ML19305A070
| ML19305A070 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 12/11/1978 |
| From: | Pollard R UNION OF CONCERNED SCIENTISTS |
| To: | |
| Shared Package | |
| ML19305A064 | List: |
| References | |
| TAC-11299, NUDOCS 7901020507 | |
| Download: ML19305A070 (63) | |
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LIMITED APPEARANCE STATEMENT BY ROBERT D.
POLLARD UNION OF CONCERNED SCIENTISTS DECEMBER 11, 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) 7 3010 2050 ~[
6 l
1 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.
UCS' primary areas of interest are the health, safety, environmental 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
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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.
The 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
o 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.
1.
I appreciate the courtesy extended by Ms. Bell in delivering i
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.
From February 1964 to April 1965, I served as the senior reactor operator and supervised the reactor 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 magna cum 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.
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I subsequently advanced to the positions of Reactor Engineer (Instrumentation) 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).
In 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:
Indian Point Unit 3 in New York; Comanche Peak Units 1 and 2 in Texas; Catawba Units 1 and 2 in South Carolina; an'd McGuire Units 1 and 2 in North Carolina.
Copies of the " job description" for
the position of Project Manager and the last NRC appraisal of my performance in that position are attached (Enclosures 3 and 4 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.
~. TECHNICAL SUBJECTS DESERVING ADDITIONAL 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 ha delivered by December lith.
Therefore, I have chosen just a few examples which support my conclusion that, if these matters are 1. eft 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 Safety-Related Electrical Eculpment Based on the Board's order, " Order Regarding Conclusion of Evidentiary Hearing on Interim Operation," date~d 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.
If 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-
-6 bility of a nuclear plant catastrophe is no greater than 10 to 10 per eactor year and since no competent witness would
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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 perfermance 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 2
transmittal letter, pages 7 and 10 of its Enclosure 1 and its
, 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.
i
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.
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Trojan Operating License was granted.4 Sections 3.10, Seismic Qualification, and 3.11, Environmental Qualification, simply refer to Section 7.8, 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'he 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.
I 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 Staff's review of equipment qualification.
With the above as an introduction, I 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 if issuance of the operating license had been sub-ject to a contested hearing on some safety issues, the license could not have been issued on the same day as the SER Supplement.
As it was, the sole issue admitted into contention during the 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 referenced the generic submittal from the vendor was directed to make no review of the generic submittal.
This is the reason I prepared the second enclosure to the April 19, 1974 letter transmitting the EI&CSB 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, NUREG-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 1.9, 1974.
The current status is as followst 1.
WCAP-7821 (NP) - The Staff received additional informa-tion by letter dated September 29, 1978.
The regiew is scheduled to be completed by January 1, 1978.
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l 5.
The obvious exception, not applicable here, is the case where l
the same individual had been assigned to review both the j
generic submittal and a license application referencing that generic submittal.
1 6.
The designations (NP) and (P) following the report numbers identify, respectively, the non-proprisoary and proprietary versions of the topical report.
1
7 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 qualified (as well as environmentally qualified) in spite of the fact that the generic qualification programs of Westinghouse remain unreviewed and/or unacceptable to the Staff.
Foi 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, 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.
Much 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).
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-13 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
'eplaced or removed from safety circuits within six months.
r 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 i
qualification (which includes radiation qualification).
If the Board adopts the first or 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 l
l by Regulatory Guide 1.89, as the method of demonstrating conformance with the Ccmmission's regulations.
The Board can note from 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 program.
I cncountered 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 I
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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 condition 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 and 1.100, respectively.
Since Regulatory Guides are not regulations, requiring con-formance with Reculatory 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 not met, the plant must remain shutdown.8 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.
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 j
8.
I discuss the inadequacy of the direct testimony (in this proceeding) on the subject of equipment qual:.fication later in this statemer.c.
See page 23.
I by tha Stoff as a basis for approving the design of the dedicated 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 ealth and safety of the public.
(See Criterion 3 of Appendix A to 10 CFR Part 50.)
The design of 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 cannot be taken up by this Board.
It is correct that UCS' Petition for Emercency and Remedial Action is before the Commission for vv
reconsideration.
However, specific details of the Trojan Nuclear 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 : c 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 sponte review of the licensing proceedings that author-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' issues in this operating license proceeding."10 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 without 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 not alert us to the existence of a generic problem bearing on the particular aspect of plant design under discussion."
(ALAB-491, footnote omitted.)
9.
Gulf States Utilities Company (River Bend Units 1 and 2),
ALAB-444, 4 NRC 760 (1977).
10.
Virginia Electric and Power Company _ (North Anna Nuclear Power Station, Units 1 and 2), ALAB-491 (August 25, 1978).
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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 took 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 existen e of the Staff's Technical Safety Activities Report that the public became aware of the existence of a large number of 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.
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 11.
I am aware of the regulations pertaining to "show cause" 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.
- 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 Staff 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 l
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- in the frequency of core plate motion during blowdown could more tha'n 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 coolable 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 systems 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 Westinghcune design is unsafe.
I am supported in this conclusion by at least some members of the' Staff.
For example, following an event in the Westinghcuse-designed Zion plant, Dr. Stephen Hanauer, Technical Advisor to the M
-.-.--.,7 e
. Executive Director for Operations, expressed his view that the Westinghouse design was " unsafe."
In the Zion event, workmen j
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 l
systems. Westinghouse and non-Westinghouse, old and new, needs I
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.12 As I noted above, I have not discussed every technical issue which I believe should be examined before the Trojan Nuclear Plant i
is permitted to resume operation.
I recommend that the Board require the Staff to identify all the unresolved safety issues l
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 l
l list of all unresolved generic safety issues that may be germane l
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 l
Activities Steering Committee."
These minutes disclose a host 1
of other safety issues which the Staff has not included in its published listsof 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.
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- METHOD OF EL1 CITING 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 original qualification of the equipment."
(Tr. 2339,' emphasis added.)
During examination by the Board, Witness White continued this theme by stating that
" essentially everything that was qualified prior to that time was still qualified." (Tr. 2351. )
The_ key point, which I addressed at. length earlier, is whether the original equipment qualifica-tion was acceptable.
It is this point which the Licensees' witnesses 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-L related electrical equipment.
He testified as follows:
l 1
i
"The effect of the postulated earthquake on such equipment as batteries, switchgear, cor. trol 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-l fications' accompanying Mr. Knight'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 Control Building for Interim Operation," undated, emphasis added.)
I have also examined Mr. Herring's most recent testimony, " Testimony I
of Kenneth S.
Herring, Office of Nuclear Reactor Regulation, on Floor Response Spectra and Qualification of Safety-Related Equip-l l
ment and SystcmL in the As-Built Control Building Complex,"
November 25, 1978.
. j The subjact of equipment qualification is mentioned three times in Mr. Herring's 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 difficulty expressing their thought:s 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, ar 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
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 innt w able 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.
i It should suffice here simply to state that my resignation from 1
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 conclusicn regarding PGE's professional employees.
Attached are copies of correspondence between me and an individual employed by PGE in a professional capacity (Enclosures 10, 11 and 12).
I have deleted all information that could aid in identifying the PGE employee and my home address and telephone number. 0 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-i sioner Gilinsky and the reply from the General Counsel.
There are at least two significant aspects of these documents.
- First, l
l 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 j
l thought that this matter should be brought to NRC's attention, r
l I
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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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~28-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.
But 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.
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.o Reccter r.nginaar (I..atrum2ntation), GS-l'.
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Fin:CTIONAL 5A~D:IC:
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.s q W 'Y Servas as a highly qualified specialis in :he :i'* Id M c:or instru=entation and control in perfor=ing technical reviaus, analyses and evaluzzions of sys:c=
and componan designs necessary to che safe opera:Lon under normal, abnor=:1 cad e=argancy conditions of pouer, testing, production, and rasearch reactors, including DOD and AIC-owned reac: ors as wall as licensed and authorised facili-ties.
REO;;*.Al Dt "*IES :
Participates as a senior =a=ber of the Directorate of Licensing, Electrical, Instru=en:ation and Control Systems 3 ranch, uhosa function is pri=arily one of perfor=ing technical renews, analyses, and evaluations of designs of sys-te=s and co. ponants necessary to the safe operation of rasctor facili:1es under nor=al, abnor=al, and e=ergency condi: ions for the purpose of (a) determining the adaquacy of the bases for such designs, (b) of datar=ining :he adequacy of such designs co = set these basas and to withstand tha li=its of environ =antal effec:s withou: loss of =ini=u= required functional capabill:1es, (c) of deter =ining the acceptabili:y of' procadures for fabrica: ion, inspection, testing, and pos -licensing su vaillanca of such designs, and (d) of developing guidelina proceduras, ma: hods and =odals for the systematic evaluation of such designs by the Division.
Reviews Safa:y Analysis Raports as to the adequacy of the presented data pertaining to ins ru=entation, controls, and electric power and to the sound-ness of conclusions made on the basis of the prasented infor=a:Lon and preparas reports of such reviews.
Develops standard procedures, =ethods, and =odels for evaluations to determine whether or not the design of reactor procaccion syste=s, controls for engineered safety features, safety aspects of regula:ing sys:e=s, and e=ergency power sys-te=s is ::eatad in an acceptabla manner.
Evaluates industry and AEC-sponsored research and develop =ent progra=s direc:ed towards establish. ant of additional basic informa: ion on reac:or instru=enta-tion and control, and to the use of such infor=ation for safety evaluation pur-poses, and correlates and interprets the resul:s of such programs for the general use of the regula:ory staff.
Preparas :echnical studies and repor:s bearing on' unique and unusual develoh:=an:i in the field of : sac:or plan: instruman:ation, control, and alac:ric powar for presentatica of the' Advisory Co
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- ea on 2aac:or Safaguards.
Raco== ands, through tha Branch Chief, safety res,earch progra=s to be sponsored by the AZO.
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Reactor Engineer (Instru=entation), CS-14 Electrical, Instrumentation anc Control systems Branen DLeectorate of Licensing Page 2 Confers periodically wich :cchnical represen:a:ivas of indus: rial organica f ona and o:hcr AIC divisions to discuss nucicar safe:y proble=s involving creas of concern related to reactor plan: ins rumen:a: ion, control and electrical power.
Par:Icipa:es, frc= tima to ti=a as tachnical represan:a:iva of the Directora:a of Licensing on various AIC commit:cas, subco--4::aas, panels and task forca assign = cats as wall as cachnical and professional society cc :intaes such as the A=erican National S:andards Ins:ituta, the Atarican Nuciaar Socia:y, tha Ins:i:u:s of Elec:rical and Elec:ronics 2nginears, and othara.
BASIC SKILLS:
Genera knowledge of the principle =, :heories, and practicas in tha field 'of nuclear engineering with specific knowledge of reactor plan: ins:ruman:ation, control, and elec:ric pcuar systa=s is required.
Competency must be suffici-eat :o independen:ly analyze and evalua:e reactor concepts and fea:ures pro-posed by organi:c: ions ppacializing in th'a nuclear ficid, particularly with respec: to :he reac:or pro: action sys:a:s, instru=antation and con:rol for engineered safety fea:uras, reactor regula:ing sys: ems, reactor plant dynamics, i
and elec:ria power systa=s.
l The basic skill requiremen:s are considerably in excass of thosa secured-thrsu h fw6 cal educacion at universt:y leval (3.S. degrea) and are comparable s
to those achieved fro = graduate laval training or from specialiced experienca in instru=enza:ica and control in 4pplica:1ons to reactor technology.
Knowledge of licansed and autho:1ced as well as DOD and AEC-owned reactor installaticas and o,perations is required.
CONTACTS:
Contacts top technical and supervisory personnel of the AEC, other Gover==an:
l agencias, AEC con:rac: ors, indus: rial organizations, research ins:1:u: ions, l
universi:ias, and professional socie:1cs :o discuss techaical =a::ers relating to reactor plan: instru=catation, con:rol and electrical power syste=s.
RESPONSIBILITY 70*t DECISIONS:
Supervision Received:
Chief, Elec:ric 1, Instru=cata: ion and Control Systa=s Branch.
Supervision is general on technical =a::ars wi:h full au:hori:y to ac: in =st:ars wi hia the fra=avork of the broad functional assign =en:.
The Ad=inistra:ive guidas are Division and overall HEC policy and precedan:.
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Electrical, Insert =tentation and Control Systems Branch Direccorate of Licensing Page 3 Independen: Ac:fon:
Responsible for esking recon =endacions for action :o be :akan by the Chief of the Electrieni, Instrumen:acion and Control Systa=s 3:anch.
Develops scandard procedures, 2: hods, a.d codels for those aspeca of safc:y evalua: ions involving physio-che=ical considera ions.
S*J?IRVISION :
None
!!O*tNING CONDICONS:
Nor=al offica conditions while at official sta: ion.
I:gosure to 11d radia-don from reac: ors =ay be encoun:ered occasionally during field trips.
ETFO:tT:
Nor=al effort involved in any adminis:ra* ive position.
Increased physical.
affor: =ay be required while ca field trips.
ENCLOSURE 2 gI *
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UNITED STATES f{
f ATOMIC ENERGY COMMISSION I
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NOTE TO Robert Pollard's Personnel File
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I have received the attached note from Mr. Herschel Specter Y;~ N comending the performance of Robert Pollard of the Electrical, j
Instrumentation and Control Systems Branch.
Mr. Specter was the Licensing Project Manager for the Indian Point Units, and
)
is, therefore, well qualified to speak to Bob Pollard'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 under the Commission's standardization policy.
I find Pollard exceptionally expert in his technical area and articulate and effective in his work as a technical reviewer.
I am pleased i
to add my coments to those of Mr. Specter and to forward them to Bob Pollard's personnel e.
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oseph M. Hendrie, Deputy Director,
for Technical Review Directorate of Licensing Attachmen t (As Stated) cc:
F. Schroeder 9
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ENCLOSURE 3 Projec: Mana?.or, GS-14 1
Direc:orace of Licensing
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FUNCTIONAL STATEMEST Experienced in the engincaring and physics aspects of nuclear re ccors, the incu ben: plans and coordinates the technical reviews, analyses, and evatus: ions of applica:icas for licenses and authcrizaticas for the cons: rue:ica and operation of reactors and :he reviews of cer: sin aspects of their design and cperacica.
REOULAR DUTIES l
Plans and coordina:es the pre-review of applications :o deter =ine if they are sufficien:ly co=plete to accept as an application.
i l
Plans and coordina:es :he review of Safe:y Analysis Recor:s as :o the adequacy of the teqhnical and engineering design da:a and inscr=ation I
contained therein, che soundness of the basis for che conclusicas of the proposa.d designs and operating procedures.
Coordina:es the pre-paration of the safety evalua:Lon in conjunction w1:h such ravicus.
Serves as projec: =anager for group evaluatten of pctrer reac:cr license applican:s for which he has been assigned responsibility.
0:nf :: vi:h c0Lr.* cal..e..d.L.Li.. v: w6asul**Liu = 9'ouvsIwg aus r
reactors to identify or resolve general questions of doubt concerning design and opera:ing characteris:ics which have a bearing on saic:y.
Coordina:es the prepara: ion of safe:y evaluation reports relating to license applica:icas for power rese:or plan:s, as well as military and AEC reac:or plan:s for presen:ation to the Advisory Cc==ittee on Reae:or Safeguards.
A:: ends such =eetings and subec==i: tee =eecings as a represen:2:ive of the Direc:orace of Licensing's evalua:Lon s:aff.
May participa:e at public hearings on reac:cr licensing proceeding:
to assis: c:her AIC representatives or testify as an AEC staff wi: ness to present technical tes:i=or.y.
Serves as a =e ber of reactor inspection tea =s as c' y be necassary to a
discharge rese:or safety evaluatica and judg=ent.
Plans and coordinates the review of nuclear safety aspects of propo als to build any AEC-owned reac ors exa=pt frc= licensing.
Assists in :he prepra: ion of technical specifica:icn: for opera:ing reac: ors; reviews cpera:ing experience repor:s during initial phases of opera:icn; and evaluates reques:s for licensa a= cad =ca:s and technical specfica:ica changes during initial phases of operatica, utilizing :he exper:ise of persons outside o d
5'-
'- adia:a division where necessary.
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Project ::anager, CS-14 Directorate of Licensing BASIC SKILL Knowledge cf the princioles, theories and practices in the field of nuclear reactor :achnology with specific kameledge of reactor and nuclear engineering. Cc=petence =us: is sufficien: to adequately evaluate various proposed reactor con:epts and codifica:icr.s primarily as related to reactor construction and operation.
Knowledge of operations at AEC-caned reactor installscions.
Experience in the field of reactor core design and operation to supplecan: basic engineerin; training.
Basic skill requirc an:s are in excess of those secured through fo: al education at the university level C3.5. Degree) and are comparable to thosa obtained frem graduate level training or specialized experience in reactor technology and related subjec:s.
CONTACTS c=O
~
Contacts are with top technical perscanel in AEC,' AEC con::accor=,
cess $
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industry, and other governcen: agencies to discuss technical cat:ers u
relating to the hadards inherent to design, opera:Lon, and sita
$3553 location of proposed new reactors or significer.c codification of G3[@)
crfj existing reactors,
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RIS?ONSIZILITY FOR DECISIONS c
Suoervisien Received (kj,
$.-0 Assigned 3 ranch Chief, Directorate of Licensing.
gggjg Supervision is general on technical =a::ers, but specific on policy and opers:ing procedures.
Ad=inistrative guides are AIC Manual, Cc==ission Rules and Regula:icas, and AIC policy. Operational guides are in the for= of =emoranda, written guides, preceden:, and verbal directives.
Incu= bent contributes to tha development of original standards, guidas, and codes in his field of endeavor.
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Direcco ste of Liccasing O
I Indeneendent Action Incu= bent is. respossible for preps:stion ci sad adheracca to review schedules and for =sking rece =andations on conventional engineeri=g matters for action to be c: ken by tha 3rsach Chief in regard to the acceptability of the hazards involved in specific reactors.
Incu= bent's jud scant, in =any.csses, is subject to only a general 3
review.
SUPCt7ISION
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WORRING CONDI"' Ibis.
- llor=mL~offica c.c :iliic G'Liy be 4:Eposed '62.nild, indih*Uca hb.[5 on fLa14 trips -
EFFORT lior: sal.
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' ENCLOSURE 4 e
fl 0:y PERFOR>uNCE APPRAISAL AND RECORD OI* INTERVIE: FOR NON-SL"dERVISORY PROFESSIONAL TECHNICAL E:'?LOYEE PROF.ILC NAME:
Robert D. Pollard CRADE/ STEP:
14/2 POSITION:
ProjectManager/
/
TIST IN GRADE:
19 months
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TI>'E IN STEP:
7 months APPRAISER.V DATE:
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, t. 'DeYjpung g N-Ai*-It TISE IN PREVIOUS CRADE:
12 conths l'.EVIEWER & DATE:
PREVIOUS APPRAISER & DATE:
D. B. Vassallo A C/NRC SERVICE (YEARS):
6-1/4 11/11/74 PROFESSIONAL EXPERIENCE (YEARS): 6
!!Jt! LONC SUPERVISED BY APPRAISER:
13 =enths DATE OF EIRTH:
2/13 /40 EDUCATION (DECREE & YEAR):
B.S. (Elec. Eng.)
1969 DISCUSSION TOPICS BACKCROUND & IXPERIENCE SUFCuRY - Bob received a B.S. in Electrical Engineerins from Syracuse University in 1969. After joining the AEC in 196?, he studied electrical and nuclear engineering at the Graduate School'of the University of New Mexico (1970-1971) in con 1 unction 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.
Af ter jo),ning the AEC in July 1969, Bob participated pri=arily 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 Cocsittees.
Bob transferred to RL as a project manager in September 1971..
KNOWLEDGE OF J03 - 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 tec!mical 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 syste=s and site related =atters), Bob is
- r
.. technically very perceptive. He has enough confidence to challenge reviewers on questionable technical matters and to pursue resolution of those in controversy..
In a very short time, Bob has developed an excellent understanding of the technical, manage =ent, and administrative aspects of project management.
He manages-to keep himself informed of current developments in technical, policy, legal, and general licensing matters.
MANAGE'IENT CAPA3ILITY - Because.cf.his. pas.t experience in TR, Bob had a good understanding of the LPM's role.
He has made a very rapid transition in assuming a proje'cc management philosophy.
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 of all aspects of his projects. He works very effectively with TR, OELD, and applicant representatives and. gets along very well with the branch secretaries and licensing assistant.
Bob has shown excellent capability to effectively organize and manage several concurrent projects. His principal assign =ent has been the OL review of the McGuire plant.
However, he was also assigned as the LPM for the completion of the Comanche Peak and Catawba CP reviews.
The latter required a considerable amount of LPM interaction with OELD because of the spplicant'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.,
issuante of a CP.
Because of the recent loss of an LPM from LWR l-1, 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.
l PROFESSIONALISM - Bob is an extremely conscientious and dependable project sanager. He conducts himself with a degree of =aturity and professionalism well beyond his age.
In his associations with applicant representatives, i
he is very fair, but firm, and can take a strong stance unen the occasion l
warrants it.
Bob does extremely well at planning and scheduling his workload. He is consistently able to complete assignments on schedule l
without needing reminders.
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l Bob does not require much supervision.
On the contrary, he seems to have a unique instinct of knowing the type of licensins action that a situation requires and then begins to take the appropriate 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 with the Catawba and Indian Point 3 projects). He is very persistent in trying to get stalled actions moving.
Bob does an excellent job of keeping his branch chief apprised of major review matters.
JUDCMENT - Bob is a careful thinker and uses goed logic in making judgments.
He has a very good understanding of the licensing program and uses good judgment consistent with regulatory objectives.
COSDIUNICATIONS Oral - Bob has very good oral co=munication 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 presentations 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 CHARACTERISTICS - Basically, Bob is a very serious minded but personable e=ployee.
He does not take rash decisions, but rather uses a more deliberative approach.
Bob manages to =aintain 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 implementstion of licensing policy.
In my opinion, this is because Bob has an exceptional understanding of the Cc==ission's rules and regulations and takes his role of regulator very seriously.
However, this has not affected his performance as a project canager.
AREAS NEEDING IMPROVEMENT - Since transferring fro: TR, Bob is becoming exposed to a nu=ber of review areas with which he did not previously have a greac deal of familiarity.
These are principally in the areas of site safety, effluent treat =ent, and so=e portions of auxiliary systems. He l
has =ade great strides in understanding what the =ajor review obj ectives are for these areas. With the continued exposure he is now obtaining in =anaging his projects, I do not forasee any problem in Bob becoming completely conversant in all review subjects.
I
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J PR0!!OTIO! POTE':TIAL - Bob has shown excellent project management capability.
He is well organi:ed, is able to keep his projects under control, and to mee: schedule =ilestones. On the basis of his previous experience in TR and with further experience in project =anagement, Bob has an excellen:
potential for attaining higher levels.
SISD!ARY - Although Bob has been in RL for about one year, he has de=enstrated excellent skills in managing safety reviews without requiring a great deal of supervision. Through his versatility, he has perforced extremely well in handling diverse assignments in a highly professional manner; e.g.,
taking on the canassment of compleii'c~ases 'such as Catauba and Indian Point 3 in the final stages of licensing effor:.
REVIETTER'S CO:t!E::TS issTof Th ;J$M..c.
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/, -/3 70 EMPLOYEE' S CO??E'rIS Acknowledgement I have read the above perfor=ance appraisal.
Date:
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ENCLOSURE 5 o7 D'3" a(
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Docket No. 50-344
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9 R.- C. DeYoung,. Assistant Director for Light Water Reactors Group 1, L 1
PORTLAND GENERAL ELECTRIC CCIPANI, TROJAN NUCLEAP. PLWI: SArr.u EVALUATION OF THE INSTRIMENTATION, CONTROL AND ELECIRIC PCRER SYSTEMS; DOCKET NO. 50-344 Plant.Name: Trojan.Nuelear P1 ant Docket Number: 50-344 Licensing Stage:
Operating License Responsible Branch and Project Manager:
LWR 1-1, '3. M. Cutchin Description of Response:
Safety Evaluation of the Instrtmentation, Control cnd Electric Power Systems Requested Cocrpletion Date: April 12, 1974 Applicant's Response Date Necessary for Completion of Next Action Planned on Project:
Not Applicable Review Status:, Complete except: for Technical Specifications and Supplemental Safety Evaluation Report
-f The enclosed report (Enclosure 1) was prepared by the L:R3, Electrical Instrumentation and Control Systems Branch for use in the Safety Evaluation Report for the Trojan Nuclear Plant. The report is based on a review of the F1 tal Safety Analysis Report through Amendment 12 and selected schematic diagrams contained in PCE-1001 and PCE-LOO 2, " Safety-Related Schematic Tiagrams". The principal reviewer was R. D. Pollard, L:EIE=CS.
~ is a list of topical reporbs ' that: 1) have been referenced in the Trojan FSAR, 2) co~ntain information necessary to support the issuance of an operating license for Trojan and 3) have either been found unacceptable as bases for a favorable staff evaluation or have not yet been reviewed by the staff.
Since topical reports are being reviewed cio a generic basis, Portland Ceneral Electric will either have to Jupply the equivalent information on the Trojan docket for review on an individual case basis or await generic resolution by Westinghouse. We recocmead that the latter approach 'oc taken because the reports not yet reviewed are scheduled for completian of their generic review prior to the decision date for Trojan, and there appears i
1
I j
j APR 1 0 1974 to be ample time for Westinghouse to supply the necessary infomation on those that have been found unacceptable.
In the course of our review it was noted that although the applicant states that PCE-1002, " Safety-Related Schematic Diagrass," contains proprietary schamatic diagrams for the Trojan Nuclear Plant, the topical report, WCAP-7671, Process Instrumentation for Westin house Nuclear Steam Supply Systems, (April 1971) is not proprietary and contains-figures illustrating essentially the same infomation. We therefore recocoend that you consider removal of the " Proprietary Information" status presently accorded PCE-1002.
Orl-inal Sl;:ng y,.
Victor Stenq j',
Victor Stello, Jr., Assistant Director Reactor Safety Directorate of Licensing
Enclosures:
1.
Safety Evaluation 2.
List of 9eferenced Topical Reports cc w/o enc 1:
W. Mcdonald, L: OPS cc w/ enc 1:
S. Hanauer, DRIA
.D..Vassallo, L: LWR-1-1 J. Hendrie, L:TR M. Cutchin L: LWR-1-1 l
A. Ciambusso, L:RP T. Ippolito, L:EI&CS l
S. Varga, L:RP F. Roes, L:EILCS D. Eisenhut, L:RP
- 1. Pollard, L:EIACS R. Klecker, L:RP J. Carter, L:RP E. Leins L. Reading File Branch Reading File L:EI&CS L:EI6CS L:EI6CS L:RS R. Pollard:me F. Rosa TN. Ippolito V. Stello
- 8's At 17 r.
A/
/74 4/ /74
ENCLOSURE 6
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UNITED STATES t' N
- i ATOMIC ENERGY COMMISSION I*j, O.)
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' July 21, 1971 3
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Mr. J. Forster
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f, Atomic Power Equip =ent Depcrt-ent
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ceneral Electric Company - M/CO37 175 Curtner Avenue San Jose, California 95125
Subject:
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 whct 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 it's, essentially impossible to determine compliance.
For these reasons the referenced document in its present form is, as I said above, without value.
Sincerely yours,
?l'
~
St phen H. Hanauer ec: Louis Costrell Sava I. Sherr 3#1
ENCLOSURE 2 Referenced Topical Reports Not Reviewed or Reviewed and Found Unacceptable 1.
WCAP-7821, Seismic Testing of. Electrical and Control Equipment (Iligh Seismic Plants),
December 1971 Status - Not reviewed - TAR scheduled for completion 6/15/74.
2.
WCAP-7744; Environmental Testing of Engineered Safety Features Related Equipment (NSSS - Standard Scope), 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.)
3.
'JCAP-7672,
' Solid State Logic Protection Sys tem Description, June 1971 Status - Found acceptable provided "the equipment is adequately qualified seismically and environmentally and the appropriate documentation has be,en provided."
(
Reference:
Letter to R.
Salvatori from D. B. Vassallo, dated March 6,1974.)
PGE has not-supplied informatien specifically applicable to the Trojan solid state logic protcstion system.
Perhaps WCAP 7821 and WCAP-7744 discussed above will be revised to include the solid state logic protection system.
4.
WCAP-7705, Engineered Safeguards Final Device or Activator Testing, March 1973 Status - Reviewed and found unacceptable (
Reference:
Letter to R. Salvatori from D. d. Vassallo, dated September 10, 1973.)
WCAP-7705, Revision,1) Testing of Engineered Safety Features Actuation Systems, February 1974, was recently submitted.
(
Reference:
Letter to D. V. Vassallo from R. Salvatori dated March 26, 1974.)-
5.
WCAP-7819 Revision 1, Test Report, Nuclear Instrumentation System Isolation Amplifier, January 1972 Status - Not reviewed. TAR scheduled for completion 8/5/74
(
Reference:
Status Report - Review of Vdstinghouse Topical Reports, D. B. Vassallo, February 20, 1974.)
l
' O 7.7 CONTROL SYSTEMS NOT REQUIRED FOR SAFETY We have reviewed the plant control system provided for Trojan and find that the design is similar to. those of other recently-licensed plants. We have concluded that the differences in the design details do not affect our previous conclusions that the design of the control systems is also acceptable for the Trojan Nuclear Plant.
7.8 SEISMIC, RADIATION AND ENVIRO: DENTAL QUALIFICATION We 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. We will report the results of the review and their applicability to Trojan in a supplement to this safety evaluation.
8.0 ELECTRIC POWER 8.1 GENEMAL 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 PO7ER Four 230 kV circuits are provided to carry the stat'.on electrical output to and supply offsite power from the transmission network.
There are tun 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 scieching station bus. With this arrange =ent 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.1. circuit.
The technical specifications will require that, as a minimum, one of; i
' the two codb'inations of two ciredits be ava'ilabic as a limiting
~ condition for operation of the plant. The other four combin.'tf as of two circuits do not meet the requirement of CDC-17 because the two circuits are either mounted on the same transmission towers or are connected to the same switching station bus.
. n 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 qualification 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 nucicar generating station.
The applicant has not yet submitted the results of the qualification test program for the Trojan diesel gener-ators. We will report the results of our evaluation of the adequacy of the diesel generator sets in a suppicment to this safety evaluation af ter we have received and reviewed the results of the test program.
8.3.2 SEISMIC QUALIFICATION OF ESF SWITCHGEAR During seismic testing of the protective relays associated with the ESF brerkers, 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 seismic occurrence si'nce the relcys might trip as if performing a protective function. Af ter 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.
We informed the applicant of 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 ppen. We also understand that the applicant is considering automatically blocking the tripping function of other relays that could misoperate and thereby.
cause tripping of the diesel generator gircuit breakers.
I 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 wi1T be 'repo'rted in a' supplement to,this safety evaluaYi'on."
i
l ENCLOSURE 7 "I
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6 RAfiCUnt FOR: Chaiman Hendrie Commissioner Gilinsky Cossrissier.er Kennedy Consissioner Bradford gg THRU:
MxecEtive Director for Operations FROM:
Edson G. Case, Acting Director, 3RR
SUBJECT:
UNICit OF COMCEPJiED SCIENTISTS' PETITIC+t TCR RECONSIDERATION DATED MY 2,1978 provides answers to the five asterisked itens identified in the Secretary's memorandum of June 21,1978, for which staff resocnse was recuested by July 5,1978. A response to the cther items will be previded by August 25, 1978.
S/
Edson G. Case, Acting Director Cffice of Nuclear Reactor Regulaticn Attacheent:
As stated c::: Secretary UCS PDR
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'#'), :\\,&,qj' id; ENCLOSURE 1 (d
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PARTIAL RESPONSES TO THE SECRETARY'S JUNE 21, 1978 *^
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NEMORANDUM CONCERMING THE UNION OF CONCERNED SCIENTISTS'Y
,~
PETITION FOR RECONSIDERATION egj
- 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)
Resconse 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 such safety systems, but would not necessarily prevent operator action frem perfortning the safety function.
Every nuclear power plant has seme flammable materials that are either fixed or transient; therefore, at least a limited potential for fires in nuclear power plants exists; hcwever, 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.
l l
l
, objective has been to reduce the severity and lower the probability of fires.
Starting imediately 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 ootential 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 (e.g., fire detection and trained fire brigades that are prepared for the prompt use of water to extinguish cable fires). This has been an important element of the basis for continued operation of these plants pending ful;l upgrading to conform to BTp ASB 9.5-1_/ and 2
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 I
in the event of a fire.
U ee pages 15-23, 33-38, and 71-72, staff report dated December 15, 1978.
S
.,The results to date of the ongoing staff evaluatio. of fire protection E
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 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 alterncte 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-t 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 ccmplete to indicate the need for an alternative method of shutdown are identified in Table 1.
U ee pages 23-25, staff report dated December 15, 1977.
S
, As shown in the table:
11 plants have been shown t'o have adequate shutdown capability and do npt 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 detennination of the need for additional shutdown capability has not been completed. For the plants not yet evaluated, we wuuld 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 sevare damaging fires is acceptably low for the interim period until staff evaluations and licensee enhan' cements are coinpleted. This conclusion is based upon the information discussed by the Browns Ferry Fire Sp' cial Review Group e
in NUREG-0500 and upon the additional defense-in-depth protection provided by the staff's overall fire protection upgrading program which provides (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.
l l
~
Table 1 Status of Review Findings for Operating Plants Plants Shown to Date to Have Adecuate Shutdown Capability (1)
Arkansas 1 i
Fort Calhoun Kewaunee Maine Yankee Oyster Creek Turkey Point 3 & 4 Vermont Yankee Browns Ferry 1, 2 & 3 Plants Which Have Installed Alternate or Dedicated Shutdown System D. C. Cook 1 & 2 Hatch 1 & 2 Ft. St. Vrain Plants Recuiring an Alternate or Dedicated Shutdewn System Brunswick 1 & 2 (To be installed by January 1979)
Haddam Neck (To be coordinated with SEP schedule)
Oconee 1-3 (To be installed by October 1980)
-l Pilgrim (2)
Rancho Seco (To be installed by December 1979) i 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 Octdber 1980.
ENCLOSURE 3 D
R. E. Tiller, Direc::r Reactor 0:crations a Progrees Division Idaho Op: rations Office - COE Idaho Falls, Idaho 83401 PWR FUEL ASSEMoLY'FECHAli! CAL RESPO!!SE NIALYSIS - Stig-316-77
.Ref:
(a)
R. L. Grubb, P'..R Fuel Assembly !4cchanical itespense Analysis, Idaho flational Encinecring Lctoratory, RE-E-77-141, furch 1977 (b)
R. L. Grubb, PUR Fuci Asse=cly Mcchanical F.cspense Analysis, Arendment No.1, Idahe ft:tional Engineerin;: Laborctory, RE-E-77-140,l!:rch. 1977 i
(c)
R. L. Grabb and 8. F. Caffell, Jr. flon-Linear Lateral 14echanical P.uspense of Pressurized Hater P.cactnr Fuci Asse:clies, AS?:E Pacer 77-UA/0E-13, Decc:ber 1977 (d)
H. !!uno, !!. Itizuta, and !!. Tsu :una, Develo: cent of Advanced Hethe For F;:1 Seisnic Analysis, 4th International Conference en Structural !*cchanics in aaactor Technology, San Francisco, Califo nia, USA, August, 1977 (e)
R. L. Grubb, Feasability Study for Dounding the Lateral ?9P.
Fuel Assembly ficchanical Res:ense Analysis, Idaho fiational Engineering Laboratocy, RE-E-77-160, Rev.1, yuly,1977 l
Dear !!r. Tiller:
A pararetric study to assess the effect of variations in cere plate cotions on fuel asscnoly spacer grid crushing loads is currently in progress.
A su=ary descriptien of tnis study inciuding arelinincry results ha: been prepared at the rceuest of the 4.: l' car Rer;ulatory Ccerission's Civisien j
of Systen Safety, C:rc Perfer-nce Brancn.
Results of this study indicate that a small variati0n in c:re plate frecuency rey nave a sir;nificant effect en spacer grid crushing loads. As the study is not cxplete, these results should be considered preliminary.
A mechanisin t. s ; :tulated in Reference (c) which indicated that the in-put core plate rotica c:uld significantly affect spacer crid crushinn
- loads.
The princry cbjective of the present study was to deternine if this rechanisin c:uld 50 shnun to exist. A seccndary nb. ice:ive is to e x pare linear end non1inear analysis techniques.
In sumary then the purpose of this study is sofold:
(1)
Statistically dotarmine the effe: of c:r ;ia:c frecuency and ragnitude en the fuel assc=cly maxi.mun s:acer grid crushing leads, and (E)
Statistically :::;are lincar and ncnlinecr analysis mthods for lateral fuct ass:ntly cctanic:1 respenso in an at c:pt to simplify the n:nitrear analysis.
b ig Page 2 The structural rnodel utilized to analy:e the fuel assemoly :echanical response is basically described in References (a) through (c). Tsio exceptions included in the present study are the use of fuel asse.mbly experimental frequencies and mode shapes and utilization of the nicthod presented in Reference (d) for calculation of spacer grid crushing
~
loadt.
The ncminal forcing function, core plate acceleratiens, are presented in Reference (h). !lhile eight variaticas on the frequency and ar.:plitudo of core plate n:ctions are to be considered, only the fcur extren:c cases are addecssed in this discussicn. The fnur cases are 10:; variation en frequency and + 10'. variatien of the arplitude.
It 6
Is noted thct all the #recuencies c:nt:1ned in the c:re riate cotinns are varied the same an: cunt. ficalicean dync.nic analysis es described in References (a) througn (d) is in cregrc:s and preliminary results are provided in Tasle 1.
- 3. linear analysis is aisc being pursued using
,the methods cutiined in Reference (e).
TABLE l' RATIO OF PEAK SPACER GRIO CRUSHP:G LOAD TO'THE I
N0lilNAL CUR 5 hit!G LOAD Haxi;::um Crushing Load /iicninal Crushing Load'
~
l Spacer Grid Elevation
-1C", Frecuency.
+10 Frecuegy,
-10 Am::li tude ^10". Amoli tude 0.804 0.979 1.00 Center 1.76
. Center-up i.45 0.Saa 0.842 1.23 1.25 Center-dcs.n 2.11 0.830 0.335 Top 1.34 0.945 0.771 1.24 Bottom 1.56 0.805 0.363 1.26 l
l 1ficminal crusliing icad is the pec'< scacer grid crushinn lead obtained frca the base case core plate.c:icns.
. Based on the result: in Ta::le 1, it coes ancear tha a variation in frequency of tan ;creen effacts a st;nificant cnange in 2e scccer grio crushing icac:.
This ir.ct:stes that a varinien in this para-eter c:ay be in or:er fnr :his t,
c of ncnlinscr cnc'. sis.
10 shculd te ;ointed cut the Fe :ccel s udie. r: -
recents a g:ne-al : r'ipration.
The :urpose Of :his st;;dy seas not a dirc:-
analysis of a scecifi.: plant tut Oc cetermir.e if trc mechanisim ;cstula*.o
- 80'.'E. Till er
[9 d(U R
0
- 4 lii/
Stig-316-77
\\ QJ)
Page 3 Reference (e) could actually be clicited in the nonitccar analysis. The =cch-anisin a;1 pears to cr.ist; thereby causing concern that scr anent defornatica of spacer grids may cccur.
Upon cocpletion of this study the conclusions present:d in Ccforence (b) will be reassessed.
yery truly y'ours, c::tc!NAt s:cN:o 37 R. R. Stiger, !!anager P.cactor Geha:itor 0,1 vision SFS:cij V. Stclic, fiRC-CGR P. S. Check, fiRC-OSS S. B.- Xia,iiRC-DSS R. J. Itattsen, :CC-CSS R. O. l'.cycr, llRC-OSS D. F. Ross, !!RC-OSS R. 11. r.ichn, EG4G Idaho bcc:
R. L. Grubb R.11. ifacek C. A. Peore v
C.F.Obenchain(Og.,'
B. F. Sarrci1 G. L. Thinnes ;, 'O T. R. Thc.mpscn P. H. Vander Hyde L. J. Ybarrondo Central File File
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UPtTGD STATES j
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NUCLEAR HEGULATOnY cOMMisslON ENCLOSURE 9
.C.
O W ASHINGTON. D. C. 20555 y ';
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August 18, 1977 C
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%[% [N MEMORANDUM FOR:
E. G. Case, Acting Director I
7 Office of Nuclear Reactor Regulation 9
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FROM:
tephe.' H. Hanauer, Technical Advisor to g
t.: acutive Director for Operations
/
Co. ito\\
i
SUBJECT:
INTERACTION BETWEEN CONTROL SYSTEM AND PROTECTION SYSTEM The Zion incident of July 12, 1977, apparently shows a design defect as well as the obvious gross management deficiency.
The 31 durany signals disabled the primary system level control, which initiated a transient involving decreasing levs!.
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 ible 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 rgiling.
What is needed is adequate independence of control functions from safety functions that provide against control malfunctions.
D
]D ]O T 9
o w m o a 11
=
'E. G. Case 2
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
' c'onsiders it advantageous to use certain infonnation 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:
t
% _.1 - (.[. x. v.D.'
N. I
/ ) Stephen H. Hanauer
.:. ; 3n r.. a. s.. -
Technical Advisor to Executive Director for Operations cc:
L. V. Gossick
- 5. Levine E. Volgenau R. Minogue r/
ENCLOSURE 10
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m 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 FROM @ IS FLOATING AROUND THE INDUSTRY AND DESERVES YOUR RAISING IT IN SOME PROCEEDING.
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OBVIOUSLY KEEP ME OUT OF IT.
(Initials Deleted]"
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ENCLOSURE 11 C1 is g
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Deesaber 2,1976 Dear 16ny tiranks for your recent nets and the enclosure.
I'll try to put it to good use.
I no longer hava an office so the liashingten address you have is out W ad.
I now werk out of my her.s which is:
h eersly, d
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ENCLOSURE 12
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January lb,1977
Dear I 1 sassed the Westin*,
house documant on fuel handling accisients inside containment to sene interve.aers and sent a copy to the f,'RC.
A cony of ny letter to Gilinsky and the resly from Strauss ars enciesed.
8s you can ces they would like nors infomation. If have any other info you want me to forward to NRC, send it along.
Contrary to Strauss' suggestion, there are r.o circumstances where I would dicciose ths soureo.
Sincersly,
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January 3, 1977 Cor=issiener Victor Gilincky U.S. IMelcar Regulatory Ccmission WashinC on, D.C. 2C555 t
Dear Comissioner:
We receivnd the enclosed docunent fron an individual who wi ' a to re.ain anonyraous. We are sending it to you ir. the hope that the l'uclear Regulatory Cer:r.issien will take prin:t action to protect the health and safoty of the public from the known risks discussed in the docut ent.
The docunent correctly indicates that the consequences of a fuel handling accident incide the reactor containnent building are not considered by the 2C in deciding whether a nuclear pcuer plant should r-ceive a license. In addition, the doce=ent indicates that Vestinghouse believes that a fuel handling accident inside contain-nont could result in radiation doses to the public in exessa of 10 CFP. Part 1C0 limits, i.e., in excess of 25 rer. to the w.~. ole b:dy and 300 rem to the thyroid.
In view of these statenents, it appears that a fuel handling accident inside containnent is an "unreviewed safety question" and a "significant safety hazard."
We recornend that the GC review the design and procedures of each enerating nuclear power nlant to detemine whether a fuel handling accident inside contain-nont will result in doses that "are well within the guideline values of 10 Cy3 Part 1CO," as soecified in Section 13.7.h of the Standard Revinw Flan. Until such reviews are cor.nloted, no believe that the CC should issue orders to halt all refueling enerations in nrogress and to urohibit 221 future refueling opera-tiens. In addition, we believe that it is 27propriate for the :iRC to initiate investigations to determine whether Section 206 of the Energy Reorganization Act of 197h has been violated by individual directors or responsible efficers of Westinghouse and other firms which received the enclosed document.
We would apsreciata hearing fre. you prcmstly regarding the action that :iRC will i
take to resolve this matter. We also weuld like an explanation of t he reasons for HRC not previeucly requiring analysis of a fuel handling accident inside containment and the stees that will be taken to correct this deficiency in the licencing process.
By cecy of this letter, na are also sending the enclosed docunent to the chaimen of the Advisory Committes on Reactor Safegur.rds, the Atemic Safety and Licens12C Board Panel and the Atenic Safety and iicensing Appeal Panel.
Sincerely, k
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a.bert a. P.llare 12c8 Massachusotts Aonue. Cameridge. Massschusetts C2103 Tetephcne ($17) 547-5552
As you are aware, a, fuel handling accident in the spent fuel storage building is analy/cd in plant Safety Analysis Reports.
The assumpti'ons utili cd for this analysis are outlined in Hegulatory Guide 1.25. " Assumptions Used for Evaluating the Potential Consequences of a Fuel llandling Accident in the Fuct llandling and Storage f acility."
The off-site consequences of this accident are compared to 10CFR100 limits of 300 rem to the thyroid and 25 rem whole body dose in the Safety Analysis Reports.
In addition, the i;2C compares the resultant doses with unofficial limits of 30 rem to the thyroid and 5 rem whole 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.
1[ is not aware of the f.RC tases for not addressing a fuel handling
- accident inside con-tainment, the bases may include:
1.
The assumption that the containment will be isolated during refueling opera tions ;
2.
that the containment could be isolated quickly enough to limit off-site consequences; or 3.
that filtration capability comparable to that in fuel storage building exhausts exists in the containment purge exhaust.
These bases are similar to the bases used to address the fuel handling accident in the fuct hundling building.
Information available to us, including resul ts of scopin.g analyses using very conservative assunstions 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 handling accident inside containment if one assumes no credit for containment isolation, iodine filtration, or mixing within containment.
In addition to using Regulatory Guide 1.25 assumotions in' the scoping analyses, we assuned operaticn of systems which would result in the most conservative dose.
For e.ta51e. it sas a*.su: ed that a push-pull tycc or exhaust only sucep ventila-tion sys cm is in operation over the refueling canal so that activity releases are routed inmediately to the purge exhaust.
Much of the infor: cation required to do an evaluation for specific plants is not available to us.
We do recomend, nowever, that you evaluate the consequences of this potential incident to assure chat unacceptacle doses are not a probable result.
Since the f;RC regulations do not require the analysis, v.e do not believe this sitt.ation requires reporting to the tiRC unicss your engineering ovaluation shows unaccentable results.
In accomplishion tr e evalua tion for your plant, we reccevend t".n you use Regulatory Guide 1.25 assumptions or oth:: con w rva ti ve,
jus ti fi atic pa raua ters.
We also believe that you shuuld not t.O.e credit for the function of ar.y system or component that is not qualified for operation during this particular incident.
For example, we think you might take credit. for equip-ment not qualified for the post accident containment environment but seismic qualification may very well be required.
Please feel free to contact us if further information or assistance is required.
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UNITED STATES 4
NUCLEAR REGULATORY COMMISSION yo 7
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Mr. Robert D. Pollard 4
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Dear Mr. Pollard:
Comissioner 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 for.varded to Mr. Lee Gossick, the Comission's Executive Director of Operations, with a request th.at he provide it to the proper Staff offices, and assure a timely, direct and appropriate response to your statements of concern and to the issues raised by the partial document you have supplied.
The Comission 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 Comission 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 cencerns you have raised and the documentary fragment you have provided.
Sincerely.
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Peter L. Strauss General Counsel
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