ML20138J825

Nuclear Regulatory Commission Issuances for February 1997. Pages 49-93

ML20138J825
Person / Time
Issue date:	04/30/1997
From:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
To:
References
NUREG-0750, NUREG-0750-V45-N02, NUREG-750, NUREG-750-V45-N2, NUDOCS 9705090049
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. NUCLEAR REGULATORYL

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, - Available from  !

Superintendent of Documents U.S. Government Printing Office RO. Box 37082  ;

Washington, DC 20402-9328

• f A year's subscription consists of 12 softbound issues, 4 indexes, and 2-4 hardbound editions fer this publication.

t 4 Single copies of this publication ,

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Errors in this publication may be reported to the Office of Information Resources Management U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 (301 - 415-6844) 1

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l NUREG-0750 i Vol. 45, No. 2 '

Pages 49-93 i

! NUCLEAR REGULATORY  ;

j COMMISSION ISSUANCES i

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i February 1997 l

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l This report includes the issuances received during the specified l period from the Commission (CLI), the Atomic Safety and Licensing l

Boards (LBP), the Administrative Law Judges (ALJ), the Directors' i i Decisions (DD), and the Decisions on Petitions for Rulemaking l (DPRM)

The summaries and headnotes preceding the opinions reported l herein are not to be deemed a part of those opinions or have any j independent legal significance.

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- U.S. NUCLEAR REGULATdRY COMMISSION i

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4 Prepared by the

( Office of Information Resources Management i U.S. Nuclear Regulatory Commission i Washington, DC 20555-0001 i (301 - 415 - 6844) j l

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l COMMISS!ONERS l l l l l

Shirley A. Jackson, Chairman l Kenneth C. Rogers  ;

Greta J. Dicus  !

Nils J. Diaz Edward McGaffigan, Jr.

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B. Paul Cotter, Jr., Chief Administrative Judge. Atomic Safety & Licensing Board Panel I

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i CONTENTS i Issuance of the Nuclear Regulatory Commission LOUISIANA ENERGY SERVICES, L P.

(Claiborne Enrichment Center) 4 Docket 70-3070-ML I ORDER, CLI-97-3, February 13,1997. . . 49 Issuance of the Atomic Safety and Licensing Board 4

RALPil L. TETRICK (Denial of Application for Reactor Operator License) I I

Docket 55-20726-SP (ASLBP No. 96-72101-SP)

(Re: Operator License)

INITIAL DECISION, LBP-97-2, Rbruary 28,1997. . 51 1ssuances of Directors' Decisions i

l ENVIROCARE OF UTAH, INC.

l Docket 40-8989 (License No. SMC-1559) l DIRECTOR'S DECISION UNDER 10 C.F.R. 5 2.206, DD-97-2, Rbruary 5,1997 . . 63

.) NORTilEAST NUCLEAR ENERGY COMPANY '

] (Millstone Nuclear Power Station. Unit 1) 7 Docket 50-245 (License No. DPR 21)

DIRECTOR'S DECISION UNDER 10 C F.R. 6 2.206,

DD-97-4, Rbruary 11,1997 . . ., , 86 4

TOLEDO EDISON COMPANY, et al.

(Davis-Besse Independent Spent Fuel Storage Installation)

Dockets 50-346, 72-1004

! DIRECTOR'S DECISION UNDER 10 C.F.R. 5 2.206, l DD-97-3, Rbruary 5,1997 . .. . . 71 l

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Cite as 45 NRC 49 (1997) CLI-97-3 UNITED STATES OF AMERICA '

NUCLEAR REGULATORY COMMISSION '

t COMMISSIONERS: >

l Shirley Ann Jackson, Chairman f

Kenneth C. Rogers Greta J. Dlcus i Nils J. Diaz Edward McGaffigan, Jr.  !

In the Matter of Docket No. 70-3070-ML LOUISIANA ENERGY SERVICES, L.P. .

(Claiborne Enrichment Center) February 13.1997 '

The Commission grants petitions filed by the Staff and Louisiana Energy Services for Commission review of the Atomic Safety and Licensing Board Partial Initial L)ecision, LBP-96-25, 44 NRC 331 (1996), and sets a bricting schedule pursuant to 10 C.F.R. 5 2.786(d).

ORDER The Nuclear Regulatory Commission Staff and Louisiana Energy Services (LES) have filed petitions for Commission review of the Atomic Safety and Licensing Board's December 3,1996 Partial Initial Decision, LBP-96-25, 44 NRC 331 (1996). This proceeding involves LES's application for a license to construct and operate the Claiborne Enrichment Center (CEC) near Homer, Louisiana. He Intervenor, Citizens Against Nuclear Trash (CANT), opposes the petitions for Commission review. In accordance with the considerations set forth in 10 C.F.R. 6 2.786(b)(4), the Commission has decided to grant the petitions and will review the issues raised in the Staff's and LES's petitions.

1. SCHEDULING OF BRIEFS Pursuant to 10 C.F.R. I2.786(d), the Commission sets the following briefing schedule:

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1. The Staff and LES shall file their briefs on or before March 13, 1997.

Each brief shall be no longer than 40 pages.

2. CANT shall file a single responsive brief on or before April 10, 1997.

Its response shall not exceed 50 pages. We allow 50 pages for CANT's responsive brief so that CANT will have adequate space to respond to sgar e approaches that may be taken in the opening briefs of the Staff and LES.

3. On or before April 24, 1997, the Staff and LES may file reply briefs.

Their replies shall not exceed 15 pages each.

To be timely, all documents must be served on the parties and on the Commission, so that they are received in the hands of the recipient no later than 4:15 p.m., Eastern Time, on the due dates forfiling. Any means is permitted, including hand delivery, facsimile transmission, or e-mail. liowever, for service on the Commission, facsimile or e-mail transmissions shall be followed by a mailed original signed copy. Briefs in excess of 10 pages must contain a table of contents, with page references, and a table of cases (alphabetically arranged),

statutes, regulations, and othes authorities cited, with references to the pages of the brief where they are cited. Page limitations on briefs are exclusive of pages containing a table of contents, table of cases, and of any addendum containing statutes, rules, regulations, etc.

11. REMAINING ISSUES BEFORE Tile IlOARD The Commission expects that the Board will be able to decide the remaining issues by May 1,1997. If the Board cannot do so, the Board should advise the Commission and parties of an alternative, reasonable schedule for deciding these issues.

IT IS SO ORDERED.

For the Commission' JOIIN C. HOYLE Secretary of the Commission Dated at Rockville, Maryland, this 13th day of February 1997.

I Conmusswners Thcus and thaz were not available for the afhrmation of this Order, sad been present.

they would have approved the (hder l

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i Atomic Safety l and Licensing i Boards issuances l

ATOMIC SAFETY AND UCENSING BOARD PANEL I

I B. Paul Cotter, Jr.,* Chief Administrative Judge

• James P. Gleason,* Deputy Chief Administradve Judge (Executive)

Frederick J. Shon,* Deputy Chief Administrative Jucfge (Technical)

Members Dr. George C. Anderson Dr. Richard F. Foster Marshall E. Miller Charles Bechhoefer* Dr. David L. Hetrick Thomas S. Moore * {

Peter B. Bloch* Emest E. Hill Dr. Peter A. Morris G. Paul Bollwerk fil* Dr. Frank F. Hooper Thomas D. Murphy

• Dr. A. Dixon Callihan Dr. Charles N. Kelber* Dr. Richard R. Parizek Dr. James H. Carpenter Dr. Jerry R. Kline* Dr. Harry Roin Dr. Richard F. Cole

• Dr. Peter S. Lam

• Lester S. Rubenstein Dr. Thomas E. Elleman Dr. James C. Lamb til Dr. David R. Schink Dr. George A. Ferguson Dr. Emmoth A. Luebke Dr. George F. Tidey Dr. Harry Foreman Dr. Kenneth A. McCollom I

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i Cite as 45 NRC 51 (1997) LB P-97-2 ,

e UNITED STATES OF AMERICA -

NUCLEAR REGULATORY COMMISSION l

ATOMIC SAFETY AND LICENSING BOARD PANEL l

Before Administrative Judges:

Peter B. Bloch, Presiding Officer {

Peter Lam, Special Assistant in the Matter of Docket No. 55-20726-SP '

(ASLBP No. 96-721-01-SP) I (Re: Operator License)

RALPH L TETRICK (Denial of Application for Reactor i Operator License) February 28,1997  ;

The Presiding Officer determined that a reactor operator should be considered to have passed the written test for senior reactor operator. -

He determined that one of the questions on the exam was ambiguous and  !

should be disallowed. He also determined, in the absence of guidance from the Staff of the Commission, that examination scores are sufficiently imprecise I that they should be rounded to the nearest integer. As a consequena, the score l on the written examination was 80%, which the Presiding Officer considered  !

a passing score. Since this was the last hurdle for the applicant in obtaining l his license, the Presiding Officer directed the Staff to issue a Senior Reactor r Operator's license to him.

I INITIAL DECISION I

Ralph L Tetrick, a reactor operator at the Turkey Point Nuclear Generating Plant, Units 3 and 4 ("Tudey Point"), operated by Florida Power & Light Company (" Florida Power"), is an applicant for a senior reactor operator's ,

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(SRO's) license. On October 21, 1996, I granted Mr. Tetrick's request for ,

j a hearing concerning whether he had pamd his SRO lianse examination.' An i SRO is defined in 10 C.F.R. 6 55.4 as "any individual licensed under this part to manipulate the controls of a facility and to direct the licensed activities of licensed operafors." (Emphasis added.)

Ac Nuclear Regulatory Commission (NRC) has jurisdiction of this request for a hearing, in which Mr. Tetrick appeals the result of his written examination .

The NRC helps to assure the health and safety of the public by requiring reactor operators to successfully demonstrate their knowledge of nuclear power plant operation before they are licensed. See Alfred J. Morabito (Senior Operator License for Beaver Valley Power Station, Unit 1), l.BP-88-10, 27 NRC 417 (1988), and LBP-88-16, 27 NRC 583 (1988); Rodger 15'. Ellingwood (Senior i Operator Licer.se for Catawba Nuclear Station), LBP-89-21,30 NRC 68 (1989). I ne Commission's regulations in 10 C.F.R. 55 55.43 and 55.45 require i that an applicant for a senior reactor operator license pass both a written examination and an operating test. Written examinations taken by applicants for senior reactor operator licenses are developed and administered by the I licensee, in this case Florida Power & Light Company, and are governed by 10 C.F.R. 5 55.43. Written examination questions test "the knowledge, I skills, and abilities needed to perform licensed senior operator duties." 10 I C.F.R. 5 55.43(a). In addition to information contained in a facility's training program, knowledge of "information in the Final Safety Analysis Report, ]

system description manuals and operating procedures, facility license and license amendments, [and] Licensee Event Reports" may properly be te.sted. Id. Written examinations for senior operators include a representative sample of questions from fourteen subject areas specified for operator license applicants in 10 C.F.R. 5 55.41(b)(1)-(14), In addition, written examinations for senior operators are to include a representative sample of questions from the seven areas specified in 10 C.F.R. 6 55.43(b)(i)-(7).2 In addition to the written test, Mr. Tetrick took and passed the operating ,

test, which involves a plant walkthrough and dynamic simulator evaluation {

during which various plant tasks, scenarios, and questions are presented to the '

applicants. See 10 C.F.R. 6 55.45.

On the written examination, Mr. Tetrick was scored by the examiner as correctly answering 78 of 100 multiple-choice questions, for a score of 78%,

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'This is an informal lwanns tmder 10 C F R. Part 2. Subpart L See 10 CF R 12120)(aK2) By letter of l Noventer 7.1996. Ow NRC Staff O' staff") subnutted the Heanng ble pursuant to 10 C f R. Il 1231. on l Decenter 30.1996 Mr. Tetnck filed tus wntien presentauon in stus proceeding, pursuant to 10 CI R. I 2.1233  !

(Tetnck Presentauon) Staff rephed, pursuant to this sane secuon of the regulauons, on hnuary 23.1907 (Staff l

Presentation). I 2

5cc NL'RLG-1021. " operator 12ceniung Emanuner Standards," for funher guidance ce de adnuiustranon and l grading of the scruor reactor operator wntien test j 52 1 i

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which does not meet the 807c minimum score required to pass. See NUREG-1021, at 5 of 6. In response to Mr. Tetrick's request, the Staff completed an informal review that confirmed his failing grade. Hearing File item 21, attachment at 2-7.

Initially, Mr. Tetrick challenged the grading of Questions 24,63,84, and 96 on his examination. In its review, the Staff determined that Question 24 was invalid and should be deleted from the examination. However, the result of this determination was that Mr. Tetrick's score was raised only to 78.8%, which is short of the 80% required to pass. Mr. Tetrick continues to contest the scoring of his answers to Questions 63,84, and 96 and he also is contesting the scoring 1

of his answer to Question 90. Mr. Tetrick must be sustained in at least one of the four remaining challenges to pass the examination. Below, the challenges are considered one at a time.

I. QUESTION 63 A. The Question f

i Examination Question 63 stated as follows:

4 Plant condstwns:

- Preparatwns are bemg mad.efor refuelmg operanons.

- The refuelmg rasstr usjnlled nuth the transfer tube gate salse open.

- Alarm unnum nators H 1/l. STP LO llXEL and G=W5, CNTh1T SUhlr HI 12XEL are in alarm.

Which ONE of the follonmg ts the required IhthfEDIATE ACTION m response to these comdstwns ?

a. Ver(fy alarms by checking temramment sump lesci recorder and spent fuel lesel snJacatum.

b. Sound the containment emc ation alarm.

c. Instiate containment sentilatiem notarwn.

d. Initsate comtrol room sentilation isolation B. Staff Position Staff contends 5 that the correct answer to this question is "b. Sound the confainment cracuation alarm." It relies on Procedure 0-ADM-219, 6 3.4.1 I

Afhdavit of Bnan Hughes and Thonus A. Ivebles, knuary 23,1997 (Staff Afhdavith Attachnent 2 to staff's Presentation, ai 8.120.

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(Hearing File #20. attachment 2), which states: " Respond to alarms on color code priority and plant coriditions." (Emphasis added.) Staff argues that:

The plant conditions and indicauons specified in this question (i c.. the refueling cavity hlled and the transfer tube gate vahe open with coincident SFP LOW LEVEL and CONTAIN-MENT SUMP lilGil LEVEL alarmo are mutually supportire and omfermatory, and require entry into Off-Normal Operating Procedure 3 ONOP 033 2. "Refuchng Cavity Seal Failure" 4

Oleanng File #24 ). [ Emphasis added l Staff further argues that there is only one immediate action specified for a refueling cavity seal failure. That action, which the operator must be able to perform from memory and before opening and reading the emergency

, procedures, is to sound the containment evacuation alarm. licaring File #24, 3-ONOP-033.2, at 5, 6 4.1; liearing File #25, 0-ADM-211, at II, 5 5.2.l; and llearing File #25,3-BD-EOP-E-O " BASIS DOCUMENT."

Staff stresses the importance of this immediate action. It states, in Staff Affidavit at 9, that:

] Significantly, the need for such imnwdiate action results fmm the fact that under the stated

( conditions, personnel located in the containment would quickly be exposed to high levels of radiation (due to loss of water which normally acts as . radiation shield) unless they are l prompdy notified by a containment alarm to evacuate the area.

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, Furthermore, Staff indicates that Off-Normal Operating Procedures have a high priority among plant operating procedures. licaring Record #25,0-ADM-211, at 25, 6 5.13.1.

Staff also points out that the question explicitly asks for "the IMMEDIATE i ACTION." Staff Affidavit at 10.

C. Mr. Tetrick's Position J

, Mr. Tetrick's answer was "a. Verify alarnu by checki"g containment sump level recorder and spent fuel level indication." lie relies on the CONTROL 4

ROOM ANNUNCI'. TOR RESPONSE procedure 3-ARP-097.CR to support his belief that, 'The annunciators should be verified by additional supportive

, information to preclude the possibility of annunciator failure." llearing File

1. 20, discussion of Exam Question #63; see also Tetrick Request for llearing, September 25,1996.

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" Staff refen to " hem 24." wiuch I have changed solely for the purpose of complying with the style used in this document 3

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D. Conclusion The Staff has persuaded me that when two concurrent annunciators sound, indicating that there is an off-normal event that could cause harmful radiation i within the containment, that the operator should take the required IMMEDIATE ACTION. Given the important safety problem that is being indicated by two different annunciators, that is not the time to verify that each of the annunciators is working properly. That they sound togelhcr is enough corroboration to act ,

immediately to prevent injury to the health of plant employees.  !

) Mr. Tetrick has had this Staff response available to him for some time and j has never directly addressed it. In consequence, he continues to argue for

an examination answer that could delay his action in preventing unnecessary exposure of his co-workers. I find that Mr. Tetrick's answer to this question was not correct.

I note, as wc!!, that Mr. Tetrick is incorrect in stating that 3-ARP-097.CR states "that for all alarms the ARP shall be consulted." See the ARP at 8,  !

" NOTES," at the bottom of the box. Step 2 in the notes requires that immediate corrective actions be taken as necessary. I interpret this to require that the immediate action of 3-ONOP-033.2 should be taken. The language quoted by Mr. Tetrick is from a bulleted paragraph that is part of paragraph "3) Daily Annunciator Response Pracedure Usage." I do not interpret that language to supersede or qualify in any way plant procedures that require immediate action.

II. QUESTION 84 A. The Question Examination Question 84 stated as follows:

Which ONE of the follonmg a the basu for ster I. " Vers]y Reactor Trip", ruf TR.S l.

Respemse to Nuclear Pouer GeneratwdATWS?

a. To eraure that only decos heat and reactor coolant pumps are adJung heat to the RCS. i h To ensure shutdown margm u utthin 1hhnical Spectfwattons hmus for HOT +

STANDBY. l

c. To alert the operator la sake further correctise action 4 the reactor u NOT try> ped.

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d. To senly that all automatte reactor protectsse features hase functwned as dessened.

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11. Stalf Position Staff states that the correct answer is "a." Staff argues that the question requests the basis (or reason) for Step 1 Verify Reactor Trip, of FR-S.1, 1 55 l

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i Response to Nuclear Power Generation /ATWS. To determine the basis for Step 1, I first examine Step i in the following table:

Verify reactor trip:

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• Rod bottom lights - ON Manually trip reactor. j If reactor will NOT '

• Reactor trip and bypass trip. T/IEN manually insert j breakers - OPEN control rods.

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• Rod position indicators - AT ZERO

• Neutron flux - DECREASING

, Staff asserts that the reason or basis for this step (e.g., the reason the step l

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is required) is: "a. To ensure that only decay heat and reactor coolant pumps are adding heat to the RCS [ reactor coolant system]." In support of this basis, Staff states that the reactor safeguard systems that protect the plant during an

accident are designed on the basis that there are no additional sources of heat other than those mentioned in the correct answer, a. Staff Affidavit at 11 12,

26-27; Hearing File #20, "Page 9," 3-BD-EOP-E-0, " Basis Document."

C. Mr. Tetrick's Position Mr. Tetrick asserts that a correct answer to Question #84 is, "C. To alert the '

operator to take further corrective action if the reactor is not tripped."

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D. Conclusion l 1 conclude that the basis or " reason" for Step I has been correctly specified by the Staff as specified in File #20,3-BD-EOP-E-0, " Basis Document." Since the pmcedure correctly states the " basis," a student could have answered correctly merely by learning what the procedure stated. The answer given by Mr. Tetrick is not the " basis" for Step 1. It is a followup action that might be taken after performing Step i but it is not the " basis" for that step.

111. QUESTION 90 A. The Question Examination Question 90 stated as follows:

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- _ ._ ._ _ ._m . . _ - m _ _ _ _ _ _ _ . _.-_ __ _ __ _ . _ _ _ . _

L When drammg the RCS using 3-Op4141.9. REDUCED lhYLNf0RY OPERATIONS. she reactor sessel heud and pressurner are both vented to s oniamment atmosphere. ,

Which one of theJallonmnr Jescrubes the egerts um reacwr sessel mdwation sf an adequate sent path is not prended? tAssume the reference leg remains fullt a A vacuum in the RCS loops well ressdt m level indi< atum bemg louer than actual levels.

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b. A vacuum in the RCS loops usl: result on lesel indication being lagher than actual levels.

c. A pusstive pressure m the RCS loops uall result m l vel LJucatum bemg lower than ,

actual les els. '

d The lesel instruments automatically compensate for posinse or negatne pressure.  ;

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11. Mr. Tetrick's Position '

Mr. Tetrick's argument is simple. He states:

The anumption that the reference leg remams full makes this questmn invahd At Turkey -

Pomt the drain down level indication has dry reference legs. This condition is venfied by >

0-PMI-041.110. Appheant requests that this question be deleted, C. Staff Position .

Staff states that the correct answer is:

i a A vacuum in the RCS lwps m ull result m lesel andwatum being loner than actuallewis. '

Staff concedes that at Turkey Point the draindown level indication has a dry reference leg and that the assumption that the reference leg remains full is contrary to fact. Staff Attidavit c.t 15, 1133, 35. Nevertheless, the Staff assens that the question remains valid because ~the fact that the reference leg is dry as  ;

opposed to filled with water is immaterial." Staff Affidavit at 17 39 l The purpose of this question, according to the Staff, was to test an utsder- I standing of a basic hydraulic prir.ciple, that if a vacuum is drawn above the l water level in the reactor pressure vessel, that will affect the instrument that l measures water level because it will reduce the pressute exerted by the watec in the pressure vessel. '

The important leg to consider here is the variable leg of the water-level instrument. When there is a sacuum above the water in the pressure vessel, there will be less pressure on the variable leg than if the space above the water were filled by air at atmospheric pressure. 'Ihe purpose of the " reference leg" of the pressure indicator is to measure the height of water that corresponds to l

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the pressure on the variable leg. Providing that there is no malfunction affecting the reference leg it does not matter whether the design uses a wet or a dry reference leg. The answer will be the same: an accurate measurement of the

height of th, water in the variable leg. Staff Affidavit at 16-17, t 37-39.

Staff states that

38 This quesuon tests upphcants on their understandmg of the hydrauhc effects on level indication dunng mid loop operanons U c. water level m the loop pipmg is less than fulh and other drainmg operanons if a vacuum is drawn while lowering water level Numerous incidents have occurred within the nuclear industry which invohed draining reactor coolant systems. A lack of understanding of the hydraulic etTects on level indications by operators has been a prime contributor to many of these esents. Therefore, it is important that apphcants demonstrate an understanding of this problem. as examined on ttus questmn.

(Emphasis added.)

D. Conclusion

On this question, I agiee with the Staff. The question is poorly worded.

containing an assumption that is different from the plant configuration. This 4 could have been somewhat confusing to Mr. Tetrick.

However, I have decided that if Mr. Tetrick had a basic knowledge of the a principles that affect water-level indication, he should have realized that the er. tire purpose of the reference leg of the water level mdicator is to measure the height of water in the variable leg. Since the pressure exerted by the column of water in the variable leg would be reduced by the vacuum above the water in the reactor pressure vessel, the water levelindicated by the instrument uvuld be tourr shart the water level in the reactor vessel. Given the importance of this principle, I conclude that Mr. Tetrick should be able to understand it and answer the question correctly. There is no explanation for the answer he gave:

, that the water level indication would be higher than actual levels.

) I conclude that, despite the contrary-to-fact predicate that makes this question

more dillicult than intended, Mr. Tetrick should have answered it correctly. The i qu. stion is valid and Mr. Tetrick's answer is wrong.

IV. QUESTION 96

A. The Question Examination Question 96 stated as follows:

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Which UNE of the following u the lowea levelposition responsablefor ensunny entrics are made in the TechnicalSpecificanon Related Equspment Out-of Senice Index?

a. Nuclear Plant Supenosor I i

& A htimt Nuclear Plant Supervisor i

c. Senior Nuclear Plar.: Operator

d. Nuclear Watth Engineer l

B. Staff Position .

Staff states that the correct answer is "b. Assistant Nuclear Plant Supervisor."

Staff states that l

Procedure 0-ADM 213. " Technical Specification Related Equipinent and Risk Sigmhcant 4

S C. Out of Service Logbook." states that the ALPS is the lowest level position responuble for entenng inoperable equipment in the subject index (ltem 24). When the NWE [ Nuclear Watch Engmeer) reheves the ALPS. he then anumes the poution of the ALPS. The NWE is not authorized to anale entnes in the subject index unless he is acung in the capacity of the ALPS. any more than he would be able to exerciw any other functions of the ALPS unless '

l.e is acting in the ALPS capacity.

C. Mr. Tetrick's Position Mr. Tetrick states that "d. Nuclear Watch Engineer" is also correct because procedure 0-ADM-200 makes the Nuclear Welch Engineer (NWE) responsible "far routinely relieving the Assistant Nuclear Plant Supervisor (ALPS) of the control rcom command and control function to enable the ALPS to ! cave the control room." [ Emphasis added.] Staff does not question Mr. Tetrick's statement that this substitution is authorind and routine.

D. Conclusion I conclude that the question is ambiguous and should be struck.

Mr. Tetrick has reasonable ground to consider his answer to be correct. I do i not think it necessary to address the following metaphysical question: Is the i Nuclear Watch Engineer still at least in part a Nuclear Watch Engineer when he relieves the Assistant Nuclear Plant Supervisor? Staff apparently thinks that the Nuclear Watch Engineer completely loses his ordinary job identity when he i acts as a substitute for the Assistant Nuclear Plant Supervisor. While that is a plausible way to view what happens, I do not think it fair to require Mr. Tetrick to adopt that view of the use of words in order to pass his examination. The question in its current form is ambiguous and invalid. f 59

Mr. Tetrick has answered correctly 78 of 98 questions. His score, rounded to the nearest tenth of a percent is 79.6L I note that for the examination question to have the unambiguous meaning given to it by the Staff, it could have said: "Which ONE vf the following is the lowest levelposition that one mu st have (or be acting as)for ensuring entries are made in the Technical Specification Related Equipment Out of-Service Index?"

V. OVERALL CONCLUSION I have determined that Mr. Tetrick was correct in 78 of 98 valid questions on his examination. Staff has not addressed the question of the number of digits in the examination score that should be considered significant. Because I have not been directed to any governing guidance or regulation I have decided that it is appropriate to round up the answer to the nearest integer. These tests are not so precise that tenths of a percent have any meaning. Consequently, Mr. Tetrick's score is 80%, which is a passing score. He shall, therefore, be granted a license as a Senior Reactor Operator.

VI. ORDER For all the foregoing reasons and upon consideration of the entire record in this matter, it is, this 28th day of February 1997, ORDERED that:

1. The Staff of the Nuclear Regulatory Commission may issue to Mr.

Ralph L. Tetrick a Senior Reactor Operator License for Turkey point Nuclear Generating Plant, Units 3 and 4.

2. Pursuant to 10 C.F.R. 5 2.1251, this initial Decision constitutes the final action of the Commission thirty (30) days after the date of issuance, unless any party petitions for Commission review in accordance with section 2.786 or the Commission takes review of the Decision sua sponte. If there is no petition for review, the date on which this Decision will become final is Monday, March 31,1997,

3. Pursuant to 10 C.F.R. 5 2.786, a petition for review must be filed within fifteen (15) days after service of this Decision. which is considered served on the date it is mailed, pursuant to 10 C.F.R. 5 2.712(e). However, since service of this Decision is by mail, five days shall be added to the prescribed period of response, pursuant to 10 C.F.R. 6 2.710, which governs the computation of time. Consequently, the date the petition for review must be served is Thuisday, March 20. Service of the petition for review must, pursuant to this Order, be made by express mail.

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4. A petition for review and a response to a petition for review must meet

the requirements of 10 C.F.R. 5 2.786.

5. {

If a petition for review is filed, the answer must be fiicd within 10 days.

I Since the petition for review shall be filed by express mail, two days shall be ,

added to the period of response pursuant to 10 C.F.R. s 2.710, which governs i the computation of time. Consequently, the date the answer must be served is 4

Tuesday, March 16, 1997. Service of the answer must, pursuant to this Order,

be made by express nuit.

\

Peter B. Bloch, Presiding Officer  :

i ADMINISTRATIVE JUDGE '

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. Rockville, Maryland I 1

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i Directors' '

Decisions i

Under 10 CFR 2.206 4

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1- Cite as 45 NRC 63 (1997) DD-97-2 l

UNITED STATES OF AMERICA

{ NUCLEAR REGULATORY COMMISSION

! OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS 1

4 l Carl J. Paperiello, Director in the Matter of Docket No. 40-8089

[ (License No. SMC-1559) i.

, ENVIROCARE OF UTAH, INC. February 5,1997 i The Director Office of Nuclear Material Safety and Safeguards, has denied a i petition filed by Dr. Thomas B. Cochran on behalf of Natural Resources Defense j Council (NRDC) requesting that the NRC take action regarding Envirocare of i Utah, Inc. (Envirocare). The petition requested that the NRC immediately revoke l any license or cause the State of Utah (Utah) to revoke any Agreement State i license or licenses held by Envirocare, its President, Khosrow Semnani, or any j entity controlled or managed by Mr. Semnani; prohibit the future issuance of

j. any license by the NRC, Utah, or other NRC Agreement State to Mr. Semnani j or any entity controlled or managed by him or with which he has a significant i affiliation; and suspend Utah's Agreement State status until it can demonstrate l that it can operate its Division of Radiation Control in a lawful manner. As a i basis for the petition, the Petitioner asserted that an article in the Salt Lake City j Tribune reported secret cash payments made by Mr. Semnani to the Director of the Utah Division of Radiation Control, and Utah's initiation of a criminal j investigation into the matter. The reasons for the denial are set forth in the Decision.

ATOMIC ENERGY ACT: ENFORCEMENT ACTION (HEARING RIGHT)

"Ihe Commission's regulations recognize that a licensee should be afforded under usual circumstances a prior opportunity to be heard before the agency suspends a license or takes other enforectrent action, but that extraordinary circumstances may warrant summary action prior to hearing.

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RULES OF PRACTICE: SIIOW CAUSE PROCEEDING Since the inception of the 10 C.F.R. 6 2.206 process, the Commission has consistently stated that the purpose of 10 C.F.R. 6 2.206 is to provide the public with the means for participating in the enforcement process.

RULES OF PRACTICE: SHOW-CAUSE PROCEEDING In accordance with the Commission's determination that the section 2.206 process should be focused on requests for enforcement action rather than an evaluation of safety concerns, petitions will be reviewed under 10 C.F.R. 9 2.206 if the request is for enforcement action, and a request under section 2.206 should be distinguished from a request to deny a pending license application or amendment.

DIRECTOR'S DECISION UNDER 10 C.F.R. 6 2.206

1. INTRODUCTION in a letter dated January 8,1997. Dr. Thomas B. Cochran, Director of Nuclear Programs, Natural Resources Defense Council (NRDC), requested, under 10 C.F.R. I 2.206 of the Commission's regulations, that NRC take action to revoke all licenses held by Envirocare of Utah, Inc. (Envirocare). Specifically, the petition requested that "NRC take the following actions:"

1) Immediately revoke the license or licenses, or cause the state of Utah to revoke its agreement state license or licenses, under which Envirocare is currently permitted to accept low level radioactive waste and mixed waste for permanent disposal.

2) Immediately revoke the NRC i1.e(2) byproduct material license under which Envirocare is currently permitted to accept uranium mill tailings for disposal.

3) Immediately revoke any other NRC license, or agreement state license, if such license exists, held by Envirocare, Khosrow Semnani, or any entity controlled or managed by Khosrow Semnani.

4) Prohibit the future issuances of any license by the NRC, the State of Utah, or other NRC agreement state, to Khosrow Semnani or any company or entity which he owns, controls, manages, or [with which he] has a significant affiliation or relationship.

5) Suspend the agreement with the state of Utah under which regulatory authority has been transferred from the NRC to the Utah's [ sic] Bureau of Radiation [ Division of Radiation Control), until the state of Utah can 64

. demonstrate that it can operate the Bureau of Radiation [ Division of Radiation Control] in a lawful manner, and without the participation of licensees, or employees of licensees, in Bureau of Radiation [ Division of Radiation Control] oversight roles."

NRDC asserts, as a basis for the request, that a December 28,1996 article in 7he Salt Lake Tribune reported that between 1987 and 1995, Mr. Semnani made secret cash payments to Mr. Larry F. Anderson, who served as Director of the Utah Division of Radiation Control (UDRC) from 1983 until 1993. The article also reported that the Utah Attorney General's office has initiated a criminal investigation into the matter.

Although NRDC's request that NRC suspend its agreement with the State of Utah, or cause Utah to revoke the license that it issued, does not squarely fall within the scope of matters ordinarily considered under section 2.206,' the Staff has evaluated the mer;ts of those requests. This evaluation is contained in a separate "NRC Staff Evaluation of Natural Resources Defense Council Request to Suspend Section 274 Agreement with the State of Utah." This Director's Decision will address the NRDC requests that relate to the license to receive, store, and dispose of certain byproduct material issued to Envirocare by NRC, pursuant to section ll.e(2) of the Atomic Energy Act of 1954 (AEA),

as amended.

11. BACKGROUND Envirocare operates a radioactive waste disposal facility in Clive, Utah, 128 kilometers (80 miles) west of Salt Lake City in western Tooele County. "

Radioactive wastes are disposed of by modified shallow land burial techniques.

Envirocare submitted its license application to the NRC in November 1989 for commercial disposal of 11.e(2) byproduct material, as defined in section 11.e(2) of the AEA. On November 19,1993, NRC completed its licensing review and issued Envirocare an NRC license to receive, store, and dispose of uranium -

and thorium byproduct material. Envirocare began receiving II.e(2) byproduct material in September 1994 and has been in continuous operation since.

To ensure that the facility is operated safely and in compliance with NRC requirements, the Staff conducts routine, announced inspections of the site.

Areas examined during the inspections include management organization and controls, operations review, radiation protection, radioactive waste management, transportation, construction work, groundwater activities, and environmental l

NRC Manual threcove 611. " Review Process for 10 Cf R 2 206 Peutsons." assued september 2A 1994 (revised December 12.1995). states that the scope of the section 2.206 process as hnuted to requests for enforcement acuan agamst beensees or enuues engaguig m NRC.hcensed actmtics But see Staic of t/tah (Agreenent Pursuant to secuon 274 of the Atonuc f.nergy Act of 1954 as Amended). DD 95 l Al NRC 43 0995r l

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monitoring. The NRC has conducted five inspections of the Envirocure facilities and has cited the Licensee for three violations. All violations were categorized in accordance with the guidance in NUREG-1600, " General Statement of Policy and Procedures for NRC Enforcement Actions" (Enforcement Policy) at a Severity Level IV.2 The first violation, issued as a result of a July 1995 inspection, and the second violation, issued as a result of a July 1996 inspection, have been adequately resolved by Envirocare. The last inspection, conducted on November 18-22,1996, resulted in the issuance of the third citation noted above.

This violation involved a failure to develop and implement, in a timely manner:

(1) site-specific standards for three constituents found in the groundwater that exceeded their baseline values, and (2) a Compliance Monitoring Plaa for arsenic after it was found to exceed its baseline value. These results of the November 1996 inspection are documented in Inspection Report 40-8989/96-02 which was issued on January 28,1997. The NRC is in the process of determining whether Envirocare has taken appropriate action to correct this violation.

In addition, the November 1996 inspection identified other areas of concern where the Staff determined that additional evaluation was necessary. As a result, a followup inspection was conducted the week of January 27, 1997. Areas that were examined during this inspection included: (1) the Licensee's quality assurance / quality control program; (2) the Licensee's review of changes made to the facility; and (3) contractor laboratory certification. The results of the January 27,1997 inspection are currently being evaluated. Once this evaluation is complete, the NRC will document the results in an inspection report. Based on a preliminary review of the inspection results, no significant violations were identified.

III. DISCUSSION In December 1996, the Salt hke Tribune published a series of articles that questioned the relationship between Larry F. Anderson, former Director of UDRC, and Khosrow Semnani, President of Envirocare, during the licensing of the low-level radioactive waste (LLW) disposal facility. Subsequently, the NRC Staff learned that on May 16,1996, Larry F. Anderson sited a complaint against Khosrow B. Semnani in the Third Judicial District Court of Salt Lake County, State of Utah, to obtain compensation for alleged consulting services in the sum of 5 million dollars. The complaint alleges that, while Director of UDRC, Mr.

Anderson recognized the need for an LLW site in Utah; incorporated a consulting 2

As explained in secuan IV of the Enforcement Pohey, vmlanons are nonnally caiegonzed in terms of four levels ,

of seventy. A Sewnty tevel IV violanon is defined as a vmlatmn of note than minor concern which. af left  !

uncorrected. could lead to a nmre menous concern 66

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4 firm, Lavicka, Inc., for the express purpose of developing a plan for siting the facility; and entered into a business arrangement to provide Mr. Semnani with l

a license application and consulting services. Mr. Anderson alleges that Mr. ,

Semnani, President of Envirocare, agreed to pay a consulting fee of 100,000 '

i dollars and an ongoing remuneration of 5% of all direct and indirect revenues that l Mr. Semnani would realize from such a facility, if the site were successful. The i complaint contends that Mr. Semnani owes Mr. Anderson unpaid compensation for consulting services in the sum of 5 million dollars.

I In October 1996, Mr. Semnani filed a counterclaim in the court, denying Mr.

Anderson's claim and alleging that, in fact, Mr. Anderson used his position as the

Director of UDRC to extort money in the sum of 600.000 dollars. Mr. Semnani l contends that all the money he paid was based on the belief that if he did not  ;

pay, Mr. Anderson would use his official position and capacity as an officer and 4 employee of the State of Utah to deny Mr. Semnani fair consideration, review, hearing, and determination on his license application and, thereby, cause the license not to be granted, or, if Envirocare was granted a license, Mr. Anderson would use his position to subject the facility to unfair and biased oversight and supervision of the operation of the facility under the license. As a result of these

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j allegations, the Utah Attorney General's office is investigating the relationship l between Mr. Semnani and Mr. Anderson. '

De NRDC petition is based on the events described above. The NRC has evaluated the NRDC's requests and found no basis to take the requested actions.

As an initial matter, NRDC requests that the NRC immediately revoke the NRC i1.e(2) byproduct material license under which Envirocare is currently permitted to accept uranium mill tailings for disposal. In addition, NRDC also l asks that the NRC immediately revoke any other NRC license, or agreement l State license, if such license exists, held by Envirocare, Khosrow Semnani, or l any entity controlled or managed by Khosrow Semnani.

! He NRC's Enforcement Policy describes the various enforcement sanctions i available to the Commission once it determines that a violation of its require-l ments has occurred. In accordance with the guidance in section VI.C.3 of the ,

l Enforcement Policy, Revocation Orders may be used: (a) when a licensee is l unable or unwilling to comply with NRC requirements; (b) when a licensec

! refuses to correct a violation; (c) when a licensee does not respond to a Notice of Violation where a response was required; (d) when a licensee refuses to pay l an applicable fee under the Commission's regulations; or (e) for any other rea-son for which revocation is authorized under section 186 of the Atomic Energy l l Act (e.g., any condition that would warrant refusal of a license on an original ,

application). Pursuant to 10 C.F.R. 6 2.202(a)(5), the Commission may issue l

! an immediately effective order to modify, suspend, or revoke a license if the Commission finds that the public health, safety, or interest so requires or that the violation or conduct causing the violation was willful. De Commission's regu- j i

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lations recognize that a licensee should be afforded under usual circumstances a prior opportunity to be heard before the agency suspends a license or takes other enforcement action, but that extraordinary circumstances may warrant summary action prior to hearing. See Advanced Medical Systems, Inc. (One Factory Row, Geneva, Ohio 44N1), CL1-94-6,39 NRC 285,299 (1994).

In this case the NRDC has not provided the NRC with specific information establishing that a violation of NRC requirements has occurred, nor provided the NRC with any other information that would provide a basis for immediate suspension of the Envirocate license. As NRDC notes in its request, the Utah State Attorney General has initiated a criminal investigation into the matter of the relationship between Mr. Anderson and Mr. Semnani. Aosent specific information supporting the existence of such extraordinary circumstances as would warrant such action, NRC believes that it would be premature to initiate immediate action pending completion of this investigation. We recognize that this matter involves potential issues of integrity, which, if proven, may raise questions as to whether the NRC should have the requisite reasonable assurance that Envirocare will comply with Commission requirements. 'NRC intends to follow the investigation of the State Attorney General closely, if NRC receives information of public health and safety concerns during the investigation or on its completion, or receives such information from other sources, including NRC's ongoing Agreement State oversight activities, it will evaluate that information and take such appropriate action at that time as may be warranted.

Furthermore, the NRC Staff has reviewed the bases for its licensing actions involving Envirocare, and confirmed that NRC did not rely on technical eval-uations performed by the State to reach a decision regarding the evaluation of Envirocare's 11.e(2) byproduct material license. The Staff conducted an inde-pendent technical evaluation of Envirocare's hcense application and subsequent amendment requests, and concluded that Envirocare had adequately demon-strated compliance with all applicable health and safety standards and regula-tions. In addition, as noted above, NRC inspections of Envirocare have not revealed significant violations that would warrant immediate action.

I Moreover, with regard to NRDC's request that the NRC immediately revoke ,

any other license, the NRC has issued no other license to Envirocare, Khosrow Semnani, or any entity controlled or managed liy Khosrow Semnani. For these reasons, this request is denied.

NRDC also requests that the NRC prohibit the future issuances of any license i by the NRC, the State of Utah, or other NRC agreement state, to Khosrow l Semnani or any company or entity that he owns, controls, manages, or with which he has a significant affiliation or relationship.

With regard to this request, we have already noted that there is no basis for NRC to take immediate action. In any event, section 2.206 is not a venue for presenting licensing contentions of the sort raised by this aspect of NRDC's 68 i

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4 petition. Section 2.206 provides for requests for action under that portion of

, the NRC's regulations governing enforcement actions, namely 10 C.F.R. Part  !

j 2, Subpart B. Subpart B is entitled " Procedure for Imposing Requirements by Order, or for Modification, Suspension, or Revocation of a License, or for Imposing Civil Penalties." Since the inception of the section 2.206 process, the Commission has consistently stated that the purpose of section 2.206 is

to provide the public with the means for participating in the enforcement i

process.) The Commission has determined that the section 2.206 process should be focused on requests for enforcement action rather than evaluations of safety concerns. In accordance with this determination, the Commission's Management j Directive 8.11 " Review Process for 10 C.F.R. 2.206 Petitions," Part 111, section

, A, states that petitions will be reviewed undcr section 2.206 if the request

is for enforcement action, and that a request under section 2.206 should be l

distinguished from a request to deny a pendi g license application or amendment.

Because this request by the NRDC ancerns licensing-type action, not i i enforcement-type action, the Staff has de ermmed that, consistent with the j guidance of Management Directive 8.11, this request is not within the scope

! of section 2.206.4 To the extent that further facts may be developed that may

! warrant consideration of this request, the matter may be raised in an individual licensing proceeding; however, no such proceeding is presently pending, as there j is no application pending for the issuance of a license to Envirocare.

IV. CONCLUSION On the basis of the above assessment, I have concluded that no substantial

! health and safety issues have been raised regarding Envirocare that would I require initiation of the immediate action requested by the NRDC, and the petition is therefore denied. As explained above, the NRDC has not provided any information in support of its requests of which the NRC was not already i aware. Moreover, NRC inspections of the Envirocare facility have not revealed f

the existence of extraordinary circumstances that would warrant immediate suspension of the Envirocare license. In addition, the Staff's review of the technical basis for its issuance of the license and subsequent amer.iments found no evidence of the existence of any substantial heahh or safety issue that would justify the actions requested by the NRDC. NRC will monitor the investigations 3 -Requeses to impose Requirenruts by order on a licensee. or to Mo#y. suspen,1 or Revoke a tieense." 39 fed.

Reg.12,353 (Apnl 5,1974L "teBoeuf. Lamb. Imby & Macrae." di led Reg 3359 dan. 22.1976). "Peutions for Review of Director's Demal of Enforcenent Requests." 42 Fed Reg 36.239 Ouly 14.1977).

4 Even if this request were iruerpreted as a request that the NRC issue an enforcenwns order prohihmng Mr.

senmam from engagmg m heensed acuvmes. and thus consutute a request for enforcenwnt acuon withm the scope of secuon 2.206. NRDC has not provided the NRC with speahe informanon such as would marrant the requested acnon, as esplamed above 69 I

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and actions being conducted by the State of Utah. If NRC receives any specific

! infonnation that there is a public health or safety concern as a result of these actions or from any other source, including the NRC ongoing Agreement State oversight activities, NRC will evaluate that information and take such action as >

it deems is warranted at that time.

I FOR THE NUCLEAR REGULATORY COMMISSION Carl .I. Paperiello, Director i Office of Nuclear Material Safety l and Safeguards l

l Dated at Rockville, Maryland, <

l this 5th day of February 1997.

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, Cite as 45 NRC 71 (1997) DD-97-3 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION 4

OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS l

4 l Carl J. Paperiello, Director

)

! In the Matter of Docket Nos. 50-346 72-1004 TOLEDO EDISON COMPANY, et al.

(Devis-Besse Independent Spent Fuel

, Storage Installation) February 5,1997 The Director of the Office of Nuclear Material Safety and Safeguards grants, in part, and denies, in part, a petition filed pursuant to 10 C.F.R. 6 2.206 on behalf of the Toledo Coalition for Safe Energy, Alice Hirt, Charlene Johnston, Dini Schut, and William Hoops. The petition is granted to the extent that the NRC has initiated a rulemaking to modify the Certificate of Compliance for the VECTRA Technologies NUHOMS-24P dry shielded canisters (DSCs) in order to require fabrication inspection. The Petitioners

• request that the NRC require the unloading of DSCs pending completion of the rulemaking is denied. The Director also finds no basis for taking any further enforcement action against VECTRA or to require the halting of the ISFSI operation at Davis-Besse.

DIRECTOR'S DECISION UNDER 10 C.F.R. 5 2.206 INTRODUCTION By a petition dated December 5,1995, filed on behalf of the Toledo Coalition for Safe Energy, Alice Hirt, Charlene Johnston, Dini Schut, and William Hoops (Petitioners),8 the U.S. Nuclear Regulatory Commission (NRC) was asked I

According to the peuuon. tte Toledo Coahuon as a grauroots anunuclear orgamzauon with nrmbers who reside wittua a M-nule radius of the Davis-Besse Nuclear Power Station. The peuuan indicates that it is also offenng the positwns of the Maryland Safe Energy Coahuon, an organizanon represemed to have nrinbers near the Calwn Chffs nuclear plant, another sue where NUHoMs.24P dry storage carusiers are bemg used.

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immediately to issue an order to prevent the loading of spent nuclear fuel into the VECTRA Technologies Inc. (VECTRA), NUHOMS-24P dry-shielded canisters (DSCs) at the Davis-Besse Nuclear Power Station until the NRC conducts a rulemaking and/or license modification hearing on all safety-related changes that have been made to the DSCs, as described in the Safety Analysis Report (SAR). Also, the NRC was requested not to authorire any loading of the DSCs until a written procedure for unloading them, in both urgent and nonurgent circumstances, was written, approved, and field tested.

Petitioners contend that the safety of the DSCs has been compromised because of certain reductions that were made by VECTRA in the thickness of the welds in the DSC metal walls. In addition. Petitioners question the legal validity of the administrative and regulatory processes used by NRC after discovery of the DSC wall-thickness issue. They assert that an agency rulemaking or other public process is required for the DSCs at the Davis-Besse site.

The petition was referred to me pursuant to NRC regulations in 10 C.F.R. 5 2.206,2 Because the petition requested immediate relief (i.e., a halt to any loading of the DSCs at Davis-Besse), it was necessary for me to give an immediate response to that portion of the Petitioners' request. By letter dated December 18, 1995, I denied the Petitioners

• request for immediate action on the petition on the basis of my judgment that there was (and continues to be) no imminent risk to health, safety, or environment such as to warrant the emergency relief sought by the Petitioners.3 By letter dated January 23,1996, to Mr. Lodge, on behalf of the Petitioners, I formally acknowledged receipt of the petition. Notice of receipt was published in the Federal Regisfer on January 30, 1996 (61 Fed. Reg. 3060).

Based on the NRC Staff's evaluation of the issues and for the reasons given below, I have now concluded that the Petitioners' request should be granted in part and denied in part.

BACKGROUND NRC regulations contain a general license that authorites nuclear power plants licensed by NRC, such as Davis-Besse, to store spent nuclear fuel at a reactor site in storage casks approved by NRC. See 10 C.F.R. 66 72.210 and 72.212. Among 2

Secuon 2.206 provies that "falny person may file a request to insutute a proceeding . . . to nahfy. suspend, or revoke a hcense. or for such other acnon as may be proper." The Director of the NRC office with responsibibty for the subject marter of the request -in dus case, the Of6ce of Nuclear Matenal Safety and Safeguards -is to decide whether to msutute the requested procerchng and. af no proceed ng as insututed, will provide the reasons for the Deciuon 3

My December 18 letter also notthed Peutmners of my intennon to treat their December 5 request as a petiuon under 10 C1 R I 2.206 and mdicated that NRC would respond to the legal and technical issues they raised within a teamnable ume.

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other things, the Licensee is required to conform to certain NRC conditions for ensuring safe storage and to notify NRC at least 90 days prior to the first storaFe of spent fuel under the general license. By letter dated June 30,1995. Toledo Edison Company (Licensee) informed NRC that it planned to use the VECTRA Standardized NUllOMS-24P dry spent fuel storage system (NUHOMS) under the general license at the independent spent fuel storage installation (ISFSI) facility at the Davis-Besse Nuclear Power Station. VECTRA's NUHOMS had previously been approved by NRC in December 1994 (59 Fed. Reg. 65,898) and as further reflected by the issuance of NRC Certificate of Compliance No.1004 (COC) to VECTRA, the cask vendor. This NRC approval was granted after notice-and-comaent rulemaking, to allow use of the NUHOMS system (subject to conditions specified in the COC) to store dry spent fuel at a nuclear power reactor site under the terms and conditions of the general license in 10 C.F.R. Part 72.

NRC regulations require cask vendors, such as VECTRA, to permit NRC to inspect the premises and facilities at which NRC-approved storage casks are fabricated and tested. See 10 C.F.R. 0 72.232. On June 20-23,1995, NRC conducted an inspection of VECTRA's contractor, Ranor, Inc., at ' trninster, MA. At (nat time, Ranor was fabricating the three NUHOMS ' and the transfer cask (TC) for VECTRA that were destined for D . esse. The objective of the NRC inspection was to confirm that activities associated with the fabrication of the DSCs and TC had been executed in accordance with the requirements of the NRC COC and commitments made by VECTRA in the

" Safety Analysis Report for the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel" (SAR).d VECTRA /Ranor was

fabricating the DSCs and the TC for Toledo Edison (Davis-Besse site).

j 'the NRC inspection identified three items of concern that required further j action by VECTRA: (1) there was inadequate documentation to demonstrate j that changes made by VECTRA /Ranor to the storage cask design described in i the SAR had been reviewed and evaluated by the cask vendor in accordance with 1 Condition 9 of the COC;' (2) cask wall-thickness measurements had not been l

taken by VECTRA /Ranor after welding and grinding operations were performed I

'Under NRC regulanons, a cask vendor who requests NRC approval of a spent fuel storage ek niusi subnut f an appheatmn that includes a SAR descnbing the proposed cask design and how the cask should be used to store j spnt fuel safely See 10 C r R. I72 UO The NRC report to VECrRA Oce July 7.1995 CAL)inaccurately desenbed the correcuve acuan as lollows-

"VLCTRA will provide to the NRC wruten nochcanon that the safety evaluanons consistent with 10 CtR 72 48 i have been cornpleted and no unresolved safety issues were idennhed pnor to shipping the DSC: and the TC." In

, fact VECrRA was required to provide (and ulumately ed provuk) safety evaluauons " consistent with Con &uun 9 of the LOC " Conatwa 9 and 10 C F R l 72 48 are substanuvely sinular in that each pernuts changes to the cask design desenbed in the SAR, without pnor NRC approsal. af cenaan specahed con &uons are met and documented by a wrmen safety evaluauon However. Con &uon 9 apphes to changes by the cak vendor (i c.. VLCIRA),

whereas secuon 72 48 apphes to changes by the beensee O c.. Toledo Edson).

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on the DSCs;' and (3) leak testing was performed on the DSCs in lieu of pressure testing.7 On July 7,1995, NRC issued a Confirmatory Action Letter (CAL) to VECTRA. confirming VECTRA's commitment to take actions to resolve the above three items of concern. Among those actions, as listed in the CAL, the following actions are related to Davis-Besse's ISFSI operation. l

1. Regarding the finding of inadequate documentation of design changes, VECTRA was to review evaluations for adequacy and complete the documentation packages. VECTRA was to provide to the NRC written notification that the safety evaluations were completed and that no unresolved safety issues were identified prior to shipping the three DSCs and E to Davis-Besse.

2. Regarding the finding on the lack of wall-thickness measurements after welding and grinding operations, VECTRA was to inspect welded areas in the DSCs to determine actual wall thickness and prepare an i engineering document providing an evaluation of the safety significance of any wall thinning below design specifications. VECTRA was not j

to ship the three DSCs affected by wall thinning until this issue was resolved with NRC.'

3. Regarding the finding on performing leak testing instead of pressure testing, VECTRA was to provide to NRC an engineering evaluation justifying the use of a leak test in lieu of a pressure test. VECTRA was ,

not to ship DSCs until this issue was resolved with NRC.

It is item 2 above - the absence of DSC wall-thickness measurements by VECTRA - that relates to the major issue of this petition.

As to item 2 of the CAL, on September 5,1995, VECTRA informed NRC that i the maximum thickness measured in the three DSCs prepared for Davis-Besse was 0.682 inch and occurred off the weld seam and in the base metal. VECTRA said that the minimum thickness measured in the three DSCs was 0.581 inch i

' VLCTRA's NUHoMS design desenbed in the S AR uses a nonunal DsC shell thickness of 0 625 inch However.

VLCrRA/Rame had noi neasured the actual thickness of the labncated DsC shells after welding and gnndmg ytanons to venfy that it conforned to the desenpuon in the SAR.

As indicated in the SAR. the DSCs are designed, with one escepuon. as F. .sure vessels an accordance with the apphcable scenons of the Anencan Society of Mechanical Engmeers ( ASML) Baler and Pressure Vessel (B&PV) Code. The ASME B&PV Code calls for proof-pressure Ecsting of the vessel The one cucepuon is the DSC top and bottom closure welds to which the ASME B&PV Code cannot pracucably be apphed "The CAL aho requned VECIRA to evaluate the potennal safety miract of the lack of wall-ttuckness nrasurements on previously fabncated DSCs which were shipped to sites other than Davn Besse and to provide an enprwering analysis and any recomnwnded acunns resulung from that analysis to NRC By letter daied August 7.1995, ViCIRA subnutted an action plan to address the issue related to those prev:wsly fabncated DSCs. such ,

as those at the Calven Chffs site. Subsequently, VLCrRA subnuned informanon for si.f review in letters dated octobei 2.1995. March 8.1996. and Apnl 25, 1996. The Staff evaluaied the subnuned enfwmanon. and by letter Jated January 3.1997, inforned VLCIRA tlut the cat. Issues were resolved and, therefore. closed. Given at - Silowup acuvines of VLCTRA and NRC already under way pursuani to the CAL and the absence of any addiuonal informauon or claims m the peution relaung specifically to Calvert Chfts, I see no basis to take any further acnon at thu ume with ergard to Calven Chffs.

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and occurred in the weld scam of one of the DSCs. VECTRA also performed calculations that demonstrated that a DSC of 0.500-inch uniform wall thickness still met all ASME Code stress allowables, although the original design shell thickness in the SAR is 0.625 inch. In essence then, when it performed the required measurements of the three DSCs fabricated for Davis-Besse, VECTRA found actual, minimum wall thicknesses in each of the DSCs that were less than the 0.625-inch nominal thickness described in the SAR and a minimum thickness in one DSC of 0.581 inch. VECTRA thereafter went on to analyze whether a thinner wall design of 0.500 inch would satisfy NRC design criteria.

The results of VECTRA's analysis submitted to NRC on September 5,1995, showed that it would.

On October 12, 1995, NRC responded to the VECTRA actions taken in response to the CAL. Regarding item 2 of the CAL (the lack of wall-thickness measurements and VECTRA's subsequent September 5,1995 reevaluation),

NRC accepted VECTRA's 0.500-inch uniform wall-thickness calculation as meeting the ASME Code stress allowables, the onginal structural design criteria for the three DSCs. NRC said the structural capability of C e DSCs would not be compromised if wall thinning from weld grinding were limited to local spots along weld seams and if the remaining shell thickness was 0.500 inch or more.

However, NRC said that, because of the limited experience in performing weld-thickness measurements, "it is prudent to require a minimum weld inspection threshold thickness of 0.563 inches," to maintain a 0.063-inch fabrication margin over the 0.500-inch minimum. The NRC Stafi prepared a safety evaluation dated October 5,1995, documenting the basis for its acceptance of VECTRA's response to item 2.

NRC's October 12,1995 response also found that VECTRA had acceptably addressed items I and 3 in the CAL. Thus, based on the actions taken by VECTRA and NRC's independent evaluation of the technical issues and review of the supplementary documentation provided, NRC found that VECTRA had acceptably completed the actions specified in the CAL and could, therefore, ship the three DSCs and the TC to the Davis-Besse site. VECTRA shipped the DSCs and TC to Davis-Besse shortly thereafter.

On November 14, 1995, one of the Petitioners (Ms. Charlene F. Johnston) wrote NRC asking for clarification on certain questions relating to the following issues: (1) whether an amendment process is required for the change in the wall thickness of the DSCs at Davis-Besse, and (2) whether the legality of a vendor's changes to a cask design can be questioned because the vendor is not a utility licensee and, therefore, cannot use the provisions of scetion 72.48 in making changes. Since the petition covers issues that are related to the two issues in the November 14 letter - and adds a third issue on cask unloading procedures -I have decided to include my response to the November 14 letter in this Decision.

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DISCUSSION He petition and associated November 14 letter raise three issues involving the DSCs at Davis.Desse. First, Petitioners contend that the reduction in the DSC shell thickness to less than 0.625 inch compromises the safety of the DSCs. Second, Petitioners question the legal validity of the administrative and regulatory processes used by NRC after discovery of the DSC shell. thickness issue and assert that an agency rulemaking or other public process is required for the DSCs at the Davis-Besse site. Finally, Petitioners contend that NRC should have reviewed and approved and field tested the procedure for unloading the DSCs both in urgent and nonurgent circumstances prior to the operation at the Davis-Besse site. In the following discussion, I will address each of these issues in turn.

A. Reduction of Shell Thickness Does Not Compromise the Safety of the DSCs Petitioners claim that "the reduction in the thickness of the DSC metal walls to less than 0.625 inch compromises the safety of the DSCs." Petition at 1. For the reasons,tbat follow, I conclude that the change will not compromise safety.

I begin by discussing the safety function of the DSC.

He DSC shell provides a key confinement barrier for the spent fuel stored inside the NUHOMS dry cask. Thus, the DSC shell helps to ensure safety for d.y cask storage and protection of public health and safety by maintaining safe confinement of the stored fuel despite the forces, pressures, and stresses that are constantly acting on the cask (including the DSC shell) dismg normal handling, as well as during anticipated occurrences or potential task accidents.

It is logical for Petitioners to conclude that, by reducing the thickness of the DSC shell, VECTRA could adversely impact the DSC's capability as a safe confinement barrier. Indeed, it may seem obvious that a DSC having a shell thickness of 0.625 inch would have more capability to withstand cask bumps, drops, and pressure extremes than a DSC shell of reduced weld seam thickness no matter how small or limited the areas of thinning might be. Thus, at the core of Petitioners

• claim is the intuitive assertion that VECTRA's change in the DSC shell thickness lessened the DSC's capability as a confinement barrier to some extent. The question raised, but not answered, by the petition is whether this teduction in capability is sufficiendy great to compromise safety. I conclude that it is not.

In NRC's original evaluation, when it certified the NUHOMS and accepted VECTRA's SAR in 1994, the NRC Staff reviewed a variety of potential cask accidents (e.g., a cask drop or tipover, vent blockage leading to cask heatup, l low temperatures, earthquakes and tornadoes, explosions, lightning, floods) that 76 i

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were thought to cover the range of cask accidents that might reasonably be assumed to occur. In the NRC review, the a;cident was assumed to occur (i.e.,

probability of occurrence was assumed to be one), and the consequences were evaluated. For each accident, the NRC Staff review found that the DSC would maintain confinement of the spent fuel without any breach or rupture of the DSC. Derefore, there could be no adverse impact on the public. As noted, the original NRC evaluation was based on a DSC nominal i. hell thickness of 0.625 inch.

In NRC's evaluation of the VECTRA September 5,1995 submittal, which l 1

used a minimum DSC wall thickness of 0.500 inch to demonstrate a bounding case, the NRC Staff review assumed the occurrence of essentially the same range of accidents. Again, the NRC Staff found that the DSC would maintain  ;

confin'ement of the speat fuel without any breach or rupture of the DSC. I When VECTRA initially sought the NRC's 1994 approval of the NUHOMS, it provided design criteria for the DSC in the SAR as a basis for NRC approval l

of the NUliOMS system. VECTRA's proposed design criteria for the DSC were '

certain portions of the ASME BP&V Code.' Materials (such as the materials that make up the DSC) have known stress values at which they will bend or break.  !

During an accident, if the stresses acting on a vessel such as the DSC exceed those values, then it can be assumed that the material will fail. To facilitate the design process, the Code prescribes design criteria in the form of " allowable l stresses" and requires that vessels such as the DSC must be analyzed under l

accident conditions to ensure that the stresses resulting from the accident do not j

exceed the allowable stresses of the materials used in the vessel. Depending ,

on the likelihood of given design loading conditions, the Code builds into the  !

design criteria and allowable stress values for each material a safety margin by I setting generally the allowable stress at a fraction of the stress at which the material is known to bend or break.

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'NRC approved the NUHOMS based on VECTRA's appbcanon and a supporting SAR which. m turn pursuant I to apphcable NRC regulatwns. mcluded appropnate deugn entena for the storage cask See 10 C F R. I 72 236(b1 A vendor's design entena in the SAR are maportant because they are to be used to analyze the acceptabihty of {

i the vendor's proposed cask design against potential stresses on the task aftet it is loaded wuh spent nuclear fuel.

The stresses to be analyzed co er a vancty of conditions that the cask nuy encounier dunng use. includmg those attritratable to the dead weight or temperature of the spent fuel in the cask, mternal pressures placed on the cask after n is loaded and sealed, normal handhng of the cask dunng onsite transport or transfer, a potential handhng accident such at a Jammed camster wher it is bemg placed in or retneved from the storage modul at a dropped cask dunng transport. seismic loads than anse from ground accelerations dunng an earthquake pt 'ulaicd flood events, and stresses frorn certain kiad combmanons The deugn entena in the SAR subnutted by VECTRA to NRC covered each of the cask condicons appheable to the proposed NUHOMS design (mcluding the DsCA VECTRA also analyzed the NUHoMS deugn agamst these design cntena in the SAR. us ng a norrunal DSC wall thickness of 0 625 inch. pnor to NRC cask approval m i December 1994 hirther. and as detaded in the NRC Starf a safety Esaluanon Report (SER) which supported the

! rulemakmg and ulumate approval of the VLCTRA NUHoMS. the NRC Staff evaluated and accepted VECTRA's design (niena and analyses before issums VECTRA the CoC as pan of the NRC's December 1994 approval 77 I

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VECTRA used the same ASME Code provisions for evaluating the DSC l designs'" and demonstrated that the Code provisions were met by a DSC shell thickness of 0.500 inch." Thus, even with the reduction in shell thickness, VECTRA demonstrated that the ASME Code provisions will be met by the DSC shell thickness of 0.500 inch?

Therefore,I conclude that the reduction in shell thickness does not compro- j mise the safety of the three DSCs at Davis-Besse. VECTRA has demonstrated that a DSC with a minimum shell thickness of 0.500 inch will provide safe confinement of spent fuel in the event of an accident.

VECTRA'S revised structural analysis assumed that the entire DSC shell thickness, including all shell plating and weld lengths, had been reduced from 0.625 to 0.500 inch. This assumption by VECTRA resulted in a calculation that underestimated the strengths of c':e actual DSCs at Davis-Besse that were measured by VECTRA and found to have the specified 0.625-inch material thicknesses for nearly all of the shell weld lengths. Thus, the actual DSCs at Davis-Besse, with nonconforming weld thicknesses on only a portion of their weld lengths, should readily perform as well as VECTRA's revised structural

'"The design entens used m VECTRA's reesaluation remamed the sanw but the coun'mg of the dead k>ad effects difleted m one respect from the SAR. In a request for additional informanon (RAl) dated Augusi 17. lW5. tir staff comnwnted that, "the deducuon of deaJ weight (DW) from normal hundhng stress (Ln) load condmon is a change m tte deugn cntena used m the SAR " later, in the october 5. lW5 Staff safety evaluanon of VEGRA's revised calculanon package and respimse to de RAl the Staff noied that. "the calculanon rackage conuders the sanw design bases and cntena as those m the SAR." to the SAR. m amupmg cert.un load combmanons. VEGRA had counted sone dead weight stresses t*o times. whereas m the reesuluauon of the DsC. It &d not The Staff agreed that the double counnng m the SAR was unnecessary and, therefore, accepted the remow_I of the dout le countmg in the revised analyus.

" As &scussed previously. after the June IWS &scovery b) NRC inspectors of the DSC wall-thickness issue.

VEGRA was asked in the July 7. lW5.NRC CAL to proude an engnwenng analyus addressmg the potennat safet) impact of the lack of wall-thickness nwasurenrnes that coscred casks m fabncatmn. most parucularly the three DSCs destined for Dans-Besse. VEGRA chose to subnut a revision to the structural analyus prenously prouded to NRC m the SAR. udng a mammum DSC shcIl stuckness of 0 500 mch. while conudenng the er deugn cnterta as those m Etw SAR. which had been found acceptable by NRC for nwenng NRC requirementi I l m61udmg secuon 72 2Wbt The Staff notes that, dunng the design process for conipiments such as (tw DSCs.

vendors commmly use conservanvc asuimptions m their calculanons to simplifv se calculatum process. (See NRC's SER 5 3 2 3 ) Therefore, it can and should be espected that it snay te posuble to use ao alternauve nrthod to perform design calculauon te g.. a more retmed calculation that elomnews some of the conservauve

! anumpnons) to denenstraic that a efferent DSC shell deugn (e g.. a deugn that u.es a thimier wall duckness) will i l also sausfy the deugn entena ensbo&ed m ttw ASME Code. As &scuued above this is exactly whas VECTRA l

&d That is, to resolve the wallahickhess awasurement issue raised m the July 7.1995 CAL VECTRA perforned a structural reanalyus of the Nt.'HOMS. VECTRA reanalyzed the D5C with a unibrm wall tluckness of 0.500 meh, whnh is ihmnet than the nonunal wall thickness of 0625 inch used in the anasyns ongmally pionded by YLCRA in the SAR. Further, the structural adequacy of the DSC was denmstrated by companng the calculated stress mienstres for the 0 500.mch Dsc =hJ1 to the same design cntena used for the 0 625anch shell o e.. ASME Code 5111 stress allowables) .

f U When dw NRC Staff rewwwed VECTRA's revised structural analyus submitted in September IWS 6 e , tlw analysis denmnstraung the structural acceptability of the DSC uung 0 500-inch DSC shcIl thickt ss). the NRC )

Sraff also rehed on compliance with the same ASME Code provtuons to estabhsh the relevant design entena )

fur deternumng whether the 0.500-mch DSC shell design would provtJe the required safety Specifically. m its safety venhcation of ttw VLCrRA calculauon package (NUH004 0213. "Standai& zed NUHOMS.24P DSC Shell Mimmum Acceptable Umform Thickness"). the NRC Staf f concurred thai all calculated stresses for etw 0 500 mch DSC shell tluckness are acceptable. (See NRC tetter to VEGRA. dated &tober 12 iW5 I i

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analysis predicted. In either case, the affected casks will perform in accordanc-with the pertinent ASME Code requirements, the operative design standard i inherent in the NRC Staff's approval. This level of performance provides 1

reasonable assurance that public health and safety will be protected. '

Thus, while VECTRA failed to comply with its SAR commitment of 0.625 '

inch, its failure resulted in no compromise of safety. Nonetheless, the failure raised an issue of poor control during the fabrication process. This deficiency was identified by NRC during the June 1995 inspection; and VECTRA was cited f

for it in the NRC Notice of Nonconformance issued to VECTRA in August l 1995." '

B. Rulemaking Should Be Conducted to Propose Changes to the >

NUHOMS Certificate in Light of the Weld Thinning Issue, and Petitioners' Claims Can Be Made in That Rulemaking l

Petitioners question the legal validity of the administrative and regulatory r processes used by NRC after discovery of the DSC wall-thickness issue. Petition  ;

at 1. Specifically, Petitioners believe an NRC rulemaking (or other public '

proceeding) should have been held. i As set forth above, my conclusion is that the DSCs at Davis-Besse are safe.

However, as I will explain below, I believe an issue remains as to whether NRC l should take some additional action with respect to VECTRA's COC for the i NUHOMS cask.  !

I have already referenced the NRC's action with respect to VECTRA's failure  !

to conform to NRC requirements. In particular, the fabrication process for the DSCs did not ensure that acceptable DSC wall thickness was maintained as i required by NRC. The process included an instruction that the operator manually flush-grind the welds, after welding the DSC shell seams. However, there

[

was no procedure that provided an adequate level of control in maintaining minimum acceptable wall thickness. Moreover, under the procedures, the '

operator did not measure the final wall thickness of the DSC in the area of the welds after grinding. Further, measurements were not taken in any subsequent  !

steps in the fabrication process to ensure that minimum wall thickness was #

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U pettu wrs' Novernber 14 letter asserts that VECTRA violated NRC regulanons when a fmled to do measure.  !

nwnes en DsC unll ttuckness and weld seams dunng fabreauon. NRCs June 1995 inspecnon of VLCrRA/Ranor  ?

und NRC) August 1995 Nouce of Nonconf armance to VLCrRA have already mdicated that VECTRA Imled to i conform to NRC regulauons, The Pecuoners* November 14 letter also quesnons whether VECrRA may have willfully failed to report a nonconfurnance or devasuon in mall thicknesses for the DSCs. The NRC inspecuon l Jid not idenufy any indicahons of a willful failure to report Rather, the falure on the part of VECrRA/Ranor '

was na failure to have adequate quahry curgrol measures in pl. ace dunng the fabncation process to measure DSC f welds after gnndmg. It appears that VECTRA /Ranar did not anticipate th.H gnadang the weld could result in l goms below the specified plate duckness. Therefore, the Pennoners' concern about a possible wellful falure to  !

report a nonconformance cannot be substanualed. I 79 I

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I maintained. VECTRA thus failed to ensure conformance to NRC's requirement that activities affecting quality must be prescribed by appropriate, documented instructions, procedures, or drawings that include criteria for determining that l important activities have been satisfactorily accomplished. See 10 C.F.R. '

6 72.150, " Instructions, procedures, and drawings." As a consequence, NRC l

issued VECTRA a Notice of Nonconformance on August 29, 1995, citing VECTRA for its failure."

i Petitioners, however, seek additional action. Specifically, in their December 5,1995 petition, Petitioners state that they believe that an NRC rulemaking (or other public proceeding) was required to permit use of the three DSCs with wall thinning at Davis-Besse. Further, in their related November 14,1995 letter, Petitioners question whether an NRC rulemaking was required because VECTRA's change of the three DSCs to a wall thickness of less than 0.625  ;

inch involved a reduction in "the margin of safety" that must be approved by an NRC amendment process.

Petitioners' November 14 questions appear to be aimed at VECTRA's im-picmentation of Condition 9 in the NRC COC issued to VECTRA in December 4

1994.0 Condition 9 permits VECTRA to make changes in the DSC design without NRC approval provided, among other things, the change does not in-l volve "an unreviewed safety question." Condition 9 states that a change shall j be deemed to insolse an unreviewed safety question "[ijf the margin of safety I

as defined in the basis for any technical specification or limit is reduced." After i

evaluation, VECTRA concluded that the wall thinning of the three DSCs at Davis-Besse did not involve a reduction in the " margin of safety" or "an unre-l 4

"NRC's Nonce of Nonconformance sited vlCTRA for several other nonconformances with NRC requirements i uarelated to DSC wall thsckness or the pennon.

"In their November '4.1995 letter Pennoners quesuoned vtCTRA's IcFal authonty to make changes to the DSC. In 1990, to fulfill the nundate of the Nuclear waste Pohey Act of 1982. the NRC arrrnded 10 C F R. Part 72 so as to put in place the regulatory procedures that authonze a nuclear power reactor heensee to store spent fuel en ute under a general heeme wnhout the need for an adenonal site-specihe Commission approval 55 fed Reg 29.181 (1990). To use the general hcense, the reactor heensee must store the spent fuel m a ca:.k that has been ceruhed urrder the proviuons of 10 C F R. Part 72. Subpart K. See 10 C.F R. I 72.2124ax2). A vendon who meets the Subpart K requirements mil be tssued a CoC by the NRC and, aher a pubhc rulenmLmg proceeding.

the cask will be added to the his of approved spent fuel storage casks at 10 C F R.172 214 These regulatory pmcedures were t. sed with respect to NRC's approval of VLCTRA's NUHOMS system. On June 2,1994, the NRC pubhshed a proposed rule addmg the NUHoMS svstem to the approved hst and gave nouce that alw draft CoC was available for mspecuon and comnrnt at the NRC Puhhc Document Room. 59 ftd.

Reg 28.496 0994). As Peunoners are aware. Condmon 9 c f the COC provides to the hokter of the CoC the same type of authority 10 make changes as is provided to heensees under section 72 48 Condiuon 9 provides to VICrRA, among other ttungs. the authonty to make changes in the cask design desenbed an the SAR without pnor Comnussion approval unless the proposed chanFe mvolves a change in the CoC. an unreviewed safety quesuon. a significani incres.se in occupanonal exposure, or a sigruheant unreviewed environnental irrynet The C(munnsion received a number of posiuve and negaine comnrnts.includmg conurents frorn some of the preszut Peuuoners, on sin proposal to incorporate sectmn 72 48 type language into Condmon 9 of the CoC but deternuned to retain th3s language in the final rule See 59 Isd Reg 65.893. 65314-15 0994L Condinon 9 was adopted through nonce-and comment rulemaking. and vlCTRA was enntled to uuhze its provismns in comidonng changes to the cask design as desenbed in the SAR.

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4 viewed safety question." By asserting that an NRC rulemaking was required, Petitioners may be effectively arguing that I should find these VECTRA con.

clusions to be wrong.* However, it is not necessary to evaluate VECTRA's conclusions in order to decide that Petitioners' request for NRC rulemaking should be granted with respect to the wall-thinning issue. As I explain below.

I beliese that rulemaking should be undertaken for different reasons.

In this regard, I note that the NRC Staff's October 5,1995 SER (issued when Staff accepted VECTRA's analysis of a minimun DSC shell wall-thickness of 0.500 inch) includes the conclusion that "it is prsdent to require" a minimum weld inspection threshold thickness. As part of its response to the CAL and to address the nonconformance with NRC requirements described above, VECTRA had proposed an inspection procrdure to ensure that DSC weld-grinding operations do not r: sult in wall thiraing below acceptable levels. Staff viewed (and continues to view) VECTRA's proposed inspection procedure, which invokes enhanced actions if grinding operations exceed a 0.563. inch threshold, as an acceptable quality control practice. Further, it was Staff's intent in the SER to reflect VECTRA's inspection plan as an important consideration in Staff's acceptance of VECTRA's response to the CAL.

However, although VECTRA implemented the inspection procedure as to the three Davis-Besse DSCS and committed to use it in fabricating future DSCs and although the NRC Staff's SER expressly relied on the VECTRA inspection procedure as a consideration in accepting VECTRA's response, nothing in the VECTRA COC explicitly requires VECTRA to conduct inspections during fabrication of the DSC. Thus, one purpose of rulemaking would be to consider whether these (and possibly other) circumstances of the NUHOMS wall-thinning issue justify the step of putting a fabrication inspection requirement in the VECTRA COC. Specifically, rulemaking could propose to amend the VECTRA COC to require that, in the fabrication of the DSC, the shell and basket assembly must be inspected to ensure that structural desiFn margins, associated with the ASME Code iIll stress allowables, are not compromised. Such a requirement would serve the purpose of helping to ensure that the DSC fabrication process, including weld-grinding operations, produces DSC components that conform to the design criteria and safety margins approved by NRC.

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" An NRC mspecnon team found VECTRA's safety analpis for the wall tlunning lasue to be m adnumstrauve comphance wnh Condiuon 9. The technical aspects of VECTRA's safety analysis were not reviewed by the

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team. Ser NRC L nection Report No 72-1004N6 201 I should note that NRC pohey m this area nught undago '

clanficanon The regulatory language in Condiuon 9 is similar to language in 10 C F R. 6 50 59, a separate and I unrelated prension mvolvmg nuclear reactors. NRC is conduenng an miernal renew of its pohcy gmdance on i identifymg " unrenewed safety quesnons" in the context of secuon 50 59 i miend to monnor that review and, l when u is complec, consider whether there is a need to dmlop clanfying gmdance for Condshon 9. as well as secuon 72 4X which governs changes by Part 72 heensees.

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At this point, I am inclined to believe that VECTRA's COC should be modified in light of the weld-thinning issue." As discussed above, I believe that changes to VECTRA's COC merit consideration as possible additional actions to ensure the quality of VECTRA's NUHOMS components in light of the history of this matter. Further, rulemaking would allow us to receive and consider ,

comments of the Petitioners and other members of the public who are interested in the weld-thinning issue. As part of the rulemaking, NRC could include in the record the entire NRC Staff safety evaluation of VECTRA's wall-thickness reesaluation and the VECTRA reevaluation itself submitted in response to the NRC July 7,1995 CAL. As noted, the Staff's safety evaluation and the acceptability of VECTRA's reevaluation both depended, in part, on a VECTRA inspection that the rulemaking would propose to require in VECTRA's COC.

In the rulemaking, as I envision it, Petitioners, as well as any other inter-ested member of the public, would be given the opportunity to comment on any aspect of the NRC safety evaluation associated with this issue. At the conclu-sion of the comment period. NRC would consider all comments and provide a response. Further, if NRC determined, after considering the comments, that it should modify VECTRA's COC or change the Staff's previous determination to accept VECTRA's 0.500. inch uniform wall-thickness calculation, the tule-making would provide a vehicle for it to do so.

'Ihis course of action, which I intend to pursue, would provide Petitioners the agency rulemaking they seek on the reduction in the thickness of the DSC metal walls to less than 0.625 inch, and it will provide them the opportunity to examine and comment on NRC's determination that the safety of the DSCs has not been compromised and to submit such other information as they wish on any I aspect of the wall-thickness issue. Therefore, to this extent, I am determining that the petition should be granted.

I have also considered whether NRC should take some additional action, pending completion of the rulemaking, with respect to the three DSCs now in service at the Davis-Besse site. In Part A of this discussion. I set forth the basis for my conclusion that -the reduction in the shell thickness of the l DSCs at Davis-Besse does not compromise their safety. Therefore, I believe that continued storage of spent fuel in the DSCs, pending completion of the rulemaking, would not pose an unreasonable risk to public health and safety  !

and that there is no technical basis to require their unloading. Further as I have previously summarized in this Part B. NRC already cited VECTRA for its failure to comply with NRC requirements in August 1995. Accordingly, to the extent Petitioners seek additional action, pending completion of the rulemaking, their request is denied.

U tJnder NRC miernal procedwes. the Statt rnost request and obtain Conunimon approval before undenalung rulenudung. Therefore. I intend to seek Comnumon approval to do so 82 l

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C. There Is No liasis to Grant Petitioners' Request That NRC Review, Approse, ar.d Field-Test Procedures for Unloading DSCs Prior to Operation Petitioners also present claims concerning the unloading of the casks at Davis-Besse. Specifically in this regard, they demand that "no loading of the canisters be authorized until there is in place a written, approved, and field-tested procedure for unloading the DSCs both in urgent and nonurgent circumstances."

Petition at 12.

Here is no regulatory requirement for NRC to review, approve, and field test a licensee's operating procedures, including unloading of spent fuel casks under urgent and nonurgent circumstances. Rather, NRC's approach is to require that licensees have a formal process for procedure development and control.

Generally, in the analogous case of a power reactor, this process is part of the facility license. NRC oversees the implementation of that process and the product (i.e., the procedures and their use) through its inspection program. This approach to overseeing licensee operations has been effectively demonstrated by successful startup of power reactors following construction and the continued safe operation of existing facilities. NRC expects that a general licensee under Part 72 will prepare ISFSI procedures in accordance with its established procedure development process and as required by its quality assurance program.

As a general licensee, Toledo Edison is required to comply with the terms and conditions of the COC issued for the VECTRA NUHOMS-24P. See 10 C.F.R. 6 72.212(b). The applicable conditions in the COC can be found in section 1.1.2 which requires that "[w]ritten operating procedures shall be prepared for cask handling, loading, movement, surveillance, and maintenance." This condition is broadly written and interpreted by NRC to require the licensee to have detailed written procedures for loading and unloading a DSC. Another related condition in the COC appears in Section 1.1.6 which requires "[p]re-operational testing" that includes, but is not limited to, a dry run of loading and unloading a DSC.

Thus, it is the licensee's responsibility to prepare, review, approve, and test written procedures for eask loading and unloading. Further, NRC requires a licensee to conduct activities related to ISFSI operation, including cask loading and unloading, in accordance with those written procedures once the licensee has approved, tested, and put procedures in place. See 10 C.F.R. 5 72.212(b)(9).

The NRC also conducts periodic audits of these activities through its inspections program.

It is not NRC's practice to review and approve a licensee's operating procc-dures. It is important to understand that, just with respect to dry cask storage activities, which are a very small fraction of the daily activities conducted at an operating nuclear power plant, the applicable written procedures of a gen-eral licensee are likely to be voluminous. Moreover, the written procedures 83

prepared by a licensee typically are site-specific in nature and thus reflect the

, licensee's special knowledge of its plant and how dry cask storage activities interconnect with plant personnel, as well as other plant activities and procc- i dures. The written procedures are prepared according to the formal procedure development process and exercised during the dry run. In my judgment, there woul<J be very little additional value to be gained from a requirement of NRC review and approval of a licensee's written operating procedures, particularly

{

give'n our existing inspection activities illustrated by the Davis-Desse example, '

below.

In particular, with regard to Davis-Besse's ISFSI operation, the Licensee de-

veloped written operating procedures for' dry cask handling including loading

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and unloading procedures. These procedures were used by the Licensee for l

] the preoperational dry-run testing at the Davis-Besse plant during November 30 through December i1,1995. The NRC Staff inspectors were present at the  ;

plant throughout the testing, conducted an onsite observation of the Licensee's i dry-run loading and unloading activities, and also inspected the detailed written '

)

, procedures used by the Licensee for cask loading and unloading. NRC Inspec-l tion Report 50-346/95-09 documents the extensive NRC inspection activities, as

, well as the inspection finding that the dry-run activities were conducted satis- l factorily and in a safe manner. Therefore, based on the circumstances reflected i in the foregoing discussion, I conclude that there is in place at Davis-Besse  !

an adequate written procedure, approved and field tested by the Licensee, for i unloading the DSCs if needed, and that the Petitioners' request - to the extent it seeks further NRC review and approval- should be denied.

, CONCLUSION

, As discussed above, VECTRA's change to the wall thickness of certain

, weld seams does not compromise the safety of the three DSCs at Davis-Besse.

i However, the NRC COC for VECTRA's Standardized NUHOMS-24P should be modified to require a fabrication inspection of the DSC. An agency rulemaking is, therefore, needed and should be conducted to accomplish this modification.

In rulemaking, Petitioners would have the opportunity to comment on any aspect i of the DSC wall-thickness issue. However, because the continued storage of spent fuel at the DSCs at Davis-Besse does not pose an unreasonable risk to j public health and safety, I find no technical basis to require the DSCs to be unloaded pending completion of this rulemaking. Further, VECTRA has already been cited for a nonconformance with NRC regulations, and I find no basis in

~

the petition to take other action in this regard.

Toledo Edison has developed loading and unloading procedures for handling spent fuels. These procedures have been applied for the dry-run testing with i

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NRC's oversight. Therefore, I find no basis in the petition for requiring halting of the ISFSI operation at Davis-Besse.

Accordingly, the petition from Toledo Coalition for Safe Energy is granted to the extent that it requests an agency rulemaking and is denied in all other respects.

FOR THE NUCLEAR REGULATORY COMMISSION Carl J. Paperiello, Director Office of Nuclear Material Safety and Safeguards I r Dated at Rockville, Maryland, this 5th day of February 1997.

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Cite as 45 NRC 86 (1997) DD-97-4 l

UNITED STATES OF AMERICA i NUCLEAR REGULATORY COMMISSION  !

t OFFICE OF NUCLEAR REACTOR REGULATION )

Frank J. Miraglia, Jr., Acting Director j In the Matter of Docket No. 50-245 I (License No. DPR-21) t i

NORTHEAST NUCLEAR ENERGY COMPANY  !

l (Millstone Nuclear Power Station,

• Unit 1) February 11,1997 l

t The Acting Director, Office of Nuclear Reactor Regulation, has granted in i

part and denied in part a petition filed by Anthony J. Ross requesting action .l regarding Millstone Nuclear Power Station, Unit 1. The Petitioner requested  !

that the Commission take escalated enforcement action against the Licensee and  ;

certain individuals based upon the deliberate failure to comply with procedures  ;

involving sign-out of measuring and test equipment, and conduct an investigation  !

into alleged procedural violations and audit the Millstone Unit I maintenance  !

department Measuring and Test Equipment folders for widespread problems  ;

regarding procedural noncompliance. To the extent that the Petitioner requested '

escalated enforcement action he *.aken, the petition has been denied; to the extent <

that the Petitioner requested an investigation into the procedural violations and .

an audit, the petition has been granted. (

ENFORCEMENT POLICY: SEVERITY OF VIOLATIONS  !

l Minor violations, as described in the current enforcement policy, are not the subject of formal enforcen ent action and are usually not cited in inspection reports. To the extent that such violations are described, they are now noted as noncited violations.

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i RUI,ES OF PRACTICE: INSTITUTION OF PROCEEDINGS UNDER 10 C.F.R. 6 2.206 The institution of a proceeding purwnt to 10 C.F.R. 6 2.2% is appropriate only if substantial health and safety issues have been raised.

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DIRECTOR'S DECISION UNDER 10 C.F.R. # 2.206 >

1 i I. INTRODUCTION l l,

] . On January 5,1995, Mr. Anthuny J. Ross (Petitioner) filed a petition with j

the Executive Director for Operations of the Nuclear Regulatory Commission (NRC) pursuant to section 2.206 of Title 10 of the Code of federal Regulations i j~ (10 C.F.R. 6 2.206). In the petition, the Petitioner raised concerns regarding '

noncompliance with Procedure WC-8, " Control and Calibration of Measuring . '

and Test Equipment," at Millstone Nuclear Power Station, Unit 1, and requested  ;

j that escalated enforcement action be taken. Specifically, the Petitioner provided 4 l several examples of what he alleged were violations of Procedure WC-8, which  !

i he stated required that measuring and test equipment (M&TE) be signed out  !

from, and returned to, a custodian upon completion of work. The Petitioner y requested that the NRC institute sanctions against his department manager, his first line super isor, and "two coworkers"' for engaging in deliberate misconduct

! in violation of 10 C.F.R. 5 50.5 in failing to comply with Procedure WC-l 8. The Petitioner also asserted that the NRC should conduct an investigation j into violations of this procedure and audit the Millstone Unit I maintenance department M&TE folders for widespread problems regarding noncompliance i with this procedure.

On February 23,1995, the NRC informed the Petitioner that the petition had I i been referred to the Office of Nuclear Reactor Regulation pursuant to section i

! 2.206 of the Commission's regulations. 'Ihe NRC also informed the Petitioner that the Staff would take appropriate action within a reasonable time regarding i the specific concerns raised in the petition. On the basis of a review of the j issues raised by the Petitioner, as discussed below, I have concluded, for the i

i reasons explained below, that the petition is denied with regard to the request f for escalated enforcement action and instituting sanctions against the department ,

a manager, first-line supervisor, and two co-workers, but granted with regard to e

s

)' I lhe "two coworkers *' are understood to be an indmdual the Pennoner alleges m'illfully faluned (tackdated) an entry nn the forrn to indicate that the meter was returned on October 13.1994, and an individual the XYutioner

) alleges winfully violated Procedure wC-8 on Novemter 17.1994. by signing cut his own M&TL A

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the requests for an " investigation into the above mentioned procedure violations" and for the NRC to " audit the Unit i maintenance department M&TE folders."

II, DISCUSSION In the petition, the Petitioner raises concerns regardmg numerous noncompli-ances with Procedure WC-8, Revision O. at Millstone Unit 1. Specifically, the Petitioner states that (1) quality assurance (QA)2 test meter 1587 was signed out on October 13,1994, for weekly battery readings, and as of October 19,1994, the user had not returned the meter or signed it in. The Petitioner states that this practice was in violation of Procedure WC-8, which stated " return M&TE to custodian upon completion of work";) (2) although he identi'fied a problern with Procedure WC-8 (specifically, who was responsible for the actual signing in and out of M&TE) to his first-line supervisor on November 7,1994, as of December 1994, the procedure still had not been changed (in accordance with Procedure DC-4, " Procedural Compliance," which requires that if a procedure conflict or interpretation problem exists, a change or revision should be made);

(3) on November 10, 1994, he noticed on a station form that someone signed in the QA meter with the return date of October 13, 1994, and that this was a willful falsification (backdating) of a nuclear record; (4) on November 17,1994, an electrician co-worker was directed by their first-line supervisor to willfully violate Procedure WC-8 by signing out his own M&TE, and signed out his own M&TE although both the supervisor and co-worker knew they were to have the custodian sign out the equipment; (5) on November 21,1994, his department manager instructed the custodian to give a spare key for the QA locker to the Millstone Unit I control room so the control room could sign out equipment at night; and (6) on November 25,1994, a mechanic signed out M&TE without a custodian.

In addition, the Petitioner states that he believes that his department manager was directly responsible for sharing the effects of a new, revised, or rewritten procedure with the employees of his department if the procedure directly af-fccted day-to-day operations. The Petitioner asserts that this individual's " lack of 2

Quahty Assurance comprnes those quahry assuranco acuens related to the physical charactensucs of a nuitenal, structure, component, or system that provide a rneans to control the quahry of the matenal. structure. component, or system to predetennined requirenants 3

Tlus procedure had becone effective on June 20.1994 It reqmred that a " designated custodian" enter tir date of assue and date or return on the custody and usage record, and that the user of the equipurnt return it to the cusiodian upon compleuon of work. In Attachment i to the procedure. " custodian" was dehned as the endnidual dengnaied t y tir depanrrent head to store, track, and issue the depanmein's M&TE.

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communications" regarding the procedure has caused a " widespread problem of procedure noncompliance."d in letters to Northeast Nuclear Energy Company (NNECO), Licensee for Millstone Units 1,2, and 3, dated December 5 and 28,1994, and February 14, 1995, the NRC Staff raised a number of maintenance-related issues. In those letters, the NRC Staff requested NNECO to review these issues and submit a written response. Among these issues, the NRC requested NNECO to review two issues associated with Procedure WC-8 that are now presently being raised by the Petitioner. These were that: (1) the Millstone Unit ! QA test meter 1587 was signed out on October 13,1994, to perform weekly battery readings, but as of October 19,1994, the user had not returned the meter or signed in the meter; and (2) many members of the Millstone Unit i Maintenance Department never received training on Procedure WC-8, Rev. O, within 60 days of the effective date of June 20,1994, as required by the documentation of training requirements form of NNECO Procedure DC-1.

In a letter dated March 6,1995, NNECO responded to the issue regarding failure to return the QA meter signed out on October 13, 1994. In its !ctter, NNECO stated that on October 13,1994, a maintenance electrician signed out

. QA test meter 1587 to perform weekly battery surveillances and signed it back in on the M&TE log on the same day. On October 19, 1994, a different maintenance electrician signed out and returned QA test meter 1587. Sometime later that day, QA test meter 1587 was signed out again and subsequently

returned the same day. NNECO stated that it was unable to detennine, based  ;

on interviews with the parties imolved and a review of the custody and usage i i record, the exact circumstances surrounding QA test meter 1587. However, w hat l

was known was that QA test meter 1587 had been signed out once on October 13 l and twice on October 19,1994. NNECO's review further concluded that strict  !

compliance with Procedure WC-8 was not being observed at all three Millstone units in that a custodian was not being used to ensure that certain actions (i.e., signing in and out M&TE on the M&TE log) were being accomplished.

However, NNECO stated that it believed it met the " intent of the procedure" in that the user of the M&TE stored, tracked, and issued the equipment as required by the procedure, except that the custodian was not involved. As a result of its review, NNECO undertook certain corrective actions. Specifically, NNECO held a site-wide meeting for all departments responsible for use or issuance of QA M&TE on February 21,1995, to determine corrective actions necessary to ,

ensure procedural compliance. Subsequently, NNECO revised Procedure WC-8

• NNI.CO Procedure DC.I requires that the L.icenwe select the tranmg requirenrnis to be used in trmning errgbyees wheneser procedures are revhed. and indicate the type of tranung that would be perfomied on Anachrnent 5 to Procedure DC.! Ftw Procedure WC-8. Revismn O. the trammg reqmred was rnarled as "trairung to be done by Drpannrnt or Nucicar Trammg Depannent wntun 60 days of the effective dale and prnw to perfortnance of procedure."

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l on April 27, 1995, to specifically allow the user of M&TE to sign QA test  ;

. equipment in and out. The custodian is still responsible for storing and tracking l M&TE. In addition, Millstone Unit I control room personnel responsible for l accessing QA M&TE were made aware of the logging requirements.

i 'lle NRC conducted a special safety inspection from May 15 through June l

l 23,1995, at the Millstone station. During this inspection, the Staff reviewed '

l a number of the concerns, including the concerns about QA test meter 1587 '

l and the other examples of noncompliance with Procedure WC-8 alleged by the Petitioner, and issued its findings in inspecimn Report (IR) 50-245/95-22, 50- J 336/95-22, 50-423/95-22 (95-22), dated July 21,1995. l During the inspection, the NRC Staff reviewed the custody and usage record sheets for QA test meter 1587 from September 27 to November i1,1994. Based on this review, the Staff was unable to determine whether QA test meter 1587 was properly logged in and out in October 1994 or if the custody and usage l record sheet was backdated. The NRC Staff discussed this issue with the workers

involved who indicated that they had no recollection of the exact circumstances 3

surrounding QA test meter 1587 and that, to the best of their knowledge, QA test l meter 1587 was logged in and out properly, Therefore, the Staff was unable to determine whether QA test meter 1587 was controlled improperly and whether the Petitioner's co-worker willfally falsified (by backdating) a nuclear record (M&TE log).

The Staff also reviewed the original procedure and determined that although l Procedure WC-8, Rev. O. was not clear in specifying who was responsible for l the actual signing in and out of equipment, NNECO was meeting the intent

! of the procedure in that M&TE was stored, tracked, and issued in a controlled l- manner. The NRC Staff further concluded that NNECO's additional corrective actions (i.e., modifying the procedure) were adequate in clarifying the procedure and should prevent interpretation problems in the future.

l Notwithstanding the findmgs of the inspection report, however, the NRC has l reconsidered this matter and determined that NNECO was not in compliance  ;

l with Procedure WC-8, Rev. O. This determination is supported by the fact that i NNECO admitted in its March 6,1995 letter that it was not in compliance with Procedure WC,8. In addition, the NRC has reviewed the custody and usage records for signing in and out M&TE on November 17 and 25,1994, and i determined that an electrician and mechanic had signed out their own M&TE, l respectively, on those dates. Accordingly, the Petitioner's assertions that the i procedure was violated when a co-worker electrician signed out his own M&TE on November 17, 1994, and a mechanic signed out M&TE on November 25, 1994, is substantiated. However, the NRC has been unable to confirm that either of these individuals had been " directed" by supervision to sign out the equipment.

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in addition, NNECO's review, as described in its letter dated March 6, 1995, and verified by the Staff in IR 95-22, determined that keys had been availaHe during this time frame in all Millstone control rooms and were in the possession of. security personnel to allow access to QA M&TE storage locations.

These Froups required access to these areas in order to properly execute their duties. Therefore, since the cunodian did not sign in and out the equipment, the Petitioner's additional assertion that the procedure was violated because security personnel and personnel in the Millstone Unit I control room could sign out M&TE at night is substantiated. However, the NRC has been unable to confirm that the department manager had instructed the custodian to give a spare key to the control room so the control room could sign out M&TE at night.

Furthermore, the Staff has determined that, since there were no safety consequences as a result of these events, the noncofnpliances with Procedure WC-8 did not constitute a violation that could reasonably be expected to have been prevented by the Licensee's corrective action for a previous violation or a previous Licensee finding that occuned within the past 2 years of the inspection at issue, adequate corrective actions were implemented regarding Procedure WC-8, and the violation was not willful, the violation would have been categorized in accordance with the enforcement policy in effect at the time of the inspection as a noncited Severity Level V violation and would not have been the subject of formal enforcement action.5 In addition, since the procedure was not clear in describing specific respon-sibilities and NNECO believed it was meeting the intent of the procedure, the NRC has concluded that the Petitioner's department manager, his first-line su-pervisor, and two co-workers did not deliberately violate NRC regulations or the Millstone Unit I operating license and, therefore, did not violate the provi-sions of 10 C.F.R. i 50.5. Moreover, NNECO revised Procedure WC-8 on April 27, 1995, and the procedure now more clearly allows the user of the M&TE to sign in and out QA test equipment. The custodian still is responsible for storing and tracking M&TE. Therefore, the Staff has determined that, although the Petitioner is correct in that the procedure was not revised as of December 1994, the procedure was subsequently revised, so that Procedure DC-4 was not violated.

5 The staff has resonsidered this violauon in accordance with the cuntnt enforcement pohey (NUREG-1600.

" General statement of Pohey and Procedures for NRC 1.nforcenent Acuan") and has concluded that the violanon is behm the level of sagruncance of Severity tesel IV siolations This deternunanon is based on the fact that NNICo was nretmg mteni of the procedure; there was neghgible impact on safery; NNECO's interpretation of the M&TL custodian's responubihties does not indicale a programmauc problem that could hase safety or regulatory unpact, if the violanon recurred, a would not be ronudered a sigmhcant concern, and the wolation was not winful. Therefore, if considered under the new enforcement pohey, tius violanon would be classified as a nunor vmlanon. Minor violanons, as desenbed in the current enforcement pobey, are not the subject of formal enforcemens acuon and are usually not ened m inspecuon reports To the entent that such vmlaimns are described, they are now reied as ncmened violanons.

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i By letter dated April 26, 1995, NNECO provided its review of whether

, members of the Maintenance Department received training within 60 days of Revision 0 of Procedure WC-8 (June 20,1994). In its letter, NNECO stated that no documentation indicating that training was conducted for Procedure WC-8, Rev. 0, had been found. While no training records were located, NNECO stated that the Millstone Unit 1 Maintenance Manager recalled that the procedure was

! discussed at a Maintenance Department meeting within 60 days of its effective date.

'Itc NRC Stalf reviewed Procedure DC-1 and determined that since NNECO could not locate the training records for Procedure WC-8, Rev. O, and that training by the Maintenance Department or the Nuclear Training Department was not conducted within 60 days of the effective date for Procedure WC-8, Rev,0, NNECO was in violation of Procedure DC-l. I 1

The Staff's review of NNECO's April 26, 1995 response to the NRC Iciter dated Rbruary 14, 1995, was documented in IR 95-22. The Staff has reviewed NNECO's corrective actions that included NNECO management i

reemphasizing the importance of training on new or revised procedures and  !

following procedures, the revising of Procedure WC-8, and training on the  !

revised procedure. Based on that review, the Staff has determined that the i corrective actions the Licensee has taken are acceptabic. The Staff has further determined that since there were no safety consequences as a result of this event, it was not a violation that could reasonably be expected to have been prevented by the Licensee's corrective action for a previous violation or a previous Licensee finding that occurred within the past 2 years of the inspection at issue, adequate corrective actions were implemented, and the violation was not willful, the violation would have been categorized in accordance with the enforcement policy in effect at the time of the inspection as a noncited Severity Level V violation and would not have been the subject of formal enforcement action.*

111. CONCLUSION The institution of a proceeding pursuant to section 2.206 is appropriate only if substantial health and safety issues have been raised. See Consolidafed Edison 6

The Staff has reconsidered this volation in accordance with the guidance in the current enforcement policy and has concluded that the vmlanon is below the level of sigmficance of sevenry level IV violanons This determmation as based on the fact that there was ntghgable impact on safety; the violanon does not indicate a programmauc problem that could have safety or regulatory impact. if the violauon recurred, it would not be considered e sigmficant concern. and the violanon was not millful Therefore this violation is classified as a nunor volanon and, as previoudy discussed. nunar vmlations are not normally the subject of formal enforcement action and are usually not cited in inspecuon repons To the extent that such vmlatmns are desenbed, they are charactenzed as noncited violaimns 92

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t Co. of New York (Indian Point, Units 1, 2, and 3), CL1-75-8,2 NRC 173,175 (1975), and Washington Public Power Supply System (WPPSS Nuclear Project No. 2), DD-84-7,19 NRC 899. 924 (1984). This is the standard that has been l l applied to the concerns raised by the Petitioner to determine whether the acuon (

requested by the Petitioner, or other enforcement action, is warranted.

On the basis of the above assessment, I have concluded that, although certain minor procedural violations occurred, no substantial health and safety issues have been raised by the petition regarding Millstone Unit I that would require  !

initiation of enforcement action. Therefore, to the extent that the Petitioner i l

requests that escalated enforcement action be taken against individuals and  !

NU for violations of Procedure WC-8 or failure to train employees on the j procedure, the petition has been denied. Ilowever, as described above, the NRC i

conducted an inspection into the alleged violations of Procedure WC-8 from May 15 through June 23, 1995, and conducted an audit of the custody and l usage record sheets. Therefore, to the extent that the Petitioner has requested an j

NRC " investigation into the above mentioned procedure violations" and for the  ;

NRC to " audit the Unit 1 maintenance department, M&TE folders," the petition has been granted. l i

As provided in 10 C.F.R. 6 2.206(c), a copy of this Decision will be filed with l

the Secretary of the Commission for the Commission's review. This Decision l will constitute the final action of the Commission 25 days after issuance unless the Commission, on its own motion, institutes a review of the Decision in that ,

time.  !

I FOR TiiE NUCLEAR l REGULATORY COMMISSION I

Frank J. Miraglia, Jr., Acting  ;

Director '

Office of Nuclear Reactor Regulation Dated at Rockville, Maryland, i this 1Ith day of February 1997. l l l 93 l I

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