ML18003A356

Forwards Limited Appearance Statement of 781211 Prepared & Submitted by R Pollard of Union of Concerned Scientists. Purpose of Statement to Identify Technical Subjects That Deserve Addl Board Inquiry

ML18003A356
Person / Time
Site:	Harris
Issue date:	12/18/1978
From:	Roisman A National Resources Defense Council
To:	Atomic Safety and Licensing Board Panel
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ML18003A354	List: ML18003A356
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NUDOCS 7901150012
Download: ML18003A356 (77)

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Natural Resources Defense Council, Inc..

917 15TH STREET, N.W.

WASHINGTON~ D Co QOOO5 202 7/7 5000 IVcstcrn Opec Jlfcat York Opec 2$ $ 5 YALE STREET 122 EAST /2ND STREET PALO A1TOg CAL1Fe 9/$ 06 NEW YORKy N.YI 10017 December 18, 1978

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+7 212 9/9 00/9 Q4$

Members of the Atomic Safety and Licensing Board Panel and r% .+@+ p Lg the Atomic Safety and Licensing Appeal Panel 4%4A U.S. Nuclear Regulatory Commission Washington, D.C. 20555 0 0o' Gentlemen and Ladies:

Enclosed is a copy of a recent limited appearance statement prepared and submitted by Robert Pollard of the Union of Concerned Scientists at hearings being held on the Trojan Nuclear Plant operating license. No single document better illustrates the serious dilemma now faced by each of you in attempting to fulfill your statutory responsibilities at hearings where a thorough substantive input from citizen opponents of the proposed action is not presented.

I sending Mr. Pollard's statement to you not because.

am I believe he is correct, although I do, but issues trates with reference to only a few of the because it illus-relevant to deciding on whether to license a nuclear plant the kind of inquiry that could be made and which is not made absent the appearance of a person of Mr. Pollard's competence, integrity and commitment. Mr. Pollard advises me that preparation of this document required more than 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, exclusive of the time he spent as an AEC employee, examining the Trojan operating license application. No citizen group can afford to hire Mr. Pollard or anyone of his calibre to competently present even a small percentage of the legitimate issues which should be resolved prior to deciding on whether to issue a construc-tion permit or an operating license.

As I did at the ASLB Training Seminar, I again raise with you the question of how you can legally and morally allow hearings to be concluded and decisions issued when you must be aware from your own experience and after reading Mr. Pollard's statement that you do not have presented to you all of the relevant information required to make a sound decision. Surely the experiences in North Anna, St. Lucie, Shea on Harris, to mention only a few, have provided ample evidence that, in the heat and pressure of a proposed construction permit and operating qq o I ized<<~

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Members of ASLB and ASLAP

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December 18, 1978 Page Two proceeding, neither the Staff nor the applicant can. be relied upon to provide you with the whole truth about the proposed action. Without the benefit of an adequately funded citizen opponent to the proposal, no board has had or can ever expect to have an adequate record for decision.

I therefore urge all of you as a body to speak out to the Commissioners and to Congress in favor of an immediate provision for funding of qualified and relevant presentations by citizens in licensing proceedings. While I fully believe similar funding is needed to provide legal assistance to citizens in these proceedings, I believe that including that in my recommendation would substantially reduce its chances of success. Because success of this recommendation is so long overdue and so desperately needed, I only urge you to take the easy course. You, more than anyone, need the benefit of funding of technical experts for citizens. Please make the effort needed to obtain that benefit.

Sincerely,

/(~

Z. Roisman enclosure

LIMITED APPEARANCE STATEMENT BY ROBERT D. POLLARD UNION OF CONCERNED SCIENTISTS DECEMBER ll, 1978 Prepared for Presentation to the Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission In the Matter of Portland General Electric Company, et al.

Trojan Nuclear Plant Docket No. 50-344 (Control Building Proceedings)

INTRODUCTION My name is Robert D. Pollard. I am presently employed as a Nuclear Safety Engineer by the Union of Concerned Scientists (UCS).

My business address is 1025 15th Street, N.W., Washington, D. C.

20005.

UCS, a non-profit, public interest organization, is a coali-tion of scientists, engineers and other professionals supported by contributions from over 50,000 members of the public.-

areas of interest 'are the health, safety, environmental UCS'rimary and national security issues posed by civilian nuclear reactor development and nuclear weapons proliferation. UCS has published numerous technical reports on various aspects of nuclear technology and has been involved in a number of proceedings. before the Nuclear Regulatory Commission (NRC). UCS is not opposed to nuclear power per se. Rather, UCS is in favor of resolving known safety hazards before nuclear plants are licensed. I have concluded that this was not the procedure used to license the Trojan Nuclear Plant.

The purpose of this limited appearance is as stated in my letter of December 6, 1978 to the Board: to identify specific technical subjects that deserve additional inquiry by the Board.

My attention was drawn to this proceeding by requests from one of the parties, Ms. Nina Bell, that I agree to be an expert witness in this proceeding. 'he principal reason I declined those requests is that my observation of many prior proceedings has dictated the.

conclusion that my participation as an expert witness would be

unlikely to change the. outcome. Future actions by the Board in this proceeding may alter this conclusion.

Officials of the Nuclear Regulatory Commission and other nuclear power advocates are fond of describing a licensing process that is open to the public, free of economic and political con-siderations, and dedicated solely to protecting the health and safety of the public. Unfortunately, such pronouncements are not statements of fact about an existing decision-making process. At best, they are statements of goals that are far from being achieved.

However, since I remain hopeful that this situation can change for the better, I decided to make this. limited appearance statement.

I appreciate the courtesy extended by Ms. Bell in delivering this statement to the hearing location. However, in the interest of fairness, I requested that the shipping carton not be opened until all parties had the opportunity to receive their copies simultaneously, as would be the case if I were present.

PROFESSIONAL QUALIFICATIONS I am aware that in making a limited, appearance it is not necessary to present my professional qualifications.. However, since I am recommending further inquiry by this Board, I believe the statement of my qualifications would help the Board decide what weight should be accorded my statements and recommendations.

My formal education in nuclear technology began in May 1959 when I was selected to serve as an electronics technician in the nuclear power program of the United States Navy. After completing the required training, I became an instructor responsible for teaching naval personnel both the theoretical and practical aspects of operation, maintenance and repair of naval'nuclear power plants.'rom February 1964 to April 1965, I served as the senior reactor operator and supervised the react'or control division aboard the U.S.S. Sargo, a nuclear-powered submarine. In 1965, I was honorably discharged from the U. S. Navy and attended Syracuse University, where I received the degree of Bachelor'of Science macCna curn laude in Electrical Engineering in June 1969.

In July 1969, I was hired by the U. S. Atomic Energy Commis-sion (AEC) and continued as a technical expert with the AEC and its successor the NRC until February 1976. After joining the AEC, I studied advanced electrical and nuclear engineering at the Graduate School of the University of New Mexico in Albuquerque.

I subsequently advanced to the positions of Reactor Engineer (Xnstrumentation) and Project Manager.

As a Reactor Engineer assigned to the Electrical, Instru-mentation and Control Systems Branch, I was primarily responsible for analyzing and evaluating the adequacy of the design of reactor protection systems, control 'systems and emergency electrical power systems in proposed nuclear facilities. It was in. this capacity that I was assigned to review the operating license application for the Trojan Nuclear Plant. The specific subject matter which was assigned to the Electrical, Instrumentation and Control Sys-tems Branch is that discussed in Sections 3.10 .and 3.11 and Chapters 7 and 8 of the licensee's Final Safety Analysis Report and the Staff's Safety Evaluation Report. Copies of the "job description" for the position of Reactor Engineer (Instrumenta-tion) and comments by Dr. Joseph M. Hendrie on my performance in that position are attached (Enclosures 1 and 2, respectively).

Xn September 1974, I was promoted to the position of Project Manager and became responsible for planning and coordinating all aspects of the design and safety reviews of applications for licenses to construct and operate several commercial nuclear power plants. When I resigned my position with the NRC in February 1976, I was serving as Project Manager for the review of the following nuclear power plants which, like the Trojan Nuclear Plant, are Westinghouse-designed pressurized water reactors:

Xndian Point Unit 3 in New York; Comanche Peak Units 1 and 2 in Texas; Catawba Units 1 and 2 in South Carolina; and McGuire Units 1 and 2 in North Carolina. Copies of the "job description" for

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the position of Project Manager and the last NRC appraisal of my performance in that position are attached (Enclosures 3 and respectively).

I am a member of the Institute of Electrical and Electronics Engineers (IEEE). I have served as the NRC representative on various IEEE committees. that developed some of the IEEE standards used by the NRC to evaluate the safety of nuclear power plants.

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TECHNICAL SUBJECTS DESERVING ADDITXONAL INQUIRY Time precludes my discussing in detail all of the specific technical subjects that I believe deserve additional inquiry by the Board if this limited appearance statement is to be delivered by December 11th. Therefore, I have chosen just a few examples which support my conclusion that, if these matters are left un-resolved, interim operation of the Trojan Nuclear Plant will pose an undue risk to the health and safety of the public.

Seismic Qualification of Safet -Related Electrical E uipment Based on the Board's order, "Order Regarding Conclusion of Evidentiary Hearing on Xnterim Operation," dated November 6, 1978, it is clear that the Board is of the opinion that there is only one issue remaining. Before turning to other issues deserving the Board's attention, I will address this one issue to the extent that it relates to the seismic qualification of safety-related electrical equipment.

The seismic qualification of safety-related equipment is of vital importance to public health and safety. Xf equipment needed to protect the public cannot withstand the effects of an earth-quake, then the probability of a catastrophic accident affecting the public is about equal to the probability of an earthquake occurring. Since the Staff's goal is to assure that the proba-bility of a nuclear plant catastrophe is no greater than 10 -6 to 10 per reactor year and since no competent witness would

1 testify that it is possible to predict earthquake probability in this range with the high level of confidence needed,- the only safeguard available is proper seismic qualification.

The regulations applicable to seismic qualification include:

(1) General Design Criterion 2 of Appendix A to 10 CFR Part 50 which requires that equipment important to safety be designed to withstand natural phenomena such as earthquakes, (2) IEEE Std 279 which is incorporated in 10 CFR '50.55a(h) and which requires qualification of safety-related equipment, and (3) Criterion III of Appendix B to 10 CFR Part 50 which requires that design control measures provide for verifying the adequacy of design such as by the performance of a suitable testing program.

The input to the Staff's Safety Evaluation Report for the Trojan plant from the Electrical, Instrumentation and Control Systems Branch was prepared by me. After review and approval by my Section Leader, Faust Rosa; the Branch Chief, Thomas Ippolito, and the Assistant Director for Reactor Safety,, Victor Stello, Jr.,

it was sent to the Project Manager, J. M. Cutchin. A copy of this transmittal letter, pages 7 and 10 of its Enclosure 1 2 , and its 1 are attached (Enclosure 5).

I direct the Board's attention particularly to the second paragraph of the transmittal letter, Sections 7.8 and 8.3.2 of the SER input, and Enclosure 2, "Referenced Topical Reports Not

2. Because of the size of that enclosure, I have copied only the two pages that are relevant to the discussion of seismic qualification. Complete copies are available to the public in the NRC Public Document Room in Washington, D. C.

Reviewed or Reviewed and Found Unacceptable.". It can be seen that the seismic, radiation, and environmental qualification areas are listed among those topics which were either found unacceptable as bases for a favorable Staff evaluation or had not yet been reviewed by the Staff.

The official (i.e., publicly disclosed) version of the Trojan Safety Evaluation Report (SER) was published by the Staff on October 7, 1974. Examination of the official SER discloses the following facts. Section 7.8 is essentially identical to the SER input supplied by the Electrical, Instrumentation and Control Systems Branch (EI&CSB) in April 1974. However, Section 3.10, "Seismic Qualification of. Seismic Category I Instrumentation and Electrical Equipment," contradicts the earlier EI&CSB input to the SER in stating that the Staff had concluded that the seismic qualification program was acceptable. Similarly, Section 8.3.2,,

"Seismic Qualification of Engineered Safety Features Switchgear,"

of the official SER differs in substantive respects from the earlier EI&CSB input to the SER.

Supplement No. 1 to the Safety Evaluation Report was issued by the Staff on November 21, 1975, the same day on which the

3. I had been promoted to the position of Project Manager in September 1974 and relieved of my previous responsibilities for review of the Trojan Nuclear Plant.,

Trojan Operating License was granted. Sections 3.10, Seismic Qualification, and 3.11, Environmental Qualification, simply refer to Section7. 8, Seismic, Radiation and Environmental Qualification, of the same document.. There are at least two significant points to be noted about Section 7.8 of SER Supplement No. 1. First, the plant is allowed to begin operation even though t'e Staff states that it cannot find that certain relays satisfy the seismic qualification criteria and that these relays will have to be replaced after the commencement. of operation (page 27A) . Second, although the title of Section 7.8 includes the subjects "Radiation and Environmental Qualification" and Section 3.11 of the SER promised further discussion of environmental qualification in a SER supplement, the text never even mentions these subjects.

consider it highly probable, if not a certainty, that a contested operating license hearing would have prevented this blatant violation of the Commission's regulations and detected the obvious gaps in the St'aff's review of equipment qualification.

With the above as an introduction, l will now proceed to explain the Staff's review process and the current status of the seismic qualification of safety-related equipment in the Trojan Nuclear Plant.

4. Note that ject to a if issuance of the operating license had been sub-contested hearing on some safety issues, the license could not have been issued on the same day as the SER Supplement.

contention during As it the was, the sole issue admitted into operating license review stage related to whether geothermal energy obviated the need for the Trojan Nuclear Plant. That one issue was promptly disposed of and the ASLB decision authorizing issuance of an operating license was rendered on February 4, 1974, almost two years before the license actually was granted.

When a subject was being reviewed by the Staff on a generic basis, the Staff member assigned to review a license application that refexenced the generic submittal from the vendor was directed to make no review of the generic submittal. 5 This is the reason I prepared the second enclosure to the April 19, 1974 letter transmitting the EIECSB input to the SER to the Project Manager.

It was intended to inform the Project Manager, among others, that (1) Westinghouse topical reports had been referenced in the Trojan Final Safety Analysis Report, (2) the topical reports contained information necessary to support issuance of an operating license for Trojan, and (3) the review of the topical reports either had not been completed or had been completed and the report judged unacceptable. It is apparent that my warnings had no effect.

According to a Staff report, NURgG-0390, "Topical Report Review Status," dated October 15, 1978, the current status of the five Westinghouse topical reports is about the same as or worse than the status described in Enclosure 2 to the EI&CSB letter of April 19, 1974. The current status is as follows:

1. WCAP- 7821 (NP) The Staff received additional informa-tion by letter dated September 29, 1978. The review is scheduled to be completed by January 1, 1978.

5. The obvious exception, not applicable here, is the case where the same individual had been assigned to review both the generic submittal and a license application referencing that generic submittal.

6. The designations (NP) and (P) following the report numbers identify, respectively, the non-proprietary and proprietary versions of the topical report.

2. WCAP-7744 (NP) and WCAP-7410-L (P) still under review.

The most recent request for additional information was sent by the Staff to Westinghouse on September 29, 1978.

Another Staff request for additional information was scheduled for October 31, 1978. The review is expected to be completed by January 1, 1979.

3. WCAP-7672 (NP) and WCAP-7488-L (P) Accepted March 6, 1974, provided the Safety Analysis Report includes a discussion of qualification, connection, independence and safety function.

4. WCAP-7705 "Not Accepted." According to NUREG-0390, this "...means that the topical report has been reviewed by the staff and has been found not to be acceptable for reference."

5. WCAP-7819 (NP) and WCAP-7506-L (P) - Accepted on Septem-ber 31, 1974. However, the Safety Analysis Report must show adequate separation of connections.

If, in its review of the Trojan application, the Staff simply overlooked the fact that review was incomplete or unsatisfactory on the subject, of equipment qualification, Trojan should remain shutdown until the Board can elicit evidence demonstrating adequate qualification. If the Staff claims that it relied on plant specific information other than these topical reports, the Board should require the production of a Staff witness capable of identifying the specific document(s) relied upon and the basis for concluding that the Westinghouse-supplied equipment is seismically qualif ied (as well as environmentally qualif ied) in spite of the fact that the generic qualification programs of Westinghouse remain unreviewed and/or unacceptable to the Staff.

For the balance-of-plant safety-related equipment (i.e.,

equipment supplied by vendors other than Westinghouse), it appears from the SER and Supplement 1 to the SER that no Staff review of the seismic qualification (or, for that matter, the

radiation and environmental qualification) has yet taken place, other than the spot,-check which I performed.

I requested information from Portland General Electric (PGE) concerning the seismic qualification of the balance-of-plant safety-related equipment. This request led to the submission of a sheet of paper which, as I recall, 7

appeared to be a page from a relay manufacturer's sales catalog. It contained two columns of data. One column listed the model or stock numbers of. the relays and the other column listed the purported seismic qualifi-cation level of the relay. Since I could not. relate the part numbers to any data I had, I could not determine which, if any, of those relays were used in safety systems. Therefore, I asked PGE to identify which relays were used in which safety circuits and to indicate the acceleration the relays would experience at their respective mounting locations during a safe shutdown earthquake.

I was surprised when PGE replied that some relays used in safety circuits would experience earthquake-induced vibrations exceeding their seismic qualification level. In the case of other relays, PGE could not determine whether they were seismically qualified. The reason given for this was that during seismic qualification testing of the cabinets, no accelerometers had been installed at the relay's mounting location in the cabinets. This

7. Huch of the activity described in this paragraph was conducted by hour-long conference calls. Therefore, documents to verify this part of my statement may have been submitted informally (i.e., not docketed and retained in files).

means they had no knowledge of the acceleration to which the relays would be subjected during an earthquake. I have. noted earlier the action that the Staff took--the operating license was issued with the condition that the unqualified relays be replaced or removed from safety circuits within six months.

t I performed no similar spot-check to determine whether the balance-of-plant safety-related equipment had undergone adequate environmental qualification. This was not then and is not now unusual. The NRC adopted the AEC's policy of "limited self-regulation." That phrase means that the principal source of assurance that the Trojan Nuclear Plant has been designed and constructed safely is Portland General Electric Company and its contractors such as Westinghouse and Bechtel. In view of the obvious inadequacy of the Staff review, I recommend that the Board conduct further inquiry into the subject of seismic quali-fication of safety-related equipment, which is a part of the sole issue the Board has determined remains. In addition, I recommend further inquiry into the subject of environmental 4e

'ualification-(which includes radiation qualification).

If the Board adopts the first'r both of these recommenda-tions, I further recommend that the Board require conformance with IEEE Std 344-1975, as endorsed and modified by Regulatory Guide 1.100, and IEEE Std 323-1974, as endorsed and modified by Regulatory Guide l. 89, as the method of demonstrating conformance with the Commission's regulations.

The Board can note'rom Section 3.10 (pages 3-24) and Appendix C (page C-7) of the official SER for Trojan that the

1971 version of IEEE Std 344 was used in the safety review 'of seismic qualification. During the brief period I was assigned to the Division of Reactor Standards, I was assigned responsibility to prepare a Regulatory Guide endorsing IEEE Std 344-1971. Several members, of the Staff had strong technical arguments that this standard did not prescribe an acceptable seismic qualification piogram. I encountered strong resistance from management officials to my suggestion that we not issue a Regulatory Guide endorsing IEEE Std 344-1971. The principal reason for this resistance was a previous agreement with the nuclear industry's standards com-mittees that if the industry developed a standard, the agency would endorse its use in the licensing process. Fortunately, at least in this specific case, the technical arguments prevailed and IEEE Std 344-1971 was never endorsed by a Regulatory Guide.

The Board can note from Regulatory Guide 1.100 that IEEE Std 344-1975 is an ancillary standard of IEEE Std 323-1974, "IEEE Standard for Qualifying Class lE Equipment for Nuclear Power Generating Stations." IEEE Std 323-1974 is endorsed, with exceptions, by Regulatory Guide 1.89. However, if this standard was used in the licensing review of Trojan (the official SER and SER Supplement No. 1 do not identify the use of any standards pertaining to environmental qualification), it had to be the 1971 version.

As in the case of IEEE Std 344, the 1971 version of IEEE Std 323 was. never endorsed by a Regulatory Guide. IEEE Std 323-1971 was issued in April 1971. In July 1971, Dr. Stephen Hanauer

wrote a letter that expresses in a succinct manner the reasons that IEEE Std 323-1971 was never endorsed by a Regulatory Guide.

(A copy of the letter is attached as Enclosure 6.) Dr. Hanauer stated, in part, "I cannot find a single redeeming feature in this worthless document." The Staff has, on several construction permit reviews, stated that unless the applicant agreed to comply with the provisions of IEEE Std 323-1974, the Staff would add a

'ondition in the construction permit requiring compliance.

I share the Staff's view that compliance with IEEE Std, 323-1974 is of vital importance to protecting the health and safety of the public. Therefore, I recommend that the Board require, prior to allowing resumption of operation, a showing that the seismic and environmental qualification of the safety-related equipment in the Trojan Nuclear Plant meets or is equivalent to the provisions of IEEE Std 323-1974 and IEEE Std 344-1975 as endorsed, with exceptions, by Regulatory Guides 1.89 hand 1,100<

respectively.

Since Regulatory Guides are not regulations, requiring con-formance with Regulatory Guides 1.89 and 1.100 (and the IEEE standards they endorse) is not an attack on the regulations.

These two Regulatory Guides define what the Staff believes to be an acceptable way of meeting the regulations applicable to equip-ment qualification. Therefore they can be used to judge the adequacy of the equipment qualification program used for Trojan.

If they are so used, I feel confident that the Board will agree with my conclusion that the safety-related equipment in the Trojan

Nuclear Plant has not been demonstrated to be seismically or environmentally qualified; Therefore, since the regulations are 8

not met, the plant must remain shutdown.

Fire Protection The inadequacy of the fire protection provided for the Trojan Nuclear Plant is another specific subject that deserves additional inquiry by the Board. The Staff has determined that a fire in the Trojan plant may destroy all methods of achieving and maintaining safe shutdown conditions. Therefore, the Staff has required that an alternate or dedicated shutdown system be installed in the Trojan Nuclear Plant. The Staff reported to the Commission that this system will be installed by June 1979. Nevertheless, the Staff intends to allow the Trojan plant to resume operation based on the alleged low probability of occurrence of a fire that would prevent achieving and maintaining safe shutdown conditions.

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Attached. is a copy of that part of the Staff's testimony before the Commission which confirms the accuracy of the above statements (Enclosure 7) . Since the Commission's regulations pertaining to fire protection and safe shutdown do not allow operation of a plant if a fire could destroy all methods of achieving and maintaining safe shutdown conditions, the Trojan Nuclear Plant should not be allowed. to resume operation until this hazard to the health and safety of the public is resolved.

Furthermore, the Board should examine the criteria being used

8. l discuss the inadequacy of the direct testimony (in this proceeding) on the subject of equipment qualification later in this statement. See page 23.

by the Staff as a basis for approving the design og the dedicated I

shutdown system to be installed in the Trojan Nuclear Plant.

Nuclear plants must be designed to minimize the probability and effects of fires and explosions in order to provide adequate protection for the health and safety of the public. (See Criterion 3 of Appendix A to 10 CFR Part 50.) The design'f systems used for achieving and maintaining safe shutdown conditions must meet the single failure criterion. (For example, see General Design Criteria 17 and 34 of Appendix A to 10 CFR Part 50.) Therefore, the design of the dedicated shutdown system must include a suf-ficient amount of independent equipment so that, after discounting all equipment that could be destroyed by a fire, the remaining equipment meets the single failure criterion. The Staff has pre-viously stated in another proceeding that its fire protection evaluations have the goal of determining "whether there is reason-able assurance that at least 'one method of achieving and maintain-ing safe shutdown is independent of the influence of the postu-lated fires." (Memorandum for Chairman Hendrie, et al., from Edson G. Case, "Union of Concerned Scientists Petition," dated December 15, 1977, enclosure 1, page 36, emphasis added.) I be-lieve that the single failure criterion requires two methods of achieving and maintaining safe shutdown which are independent of the influence of the fire. Therefore, I believe the Board should elicit evidence to determine whether the single failure criterion has been properly applied in the design of the dedicated shutdown system being installed in the Trojan Nuclear Plant.

The Staff and Licensees may argue that the issue of adequate fire protection is before the Commissioners and therefore canno" be taken up by this Board. It is correct that UCS'etition for Emer enc and Remedial Action is before the Commission for

reconsideration. However, specific details of the Trojan NQclea'r Plant, such as the inadequacy of the present fire protection, the design details of Trojan's dedicated shutdown system, the Staff's bases for requiring the additional system and the Staff 's bases for approving its design, are not before the Commission. These specific details are subjects that this Board can and, I believe, must, examine.

Generic Unresolved Problems In the Appeal Board's River Bend decision last fall, the sig-nificance of unresolved. generic safety issues in a construction permit proceeding was dealt with at some length. More recently, in its sua onte review of the licensing proceedings that author-

~s ized issuance of operating licenses for North Anna Nuclear Power Station, Units 1 and 2, the Appeal Board undertook "to ascertain whether the staff dealt appropriately with the 'unresolved'ssues in this operating license proceeding." The Appeal Board discussed the facts that the SER identified some of the ACRS generic issues germane to the North Anna reactors but did not do so for other generic issues contained in the Staff's Task Action Plans. The Appeal Board stated:

."And, equally important, for some of the ACRS issues the statement in Supplement 7 [of the SER] was inade-quate on its face. In particular, we found it unhelp-ful for the staff simply to note that a search for a generic solution was still underway withou" analyzing why the absence of a generic solution did not call into question the safety of current operation. Similarly, there were instances in which the main body of the SER did'ot alert us to the existence of a generic problem bearing on the particular aspect of plant design under discussion." (ALAB-491, footnote omitted.)

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9. Gulf States Utilities Com an (River Bend Units 1 and 2),

ALAB-444, 4 NRC 760 1977).

10. Vir inia Electric and Power Com any (North Anna Nuclear Power Station, Units 1 and 2 , ALAB-491 (August 25, 1978).

s The problems noted by the Appeal Board in North Anna are present in this proceeding.

The entire operating license review for the Trojan Nuclear Plant to'ok place when the Staff. was withholding from public dis-closure the existence of unresolved generic safety issues germane to the Trojan Nuclear Plant (except for those issues the ACRS, identified). It was not until I resigned and disclosed the existence of the Staff's Technical Safety Activities Report that the public became aware of the existence of a large number of I

unresolved safety issues. However, Trojan's operating license was issued before I resigned and, therefore, the opportunity for the public to raise. these issues was, for all practical purposes, foreclosed. 11 With respect to the ACRS issues germane to Trojan, examination of Supplement No. 1 to SER for the Trojan Nuclear Plant demonstrates that, at least for some items, the Staff simply noted that a search for a generic solution was underway. This is precisely the same treatment that the Appeal Board found "unhelp-ful" in North Anna. (For example, see Section 7.2.2, "Anticipated Transients Without Scram," and Section 18.2.9, "Generic Problems,"

of Supplement No. 1 to the Trojan SER.)

There are several other unresolved safety issues germane to the Trojan Nuclear Plant which the Staff neglected to mention in the SER. It also appears that the Staff does not intend to bring these issues to the attention of the Board during this proceeding. Therefore, I will give the Board some examples of issues that I have concluded bear on the question of whether operation of the Trojan Nuclear Plant will pose undue risk to the I am aware of the regulations pertaining to "show cause" II proceedings. However, I am of the opinion that these place such enormous burdens on a member of the general public as to render them useless.

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health and safety of the public.

The inadequate fire protection at the Trojan Nuclear Plant was discussed earlier. I mention it. again now because, even though the Staf f has determined that fire protection modifications "will provide substantial, additional protection which is required for the public health and safety", (10 CFR 50.109), .the Staff sup-ports operation of the plant before the modifications are completed.

This is particularly significant because it indicates a reluctance on the part of the Staff to give the highest priority to protection of the public if that would interfere with continued operation of the Trojan Nuclear Plant.

Another example of an unresolved generic safety issue germane to Trojan which the Staff may have failed to bring to the Board's attention is the forces on core internals during a loss-of-coolant accident. This problem arose in the course of attempting to resolve the generic issue identified in NUREG-0410 as "Task A-2, Asymmetric Blowdown Loads on PWR Reactor Vessel." This subject was discussed in Section 3.9 of Supplement No. 1 to the Trojan SER. In addition to the fact that this is another instance where the Staff simply noted the search for a generic resolution, new information has been brought to the Staff's attention, but perhaps not to this Board's attention.

In November 1977, a consultant notified the Staff that the impact force on fuel assembly spacer grids, caused by asymmetric loads during blowdown following a loss-of-coolant, accident, could be more sensitive to core plate motion than it was originally believed. Specifically, the consultant found that a 10$ variation

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in the frequency of core plate motion during blowdown could more than double the crushing load experienced by the fuel assembly spacer grids. This raises the question of whether there would be permanent deformation of the spacer grids during a loss-of-coolant accident and, therefore, whether the reactor core would have a eoolable geometry, as required by 10,CFR 50.46(b)(4).

The attached letter (Enclosure 8) has recently been introduced by the Staff in other proceedings, with the notation that the in-formation is preliminary. I recommend that this Board inquire further into the matter of asymmetric loads to determine whether there is a basis for concluding that Trojan meets the requirement of 10 CFR 50.46(b)(4), "Coolable Geometry." The Staff should not be permitted to call year-old information "preliminary," and allow operation of the Trojan Nuclear Plant.

The last example of a generic issue germane to Trojan that I will bring to the Board's attention is the adverse interaction between non-safety control'ystems and safety-related protection systems. Interaction between control and protection systems can result in an event which creates a situation requiring protective action and concurrently disables all the protection systems designed to perform the required protective action. Based on my review of several plants, including Trojan, and events that have occurred in operating plants, I conclude that the Westinghouse design is unsafe.

I am supported in this conclusion by at least some members of the Staff. For example, following an event in the Westinghouse-designed Zion plant, Dr. Stephen Hanauer, Technical Advisor to the

Executive Director for Operations, expressed his view that. the Westinghouse design was "unsafe." In the Zion event, workmen had disabled 31 instruments monitoring the reactor; The result was that water was being drained from the reactor cooling system and all systems capable of detecting the loss of water were in-operative. Dr. Hanauer stated that: "The acceptability of all systems, Westinghouse and non-Westinghouse, old and new, needs to be reviewed in the light of the Zion event and any unacceptable interactions removed." (Enclosure 9)

I agree with Dr. Hanauer and recommend that the Board require that such a review be performed before the Trojan Nuclear Plant is permitted to resume operation. The results of that review should be tested by cross-examination of witnesses under oath.

As I noted above, I have not discussed every technical issue which I believe should be examined before the Trojan Nuclear Plant is permitted to resume operation. I recommend that, the Board req'uire .the Staff to identify all the unresolved safety issues germane to this plant and to explain why the plant can be permitted to operate in the face of each unresolved issue. To have a complete list of all unresolved generic safety issues that may be germane to Trojan, the Board should not rely solely on NUREG-0410 or some other list of the Staff's Task Action Plans. The Board should also examine the minutes of meetings of the Staff's "Technical Activities Steering Committee." These minutes disclose a host of other safety issues which the Staff has not included in its published li'stsof unresolved safety issues.

12. After an anonymous, member of the Staff provided UCS with a copy of Dr. Hanauer's letter and UCS disclosed it to the public, Dr. Hanauer changed his opinion of the urgency of this problem.

METHOD OF ELICITING EVIDENCE In addition to identifying subjects which I believe deserve additional inquiry by the Board, I want to recommend a technique of eliciting evidence that could result in compiling a more complete and accurate record.

After reviewing some of the recent direct, testimony in this proceeding, I detect a reluctance by the witnesses to address directly the question of equipment qualification. Witnesses for the Licensees, Richard C. Anderson and William H. White, limit their testimony to the new response spectra and conclude "that there would be no effect on previously qualified equipment, piping, and electrical equipment." (Tr. 2337, emphasis added.) These witnesses go on to state that, "we find that the equipment still remains qualified, based on the ori inal uglification of the the Board, Witness White continued this theme by stating that "essentially everything that was qualified prior to that time was still qualified." (Tr. 2351.) 'he key point, which I addressed at length earlier, Xs whether the original equipment qualifica-tion was acceptable. It is this point which the Licensees'itnesses apparently would rather not discuss.

The Staff's testimony on this point is even more circumspect.

In his testimony dated October 13, 1978, James E. Knight specifically disclaims any evaluation of the. seismic qualification of safety-related electrical equipment. He testified as follows:

"The effect of the postulated earthquake on 'such equipment as batteries, switchgear, control panels, etc., located within the control, building was not a part of my evaluation, but has been addressed as a part of Mr. Herring's testimony."

(Testimony of James E. Knight, page 2, October 13, 1978.)

I find this curious because, according to the Professional Quali-fications accompanying Mr. K'night's testimony, he holds a degree in electrical engineering and from 1975 to the present has been employed by the Nuclear Regulatory Commission as a Reactor Engineer (Instrumentation). Although now assigned to the Plant System Branch in the Division of Operating Reactors, I believe Mr. Knight was previously assigned to the Electrical, Instrumenta-tion and Control Systems Branch. It was this Branch that had responsibility for reviewing the seismic qualification of safety-related electrical equipment.

In contrast, Kenneth S. Herring holds a degree in civil engineering and, in his statement of Professional Qualifications, he describes his "duties and responsibilities" as involving "the review, analysis, and evaluation of structural and mechnical aspects related to safety issues for reactor facilities...."

(Professional Qualifications of Kenneth S. Herring, undated, attached to "Testimony of Kenneth S. Herring, Office of Nuclear Reactor Regulation, on Structural Adequacy of the Trojan Contxol Building for Interim Operation," undated, emphasis added.) I have also examined Mr. Herring's most recent testimony, "Testimony of Kenneth S. Herring, Office of Nuclear Reactor Regulation, on Floor Response Spectra and Qualification of Safety-Related Equip-ment and Systems in the As-Built Control Building Complex,"

November 25, l978.

~ 0 The subject of equipment qualification is mentioned three times in Mr. Herring ' November 25, 1978 testimony: in the title; in the first paragraph, which describes the purpose of the testimony; 'and in the first "sentence" on page 2 of the testimony.

In each instance, the same amount of information is conveyed. For the Board's convenience, the "sentence" on page 2 is. repeated in its entirety:

"Further, given that the appropriate modifications are performed to assure conformance of the equipment, systems, piping and components with the .spectra as defined in the October 27 and November 2, 1978 sub-mittals and further widened as indicated in the November 22,, 1978 submittal, these investigations are adequate to make the determination that there is reasonable assurance that the safety-related equip-ment, systems, piping and components in the Control/

Auxiliary/Fuel Building complex will withstand an earthquake up to and including the 0.25g SSE."

I am aware that engineers, myself included, sometimes have I

difficulty expressing their thoughts in clear language, but Mr.

Herring's "sentence" is total gibberish. If it, is comprehensible at all, this "sentence" seems to say nothing more than that, if something is done correctly in the future, someone might be able to determine whether the safety-related equipment is seismically qualified. Mr. Herring could have simply stated, as Mr. Knight did, that he has not evaluated the seismic qualification of the safety-related equipment used in the Trojan Nuclear Plant.

Why does the Board face a situation where not a single witness is willing to answer a direct question--"Is the, safety-related

'equipment seismically qualified?" I believe this situation arises because the equipment never was properly qualified and neither the professional employees of the Staff nor PGE are free to so testify. The bases for these conclusions are my own review

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of the Trojan operating license application and my personal knowledge of the pressures faced by professionals employed by the NRC and PGE.

I can give innumerable examples of the actions by management officials that inhibit members of the Staff from expressing view-points that could delay or preclude operation of a nuclear plant.

It should suffice here simply to state that my resignation from the Staff was caused by such actions. However, since I have never been employed by PGE, I should explain the reasons why I reach the same conclusion regarding PGE's professional employees. Attached are copies of correspon'dence between me and an individual employed by PGE in a professional capacity (Enclosures 10, ll and 12). I have deleted all information that. could aid in identi fying the PGE employee and my home address and telephone number.

Enclosure 10 is a memo I received from the PGE employee.

Since the memo was handwritten, I have retyped the text in its entirety. The enclosure to that undated PGE memo (the envelope was postmarked in Portland, Oregon on November 23, 1976) is attached to Enclosure 12, together with my letter to NRC Commis-sioner Gilinsky and the reply from the General Counsel. There are at least two significant aspects of these documents. First, it can be observed that Westinghouse believed "that site boundary doses in excess of exposure guidelines set forth in 10 CFR 100 could result from a fuel handling accident inside containment..."

(Second attachment to Enclosure 12.) Second, the PGE employee thought that this matter should be brought to NRC's attention,

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but apparently was afraid to do so except. through a third party.

Therefore, I conclude that there is evidence that neither Westing-house employees nor PGE employees are free to bring potential safety hazards to the attention of NRC.

If the Board, adopts my recommendations and requires further testimony on the specific technical subjects I have identified, the witnesses proferred by the Staff and Licensees should be carefully chosen and instructed. With respect to the Staff's witnesses in particular, the Board should require testimony by the individual professional who performed the review, rather than by a supervisor of the reviewer. Furthermore, such testimony should be given orally or the Staff should be instructed that written testimony should be prepared without review or concurrence by supervisory personnel until after the testimony is filed. I believe these precautions are absolutely essential in order to assure candid testimony by professional employees of the Staff.

Furthermore, all prospective witnesses should be provided with guidance from the Board concerning the meaning of their oath or affirmation that their testimony represents the truth, the whole truth, and nothing but the truth. In addition, the Board should explain to each prospective witness the protection, if any, available to prevent reprisal by his employer should the witness offer testimony unfavorable to a decision authorizing operation of the Trojan Nuclear Plant.

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CONCLUSION I am aware that, adoption of my recommendations may delay the prompt resumption of operation of the Trojan Nuclear Plant envi-sioned by the Staff and Licensees. However, since modifications to the control building are needed and since the dedicated shut-down system determined to be needed by the Staff will not be installed until June 1979, there appears to be some time available to conduct the additional inquiry which I recommend.'ut the principal test of whether the additional inquiry should be under-taken is not the time available. Rather, the test is whether, in the absence of such inquiry, there is adequate assurance that the health and safety of the public will be protected. I have concluded that such assurance does not now exist.

Reactor t:ng luce r~ a' ~me u tat ion) ~S-l~ ENCLOSURE 1 Electrical, Instrumentation and Control Systems'ranch Direc'borate of Licensing Pl;.:CTIOiPL S .'l7;=.ice"NT:

Se".v s as a hi hly quali'ied specialist in the field of reactor instrum ntation and control in performing technical reviews, analyses and evalua.tions o sys em

-nd co"..yon nt designs necessary to the sa=e o~erztion unde Qo uzi abno ~ Q.

-nd emergency condit'ons o power, testing, production, and resea"ch reac"ors, includ-n" DOD and AEC-owned reactors as well as licensed and authorized xaci1i-,

ties.

?"- "' Db'ES:

Participates as a sen'or member of the Directorate of Licensing, Electrical, Instrumentation and Control Systems Branch, whose function is primarily one oX "rr oning tecnnical res'ews, analyses, and evaluations of designs of sys-tems and components necessary to the safe operation of reacto- fac-lit'es under normal, abnor el, and emergency cond tions for the purpose of (a) dete~ning the adeauacy of the bases for such designs, (b) of determin ng the adequacy of such designs to r;eet these bases and to withstand the limits o-.".

environmental e"fects witnout loss of m='nimu-. required functional capabi ities, (c) of determ-ning the acceptaoility of'rocedures for fabrication, inspection, testing, and post-licensing surveillance o such desig..s, and (d) o-. develop=ng guideline procecu" s, methods and models fo" the systematic evaluation of such designs by tne Divis'on.

Reviews Safety Analysis Reports as to'the adequacy of the presented data pertaining to -nstrumentation, controls, and electric power and to the. sound-ness of conclusions made on the basis of the presented information and prepares reports of such reviews.

Develops standard procedures, methods, and models for evaluations to deter;ice whether or not the design of reactor protection system, controls =or eng~weered sarety features, sa ety aspects of re ulat'ng syste-s, and, emergency'powe= sys-tems is treated in an acceptable manner.

Evaluates industry and AZC-sponsored research and developmen" programs directed towards establishment of additional basic information on re"ctor instrumenta-tion and control, and to the use o" such information for safety evaluation pur-poses, and correlates and interprets the results oi such pro rams for the gener&

use o the regul- tory staff.

Prepares technical'tudies and reports bearing on un.que and unusual. develop-..en "s in ta-.e field o". = actor plant inst"umentation, control, and electric power for presentation of tne Advisory Com.attee on Reactor Safeguards.

Reco ends throu h the Branch Chief, safety research programs to be sponsored.

by the AEC.

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p, actor Engineer (Ins t t'umentatioL1), CS-I(>

Blcctrical, Instru"Lentation ana Cont "o1 iys teLLLs Branch Directorate of LicensiLag Page 2 Confc" pe od c . ay ti'l CC .n ca ropres"ntativas I o xncustrial. 0 g -TLizat-Gns

- nd other A C d-':isio"..s to discuss nuc3car safety pronla-...s 'nvo ving areas o=

concern related to reactor plant inst umentation, control and elect ca] pol'a -.

l Participates, frorl time to tir;;a as tachn ic" representative or the D'acrorata of L-'ensing on va"'us a:"C co;..",atteas, subco-..~.xtteas>> pana3. ana tasx -orca assignments as wall as technical and pro feSSiOna SOC- ety CO-...." taas SUC aS tancar" s Ins ti cut a

tha American 'National S t.aa AmariCa.l (auc aai. SOC aty>> taae Ins'tituta nf "-'c"rical and .Electronics "'ilgwu(Laa S>> aaaa oa.a(ai~( \

BASIC SKILLS:

General ~ow ed e of tha principles, theories, and pract cas in tha iield 'of nuclear engineer'nig with spacif'c snot.edge of reactor plan" -nstru control, and electric pow r sys tams is required. Competency must be su ent to independently analyze and evaluate reactor concepts and features pro-f'i-

...antation, posed by organizations special zing in tna nuclear f 'eld>> particularly wit.".

respect to the reactor protection systa-.s, 'nstrumantation and contro3. =or engineered safety features, reactor regu'ating systemas, reactor plant dynamics, and electric power systems.

Tha basic s~i~l requirements ara considerably in excess o f those sec .red.

taarv'algal fvLl84 education at universe,ty lava 'B. S. jiegrea) znc. are co(..paraola to those achieved irom graduate lev 3. training or from specialized experience in instrumentation and control in applications to reactor technology.

Knowledge of licensed and authorized as well as DOD and ABC-owned reactor installations and operations is required.

CONTACTS:

Contacts top tecnnical and supervisory personnel of tha A"=C, other Govern-..ant agar cias, A""C contractors, industrial or"anizat'ons, research institut- ons, universities, and professional societies to discuss technic& mtte=s relating to reactor plant ialst u=,antation, control and electrical power syste-.s.

RESPOND'SXBILZTY rO.C DECISIONS:

Supervision Received:

Chief, ilactric=l, Instrumentation -nd Contro'ystems Era(lci1 S up arv s 0 il

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is ge-.,eral on t c'".nical rattars wi "n u' authority to aCt iil VZta.ars Witaaa n the fra-aworw of the broad unctional assi nment.

istrat ye guides ara Div'-'ion and overall AZC pol'y and pracadan".

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Reactor Engineer, (Enstrum~ntnti on), GS-l~

9'lect@ical, Instr~~entation and Control Systens Branch Directorate of Licensing Pa e 3 End ~endent Action:

Respons=bla for @awing reco:.... ndations =or aetio. to ba tacan by t..a Chia=

of t'na =lect"ical, lnstru-..antation and Control Syste=s 3ranch<

Develops standard procedures, ..e"hods, and ...odels for those aspects o saf"ty evalu-tions involving physio-cha-, 'cal considerations.

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'..'MZ~YG COYDITZO'AS:

Zoril office conditions w..ila at ofiicial station. inclosure to n-Xd radiaaon from reactors cay ba encountered occasio"ally dur-'ng field trips.

EFr ORT:

Zoril effort involved in any administra'va position. ~

increased physical, e fort cay ba required while on field trips.

y0 ~ ~ ENCLOSURE 2

<<(<<CS <o,y

+p UNITED STATES ATOM IC ENERGY COMMISSION

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<<+ March 20, 1974 ilff S

NOTE TO Robert Pollard's Personnel Pile I have received the attached note from Mx. Herschel Specter commending the performance of Robert Pollard of the El.ectrical, Instrumentation and Control Systems Branch. Mr; Specter was the Licensing Pro)ect Manager for the Indian Point Units, and is, therefore, well qualified to speak to Bob Pol.lard's per-formance on those reviews.

I' have had occasion myself to observe Pollard's work in the review of several plants and in the review of reference designs undex the Commission's standardization policy. I find Poll.ard exceptionally expert in his technical area and articulate and effective in his work as a technical reviewer. I am pleased to add my comments to those of Mr. Specter and to forward them to Bob Pollard's personnel e.

oseph M. Hendri.e, Deputy Director .

for Technical Review.

Directorate of Li.censing Attachmen t (As Stated) cc: F. Schroeder V. Stello

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ENCLOSURE 3 P:ojecc ~!ana.";er, GS-14 Directorate o Licensing l ~ 'a (+ hat I 0< ~

I Exp ". en"ed in che ergin"ering and physics aspects or" ruclea= xeacto"s, the inC ~ant planS ond.COOrdinaiaS Cne teChniCal rericWS, Zna ySaS, and evaluaiions of applications for. 1'censas and authorizations xor tha construction ard operation oi reactors and the rav. ews of carta'n aspects. of the== design and operat'n.

Rr.GUI 4R DvxZ" S Plans and coordinates the pre-xeview of applications to determine i""

they are surf ciencly complete to accept as an application.

Plans ard coorcinoies the review o" Safety Analysis Rapox'cs as co t..e adequacy of ti.e technical and ang nearing desi"n data and inrorma"=on contained therein, the soundness of the b s's for the conclus'ons o=

the proposed designs and operacing procedures. Coord-nates the pre-paration of rha safety evaluation in conjunctiora w-'th such re'ia<<s.

Serves as proj ect ;..anagar for group evaluation of powa reactcr l cc-.nse applicants for which he has bean "ssigned responsioi" ity.

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vL vL )claLcrdl Lva 0' L UVULA Lug~ nl danti&j o" resolve gen ral que'scions o" "o 'ot conce

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reactors to ...ir' design and operating characterist.cs which have a ba-ring on sa"e"y.

Coordinates the preparation of sa ety evaluation reports alar.'ng to 1icease applications o" powex'eactor plants, as well as r.:il ""'; and AEC reactor plants for presentation to the Advisory Cc ttae on Reacio Sofaguords. Acta..cs such ..ea rigs ond subcc .i cceP.'..eec r. s a raprasantac've o= tria Diractox'ace o= icensin s evaluation s""==.

Qy part cipaie at public hearings on reactor 1'censing proceedings to ass-st ocher A:.C representatives or testify as an 9ZC sta f w'tness to present technical testimony.

Serves as a, ~a=~ r of reactor inspac" ion teams as may oe nec ss-xy to discharge reactor sa ety evaluation and judg. ant.

Plars ond coordinates the review oi nuc'ar safety spects o" propo als to build any r'"-C-owned reacrors a:tempt from licens'n Assists in t..o prepracion of technical speciz'catic..s =cr o"arac.n reactors; reviews operat'ng a@peri nce reports dur'ng initi"1 phase" ox opa"scion; and avaluaces rcquesis for license amancmancs arid technical spec ication c',".angas during initial pha'sa" of oper"t'on, util "'"..". che expertise o= persons outs.'de of h' '=. diata d'vis'-..

whare necessary.

Project "nager, CS-14 D'ectorate o" Licensing BASTC SN LL~

Knowledge c the principles, theories and practices in the field o nuclea" reactor technology with speci=ic 'ccwledge c=: reactor and nuclea" eng ".eeri.".g. Co-.petence ."..us= - su==.c.'en" to adequately evaluate various proposed reacto" co.".c p"s and rrodifications prir.:arily as related to reactor construction an= operation.

Knowledge of operations at A"C-owned "eactor installations.

Experience in tne Iield of reactor co"e design and operation to supplen nt basic eqgineering training.

Basic s~iLL requireiants .are in excess o those secu-ed throug:. for-...al education at tne university level(B.S. Degree) and a=e compa'rable to those obtained from graduate level t"a.ning o" speci-.lized experience in reactor technology and related sub=- cts, CoiiTACTS Contacts are with top technical perso"nel in A"-C, A=-C contractor, industry, and other governr:,ent agencies to discuss tachnica natter relating to the ha~rds inherent to 'des'n, opcrat on., an/,site:

location of proposed new reactors or significant aoi= fication of existing reactors.

P~SPO~!SV~ TiiTY FOR D:"CISIOYS Su ervis'on Received Assigned Branch Chief, Directorate of icensing.

Supervision is ger.eral on technical ~tters, but spacific on policy and oper ting procedures.

Adrinis'ative gu'des are, AZC manual, Co~ission RuLes and Regulations, and AcC policy. Operational gu des -" in the Eow of ...emoranda, wri=ten guides, precedent, and verbal d'rectives. Tncuchent contri .tes to the developr.ent of original stand---'s, guides, aud codes in his fieid oi endeavor,

Project t~QMe GQ ]Zf ~ 0 Directorate of L c"nsing Ind nosed ol ~

Xncu=> nt is. -"sponsible for preparation o'".." adherence to rex.eu sch dules and fo" r.:a! in~ reco=-endations on coaveational engineeri"g

u. tters for aetio to be t&en by tbe Branch Chic= ia. -edward. to th acceptability o= th hazards involv d in spec fic reactors.

Incident's jud=e ea.t, ia aany cas s, is subject ta only a gene-al, efQ SUP~WViSIGN Yon~

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SORUSG CO~iDZ'ZX0'.iS.

UoaraT. of:=~ce: ~it,ious .1>y be- e moss to.aQd. Xad Lclt o R 4gl~ l~~

on field trips=

EP:"QW'or~i

ENCLOSURE 4

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PERFORifglCE APPPCISAL AiD RECORD OF INTERVIE! FOR NON-SUPERVISORY PROFESSIONAL TECHNICAL EPLOYEE PROFILE hA.'K: Robert D. Pollard GRADE/STEP: 14/2 POSYTTOÃ: . Project >Ianager g z/ TI~K IN GRADE: 19 months cc44 APPRAISER & D;Y"'E: D.. ,Jg assallo TI~E IN STEP: 7 months VIEWER DATE: E. D. DeYPung V "JS-Z> TI~K IN PREVIOUS GREGE: 12 months 5

g PREVIOUS APPRAISER & DATE: D. B. Vassallo AEC/NRC SERVICE YEARS: 6-1/4 11/11/74 PROFESSIONAL EXPERIENCE YEARS): 6 Holi LONG SUPERVISED BY APPRAISER: 13 months DATE OF BIRTH: 2/13/40 EDUCATION (DEGREE & YEAR):

B.S. (Elec. Eng.) 1969

'ISCUSSION TOPICS BACKGROUND & LWERIENCE SIMERY Bob received a B.S. in Electrical Engine'ering from Syracuse University in 1969. After joining the AEC in 1969, he studied electrical and nuclear engineering at the Graduate School of the University of New Hexico (1970-1971) in conjunction with the AEC Intern Program.

Bob served for six years with the U. S. Navy as an electronic technician.

He served as an instructor, reactor operator, and was in charge of the reactor control division aboard a nuclear-powered submarine.

After joining the AEC in July 1969, Bob partic'pated primarily in technical review groups in the review of instrumentation, control, and electrical systems of nuclear power plants. For a brief period, he was a member of the Standards group and participated in developing standards and safety guides. He also served as a member of IEEE Committees. Bob transferred to RL as a project manager in September 1974.

KNOtJLEDGE OF JOB - Although Bob has excellent expertise in the instrumentation, control, and electrical systems of nuclear power plants, he has also developed very good overall knowledge of nuclear power plant design. Since transferring to RL, he has shown the capability to rapidly expand his knowledge and under-standing of the diverse technical review areas with which a project manager must be familiar. Although he may require a little more exposure in certain review 'areas (e.g., auxiliary systems and site related matters), Bob is

technically very perceptive. He has enough confidence to challenge reviewers on questionable tecnnical matters and to pursue resolution of those in controversy..

In a very short time, Bob has developed an excellent understanding of the technical, management, and administrative aspects of project management.

He manages to keep himself informed of current developments in technical, policy, legal, and general licensing matters.

MANAGE!E'AT CAPABILITY . Because~f. his. past experience in TR, Bob had a good understanding of the LPtf's role. He has made a very rapid transition in assuming a project management pnilosophy. In the year that he has been in RL, he has demonstrated an excellent capability to manage radiological safety reviews. Bob is an exceptionally thorough project manager who performs his tasks with a very critical view and in a very organized manner.

He is very knowledgeable of, and quick to grasp and implement administrative procedures He is very effective in maintaining full cognizance o all aspects of his projects. 1ie works very effectively with TR, OELD, and applicant representatives and gets along very well with the branch secretaries and licensing assistant.

I Bob has shown excellent capability to effectively organize <<nd manage several concurrent projects. His principal assignment has been the OL review of the ~fcGuire plant. However, he was also assigned as the LP~I for the completion of the Comanche Peak and Catawba CP reviews. The latter required a considerable amount of LPH interaction with OELD because of the applicant's request for. an exemption from meeting the ECCS criteria. Bob showed great adeptness at understanding and handling the unique technical legal aspects for bringing the Catawba project to an end; i.e., issuance of a CP. Because of the recent loss of an LP"f from LHR l-l, Bob was also assigned the task of completing the OL review of Indian Point 3, another project with a long history of complexities. In handling all of these projects, Bob has shown a great deal of resourcefulness in moving these projects forward concurrently without diminishing his efforts in any one of them..

PROFESSIONALISH Bob is an extremely conscientious and dependable project manager. He conducts himself with a degree of maturity and professionalism well beyond his age. In his associations with applicant representatives, he is very fair, but firm, and can take a strong stance when the occasion warrants it.. Bob does extremely well at planning and scheduling his workload. He is consistently able to complete assignments on schedule without needing reminders.

Bob does not require much supervision. On the contrary, he seems to have a unique instinct of knowing the type of licensing action that a situation requires and then begins to take the appropri te action without waiting for direction from the branch chief. In this regard, Bob has an out-standing knowledge of the Regulations and works very effectively with lawyers (e.g., has prepared some quite involved technical - legal documents in conjunction wi.tn the Catawba and Indian Point 3 projects). He is very persistent in trying ro get stalled actions moving. Bob does an excellent job of keeping his branch chief apprised of major review matters.

JUDGMENT Bob is a careful thiater" and-uses good logic in making judgments.

He has a very good understanding of tne licensing program and uses good judgment consistent with regulatory objectives.

CO~LQIUVi KCATIONS Oral Bob has very good oral communication skills. He speaks clearly, with thought, and is very easily understood. He handles meetings extremely well. When he was a member of TR, he had considerable experience and was very effective in pr<<sentations before the ACRS and was also exposed to public hearings.

Written Bob writes extremely well. The documents he prepares are concise and clear. As mentioned above, he has a decided instinct for knowing the type of action required and can translate this in writing without any apparent difficulty. His written work requires very little editing.

PERSONAL CH~CTERISTICS Basically, Bob is a very serious minded but personable, employee. He does not make rash decisions, but rather uses a more deliberative approach. Bob manages to maintain a rather even composure no matter how difficult a situation may get.

Bob is an extremely conscientious, responsible, and dependable employee.

Occasionally he appears to become somewhat perplexed in rationalizing the implementation of licensing policy. In my opinion, this is because Bob has an exceptional understanding of the Commission's rules and regulations and takes his role of regulator very seriously. However, this has not affected his performance as a,.proj ect manager.

AREAS NEEDING DPROVEiIENT Since transferring =rom TR, Bob is becoming exposed to a number of review areas with which he did not previously have a great deal of familiarity. These are principally in the areas of site safety, effluent treatment, and some portions of auxiliary systems. He has made great strides in understanding wha" the major review objectives are for these areas. With the continued exposure he is now obtaining in managing his projects, I do not foresee any problem in Bob becom'ng completely conversant in all review subjects.

PRO':lOTIOÃ POTE';!TIAL Bob has sho w excellent project management capability.

He is well organized, is able to keep his projects under control, and to meet schedule milestones. On the basis of his previous experience in TR and with further experience in project management, Bob has an excellent potential for attaining higher levels.

SK'DLkRY - Altnough Bob has been in RL for about one year, he has demonstrated excellent skills in managing sazety reviews without requiring a great deal of supervision. Through his versatility, he has perzormed extremely,well in handling diverse assignments in a hignly prozessional manner; e.g.,

taking on tne management of comple:c cases such as Catawba and Indian Point 3 in the final stages'f licensing effort.

REVIE':F'R'S CO!CfE".ITS

/i -J8- >u E'EMPLOYEE' CO~iMi';ITS Acknowled ement I have read the above performance appraisal.

Comments b Employee

.. 0 ENCLOSURE 5 QPn j ~ l374 Docket No. 50-3W R.- C. DeYoung, Assistant Director for Light Rater Reactors Group 1, L POKrLAHD GENERAL ELECTRIC CC'fpAHY, TROJAN NUCLEAR PLANT; SAEfENTAL QUALIFICATION Me have not completed our review of the qualification test programs applicable to the Trojan instrumentation systems. Major portions of the review are being conducted on a generic basis with the nuclear steam system supplier. Me will report the results of the review and their applicability to Trojan in a supplement to this safety evaluation.'.0 ELECTRIC POWER 8.1 GENE'VQ The Commission's General Design Criteria 17 and 18, Regulatory guides 1.6, 1.9 and 1.41, and .IEEE Std 308-1971 were utilized as the primary bases for evaluating the adequacy of the electric power systems of the Trojan Nuclear Plant.

8.2 OFFSITE POn'ER Four 230 kV circuits are provided to carry the station electrical output to and supply offsite power from the transmission network.

There are two sets of double-circuit transmission towers 'located on separate rights-of-way with two 230 kV circuits mounted on each set of towers. One circuit of each set is connected to one switching station bus on the site and the other circuit is connected to the second sQitching station bus. With this arrangement there are only two combinations (of the six possible combinations) of two circuits that meet the requirement of GDC-17 for two independent circuits to supply power from the transmission network. These two combinations of two circuits are: 1) Trojan-St. marys circuit and Trojan-Allston No. 2 circuit and 2) Trojan-Harborton circuit and Trojan-Allston No. l.circuit. The technical specifications will require that, as a minimum, one of, the two combinations of two circuits be availablc as a limiting condition for operation of tne plant, The other four combir>>i-ns of two circuits do not meet the requirement of GDC-17 because the two circuits are either mounted on the same transmission towers or are connected to the same switching station bus.

operation until temperature equilibrium is attained. If during this performance testing a failure rate in excess of one failure per one hundred tests is experienced, further testing as well as an evaluation of the system design adequacy will be required. This qualificat'ion testing program has

'been established on other recent license applications using diesel generators of a type or in a configuration not previously qualified for standby power service at a nuclear generating station. The applicant has not yet submitted the results of the qualification test program for the Trojan diesel gener-ators. He will report the results of our evaluation of the adequacy of the diesel generator sets in a supplement to this safety evaluation after we have received and reviewed the results of the test program.

8.3.2 SEISMIC QUALIFICATION OF ESF Sh'ITCHGEAR

.During seismic testing of the protective relays associated with the ESF breakers, some relays were found to misoperate intermitt-ently during application of the accelerating forces. This meant that some of the ESF circuits could be inoperative during a seism"'c occurrence since the relays might trip as if performing a protective function. After the seismic occurrence, it. would have. been necessary to manually reset those circuits that had tripped. The applicant had concluded that since the relays were not damaged, this was acceptable.

Ne informed the applicant bf the staff's position that all safety-related electric equipment is required to be designed to withstand the effects of the Safe"Shutdown Earthquake without either malfunction or loss of capability to perform the intended function without operator action. The applicant subsequently stated that some types of relays will be replaced by relays of a different manufacture for which test data indicate the ability to withstand forces greater than required without causing the ESF circuit breakers to open. lee also understand that the applicant is considering automatically blocking the tripping function of other relays that could misoperate and thereby.

cause tripping of th diesel generator gircuit breakers.

Additional information on the seismic qualification program can be found in Section 3.10 of this safety evaluation. The results of our evaluation of the adequacy of the seismic testing of the ESF switchgear wi11 be rep'orted in a supplement to this sa fe ty evaluation ."

~~ ~0 ENCLOSURE 2

.Referenced To ical Re orts Not Reviewed or Reviewed'nd Found Unacce table WCAP-7821, Seismic Testin of. Electrical and Control Equi ment (fli h Seismic Plants) December 1971 Status - Not reviewed - TAR scheduled for completion 6/15/74.

WCAP-7744; Environmental Testing of En ineered Safet Features Related Equi ment (NSSS - Standard Sco e), August 1971.

Status - Reviewed and found unacceptable because "it can not be determined that the equipment and systems tested under the subprograms can complete their safety functions for the time required following a design bases accident." (Ref: Letter to R. Salvatori from D. B. Vassallo, dated March 12, 1974.)

WCAP-7672, Solid State Lo ic Protection System Descri tion, June 1971 Status - Found acceptable provided "the equipment is adequately qualified seismically and environmentally and the appropriate documentation has been provided." (

Reference:

Letter to R.

Salvatori from D. B. Vassallo, dated March 6, 1974.) PGE has not.supplied information specifically applicable to the Trojan solid state logic protection system. Perhaps WCAP 7821 and WCAP-7744 discussed above will be revised to include the solid state logic protection system.

WCAP-7705, En ineered Safeguards Final Device or Activator T~esc1e Beech 1973 Status - Reviewed and found unacceptable (

Reference:

Letter to R. Salvatori from D. B. Vassallo, dated September 10, 1973.)

WCAP-7705, Revision,l,'esting of Engineered Safety Features Actuation S stems, February 1974, was recently submitte

(

Reference:

Letter to D. V. Vassallo from R. Salvatori dated March 26, 1974.) ~

WCAP-7819 Revision 1, Test Re ort Nuclear Instrumentation

,S stem Isolation Am lifier, January 1972 Status - Not reviewed. TAR scheduled for completion 8/5/74

(

Reference:

Status Report - Review of Westinghouse Topical Reports, D. B. Vassallo, February 20, 1974.)

OO UNlTKD STATES

~0 ENCLOSURE 6 ATOM)C ENERGY COM'ldlSSION WASHINGTON. D.C. 20S4$

July 21, 1971 Mr. J. Forster Atomic Power Equipment Department General Electric Company - M/C037 175 Curtner Aven e San Jose, California 95125

Subject:

IEEE-323

Dear Jay:

My comments on this document were solicited by Mr. Sherr in his letter of June 24, 1971. He should not have done it.

I cannot find a single redeeming feature'in this worthless document.

Far from being what its title suggests, it contains only the most gener-al kind of stuff on how to qualify something - anything. The body of the document is not even specific enough to be related to electrical equipment.. Furthermore, the various clauses are so general that essentially imoossible to determine compliance. For these reasons the it'.

referenced document in its present form is, as I said above, without value..

Sincerely yours St phen H. Hanauer cc: Louis Costrell Sava I. Sherr

~0 FNCLOSURE 7 HE,"-.ORAHDUht FOR: Chaf roan Hendrf e Ccemfssf oner Gf1fnsky Ceurrfssf oner Kennedy Comwf ssf oner Bradford

~wg X. a,~

THRU: xecutfve Director for Operations FROH: Edson G. Case, Acting Dfr ctor, MR

SUBJECT:

UNIOlt OF CO>fCEP3(ED SCIBlTISTS'ETITION FOR RECONSIDERATION DATED ICY 2, 197S provides answers to the five asterisked items fdentfffed fn the Secretary's mnorandum of June 21, 1478, for Wfch staff response was reouested by July 5, 1978. A response to the other ftems xill be provided by August 25, 1978.

/~j'dson G. Case, Acting Director Cfffce of Huclear Reactor Regulatfon

Attachment:

Ps stated cc: Secretary

~JCS PDR

OO ~

ENCLOSURE 1 RESPONSES TO THE SECRETARY'S JUNE

'ARTIAL 21, 1978 ~"',.R ..--" '-'.

' i MEMORANOUH CONCERNING THE UNION OF CONCERNEO PETITION FOR RECONSIDERATION SCIENTISTS'+ "" '/

• l. UCS quoted the staff as stating that in at least some presently operating plants, a fire could have the same effect as the Browns Ferry fire. Identify those plants, if any, and the basis for permitting such plants to continue operation. (pages 1,2)

~Res ense The Brown's Ferry fire was an unmitigated fire which burned in excess of seven hours, disabling redundant safety systems.

The staff has recognized since the Brown's Ferry fire that there are certain locations in some operating plants in which an unmiti-1/

gated fire could affect redundant systems. Fires causing cables to burn in such locations could disable the normal operation of r

such safety systems, but would not necess ari 1y prevent operator action from performing the safety function.

Every nuclear power plant has some flammable materials that are either fixed or transient; therefore, at least a limited potential for fires in nuclear power plants exists; however, such fires should not adversely affect safe plant shutdown. Since the total elimination of unforeseen fires is not an achievable goal, the staff 1/ See page 34, staff report dated December 15, 1977.

%2a objective has been to reduce the severity and lower the probability of fires.

Starting iamediately after the Browns Ferry fire, even in the absence of a complete analysis of the potential effects of fires at each individual plant, the staff began its program of fire protection upgrading at all operating nuclear power plants. This program has significantly reduced the potential for severe damaging fires through actions taken to control combustibles and ignition sources, to control access to areas of concern, to install fire retardant coatings and blankets, to improve detection and the ability to extinguish fires fire detection and trained fire brigades that are prepared for

'e.g.,

the prompt use of water to extinguish cable fires). This has been an important hlement of the basis for continued operation of these 2/ and 9.5-1 plants pending ful.l upgrading to conform to BTP ASB E

Appendix A to the BTP.

Further, since the Browns Ferry fire, each operating plant has been subjected to a specific fire hazards analysis by the licensee.

The adequacy of fire protection for any particular plant safety system or area is determined by such an analysis of the effects of the postulated fire on the plant 's ability to safely shutdown and the ability to minimize radioactive releases to the environment in the event of a fire.

See pages 15-23, 33-38, and 71-72, staff repor. dated December 15, 1978.

M3w

+he results to date of the ongoing staff evaluations of fire protection programs- show that each plant contains a few fire areas where a postu-lated unmitigated cable fire may affect both divisions of redundant safety systems. Such potential fires are then carefully analyzed to I

determine whether the existing and proposed fire protection features would assure compliance with BTP ASB 9.5-1 and its Appendix A. Particular attention is given.to assuming for each critical area, that either the fire protection systems would mitigate the fire so that the fire would disable no more than a single division, or that an alternate method of shutdown is available independent of the systems affected by the fire in a particular area. Where it is not clear that a potential fire would be limited to one division, or that at least one method of achieving and maintaining safe shutdown is independent of the postulated fire, additional fire protec-tion features or additional shutdown capabilities, or both, have been required by the staff.

The operating plants for which our fire protection evaluations are suffi-ciently complete to indicate the need for an alternative method of shutdown are identified in Table I.

See pages 23-26, staff report dated December 15, 1977.

As shown in the table:

ll plants have been shown to have adequate shutdown capability and do ngt require installation of an alternate or dedicated system.

5 plants have installed an alternate or dedicated system as a result of our evaluation.

12 plants require an alternate or dedicated shutdown system which is to be installed between now and October 1980.

The remaining plants are under review, and a determination of the need for additional shutdown capability has not been completed. For the plants not yet evaluated, we would expect similar findings.

for those plants not yet evaluated, and those plants for which the staff has required enhancement of the fire protection systems, the staff believes that the probability of occurrence of sev re damaging fires is acceptably low for the interim period until staff evaluations and licensee enhancements are completed. This conclusion is based upon the information discussed by the Browns Ferry Fire Special Review Group in NUREG-0500 and upon the additional defense-in-depth protection provided by the staff's overall fire protection upgrading program which prpvides (1).

controls over ignition sources, combustibles and access to the areas, (2) physical separation and use of flame retardants to delay or prevent propa-gation, and (3) fire detection, fire suppression and trained fire brigades to effect prompt manual suppression of fires.

Table l of Review Ffndin s 'tatus for 0 eratin Plants Plants Shown to Date to Have Ade uate Shutdown Ca abilit (1)

Arkansas 1 Fort Calhoun Kewaunee Maine Yankee Oyster Creek Turkey Point 3 5 4 Vermont. Yankee Browns Ferry 1, 2 5 3 Plants Mhich Have Installed Alternate or Dedicated Shutdown S stem D. C. Cook 1 5 2 Hatch 1 8 2 Ft. St. Vrain Plants Reouirin an Alternate or Dedicated Shutdown S stem Brunswick 1 5 2 (To be installed by January 1979)

Haddam Neck (To be coordinated with SEP schedule)

Oconee 1-3 (To be installed by October 1980)

Pilgrim (2)

Rancho Seco (To be installed by December 1979)

Robinson 2 (2)

Three Mile Island 1 (2)

Three Mile Island 2 (To be installed by March 1980)

Trojan (To be installed by 'June 1979)

(1) All except Browns Ferry subject to verification analysis.

(2) Schedule for implementation not yet determined; staff expects it to be prior to October 1980.

~0 1 i' '7 p

HOV -4 1977 P.. E. Tiller, Ofrector Reactor Opera fons ~ Procurer.;s Ofvfsfon Idaho Opera tf on s 0, i ce - OOE

=

Idaho Falls, idaho 83~01 M FUEL ASSElloLY l'HAEC!Llh'C.l

~ ~

RESPO>tSF AtlALYSIS Stir316 77

, Ref: (a) P,. L. Grubb, P',.'0 Fuel Assembly Hcchanical l?espons Analysis, Idaho llatfonal Erqfneerfnq Laboratory, RE-E-77-1'il, t! rch 1977 (b) P.. L. GruLb, P!.'R Fuel Assr;..bly ]',echanical P. s,",onse Analysis, A. nC,"..ent tlo. 1, Idaho !l tfonal Enoine ring Laborato-,v, RE-E-77-1>10, l! rch 1977 (c) A. L. Gr Jbb and 0. F. Safiel1, 3r, llor>-Linear Lat=ral tlechanfcal P. sponse of Pr"ssuri-ed 1",ater Pcactor F el Asse;..blies ASl'E Paper 77-le/i:E-13, Oece:-..ber 1077 H. ltuno, ll. l'l'zJ.a, and >l. Ts>>-.una> Oevelo-.-..ent of Advanced flethod For Fu"1 'eis.-.:ic Ana';.':sis, c >", 1n.ernational.Con-.eronce on Struc u. a> "'charics in >".eac qr Technology, San Fr=ncisco, Cali ornia, U.", August, 1J/7 (e) R. L. Gr >bb, Feasability S 'dy for Dourdirc the Lateral PHP.

Fuel Asse;.,bly llechanical Response Analysis, Idaho liational Engfneerfng Laboratory, RE-E-77-160, Re(. 1, 'uuly, 1977

Dear llr. Tiller:

A para-~tric s.uCy to assess the e fec" o varfatfons fn core plat ...otions on fuel asse.ably spacer cr >C cru-hinc loads is curr . -ly la prorress.

su A

pl epare a the ro "est 0 the " ~ 'al

".arv . scrfptfcr> of th>is study including prelii-.,-irarr resul:s has been

.et'uIatol " C>J i lssiion s C v sf on

o. Sys ~ Sa-, ty, Ccr" Perfcr..-nco "ranch. Pesul ts of this stu"y indicate a s-..all .arfa.ion in c"re plate,r ",uency ray have a sir,nf. icant

'hat

.effect on spac r rrfd crushing loads. As the s.u".'y is not cor.piete,, these results should be considered preliminary.

A necharisfn i. s p"stulate" fn Reference (e) i;hfch indicated t1 at t>"e fn-put core plate rotfcn c"uld fonf icantly a fcct sp c r arid crushinn

'loads. The prfr~ry objective o. the present s.udy 'zs to deto~iine if

~

this rachanis'n could ho shown to evis-. A secondary nbfective fs to compare linear end nonlin ar analysis ". cnniq> cs. in sur.1ry tl:en the purpose of this s. Cy is ".~ofold:

(1) St: >5 ca11 J dete~ifne, 4'e oft t of cor~ plate f nrunr and,".acr.f "u o on "..'".e . >el assc;..bly czxf>>~u~ s-acer qrfd crushfna lca" , and 5~i> f st'cz1 1 / c '., .re 1 ne>r +nC onl One r a Qlvsis ~. I)Q.

~ 5 for late>r.'.1 iuol:ssc.-,."1y wc.an>cal ruspcnse foal an atte.">

to sfr;pl f-.y ..".c n".nl inear analysis.

R. E, T:lier llov. -4 1077 Stig-316-77 page 2

The structural model uti',ized to analyze tl.e uel assembly mechanical response is basically described in Refer.ences (a) throuqh (c). IIIo exception included in tne present s .udy are the use of fu asse..Ibly 1 experimental frequencies and mode shapes and utilization o, the IIIethod

. presented in Peference (d) for .calculation or soacer qrid crushinri loads. The nomina forcing function, cnre plate accelerations, are 1

pIesente'r. I',eference (b). tlhile eir.h: varia icns on tile Treque:Icy and ar plitude of coI.e plate IIotiors are to 4e considered, only the four extrer.;e cases are aCdressed in t.lis discussinn. The four cases aI e l 10".. var iation on lrequ ncy ilnd + 1IO'.: variation or tile anplitude. It

'5 noted tlIat all the frequencies cc~tainerl in the cole plate motior.s are varied the sa.".e amount. t<on'irear. dynamic analysis as describer'.

in Peferences (a) through (d) is in prncre:s and preliminary results are plovideC in Table l. A lineal 3'lal"5'ls ls also bc Ag pulsued us lAQ the methods outl)ined in Referen=e (e).

TABLE 1 OF PEA< SPACER GP.ID CRIISH HG LOAD TO T}IE

'ATIO 1

HOHIHAL CURSiIHG LOAD Haximum Crushing Load/Ilomina 1 'Crushing Load

.:Spacer Grid El'eva t 'I GI'. -1C!; Frecuency +10" F'recue. cy -10" Amplit de ~lG".. Ampli ude

'Cen. er 1

) ~ Ib 0). 804 0 Ojc 1.G4 Cen'er-);p 1.<5 O.S<< 0.84)2 1.23

'enter-dc'.n 2.11 0. 830 0. 335 1.25 IOP 1.34 O.c45 0.771 1.34 Bottom 1.56 0.808 0.863 1.26 Hominal cI ushin. Ioiid is the peak splcer gr'id crushing lead obtained from the base case core plate .-,.oticns.

Based on the re suit 5 lA Iab Ic 1 1 t do 5 appeal ifIar. 3 vaI atiol 1 TI e.,llency 1 1 of ter. -erceni eTTe ts 3 51 cAl; '.cant chance ir. tl:e soacer grid crushirg 1)cacs.

.his irCicates that :ar t 1 on ir, .his para,etar II'ay be in nr='er -:nr ;his t -..e oT non! Ir="'I. an is ~ li =hou !C e

ile "oir ted ou "lla

the ;:coel 5:u<<e'er.'.","-

"ur,".ose cI:h 5 s..='y iras r.ot a dir"".

resents a ::=ner 1GIl ana)ysls 0 ce 1 1 i g) cfi t, i'J : to ce'erlI re 1 Ir t.".e;ec'.-.anisim pcstu!a'.ed

R E. ?il>er N6" -1 I"il St i g-316-77, Page 3 Reference (e) could actually be elicited in :he nonlinear analysis. The r. ch-anisin appeai s .o e..ist; hcreLy causing concern that per.. nent de=orna.ticn of spacer grids ray cccur.

Upon cor.:pletion oT this study the conclusions presented ln f',nference (bj

>rill bd reasscssed.

'ery truly yours, R. R. Sticer, i'anager Peactor Dtiavior Div'sion BFS:clj

.": ~

Y. Stcl lo, tiRC-DGi~

~

'. B.'in, P. S. ChecI:, !;.-.C-rsS S

R.

~ ttRC-DSS J: t~a.'scn...t:C-".Ss

0. tleyer, <>RC-"SS D. F. Ross, t!RC-iSS R. 1t. Yiehn, EGiG Idaho bcc: R. L. Grubb R. 'W. Hac k C. P. !!core C. F. Obenchain+':.~

G. L. Thinnes v'.

R. Thcmpscn P. H. Yander Hyde L. J. Ybarrondo Central File File

P~

p ~ ~

~0 ~y Uh'ITEO STATES NUCLEAR REGULATORY COf1MISSlON ENCLOSURE 9 WASHINGTON, D. C. 20555 August 18, 1977 HEtlORAHDUtl FOR: E. G. Case, Acting Director

~

Office of Nuclear Reactor Regulation IIo;.~

C'" >.0~'jul '

FROtl: Stephen H. Hanauer, Technical Advisor to.

Director for Operations ~'xecutive

~ $ g ~

SUBJECT:

INTERACTION BETWEEN CONTROL SYSTEM AND PROTECTI I SYSTEN The Zion incident of July 12,'1977, apparently shows a design defect as well as the obvious gross management deficiency. The 31 dumny signals disabled the primary system level control, which initiated a transient involving decreasing level. Concurrently, the same sequence of events disabled portions of the protection functions associated with the same level. Thus a single sequence of events caused the transient and paralyzed the safety provided for that very transient.

Westinghouse designs are characterized by the large number and types of interactions between control systems and related safety'systems.

They think this is great. I think it is unsafe. This feud has been going on for years.

I have not so far been able to find out whether a single signal or group of signals went to both control and safety, or whether the interaction was more obscure. It almost doesn't matter. I also don't know (and don't much care) whether the interaction, whatever its nature, is allowed by the various meticulously crafted clauses in IEEE-279.

For existing plants, I believe the lesson of the Zion incident should be taken to heart and acted on constructively. The fact that, this time, nothing bad happened is a tribute to good operator action and defense in depth, and should not keep us from learning the lesson.

All interactions between control functions and safety function should be reviewed in the light of this experience. A statement that no such dummy signals are allowed is not to the point; next time, some different and not now foreseen sequence of events may start the ball rolling.

What is needed is adequate independence of control functions from safety functions that provide against control IIIalfunctions.

~0 ~e I ('~~ ~

0 E. G. Case August 18, 1977 For future plants, we have RESAR-414, with a new "Integrated Protection System," which includes interactions between safety channels and between safety and "non-safety systems for monitoring and control" (PSAR, p. 7.1-27). Such interactions seem to be on a scale far beyond present practice and involve a complexity (multiplexing, data links between computers) not previously encountered.. The philosophy (old and new) is, "Westinghouse considers it advantageous to use certain information derived from protection channels to control the plant" (PSAR, p. 7.1-62).

The acceptability of all systems, Westinghouse and non-Westinghouse, old and new, needs to be reviewed in the light of the Zion event and any unacceptable interactions removed.

/

I:

(

(

Stephen H. Hanauer i.i CC. ~

Technical Advisor to

• Executive Director for Operations cc: L. V. Gossick S. Levine E. Volgenau R. Ninogue

~0 ENCLOSURE 10 q7 5

O The complete text of this memo, with the exception of the writer's identity, is as follows:

"Memo from [Name Deleted]

BOB POLLARD ENCLOSED NOTE FROMM")IS FLOATING AROUND THE INDUSTRY AND DESERVES YOUR RAISING IT IN SOME PROCEEDING.

OBVIOUSLY KEEP ME OUT OF IT.

[Initials Deleted]"

Ponfond Gtnesol El~Wc f i ~~{

~0 ~ l

) 4

~y ENCLOSURE 11 ll Q

CO D.-.csri~mr 2~ 1976 Dear liny tlmnks for an6 the enclosure.

V33. try to put it to recent yeux nets good use.

X ne longer have an office eo .the lbshington address you have is ouMateR. X now vora out of my hone which S.s:

.'incoroly>>

Robert D. "ol3.ar6

January 14, 1977 I passed the Vestin~house decmnent on fuel handling acciifents inside coataimnent to sono interro~ers and. p sent a copy to the %C ~

h cony of ny'etter to Gilinsky anti the re>ly from QO Strauss are enclosol, >s you can sos they wou1d like nore im"oz...ation, If havo any other info you w~t mo to forward to HRC~ send. it aloe@,

Conti ary to Strauss'urgostion~ th re are ro circu;,stuccos where I would dimloso th~ souroo Sincerely~

Robert D, Pollard.

UNfON OF COf! CERNED SC)=NTfSTS January 3p 1977 Commissioner Victor Gilinsky U.S. i%clear Ro~atory Cc.,aission Washington, D.C. 20555

Dear Cemmissionor:

We received thc enclosed docunent from an individual who wishes to remain anonyrttous. He are sending it to you in the hope that the liuclcar Rem3.atory Conrtssion will take prert"t action to protect thc health and safoty of the pubs.c from the known risks di cussed in t.'tc document.

The document correctly indicates that the consequences of a fuel handlin accident inside the reactor containmcrt bui ding are not considered by the 'iRC in deciding whether a nuclear power plant should r-..cc9.vc a license. Zn addition~ the document indicates that hhstinghouse believes th-t a fuel handling accidort insioe centain-mont could result in radiaticn doses to'he public in excess of 10 CFR . art 100 limits~ i.o.~ in excess of 25 rct.. to thc wnole body and 300 rcm to the thyroid, In vic~r of these statements, it ppears that a fuel hmdling accident inside containmcnt is an "urwcvicwcd safety question" and a "significant safety hazard,"

P Ue recommend that th" 11RC review the desi'nd procedures of each operatin>>

nuclear power plant tc determine whether a fuel handlin~ accident inside contain-mont will result in doses that "arc well withia thc pCidcline values of 10 CFR Part 100~u as specified in Section 15.7,4 of the Standard Review H.an, Until such rcvietrs are completed, we be3i.eve that the 3',2C should issue orders to halt all refueling operations in pro~ress and to urohibit a3't future re&soling op "a-tions, In a~Mition, we believe that it is a~proyriate for the iiRC to initiate investi"atiens to doterrn*.ne whether Section 206 of the Energy Reorganization Act of 1974 has bcon violated by individual directors or responsible officers of Westinghouse and other firms which received tno onclosod docuntent We would ap~reciate hearing from you prcmatly rc~arding the action that:tHiC will tmc te resolve this matter. We also would like an explanation of the reasons for 1BC not previous'ty requiring analysis of a fuel hartdling accident inside containmcnt and the ste~s that will bc taken te correct thio" deficiency in the 1- censin+ process, By co~ of this letter, " are also scnmn~ the enclosed document to the chairmen ef the .>dtvisory Cori. 'ttc,; on Reactor Safe,";u."zds, the Atomic Safety and Licensing Board panel and the Atomic Safety and Liccnsin~ Appeal Parol.

Sincerely~

Robert D. Pell"rd 12C8 Massachusetts Avenue ~ Cambridge, Massachusetts 02138 ~ Telephone (617) 547-5552

g

~0 0 2 4. It Q ~

~y As you, are aware, a tuel handling accident in the spent. fuel storage building

$ s anh'ly'zed in plant Safety Analysis Report:. Th>> assumptions uti) ized for this analysis are outlined in }regulatory Guide 1.25, "Assumptions Used for Evaluating the Poteiitial Consequences of a Fuel liandling Accident in the Fueland}ing and Storage Facility." The off-site consequences of this accident are compared to 10Cf R100 limits of 300 rem to the thyroid and 25 rein whole body dose in t}ie Safety Analysis Reports. In addition, the f'.C co.".ipar s the resultant doses with unofficial limits of 30 rem to the thyroid and 5 rem w}iole body dose. However, a fuel handling accident inside the containment is not addressed in the Safety Analysis Reports, other than indirectly in Standard Tech Specs. M is not aware of the t'-RC bases for not addressing a fuel handling'accident inside con-tainment, the bases may include:

1. The assur:.ption that the containment will be isolated during refueling operations;

2. that the containH:ent could be isolated quicLly enough to limit off-site co>>s quences; or

3. that filtration capability coi".parable to that in fuel storage building exhausts exists in the containment purge exhaust.

Tho se bases are similar to the bases used to address the fuel handling accident in the fuel handling building. Information a;ailable to us, including results of scoping analyses using very consei vative assu".:ptions based upon Regulatory Guide 1.25, indicates that site boundary doses in excess of exposure guidelines set forth in 10CFR100 could result from a fuel }iandling accident inside containmen. if one assumes no credit for containment isolation, iodine filtration, or mixing within containment. In addi tion to using Regulatory Guide 1.25 assumptions in'he scoping analyses, we assum d op ration of syste.".is which would result in the most conservative dose. For exa~i:.=.le, it was assumed that a push-pull type or exhaust only swee;p veritila-tion system is in operation over the refu=ling canal so that aci.ivity releas s are routed inv:ediately to the purge exhaust. Much o the infor;.ation required to do an evaluation for specific plants is not availab)e to us. 'he do recoi"iiiend, however, that you evaluate the consequences of this pot ntial incident to assure that unacceptable doses are not a probable i'esult. Since the t'RC regulations do not require the analysis. we do iiot believe this situation requires reporting to the NRC unl ss youi. engineering evaluation shoi:s unacceptable results. In acco::.plishing tr>> evaluation for your plaiit, we recc ':::end t...'.t you us>> Regulatory Guid~ 1.25 as'.uii'ptions oi ot!i;": con.-.< rvative. justifiable parai::eters. ke also believe that. yr u s}ould not talc. credit'for the function of ary system or component that is not. qualified for op:ration during this particular incident. For examiile, we thin}: you might take credit for equip-ment not qualified for the post accideiit contaiiiment environment but seismic 'ualification may very well be required. Please feel free to contact us if further 'nformation or assistance is required.

C'

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1

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 January 6, 1977 Mr. 'Robert D. Pollard Union of Concerned Scientists 1208 Massachusetts Avenue Cambridge, Massachusetts 02138

Dear Mr. Pollard:

Commissioner Gilinsky has asked me to respond to your letter of January 3, 1977, forwarding an otherwise unidentified .attachment raising certain questions regarding fuel handling accidents. The letter and its enclosure have. been forwarded to Mr. Lee Gossick, the Comnission's Executive Director of Operations, with a request that he provide it to the proper Staff offices, and assure a timely, di rect and appropriate response to your statements of concern and to the issues raised by the partial document you have supplied. The COIInIission has asked to be promptly informed of the outcome. If there are circumstances which would permit you to supply further contexting information regarding the attachment forwarded with your letter, the Corrmission would appreciate having that information. . I am sure that procedures can be devised to provide the necessary con-fidentiality for your source, should that be an issue in this regard. Obviously, knowledge of the full context will assist the Commission's staff in evaluating the concerns you have raised and the documentary fragment you have provided. Sincerely, Peter L. Strauss General Counsel

y I

UNITED STATES OF A!AFRICA CO.'DEMISSION NUCLEAR REGP ATORY In the Hatter of )

                                         )

CAROLINA POi'ER AND LIGHT CO&'ANY ) Docket No. (s) 50-400

                                         )                         50-~oa (Shearon Harris Nuclear Power           )                         50-402 Plant, Units 1, 2, 3, and 4)         ).                        50-403
                                         )
                                         )
                                         )

CERTIFICATE OF SERVICE I hereby certify that I have this day served the foregoing document(s) upon each person designated on the official service'list co..piled by the Office of the Secretary of the Conmission in tn s proceeding in accordance ~"ith the requirements of Section 2.712 o 10 CFR Part 2 = Rules of Practice, of the Nuclear Regulatory Commission's Rules and Regulations. Da d at 4'ashingto, D. this C.

                                   . ~ +A/47" iOffice of the Secreta        y  of the Commission

UNITED STATES OP AMERICA NUCLEAR REGULATORY CO'MISSION In ehe Matter of )

                                        )

CAROLINA POTPi.'R AND LIGHT CO:PANY ) Docket No.(s) 50-400

                                        )                      50-401 (Shearon-Harris Nuclear Paver            )                      50>>402 Planes, Units 1-4)                  ~
                                        )                      50-403
                                        )

SERVICE LIST Ivan H. Smith, Esq., Chairman Richard E. Jones, Esq. Atomic Safety and Licensing Board Carolina Power and Light Company U.S. Nuclear Regulatory Commission P.O. Box 1551 Washington, D.C. .20555 Raleigh, North Carolina 27602 Mr. Glenn 0. Bright George P. Trowbridge, Esq. Atomic Safety and Licensing Board Ernest L. Blake, Jr., Esq. U.S. Nuclear Regulatory Commission Shaw, Pittman, Poets 6 Trowbridge Washington, D.C. 20555 1800 "M" Street, N.H.

                                             >tashington, D.C. 20006 Dr. J.V. Leeds,    Jr.

Rice University Thomas S. Erwin,. Esq.. P.O. Box 1892 P.O. Box 928 Houston, Texas 77001 Raleigh, North Carolina 27602 Counsel 'for NKC."-Ft'a'ff" Office of the Executive Legal Director Dennis P. Myers, Esq. U.S. Nuclear Regulatory Commission Attorney General's Office 7',ashington, D.C. 20555 P.O. Box 629 Raleigh, North Carolina 27602 Alan S. Rosanthal.. Esq,. ,Chairman Atomic Safety and Licensing Appeal Mr. 0. Gene Abston, Acting Director Board Office of Inspector and Auditor U.S. Nuclear Regulatory Commission U.S, Nuclear Regulatory Commission Washington, D.C; 20555 Uashingeon, D. C. 20555 Dr. John H. Buck Atomic Safety and Licensing Appeal Board U.S. Nuclear Regulatory Commission Hashington, D.C. 20555 Michael C. Parrar, Fsq. Atomic Safety, and Licensing Appeal Board V.S. Nuclear Regulatory Commission Washington, D.C. 20555}}