ML19257C379
| ML19257C379 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 01/24/1980 |
| From: | Hoefling R, Lewis S NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | Atomic Safety and Licensing Board Panel |
| Shared Package | |
| ML19257C377 | List: |
| References | |
| NUDOCS 8001290006 | |
| Download: ML19257C379 (84) | |
Text
1/24/80 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
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SACRAMENT 0 MUNICIPAL UTILITY
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Docket No. 50-312 (SP)
DISTRICT
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(Rancho Seco Nuclear Genercting
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Station)
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NRC STAFF RESPONSE TO F0E DISCOVERY REQUESTS AND MOTION FOR A PROTECTIVE ORDER The NRC Staff Response to the recent discovery requests 1/ of Friends of the Earth (F66) is contained herein.
With respect to F0E's Request for the Production of Documents, a number of documents are being provided as specified in the individual Staff Interro-gatory Responses.
One document, NUREG-0660 enticled " Action Plans for Implementing Reconmendations of the President's Commission and other Studies of the TMI-2 Accident," is referred to frequently and copies of this document are herewith served on all of the parties.
Certain ANSI /ANS Standards referenced in the Staff Interrogatory Responses may not be reproduced without permission of the publisher.
Copies may be obtained from:
American Nuclear Society 555 North Kensington Avenue La Grange Park, Illinois 60525 l0 g-
~~1/ The Staff is responding to both F0E's First Request for the Production of Documents and F0E's First Set of Interrogatories directed to the NRC Staff.
Affidavits of the Staff members who prepared the Rosponses follow the Responses.
8001290 9
Also, the Licensee's procedures either are not in the Staff's possession or, where the Staff does have copies, may not be the latest version. To the extent F0E seeks to examine such procedures, the Staff would suggest that F0E contact the Licensee.
With respect to '0E's interrogatories, these interrogatories were filed by F0E pursuant to 10 CFR 52.720(h)(ii) which permits the filing of such Interrogatories with the presiding officer of a proceeding to permit a deter-mination by the presiding officer that the answers to the intarrogatories from the NRC Staff are necessary to a proper decision in the proceeding and that answers to the interrogatories are not reasonably obtainable from any other source.
Without awaiting a Board finding relative to the F0E interrogatories, the NRC Staff is responding to many of the interrogatories.
With regard to certain of F0E's interrogatories, however, the Staff has objected.
The rules of discovery in NRC proceedings applicable to all parties are set forth in 10 CFR 52.740.
This section allows discovery regarding any matter not privileged as long as it is relevant to the subject matter involved in the proceeding or wil' lead to the discovery of relevant evidence. The rules of discovery add additional requirements when discovery is directed against the NRC Staff.
These requirements are set forth in 10 CFR 52.720. With regard to interrogatories directed to the Staff, 52.720(h)(2)(ii) requires a finding by the presiding officer that:
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... answers to the interrogatories are necessary to a proper decision in the proceeding and answers to interrogatories are not reasonably obtainable from any other source...
Additionally, pursuant to 10 CFR 52.740(c):
Upon motion by a party or the person from whom discovery is sought, and for good cause shown, the presiding officer may make any order which justice requires to protect a party or person from annoyance, embarrassment, oppression or undue burden or expense...
It is in this context that NRC Staff objections to F0E's interrogatories must be viewed.
The Staff responses and objections are presented below following the applicable F0E interrogatory.
In some instances, a partial objection and response are presented together.
Staff Request for Findings Under 10 CFR 12.720(h)(2)(ii) and Staff Request for a Protective Order For the reasons set forth in its objections, the Staff requests that the Licensing Board find, under 10 CFR 52.720(h)(2)(ii), that the responses to certain interrogatories of F0E, or portions thereof, specifically Interrogatories 34, 37 through 47, 52 and 54 are not necessary to a proper decision in this pro-ceeding and/or are reasonably obtainable from other sources. The Staff further requests that the Board issue a protective order, under 552.720(h)(2)(iv) and
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e 188) 100 2.740(c), that, as to the interrogatories objected to by the Staff, discovery either not be had or be had only ca the terms and conditions (including a designation of the place) suggested by the Staff.
Respectf y submitted, ick c-R chard K. Hoefling I
C Cou g l for NRC Staff
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ephen H. Lewis (3
Counsel for NRC Staff Dated at Betnesda, Maryland this 24th day of January, 1930.
9 18$101
Interrogatory 1 Provide responses to the questions and requests for information contained in the documents "First Set of CEC Interrogatories to the NRC Staff," dated November 15, 1979.
Response
A copy of the "NRC Staff Response to California Energy Commission's First Set of Interrogatories to the Nuclear Regulatory Commission," dated December 11, 1979, was served upon counsel for F0E.
Interrogatory 2 Following the substantive response to each of the follow-ing interrogatories, identify by name and affiliation each individual who has knowledge which served as the basis for that interrogatory.
Response
Fbliowingeachresponse,thepersonwhohasknowledgewhichservedasthe basis for the response is identified.
Interrogatory 3 Following the substantive response to each of the subsequent interrogatories posed by F.0.E., identify all documents and studies relied upon by NRC Staff in providing the answers to that interrogatory. The identification should be specific to the portion of the document or study relied upon. Studies s' hall include observations, calculations, literature and other types of work, whether recorded or not, which consist of an examination or analysis of a phenomenon.
Response
Every document directly relied upon by the preparer of a response is identified in that response.
If a specific portion (s) of the document.was relied upon, that portion (s) is identified.
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Interrogatory 4 Describe in detail the current process, procedures, standards or other criteria and evaluations that NRC uses to certify the competency and fitness of licensees.
In what manner have these been applied to management personnel at SMUD.
Response
Commission regulations (10 CFR 50, Section 50.34) require an applicant to submit information concerning organizational structure, personnel requirements, and technical qualifications to engage in the proposed activities. The NRC is required (10 CFR 50, Section 50.40) to make a finding of technical competence of the applicant before issuing a license. At the present time, an applicant mest submit the pertinent information in its Safety Analysis Reports in accordance with Regulatory Guide 1.70, Section 13.1, " Organizational Structure of Applicant," and Section 17, "Cuality Assurance." A description of the NRC review procedures and general acceptance criteric are strted in the Standard Review Plan, Sections 13.1.1, 13.1. 2, a nd 13.1. 3.
The appli-cant's submittals are reviewed by the NRC staff against the Standard Review Plan.
If any deficiencies are notEd, the applicant is requested to supplement its submittals by responding to specific questions or comments. When the applicant has provided acceptable provisions for the implementation of the requirements, documentation of acceptability is provided in the staff safety evaluation report issued in conjunction with a construction permit or operating license. After acceptability is established and a construction permit or operating license is granted, contact is maintained by members of the Cormiission's Office of Nuclear Reactor Regulation involved in continuing licensing activities 183d 103 (principally the respective Project Manager) and by the Office of Inspection and Enforcement (I&E) inspectors. As a result of this contact, or as a result of a specific event at a facility, any significant deficiencies in a utility's management and technical support capabilities would likely be dis-covered.
If this were to occur, I&E would have the lead responsibility of initiating action on the part of the utility to resolve the deficiency.
With respect to SMUD, the Rancho Seco Final Safety Analysis Report (FSAR) was submitted on May 1, 1971, which described the SMUD organization and management responsibilities and quality assurance program for Rancho Seco. This information was submitted in accordance with requirements and guidelines in existence at the time.
Questions were forwarded to SMUD in this area by the Atomic Energy Commission Regulatory Staff (the predecessor to the NRC staff), and responses are documented in Appendix 12 of the FSAR. The SMUD organization, personnel qualifications, and quality assurance program were found acceptable to the Regulatory Staff as documented in a Safety Evcluation dated June 8,'i973, which was issued in support of the operating license granted for Rancho Seco.
Subsequent to the issuance of the operating license for Rancho Seco, SMUD submitted a proposed amendment to its operating license that reflected a reorganization of its personnel and assignment of responsibilities (SMUD letter dated February 21,1978). This new organizational structure was again reviewed to ensure that NRC requirements and guidelines were being met.
The NRC Staff found the reorganization acceptable and documented this acceptability in a safety evaluation forwarded with a license amendment in a letter dated November 14, 1978.
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1830 104 To date, no significant deficiencies in SMUD's capability to operate the Rancho Seco facility have been noted. However, as a result of the accident at Three Mile Island, the NRC has initiated efforts to upgrade and more clearly define acceptance criteria in the area of management and technical support capability. These efforts are described in The Action Plan, NUREG-0660 (DRAFT) dated December 10, 1979, as Task I.B.l.
Although the schedule for Task I.B.1 states that acceptance criteria will be developed by January 1,1980, the criteria are still under development. When fully developed and applied to operating reactor licensees, these criteria could result in an upgrading of SMUD with respect to its management and technical support capabilities.
In connection with this effort, operating reactor licensees were requested by letter dated June 29, 1979, to submit information regarding utility capabilities to respond to events such as the TMI-2 accident.
SMUD's response, dated July 30, 1979, is attached.
This response was prepared by Daniel Garner.
iB2 105 Interrogatory 5 Describe in detail the current process, procedures, standards, or other criteria and evaluations that NRC uses to determine or judge that the competency of a licensee is deficient.
In what manner have these been applied to management personnel at SflUD.
R_esponse The response to this interrogatory is covered in the response to Interrogatory 4.
This response was prepared by Daniel Garner.
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_I_nterrogatory 6 Describe the reviet mechanism required by NRC of utility training programs for all operational personnel, iNlud-ing maintenance and technical personnel.
Response
Utility training programs are submitted by Applicants as Section 13.2 of the PSAR at the Construction Permit (CP) stage of licensing and Section 13.2 of the FSAR at the Operating License (OL) stage. These programs are rev ewed by Operator Licensing Branch Examiners in accordance with Section 13.2 of NUREG-75/
087, " Standard Review Plan for the Review of Safety Analyses Reports for Nuclear Power Plants."
(Enclosure).
This response was prepared by Bruce Wilson.
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'l83f107 Interrogatory 7 Regarding Interrogatory 6, describe the process and basis by which such training programs are judged to provide sufficient assurances that safety-related functions will be effectively carried out. Define explicitly how suffi-cient assurances are determined.
Response
See response to Interrogatory 6.
The process and basis for judgement of the training programs is contained in Section II, " Acceptance Criteria," of Section 13.2 of NUREG-75/087.
This response was prepared by Bruce Wilson.
183f108 Interrogatory 8 Describe the extent and nature of emergency duties included in NRC training program reviews, and their applicability to Rancho Seco.
Response
The term " emergency duties" is unclear.
If it means " emergency procedures,"
the answer is that the NRC reviews licensee's training programs to determine that the emergency procedures listed in Appendix A to Regulatory Guide 1.33 (Revision 2, February 1978) are included.
If it means " duties" under a licensee's " Emergency Plan" (involving both onsite and offsite actions), the answer is that it is reviewed as part of Section 13.2 of the FSAR under the heading of general employee training. All of these reviews have been con-ducted for Rancho Seco.
TdisresponsewaspreparedbyBruceWilson.
i81/109 Interrogatory 9 Describe all procedures and documents related to fJRC of in-plant drills.
Include all criteria used to assess acceptability and success of such drills as they specifically relate to Rancho Seco.
Response
The f4RC does not presently review in-plant drills aside from fire and evacuation drills. Each operator and senior operator must participate in an oral examination with the Pancho Seco Plant Superintendent or his designated representative as one of the requirements of Requalification Program.
This oral examination includes a simulation of abnormal and emergency conditions while in the control room.
This response was prepared by Bruce Wilson.
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i Interrogatory 10 Describe all efforts to date, on-going and in the future, to implement SECY-330E. Describe in detail impediments that may exist to implementation at the Rancho Seco facility.
Response
A draf t of NUREG-0660, " Action Plans for Implementing Recomendations of the President's Commission and Other Studies of the TMI-2 Accident," has been completed. The requirements of SECY-330E (enclosure) are included in this draft Action Plan. Once the final plan is approved by the Comission, the requirements of SECY-330E will begin to be implemented. The initial imple-mentation date is expected to be March 1, 1980.
At the present time, we foresee no impediments that may exist to implementation at the Rancho Seco facility. The only impediment is due to manpower resource limitations at the NRC.
It is anticipated that it will be at least two years before the NRC will have sufficient manpower to administer all simulator certi-fication and requalification examinations.
It is assumed that Rancho Seco per-sonnel will continue to use the B&W Simulator in Lynchville, Va., which is representative of their control room design.
This response was prepared by Bruce Wilson.
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Interrogatory 11 Describe all documents, processes, and procedures used by NRC with regard to the renewal of operator licenses.
When have these been applied at Rancho Seco?
Response
The requirements for renewal of operator licenses are contained in 10 C.F.R. 555.33 and NUREG-00is4 (enclosure). The application for renewal is reviewed by the Chief, Operator Licensing Branch and the certificate of medical examina-tion is reviewed by a consultant M.D.
If the information required by 10 C.F.R. 550.33 is satisfactory, a renewal of the license is issued. Renewals of Rancho Seco operator licenses were issued in June and December 1976, February 1977, June, August arid December 1978, and February and July 1979.
This response was prepared by Bruce Wilson.
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181 112 Interrogatory 12 Describe all processes and procedures used by NRC to identify individuals committing operational errors identified in licensee event reports.
Identify all such individuals identified in Rancho Seco L.E.R.'s in the past three years.
Response
NRC does not currently require identification of individuals committing operational errors leading to licensee event reports (LERs). No individuals have been identified to the NRC in connection with Rancho Seco LERs.
Although the NRC does not currently require the identification of individuals committing operational errors in LERs, the Lessons Learned Task Force has recommended that the NRC establish such an approach.
(See NUREG-05S5 Recommendation 1.4(2), p. A-5).
This response was prepared by Daniel Garner.
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183 111 Interrogatory 13 Describe how such processes and procedures, referred to in Interrogatory 12, will or do affect the quality of reports received by NRC.
If NRC believes that the quality of the L.E.R.'s is unaffected, please explain why this is so.
Response
No f;RC position has been established with respect to whether or not quality of LER's would be affected by this procedure.
This response was prepared by Daniel Garner.
183[l14 Interrogatory 14 Describe all documents, processes, and procedures used by NRC to determine the level of the individual shift super-vision or shift technical advisor's technical knowledge in the area of transient and accident response and, in the case of a shift supervisor, the managerial ability to command and control the activities of shift personnel.
How have these determinations been made with regard to Rancho Seco?
Response
ine present process and procedures for determining the level of knowledge of Shift Supervisors are contained in 10 C.F.R. Part 55 and NUREG-0094; "NRC Licensing Guide." There are presently no such guidelines for Shift Technical Advisors. However, these are being developed.
I & E will develop prncedures for auditing the process used by licensees in select'ing and certifying shift supervisors and shift technical advisors. I & E is scheduled to begin this audit by January 1, 1981.
This response was prepared by Bruce Wilson.
Interrogatory 15 Regarding Interrogatory 14, explicitly describe the criteria used by NRC to determine the level of accept-able performance of individuals.
Response
The present criteria for Shift Supervisors is the same as for Senior Reactor Operators and is well documented in all the material pertaining to licensed personnel, particularly by 10 C.F.R. Part 55, fiUREG-0094, AriSI 18.1-1971, and Standard Review Plan Section 13.2.
The new criteria for Shift Supervisors and Shift Technical Advisors has not yet been developed.
This response was prepared by Bruce Wilson.
}fb Interrogatory 16 Describe all documents, processes and procedures currently utilized by fiRC to incorporate, collect, disseminate, or otherwise learn from the operating experience of utility shift personnel.
Response
Licensee requirements for reporting information to the fiRC concerning operational data, operating events, and other reportable information are contained in the Commission's regulations, as outlined in Regulatory Guide 10.1, and in the applicable portions of a licensee's Technical Specifications covering reporting requirements. Regulatory Guide 1.16 provides guidelines on the format of reports. Other methods of collecting data include I & E Bulletins and letters issued by the f:RC.
f;RC procedures for dissemination and use of information obtained from licensees are outlined in " Program Plan -
Use of Inspection Data in the Licensing i'rocess" (SECY-77-229) dated May 4, 1977, and a memorandum from Harold R. Denton to fiRR Division Directors dated flovember 7,1979.
This response was prepared by Daniel Garner.
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Interrogatory 17 Describe all processes and procedures f6C requires and deems acceptable for requalification of utility shift personnel and instructors. When have these been applied with regard to Rancho Seco?
Response
Requalification programs are reviewed by NRC on a case-by-case basis. There are no offically adopted standards for acceptability. Rancho Seco procedure, AP-25, " Licensed NRC Operator Retraining" addresses tne requalification of licensed operators and senior operators. The original program was approved in June 1974 by an OLB examiner. A revised program was approved in June 1978.
I & E is responsible for assuring that all licensed individuals comply with the requirements of the Requalification Program. The Operator Licensing Branch is responsible.or assuring adequate quality of the written examinations and grading.
This response was prepared by Bruce Wilson.
Interrogatory 18 Describe all processes and procedures NRC requires and deems acceptable to determine the quality, efficacy, and relevancy of simulator training programs.
In particular, describe the basis for determining the acceptability of B&W simulation training, if such a determination has been made.
Response
Applicants for " cold" examinations must have extensive experience at a reactor facility which is generally classified as comparable in complexity and operating characteristics to the nuclear plant at which tr.c n aminations are requested.
One method by which this experience may be acquired is by certification of satisfactory completion of an AEC (NRC)-approved training program which uti-lizes a complete and accurate nuclear power plant simulator as part of the program. This recommendation is contained in ANSI-18.1 - 1971, " Standard for Selection and Training of Personnel for Nuclear Power Plants." Applicants for Operating Licenses (0Ls) are required to commit to meeting the require-ments of this standard in Section 13.2 of their Safety Analysis Report or to provide acceptable alternatives.
This information is also contained in NUREG-0094, "NRC Operator Licensing Guide."
These vendor / utility training programs must be submitted to the NRC for review and approval. The criteria for acceptability of the simulator training programs are contained in an internal OLB guide. Generally, the criteria include Phase I - Normally 12 weeks of basic fundamentals including 10 reactor start-ups on a research reactor. Phase iia - Observation training on an operable nuclear power plant - normally 2 months. Phase iib - Simulator training -
normally 2 months. Phase III - Design Lecture Series - norm ly 6 weeks.
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In addition, all cold applicants must successfully complete an approved on-site training program which is a minimum of 6 months long.
The quality, efficacy and relevancy of the simulator training programs are determined to be acceptable on a case-by-case basis by examiners in the Opera-tor Licensing Branch. The scope and fidelity of the simulator is evaluated on the basis of meeting its specifications and on the basis of the enaineering judgment of the NRC examiners and the utilities or reactor vendors purchasing the simulator.
In the future, this evaluation will be based on more explicit requirements as set forth in ANSI /ANS-3.5.
Further discussion of this stan-dard is included later in the response.
Comission accepted recomendations for changes to simulator training programs are contained in SECY-330E and SECY-330F.
In the case of B&W training programs, discussions were initiated between the AEC's Division of Reactor Lice sing (DRL) and B&W as early as June,1967.
Numerous meetings and discussions were held to cstablish what requirements DRL would impose on the B&W training program in order for applicants to meet the " cold" eligibility requisites set forth in 10 C.F.R. 155.25(b).
At that time, B&W had a considerable interest in developino nuclear power punt training programs since five Construction Permits (cps) were issued for B&W plants between November, 1967 and November, 1968.
(Indian Point Unit 1 was the only operating B&W plant at that time but it was a radically different design.)
When B&W decided to procure a simulator, DRL was closely involved with them in the development of the capabilities and specifications of the simulator 8d120'-
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and how it was to be incorporated into the training programs.
In fact, in December 1969, two members of the Operating Licensing Branch visited the manufacturer, Singer Link Company in Sunnyvale, California to witness the B&W acceptance testing of the simulator prior to it being moved to Lynchburg, Virginia. The same OLB examiners also visited the B&W training center in Lynchburg the following August for progress reports on the B&W training pro-grams and to observe operation of the simulator.
To evalute the B&W training program, approximately 10 SMUD cold eligible license applicants were examined on the simulator in the Fall of 1971. The purpose of the examinations was not to evaluate the individual applicants, thus no results were retained.
However, on the basis of their performance a letter was issued on March 7,1972 from the Chief, OLB to B&W stating that the use of the simulator was acceptable in B&W's training program which they submitted to the AEC dated December 18, 1968. Since 1972, OLB examiners have made numerous visits to the B&W training center for the purposes of receiving training and conducting audits of the training programs.
Several revisions of and additions to the B&W training programs have been submitted to the AEC (NRC) for review and approval.
Simulators are also used to provide startup certification eligibility for
" hot license" applicants. The criteria for eligibility for examination with no reactor startup demonstration is contained in Appendix F of NUREG-0094.
The applicant must have satisfactorily completed an NRC approved training pro-gram for at least one week duration at a nuclear power plant simulator. The basis for determining acceptability of these programs is their similarity
'18f121 to the cold eligibility programs.
The Rancho Seco approved hot license training programs (Topographical Report T1-76) includes a three-week simu-lator course in which the reactor startup certification is incorporated.
An integral part of simulator training programs is the quality of the instruc-tional staff.
Although now i+ is an adopted recommendation of SECY-330E, there was never a requiremer.t that simulator instructors hold or have held licenses.
The attached letter from the B&W Manager of Training Services to tne Chief, Operator Licensing Branch demonstrates their effort to provide a quality training program.
In January, 1979 NASI/ANS 3.5-1979, " Nuclear Power Plant Simulators for Use in Operator Training" was published by the American Nuclear Society. The pt rpose of the standard was to establish the minimum requirements for nuclear power plant simulators for use in operator training and requalification pro-grams.
As a result of the TMI-2 accident, it was recognized that the require-ments set forth in the standard were inadequate and therefore it is presently undergoing revision.
In the future, all new and existing simulators will be required to comply with this standard in order to be considcred acceptable for use in training and requalification programs.
This response was prepared by Bruce Wilson.
183 122
(Attachment to Interrogatory Response 18)
IFUCOCh 88/i[COX po.er ceneraan croup P.O. Box 1260, L>nthburg, Va. 24505 Telephone: (S?4) 384 5111 9H-14 KRCl-77 March 14, 1977 Mr. P. F. Collins, Chief Operator Licensing Branch
';uclerr Regulatory Coc: mission Room 330 Philips Euilding 7920 Norfolk Avenue F.e th esda, MD 20014
Dear Fa. Collins:
Enc 1csure I has been prepared and is. submitted as a proposed revision to the Sabcock & Wilcox Training Program to establish eligibility to sit for license examination prior to initial criticality.
Enclosure I is essentially the sare as the draf t program informally presented to you on January 18, 1977.
The caly changes have been the correction of ninor typographical errors.
It is requested that you review our program and provide us with your cc aents or reconcendations.
Sir.ce cur last submission, we have centinutd our effort to upgrade the technical cualifications of our training staf f, primarily through the hiring of technically cc ;etent S?O Liccnsed individuals and training and licensing of our own A:sociate Instructors.
The current status of cur training staff is as fo11cus:
' sME POSITION LICENSE
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Senior In's'tructor SRO Oconee I, II, III Senior Instructor SRO Ecaver Valley, Unit I Instructor SRO Crystal River, Unit III Instructor SRO Crystal River, Unit III
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Instructor SRO Three Mile Island, Unit I Instructor SRO L> chburg Pool Reactor + on-shift experience at Crystal River, Unit I.
Associate Instrue.
None Associate Instrue.
None Associate Instrue.
Eone Sincere yr ?,/,
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~ / y' ( / /. 7 s
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'N.[S. Elliott, nager Training Services & Special Frejects
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~ 18 123 Tr e Ecb:::k & V :ci Ccr ;3ny / Ecst's'ec )E67 Interrogatory 19 Regarding Interrogatory 18, describe the criteria and basis for determining that a simulator training program is sufficient to provide reasonable assurances of utility shift personnel competency to respond to off-normal or transient conditions.
Response
This information is substantially contained in the response to Interrogatory 18.
We note that simulators devute approximately one-fourth to one-third of their time to transient and off-normal conditions.
This response was prepared by Bruce Wilson.
i83 l'24 Interrogatory 20 Describe current NRC minimum requirements, their basis and enforceability, for shift supervisor and senior reactor operator qualification.
Response
The NRC does not have any regulations pertaining to " shift supervisors" per g.
All shift supervisors must, however, hold a Senior Operator license.
10 C.F.R. 550.54(l). The minimum requirements for Senior Operators are set forth in Regulatory Guide 1.8.
In Section 6 of the Technical Specifications the licensee has committed to meet or exceed these minimum qualifications.
This response was prepared by Bruce Wilson.
e Interroga+.ory 21 R!garding Interrogatory 20, describe all on-going efforts to up-grade requirements of utility shift personnel qualifications.
Response
All on-going efforts to up-grade requirements of utility shift personnel are sumarized in Draft NUREG-0660, " Action Plans for Implementing Recommenda-tions of the President's Commission and Other Studies of TMI-2 Accident."
This draft Action Plan incorporates the recommendations and establishes a schedule for implementation of SECY-330E, Short Term Lessons Learned, Long Term Lessons Learned, and the Kemeny Commission Report. Also, ANS 3.1 and Regulatory Guide 1.8 are currently being revised to reflect the changes in rcquirements of utility shift personnel qualifications.
Proposals have been rdceived and are presently being evaluated in response to RFP-NRR-80-117,
" Study of Requiremente for Operator Licensing."
This response was prepared by Bruce Wilson.
' 18p 126 Interrogatory 22 Regarding Interrogatory 21, explain the nature of the up-graded requirements and the schedule for implementa-tion of said requirements.
In particular, provide the Rancho Seco implementation schedule.
Response
This information is contained ir. Draft NUREG-0660.
This response was prepared by Bruce Wilson.
Interrogatory 23 Regarding Interrogatory 22, describe the instructional guidelines for such up-grading of requirements.
Response
The up-grading of requirements of shift supervisor and senior reactor operator qualifications are in two areas - formal education and experience. All schedules, i'RC Actions, priorities, resource and budget estimates, and " instructional guide-lines," if any, are contained in NUREG-0660.
This response was prepared by Bruce Wilson.
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'18 128' Interrogatory 24 Describe all documents, processes, and procedures related to NRC's ability to monitor and verify the licensee's management and technical support during normal operation and during an emergency.
Include in this response all criteria used by NRC to determine the acceptability of licensee management and technical support capabilities at Rancho Seco.
Response
The response to this interrogatory is co red in the response to Interroga-tory 4.
This response was prepared by Daniel Garner.
m Interrogatory 25 Describe all documents, processes, and procedures, to deter-mine the qualifications and level of adequate training for non-licensed personnel such as managers, engineers, auxiliary operators, maintenance personnel, and technicians. On what basis are present NRC guidelines regarding this matter deter-mined to be suitable for an operating licensee.
Response
Currently, qualifications and training requirements are applied only to the personnel within the operating organization of the plant.
Unlike reactor operators, persons employed as managers, technical support personnel, auxiliary operators, maintenance p;rsonnel, and technicians are not licensed by the NRC. We do, however, review an apulicant's proposed qualifications and training requirements using the Standard Review Plan, Sections 13.1.3 and 13.2.
The standards for acceptability are provided in ANSI 18.1-1971 as endorsed by Regulatory Guide 1.8.
The basis for this endorsenent is a best concensus opinion among the staff cognizant of these matters.
This response was prepared by Daniel Garner.
18 130 -
Interrogatory 26 Describe all documents, processes and procedures which refer to minimum shift staffing of licensed reactor operators.
Response
10 C.F.R. 550.54 paragraph., (k), (1) and (m) specify the plant conditions for which licensed operators and senior operators must be present, at the control, and/or readily available on-call.
Table 6.2-1 of the licensee's Technical Specifications, " Shift Crew Personnel and License Requirements" specifies the minimum shift staffing of licensed reactor operators.
I'& E Bulletins79-05c and 79-06c directed that each PWR provide two licensed operators in the control room at all times during operation to accomplish the reactor coolant pump trip requirement should it be necessary.
The Lessons Learned Task Force Final Report stated "... consideration should be given to requiring the presence in the control room at all times during normal operations of two reactor operators and one senior reactor operator."
NUREG-0585, p. A-9.
This response was prepared by Bruce Wilson.
18 131-Interrogatory 27 Describe all processes and procedures which relate to NRC review of licensee administrative procedures re-garding utility shift personnel, including criteria for determining the acceptability of said administrative procedures at Rancho Seco.
Response
All procedures, including administrative procedures, are reviewed by the NRC staff in accordance with Regulatory Guide 1.33 (enclosure) and ANSI Stan-dard N13.7-1976. The NRC staff reviews each applicant's PSAR to ensure a commitment to meet Regulatory Guide 1.33 or an acceptable alternative.
In the FSAR, the procedures identified are compared against those listed in the Regulatory Guide to assure conformance with it.
These two functions are performed in the Operator Licensing Branch, NRC.
Once the procedures are written, it is an I&E function to assure compliance with the applicant's commitments.
Inspectors use their technical judgment in evaluating the contents of the licensee's procedures.
Any comments the inspectors have regarding their judgment of the procedures are resolved with the management of the facility.
This response was prepared by Bruce Wilson.
18 132 Interroaatory 28 Describe the criteria used by NRC to determine the acceptability of emergency operating procedures for nuclear plants.
In particular, how have Rancho Seco emergency operating procedures been determined to be acceptable.
Response
See response to Interrogatory 27. The same processes, procedures and criteria are applied to emergency procedures.
Two of Rancho Seco's emergency procedures were specifically reviewed by the NRC staff to insure compliance with the May 7, 1979 Commission Order. These procedures were D.5, " Loss of Reactor Coolant / Reactor Coolant Pressure," and D.14, " Loss of Steam Generator Feed."
As a result of a recommendation of the Lessons Learned Task Force, a special reyiew group will soon be established in the NRC to review and approve emer-cency procedures for all nuclear power plants.
This response was prepared by Bruce Wilson.
O Interrogatory 29 Regarding Interrogatory 28, describe the criteria by which NRC determines the acceptability of a licensees consideration of the compatibility of the design bases of the systems involved, with the discipline of human factors, in the development of emergency operat-ing procedures. Describe how this has been applied at Rancho Seco.
Response
The "TMI-2 Lessons Learned Task Force Final Report," NUREG-0585, stated on page A-9 that " Previous NRR reviews and I&E reviews of emergency operat-ing procedures did not specifically investigate their compatability with the design basis of the systems involved nor was the discipline of human factors included." We therefore have not yet established criteria for determining acceptability. The recommendation of NUREG-0585 is addressed in Action Item 1.C of Draft NUREG-0660.
This response was prepared by Bruce Wilson.
8
- 9 g,
y f8 134
Interrogatory 30 Describe all documents, processes and procedures utilized by NRC in developing minimum acceptable criteria for opera-tions verifications procedures.
Include the status and time table for implementation of the installation of status monitoring equipment in conforming with Regulatory Guide 1,47 at Rancho Seco.
Response
The NRC Lessons Learned Task Force, in its final report - NUREG-05SS, identified the need for a more effective system of verifying the correct per-formance of operating activities. See Section 2.3.6, Table A-1 (Recommendation 5), and p. A-10.
The Task Force recommended that automatic status monitoring be required for all plants not presently having such capability and that this could be accomplished by requiring such plants to meet the recommendations of Regulatory Guide 1.47 " Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems."
Following-up on the recommendations in NUREG-0585 and in other reports spawned by the TMI-2 accident, the NRC Staff has now proposed to the Conmission a Task for improving control room design NUREG-0660, Task I.D.1.
The Staff's proposals are still only in the draft stage, but would require installation of automatic system status monitoring (by means such as those described in Regulatory Guide 1.47) at operating reactors by December 1981.
See NRC Actions 4 and 7 and Licensee Action 4.
The Staff expects to issue two further drafts of NUREG-0660 before it is presented to the Commission for final approval. Therefore, no plan for implementation of Regulatory Guide 1.47 (or similar requirements) has yet been adopted f'or operating plants (such as Rancho Seco).
183/135 The above response was prepared by Dale Thatcher.
The only additional documents, processes, and procedures related to criteria for " operations verifications procedures" are NUREG-0578 (Section 2.2.1.c) and Draft NUREG-0660 (Task I.C.1(2)), both of which deal with shift relief and turnover procedures.
The latter portion of this response was prepared by Bruce Wilson.
18,3;h 136 Interrogatory 31 Pursuant to the Recommendations of NUREG-0578, each licensee is now required to have an operations experience evaluation group. Describe all processes and procedures utilized by NRC to determine the competency and acceptability of the personnel, procedures, analyses, or other efforts of SMUD's OEEG.
Response
NUREG-0578 does not specifically require the assembly of an " Operations Experience Evaluation Group." However, recommendation 2.2.1.6 of NUREG-0578 requires each licensee to provide an on-shift technical advisor to the shift supervisor who would be assigned normal duties to include review and evaluation of operating experience.
In a letter dated September 13, 1979, which forwarded NUREG-0578 to each licensee, alternatives to the specific requirements 2.2.1.6 were suggested which would allow some flexibility in the implementation of th'is recommendation by licensees.
By letter dated October 30, 1979, clarification of the NRC position on recommendation 2.2.1.6 was forwarded. The NRC staff will use information contained in the above three cited documents in determining the acceptability of SMUD's proposed method of implementing this requirement.
This response was prepared by Daniel Garner.
b 18Bi
~13 7 Interrogatory 32 Describe all documents, processes and procedures utilized and proposed by NRC, to assure that operators and other operations personnel are continually provided with lessons learned from operating experience at other reactors in the United States.
Response
Procedures utilized to collect data from licensees are described in the response to Interrogatory 16. Current procedures for dissemination of operating experience to licensees is also discussed in that response.
The long-term efforts to be taken to ensure feedback of operating experience at other facilities include actions by both the NRC and the nuclear industry. The efforts that will be taken are summarized as Task I.E of NUREG-0660 (DRAFT) of December 10, 1979.
Th'is response was prepared by Daniel Garner.
.h j}h Interrogatory 33 Describe all documents, processes and procedures utilized by NRC to determine explicit criteria for control room design review.
Include all specific requirements for backfitting existing control rooms to correct deficiencies.
Response
The following documents set forth the Staff's efforts to establish criteria for control room design review:
1.
The NRC Lessons Learned Task Force, in its final report - NUREG-0585, identified the need for better man-machine interf ace (Recommendation 7).
The Task Force made a number of recommendations regarding control room design reviews. See pp. A-11 through A-13.
2.
The NRC has prepared or funded a number of studies concerning control room design. These were identified in the response to Interrogatory 13 of the California Energy Commission's First Set of Interrogatories.
3.
In addition, Task I.D.1 of NUREG-0660, which is a draft set of action plans for implementing recommendations of the Presiderit's Commission and other studies of the TMI-2 accident (including NUREG-0585), deals with control room design. Pursuant to this proposed Task, licensees of operating plants (such as Rancho Seco) would have to complete comprehen-sive reviews of their control rooms using NRC human factcr guidelines and make modifications to correct significant deficiencies by February 1981 (Licensee Action 1). Licensees would also have to install a safety monitor console by June 1, 1931 (Licensee Action 3). As noted in response to Interrogatory 30, the Staff's Task Action Plans are still only in a draft stage and have not yet been presented to the Commission for final approval.
This response was prepared by Dale Thatcher.
18jl 140 Interrogatory 34 Describe all documents, processess and procadures utilized by NRC to evaluate the technical basis for definitive li-censing criteria for manual and automatic operations for systems which execute plant safety functions and safety-related functions.
Objection We object to this interrogatory as it is overly broad. The interrogatory basically requests all information supporting all NRC licensing criteria.
~
Without waiving our objection, we provide the following response.
Response
The above interrogatory is extremely broad in nature.
Numerous plant systems and components are related to the execution of plant safety functions. Li-censing criteria for such plant systems and components, be they manually or automatically operated, are extensive and must be viewed as a whole to judge plant safety. Generalized licensing criteria for the operation of safety related systems are found in 10 C.F.R. Part 50, while more detailed cr:teria are found in the USNRC Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, NUREG 75/087, and Division 1 Regulatory Gu ides. Evaluation and review procedures are described in the Standard Review Plan referenced above.
This response was prepared by Mark Rubin.
} 4l Interrogatory 35 Regarding Interrogatory 34, describe all documents, processes, and procedures related to the determination of the feasibility of backfitting existing plants, in particular, Rancho Seco.
Response
Pursuant to 10 C.F.R. 550.109, the Commission may require the "backfitting" of a facility "if it finds that such action will provide substantial, addi-tional protection which is required for the public health and safety or the common defense and security." Backfitting is defined in the regulation as "the addition, elimination or modification of structures, systems or compo-nents of the facility after the construction permit has been issued." A pro-cedure has been developed whereby any proposal to impose a backfit require-ment upon a facility (ies) must be approved by the Staff's Regulatory Require-ments Review Committee.
This response was prepared by Mark Rubin.
183[.
l42~'
Interrogatory 36 Describe all documents, processes and procedures related to the establishment of standard design requirements for control rooms.
Describe the evaluation of standard design requirements with reference to their applicability to existing control room designs.
Response
The response to this interrogatory is contained in the response to Interrogatory 33.
This response was prepared by Dale Thatcher.
Interroaatory 37 Describe all documents, processes, and procedures related to evaluating the interaction of non-safety and safety grade systems during normal operation, transients, and design basis accidents to assure that any interaction will not result in exceeding the acceptance criteria for any design basis event.
Include an explicit definition of the acceptance criteria.
Objection We cbject to this interrogatory insofar as it relates to matters other than the ability of the facility to respond to feedwater transients. With respect to feedwater transients, we provide the following response.
Response
A study has been underway since 1978 to study the impacts of systems on other systems. A fault-tree method has been developed and is being applied to a reference plant, as part of the project which is known as Task Action Plan A17.
In the Lessons Learned Task Force Final Report (NUREG-0585), it was recom-mended that additional work be performed to determine the possible effect from the interactions between non-safety grade and safety grade equipment.
The December 10, 1979 revisions to Task Action Plan A17 include provisions for applying system interactions considerations to operating plants.
Presently, the schedule calls for the extension of fault-tree systems-interactions methodology by June 1930.
If requirements for various reactor designs are identified, letters will be sent to operating license holders to implement modifications emanating from the systems interactions study.
18;7 144 Additionally, a Regulatory Guide is scheduled for issuar.ce by December,1980 to provide the NRC Staff position on the application of Systems Interaction methodology, and to present the acceptance criteria for the staff review in this area.
Prior to the completion of the above work, a continuing study on the reliability of the Rancho Seco auxiliary feedwater system is being performed.
The effect of the non-safety. grade integrated control systen (ICS) on the safety quali-fied auxiliary feedwater syste") is one of the areas of review. Any system modifications recomended by the staff as a result of this study will be required of Rancho Seco.
In addition, on September 17, 1979, all operating light water reactors re-ceived interaction between non-safety grade systems and safety grade systems.
The information request was directed towards the interaction of these systems during a ligh energy line break, where environmental conditions might effect various system components.
On September 20, 1979, a sumary of this meeting was issued. Also on October 5, 1979, P,ancho Seco submitted their evaluation of potentially adverse environmental effects on non-safety grade control systems in response to the staff's request. On December 19, 1979, the staff issued its status report on B&W reactors in regards to this issue.
Copies of these materials are enclosed.
This response was prepared by Mark Rubin.
181
'l45-Interrogatory 38 Describe all design features necessary to mitigate the consequences of a core melt or severe core damage which provide reasonable assurances that the health and safety of the public are protected. On what basis does flRC determine what assurances are reasonable. Likewise, on what basis does !JRC determine what design features are necessary to provide such assurances.
Interrogatory 39 Regarding Interrogatory 38, in lieu of such features, describe additional and supplemental means of prevent-ing core damage or core-melt accidents, through improved engineered safety features.
Interrogatory 40 In reference to Interrogatory 38, describe the objective of such design features. Describe the design objectives by a set of specific acceptance criteria.
Interrogatory 41 In reference to Interrogatory 38, describe the characteristics and functions of such design features.
Interrogatory 42 In reference to Interrogatory 38, describe in detail:
A.
The probabilities and consequences of the various event sequences that might result in releasing significant amounts of radioactivity to the environment. Which sequences are amenable to interdiction and Dy what means?
B.
The expected effectiveness and performance of suggested means of reducing the consequences of events in which severe damage or substantial melting of the core occurs, in particular, systems for controlled, filtered venting of the contain-ment and for preventing the uncontrolled combustion of hydrogen.
C.
Other requirements, and in particular those of siting, emergency plans and procedures, training or other re-lated areas, which would be modified if such design features were required.
18p 146
e
. 45 D.
Additional information required or desirable before setting requirements.
E.
The final form of the requirement. What will be the implementation schedule for new plants, plants under construction, and operating plants?
Objection to Interrogatories 38 through 42 We object to these interrogatories on the ground that they are not relevant to the issues in this proceeding, which are limited to the ability of the facility to respond to feedwater transients.
Interrogatory 43 Describe in detail the approach, methods, and organiza-tion of the NRC staff in performing post-licensing reviews of nuclear plants with particular attention paid to the following specific items:
A.
An overall system level, integrated review that gives full consideration to operational safety aspects and provides for a design basis accident assessment function from event initiation through consequence mitigation, including the review of emergency operating procedures.
B.
Timely analysis of operating experience and imple-mentation of needed changes derived fron, operating experienca.
C.
Discipline in the application of a single overall safety goal.
D.
Technical oversight of Safety Evaluation Reports to assure increased emphasis on safety while still satis-fying the requirements of the administrative process of regulation.
E.
Assurance of adequate operations experience and training for the NRC technical review staff, especially those staff members assigned responsibility in accident response situations.
F.
Dedication of adequate resources to the three principal functions of the Office of Nuclear Reactor Regulation:
reactor licensing, oversight of operating reactors, and resolution of generic safety issues.
G.
Use of a formal procedure for follow-up on ques-tions and requests from the Advisory 'ommittee on Reactor Safeguards and its individual members.
Objection We object to this interrogatory insofar as it relates to the NRC Staff's con-duct of " post-licensing reviews" in general.
Such reviews are not relevant to the issues in this proceeding. We are, however, previding the following response with respect to the " approach, methods and organization" of,the '
Staff for review of the ability of B&W facilities to respopd to fe.edwater.
O l.kb
- 47.
Response
It is noted that this interrogatory essentially repeats recommendation 12 of the TMI-2 Lessons Learned Task Force Final Report (NUREG-0585). The Task Force recommended that the NRC revise the process of licensing reviews to emphasize the objectives itemized in the interrogatory. Commission action on the recomendations of NUREG-0585 is still pending, and it is not possible to provide a response which addresses the listed objectives at this time. However, a general description of how the staff conducts post-licensing reviews is provided below.
Post-licensing reviews can originate from a number of sources including license conditions, events at operating plants, office of Inspection and Enforcement (I&E) Bulletins, notifications under 10 CFR 21, and concerns ge_nerated among the NRC staff. The normal organization that would be utilized in performing the review in connection with Rancho Seco's capability to respond safely to feedwater transients is shown in the attached chart.
Management of the review for Rancho Seco would be carried out by Operating Reactors Branch #4, and technical input would be provided by one or more of the technical branches. Depending on the origin of'the review, the licensee may be requested one or more times to provide information on how the plant is designed or operated to conform to acceptance criteria as provided in a general design criteria, regulatory guide, standard review 18$/149 plan, or a newly developed staff position.
If necessary, NRC personnel involved in the review would visit the plant to clarify information received from the licensee. When sufficient information is gathered, a safety evaluation report (SER) would be written describing the nature of and reason for the review, and how the licensee does or does not conform to the acceptance criteria.
In the case of non-conformance, commitments from the licensee to modify the plant would be requested, and these commitments would be documented in the SER. Verification that modifications have been accomplished, if any are required, would be carried out by I&E inspectors.
On occasion, special events (such as the TMI-2 accident) may warrant the establishment of interdisciplinary review groups, or task forces, to conduct a post-licensing review. These task forces are formed by assembling personnel from a variety of branches within the Commission. The Bulletins and Orders Task Force is an example of such an interdisciplinary review gr6up.
This response was prepared by Daniel Garner.
Interrogatory 44 For each of the TMI lessons learned task force short term recommendations describe the basis or criteria utilized by NRC to determine:
A.
Compliance; B.
That recommendations explicitly provide reasonable assurance that Rancho Seco can be operated safely.
Objection We object to this interrogatory as not relevant to the issues in this pro-ceeding. The Commission's Orders of May 7 and June 21, 1979 preceded the short term recommendations and defined the issues in this proceeding.
9 M
182/151 Interrogatory 45 Will SMUD adhere to the implementation schedule required by flVREG-0578?
If not, what sanctions will tiRC impose?
If none, why not?
Objection We object to this interrogatory for the reasons noted in our response to Interrogatory 44. Without waiving our objection, we would direct F0E's attention to the letter of January 2, 1980 from Mr. Denton (f;RC) to Mr. Mattimoe (SMUD), a copy of which was sent to the service list. This letter encloses an Order to Show Cause with regard to the implemention of the Category A (i.e., short-term) Lessons Learned set forth in flUREG-0578.
51.
Interrogatory 46 Describe any additional measures not included in the May 7 Order, NUREG-0578, or I & E Bulletins issued, identified by MC. SMUD, cr others, that would enhance the safety and reliability of Rancho Seco in responding to various trans-ient events or provide greater assurance of the safe and reliable operation of the facility.
Interrogatory 47 For any improvement that could increase safe and reliable operation of Rancho Seco identified in response to Interrogatory 46, what criteria and procedures are used by SMUD and the NRC in determining whether to implement them and the timetable for implementation, and whether to shut down or derate the plant pending successful implementation.
Objection to Interrogatory 46 and 47 We object to these interrogatories to the extent they relate to other than th'e provision of reasonable assurance that the facility can respond safely to feedwater transients. Any other transients or any other aspects of the safe and reliable operation of the facility are not relevant to the issues in this proceeding. With respect to feedwater transient:,, we provide the following response.
Response
All actions which the Commission considered necessary to provide reasonable assurance that Rancho Seco could safely respond to feedwater transients pending completion of the long-term actions identified in the May 7, 1979 Order were required to be completed prior to the restart of the facility. See June 27, 1979 Staff Evaluation. The Licensee's compliance with I & E Bulletins79-05A and 79-05B is also documented in the staff's Evaluation of November 23, 1979.
Other actions, also denominated "short-term," were recommended in NUREG-0578 i8sp I53 --
and have now been imposed on SMUD, as well as all other licensees. See January 2, 1980 Order to Show Cause.
fiumerous other reports containing recommendations for additional steps to enhance safety have been prepared by groups both within and outside the f1RC.
The Staff has reviewej these various recommendations and has prepared a draft set of " Action Plans for Implementing Recommendations of the President's Commission and other Studies of TMI-2 Accident" (fiUREG-0660, December 10,1979).
This document is expected to be revised before being presented to the Commission for its final approval.
The basic criterion that would be used by the Commission in determining whether to impose certain requirements would be whether they are required to provide reasonable assurance for the protection of the public health and safety. To the extent that backfits would be required, the criterion of 10 C.F.R. 150.109 (see response to Interrogatory 35) :muld also be applied.
In determining time-tables for required changes the Commission would have to consider the urgency for the changes and such additional matters as the availability of the equipment.
This response was prepared by Mark Rubin.
1 81/d154
~
Interrogatory 48 Are the safety design and operation requirements for Rancho Seco as stringent as those for new plants apply-ing for a C.P. or 0.L.?
Interrogatory 49 Regarding Interrogatory 43, if requirements differ, are there two different standards for design and operation?
Response to Interrogatory 48 and 49 The stringency of design and operational requirements for Rancho Seco necessary to assure safe plant shutdown following postulated feedwater transier.ts have been reviewed relative to corresponding requirements for new plants applying for a C.P. or an 0.L.
It is concluded that the Rar.cho Seco requirements already in effect along with those committed by the licen-see forifuture implementation and those still under review by the staff as discussed below are as stringent as those for new plants.
The basis for this conclusion is discussed below.
Rancho Seco was issued an operating license in August 1974. Since that time there have been no changes in any NRC regulations, i.e. General Design Criteria in 10 CFR Part 50, Appendix A, which establish design and perfor-mance requirements of safety systems whose operation is required to assure safe plant shutdown following postulated feedwater transients.
Thus, the applicable General Design Criteria are the same for Rancho Seco as for new plants.
T83/155-
Sa -
Further, the additional requirements with respect to feedwater transients imposed by the NRC staff since the THI-2 accident are at least as stringent
(
as those for new plants. These requirements include those in IE Bulletins79-05A, 05B, and 05C; the requirements of the Commission Shutdown Order of May 7,1979; and the short and long term Lessons Learned requirements (NUREG 0578 and 0585).
In fact, some of the requirements of the above documents may be more stringent than curre,t licensing review guidelines (e.g. Regulatory Guides, Standard Review Plans) for new plants.
In those cases, such requirements will be incorporated into the review of new plants, as applicable.
Also, the licensee (SMUD), as part of the long term requirements of the Commission Shutdown Order of May 7,1979, submitted for NRC review by SM'JD letter dated December 17, 1979, a report entitled " Auxiliary Feedwater System Reliability Analysis for the Rancho Seco Nuclear Generating Station Unit No.1" dated December 1979 As a result of this analysis, the licen-see has committed to additional Am system design and procedure modifica-tions to further improve overall system reliability, This analysis is still under review by the staff. Any additional staff recomendations for improving AFW system reliability resulting from the review of this analysis will be sent to those parties on the Rancho Seco Hearing document service list.
In regard to different requirements for operating plants versus plants applying for C.P. or 0,L,, the Staff thus far has required AFW system reliability evaluations for operating PWR plants. This requirement will be imposed on PWR plant 0.L. applicants,
55 -
Thus, it is concluded (1) that the design and operational requirements for Rancho Seco safety systems necessary to assure safe plant shutdown follow-ing postulated feedwater transients e;e at least as stringent as, if not more so, than those for new plants applying for C.P. or 0.L. and (2) that the Rancho Seco design of those systems provide reasonable assurance of public health and safety following such events.
This response was prepared by Philip Matthews.
1
56 _
Interrogatory 50 Based on present licensing criteria and regulations applicable to plants seeking a C.P. or 0.L., would the NRC approve the present Rancho Seco design con-figuration for an 0.L.?
Interrogatory 51 Regarding Interrogatory 50, if not, why not?
If so, who so?
Response to Interrogatory 50 and 51 The response to Interrogatories 48 and 49 concluded that the design and operational requirements for Rancho Seco systems necessary to assure safe plant shutdown following postulated feedwater transients are as stringent as those for new plants and thus the design of these Rancho Seco systems provide reasonable assurance of public health and safety following such events.
However, the staff cannot definitely state that the present Rancho Seco design configuration would be recommended for issuance of an 0.L. at this time.
The reason for this uncertainty is the possibility of the staff establishing additional plant requirements associated with feedwater tran-sients that might occur as a result of the staff review of B&W plant sensitivity to changes in feedwater flow rate or temperature as described in NRC Memorandum, dated November 16, 1979 from D. Eisenhut to S. Scott, Subject " Rancho Seco Board Notification 10 CFR 50.54 Request Regarding Design Adequacy of Babcock and Wilcox NSSS." The staff review of this issue is in progress.
If additional requirements are established which are considered to provide substantial, additional protection required for public health and safety, they would be backfitted to plants with operating
~
18'30 158 f
57 -
licenses in accordance with Section 50.109 of 10 C.F.R. Part 50 of the Commission's regulations.
This response was prepared by Philip Matthews.
58 -
Interrogatory 52 Which, if any, of the following are necessary to pro-vide a reasonable assurance of Rancho Seco's safe and reliable operation.
A.
Redundant power operated relief valves that can override releases of primary system radioactive coolant.
B.
A recombiner to mitigate hydrogen formation.
C.
Better radiation monitoring devices at Rancho Seco and surrounding areas to properly quantify radiation releases in the event they occur.
D.
Use of other reactor systems that would provide less risk to the public in the event of feedwater transients.
E.
Venting of hydrogen from the reactor core at Rancho Seco if it is created by circumstances similar to those that occurred at Three Mile Island.
F.
A revised evacuation and emergency response plan for Rancho Seco and surrounding communities.
G.
An automatic accident notification system.
H.
A controlled, filtered venting system to mitigate unavoidable releases of radionuclides.
I.
A revised measurement system to better inform Rancho Seco operators of hydraulic conditions in the steam generator, pressurizers, and reactor vessel, J.
Redesign of Rancho Seco's control room to be con-sistent with modern principles of human engineering.
K.
Revised consideration of the possibility of multiple and common-mode failures in Rancho Seco's design and operating procedures.
Objection We object to Interrogatory 52 to the extent it relates to provisions that are not related to the ability of the facility to respond to feedwater tran-aients. Without waiving our objection, we provide the following response.
59 -
Response
The staff required by its letter of September 13, 1979, that plant design be modified to satisfy the requirements of NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations." Additional require-ments were also specified by the same letter regarding the capability for high point venting of the reactor coolant system. The staff has also published NUREG-0585, "TMI-2 Lessons Learned Task Force Final Report," which is being considered for implementation by the Commission in its review of NUP.EG-0660,
" Action Plans for Implementing Recommendations of the President's Commission and Other Studies of the TMI-2 Accident."
These requirements identify the need for improvement in areas related to Items C, E, F, G, H, I, J and K.
This response was prepared by Thomas Novak.
i83[l61-
_ 60 _
Interrogatory 53 Describe and identify any outstanding long-term generic issues related to the Commission Order of May 7, as it specifically concerns Rancho Seco.
Response
The Commission Order of May 7, 1979, issued to the Sacramento Municipal Utility District, identifies on page 5 (of the Order) four long-term modifications which were to be implemented by the licensee to "... further enhance the capability and reliability of the reactor to respond to various transient events." This response will quote the requirements of the Order and discuss the status of each of the requirements.
L0r:G-TERM ITEM 1 "The licensee will provide to the f;RC staff a proposed schedule for implemen-tation of identified design modifications which specifically relate to items 1 through 9 01 Enclosure 1 to the licensee's letter of April 27, 1979, and would significantly improve safety."
Items 1 through 9 of Enclosure 1 to the licensee's letter of April 27, 1979, identify the 9 specific actions which the licensee took in the short-term to upgrade the timeliness and reliability of delivery from the auxiliary feedwater system (AFW). The licensee completed these 9 items prior to the restart of Rancho Seco following its shutdown on April 28, 1979. The staff verified the completion of and the adequacy of these items in its " Evaluation of Licensee's Compliance with the fiRC Order Dated May 7,1979."
In our letter from H. R.
Denton to J. J. Mattimoe, dated June 27, 1979, we directed the licensee to submit to the staff within 30 days a schedule for completing the long-term modifications described in the Order.
In its response, from W. S. Bossenmaier to H. R. Denton dated July 26, 1979, the licensee stated that "The District carefully reviewed Auxiliary Feedwater System operating procedures and verified that system design requirements are satisfied. The system was modified to improve both system flow indication and actuation information in the control room. The District will continue to review the design of the Auxiliary Feed-water System and if additional analysis or system enhancements are required, they will be made."
The staff desired that this matter be pursued on all B&W facilities, includ-ing Rancho Seco. Therefore, in a meeting with the B&W Owners' Group on July 19, 19'79, the staff directed the B&W licensees to perform an AFW reliability study.
The purpose of the study would be to identify any additional system or component modifications which could be made to the AFW system to further upgrade its time-liness and reliability of delivery of water to the steam generators upon demand.
Several meetings have taken place subsequent to the July 19, 1979 meeting between the staff and licensee to monitor the progress of the study.
On December 17, 1979, the licensee forwarded its final Auxiliary Feedwater System Reliability Analysis for Rancho Seco Unit No. 1.
Included with this report was a list of items which the licensee committed to perform to further improve the reliability of its AFW system along with a schedule for completing the identified items. The staff is presently reviewing the reliability analysis 18p 165 -
. 62 _
as well as the identified modifications and implementation schedule.
This review should be complete by the end of January 1980. Following this review, the staff will provide an evaluation to the licensee which will address the acceptability to the staff of the proposed modifications and implementation schedule.
In addition, the evaluation will address any additional system modifications which it deems are necessary and appropriate.
L0tlG-TERM ITEM 2 "The licensee will submit a failure mode and effects analysis of the Integrated Control System to the flRC staff as soon as practicable."
By letter fron J. H. Taylor (B&W) to D. F. Ross (t4RC), dated August 17, 1979, B&W submitted this report entitled " Integrated Control System Reliability Analysis," BAW-1564. This was a generic report, prepared by B&W for all B&W operating plants; however, the facility chosen for the study was Rancho Seco.
On August 31, 1979, SMUD endorsed the conclusions and recommendations of the report.
Oak Ridge flational Laboratory (ORiiL) was contracted by the f4RC to review the report. A preliminary review of the report indicated that the recommendations made by B&W in the report were reasonable and should be addressed by the licensees. On tiovember 7, 1;/9, the licensee was requested to state its position with respect to the recommendations contained in BAW-1564.
In a letter from J. J. Mattimoe (SiluD) to R. W. Reid (f4RC) the licensee committed to providing its position by January 8, 1980.
f83/164 The final evaluation of BAW-1564 from ORf!L is scheduled to be received by mid-January 1980. At that time the staff will review its conclusions and recommendations along with SMUD's response to our flovember 7,1979 letter.
Following the staff's evaluation, the licensee will be informed what additional work and/or modifications, if any, will need to be done to improve the reli-ability and performance of the ICS or its interfacing systems.
L0f1G-TERM ITEM 3 "The reactor trip following loss of main feedwater and/or trip of the turbine to be installed promptly pursuant to this Order will thereafter be upgraded so that the corponents are safety grade. The licensee will submit this design to the f;RC staff for review."
The staff has reviewed the licensee's submittals of May 21 and October 5, 1979, in which it forwarded a preliminary design for upgrading the present control-grade anticipatory reactor trip for loss of feedwater and turbine trip to safety-grade.
By our letter from R. W. Reid (flRC) to J. J. Mattimoe (SMUD) dated December 20, 1979, the licensee was given preliminary design approval for the proposed upgrade. The staff Safety Evaluation which documents the basis for this approval was attached as an enclosure to the letter.
It is anticipated that final design, procurement of equipment and installation will take approximately 6 months from the date of our preliminary design approval.
183 165
. 64 _
LONG-TERM ITEM 4 "The licensee will continue operator training and have a minimum of two licensed operators per shift with TMI-2 simulator training at B&W by June 1, 1979. 1979. Thereafter, at least one licensed operator with TMI-2 simulator training at B&W will be assigned to the control room. All training of licensed personnel will be completed by June 28, 1979."
As documented in the licensee's letter from W. S. Bossenmaier (SMUD) to H. R. Denton(NRC) dated July 26, 1979, all Rancho Seco licensed operators (5 reactor operators and 17 senior reactor operators) completed TMI-2 simulator training on the B&W simulator at Lynchburg, Virginia on June 21, 1979. Satis-factory completion of this long-tern portion of the Order was indicated in the staff's " Evaluation of Licensee's Compliance with the NRC Order Dated May 7, 1979."
With regard to continued operator training, by letter dated September 21, 1979, from J. J. Mattimoe (SMUD) to D. F. Ross (NRC), the licensee documented that it had modified Administrative Procedure No. AP25 (" Licensed NRC Operator Training") to incorporate the requirement that TMI-2 type accident training would become part of the licensee's regular operator training and requalifi-cation program. The criteria used to determine sufficiency of both the Requalification Training Program and the Hot License Training Program at each of the B&W facilities, including Rancho Seco, were to assure that these pro-grams incorporated the following items:
18$f166~
. 65 _
A.
The following lecture subjects are to be included or expanded, as applicable, in each of the programs:
(1) Thermodynamics (2) Hydraulics (3) Fluid Flow (4) Heat Transfer (5) Small Break LOCA Phenomenon (6)
Inadequate Core Cooling, and (7) Transient Training including Loss of Feedwater.
In addition, the TMI-2 sequence of events is to be included in at least the first year's Requalification Program and all Hot License Programs.
B.
All programs are to include simulator training in which the operators or applicants are '.o be provided with hands-on experience in handling small breaks and other transients that could lead to loss of heat removal, inadequate core cooling and natural circulation.
This response was prepared by Robert Capra.
66 -
Interrogatory 54 Describe all documents related to the risk inolications of the sensitivity of the B&W design and on the poten-tial interactions arising from the I.C.S.
Objection We object to this interrogatory to the extent that it relates to the "sensi-tivity of the B&W design" to events other than feedwater transients. With respect to feedwater transients, we provide the following response.
Response
Risk studies, as properly defined, have generally not been performed in rela-tion to the sensitivity of the B&W design.
Rather, staff effort has largely been to study B&W system sensitivity from a perspective of system response and reliability.
A risk analysis includes the additional component of possible consequences to the public for system failures.
However, recently a proaram has been undertaken to conduct a limited risk assess-ment of a B&W reactor aimed at identifying any unique risk-impactino secuences relative to the Reactor Safety Study.
It is expected that this study will provide some insight into possible risk levels resulting from the sensitivity of the B&W desian.
This work is being done by the Probabilistic Analysis Staff of the Office of Research, and is expected to be completed by March 1930. Additionally, a failure modes and effects analysis has been performed on the B&W ICS, and is documented in BAW 1564.
This study does present various ICS interaction possibilities.
For further documents related to B&W plant sensitivity, see the response to F0E Interrogatory 55.
This response was prepared by Mark Rubin.
T87 168
. 67 _
Interrogatory 55 Describe any evidence or documents which suggest whether B&W plants are overly sensitive to feedwater transients due to the OTSG concept, es coupled with the pressurizer sizing, I.C.S. design, and PORV/ Reactor trip set points.
Response
Material which is related to the sensitivity of B&W plants can be found in NUREG-0560, the Commission's May 7 Order, and the 10 C.F.R. 150.54 request Regarding the Design Adequacy of Babcock and Wilcox Nuclear Steam Supply Systems Utilizing Once Through Steam Generators, which were sent to all B&W construction permit applicants. Analytical studies providing insight into the sensitivity of a plant similar to Rancho Seco to feedwater transients, is provided in the December 4, 1979 response of the Midland plant to the 150.54 request. A copy of this document is enclosed.
This response was prepared by Mark Rubin.
18 169 Interrogatory 56 Regarding Interrogatory 55, describe any increase in the frequency in reactor trips or AFW actuation as a result of modifications in the May 7 Commission order.
Response
The staff has not performed an in-depth statistical study on the number of plant trips which have occurred at B&W plants before and after the May Order.
The limited c3ta base available is probably not sufficient to arrive at any conclusive findings, until more operating experience is obtained.
- However, a review of the operating history of B&W plants does indicate the presence of some reactor trips which probably would not have occurred if not for the modifications of the May 7 Order.
Table F0E 56-1 lists the significant feed-water related transients from April through November 1979.
Specifical'.y, it appears likely that the four reactor trips that occurred at Crystal Piver on Auaust 16 and 17 would not have taken place if not for the Order modifica-tions requiring a reduction in the high pressure reactor trip setpoint.
With the presently available information, it is not possible to say with cer-tainty which events would not have occurred without the modifications.
How-ever, the staff feels that B&N operating experience does point strongly to this possibility.
Monthly operating reports for 1979 for Rancho Seco are also included in the response to this interrogatory. They will provide information on all reactor trips which have occurred over the past year.
This response was prepared by Mark Rubin.
e0
'4 181f170'
~
Table F0E 56-1 B8W OPERATING EXPERIENCE APRIL - NOVEMBER 1979 (Attachment to Interrogaton/
Response 56)
PLANT DATE POWER CAUSE TRIP SIGNAL RESULTING OVERFEED CRYSTAL 08-16-79 72%
RCP SECURED HP YES 08-16-79 45Z FN CONTROL TRANSFER HP YES 08-17-79 i48%
FW CONTROL PROBLEMS HP 08-17-79 26%
FW CONTROL PROBLEMS IIP 08-02-79 10%.
?
SPURIOUS FW PUMP RUNBACK HP ANO 08-13-79 75%
RELAY FAILURE ilP YES OCONEE 1 OC-11-79 99%
LOSS OF B00 STER PUMPS
FWP TRIP
~
?
FWP TRIP
?
.'0CONEE 2 05-07-79 28%
FW CONTROL PROBLEMS HP BEFORE TRIP c2?'
06-03-79 30%
FW CONTROL FAILURES HP BEFORE TRIP
__,OCONEE 3 11-10-/9 100Z REDllCED FW FLOW llP
~3MANCHO SEC0 07-12-79 100%
TURBINE TRIP ARTS 04-22-79 100%
INVERTER FAILURE FW MISMATCH HP Interrogatory 57 Describe any documents or criteria which define the point at which a given frequency rate for reactor trips or AFW actuation is:
A.
Desirable; B.
Undesirable; C.
Unacceptable.
Response
This information is provided in the NRC Staff Response to CEC Interrogatory 1 dated January 17, 1980.
This response was prepared by Mark Rubin.
9 18J/172
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70 -
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
SACRAMENTO MUNICIPAL UTILITY DISTRICT Docket No. 50-312 (SP)
(Rancho Seco Nuclear Generating
)
Station)
)
AFFIDAVIT OF ROBERT A. CAPRA I, Robert A. Capra, depose and state under oath as follows:
1.
I am a Project Manager (Nuclear Engineer) in the Nuclear Regulatory Comission Staff's Project Management Group of the Bulletins and Orders Task Force.
I am responsible for coordinating the review and evaluation of accions taken by the Babcock & Wilcox operating plant licensees in response to the Three Mile Island Unit 2 accident-related IE Bulletins ano Comission Orders. My professional qualifications are attached to the NRC Staff response to California Energy Comission Interrogatory 17 filed in this proceeding.
2.
The response to Friends of the Earth Interrogatory 53 was prepared by me.
I certify that the answers given by me are true and accurate to the best of my knowledge.
Robert A. Capra Subscribed and sworn to before me this day of January, 1980.
i8 173 Notary Public My Comission expires: July 1,1982 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
SACRAMENTO MUNICIPAL UTILITY
)
Docket No. 50-312 (SP)
DISTRICT
)
)
(Rancho Seco Nuclear Generating
)
Station)
)
AFFIDAVIT OF MARK P. RUBIN I, Mark P. Rubin, depose and state under oath as follows:
1.
I am a Reactor Engineer, Reactor Systems Branch, Division of Systems Safety, U.S. Nuclear Regulatory Commission.
I am responsible for evaluating the capability of reactor systems needed for safe shutdown during normal and accident conditions, including the performance of emergency core cooling systems.
In addition, I was on temporary detail to the Bulletins and Orders Task Force where I was involved in the evaluation of operating reactor responses to the bulletins issued follow-ing the accident at TMI. My professional qualifications are attached to the NRC Staff response to California Energy Comission Interrogatory 17 filed in this proceeding.
2.
The answers to Friends of the Earth Interrogatories 34, 35, 37, 46, 47 and 54 to 57 were prepared by me.
I certify that the answers given by me are true and accurate to the best of my knowledge.
/ '.
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Mark P. Rubin Subscribed and sworn to before me this 24th day of January, 1980.
S'E u i db
~18f174 Notary Public My Comission Expires: July 1, 1982.
72 -
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
SACRAMENTO MUNICIPAL UTILITY DISTRICT )
Docket No. 50-312 (SP)
)
(Rancho Seco Nuclear Generatina
)
Station)
)
AFFIDAVIT OF BRUCE A. WILSON I, Bruce A. Wilson, depose and state under oath as follows:
1.
I am a reactor engineer in the Nuclear Regulatory Comission Staff's Operator Licensing Branch.
I am responsible for the preparation and administration of written, oral, and practical exams for operators' and senior operators' licenses at production and utilization facilities.
Since May, 1979, I have been assigned to the Systems Group, Bulletins and Orders Task Force. My professional qualifications are attached to the NRC Staff response to California Energy Commission Interrogatory 17 filed in this proceeding.
2.
Identified portions of the responses to Friends of the Earth Interroga-tories 6 to 11, 14, 15, 17 to 23, and 26 to 30.
I hereby certify that the answers given by me are true and accurate to the best of my knowledge.
/ w 4.lAl -
/ / Bruce A. Wilson Subscribed and sworn to before me
's this 24th day of January,1980.
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In 1o ry u bTi~c- / ~
My Commission expires:
July 1, 1982 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
SACRAMENTO MUNICIPAL UTILITY
)
Docket No. 50-312 (SP)
DISTRICT
)
)
(Rancho Seco Nuclear Generating
)
Station)
)
AFFIDAVIT OF DALE F. THATCHER I, Dale F. Thatcher, depose and state under oath as follows:
1.
I am a reactor engineer in the Nuclear Regulatory Commission Staff's Instrumentation and Control Systems Branch.
I am responsible for the review and evaluation of the instrumentation and control systems of nuclear power generating stations.
My professional qualifications are attached to the NRC Staff Responses to California Energy Commissica's First Set of Interrogatories.
2.
The answers to the Friends of the Earth Interrogatories 33 and 36 and part of Interrogatory 30 were prepared by me.
I hereby certify that the answers given by me are true and accurate to the best of my knowledge.
Dale F. Thatcher Subscribed and sworn to before me this 24th day Januar. 1980.
M
]Qj[f g
Notary Public /
My Commission Expires: July 1, 1982 UNITED STATEE OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
SACRAMENTO MUNICIPAL UTILITY
)
Docket No. 50-312 (SP)
DISTRICT
)
(Rancho Seco Nuclear Generating
)
Station)
)
AFFIDAVIT OF PHILIP R. MATTHEWS I, Philip R. Matthews, depose and state under oath as follows:
1.
I am a Section Leader in the Nuclear Regulatory Commission Staff's Auxiliary Systems Branch.
I am currently responsible for the review and evaluation of nuclear M',it auxiliary systems. My professional qualifications are attached to the NRC Staff response to California Energy Commission Interrogatory 17 filed in this proceeding.
2.
The answers to the Friends of the Earth Interrogatories 48, 49, 50, and 51 were prepared by me or under my supervision.
I hereby certify that the answers given by me are true and accurate to the best of my knowledge.
F Philip R. Matthews Subscribed and sworn to before me this,w71. day of January, 1980.
../
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/ Notary Public My Comission Expires:
July 1, 1982 77
75 -
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
SACRAMENTO MUNICIPAL UTILITY DISTRICT )
Docket No. 50-312 (SP)
)
(Rancho Seco Nuclear Generating
)
Station)
)
AFFIDAVIT OF DANIEL J. GARNER I, Daniel J. Garner depose and say under oath as follows:
1.
I am a Project Manager in the Nuclear Regulatory Commission Staff's Operating Reactors Branch 4.
I am responsible for the overall coordina-tion of licensing actions as they apply to the operating license of the
, Rancho Seco Nuclear Generating Station.
My professional qualifications are attached to the NRC Staff response to California Eneray Commission's First Set of Interrogatories.
2.
The answers to Friends of the Earth Interrogatories 4, 5, 12, 13, 16, 24, 25, 31, 32, and 43 were prepared by ne.
I hereby certifv that the answers given are true and accurate to the best of my knowledge.
~
s ja LPCL d Daniel
. Garner Subscribed and sworn to before me this 23rd day of January, 1980.
L j' l }lI.
'I f-Notary Public 183 178 My Commission expires: July 1, 1932.
76 -
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
SACRAMENTO MUNICIPAL UTILITY
)
Docket No. 50-312 (SP)
DISTRICT
)
)
(Rantno Seco Nuclear Generating
)
Station)
)
AFFIDAVIT OF THOMAS M. NOVAK I, Thomas M. Novak, depose and state under oatn us follows:
1.
I am a branch chief in the Nuclear Regulatory Commission Staff's Division of Systems Safety. The branch that I supervise is responsible for the review and <. valuation of a variety of safety systems in addition
- to the review and evaluation of a large number of transients and accidents described in Chapter 15 of applicants' Safety Analysis Reports. Currently, I am assigned to the Staff's Bulletins and Orders Task Force serving as Deputy Director. My professional qualifications are attached to 6.le NRC Staff response to California Energy Comission Interrogatory 17 filed in this proceeding.
2.
The response to Friends of the Earth Interrogatory 52 was prepared by me.
I hereby certify that the answers given are true and accurate to the best of my knowledge.
m Thomas M. Novak Subscribed and sworn to before me this A4tday of January,1980.
& /
.._ i. 1 g "18d 179 kotal y Public My Commission Expires: July 1, 1982
UNITED STATES OF AMERICA N'J C L Ei.T I'E T!_ T.T R Y CE ' :::IU PErnne Tur f.Tn9]c FrcrTv r"9 L!rca:S P'c Penop In the Matter of
)
)
SACRAMENTO MUNICIPAL UTILITY
)
Docket No. 50-312 DISTRICT
)
)
Rancho Seco Nuclear Generating
)
Station
)
h N fh 0
D CERTIFICATE OF SERVICE U
'J UbL L A u d _,
I hereby certify that copies of "NRC STAFF RESPONSE TO F0E DISCOVERY REQUESTS AND MOTION FOR A PROTECTIVE ORDER" in the above-captioned proceeding have been served on the following by deposit in the United States mail, first class, or, as indicated by an asterisk, through deposit in the Nuclear Regulatory Commission's internal mail system, this 24th day of January, 1980:
- Elizabeth S. Bowers, Esq., Chairman Atomic Safety and Licensing Board Panel Gary Hursh, Esq.
U.S. Nuclear Regulatory Commission 520 Capitol Mall Washington, D.C.
20555 Suite 700 Sacramenta, California 95814
- Dr. Richard F. Cole Atomic Safety and Licensing Board Panel Mr. Richard D. Castro U.S. Nuclear Regulatory Commission 2231 K Street Washington, D.C.
20555 Sacramento, California 95816
- Mr. Frederick J. Shon James S. Reed, Esq.
Atomic Safety and Licensing Board Panel Michael H. Remy, Esq.
U.S. Nuclear Regulatory Commission Reed, Samuel & Remy Washington, D.C.
20555 717 K Street, Suite 405
~
Sacramento, Califorriia 95814 David S. Kaplan, Esq.
General Counsel Christopher Ellison, Esq.
Sacramento Municipal Utility District Dian Grueneich, Esq.
P. O.
Box 15830 California Energy Commission Sacramento, California 95813 1111 Howe Avenue Sacramento, California 95825 l82[l80
if
[
~
~
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- Atomic Safety and Licensing Mr. Michael R. Eaton 5 :rd Panel E...cgy I::::: C::cdin:ttr U.S. Nuclear Regulatory Commission Sierra Club Legislative Office Washington, D.C.
20555 1107 9 Street, Roo,1020 Sacramento, California 95814
- Atomic Safety and Licensing Appeal Board Panel Thomas A. Baxter, Esq.
U.S. Nuclear Regulatory Commission Shaw, Pittman, Potts & Trowbridge Washington, D.C.
20555 1800 M Street, N.W.
- Docketing and Service Section Office of the Secretary U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Herbert H. Brown, Esq.
Lawrence Coe Lanpher, Esq.
Hill, Christopher and Phillips, P.C.
1900 M Street, N.W.
Washington, D.C.
20036 (Ck kichard K. Hoefling
/
CounselforNRCStaffh 185[l81
'