became interestod in the calculation of accident source terms, sin? the "J4I data strongly suggested that the then current :rethods of calculaticn ever-predicted the actual source term.

Since then he has prepared for Iong Island Lighting Company an extensive review of nearly 30 years of literature which pertains to release of fission products from overheated er molten fuel. Pertinent -

accident data were also reviewed and compared to calculation. He is currently serving On the Peer Review Committe for the Shoreham PFA study and is in charge of Section ll.3--Fission Product Transport--of the industry's degraded core study--IDCDR. He is alsc doing studies for the Electric Power Fasearch Institute of the release of fissien products frem both degraded and molten ccres.

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@ safety nuciocr-associates WAL m 3 A PCDGER page 3 17tILITY In addition to the Consolidated Utility Group for 'Acm CLIE m the Appendix I work was done, utilities with whom Dr.

Rodger has consulted during the past decade include:

General Public Utilities Iowa Electric Light & Power Pochester Gas & Electric Co Pacific Gas & Electric Co Boston Edison Co Iong Island Lighting Co Florida Power & Light Co Houston Lighting & Power Co Carolina Power & Light Co San Diego Gas & Electric Co Vermont Yankee Metropolitan Edison Co Jersey Central Power & Light Co PPCFESSIONAL Dr Pcdger is a Fellcw of AIChE 'and the 1981 recipient AFFILIATIONS of AIChE's Pdoert E Wilson Award. In 1960 he was Chairman of the Nuclear Engineering Division of the Institute. He is also a member of ANS and AIF. In 1959 he served as Technical Censultant to the Joint Comittee en Atemic Energy of the 86th Congress at the Hearings en Industrial Radioactive Waste Disposal, and in 1972 prepared testimony before the Joint Comittee on Atcaic Energy's Hearings on Ultimate Disp sal of Nuclear Waste. Until mid-1975, he was Chairman of the MEI N-48 Co.mittee en Ultirate Waste Disposal. He is also a merber of the MS Ad hoc Ccimittee en Reacter Accident Source Terms.

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Samuel Rothstein Chemical and Metallurgical Engineer SPECIAL QUALIFICATIONS:

over thirty years in applied metallurgy including selection of materials,-fabrication of pressure vvssels and piping, and studies of corrosion of materials. ,

EDUCATION: ,

B .S . - College of City of New York - 1940 Graduate courses and seminars in metallurgy and associated subjects gt: Brooklyn Polytechnic Institute, Columbia University,: M.I.T., Pennsylvania State University, and University of Illinoir. Professional Engineer, licensed in State of New I

York.

EXPERIENCE: '.y '

-/ r From 19711 Chemical and Metallurgical Engineer Mechanical Studies Section, Mechanical Engineering

- Dept., Consolidated Edison Co. of New York, N.Y.

\ ,

Evaluation and specification of materials for use in nuclear and conventional power plant equipment and structures. Specification of testing \and non-destructive ~ examination of components and evaluation of results. Specification and control of welding procedures. Investigation and correction of problems

/ and failures in materials.

,1970- Consulting Me tallurgist 1971 (Self-employed) New Hyde Park, N.2.

1958- Chief Metall'argist AMP Inc., York, Pa.

j ;1970 Evaluation and specification of fabrication pro-cesses, materials and surface finishes for use in industrial, space and military equipment. Monitoring metallurgical manufacturing processes. Investigation and correction of-problems and failures in materials.

1944- Materials. Consultant 1958 Fairchild Camera & Instrument Corp., Syosset, L.I., N.Y.

Evaluation and specification of materials and finishes for use in military airborne equipment.

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/ Samuel Rothstein page 2 Consultant to engineering and quality control on

- materials testing and examination. Technical direc-

/ tion of heat treating, plating and finishing opera-tions. Investigation and correction of materials and manufacturing problems and material failures.

1940- Instructor 1944 Air Forces Technical School, Biloxi, Miss.

Classroom instruction re: materials in aircraft and aircraf t engines.

PROFESSIONAL ACTIVITIES:

, American Society for Metals American Welding Society EEI - Metallurgy & Piping Task Force Metals Properties Council - Technical Advisocy Committee and Subcommittees on Nuclear Materials, Pressure Vessel Metals and Fracture Toughness Lis,ted in "American Men of Science" PU LICATIONS:

5 Years of Ion Exchange - Chemical Waste Treatment, ,

Plating, Vol. 45, No. 8, August, 1958 Gravity Gradient Stabilization System Antennae Structures, NASA 9596, December 1966 Materials Performance at Indian Point Nuclear Power Station, Symposium on Materials Performance in Operating Nuclear Systems, USAEC Conference - 730801, Nucl. Me tall. , Vol . 19, 1973 (co-author)

Contributor of chapter (s):

ASME Handbook on Water Technology (in preparation) plus many others COPYRIGHTS:

Calculator, Determination of Ms Point in Steel, 1946 PATENTS (pending):

Tubing inspection device and tube plugging device

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inTF. OF IGETH: 15 LMy 2922

-ia rri ed i.%RITAL ST ATt'S

r,'0F351CN On TRADE: Physicist "n:1, r . AT I O.N:

H}GHEST DF. GREE LOCAT]ON AND DATE FIELD OF STUDY Sci!OOL P.is er Tal 3 s liigh F.iver Fa)I s ,10 1940

" " " nene undergraduate U. of Wiscensin

1inneapolis, I1. A.E., 1947 Physics, ;bth U. of L;i nnesota

" " Ph.D., 1952 Physics,

• nth U. of *:innrsota .

d.c ! S ADi i 50R - .ichr. H. '.i i i i ams Sc}id

' i;i 5 ! 5 T I T LL : s ne Experi.:;cn:c) Determination of Cross Section Per . Unit .

.eigl e .,or the c.last) c Sc at t erm; or. Deuterons .n) ar) ions anc Inc r.3 :stic St att ering os, Protons b,. Dentrons ::s a runct :en c:,

Scattersr; An;1c and Incidcn; ilevieron Energy.

I E: PLOD!F.NT AND E). PERI E!:CE:

Pi ca: 2 No.c=ber 19// :o: Present 1

i L:ste and Address of repiover: LASL, Grcup TD-7 -

.slarec. . r ea sl ee, -r.

Supervisor.

1 4 t

.usiness:

r. Intelligence

[

I Your Position: 51 a f f ":nber Wori. per formed and er,ui, r.cnt used: Project rianager for foreign technology

na2ysis.

l F r c.n- 1 !iarch 1977 Tc. : 3 November 1977 1

~ :e n d Ad d r e s s o f E r.ni o.v e r :

LASL, Group Q- 7 l

Super.isor:  ;.;i ch a e l Stercrsen

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F.u s i r.e s s : F.cacior Saf e;y Dsvid Hali Your Positien: Sinff'% :hcr A

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Fro -  ;'ay 1957 To:  ?' arch 1977 -

nce and ;.ddres s of E ;1 oyer- tr,SL, Gre::p ?:- 2 . .__

Business: Supcrvisor- Hugh Paxton David Hall .

T.our Posa. tion: Stati. . , .v.c m.oer

!!ori. performed and ecuipment used: Theoretical studies connected with cri-ticality sa fety, paramet ers as socint ed s:ith cri tical and nearreacting critical systems, and studies of the dynamic h:havicr of supercritica systems; occasional efforts ccnnected 1 ith crit ical ar se-M ies. Safety of Rover reacto2s and test cperations. Design of and prediction snd a n a l y s :. s o:. in e L. .; ui- J ,o_ . _ experirent.

From- August 1952 To: 3957

.:ame and Address of Employer: LASL, Group -l-4 Tusiness: :: capon Design Supervisor: Art Sayer Your Tosition: Staff I' ember V ort per formed and equipment used: Thecretical s, capon design.

Frcr- Ar-il 1952 To: Julv 1952 Name- and Address of Emnloyer: Univ. of I:innesota Eusiness: 'Jucl ear Research Supervisor- John H. )])inus Your Position: Research Associ:2_t e Research witn the Univ ef "f.:wsc,tc Lori perforced and equipment u c.ed :

Van de Graaff accelerator, assisting graduate studcats.

PROFESSIONAL REFSRE."SS: David Hall H. C. Paxton R. E. Schreiber

? orris Eradbury OTHEP. RE LEVANT INF0FJ'.ATION.

K; her, Advisory Cor_ittee on Reactor Safeguards, 12/01/66 - 12/31/75

% ,her, Les Alanos Criticality Sa fety Ccati::ee, 1960 -

U. S. Ecpresentative, C:sdarache Laberat ory, France, 1905-1966; special c' 'Tri men t i .

in*erests: fast TCat:Cr safety, criticFlity f.Pfety, critica]

M. ..er, Thc "1.231i:c." C. .ittee, i.'a r b i ry t t n , iC , 9/ L.:- 2/ E i, Ti e * - ' - n Cemmi:: ee un'. a : peci ni ad hoc ;:dv: tery 1.rcup tri t;ie President 's Science Advi: or;, C c:..n i ; e e .

IM.ber, An.erican Physical Society Member, Society of the Sigma Xi Mecher, American Association for the Advancement of Science Fellow , Arerican "uc! car Society Scard of Governors, / .crict.n Nuc1 car Society 1975-397S Censultant to Advisory Cc =ittee on F.enctor Safc;:ards.

Consultant :o the Division of Reactor Development and Demonstration.

Past activities at the LASL have included theoretical nuc2 car weapon . . . .

cesarn, .r.cver reactor ces 2;n and sar. ty, e w2 ce e.xper2 cnce in cra t2 ca,. 2ty sa,ety nnd s:udies in parameters appropriate to criticality. A con:ir. scus effort s i n c e 1 c. a- i t..as been in tne general arca of tast reactor sa:ety.

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1. Lat e Janua ry- Febru;,ry, 157E, I was a r ember cf the special Emergency Starch 7 tam that was scn ic rwr:Lurn Canada to nitist she Soviet < ate 11ste in the search for and recovery of frag ents 02This Lat e11ite contained a nucl e "Cosmes 954". of created a special hazard when it returned toI the wascarth cast an advisor Yellowknife Nor thwest Territories , Canada.

for criticality and safety problems. ,

Cn May 20, 1979, I was esied to wori with the staff supporting and .. _.

2. .. ne .-ccacent at a r.r e e rcport ing . o the s

_e res 2 cent 's Cc rri s i cn on t.The chairman of this Cc mi "ile Island. This activity was practically full president of Dartmouth College. I was the principal author of time until the end of Ocicher, 1979.

the staff document "frca Alternative Event Secuences" and contributed Assessnen:

the group inowm as the Technical to other documents Task Force.

3. Publicatiens:

An assessment (a)

"US-USSR Cocperation in Fast Ereeder Ecacters:

Gil es , John L. Rand, and Killiam of the Agreement; by Paul M. 197E.

R. Stratton, Les Alames Scientific Laboratory, Sepicmher, of Energy.

Spon'.ored by the US Department cf the Destructive Energy Creat ed (b) "Estimat es of the L*prer Limit Keactor Tower E>cursicns, John S;bsequent to Festulnied Fast Straticn, Les T.

Lariins, Th mas P. McLau,chlin and Killiam LA-URR.79-534, 1979.

Alamos Sci ent i fic Laborat ory document ,

Ocicher 26, 1979, "Are Portions (c)

An article in Science magatine, R. Stra: ten, Danny of the Urals Ecally Contaminat ed?" by W.

St il l man , Sunner Barr and Harold Agnew.

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. .. .n . Cu-Accidents Involving Postu3ated Large Changes in Ecactivity (EN). Stratton, H. R. Lu ll e t i n o f t h e Am e r i c a n ? ?.v._s__i_-

cal Society, Vol. 21, f4, pp. 604-605, Los A3ames Scientific Lab. (1976).

M. R., and Analysis of the Kivi-TNT Experinent. Stratton, King, L. D. P. Trans. Amer. Eucl. Soc._, Vol. E, #1, pp. 126-127 (June 1965).

Analysis of the Ki.ri-TNT Escursion and its r.elaticn Stratten, to Forer U. R.; .L d l er ,

Reactor Accident Predictions. AIAA F. T.; Altomare, P. M.; and Peterson, D. ".

Propulsion Specialists Conference, Colorado Surines, .

CO (June 14-38, 1965).

Analysis of the Kiwi-TNT Experiment and P:noebus Conceptral Accident Study. Stratton, M. R.; Altcmare, P.; and Peterson, D. M. Chcpter IV; Sections S r.nd D, LA- 3 2 5 5-?:S .

Stratten, Analysis of Prompt Excursions in ;Simp]e Systems.

cst Pecctor Information W. R., and Colvin, T. .

p. 14, DTIE

le e ting , Chicago, Scmmaries of Papers, (TID- 7 5 4 8 ) (1957).

In: Iysis of Prompt Excursions in Simole Sy s t c:rs H.;and snd Idea 3ined Lazarus, Reactors. Stratten, U. R.; Colvin, T.

R. E. Internaticnal Conference on the Pcaceful Uses of Atomic Energy, 2nd, Geneva (1945), Proceedincs. Vol. 12, l

pp. 196-206 (1958).

1 l

Ancular Distribution of the Recction 1:e3 (d,n) Hc4 Eetween .

F.; and240 Yarnell, Y. L.; Lovberg, R.

KV and 3.56 Mev.

Phys. Rev. 90, pp. 292-297 (1953).

I Stratton, N. R.

Applications of the PAD Code to LMPER Power Transient Studies.

R.; and Peterson, D. M.

Engle, L. E.; Stratton, W.

Transactions of the ANS, 15, f2, p. E20 (N cvember 32-17, 1972).

Conceptual Prompt Pcuer Excursions in Propulsion4,Reactors.

'mer. Mucl. Soc., il, pp.

Stratton, H. R. Trans. -

1949-50 (June 1961).

Corre2 aticns of Expor:n.cntal and '.heoretica] Critical ; ta.

Comparative Rel: ability, Safety r ectors fcr Criticality E. G.; Geodwin, A.,

~:.;

Control. Taxton, H. C.; Carlson, T< o a ch , E. H.; Safonc , G . ; - __

Hansen, G. F.; Mills, C. E.;

and Stratton, M. E. LA.MS-2537 (March 1961).

Strattcn, Correlations of Experiments and Ca3culations.

W. E. Nuclear Criticality Safety Hational Topical J_ elican Mceting, Trinity and Southern Fevada Recticns,.c c .91-104,

'luclear Soci ety , Las 'lecas, HV, - roceefincs, Sandia Corporation, SC-DC-67-1305 ( De c c.~.be r 13 - 15 ,

1966).

Lazarus, R. E., and Coup 3ed Neutronic-Dynamic Froblems.

comuter nethods in Eeactor Physics, Stratton, M. R. ,2

# , X f

H. Greenspan Ed., Chapter 7, pp. 509-533, Gordon and

- Breach (1968).

Critical Dimensions of Uoreogeneous Spheres Containing 2350, 23SU, and Carbna for various C/235U Rctios W. an6E.235U Engle, L. E., and Stratton,

_nrichments.

i.A-3SS3-MS (May 15, 1968).

Critical Dimensions of Ura.nium (9 3. 5)-Graphite-!:a te r Epheres ,

L.'u':5 - 2 9 5 5 (Ma_v Stratten, U. R.

Cyllncers, anc Elabs.

1962).

Criticality Data and Facters Af f ecting Criticc2ity LA-3612 (Septem- of Single Stratton, E. R.

licmogeneous Units.

ber 22, 1967).

Laboratory Criticality Research at the Los Alamos Scientific M. R. ANS, and at the Rocky Flats Plant. Stratton, Transactions, Vol. 11, ?2, p. 690 (1968).

Crcss Section of the G12 (p,y)N13 Reaction ct LcwEev. Energies.

77, and Stratton, W. R. F

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_ v s_.

Bailey, C. L.,

c.o . 194-196 (19 5 0) .-

for the Scattering Dif ferential Cross-Section ecsurtments Erown, R. J.; Freier, G. D.;

of Protons by Deuterons.

R.; and Yarnell, J. L.

Holmgren, H. D.; Stratton, W.

Phys. Rev. 88, pp. 253-256 (1952).

Stratton, The Elastic Scattering of DeuterensG.byR.; Tritons.

Eanhin, D.; and W. R.; Freier, G. D.; Ucepin, Stratton, T. 1. rhys. Eev. SS, p. 257 (1952).

e-m

D Ct: u . ; r. , ,. R.;

Energy Tieleast from MeltCc.n Jiccidents. D. J '. . I .o r. Is l u - : s r e f : r.t i-Fngle, L. E.; and Peterson, ,- ... .

1,. 0 c ,. . Soc., .L i t . . : c :, :- r 12 c L,D n . , 1,. . .. , ,..* r e n s . ,.

• n._e r .

1973). . - -

Encrev-Release f rom Lioiten-Fuel s.

P.ecriticality accidents P.; Peterson, D. M.; and Stratton, (Ell) . McLaughl, T.

E. R. Trann. Amer. Nucl. Soc._, Vol. 18, pp. 195-199, "'

Los Alamos Scientific Lab., POS 1663, Los Alamos, (June 23, 1974).

Stratton, Explosive Energy Release from Meltdo.cn Accicents. .LNS Meeting, D. I-3 .

E. R.; Enc}e, L. B.; and Peterson, San Francisco, CA (Eovember 12-16, 1973).

Concept for Fost Feasibility Study of the :iuciecr Destruct King, Operational Disposal of the Rover ItIhAType Rocket.

Propulsion Joint P., and Stratton, U R.

L. D. Colorado S.orine.s, CO (June 14-15, Specialist Conference, 1965).

Godiva II. An Unmodernted Pulse-Irradiation H.; Stratton, Reactor.

H. F. . ; and 'Icod, T. F.; Uh ce, R.

Muclear Sci. and Enc. 8, pp. 691-703 (1900).

K ir.ie t t ,

D. P. .

Eiui-TET Experiment. Strattcn, W. R. 'ir Force /Puhlic C '..~.-a=..i"....,

r .= 3 9. c ,, ,- v. ice v".# .Cau.

.x - --

a r.'~.H."......-u-'.

., < - u d _v Cocoa Scach, FL (Decc=h e r 1-3, 1964).

cest. Tenstermacher, C. 7.; King, Kiwi Transcient Puclear M. R. AIAA Bul2etin, Vol. 2, L. D. P.; and Strettcn,

.c . 226 (1965).

Fensternacher, C. A.;

Kiwi Transient Nuclear Test (U).

P.; and Stratton, E. R. Confidential F.D ,

King, L. D.

LA-3325-MS (June 17, 1965).

AMS Trans. 13, t 2, Large Accidents. Encle and Peterson. .

o.o . 721-722 (Kovember 1970) .

, T. P.; Jackson, L.M FBR Disassembly Analyses. 14cI.a ug hl in J. J.; and Stratten, J. F.; Forehand, H. M.; Hoelling, W. R.

Los Alames Scientific Lab., Report LA-UR-70-2208, Ai:S Iieetino on Tast T.ca c tor Saferv (October 1976).

s Tem-ECMLACC--A Computer Programc Concerning the Transient .

-. r:ssioninc apneres . n.oencec LA-3115 in ':SGraphite.

(?;2 r ch

.ceratures or C. G., and Stratton, s

N. R.

Chezem, 1o,E4 2 .

A e

. . - 6 1:cutrcni.v idhoo} for r i .:i- A ' and Mi i-A2 Cperation at

- UTC r r t e r. , D. M.; G? aves, G. A.; Orndoff, J. D.;

'cr, H. C.; and Stratton, M. R.

?. u c l e r. - F -

y Aspects of the Rover Program (U). Graves, G.;

O A , arrisl P. S.; Langham, M. H.; Anderson, E.

Bei de . '

Stritton, W. P. . ; Van Dilla, M. A.; Ander-son, t. C ; Ha6 dad, E.; Pansen, G. E.; Kerr, T. C.; and Walton, E. E Secret RD, L7.- 2 4 0 9 (I*ar ch 1960) .

PAD: A One-C.Tenric.33, Coupled Ueutronic-Thernodyr.anic-Eydrof yr ' :c C _ puter Code. Fcterson, D. M.; Strat:cn, h' . R.; and * : c o. ',14n, T. P. LA-6540-;;S (Dece er 1976).

The Pajarito Dyna.mics Code with Applications to Recctor Experimentc. Strctton, W. R.; Peterson, D. M.; and Engle, L. E. Transactions of the AMS, Vol. 15, 62,

p. S19 (Hovember 12-17, 1972).

Proposal for a Pulse 6 P.cactor. Stratton, M. P. . 10-2-3117

( Au gu s t 31, 1966).

F.LC--A Computer Program for .eactor Accident Calculations.

Chezem, C. G., a.nd Stratten, W. R. LA-2920 (January 2

3n6s).

.7cactor Power Excursion Studies. Stratten, 7. P.; Fngle, L. B.; and Peterson, D. :1. International Conf erence on the Engineering of Test 7cactors for safe and Kel:cble Operation, Karlsruhe, Germany (Octobcr S-13, 1972).

Reactor Power Transient Studies. Stratton, W. R.; Engle, L. B.; and Peterson, D. M. Los Alaw.os Scientific Lab.,

1: * . LA-DC-72-660 (June 2, 1972).

Eeport on Integrity of Reactor Vessels for Light Cater Power Reactors (EN). Stratton, W. R. Euc1cer Engir,eerinc and Desian, vol. 28, -J 2 , pp. 147-195, Atom. Energy C o.T m . , Washington, D.C. (1974).

Revic. of Criticality Accidents. Stratton, W. R. LA-3611 (Ja n u a ry 1967).

Review of Cri ticality Accidents. Stratton, W. R. Procress s.

in 1:uc3 ear Encrcy, Series V. Vol. 3, pp. 163-205 (1960).

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3 Feview of Cri tica): ty :nci6ents. Criticality Control in Chemical and Metaurgical Plcnt. Stratton, W. R. Paris, Eurcoeun 1;uc'; e a r Ene rc.v. Acene), c. o . 491-533 (1961). Stratton, Rover Peactor Pouer Transient Calculations. Proceedings, nuc3 ear Propul-W. E., and Chezen, C. G. - sion Conference, ?'aval Post Graduate School, Monterey, CA. (_lo62).

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EcVer P.eactor Transient Analyses. Stratton, W. R., and Chenem, C. G. Proceedings, .'scrcspace Ruclear Safety Eational Topical :leeting, J.nerican Nuclear Society, Albucuercue, ::M (October 1-4, 1963). Classen, R.'S.; The Scattering of Protons by Tritons. D.; and Stratton, M. R. 3rown, R. J. S.; Freder, G. 7bys. Rev. 82, pp. 589-596 (1951). Sore Unstable Reactors. Stratten, W. R.  ?:-2-8535 (Jan-uary 14, 1972). Summary of Com. rents Of fered at the ::cetinns on .7ecriticality Energetics. Stratten, M. R. Los Alamos Scientific Lab., MM. LA-UR-76-SS7 (1976). Thermal Reactor Safety. .'S S Trans. 12, il, p. 11 (June

                   , cJ 6 a, l

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 ;_   e ARTHUR H. TUTHILL, P.E.

3 CHAPEL LANE RIVERSIDE, CT 06873 Materials and Corrosion Consultant Extensive experience in fabrication, use and performance of cast and wrought a!!oy steels in submarines, naval, commercial and fishing vessels, nuclear and fossil power plants, desalination plants, pumps, ofIshore oil and waterflood, refineries and sour gas well production, chemical and pulp and paper plants. Marine Assisted Naval architects, owners and shipbuilders in evaluation and selection of materials of construction for, and the solution of, troublesome galvanic and other corrosion problems on nuclear submarines, Alvin, deep submergence rescue vessels, aircraf t carriers, destrovers, LNG tankers, chemical tankers, deep tanks, offshore oil platforms, fishing vessels and pleasure craf t. Organized and led two shipbuilding industry materials seminars. Authored " Guidelines for Selection of Marine Materials", " Materials for Sea Water and Brine Recycle Pumps" and "CA706 Copper-Nicke! A!!oy Hull:: The Copper Mariner's Experience and Economics". Initiated reports from shipowners summarizing experience with materials usage to provide better feedback to marine industry on service experience. Inspection and quality assurance of naval ordnance-gun mounts, guns and turrets at Naval Gun Factory and assembly and alignment aboard CVLs during construction. Combat experience USS Miami - Pacific. Desalination Initiated and coordinated materials test programs and inspection of OSW demonstration ! plants and operating desalination plants at Freeport, Chula Vista, Key West and Caribbean. Persuaded OSW to fund and undertake the four A.D. Little Surveys of materials usage in actual operating plants to provide this industry with better feedback on service experience. Assisted domestic, European and Japanese design firms and manufacturers in selection of materials and in solution of corrosion problems on shipboard and land based desalination i plants for mid-east and western hemisphere. Co-authored " Desalination, Lower Cost Water by Proper Materials Selection" and other papers on materials for desalination. Power Provided design engineering firms, utilities and equipment manufacturers with feedback of overall industry experience, individual case histories and pertinent research data on materials for, and solution to, corrosion problems on condenser tubing, feedwater heaters, cooling towers and other plant components.

Developed and stimulated use of guidelines for designing new condensers to take full advantage of the properties of each different tubing alloy. Guidelines were fundamental to the successful introduction and promotion of Type 304 stainless steel for inland power plants, C70600 alloy for coastal plants and AL6X for severely po!!uted waters. Developed practical solutions to galvanic corrosion and biofouling problems in cooling water systems. Pulp and Paper Assisted in organizing cooperative alloy producer - pulp and paper mill corrosion test program to identify materials needed to meet more aggressive corrosion conditions arising from current government regulations regulating effluent water quality. Analyzed and reported data for Metals Subcommittee of TAPPIin five technical pafers on bleach plant materials and corrosion. Organized and conducted two pulp and paper industry corrosion and materials seminars. Nuclear (DuPont) Developed welding procedures and evaluated materials for high purity and reactor grade water. Developed procedures for minimizing and removing contamination (embedded iron) from surface of stainless steel that were the forerunner for ASTM A380. Evaluated and developed materials for water lubricated bearings and components of machinery for slitting and handling spent fuel elements in submerged processing and storage area. Chemical (DuPont) Developed welding procedures for stainless steels and nickel base alloys. Inspected and made initial analysis and reported on stress corrosion cracking of jacketed stainless steel pipe under insulation. Evaluated materials and corrosion problems in nylon, neoprene, HF and explosives plants. Refinery (Standard Oil, NJ) i Initiated cathodic protection of pipe coolers, deep well pump casings and heat exchangers and use of tantalum tubes in sulfuric acid reboilers. Qualified steels for low temperat':re service - leading to ASTM A300. In charge welder qualification and refinery weld quality. Responsible for solution of refinery corrosion and materials problems.

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     -                                   ARTHUR H. TUTHILL, P.E.

3 CHAPEL LANE RIVERSIDE, CT 06878 MANAGEMENT EXPERIENCE Managed group of nationally recognized professional materials and corrosion specialists responsible for developing markets and projects that directly supported sales function and kept nickel markets growing despite aggressive promotion of competitive materials (Inco). Organized, financed and operated small business enterprise manufacturing high alloy spare parts for refineries, chemical plants and paper mills in Gulf Coast area (Valco Engineer *mg). Supervised team of engineers and technicians in inspection of safe operating condition of refinery equipment (Exxon). Commended by Office of Defense Mobilization for excellent services rendered while in charge stainless steel section Iron and Steel Division - procurement of stainless steel to industry and Savannah River Plant (DuPont - Government: National Production Authority). EMPLOYMENT 1949 - 1941 - Tennessee Eastman Corporation, Kingsport, TN 1941 - 1946 - U.S. Navy, Naval Gun Factory, Cramp Shipyard, Pacific Theater 1946 - 1950 - Standard Oil (N3), Baton Rouge, LA 1950 - 1954 - E.1. DuPont, Wilmington, DE 1954 - 1959 - Valco Engineering, Baton Rouge, LA 1959 - 1982 - International Nickel Company, Hartford, CT and New York City EDUCATION MS. Met. Eng. - Carnegie Tech.,1946 B. ChE. - University of VA,1940 i Numerous AMA Management and Financial Educational Courses PROFESSIONAL National Association of Corrosion Engineers Society of Naval Architects and Marine Engineers Technical Association of Pulp and Paper Industry Professional Engineer (Metallurgy), LA License No. 4580 Accomplished public speaker and discussion leader on corrosion and materials l PERSONAL 1 l Born March 27, 1919, Staunton, VA, Married - 5 Children; Health - Excellent Lt. Cdr. USNR WWil Pacific Application Mountain Club - White Water and Wilderness Canoeing Rotary International l 1

U PUBLICATIONS

1. LaQue, F.L. and Tuthill, A.H., " Economic Considerations in the Selection of Materials for Marine Applications", Transactions of the Society of Naval Architects and Marine Engineers, Vol. 69,1961, pp. 619-639.

2. Weldon, B.A. and Tuthill, A.H., "The Cupro-Nickels in Desalination Plant",

Proceedings of Conference, " Role of Copper and Its Alloys in Desalination Equipment", London, 1966, Copper Development Association, 1968, Paper # 5, pp.39-47.

3. Tuthill, A.H. and Schi!!moller, C.M., " Guidelines for Selection of Marine Materials",

Presented at Ocean Science Engineering Conference, Marine Technology, Washington, D.C., June 14-17, 1965.

4. Tuthill, A.H. and Sudrabin, D.A., "Why Copper-Nickel Alloys for Desalination",

Metals Engineering Quarterly, Vol. 7 August 1967, pp.10-26.

5. Tuthill, A.H., " Marine Corrosion", Machine Design, Vol. 40, December 19, 1968, pp.117- 122.

6. Tuthill, A.H., " Future Trends in Tubing for Desalination", Chemical Engineering Symposium Series, Vol. 67, #107, 1970.

7. Todd, B., Tuthill, A.H., Bailie, R.E., " Desalination Lower Cost Water by Proper Materials Selection", Presented at Third European Symposium on Fresh Water from the Sea, Dubrovnik, Yugoslavia, September 14-17, 1970.

S. Tuthill, A.H.," Materials for Sea Water and Brine Recycle Pumps",Inco publication.

9. Tuthill, A.H., "Nsign and Installation of 90-10 Copper-Nickel Sea Water Piping Systems", Inco puulication.

10. Manzolillo, J.L., Thiele, E.W., Tuthill, A.H., "CA706 Copper-Nickel Alloy Hulls: The Copper Mariner's Experience and Economics", Trans., SNAME Vol. 84, 1976, pp. 408-432.

11. Tuthill, A.H., " Progress Report - Corrosion Test Results Phase i Bleach Plant C, D and H Stages", TAPPI Engineering Conference, New Orleans, November 1979, TAPPI Press Report-84.

12. Rushton, J.D., Geister, 3.3., Heasley, R.H., Tuthill, A.H., Edwards, L.L., " Statistical Analysis of Ef fects of Chloride, Residual Chlorine Temperature and Exposure Period on Corrosion of Bleach Plant Materials", TAPPI Engineering Conference, New Orleans, November 1979, TAPPI Press Report-84.

13. Tuthill, A.H., " Performance of Types 316L and 317L Stainless Steels in C and D Stage Bleach Plant Environments", Third International Symposium on Corrosion in l Pulp and Paper Industry Corrosion Problems, Atlanta, May 1980.

14. Tuthill, A.H., Rushton, J.D., Geisler, 3.3., Heasley, R.H., Edwards, L.L., " Corrosion Resistance of Alloys to Bleach Plant Environments, TAPPI Journal, Vol. 62, No. !!,

Novernber 1979.

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15. Tuthill, A.H.," Performance of Types 316 and 317L Stainless Steels in C and D Stage Bleach Plant Environments, Part 11 TAPPI Engineering Conference, Washington, D.C., September 1980.

16. Tuthill, A.H., " Corrosion of Types 316/316L and 317/317v Specimens in Aggressive Environments Compared to Actual Performance", Presented at Corrosion '81, Houston, Texas.

17. Lee, T.S., Tuthill, A.H., " Guidelines for the Use of Carbon Steel to Mitigate Crevice Corrosion of Stainless Steels in Sea Water", Paper #63, NACE Corrosion '82.

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0. H. Walker - Chief Huclear Engireering - Westinghouse WRD Offshore

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Powar Systeas (OPS) 1953 University of Utah.. B.S. Chemical Engineering Oak Ridge School of Reactor Technolgy, M.S. Nuclear Engineering 1954 University of Pittsburgh, 1963 PhD. Chemical Engineering Dr. Walker is currently Chief Nuclear Engineer for OPS. His responsi-bilities have included plant licensing activities, coordination of the Floating Nuclear Flant Design Report, plant safety evaluation, defini-tion of radiation scurces throughout the plant, design of plant shield-ing, definition of offsite radiation doses, and mom mcently dimetion of Probabilistic Risk Assessment work including the Zion /Indtan Point Prtbabilistic Safety studies. He has also directed the GPS evaluation of the consequences to man and environment of radioactivity released to liquid pathways as a result of postulated core melt accidents. Prior to joining OPS, Dr. Walker participated in the LOFT Program (on the staff of Phillips Petroleum Co.), where his responsibilities included expertoental and analytical development work concerning the release and subsequent behavior of fission products followf og a loss-of-coolant accident. He was involved in program canagement and planning for the Water Reactor Safety Program with Aerojet Nucler.r Ccapary, planning of the early environmental report on the LOFT Prograc, and planning of involvecent of Aerojet Muclear Company participation in the Reactor Safety Study, WASH-1400. Other experience includes work in reactor systens coolant control, raate-rial corrosion, radiation levels resulting from activated corrosion products, participation in the startup testing of Nautilus and other early naval nuclear subcarines and perfomance of these plants, while working at Bettis Atoaic Power Laboratorj for Westinghouse. Dr. Walker is a mecber of the American Institute of Chemical Engineers and it's Nuclear Division. He has served as President of the Peninsular Florida Section of the Acerican Institute of Cheafcal Engineers. He was a member of the ANS Cocs:tittee preparing ANSI Standard N-535 and has served as a ceraber of AWS Progra: Committees and the FRA Guide Review Cocnittee.

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