ML20028E163

Testimony of Bj Garrick & LS Gibson Re Analysis of Class 9 Accidents in Plant Fes.Conclusion in Fes That Risk to Public from Potential Accidents Is Small Not Invalidated by Precursor Study.W/Certificate of Svc.Related Correspondence

ML20028E163
Person / Time
Site:	Midland
Issue date:	01/13/1983
From:	Garrick B, Lauren Gibson CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.), PLG, INC. (FORMERLY PICKARD, LOWE & GARRICK, INC.)
To:
Shared Package
ML20028E139	List: ML20028E138 ML20028E147 ML20028E156 ML20028E163
References
ISSUANCES-OL, ISSUANCES-OM, NUDOCS 8301210039
Download: ML20028E163 (22)

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UNITED STATES OF AMERICA fhfM NUCLEAR REGULATORY COMMISSION E3 JM 19 /!10 :5 4 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of ) M"-

) Docket Nos. 50-329-OM CONSUMERS POWER COMPANY ) 50-330-OM

) 50-329-OL (Midland Plant, Units 1 ) 50-330-OL and 2) )

TESTIMONY OF B. JOHN GARRICK AND LOUIS S. GIBSON CONCERNING THE ANALYSIS OF CLASS 9 ACCIDENTS IN THE MIDLAND FES This is the testimony of B. John Garrick and Louis S. Gibson. Dr. Garrick is a principal of Pickard, Lowe and Garrick, Inc. and has been retained by Consumers Power Company as a consultant in the area of risk analysis for the Midland facility. Dr. Garrick's resume outlining his professional and educational qualifications is Attachment 1 to this testimony.

Mr. Gibson is employed by Consumers Power Company as the Section Head of the Nuclear Safety and Analysis Section of the Midland Project Safety and Licensing Depart-ment. He is responsible for the nuclear safety analysis for the Midland Plant including accident analysis, transient analysis and probabilistic risk assessment. Mr. Gibson's resume outlining his professional and educational qualifi-cations is Attachment 2 to this testimony.

8301210039 830117 PDR ADOCK 05000329 T PDR

INTRODUCTION AND SCOPE OF TESTIMONY The purpose of this testimony is to address Mary Sinc 2 air's Contention 13 (formerly revised new contention 3).

This contention asserts that the Midland FES is inadequate because the NRC staff utilized, in assessing the potential environmental risk associated with severe accidents, the results of a rebaselined Reactor Safety Study ("RSS") analysis performed by the NRC rather than the preliminary results of NUREG/CR-2497 " Precursors to Potential Severe Core Damage Accidents: 1969-1979" published by the Nuclear Regulatory Com-mission (the " Precursor Study").

The assessment of the potential environmental risks resulting from accidents contained in section 5.9.4 of the Midland FES is utilized to satisfy the requirements of the National Environmental Policy Act of 1969 (NEPA) and NRC commission policy (45 FR 40101) for a reasoned consideration of the environmental risks (impacts) attributable to accidents.

The results of the FES probabilistic assessment of severe accidents found in section 5.9.4.5(2) are utilized in section 5.9.4.6 as one of three justifications for the conclucinn l

that the potential environmental impacts from accidents at

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Midland are small and that there are no special or unique radio-logical circumstances about the Midland site and environs that would warrant special mitigating features for the Midland Plant.

The other two reasons are:

(a) "the fact that considerable experience has been gained with the operation of similar facilities without significant degradation of the environment."

(b) "the fact that in order to obtain a license to operate the Midland facility, Consumers Power must comply with the applicable Commission regulations and requirements."

The results of the environmental risk analysis are thereaf ter used as an input to the summary cost benefit analysis presented in chapter 6 of the FES.

The balance of this testimony is an evaluation of whether the conclusions regarding the potential environmental impacts from accidents at Midland as stated in FES sections 5.9.4.6 and 6.4.3 remain valid in light of the information contained in the Precursor Study.

DESCRIPTION OF PROBABILISTIC RISK ASSESSMENT AND ITS USE IN NUCLEAR POWER PLANT ACCIDENT ANALYSIS.

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Probabilistic risk assessment (PRA), as this term is l

used in this testimony, is the quantification of the frequency l of occurrence of different levels of damage from accidents in-volving nuclear power plants. In principle, probabilistic risk analysis admits into consideration any scenario that can be conceived involving any number of simultaneous failures

! and physical processes. Component failure rate distributions

are determined. Equipment reliability is calculated. Physical phenomena are evaluated. Finally, the probability of any accident with any number of safety system failures can be l

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calculated along with the associated release of radioactive material. Probabilistic risk assessment is a logical approach to the evaluation of radioactive hazards from nuclear reactors.

It attempts to quantify the likelihood of each possible release and, thereby, provides a means for comparing the risk of nuclear reactors with other public hazards and for identifying those components of the nuclear reactor system which most strongly affect that risk.

The NRC Staff utilized elements of PRA in Chapter 5 of the FES as one basis for evaluating the environmental effects of severe accidents.

Interest in probabilistic safety analysis has evolved since the 1960s. Analysis techniques were borrowed from statisticians and reliability engineers and developed into detailed tools for predicting failure probabilities for large, complex power plant systems. In 1972, the U.S.

Atomic Energy Commission undertook the Reactor Safety Study under Professor N. C. Rasmussen of MIT. This mammoth project (70 man-years and $4 million) tcok 2 years to complete and was definitely a turning point in the way we think about nuclear safety. It was clearly the most thorough in-vestigation of reactor safety ever conducted and produced an enormous body of technical work that will influence safety analysis and understanding for ye'ars to come. It calculated the risk probabilities and consequences from the operation of 100 current design light water reactors in the United States.

9 e-The complete report demonstrated that it is possible to methodically analyze results for policymakers and analysts alike. The finished document formed a basis for thorough discussions of risk methodology; i.e. , a center for criticism, review and improvement.

Though it is a seminal work, initially the Reactor Safety Study was widely criticized. Between release of a draft report in August, 1974, and the final version in October, 1975, comments were received from 87 organizations and in-dividuals representing government, industry, "public-interest" groups, and universities. Many of these comments had a signifi-cant impact on the final report.

Congressional hearings on the Reactor Safety Study were held in 1976, " seeking from experts with disparate views assessments of the validity of the Reactor Safety Study's conclusions and the usefulness of the study as an aid to policy-making. Congress wanted to know how the study has increased our understanding of nuclear safety and how it might be improved upon." The study's directors and authors, representatives of many of the groups who had commented on the study, and other ,

knowledgeable and interested people testified. Many criticized various aspects of the study, particularly its estimates of the uncertainty in specific results, but most reaf firmed that it was a valuable contribution to the understanding of nuclear risks.

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The most complete and even-handed ~ review of the Reactor Safety Study was conducted by the Risk Assessment Review Group chaired by Professor H. W. Lewis of the University of California, Santa Barbara. The group was organized by the NRC on July 1,1977, at the request of Congressman Morris K. Udall, Chairman of the Committee on Interior and Insular Affairs, who had held hearings on the Reactor Safety Study. It was chartered to evaluate the achievements and limitations of that study, the peer comments on the study, and the state-of-the-art of risk assessment methodology and to advise the NRC on the use of such methods in the regulatory process. The Lewis report concluded that the fault tree / event tree methodology was sound. The authors of that report looked carefully into certain statistical questions and identified several areas where in their judgment there was a lack of mathematical rigor. On the other hand, the Lewis group agreed that the lack of mathematical rigor would have little effect on actual calculations.

The Lewis report was a competent technical review of the Reactor Safety Study. It provided constructive suggestions that would enhance future PRAs. The comments favorable to the Reactor Safety Study, particularly in regard to the basic methodology, more than offset the negative criticisms. While the list of criticisms was substantial, the " bottom line" seemed to be that the methodology was sound except in the area of the quantification of the uncertainty of the results. The observation

on uncertainty has had a major influence on the methodology employed in current PRAs. It is important that neither the

Lewis report nor the NRC disavowed the fault tree / event tree methodology; rather, both find the methodology sound and encourage its further use in the regulatory process.

Of course, the Reactor Safety Study, together with its thorough review, especially the Lewis review, provided an excellent background for further development of the PRA methodology.

Contemporary full scope PRAs such as those perforraad on the specific Zion and Indian Point plants have further advanced the development of PRA techniques.

The result of these advancements is a much greater insight into the components of risk associated with the plants investigated. Perhaps the most important insight provided by the full scope risk studies is that the risk associated with nuclear power plant operation is quantifiable and it appears to be small. It is believed that the major breakthroughs in risk analysis methodology and the ability to identify contributors to risk have already occurred. The idea now is to apply the methods and fine tune the process. Most of the advances of the future should be in the quality of information supporting risk analysis, e.g. , a better understanding of the physical processes following the onset of a damaged core.

Among the important observations from the major studies are the following:

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l Public risk from nuclear power is less than the financial risk to the owner / operator of the plant.

The evidence takes the form of a demonstration that 4

nuclear plants are much more able to cope with a damaged core or even a neltdown than had been perceived. Thus, although the possibility of a melt-down remains an economic risk, the likelihood of a radioactive release threatening the public health is much smaller than had been thought previously.

Sequences contributing to risk vary depending on the figure of merit adopted. Not only is there a difference between contributors to core melt fre-quency and to health risk, but there are even dif-ferences for different types of health risk.

The risk studies have moved a long way toward dis-pelling the " China Syndrome" scenario. The evidence is very strong that basemat melt-through is not a failure mechanism that contributes to risk.

Full scope PRAs have indicated the importance of a probabilistic treatment of such external events as earthquakes, fires, flood, and high intensity winds.

In many cases, the external events are the major contributors to risk by virtue of the large uncer-tainties associated with their frequencies.

The emphasis in newer plants on safety system train independence and separation criteria is not necessarily favorable to risk reduction. While such designs favorably impact such rare events as pipe ruptures, large fires, and extensive floodings, they make it more difficult to compensate for more frequently occurring failures. That is, the absence of crossties between systems denies access, for example, to alternate supplies of coolant water. Such limitations often turn out to be more important to risk.

THE ANALYSIS OF SEVERE ACCIDENTS IN THE MIDLAND FINAL ENVIRONMENTAL STATEMENT The FES for Midland utilizes the RSS to analyze the probabilities of severe accidents. The RSS has been "re-baselined" for this purpose, as described in Appendix E to the FES.

Risk estimates for severe accidents, early fatalities and latent cancers are made in the FES, and the overall conclusion of that document is that the environmental costs associated with radiologjeal health impacts from severe accidents is "small". The remainder of this testimony discusses whether studies subsequent to the RSS, particularly the Precursor Study referred to in the contention, invalidate this overall conclusion.

THE " PRECURSOR STUDY" The Precursor Study published by the U. S.

Nuclear Regulatory Commission was in direct response to questions raised by the previously mentioned Lewis group. What

the Precursor Study attempts to do is to make a calculation of the frequency of severe core damage (SCD) accidents in nuclear power plants as a whole.

It does this by sorting through the Licensee Event Reports (LERs) for the period 1969-1979 and identifying incidents which it calls " precursors for potential severe core damage .ccidents." It then puts these precursors through a fairly 31aborate event tree type calculation procedure and comes up with an SCD frequency of 1.7 to 4.5 x 10-3 per year. This has caused considerable stir, for this number is significantly higher than that calculated in many of the industry's probabilistic risk assessments. In particular, the RSS calculated a point estimate core melt frequency of 6 x 10-5, a factor of almost 100 less. The key qu~estion then is whether the Precursor Study number invalidates the RSS and more generally the process of PRA.

To examine this, let us paraphrase the methodology and line of argument of the Precursor Study. The essence of it boils down to the following:

Up through 1979 we have had 432 years of reactor operation and one SCD accident; namely TMI. We have also had a number of "near misses"; e.g., Browns Ferry and Rancho Seco.

We assign each of these near misses a " severity factor,"

which we get from the event trees. Adding these up we 4

consider that the near misses all together are the equivalent of about one more SCD accident. So we consider that the statistical experience, through 1979, is about two SCD

events in 432 years which gives a frequency of 2/432 = 4.6 x 10-3, The way in which the Precursor Study handled the near misses can be subject to much criticism both on an engineering modeling basis (i.e., the details of the event tree / severity factor work) and on the basis of statistical logic (e.g.,

whether near misses should be counted at all). We consider much of this criticism to be valid. However, for our present purposes this is not a key issue. What is pertinent is that as of 1979 we experienced one SCD, TMI, in 432 years, whereas, RSS calculated a core melt frequency of one in 20,000 years.

The question now is what does this experience tell us about the RSS, about PRA, and about nuclear safety in general.

Evaluating the RSS result, we would point out firstly that counting all free world power reactors, there is today about 1,500 plant years of experience. Thus, our statistical evidence is now 1/l,500 rather than 1/432.

Secondly, the RSS result is for a different event than that I analyzed in the Precursor Study, core melt rather than severe core damage. Severe core damage is a superset of, and a much l

l more frequent event than, core melt. Thirdly, the RSS result is for one plant, Surrey, and cannot be used to 1

quantify contributors to risk at any other specific plant.

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With respect to the PRA process itself, we would note that what a PRA is, basically, is a way of calculating 1

the frequency of compound events from the frequencies of the

" elemental" events, which together make up the compound event.

The way in which this is done is pure logic and as such is incontestible. Thus, PRA in general, in our view, is not called into question by the Precursor Study. In any particular application of PRA, however, there can, of course, be errors in logic or arithmetic.

With respect to this latter point, it is interesting to note that all three of the events identified as primary contributors to the estimated frequency of severe core damage in the Precursor Study (Three Mile Island, Browns Ferry, and Rancho Seco) involved human errors of commission events, which are inherently difficult to anticipate and predict. At Three Mile Island, cooling systems were mistakenly turned off by an operator; at Browns Ferry the initiating event was a fire accidentally caused by a technician; and at Rancho Seco, a light bulb dropped into a circuit assembly caused the failure of nonnuclear instrumentation. In particular, the Precursor Study finds that about 38 percent of all significant precursors involved human error of some kind.

Thus, although its calculations of frequency should not be regarded as gospel, the Precursor Study does do a valuable service in calling attention to the human element in the incidents of the past. Recent policies of the Commission i

and the industry have put much attention on the human aspect

- of recator safety. Thus, assessment of risk today would include the substantial design and operational changes brought about by the three top-ranked events of the Precursor Study with a corresponding decrease in calculated risk. Indeed, the foreward to the precursor Study makes just that point.

CONCLUSION From the foregoing discussion, it is apparent that PRA is an analytical tool which is useful in a variety of contexts. One use is that found in the FES, as a technique which will aid the NRC Staff in reaching a judgment as to the environmental effects of operating.the Midland facility.

Another use of PRA is to assist utility management in de-l termining the contribution to risk of severe accident scenarios, keeping in mind that all NRC safety regulations have been satisfied. For this latter purposo, a site specific PRA, rather than the generic use of the RSS by the NRC in the Midland FES, must be conducted. Such a study is underway.

In contrast to the Midland site-specific PRA and i

i other studies subsequent to the RSS, the Midland FES uses a i generic risk study and point estimates, with a certain amount of "rebaselining", in order to draw the conclusion that the risk to the public, i.e. adverse radiological health effects, from potential accidents is "small". For i

the reasons given in the preceding sections, we believe 4

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that this conclusion is not invalidated by the Precursor Study and will be supported by the more definitive Midland PRA study now being conducted.

In addition, it should be emphasized that the results of the FES risk analysis are but one of the three bases relied on by the Staff in its evaluation of the effects of accidents at Midland. The other two bases for the Staff's conclusion are (a) "the fact that considerable experience has been gained with the operation of similar facilities without significant degradation of the environment."

(b) "the fact that in order to obtain a license to operate the Midland facility, Consumers Power must comply with the applicable Commission regulations and require-ments."

Finally, it is worthy of note that the collective actions taken to improve the safety of nuclear power plants as a result of the investigations of the TMI accident and other events have reduced the risk of severe accidents identified in both the RSS and the Precursor Study.

Recently published PRAs that reflect the latest method-ology and modifications made in response to the TMI event demonstrate that the risk to the public has been reduced and is sharply lower than was previously believed.

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. .4 Attachment 1 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,1952.

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, Browns Ferry, Sequoyah, and Seabrook. Extensive experience with hearings and the <,eneral 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 numerous workshops and training programs on maintenance, reliability, and availability for EPRI, DOE, and many utilities. Author of over 90 papers and reports on reliability and risk, nuclear power, power plant siting, and energy technology.

Adjunct Professor, University of California, Los Angeles; member of several institutional committees including the UCLA Radiation Committee, the Select Review Committee for the Clinch River Breeder Reactor, Design Review Board for the Midland Nuclear Power Plant, Direction and Control System Advisory Committee of the Governor's Emergency Task Force on Earthquake Preparedness, and Boston Edison's Nuclear Safety Review and Audit Committee. Peer reviewer of such national efforts as: (1) PRA Procedures Guide; (2) NRC human relia-bility research project; and (3) NRC NREP Procedures Guide.

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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B. JOHN GARRICK --- page 2 1955-1954 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 Professional Engineer, State of California.

Leaders in American Science (Eighth Edition).

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Attachment 2 LOUIS S. GIBSON Position: Senior Staff Engineer - Midland Project Education: - Michigan State University, 1974, Masters in Business Administration

- U.S. Navy Training, Nuclear Engineering (12 months) 1964

- University of Notre Dame, 1963, BSEE Experience: Censumers Power Company:

1980-1983 Section Head - Midland Project - Safety and Analysis Section. Responsible for ensuring proper performance and evaluation of the nuclear safety analysis aspects of the plant design including accident analysis, transient analysis and probabilistic risk assessment.

Also responsible for development of conceptual design options for resolution of fluid mechanical and nuclear safety system issues.

1977-1980 Section Head - Reactor and Control Systems.

Responsible for design' review for new power plant instrument and control systems and nuclear reactor and support systems. Conduct detailed analysis and evaluation of certain critical areas of new plant design and modifi-cations to existing plants. Supervise members of the section in carrying out the above section responsibilities.

1975-1977 Supervisory Engineer - Perform design review for new power plant systems and equipment in the mechanical and instrument and control areas.

Have conducted detailed design analysis of certain critical aspects of design of new plants and of modifications to existing plants. Pro-vide advice and assistance to the department head in the area of Quality Assurance.

i 1972-1975 Administrator of Quality Assurance - Responsible

! for developing and implementing a program for operational quality assurance for the company's central office and its two nuclear power plants.

Served as a member of the off-Site Safety and Audit Review Board for two operating nuclear plants. Participated in the review of design specifications for new nuclear plants.

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L Attachment 2 page 2 1970-1972 General ~ Engineer - Supervised preoperational, hot functional and startup testing of a large PWR nuclear plant. Qualified as Senior Reactor Operator by successfully completing examinations administered by the Atomic Energy Commission; supervised shift operations.

1967-1970 Lieutenant, U.S. Navy - Assigned to a prototype nuclear power plant which was utilized for testing and training of personnel. Certified as senior reactor operator. Responsible for electrical and instrumentation systems at the prototype, including supervision of forty technicians. I later shared responsibility with my civilian counterpart for planning and scheduling the routine repair effort; partici-pated in a major refueling and overhaul of the facility. I also served as a member of the Senior Operator Qualification Board.

1964-1967 Lieutenant, Junior Grade, U.S. Navy - Served as Reactor Control Division Officer, USS Long Beach; responsible for the maintenance, operating and testing of all the nuclear reactor in-strumentation and control systems; supervised twenty-five electronic technicians and partici-pated in a major refueling, overhaul and testing of the nuclear plants. I was qualified as senior watchstander on shift.

Additional Training 1973-1974 Graduate level courses - Thermodynamics, Systems Analysis, Michigan State University.

1981 One week training course on B&W Simulator 1981 One week CP Co. course on Effective Management

- i UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the matter of CONSUMERS POWER COMPANY Docket Nos. 50-3290M (Midland Plant, Units 1 and 2) 50-3300M 50-3290L 50-3300L I, B. John Garrick, being first duly sworn, state that my accompanying testimony addressing Mary Sinclair's contention 13 (analysis of Class 9 Accidents in the FES) is true and correct to the best of my knowledge and belief.

B. q 'Garrick a.<2 f M sa:

Subscribed and sworn to before me this /3M day of January,1983.

$5 7

b Y Notary Psblic My commission expires: 1// Y[F 7

_- _. _ . _ . )

UNITED STATES OF AP.RICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD In The Matter of ) Docket Nos. 50-329 OM

) 50-330 OM CONSUMERS POWER COMPANY )

) Docket Nos. 50-329 OL (Midland Plant, Units 1 and 2) ) 50-330 OL I, Louis S. Gibson, being first duly sworn, state that my accompanying testimony addressing Mary Sinclair's Contention 13 (Analysis of Class 9 Accidents in the FES) is true and correct to the best of my knowledge and belief.

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' Louis S. G'ibson l

l Subscribed and sworn to before me this /3* day of January, 1983. ,

0))LO . O Notary Public My consnission expires /i[b .

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gs) COME UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION g BEFORE THE ATOMIC SAFETY AND LICENSING BOARD l

l In the Matter of )

'83 JAN 19 A10 :53 .

l ) Docket Nos. 50-329-OM 7 CONSUMERS POWER COMPANY ) 50-330-OMy .. .i V LiA:

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) 50-329-OL E - N!$(C.

(Midland Plant, Units 1 and 2)) 50-330-OL CERTIFICATE OF SERVICE l

f I, Philip P. Steptoe, one of the attorneys for Consumers Power Company, hereby certify that copies of

! " Testimony of Cris Hillman and Terry Postlewait on Sinclair Contention 4", " Applicant's Exhibit 29R", " Testimony of  ;

1 l Richard D. Woods Concerning Seismic Shakedown Settlement at I the Midland Site except Diesel Generator Building", "Appli-cant's Suggestion of Mootness or, in the Alternative, Motion to Close the Record with Respect to Material False State-ment", and " Testimony of B. John Garrick and Louis S. Gibson Concerning the Analysis of Class 9 Accidents in the Midland FES" were served upon all persons shown in the attached service list by deposit in the United States mail, first '

class, this 17th day of January, 1983. l dk ii

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Thilip P. St&ptd e' i

. - - - . . - . _ . . =- .. .

l w SERVICE LIST Frank.J. Kelley, Esq. Steve Gadler,-

Attorney General of the 2120 Carter Avenue 4

State of Michigan St. Paul, Minnesota 55108 Carole Steinberg, Esq.

Assistant. Attorney General Atomic Safety &' Licensing I Environmental Protection Div. Appeal Panel i 720 Law Building U.S. Nuclear Regulatory Comm.

Lansing, Michigan 48913 . Washington, D.C. 20555 Cherry & Flynn Mr. C. R. Stephens l Suite 3700 Chief, Docketing & Services

~

3 First National Plaza U.S. Nuclear Regulatory Comm.

i Chicago, Illinois 60602 Office of the Secretary

_. ' Washington, D.C. 20555

)

Mr. Wendell H. Marshall 4625 S. Saginaw Rd. Ms. Mary Sinclair Midland, Michigan 48640 5711.Summerset Street Midland, Michigan 48640 Charles Bechhoefer, Esq.

t Atomic Safety & Licensing William D. Paton, Esq.

Board Panel Counsel for the NRC Staff l U.S. Nuclear Regulatory Comm. U.S. Nuclear Regulatory Comm.

Washington, D.C. 20555 Washington, D.C. 20555 A

Dr. Frederick P. Cowan Atomic Safety & Licensing 6152 N. Verde Trail Board Panel l Apt. B-125 U.S. Nuclear Regulatory Comm.

i Boca Raton, Florida 33433 Washington, D.C. 20555 i Lae L. Bishop Barbara Stamiris i

Harmon & Weiss -5795 North River Road 1725 I Street, NW #506 Route 3 Washington, D.C. 20006 Freeland, Michigan 48623

{ Mr. D. F. Judd Jerry Harbour Babcock & Wilcox Atcmic Safety & Licensing P.O. Box 1260 ' Board Panel ,

t Lynchburg, Virginia 24505 U.S. Nuclear Regulatory Comm.

I Washington, D.C. 20555 l James E. Drunner, Esq.

Consumers Power Company 2

212 West Michigan Avenue i Jackson, Michigan 49201 a

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