ML20054D149
| ML20054D149 | |
| Person / Time | |
|---|---|
| Site: | Millstone, Hatch, Peach Bottom, Oconee, Palisades, Indian Point, Arkansas Nuclear, Brunswick, Turkey Point, Crystal River, Zion, 05000000, Fort Saint Vrain, Crane |
| Issue date: | 11/12/1981 |
| From: | Felton J NRC OFFICE OF ADMINISTRATION (ADM) |
| To: | Sholly S UNION OF CONCERNED SCIENTISTS |
| Shared Package | |
| ML19239A281 | List: |
| References | |
| FOIA-81-405, RTR-NUREG-0578, RTR-NUREG-0660, RTR-NUREG-0737, RTR-NUREG-578, RTR-NUREG-660, RTR-NUREG-737 NUDOCS 8204220385 | |
| Download: ML20054D149 (6) | |
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- Mr. Steven C. Sholly 8D
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1384 Massachusetts Avenue TO F01A-81-405 ' )M(pc(' '
IN RESPONSE REFE Cambridge, MA 02238
Dear Mr. Sholly:
This is in response to your letter dated October 7,1981 in which you requested, pursuant to the Freedom of Information Act, seven categories of information.
Our response to each category is noted below.
1.
A copy of the requested audit report by the Office of Inspect 0c and Auditor (0IA), as identified on the enclosed appendix, is being placed in the NRC Public Document Room, 1717 H Street, N.W., Washington, D.C. 20555, for your inspection and copying.
This report will be filed in folder F0IA-81-405 under your name.
2.
A copy of the requested 01A report, identified on the enclosed appendix, is being placed in the PDR, as noted above.
3.
Copies of the requested memoranda, identified on the enclosed appendix, are being placed in the PDR, as noted above.
4.
During a telephone converation between you and Linda Robinson on October 27, 1981, you were informed that the requested 0IA memorandum has already been placed in the PDR under docket file 50-220 and you stated your intent to inspect the memorandum at the PDR.
5.
A copy of the requested 0IA report, identified on the enclosed appendix, is being placed in the PDR, as noted above.
6.
The TMI Action Plan Steering Group was a short-lived group which was formed in December 1979 to coordinate staff develop-ment of the TMI-2 Action Plan, which was developed through a series of successive drafts and then published in May 1980. Upon an extensive search of pertinent NRC central files by staff of the Office of Nuclear Reactor Regulation, the complete files of the steering group could not be located,but some miscellaneous pertinent records were found.
In addition, several individuals who served as members of the group and its staff retained some relevant records.
These records, identified i
on the enclosed appendix, are being placed in the PDR, as noted above.
In light of our inability to provide all requested records, Mr. Warren Minners, who served as a member of the i
820t'220385 811112 PDR FOIA SHOLLY81-405 PDR l
', Mr. Steven C. Sholly,
technical support staff to the steering group, called you on November 10, 1981 to discuss the activities of the steering group.
7.
On October 27, 1981, you informed Ms. Robinson that subsequent to submitting your October 7th request you located the requested report in the POR and withdrew your request for this report.
This completes NRC's action on your request. We wish to take this opportunity to thank you for your cooperation in agreeing to an extension of time for our response to your request.
Sincerely,
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J. M.Jelton, Director Divi ~sion of Rules and Records Office of Administration s
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Re:
F01A-81-405 Appendfx Category 1 1.
10/3/80 Memo to Com from Cummings, " Audit of the Implementation of NRC-Related TMI Lessons Learned Concerning Utility Personnel Licensing and Training", with attachment:
(a) 9/29/80 Memo to Cummings from Dircks, " Comments on Draft DIA Report ' Audit of the Implementation of flRR-Related Till Lessons Learned Concerning Utility Personnel Licensing and Training'."
Category 2 1.
11/12/80 Memo to Comm. from Cummings, " Inquiry Into NRR's Implementation of Short Term Lessons Learned Category A Requirements", w/ attachments:
(a) Report entitled " Inquiry Into NRR's Implementation of Short Term Lessons Learned Category A Requirements".
(b) 12/31/79 Letter to Denton from Dise, Niagara Mohawk Power Corporation with stated attachment.
(c) 1/2/80 Letter to Dise from Denton w/ stated enclosure: Order to Show Cause.
(d) 1/22/80 Answer to Show Cause, w/ stated exhibit.
(e) 3/21/80 Letter to Dise from Ippolito w/ stated enclosure, " Evaluation of Licensee's Compliance with Category 'A' Items of NRC Recommendations Resulting from TMI-2 Lessons Learned" dated 3/21/80.
(f) 1/10/80 Memo to Schwencer, et. al., from Eisenhut,
" January Site Visits for Lessons Learned Verification", w/ stated enclosures.
(g) 2/21/80 Memo to Eisenhut from Moseley, "IE Assistance nn TMI Lessons Learned Post Implementation Reviews".
(h) 3/7/80 Memo to Moseley from Eisenhut, " Completion of Lessons Learned".
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Re: F0!A-81-405 o
Appendix Cateogry 3 i
1.
1/26/81 Memo to Eisenhut from Moseley, "NUREG-0737, Implementation j
Plan for Operating Reactors Memorandum dated December 18, l
1980", with enclosed memo:
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12/29/80 Memo to Eisenhut from Moseley, "NUREG-0737 Implementation Plan for Operating Re3ctors:
OIE Coments".
2.
3/3/81 Memo to Eisenhut from Moseley, " Draft T1, TM1 Action Plan, IE Post-Implementation Review Items".
3.
9/21/81 Memo to Sniezek from Eisenhut, " Proposal for IE Review of Selected NUREG-0737 Items at Operating Reactors".
4.
9/21/81 Memo to Sniezek from Eisenhut, " Communication Responsibilities for TMI Action Plan Information", w/ stated enclosure.
Category 5 1.
9/18/80 Memo to Comm. from Cummings, " Improvements Needed in Coordinating the Development of Related Rules" w/ attachments:
(a) " Table of Tasks Listed in the TMI Action Plan Requiring RES Support".
(b) 8/14/80 Memo to Cummings from Dircks, " Response to Draft DIA Report: Improvemcnts Needed in Coordinating the Development of Related Rules".
Category 6 1.
12/11/79 Memo to Comm. from Denton, " Draft Action Plans for Implementing Recommendations of the President's Commission and Other Studies of THI-2 Accident".
2.
12/14/79 Memo to Denton, et. al., from Gossick, "TMI-2 Action Plan",
w/ stated enclosure.
3.
12/21/79 Memo to Task Managers from Mattson, " Update of NRC Draft Action Plan (NUREG-0660)".
4.
1/5/80 Memo to Comm. from Gossick, "TMI Action Plan - Prerequisites for Resumption of Licensing", w/ stated enclosures.
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Re:
F01A-81-405 Appendix 5.
1/15/80 Letter to Chairman Ahearne from Plesset, ACRS, " Draft NUREG-0660, ' Action Plans for Implementing Recommendations of the President's Commission and Other Studies of the TMI-2 Accident'".
6.
1/22/80 Memo to Mattson from Moore, "TMI Task Action Plans, Criteria for Automatic and Manual Actions".
7.
1/23/80 Memo to TMI Action Plan Task Managers from Mattson, " Future Activities (NUREG-0660)", w/ stated enclosure.
8.
2/12/80 Memo to Moore from Mattson, "TMI Task Action Plans - Criteria for Automatic and Manual Actions".
9.
2/22/80 Memo to Plesset from Mattson, " Inclusion of ACRS Recommendations in the NRC Action Plans".
10.
3/5/80 Memo to Fraley and Mattson, " Transmittal of Draft 3 of Action Plan".
11.
3/11/80 Letter to Chairman Ahearne from Plesset, "ACRS Report on Near-Tenn Operating License Items from Draft 3 of NUREG-0660, NRC Action Plans Developed as a Result of the TMI-2 Accident".
12.
3/12/80 Memo to Minogue, et. al., from Dircks, " Management Review of Draft 3 of TMI Action Plan", w/ stated enclosures, 13.
4/1/80 Letter to Chairman Plesset from Ahearne.
14.
4/1/80 Memo to Chairman Ahearne from Dircks, "ACRS Report on Near-Term Operating License Requirements".
- 15. 4/16/80 Memo to Denton, et. al., from Mattson, " Steering Group Review of Office Director Comments on TMI Action Plan" w/ stated enclosure.
16.
4/17/80 Letter to Chairman Ahearne from Plesset 17.
4/22/80 Memo to Dircks from Minogue, " Additional SD Input Concernino TMI Action Plan", w/ stated enclosures.
18.
4/23/80 Memo to Dircks from Denton, "NRR Concurrence in Three Mile Island Action Plan, w/ stated enclosure.
19.
5/5/80 Memo to Chairman Ahearne from Dircks, " Response to ACRS Letter of April 17, 1980 on NUREG-0660, 'NRC Action Plan Developed as a Result of the TMI-2 Accident,' Draft 3",
w/ stated enclosure.
Re:
F0!A-81-405 Appendix
- 20. 5/15/80 Memo to Comm. from Hanrahan, "THI-2 Action Plan (SECY-80-230)".
21.
6/10/80 Letter to Chairman Ahearne from Plesset, " Additional Information Concerning NT0L Items from Draft 3 of the NRC Action Plan".
22.
Undated Memo to Mattson from Milhoan, " Review of TMI-2 Action Plan (Draft NUREG-0660) for Incorporation of Recommendations of NUREG-0578, NUREG-0585, and SECY-79-330E", w/ stated attachments.
23.
Undated Memo to Mattson from Milhoan, "AIF Proposed Schedules for Action Plan Line Items Applicable i.o Licensees",
w/ stated attachments.
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DE, 31960 MEMORANDUM FOR:
Chaiman Ahearne i
Commissioner Gilinsi;y Commissioner Hendrie Commissioner Bradford 1
James J. Cummings, Director FROM:
Office of Inspector and Auditor AUDIT OF THE IMPLEMENTATION OF NRR-RELATED TMI LESSONS LEARNED CONCERNING UTILITY PERSONNEL SUSJECT:
LICENSING AND TRAINING We have completed the first phase of our audit of the utility personnel licensing and training.
4 Scope This phase of our audit included a review of the recommendations of the major THI studies and how they were brought together in the action plan.
We evaluated the process used to identify and collate recommendations from the major TMI studies, to cross reference and index all action items to the source documents, and to respond to and satisfy the intent To accomplish this, we defined, identified, of the recommendations.
collated and cataloged 108 recommendations concerning utility personnel training and licensing', and tracked them to the action plan, in its draft and final foms. 'We made a detailed analysis of a 20 percent sample of these recommendations (22 recommendations We also evaluated whether reflected the intent of the\\ recommendations.nces were adequate for the reader to the action plan's cross refe h action item and the original ' recommendations j
detemine the full intent of e each was implementing.
We did not judge the adequacy of he recommendations to ' solve the many Rather, we problems and uncertainties raised.by the.TMI accident.
concentrated on. how well the plan brought.together the various proposed recommendations into a coherent package fo management action.
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O Findin;s Based on our review we believe that:
the overall action plan was well done and satisfies the intent of the many recommendations of the THI studies; and
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The group responsible for putting together the action plan did a very commendable job.
Our review disclosed several concerns which dealt generally with assuring that all recommendations were accounted for, and improving the action plan references to ensure that the intent of the actual recommendations was clear to those who will be responsible for implementing the action items. We discussed our analysis and evaluation with Dr. Mattson, the chairman of the group responsible for the action plan.
For the most part, our concerns have been resolved through his clarifications of the items in question or through changes that he agreed to make in the action plan.
However, there were two recommendations which we believe were not adequately addressed:
The Commission's response to SECY-79-330E, Qualifications of Reactor i
Operators, directed the staff to increase the scope of reactor operator and senior reactor operator examinations and to devel@
and administer all annual requalification examinations, certification examinations and audits of training programs. The NRR staff had recommended that NRC administer about 10 percent of the requalification examinations and oral evaluations but the Commission decided that NRC should perfonn all of them in-house.
The action plan included the Commission's recommendations and estimated ari expenditure of 0.7 and 1.0 manyears for fiscal years 1980 and 1981 to be used for the development.of criteria and proposed rule changes that included the requalification examination program.
The plan stated that funds to implement and administer the requalification program would be included in the fiscal year 1982 budget. In their FY 1982 budget. NRR requested an increase in operator licensing' staffing from 26 to 37 manyears, however, none of these positions were for administering the requalification examinations. Alternatively.
NRR's budget request for program support for operator licensing increased from $170,000 in FY 1981 to $4 millien in FY 1982.
This increase was to administer the annual requalification examinations by way of contract.
In spite of NRR'.s FY 1982 budget request regarding requalification examinations, they are continuing to examine alternative methods for fulfilling the Commission's directive through a contract with Analysis and Technology, Inc. The results of that study should be presented to the Commission as soon as they become available.
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The Executive Director for Operations (EDD) comments on a draf t of this report stated that it was impractical to request an increase irl the NRC staff necessary to conduct all requalification examinations As a result, NRR requested staff increases to administer in-house.
replacement examinations and technical assistance money to contract Following completion of the out the requalification program.
Analysis and Technology, Inc. contract, however, NRR plans to reapportion the people and technical assistance money available The EDO between the replacement and requalification programs.
expects to make a recommendation to the Commission on the plans and.-
budget for the requalification program in early Spring 1981.
IE's Special Review Group recanmended that if a licensee does not meet NRC's criteria for testing the adequacy of emergency preparedness, The action plan does not address NRC should conduct its own test.
the conduct by NRC of its own test should the licensee fail to conduct an adequate test.
I In commeiting on a draft of this report the EDO stated that NRC's participation with the Federal Emergency Management Agency in reviews and evaluations of licensees' emergency exercises, and the i
fact that a licensee's failure to comply with Part 50 requirements l
for conducting drills and exercises could result in suspension or modification of his license, satisfied the IE Special Review Group's 2
recommendation.
Other Observations Based on our review to date of the actions related to utility personnel licensing and training, we have the following observations relating to the action plan as a whole that we feel are important to a better understanding of the action plan. -
Many recommendations of the THI studies are not addressed by action When the committee collated and consolidated recommendations items per se.
for the action plan they had to make many detenninations regarding g-which to address by action itens. The problem arose with recommendations g.
which were very sinilar, contradictory, provided two different solutions to the same problem, or were beyond the scope of NRC's We believe however that the recommendations from the authority.
major studies were adequately considered in the plan.
The committee's charter was unclear and changed considerably since r
it was established in November 1979, causing the expenditure of At first additional time and resources to complete the project.
the charter was simply to collate or catalog all the recommendations It later made by the various groups investigating the accident.
grew to include an evaluation of each recommendation as to its relative impact on :3fety, its requirement for NRC and licensee resources, and its immediacy. Despite the difficulties arising from this situation, the committee did a very good job in bringing the varicus shidics together into c =aningfol and managenle package, omcc).........................................'............................
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A clear, fim charter would h:ve allowed a more straightfontard For example, the development and presentation of the action plan.
crosscuts, which would assure that the major recommendations have been adequately covered, were done as an afterthought--after the plan was in its third revision--rather than as the action plan took form.
The plan started out as a shopping list with everyone putting in Plany were deleted based on input from the his favorite project.
various office directors or on the committee's own initiative, because they were not TMI related or the pros and cons of the Those decisions are not h
proposed project had been argued before.
always set forth clearly in the action plan although many are on P
the record of Commission meetings and meetings of the Advisory Committee on Reactor Safeguards.
To obtain the additional data The action plan is brief, by design.
necessary to determine the full intent of each action item and the recommendations that each is implementing, the reader and action official must use the references, crosscuts, and comparisons ofThey action items to recommendations provided in the action plan.
The introduction to the action are an integral part of the plan.
plan attempted to convey this message, but in our opinion, failed.
The first printing of the final report states that~"The references...had to be considered in the process of developing the requirements, This indicates that the studies and other actions in the plan."
references have already been considered and the reader need not go The. appropriate wording should have further than the action plan.
been that the references "must be considered," thus showing the It must reader that he must go back to the referenced documents.
be clearly understood that the reference material is a critical and integral part of the plan. Dr. Mattson told us this was an editorial error which was corrected in Revision 1 of the action plan. ' However, because an errata sheet with this correction was not issued for the first printing of the action plan, there are many copies of the action clan in circulation in which the correction has not been made.
There is feedback from industry that some remedial actions as set forth in the action plan are too prescriptive and that industry may An example heard l
have other " fixes" that are just as reasonable.
most often concerned inerting containment. Generally speaking, both NRC and industry agree that there is a problem, but industry believes NRC should set general requirements and let industry We believe that all proposed decide how to meet those requirements.
alternatives should be considered by the responsible office, and if
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it deems that an alternative is better than that set forth in the action plan, it should bring that alternative to the attention of The action plan should not be so inflexible as to the Commission.
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deny NRC the ability to adapt as new infonnation becomes available.
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The action plan is only the first step toward sat sfying'the i
The i:nplementation of reccrnmendations of the major THI studies.
the action items containec' in the plan is the key to NRC's response And, by implementation we mean more than making to the accident.
open-ended studies; we mean the action taken based on of those studies.
manage or monitor, and because of the significance of the issues involved, we will review that implementation during the second phase of our audit.
EDO Comments d The A copy of the ED0's comments on *
- aft of this report is attache.His specifi EDO generally agreed with the conte..cs of this report.which we believed were comments relating to THI study recommendatinnsare included in the report not adequately addressed in the Action Plan We have no objections to sections dealing with those recommendations.
the ED0's comments but believe the Commission should closely monitor NRR's proposals for meeting the Commission's directive relating to reactor operator requalification examinations.
We are continuing our review of actions being taken by NRR as a of the accident at Three Mile Island.
NRR's implementation of items included in the action plan relatedsto utility personnel licensing and training including an evaluation of the management structure within NRR to monitor and assure timely completion of the recommended corrective actions.
Attachment:
As Stated cc: W. Dircks, EDO '
H. Denton, NRR
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MEMORANDl!4 FOR: Jaues J. Cummings, Director
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FROM:
William J. Dircks, Executive Director for Operations
SUBJECT:
COMMENTS ON DRAFT OIA REPORT " AUDIT OF THE IMPLEMENTATIO!!
0F NRR-RELATED TMI LESSONS LEARNED CONCERNING UTILITY PERSONNEL LICENSING AND TRAINING" Your memorandum of September 5,1980 requested comments on the subject draf t 01A report.
He are in general agreement with the facts and findings of the report, with the following comments.
The report notes two recommendations arising out of the various TMI studies that DIA believes were not adequately addressed in the Action Plan.
The first of these concerns the operator requalification examination pro-gram. As the OIA report acknowledges, NRR has a study undenvay with Analysis and Technology, Inc., to consider guidelines for requalification examinations and alternative methods for fulfilling the Commission 's directive to administer all annual requalification examinations.
Since the FY 1982 budget package had to be prepared before any results were available from this study, we estimated the resources that would be required to administer requalification examinations added up to approximately fifty man-years per year.
These estimates were based on administering the requali-fication examinations in a manner s:nyilar to the procedures used in administering the replacement exams.1.
Since it would be impractical to think we could increase the NRC staff by that amount for budgeting purposes, we used professional man-years for the replacement examination program and allocated money to contract out the implementation of the requalification examination program.
Rather than use all of the money and none of the people in the FY 82 for requalification examinations, we cxpect to reapportion money and people available after the planning has progressed further. The resources requested include an increase
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of eleven professionals for 1982; this seemed, based on experience, to be the maximum number we could expect to recruit in one year.
We anticipate that we will make a recommendation to the Commission on plans and budget for this program in early spring of 1981.
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Lf) #7v 1/ Replacement examinations are those that we administer routinely at licensee facilities in order to equalize the number of licensed operators availabic on-site due to attrition and promotions.
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The second recomnendation noted in the OIA report. appears to. refer to Section 3.13.8 of the IE Special Review Group report.
The recommendation states that provisions should be made to allow NRC to conduct tests of emergency prepared-ness capability if the licensee fails to conduct such a test.
The recent revisions to Part 50 require licensees to conduct various drills and exer-cises.
Failure to comply with the regulations could result in suspension or modification of the facility license.
In addition, NRC will observe and critique certain of the exercises jointly with the Fede,al Emergency Management Agency.
In the sense that NRC participates in the evaluation of an exercise, we believe that the cited recommendation is met.
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- 4 November 12, 1980 7
MEMORANDUM'FOR:
Cheirman Ahearne Commissioner Gilinsky a
Commissioner Hendrie 4fgy%1 sig d Commissioner Bradford
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Jares J. Cur:raings, Director Office of Inspector and Auditor SUSJECT:
INQUIRY INTO NRR'S IMPLEMENTATION OF SHORT TERM LESSONS LEARNEDCATEGORYAREgUIREMENTS
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In October 1980 an Office of Inspection and Enforcement (IE) Health Physics appraisal team discovered that the Short Tem Lesson Learned-Category A requirement relating to increased range of effluent monitors had not been implemented at the Nine Mile Point Nuclear Station. The licensee's implementation of Category A requirements had been reviewed by an Office of Nuclear Reactor Regulation (NRR) review team'which had visited the site in March 1980 and had concluded that the licensee had satisfactorily met s11 Category 'A requirements. As a result of the Heaith Physics appraisal team's finding, the question arose generally as to what had actually been done by the NRR review teams which visited reactor sites to review these new requirements.
Attached is our report on the results of our inquiry. Th'escopeofour inquiry was limited to the. reasons NRR established the review teams, -
what they did at the reactor sites', and the'IE responsibilities for Y-
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verifying licensee compliance with the Category A requirements. We did
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criminal referral.--Additionally, he intend to follow up on IE findingsMM-F ei with respect to.their verification of licensee'Fimplem
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INQUIRY INTO NRR'S IMPLEMENTATION OF SHORT TERM LESSONS LEARNED CATEGORY A REQUIREf1ENTS I. INTRODUCTION In early October 1980 a Region I Health Physics (HP) appraisal team conducted a radiological assessment of the plant and equiptrent at the Nine Mile Point Nuclear Station, Unit 1, owned and operated by the Niagara Mohawk Power Corporation (NMPC).
In the course of their inspection, the HP appraisal team found that the licensee's implementation of one of the NUREG-0578 -- TMI-2 Lessons Learned Task Force Status Report and Short Tem Lessons Learned -- Category A items dealing with high range effluent monitoring, was inconsistent with the written commitments previously made by NMPC to the U.S.
Nuclear Regulatory Commission (NRC). NMPC had submitted documentation to NRC in November and December 1979 and January 1980, indicating that with one exception, all of the Category A items of NUREG-0578 had been implemented. As a result of the HP appraisal team's finding, the question arose as to whether NMPC had deliberately made false statements to NRC regarding their implementation of the NUREG-0578 reauirements.
The Office of Inspection and Enforcement (IE) was aware that early in 1980 the Office of Nuclear Reactor Regulation (NRR) had sent a review team to Nine Mile Point to review implementation of the Category A requirements.
In view of the finding at Nine Mile Point, the Office of Inspector and Auditor (OIA) was requested by the Director, IE, to detemine the charter and actual perfomance of the review teams established by NRR to perfom the NUREG-0578 implementation reviews so that those facts could be considered in any enforcement action taken by NRC.
Subsequently, Chaiman Ahearne asked that we deterrine what the understanding was within IE and each of the regional offices with respect to follow-up responsibility for the NUREG-0578 requirements.
In order to make those deteminations, it was first necessary to review the sequence of events and exchange of correspondence specifically relating to NRR's review of the Category A requirements at Nine Mile Point.
Following are the details of those events.
II. DETAILS OF NINE MILE POINT IMPLEMENTATION OF CATEGORY A REQUIREMENTS In July 1979 NRR issued its TMI-2 Lessons Learned Task Force Status Report and Short Tenn Recommendations (NUREG-0578).
This report identified a number of actions that should be taken in the short tem to reduce the likelihood of a nuclear accident and improve emergency preparedness in responding to such events.
In a September 13,
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2 1979, letter to all operating nuclear power plants, NRR directed all licensees to implement the requirements set forth in that report.
For a certain set of these requirements -- those labeled Category A -- implementation was to be completed by January 1, 1980, or prior to issuance of the operating license for new plants.
This letter was followed by regional meetings in late September 1979 held by NRR with licensees to further explain and encourage industry feedback on each of the short term requirements.
As a result of the feedback obtained through these meetings, NRR issued another letter to all licensees on October 30, 1979, which further clarified and provided an implementation schedule for the staff requirements.
It was thus established that NRR's review of licensees' implementation of the short tenn requirements would be based on NUREG-0578, coupled with NRR's subsequent clarification letters, dated September 13, 1979, and October 30, 1979.
In view of the requirement for licensees to comply with all Category A items by January 1,1980, NMPC, the licensed owner of Nine Mile Point, Unit 1, sent letters to the Director of HRR on November 19, December 19, December 20 and December 31, 1979, to document their compliance with certain NUREG-0578 requirements and to provide design details and the status of a number of outstanding commitments at Nine Mile Point.
For one particular requirement, item 2.1.8.b. --
Increased Range of Radiation Monitors -- the December 31 letter from Nf1PC advised NRR of provisional steps taken by January 1, 1980, and indicated final compliance would be accomplished by January 1,1981 (see Attachment 1).
On January 2,1980, NRR issued an Order '.o Show Cause to various licensees, including Nf1PC, who~ had indicated that they did not intend to implement some of the Category A requirements until after January 31, 1980 (see Attachment 2).
NRR stated its position that implementation of the Category A requirements by January 31, 1980, was necessary to provide continued assurance of public health and safety.
The Show Cause Order required these licensees to implement the requirements by January 31, 1980, or show cause why they should not. The Order also required implementation of the requirements by that date or the plant to be shut down until such implementation is complete.
In addition, the Order stated that "The licensee may file a written answer to this Order under oath or affirmation within twenty (20) days of the date of the Order."
NMPC responded to NRR's Show Cause Order by letter dated January 22, 1980, with a notarized statement, signed by the Executive Vice President, which indicated the date of compliance for each Category A requirement (seeAttachment3).
According to Nf1PC's statement, only one of the Category A requirements -- item 2.1.3.a., Direct Indication of Value Position -- would not be completed by January 31, 1980, due to pending delivery of necessary equipment.
All other Category A items were designated as not applicable to Nine Mile Point Unit 1 or as completed as of a given date. For item 2.1.8.b.,
implementation was reported to be completed on December 31, 1979.
/
t 3
The NRR review team looking at General Electric (GE) plants performed a site visit at Nine Mile Point on March 12, 1980.
The meeting was attended by representatives of idPC corporate and plant management, the NRR review team, the NRR Project Manager, and the IE resident inspector assigned to Nine Mile Point.
The results of the implementa-tion review, reflected in an evaluation report dated March 21, 1980, were provided to NRR management, IE Headquarters and NMPC (see ). The cover letter to the evaluation report concluded that, based on NMPC's submitted documentation and the discussions between the two staffs at the March 12, 1980, site visit, NMPC had satisfactorily met all Category A requirements.
The letter included a statement that "The adequacy of certain implemented procedures and modifications will be verified by our Office of Inspection and Enforcement.
Each of these items is discussed in our evaluation."
However, for item 2.1.8.b., the NRR review team stated in the evaluation report that " Based on our review, we conclude that the licensee has satisfied the Category A requirements of this item."
No statement was included with respect to IE verification of that item.
In early October 1980 a Region I HP appraisal team conducted a radiological assessment of the plant and equipment at Nine Mile Point.
This appraisal covered a two week period, ending October 10, 1980.
Although the charter of the HP appraisal team did not require verification of the NUREG-0578 Category A requirements, the team leader decided to inspect the licensee's implementation of the high range monitor requirement (item 2.1.8.b.).
The HP appraisal team found that NMPC's actions on item 2.1.8.b.
did not satisfy the intent of that requirement and were incon-sistent with the commitments made in their written submittals to NRC.
Specifically, fNPC's actions were unsatisfactory because:
1.
Instead of installing a " lead cave" around the radiation detector as NMPC had committed to do, they had strapped a lead plate in front of the detector. This would not provide shielding to the detector as the cave was intended to.
2.
The calibration of the radiation detectors did not satisfy the l
intent of the NUREG-0578 Category A requirement.
3.
Required procedures had not been developed.
4.
Although conversion factors had been calculated they did not satisfy the intent of the NUREG-0578 Category A requirement.
NMPC was apprised of these deficiencies at the conclusion of the HP appraisal team's visit. This conversation with NMPC took place without the HP appraisal team being aware of the January 2,1980, Show Cause Order or fNPC's response. On discussing the problems with the NRR Project Manager for Nine Mile Point and on learning of
?
e 4
NRR's Show Cause Order of January 2,1980, regarding the Category A requirements, the Region I Section Chief in charge of the HP team elevated the matter to higher Region I management. The Section Chief also communicated IE's concerns to HMPC management.
On October 17, 1980, IE issued an.Immediate Action Letter to NMPC to confim the commitments made orally by the Vice President for Nuclear Generation, NMPC, that the deficiencies noted by the HP appraisal team were being corrected.
III. SCOPE OF 01A'S REVIEW In pursuing the specific objectives of our review, we met with various representatives of NRR and IE management and staff.
From NRR, these included the Director, Division of Licensing (DOL),
fomerly the Acting Director, Division of Operating Reactors (DOR);
the Chaiman of the Lessons Learned Steering Committee; the team leaders of three of the four NRR review teams and selected members of these teams; and the project manager assigned to Nine Mile Point. Although we did not talk to the fourth NRR team leader, who is no longer employed by NRC, we did meet with two of the members of that team.
In IE, we met with the Assistant Director for Field Coordination, Division of Reactor Operations Inspection, and the designated IE representative on the NRR Lessons Learned Steering Committee.
In addition, we telephonically contacted the Chief, Reactor Operati' :
and Nuclear Support Branch, in each of the regional offices and the Section Chief in charge of the Region I HP appraisal team.
Finally, we reviewed pertinent NRR and IE correspondence relating to the NUREG-0578 requirements, the respective NRR and IE responsi-bilities for review and follow-up, and specific documents relating to the Nine Mile Point review.
We did not talk to any representatives of NMPC corporate or plant management.
Following are the results of our review.
IV. DETAILS OF REVIEW NRR Review Teams The review of the Category A requirements was given a high priority by NRR management because of their potential for further ensuring public health and safety.
As a result, it was determined that all Category A evaluation reports should be completed by April 15, 1980. The NRR review of licensees' actions on Category A require-ments was to be a post' implementation review. That is, the licensees
5 were to implement the Category A requirements before their approaches were reviewed and approved by NRR.
In September 1979 a Steering Committee on Lessons Learned was fonned to oversee the implementation of the NUREG-0578 requirements and to develop an approach which would enable completion of all reviews within the scheduled time frame.
Various approaches were considered for the implementation reviews of the Category A require-ments.
The aim was to complete implementation of the Category A requirements by January 1,1980, but to utilize a review approach which could continue to function after that time since some NUREG-0578 items (Category B) would not be implemented or reviewed until 1981. The approach finally recommended by the Steering Committee and approved by NRR was to fann four multi-disciplinary review teams which would review documentation submitted by the licensees and hold discussions with the licensees at the plant sites.
It was believed that the implementation reviews could be accomplished much faster using this approach rather than the traditional in-house review.
The Lessons Learned Steering Committee was charged with providing guidance to the review teams to ensure consistency in their review approaches and providing specificity for each requirement through documentation, meetings and infonnal discussions.
The Steering Committee was composed of a Chairman, the four review team leaders, an IE representative and eight other representatives from various NRR organizations.
The Steering Committee established a separate review team for each of the four major reactor types.
Each review team had a team leaaer and three or four members with specialties in technical areas, e.g., plant systems, electrical systems, containment, and radiological protection.
The NRR project manager for each operating reactor was also a review team member for the site visit to his plant.
Because this was a new approach to licensing reviews, the Acting Director, D0R, NRR, sent a memorandum on January 10, 1980, to the D0R Branch Chiefs, the IE representative on the Steering Committee and the Fort St. Vrain Project Manager to establish the schedule for review team site visits and to set out the proposed NRR and IE review effort during the site visits (,see Attachment 5).
The memo l
called for specific items to be reviewed by IE, using the acceptance l
criteria contained in NRR's October 30, 1979, clarification letter for the Category A requirements.
IE representatives told us, however, that at the time IE did not believe these criteria were clear or well enough defined to pennit IE to inspect the adequacy of licensees' actions in meeting the Category A requirements.
Therefore, in a February 21, 1980, response to NRR (see Attachment 6), the Director, Division of Reactor Operations Inspections, IE, established a more limited role for IE, which included the following:
6 "1.
00R provide IE w!th one week advance notification where IE participation in meetings or site visits is desired.
'2.
Prior to each meeting or site visit the NRR team leader or Project Manager discuss the review plan with the appropriate regional section leader and inspectors.
'3.
At the conclusion of the meeting or site visit the NRR team leader provide a list of outstanding items to both the licensee and the IE inspector and review with the inspector those items where IE followup is requested.
'4.
NRR identify in SER [ Safety Evaluation Report] those items for IE followup.
'5.
The inspector document the results of his follow up review in an inspection report and transmit the applicable portions to D.Eisenhut, Acting Director, DOR."
The Acting Director, DOR, NRR, responded to IE's position by memo-randum dated March 7,1980, stating that "Our plans for completing the short tem Lessons Learned by April 15, 1980, are consistent with your understanding provided in your February 21, 1980, memo-randum."
(See Attachment 7.)
The NRR review teams began to make site visits in mid January 1980 and continued through March 1980. The initial plan was to visit all plant sites; however, for various reasons (time constraints and other high priority activities competing for NRR staff resources as well as the desire to minimize the impact on licensees) management detemined that for some licensees a review of submitted commitments and documentation of actions taken, in conjunction with discussions with the licensee either telephonically or at NRC Headquarters, and discussions with IE resident inspectors would provide sufficient infomation to complete the reviews.
At the completion of each site visit the review teams prepared an evaluation report to document the adequacy of licensees' stated actions. It is important to note that these evaltation reports were not the conventional SERs nomally prepared for licensing reviews.
A team leader told us that because the reports did not support a licensing action and did not contain a. finding regarding significant hazards, they could not be SERs. The evaluation reports were sent to the licensees and given varied distribution to IE by the various teams.
The Chaiman of the Steering Committee told us, however, that the reports were supposed to be sent to IE Headquarters for further distribution to the regions.
The letters transmitting the evaluation reports to the licensees stated, in various tems depending on the style of the individual f
t 7
review team, whether the licensee had satisfactorily met all Category A requirements and that IE would verify the adequacy of certain f
. implemented procedures or modifications.
The letters further stated that each item requiring IE follow-up was discussed in the
' " ~ ~ ~ ~
evaluation reports.
The fonnat and content of the evaluation reports for all four teams was basically the same, with each Category A requirement being separately addressed. The evaluation reports stated in general terms what the licensee was doing to comply with each Category A 3
requirement and whether that would satisfy the NRR requirements.
IE follow-up requirements were identified in many, but not all, of the individual report sections. For example, the evaluation report for Nine Mile Point did not identify any IE follow-up for item 2.1.8.b., Increased Range of Radiation Monitors.
The Category A reviews were completed in March 1980.
A total of 46 evaluation reports were prepared by the review teams.
The actual charter of these review teams has proven to be a matter of some confusion and controversy. There is general agreepent, i
however, that the NUREG-0578 Category A requirements could be l
reviewed considerably faster by the review teams at the sites than I
the conventional in-house review approach would permit.
This approach was chosen because it was thought to be the only way all i
Category A evaluation reports could be completed by April 15, 1980, r
as directed by NRR management. We discussed the charter and functioning of the review teams with various NRR personnel who were involved in the program at that time.
The Director, DOL, NRR (who was the Acting Director, DOR, prior to the NRR reorganization), told us the purpose of the review teams was to expedite NRR's review of licensee submittals.
He said that although some of the review teams actually verified that licensees had taken the actions they were committed to, the teams did not go to the sites for that purpose.
The evaluation reports resulting from the site visits were to document the licensees' approach for implementing the required actions; they were not intended to be inspection reports. The Director further stated that IE had responsi-bility for verifying during routine IE. inspections that the licensees had actually implemented their commitments.
The Chairman of the Lessons Learned Steering Committee. agreed with the Director, D0L.
He said the purpose of the site visits was to facilitate the NRR review of licensees' commitments and ensure that l
the licensees understood what the NUREG-0578 requirements were. He said he'had specifically directed the review teams not to verify that licensees had in fact implemented what they said they were going to do. He had emphasized to the review team leaders that IE was responsible for conducting inspections to assure -that licensees had fulfilled their canmitments and that these inspections would be r----
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,,,.,.,. - - - -, ~,.,,,, _ -,. -
8 done as part of IE's routine inspection program.
He also told us that IE was specifically going to follow up on items identified in the review teams' evaluation reports, however, this was in addition to IE's routine inspection responsibilities.
He said that the IE representative working with the Steering Committee at that time understood and agreed with this position.
D0R management's intentions in sending the review teams to the reactor sites was apparently not clearly communicated to the review team leaders, however, since there were different understandings among the review teams as to what they thought they were supposed to do.
The leader of the review team that visited sites with 'GE reactors, including Nine Mile Point, told us the review teams were not supposed to do any verification at the sites.
He said site visits were made to expedite NRR's review of licensees' commitments to meet NUREG-0578 Category A requirements, and by going to the sites the traditional exchange of letters between NRR and the licensees could be eliminated.
Although plant tours were made by this review team at every site visited and some verification was accomplished during the course of these tours, these verifications were not referenced in the team's evaluation reports.
The verbal and written statements made by the licensees were accepted as the truth. This team leader told us he understood IE follow-up of licensee commitments would be based on a Temporary Instruction (TI) to be issued by IE Headquarters.
It was not his understanding that the resident inspectors would follow up on licensee actions based on the review team's evaluation reports.
As a result of this understanding this team leader distributed copies of his evaluation reports only to IE Headquarters, not to the regions or the resident inspectors.
~
The leader of the review team that visited Westinghouse reactors told us the purpose of the visits was to review the documents submitted by the licensees and to verify, to the extent possible, that the licensees had actually implemented the commitments they had made.
He said he was not directed to verify licensee actions but did so based on his own understanding of what he was supposed to do.
He understood IE would follow up on items at the sites based on the review team's identification of items in the evaluation reports.
He did not believe that IE had a general responsibility to verify licensee implementation of all Category A requirements.
This team's evaluation reports were initially sent to IE Headquarters but were later sent directly to the regions and resident inspectors.
Two team members on the teain responsible for Babcock and Wilcox reactors said the purpose of the visits was to expedite NRR's paperwork review of licensee submittals and to verify that the licensee had actually taken the actions, especially installation of equipment, if it was possible to do so.
In addition, if, during the site meetings, the i
l
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9 I
licensee was evasive in responding to NRR's questions or the licensee's i
documentation was not complete, the team would verify the licensee's actions.
They said that at the completion of each site visit a list of items for IE follow-up was provided to the IE inspector.
They understood IE would specifically follow up on the items on this list rather than on the team's evaluation reports, although all items identified in the evaluation reports for IE follow-up were included in the lists provided to inspectors. In some cases, however, there were additional items on the lists which were not i
included in the evaluation reports.
They also told us that IE continued to have a general responsibility for inspecting licensees' compliance with all Cate' gory A requirements.
The fourth team leader, who was responsible for visiting Combustion Engineering reactor sites, also told us the site visits were for the purpose of expediting NRR's paperwork review.
He said that some verification of licensee actions was perfonned at the sites but that this verification was not to the extent nonnally done by an IE inspector.
He saw this review team verification as a side benefit of visiting the sites and not the purpose for making the visits.
He said he had not been directed to verify licensee actions but he did so based on his perception of what the team should do while on site.
He understood IE would follow up on licensee actions as directed by the review teams fa the evaluation reports, but he was not aware of any general responsibility IE had to verify all Category A items.
Distribution of evaluation reports by this team was handled as part of NRR's normal distribution of documents to IE.
In some cases this team leader sent copies of the evaluation 1
reports to the inspectors at their request. Generally, however, he thought that IE Headquarters had routinely sent copies of the evaluation reports to the regions.
In summary, it appears that the NRR review teams did not have a clear, consistent understanding of what they were supposed to do at the reactor sites. While NRR management and the Steering Committee Chairman did not intend for the review teams to verify licensee actions, the extent of the various teams' reviews during the site visits seemed to be largely based on the team leader's perception of what could adequately be accomplished, given the time they were going to be on site, and what should be cone, given their under-standing of issues such as the traditional division of responsibilities between NRR and IE.
Every team leader understood that some IE follow-up of licensee actions would be. required although there are some differences in these understandings as to the basis on which l
the inspectors would initiate follow-up reviews.
Inspection and Enforcement Follow Up i
Traditionally, IE is responsible for verifying licensee compliance with NRC requirements as approved by NRR.
When the short tenn Lessons Learned (STLL) requirements were issued it was anticipated by NRR t
. _ _ _ _ _.. ____ - _ _ _ _ _ _ _ __ _ ~ _,
10 that these traditional roles would be retained even for the Category A items, which required post implementation reviews.
This was reflected in the January 10, 1980, memo from the Acting Director,
~
DOR, discussed above.
Based on our discussions with the Assistant Director for Field Co-ordination, Division of Reactor Operations Inspections, IE, the IE representative on the Steering Committee and the IE regional offices, there did not appear to be a clear understanding of what IE's role would be in verifying licensee implementation of Category A require-ments.
The Assistant Director for Field Coordination initially told us that IE's role was to follow up on items identified for IE follow-up in the review teams' evaluation reports.
He said IE did not have a general responsibility to inspect licensee implementation of all Category A requirements.
During a later discussion, however, he told us IE recognized their general responsibility to inspect implementation of all Action Plan items, including the Category A requirements, and, in fact, issued a series of tis on October 1, 1980, to accomplish this.
The IE representative on the Steering Committee told us that while IE was going to follow up on specific items identified in the review teams' evaluation reports, they also retained their nonnal responsibility to inspect all licensee actions required by NRC, including the Category A requirements.
He also told us that he was supposed to receive copies of all evaluation reports from NRR for distribution to the regional offices, but he only received copies of the 'first few. Af ter that, he assumed that the regional offices would get copies through the normal distribution system.
To determine the IE regional offices' perceptions of their responsi-bilities for following up on NUREG-0578 requirements, we spoke to the Reactor Operations and Nuclear Support Branch Chief in each of NRC's five regions. The responses from each of the five regions were generally consistent with the February 21, 1980, memorandum fron the Director, Division of Reactor Operations Inspections, IE, to NRR.
That is, each region understood that they would be asked to follow up on items identified by the NRR review teams in their evaluation reports prepared at the conclusion of the site visits.
However, there was some variation in the details of what the regions actually did.
One region told us they understood they were to follow up on items identified in the review teams' SERs but they were not aware that any SERs had ever been issued.
This region did not even recall that evaluation reports had been issued, although they said they would not have followed up on items identified in evaluation reports anyway because they had been specifically directed to follow up on l
1 11 SERs.
The other four regions did follow up on items identified for IE follow-up in the review teams' evaluation reports.
Two regions said that in addition to following up on the specific items identified in the evaluation reports, they in fact did much more verification on their own initiative. These two regions said they had a general responsibility for inspecting what was being done by the licensees in their regions. They therefore inspected much of what the licensees were doing to implement NUREG-0578 requirements even before the NRR review teams made their site visits. The other three regions did not believe they had a general responsibility to inspect implementation of all Category A require-ments.
Problems experienced by the regions appear to have resulted primarily from the differences among the review teams' approaches.
Each review team utilized the resident inspectors differently during the site visits and various understandings seemed to have been reached with inspectors by the review teams regarding follow up requirements.
Several regions also commented that they did not receive copies of all the evaluation reports for reactors in their regions, although they thought the resident inspectors had had better luck in this regard.
In summary, there did not appear to be a consistent understanding between IE Headquarters and the regions regarding IE's role in verifying NUREG-0578 requirements.
The greatest confusion within the regional offices seems to have been caused by NRR's issuance of evaluation reports rather than SERs and by the variations among the review teams' approaches.
As a result of the finding at Nine Mile Point, on October 24, 1980, the Assistant Director for Field Coordination, Division of Reactor Operations Inspection, IE, sent a memo to all regional directors requiring the immediate verification of all licensees' implementation of Category A items. This action was to be completed by November 5, 1980.
Conclusion The Nine Mile Point scenario underscores the fact that " coordinating" interoffice projects does not, of and by itself, guarantee (a) effective coordination or (b) that fundamental issues will be addressed and resolved. The critical issue of fixing individual office responsibilities for ensuring licensee compliance with the STLL was discussed to some extent, and letters were exchanged, but the issue was then put aside without appropriate resolution. No one within NRC took full responsibility for ensuring that licensees were in compliance with the STLL by January 31, 1980 - in fact, nine months later that work is still going on.
12 Rather than focusing on who is to blame within NRC for this situation we believe it is more important to determine whether the problems have been resolved. We think they have not.
For example, during a
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Janua ry 29, 1980, IE/NRR interface meeting it was agreed that a document should be developed delineating NRR and IE contacts and responsibilities for verification of licensee actions for items in the NRC Action Plan.
This document has never been developed.
This becomes critical when it is understood that NRC is currently in the process of implementing the Action Plan and that many items in the Action Plan require a post implementation review, as did NUREG-0578 for Category A items.
In addition, there were indications within NRR during our review that the site visits by the review teams were very beneficial for the HRR representatives that made the visits and that this technique may be used again in the future.
If the problems within NRC, surfaced as a result of the occurrence at Nine Mile Point, are not corrected, it is likely that the confusion that existed during implementation of the STLL Ca.tegory A items will be experienced again as the remainder of the Action Plan is implemented.
9
ffYMIAGARA IMUMOHAWK NIAGARA MOHAWK POWER CORPORATION /300 ER.E BOULEVARD WEST. SYRACUSE N Y.13 02/ TELEPHONE (31N C u_.
December 31, 1979 Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. NJclear Regulatory Comission Washington, D. C. 20555
Dear Mr. Denton:
Re: Nine Mile Point Unit 1 Dod<et No. 50-220 DPR-63 Our letters dated November 26, 1979, December 19, 1979 and December 20, 1979 document our compliance with NUEG-0578 Recommendations 2.1.1, 2.1.2, 2.1.3. a, 2.1.5. c, 2.1.7.a, 2.1.7.b, 2.1.9, 2.2.1.a, 2.2.1.b, 2.2.1. c, 2.2.2. a, and 2.2.2.c.
Attached are the design details and the status of the outstanding comitments to the Recommendations of NUREG 0578 for Nine Mile Point Unit 1.
Very truly yours, NIAGARA M01AWK POWER CORFORATION
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Donald P. Dise Vice President - Engineering EF:jk i
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///
40 0-1-070"Q 7 9 ATTACHMENT 1
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POA
NUREG 0578 RECOM EWTIOi, 2.1.8.b - ItCFE ASED RArGE OF RADIATION M3NITORS FOSITION The requirements associated with this recommendation should be considered as advanced implementation of certain requirements to be included in a revision to RegJlatory Guide 1.97, " Instrumentation to Follow the Course of an Accident," which has already been initiated, and in other Regulatory Guides, which will be pro.nulgated in the near-term.
1.
Noble gas effluent monitors shall be installed with an extended range desigled to function during accident conditions as well as during normal operating conditions; multiple monitors are considered to be necessary to cover the ranges of interest, tbble gas effluent monitors with an upper range capacity of 105 a.
Ci/cc (Xe-133) are considered to be practical and should be installed in all operating plants.
b.
tbble gas effluent monitoring shall be provided for the total range of concentration extending from normal condition ( ALARA) concentrations to a maximum of 105 C1/cc (Xe-133). Multiple monitors are considered to be necessary to cover the ranges of interest. The range capacity of individual monitors should overlap by a factor of ten.
2.
Since iodine gaseous effluent monitors for the accident condition are not considered to be practical at this time, capability for effluent monitoring of radiciodines for the accident condition shall be provided with sampling conducted by absorption on charcoal or other media, followed by onsite laboratory analysis.
3.
In-containment radiation level monitors with a maximum range of 108 rad /hr shall be installed.
A minimum of two such monitors that are physically separated shall be provided. Monitors shall be designed and qualified to function in an accident environment.
RESm NE By January 1,198] the following provisional steps will be taken:
The existing in-line stack monit ors are capable of detecting 50 C1/sec. or i
approximately 0.55 (Ci/cc (Xe-133) with normal ventilation flow of 180,000 ~
ft.3 minute.
These monitors Tiave read out and alarm capability in the
/
main control room. Quantification of higher level noble gas releases will be provided by means of a portable gamma survey instrument. This instrument will be installed such that it will monitor a portion of the sample line to the existing stack monitors. This line comes from an isokinetic probe in the main stack.,
ESP 3 f C- ( c mt. )
Ba& ground radiation will be shielded by means of a lead cave built around the detector.
The instrument has an upper limit of at least 1000 R/hr.
It will be calibrated with a Xe-133 source such that the reading can be related from R/hr. to UCi/sec stack release rate.
Since all station
~ " - ~ ~ ~ ~
effittents are discharged via the stack, the effluents monitored in this line are representative of the stack disdiarge. Until the Xe-133 calibration can be accomplished, the existing stack monitor ca'.ibration dependence data will be utilized to establish a calibration factor.
Readings on the interim monitor will be taken locally and the results verbally communicated to the main control room. This method wculd be used only in a case where the existing monitors were off-scale (high).
Communication will be by means of a headset and will be taken approximately every fifteen minutes, when required.
The in-line monitors are powered from redundant AC power sources. These monitors are not presently powered from emergency sources. Power to the interim monitors will be from a DC battery source, capable of eight consecutive days of continuous readout.
By January 1,1981 the following modifications will be performed:
1.
A high range effluent monitor will be installed. This monitor will either extend the range of the existing in line monitors or will provide for monitoring the entire range from normal concentrations ( ALARA) to the upper range defined in NUEG 0578 (or equivalent). The range of this monitor will consider dilution from ventilation sources that would be operating during an accident. Power to the monitor will be from a vital instrument bus. This monitor will meet the requirements of Regulatory Guide 1.97.
2.
Presently, charcoal canisters and particulate filters are taken to the lab and are analyzed by GeLi spectrometer. This method will continue to be used under accident conditions.
If'necessary, remote handling tools and lead pigs will be used. Canisters will be purged of noble gas in the hot lab ventilation hood to reduce interference with iodine analysis.
Collection times will also be reduced, if necessary, to control the amount of activity on the canisters and filters.
The charcoal canister and filte.r are located in the sample line to the stad< monitors. Sanples are representative of the main stack discharge through an isokinetic probe in the stack.
Continuous in line monitoring capability of iodine and particulates is being considered. These types of monitdrs.will become available and may be part of the high range monitor addition.
3.
Two independent containment radiation monitors will be installed. These monitors will be installed in existing spare penetrations sleeves of the containment.
The detectors will be located in sleeves which will extend into the free space of the containment, thus increasing the detectors reliability. -
i t
RESULTS (coat.)
The detectors will meet the requirements of RegJlatory Guide 1.97,
_ ~.-- ~.--
including seismic and environmental qualifications.
The range will be up to 108 R/hr and power will be provided from a vital instrument bus.
Display will be continuous with recording capability in the control room.
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UrdITED STATES
- g();, f $ i NUCLEAR REGULATORY COMMISSION
,y WASH WGT ON, D. C. 20555
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January 2,1980 C.'
_ _ Docket No. 50-220 Mr. Donald P. Dise Vice President - Engineering Niagara Mohawk Pwer Corporation 300 Erie Boulevard West Syracrse, New York 13202
Dear Mr. Dise:
The Cocrnission has issued the enclosed Show Cause Order for the Nine Mile Point Nuclear Station.
Your submittals of October 18 November 26, and December 19, 1979, regarding implementation of the short-tenn lessons learned' requirements indicate that you do not intend to implement some of the " Category A" requirements until after January 31, 1980. We have detemined that inplementation of the " Category A" requirements by January 31,19C2, is necessary to provide continued assurance of public health and safety.
The Show Cause Order requires that you implement, by January 31, 1980, the
" Category A" requirements of NUREG-0578 regarding short-tem lessons learned, as supplemented by our' letters of September 13, and October 30, 1979; or show cause why you should not.
Because this Order is imediately effective it also requires that the " Category A" requirements be implemented by January 31, 1980, or the plant be shut down.
A provision regarding equipment av !1 ability problems is included.
In your submittals, you did indicate that many of the " Category A" requirements would be implemented by January 1,1980.
It is not the Con::ission's intent, in issuing this Order, to encouraga delays in implementation of those
" Category A" items you currently have scheduled to complete by January 1,1980.
Therefore, you should still submit, on or shortly after January 1,1980, a description of the methods used to implement the " Category A" requirements completed by that time.
" Category B" lessons learned requirements, those scheduled for implementation by January 1,1981, will be the subject of future correspondence.
A copy of this Order is being filed with the Office of the Federal Register for Publication.
Sincerely, I
Harold R. Denton, Director Office of Nuclear Reactor Regulation Encl osure:
Order to Show Cause f._
ATTACHMENT 2
9 Mr. Donald P. Dise 2-Niagara Mohawk Power Corporation
.m cc:
Eugene B. Thomas, Jr., Esquire LeBoeuf, Lamb, Leiby & MacRae 1757 N Street, N. W.
Washington, D. C.
20036 Anthony Z. Roisman Natural Resources Defense Council 917 15th Street, H. W.
Washington, D. C.
20005 T. K. BeBoer, Director Technological Development Programs State of New York Energy Office Swan Street Building CORE 1 - Second Floor Empire State Plaza Albany, New York 12223 Mr. Robert P. Jones, Supervisor Town of Scriba R. D. !4 4
Oswego, New York 13126 Niacara Mohawk Power Corporation ATTN: Mr. Thomas Perkins Plant Superintendent Nine Mile Point Plant 300 Erie Boulevard West Syracuse, New York 13202 Director, Technical Assessment Division Office cf Radiation Programs (AW 459) i US EPA Crystal Mall #2 Arlington, Virginia 20460 U. S. Environmental Protection Agency Region II Office ATT!.:
EIS COORDINATOR 26. ederal Plaza New York, New York 10007 Oswe;o County Office Building 46 E. Bridge Street Oswego, New York 13125
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of
)
)
NIAGARA MOHAWK POWER CORPORATION
)
Docket No. 50-220
)
(Nine Mile Point Nuclear Station)
)
)
0 ORDER TO SHOW CAUSE
(
I The Niagara Mohawk Power Corporation (the Licensee) is the holder of Facility Operating License No. DPR-63 which authorizes the Licensee to operate the Nine Mile Point Nuclear Station at power levels not in excess of 1850 megawatts thermal (rated power). The facility is a boiling water reacter located at the Licensee's site in Oswego County, New York.
II Following the Three Mile Island Unit 2 (TMI-2) accident on March 26, 1973, a TMI-2 Lessons Learned Task Force of the Nuclear Regulatory Counission ( ARC)
Staff concucted an intensive review of the design ano operational aspects of nuclear power plants and the emergency procedures for coping with potential accicents. The Task Force identified measures to be taken in the short-term to reduce the likelihooo of accidents and to ~ improve emergency prepareaness in responcing to accidents.
These measures are set forth in HUREG-0578, 'T..I-2 j
V Lessons Learned Task Force Status Report and Short-tena Reconnendations".
Tre
2-NRC has concluded'that pro::ct icplementation of the actions denominated
'~~ ' Category A* requirements at operating nuclear power plants is necessary to These " Category A" provice continued assurance of public health and safety.
requirements were transmitted to all licensees operating nuclear power plants by letter dated Septemaer 13, 1979. By letter to affected licensees dated further clarification of these requirements was provided.
October 30, 1979,
.III The Licensee has cornittec to implementation of each " Category A" require-ment albeit not in all cases prior to January 31, 1980. NUREG-0578 and rrry letters of September 13 anc Octooer 30, 1979, which are hereby incorporated into this Order by reference, cescri:e in detail the basis for implementing " Category A" The majority of licensees have corxtitted to implement the requi rements.
the " Category A" requirements ey January 31, 1980 or the reactor wf'l shutdown However, other licensees Itave until such implementation is ccmplete.
indicated that additional.ecessary equipment, which is on order, will be delivered after this date. Thirty days af'ter delivery of equipment is a Based on practical time period during wnich the equipment can be installed.
available information, all equipment should be delivered and capable of being installea by June 1,1980. Licensees are required to meet the January 31, 1980 schedule unless they adequately demonstrate, in accordance with this Order, that delay based on equipment availability is justi.fied. For reasons discussed, timely i
e rovice continued assurance implementation of these req:.f rerents is necessary I
i of public health and safety.
ith incomplete " Category A" practicabla and in no instarce shall a licensee w actions continue operation teyond June 1,1980 IV f 1954, as amended, and Accordingly, pursuant o the Atomic Energy Act o2 and 50, I l
the Conrdssion's regulations in 10 CFR Partsir. the manne THAT the Licensee sr.ow :a;st, i ements (except shoulo not:
inplecent all " Category A" requ r By January 31, 1953,
- 76) referreo to in Part II i
the requirement of 2.1.7.a of NUREG-05 ipment is shown, r
of this Order, except tncse for which necessary equ justification to the Director, by appropriate anc tinely cocumentary lace and Office of NRR, to se u availaole, or in the alternative, p fueling mode of caintain its f acilities in a cold shutdown or re l
Category A" requirements not implemented by operation.
y owin; to the unavailability of necessar J anuary 31, 1930, f the date such f
equipment shall ne isp:ecer.ted within 30 days o June 1,1980.
equipment becomes aYai'able Dut no later than f " Category A" e
In view of the importa-ce of the prompt implementation o I have determined that requirements to the healtn anc safety of the public,s that th i
I the public health, safety c-interest requ re f the Commission.
effective as of this cate :encing furtner Order o i
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The Licensee nay file a written answer to this Order under oath or The Licensee affirmation within twenty (20) days of the date of the Or, der.
or any other person whose interest nay be affected by this Order nay request Any request for f
a hearing within twenty (20) days of the date of the Order.
Any a hearing will not stay the terporary effectiveness of this Order.
r request for a hearing shall be addressed to the Director, Office of Nuclear l
Reactor Regulation, U. 5. Naciear Regulatory Commission, Washington, D.
C.,
If a hearing is requested by a person whose interest may be affected 20555.
by this Order, the Commission will issue an Order designating the time and i
place of any such hearing, t
In light of the Licensee's expres:ed willingness to implement " Category A" l
requirements, except as incicated in Part III of this Order, in the event a the issue to be considered at such hearing shall be:
hearing is requested, whether all " Category A" requirements (except the requirements of j
2.1.7.a of NUREG-0578) should be implemented in accordance with the schedule prescribed by this Order.
Operation of the facility on terms consistent with this Order is not e
stayed by the pendency of any proceedings on the Ordar.
FOR THE NUCLEAR REGULATORY C0KMISSION f
d
,fe M d2 Harold R. Denton, Director Office of Nuclear Reactor Regulation Dated at Bethesda, Marylano l
tnis 2 day of January,1953 i
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION
)
In the Matter of NIAGARA M0 HAWK POWER CORPORATION
)
(Nine Mile Point Nuclear Station
)
Docket No. 50-220 Unit 1)
)
ANSWER TO SHOW CAUSE Pursuant to Part 2.206 of the Nuclear Regulatory Commission's Rules of Practice Niagara Mohawk Power Corporation (Niagara Mohawk or Licensee) hereby answers the Commission's Show Cause Order issued to Niagara Mohawk on January 2, 1980.
The Task Force which was assigned to the Three-Mile Island Unit 2 accident has issued its report, NUREG-0578, "TMI-2 Task Force and Short-tem Recomendations." A list of the Category A Recommendations, their applicability or non-applicability to Niagara Mohawk's Nine Mile Point Unit I and the date of compliance with those Recommendations are shown in Exhibit A to this Answer. Exhibit A is hereby incorporated by reference.
Licensee calls to the Commission's attention that only one item of the Task Force's Category A short-tem recommendations will not be complied with by January 31, 1980.
This item concerns the direct indication of Safety and Relief Valve Position.
Niagara Mohawk expects to receive from its vendor the necessary equipment on or about February 29, 1980.
The unit will be shut down within thirtysdays of receipt of equipment and will not resume operation until installation is complete.
Information regarding the schedule for procurement and delivery of this equipment is contained in Exhibit B, which is NOYO hereby incorporated by reference.
Slll Apa' sd
- oowe V} [ L
. D w Ton i H
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ATTACHMENT 3 i
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2 n
~ This schedule is well within the June 1,1980 deadline established by the Commission in the Show Cause Order.
Respectfully submitted, NIAG RA M0 HAWK POWER CORPORATION W *. // -
By:
t aM/(//
ames Bartlett
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xecutive Vice President STATE OF NEW YORK.
)
)
COUNTY OF ONONDAGA
)
SUBSCRIBED AND SWORN T0, before me, this r2 4 day of
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, 1980.
I lt,.rs.... ( x. '9 3..
Notary Public, New York My Commission Expires:
% 4 s., *s C.
e CTHTHIA A. PETTA t's lit e et N*. Ted
.e*sey totis et
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a;.eit. d to Cne ute Co. 88 -
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,L t
t NINE MILE POINT UNIT 1 l
IMPLEMENTATION OF " CATEGORY A" NUREG-0578 REQUIREMENTS IMPLEMENTAT(0N DATE IMPLEMENTATION REQUIREMENT TITLE CATEGORY,I)
COMPLETED I
2.1.8.b High Range Radiation Monitors Effluents - Procedures A
December 31, 1979 2.1,8.c Inproved lodine Instrumentation A
December 31, 1979 No Number Containment Pressure Monitor Containment Water Level Monitor Containment Hydrogen Monitor RSC Venting Design Description Submitted A
December 31, 1979 2.2.1.a Shif t Supervisor Responsibilities A
December 7, 1979 2.2.1.b Shift Technical Advisor Advisor on Duty A
January 7, 1980 2.2.1.c Shift Turnovar Procedure A
December 31, 1979 2.2.2.a Control Room Access A
December 31, 1979 (1)
CATEGORY-A:
IMPLEMENTATION COMPLETE BY JANUARY 1, 1980 Page 4 of 5
UMTED STATES j
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NUCLEAR REGULATORY COMMISSION l
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WASHINGTON, D. C. 20555 e,
/
March 21, 1980 f
- ...?
Docket No. 50-220 l
1 Mr. Donald P. Dise Vice President - Engineering i
Niagara Mohawk Power Corporation 300 Erie Boulevard West Syracuse, New York 13202
Dear Mr. Dise:
Enclosed for your information is the Staff's evaluation for the Nine Mile l
Point Nuclear Station, Unit No.1 of the actions you have taken to satisfy l
1 the Category "A" items of the NRC recomendations resulting from TMI-2 1
Lessons Learned. This evaluation is based on your submitted documentation and the discussions between our staffs at a site visit on March 12, 1980.
A list of meeting attendees is also enclosed.
Based on our review, we conclude that you have satisfactorily met all I
Category "A" requirements. The adequacy of certain implemented procedures i
and modifications will be verified by our Office of Inspection and Enforce-ment.
Each of these is discussed in our evaluation.
l 4
Should you have any questions regarding our evaluation, please contact us.
Sincerely i
.s 1
Thomas A. Ipp i
, Chief l
Operating Rea tors Branch f3 Division of Operating Reactors
)
Enclosures:
i 1.
Evaluation 2.
Meeting Attendees cc w/ enclosures:
l See next page
- gro" g-4 4
ATTACHMENT 4 3f t
i
' ?!r. Donald P. Dise 2-
,_ Niagara Mahawk Power Corporation
- - cc:
Eugene B. Thomas, Jr., Esquire LeBoeuf Lamb, Leiby & MacRae 1757 N Street, N.W.
Washington, D. C.
20056 State University College at Oswego Penfield Library - Documents Oswego, New York 13126 9
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. EVALUATION OF LICENSEE'S COMPLIANCE WITH CATEGORY "A" ITEMS OF NRC RECOM4ENDATIONS RESULTING FROM TMI-2 LESSONS LEARNED l
Niagara Mohawk Power Corporation Nine Mile Point Nuclear Station Unit No. 1 Docket No. 50-220 Date: March c./, 1980 l
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INTRODUCTION
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1 31(5ftters dated October 18((l), November 26(2), December By 1 e-~
, 1979, and January 6), 1980, Niagara Mohawk Power Corporation (licensee)' submitted comm,31 itments and documentation of actions taken at Nine Mile Point Power St'ation, Unit No.' 1, to implement our requirements resulting from TMI-2 Lessons Learned. To expedite our review of the licensee's actions, members of the staff visited the licensee's facility i
on March 12, 1980. This report is an evaluation of the licensee's efforts to implement each Category "A" item which was to have been completed by January 1980.
II.
EVALUATION Each of the Category "A" requirements applicable to BWRs is identified below. The staff's requirements are set forth in Reference 7; the acceptance criteria is documented in Reference 8.
The numbered designation of each item is consistent with the identifications used in NUREG-0578.
2.1.1 EMERGENCY POWER SUPPLY The NRC requirement, as it is applicable to BWR's, is that provisions must be made such that the power-operated relief valves can be supplied emergency power when off-site power is not available.
Further, for air-operated valves, emergency power must be available to the air compressors in order to provide a long term supply of air. The reactor water level instrumentation must also be capable of operating from emergency power.
l The licensee has stated (4) that the relief valves at Nine Mile Point are electromatic operated relief valves, supplied by emergency electrical i
power. This type relief valve does got require instrument air for operation.
The licensee also statedU1 that vessel level indication instrument channels for safety system activation and control are also powered by emergency power.
Based on our review we have determined that no modifications are necessary to satisfy the requirements for this item.
2.1.2 PERFORMANCE TESTING FOR BWR RELIEF AND SAFETY VALVES The staff's position is ttat 3 oiling Water Reactor licensees shall functionally test the reactor coolant system relief and safety valves to demonstrate operability under expected operating and flow conditions.
The Category "A" requirement is for the licensee to commit to perform an appropriate test program.
The licensee is a member of a GE BWR Owners Grpup and has conunitted(2) to a test program adopted by this Owners Group.19J i
L.
5 We conclude that the licensee has satisfied 'the Category "A" requirements f __*
for this item.
i Direct indication of Pcwer-Operated Relief Valves and Safety Valve _
2.1.3.a Position for BWR's_
l The staff's position is that BWR licensees shall provide a positive indi-The valve cation for reactor coolant system relief and safety valves.
position should be indicated and alarmed in the con of flow in the discharge pipe so that the operator is provided with anIf th unambiguous indication of valve position.is not safety grade, a relia l
from the emergency bus may be provided if backup methods of determining i
Further, the valve position indication valve position are available.
should be seismically qualified consistent with the components or system If seismic qualifications are not feasible by to which it is attached.
January 1,1980, then justification should be provided and a schedule i
submitted for upgrading the system to meet the seismic requirements.
To meet the above position, the licensee has provided an acoustical system to monitor the position of each of the relief and spring safety valves.
The acoustical system consists of a hermetically sealed piezoelectric sensor mounted on the downcomer piping of the relief valve and on the flange The sensor is held in place by a special of the spring safety valves. The sensor is connected to the preamplifier stainless steel band clamp.
The signal
~
through the use of high temperature, low noise coaxial The open or closed provide indication of the position of each valve.
position for the relief valves is also indicated on the main control panel.
For the twenty two relief and safetf valves, the preamplifiers are located outside containment, where the temperature during an accident is lower Each Flow Detector module is located on an than inside containment.
auxiliary rack below the main control room and provides indicator lights The " memory for " closed," "open" positions, and a " memory circuit."
circuit" for each valve when ectivated, stays on until manually reset; thus it provides an indication of valve actuation even though the valve If any of the flow detectors indicate a valve in may have since closed.
the operi position, a comon dedicated single window of the plant annunciator is activated and when modifications are completed its signal will be inputed An on-line system test to the plant process computer and event recorder.
circuit for alarm has been provided.
The Nine Mile Plant has 6 electromatic relief valves and 16 spring safety The relief valves have the capability to be operated manually; however, three of these valves have been dedicated to the ADS function while valves.
The acoustic monitoring system the remaining three provide redundancy.
- However, installed on each relief or safety valve is not fully safety grade.
e
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o the licensee has stated and we agree that the system is a reliable single channel system that provides direct indication of valve position. The g._
system is powered from the emergency AC bus that has automatic transfer capability to a DC power supply in the event of a failure of AC power system.
Back-up valve, position indication infomation is provided and discussed l
in the emergency procedures so that the operator can make a diagnosis and take appropriate action. The back-up valve position indication is pro-vided by temperature indicators.
Each individual valve has an amheddhd type thermocouple attached to the tailpipe downstream of the valve discharge point. Signals derived from the embedded themocouple are readout and alarmed on the process computer. The power for the back-up temperature monitor position indicators is provided from a non-class lE instrument bus.
Therefore, in the event of a single failure of a power supply, at least one indicating system is available to provide the reactor operator with valve status.
The temperature indication instrumentation is already seismically and environmentally qualified and is available for backup verification of valve position.
The acoustic monitoring system valve position indicators have not yet been seismically or environmentally qualified. The licensee stated that the position indication system and components both inside and outside the containment is presently being environmentally and seismically qualified by participating in the BW qualification program.~~. The.. position indication' system components will'blirseisinidilly~qda'lified in accordance with IEEE 344, 1975 and qualified for their appropriate environment in accordance with IEEE 323, 1974 by January 1981. This schedule meets our requirements.
i Based on our review of the licensee's submittal. we conclude that the licensee is in compliance with the direct indication of power-operated relief valves and safety valve position and schedule requirements for upgrading the system to meet the seismic requirements as outlined in NUREG-0578, and is, therefore, acceptable.
2.1.3.b Instrumentation for Inadequate Core Cooling The NRC requirements, licensee actions and our evaluation thereof for this item will be evaluated separately by the NRC Bulletins and Orders Task Forc9 and reported in NUREG-0645 which is incorporated herein by referencet10),
2.1.4 CONTAINMENT ISOLATION The NRC requirements are that the licensee is to:
(a) carefully reconsider the determination of which systems should be considered essential or non-essential for safety. (b) modify systems as may be necessary, to isolate all non-essential systems by automatic, diverse, safety-grade isolation signals, and (c) modify systems, as may be necessary, to assure that the resetting of the containment isolation.
signals does not cause the inadvertent re-opening of containment isolation valves.
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The licensee's classification of essential systems was based on determination of which are required or could be of direct aid in mitigating the conse-quences of an accident.
All other systems which penetrate the primary contginment are non-essential. ^ The licensee's letter dated November 28, 1979til) includes an identification of each penetration, classification as essential (engineered safety function) or non-essential, and identification of the isolation signals for each.
As stated in NUREG-0578 our goal is to use information provided by licensees to develop a consistent set of guide-i lines for the selection of essential and non-essential systems. Accordingly.
the licensee has satisfied this aspect of the requirements for this item.
The licensee stated (5) that certain systems (reactor building closed loop.
cooling to the recirculation pump coolers and dry well coolers) do not require containment isolation signals since they are closed systems that do not comunicate with the reactor coolant pressure boundary or free space of the containment.
Two systems (the recirculation sample and suppression chamber to waste j
system lines) are normally closed during operation. The licensee has comitted to install automatic isolation valves in these systems.
In i
the interim the licensee comitted at the site visit, that when such systems are in use, an operator will be dedicated to the function of assuring that isolation valves are closed in the event of a containment isolation signal.
Based on our review of References 5 and 11 as well as are either (1) automatically isolated by diverse signals, (2) penetration discussions at the site visit we conclude that non-essential adequate compensatory measures have been instituted or (3) a rationale for deviation from the general requirements has been provided. We find that the licensee has satisfied the basic intent of this aspect of the requiremente for this item.
The licensee's design of control switches for containment isolation valves was discussed at the site visit. The design involves the use of spring-return-to-neutral control switches and holding relays.
For this type design resetting of a containment isolation signal does not result in the automatic reopening of containment isolation valves. We find that no modification to the design is required.
Based on the above, we conclude that the licensee has adequately conformed to the requirements of this t item The Office of Inspection and Enforcement n will verify the adequacy of proceliures and the completion of design V changes as discussed above.
2.1.5.a Dedicated Penetrations for External Recombiner of Post-Accident External Purge System i
The staff's position is that licensees whose plant uses external recombiners or purge systems for post-accident control of combustible gas in the contain-
o Ie 5-1 ment atmosphere should provide a containment isolation system that is l
dedicated to that function only. The system's design should be redundant
~~
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and meet out single failure requirements to that criterion 54 and 55 of the General Des-ign Criteria are met and that the system is sized to satisfy the flow requirements of the recombiner or purge system. This requirement is applicable to those plants whose licensing basis includes requirements for external or purge systems for post-accident control of combustible gas in the primary containment.
The Nine Mile Point Unit is designed to use a Containment Atmosphere
}
Control (CAC) system prior to each startup and during routine operations to maintain the oxygen concentration in the primary containment atmos-i phere to less than 5 percent to ensure that combustion of the hydrogen and oxygen cannot occur. We have detemined that the CAC system consists of the following major subsystems: The normal containment purge and exhaust subsystem, the containment inerting subsystem and the containment atmospheric make-up subsystem. These subsystems do not perform any safety 4
function.
Only those components associated with maintaining the containment isolation integrity (up to and including second containment isolation valve) are safety related and have been designed to seismic i
Category I requirements.
The Containment Atmospheric Dilution (CAD) system performs the safety function of limiting initial oxygen concentration to less than 5 percent in order to preclude a flamable mixture in the containment imediately following a LOCA and to maintain this inerted primary containment mixture on a long tem basis following a LOCA. The CAD system is used during emergencies and as such has been designed to seismic Category I require-ments; electrical components meet applicable portions of IEEE-279, and have suitable redundancy and interco,nnections so that a single failure of an active component will not render the system inoperable. The CAD system is functionally independent from the normal inerting system and its components include a storage vessel, electric vaporizers, redundant lines and valves and associated instrumentation. The nitrogen from the CAD system is injected into the drywell or torus using the purge air system lines. The CAD system branch lines are connected to the purge lines downstream of their redundant containment isolation valves. Two solenoid actuated isolation valves for each of the reduridant torus and drywell CAD lines have remote control switches located in the main control room.
In addition, two analyzers for hydrogen and oxygen have been provided for the containment drywell/ torus that are redundant to each other and are designed to meet seiismic and IEEE 279 requirements.
Initially, the licensee stated that the CAD system at Nine Mile Point was reviewed to verify that isolation provisioh's for piping and inte'r:-~~ ~ ~ ~ ~
connected lines are single failure proof during CAD operations. The licensee reported the results of the review in the December 31,1979 submittal.
They have verified that the containment isolation provisions for all lines (including vent and purge) are single failure proof during n.-
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t e
CAD system operations except for a single pathway. During this operation a single failure of a blocking valve could result in an uncontrolled
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pathway from containment for radioactivity to escape through the ventilation system. The licensee proposed a modification to interlock nonnally open valves to assure that they automatically close during nitrogen addition to the containment, so that two isolation valves in series exists for all pathways. The modification will be completed as soon as practicable but prior to January 1,1981.
In the interim, the licensee has committed to modifying procedures to assure that the pumpback system is not in operation during CAD operations. With the pumpback system not in operation, two additional isolation valves are automatically closed.
This prevents a single failure pathway from existing during the nitrogen
- addition operations.
Based on the above, we conclude that the licensee has satisfied the Category "A" requirements related to this item.
2.1.5.c Recombiner Procedures The NRC requirements' for this 1. tem apply only to those plants that include hydrogen recombiners as a design basis for licensing. We have determined that this item is not applicable to the Nine Mile Point Plant.
2.1.6.a Systems Integrity The NRC objective is to eliminate or prevent the release of significant i
amounts of radioactivity to the environment via leakage from engineered safety systems and auxiliary systems, which are located outside reactor containment. The requirements are to implement practical measures to reduce leakage, report leakage measurements to the NRC and establish a preventive maintenance program to maintain leakage at as-low-as practicable levels.
Based on our review of licensee submittal and discussion with licensee during the NRR/01E site visit, we find that the licensee has tested and measured leak tightness of systems, developed leak reduction, and initiated a preventive maintenance program.
We conclude that the licensee has satisfied the requirements of this Category "A" item. The Office of Inspection and Enforcement (0IE) will review the licensee's procedures to verify adequacy. OIE will also U
verify the implementation of a preventative maintenance program and the completion of personnel training.
Results will be reported in appropriate
' inspection reports._
2.1.6.b Plant Shielding Review The Category "A" requirements for this item are to perfonn a design review of current plant shielding to identify where corrective actions are needed to pennit personnel access to vital areas, and to protect safety equipment.
r The licensee has completed a general plant shielding review of the vital l
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areas requiring continuous and infrequent access. Problem areas such i
as coolant sampling systems will be relocated from the containment, or a new, shielded sampling sink will be provided in the turbine building.
The present shielding of the control room is sufficient to maintain dose to personnel within required limit. The TSC and OSC may require additional shielding; detailed evaluation is in progress.
Shielding review of safety system components for degradation from TID sources is presently in the final stages of completion.
The results will Be reported as part of the actions under OIE Bulletin 79-01B.
Based on the above, we conclude that the licensee has satisfied the Category "A" requirements for this item.
2.1. 7. a Auto Initiation of AFW 2.1.7.5 AFW Flow 4
These items (2.1.7.a and 2.1.7.b) are unique to PWRs and are not applicable to the Nine Mile Point Plant.
j 2.1.8.A Post-Accident Sampling The NRC objective is to quantify the degree of core damage in the course of an accident by radiological and chemical analysis of samples of reactor coolant and containment atmosphere. The Category "A" requirements are:
(a) to review the design of reactor coolant and containment sampling system to determine the capability of personnel to obtain a sample (within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) under accident conditions without exposing an individual in excess of 3 Rem and 18 3/4 Rems to the whole body or extremities; (b) to review operational procedures of the radiological spectrum and chemical analysis facilities to determine the capability to quantify radioisotopes that l
are indicators of the degree of core damage; and (c) to describe proposed plant modifications.
The licensee'has completed a design review of reactor coolant and contain-ment atmosphere sampling systems. Additional shielding has been provided, however, because personnel radiation exposure may exceed the required limit further modifications may be necessary. The capability exists for and 0 obtaining containment atmosphere samples from the containment H2 2
monitoring system.
The licensee stated that all samples can be obtained and analyzed for required isotopes within 2. to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
Interim procedures have been developed for collecting, transporting and analyzing samples.
4 I'
Based on our review we conclude that the licensee has satisfied the requirements of this Category "A" item.
2.1.8.8 'High Range' Radiation Monitors The NRC objective is to have available adequate instrumentation to follow the course of the accident.
The Category "A" requirements are to have procedures quantifying effluent releases in case existing instrumentation would go off scale (" provisional fix"). This includes a description of System / Method employed, and description of procedures for conducting all aspects of the measurement / analysis for noble gases.
radiciodines, and particulate effluents.
The existing in-line monitors are capable of detecting 50 Ci/sec, and have read out and alarm capability in the main control room. Quantification of higher effluent releases is provided by portable gamma survey instru-ments, positioned at a predetermined location to monitor a portion of the effluent sample line.
This " provisional fix" has been installed and calibrated. Conversion factors have been calculated, for detecting up to 10,000 Ci/sec NG effluent releases at the stack. Since all station effluents are dis-charged via the stack, only one itigh range effluent monitor is required.
Interim methods and appropriate procedures have been written, approved j
and implemented.
Personnel training has been conducted for sampling.
quantifying and analyzing effluent releases.
Based on our review, we ' conclude that the licensee has satisfied the Category "A" requirements of this it.em.
2.1.8.C Improved Iodine Instrumentation The NRC Category "A" requirements are that each licensee shall provide equipment and associated training and procedures for accurately determining the airborne iodine concentration in areas within the facility where personnel may be present following an accident.
The licensee has available approximately 12 portable air samplers, using charcoal cartridges. These samplers are not equipped with single channel analyzers. The charcoal cartridges wil1~ be analyzed in the counting room which is located approximately 1 minute walking distance from the control room, TSC and OSC. The licensee stated that the air samples can be collected,_transpor.ted,_ flushed and analyzed for iodine concen-tration within 10 minutes.
Interim procedures for obtaining, tiransporting,' ~ ~
l preparing and analyzing samples have been developed and implemented.
Based on our review we conclude that the licensee has satisfied the requirements of this item.
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-g-J.__2.2.1.A Shift Supervisor _ Responsibility The NRC requirement for this item is to revise, as necessary, the responsibilities of the Shift Supervisor such that he can provide direct, connand oversight of operations ~ and perform management review of ongoing operations that are important to safety and not be dis-tracted from these important responsibilities by administrative details.
The licensee har revised Plant Procedure APN 2A to satisfy this requirement.
We conclude that the licensee has satisfied the requirements of Item 2.2.1.A to provide revised responsibilities and authority for the Verification of the adequacy of the licensee's d(
Shift Supervisor.
procedures will be performed by the Office of Inspection and Enforcement and will be documented by appropriate Inspection Reports.
2.2.1.b Shift Technical Advisor _
The NRC requirement is for the licensee to provide an on-shift technical advisor (STA) to the shift supervisor to serve the two functions of accident assessment and operating experience assessment. As a supplement to.the operating staff, the STA must be able to report to the control room within 10 minutes to assist in diagnosing an off-nonnal event.
The licensee stated (4) that shift manning will be augmented by an Assistant I
Shift Supervisor to satisfy the accident assessment function of the Shift Technical Advisor. The operating experience as,essment function is performed by station professional personnel and corporate level engineering personnel coordinated by a person on the station technical staff whose primary duty will be operating experience assessment.
We conclude that the licensee has satisfied the Category "A" requirements for this item.
2.2.1.c Shift and Relief Turnover Procedures The NRC requirement is for the licensee to assure that procedures are adequate to provide guidance for a complete and systematic turnover between tne off-going and on-coming shift to assure that critical plant parameters are within limits and that the availability and alignment of safety systems are made known to the on : coming shift.
The licensee has revised Plant procedure APN 2A to implement this item.
We conclude that.the licensee has satisfied the requirements of Item 2.2.1 to provide new procedures. Verification of the adequacy of the d
implemented checklists and logs will be performed by the Office of Inspection and Enforcement and will be documented by appropriate Inspection Reports.
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~~ 2.2.2.'A Control Room Access The NRC requirement includes implementing procedures to limit access to the control room and establishing clear lines of authority in the control room in the event of an emergency.
The licensee has revised Plant Procedure APN 2A to implement this item.
We conclude that the licensee has satisfied the requirements of Item 2.2.2.A.
Verification of the adequacy of the implemented procedures will be performed by the Office of Inspection and Enforcement and will be documented by appropriate Inspection Reports.
2.2.2.b Technical Support Center The NRC requirement is that each licensee establish and maintain an onsite technical support center (TSC) separate from and in close proximity to the control room. The TSC should have reliable communi-cation systems and plant as-built technical data to provide information to those individuals knowledgeable and responsible for engineering and management support to reactor operations in the event of an accident.
Further, the licensee must describe the long range plan to upgrade the TSC to meet the Category "B" requirements.
The licensee has designated the Training Room in the Administration Building as the Onsite Technical Support Center (TSC).
During the NRR/
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OIE site visit we toured the TSC.
The center is habitable to the same degree as the control room. Direct telephone comunications and airborne and radiation monitoring capability have been provided. Access to permanent plant records, as-built drawings and procedures is available.
The TSC staff has access to technicql data by two selector typers which print out the same information available in the control room.
In addition, display of plant parameters can be provided by means of a camera with focus, zoom and pen and tilt controls. The existence and staffing of the center are included in the Nine Mile Point Unit 1 Emergency Plan. The licensee's submittal dated January 31,1979 also includes a discussion of his plans to upgrade the Center to satisfy our Category "B" requirements.
We conclude that the licensee has satisfied the Category "A" requirements for this item.
2.2.2.c Operational Support Center The NRC requirement.is to establish an area in which_ shift pers_onnel can report for further instructEns from the operations staff.
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During the NRR/01E site visit the licensee stated that the Plant lunchroom has been designated Onsite Operational Support Center.
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The center has telephone connunications. The licensee's Emergency Plan covers this Center.
We conclude that the licensee has satisfied the requirements of 2.2.2.c.
f NRR ITEM:
REACTOR COOLANT SYSTEM VENTING As specifically related to BWRs, the Category A requirements of this item is to provide current design information to demonstrate that non-condensable gases can be vented from the primary coolant system, includin,g isolation condensers.
The licensee's submittal dated December 31, 1979, provided design infor-mation on the capability of the Nine Mile Point design for remotely venting non-condensables from the reactor coolant system. Reactor vessel head high points can be vented by relief valves and the head vent system.
The licensee's review of the capability to vent the isolation condensers indicated'that modifications are necessary to assure venting capability during accident conditions. The submittal described the modifications.
The schedule for completion is consistent with our requirements for a Category "B" item.
Based on our review we have determined that the licensee has satisfied
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the Category "A" requirements for this item.
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r 1.
Letter, HMPC (Rhode) to NRC (Eisenhut) October 18, 1979.
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Letter, NMPC (Bartlett) to NRC (Denton). November 26,1979.
3.
Letter, NMPC (Dise) to NRC (Denton), December 19, 1979.
4.
Letter, NKPC (Rhode) to NRC (Denton), December 20, 1979.
5.
Letter, NMPC (Dise) to NRC (Denton), December 31, 1979.
6.
Letter, NMPC (Dise) to NRC -(Denton), January'31,1980.
7.
' Letter, NRC (Eisenhut) to ALL OPERATING NUCLEAR POWER PLANTS, September 13, 1979.
8.
Letter, NRC (Denton) to ALL OPERATING NUCLEAR POWER PLANTS, October 30, 1979.
9.
Letter, BWR OWNERS GROUP (Keenan) to NRC (Eisenhut), December 14, 1979.
10.
NUREG-0645 Report of the Bulletins and Orders Task Force, January 1980.
11.
Letter, NMPC (Dise) to NRC (Ippolito), November 28, 1979.
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e Lessons Learned Site Visit Nine Mile Point 1
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NRC P. J. Polk NRC Louis B. Riani NRC Frank C. Skopec NRC Peter Francisco NMPC Melvin A. Silliman NMPC Walt Baumack NRC T. J. Perkins NMPC D. M. Verrelli NRC E. Leach NMPC B. Taylor NMPC t.
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