Text

Nuclear Regulatory Commission Issuances for June 1997. Pages 437-495
	ML20210L101
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Issue date:	08/31/1997
From:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
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References
	NUREG-0750, NUREG-0750-V45-N06, NUREG-750, NUREG-750-V45-N6, NUDOCS 9708200220
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! NUREG-0750 Vol. 45, No. 6 Pages 437-495 l

l N UCLEAR REGULATORY-

! COMMISSION ISSUANCES l

l June 1997 ;

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Avahable from Superintendent of Documents U.S. Government Printing Office RO. Box 37082 Washington, DC 20402-9328 A year's subscription consists of 12 softbound issues, 4 indexes, and 2-4 hardbouri editions for this publication.

Single copies of this publication are available from National Technical Information Service Springfield, VA 22161 Errors in this publication may be reported to the Office of Information Resources Management U.S. Nuclear Regulatory Commiasion Washington, DC 20555-0001 (301 - 415- 6844) i m A

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i NUREG-0750 Vol. 45, No. 6 Pages 437-495 4

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i NUCLEAR REGULATORY l

COMMISSION ISSUAN CES l

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i June 1997 I

i i This report includes the issuances received during the specified period from the Commission (CU), the Atomic Safety and Ucensing

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

(DPRM)

! The summaries and headnotes preceding the opinions reported j herein are not to be deemed a part of those opinions or have any 4

independent legal significance.

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! U.S. NUCLEAR REGULATORY COMMISSION s l Prepared by the 4 Office of Information Resources Management i

U.S. Nuclear Regulatory Commission

' Washington, DC 20555-0001 (301 - 415 - 6844) 1 j

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't COMMISSIONERS t Shirley A. Jackson, Chairman s

, Kenneth C. Rogers s Greta J. Dieus . ,

Nils J. Diaz ;

Edward McGaffigan, Jr.

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. B. Paul Cotter, Jr., Chief Administrative Judge', Atomic Safety & Licensing Board Panel t

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CONTENTS Issuance of the Nuclear Regulatory Commissian LOUIStANA ENERGY SERVICES, L.P.

- (Claiborne Enrichment Center)

Docket 70-3070-ML ORDER, CLI.97 7, June 30,1997 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 437 Issuance of the Atomic Safety and !Jcensing Board

- RALPH L TETRICK (Denial of Application for Reactor Operator License)

Docket 55-20726-SP (ASLBP No. 97 727-01.SP R)

(Re: - Senior Rea' tor Operator License)

MEMORANDUM AND ORDER, LBP-97-II, June 25, 1997 ....... 441 Issuances of Directors' Decisions ADVANCED MEDICAL SYSTEMS,INC.

(Cleveland, Ohio)

Docket 030-16055 DIREC. T OR'S DECISION UNDER 10 C.F.R.12.206, L DD-97 13, J une 13, 1997 . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . 460

' CONSUMERS POWER COMPANY (Palisades Nuclear Plant)

Dockets 50 255,72-7 DIRECTOR'S DECISION UNDER 10 C.F.R. 6 2.206, DD-97 15, J une 18, 1997 , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 475 ENTERGY OPERATIONS. INC.-

(Arkansas Nuclear One, Units 1 and 2)

Dockets 50-313,50-368,7213 DIRECTOR'S DECISION UNDER 10 C.F.R. 6 2.206 DD-97 15, J une 18, 1997 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

. 475 111

l GENERAL PUBLIC UTILITIES NUCLEAR CORPORATION (Oyster Creek Nuclear Generating Station)

Docket 50-219 DIRECTOR'S DECISION UNDER 10 C.F.R. 9 2.206 DD-97 14. J une 16, 1997 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 472 GEORGIA INSTITUTE OF TECHNOLOGY (Georgia Tech Research Reactor, Atlanta, Georgia)

Docket 50-160 FINAL DIRECTOR'S DECISION UNDER 10 C.F.R. 5 2.206, DD-97-16, June 27,1997 . . . . . . . . . . . . . . . ... ... . . . . 487 SHIELDALLOY METALLURGICAL CORPORATION (Newfield, New Jersey)

Docket G40-8948 (License No. SMB 1507)

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

' DD-97 12, J une 6, 1997 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 449 SIERRA NUCLEAR CORPORATION Docket 72-1007 DIRECTOR'S DECISION UNDER 10 C.F.R. 9 2.206, DD 15, J une 18, 1997 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 475 WISCONSIN ELECTRIC POWER COMPANY (Point Beach Nuclear Plant. Units 1 and 2)

Dockets 50-266,50-301,72 5-DIRECTOR'S DECISIC-N UNDER 10 C.F.R. 5 2.206, DD-97-15, June 18,1997. . . . . . . .......... ....... .... 475 iv

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Cite as 45 NRC 437 (1997) CLl-97 7 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Cs 'NilSSlONERS:

Shirley Ann Jackson, Chairman Kenneth C. Rogers Greta J. Dicus Nils J. Diaz Edward McGaffigan, Jr.

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

(Claiborne Enrichment Center) June 30,1997 l

The Commission grants petitions filed by the Staff and Louisiana Energy Services for Commission review of the Atomic Safety and Licensing Board's May 1,1997 Final Initial Decision, LBP-97-8,45 NRC 367 (1997), and sets i a briefing schedule pursuant to 10 C.F.R. 5 2.786(d). 'Ihe Commission also denies Nucles Energy Institute's (NEl's) motion for leave to file an amicus curiac brief in support of the petition for review, RULES OF PRACTICE: AMICUS CURIAE Our rules contemplate amicus curiac briefs only after the Commission grants a petition for review, and do not provide for amicus briefs supporting or opposing petitions for review. See 10 C.F.R. 6 2.715(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 May 1,1997 Final Initial Decisioc, LBP-97-8, 45 NRC 367 (1997), concerning contention L9 (raising " environmental justice" claims).

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This proceeding involves LES's application for a license to construct and -

operate the Claiborne Enrichment Center (CEC) near 11omer, Louisiana. The Intervenor, Citizens Against Nuclear Trash (CANT), opposes the petitions for Commission review, In accordance with the censiderations 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.'

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1. _ SCHEDULING OF BRIEFS Pursuant to 10 C.F.R. 6 2.786(d), the Commission sets the following briefing schedule:2

1. The Staff and LES shall file their briefs on or before August 8,1997. l

. Each brief shall be no longer than 30 pages. 4

- 2. - CANT shall file a single responsive brief on or before September 18, l 1997. Its response shall not exceed 40 pages. We allow 40 pages for )

CANT's brief so that CANT will have adequate space to respond to separate approaches that may be taken in the opening briefs of the Staff and LES. It is also possible that CANT will face an amicus curiac brief filed by NEl. Sec discussion below.

3. The Staff anri LES may file reply briefs on or before September 30, 1997. Their replies shall not exceed 10 pages each.

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 other 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. MOTION TO FILE AMICUS CURIAE BRIEF IN SUPPORT OF PETITIONS ,

The Nuclear Energy Institute (NEI) has sought leave to file an amicus curiac brief in support of the petitions for review, We deny the motion. Our rules 8

The Conumssion also has before it three peuuons for review, two by CANr and one by 1.Es. raams vanot s challenges to the Board's handhng of waste disposal issues, including us decismo in 1.BP 97 3. 45 NRC 99 (1997). to a&huon. the Conmussion is considenas the bnefs 61ed by the parties after the Comnusson gramed earher peuuons for review raung NEPA and 6aancial quahncations luues See CLI-97 3,45 NRC 49 (1997k The Commission will act on those maners in due course.

2 In a tener dated June 5,1997. CANr's lawyers asked the Comnussion. in setung a bnenng schedule. to take' two consideranon their "previously scheduled fanuly obhganons out of town dunng the enure snonth of August!'

Lf5 opposes any driay in the proceeding The Comnussma has taken into accous both concerns in estabbstung the bnehng schedule in this case, 438

contemplate amicus curiar briefs only after the Commission grants a petition for review, and do not provide for amicus briefs supporting or opposing petitions for review. Ser 10 C.F.R. 6 2.715(d); cf. Sequoyah fuels Corp. and General Atomics (Gore, Oklahoma Site), CLI.96 3,43 NRC 16,17 (1996). No special circumstances here warrant an exception to our rules.

Without further motion, however, we will permit NEl to file an amicus brief on the merits, not to exceed 20 pages, should it choose to do 50. Ser Sequoyah Furls Corp.,43 NRC at 17. NEl must file its amicus brief no later than the filing date of the briefs for the parties whose position NEl supports. See 10 C.F.R. I 2.715(d).

IT IS SO ORDERED.

For the Commission' JOHN C. IiOYLE Secretary of the Commission Dated at Rockville, Maryland, this 30th day of June 1997.

3Comni.woner thcus was act available for the artrmanon of dus order If she hat been tusent, she would have approved me order 439

Atomic Safety and Licensing Boards issuances ATOMIC SAFETV AND LICENSING BOARD PANEL B. Paul Cotter, Jr.,* Chief Administrative Judge James R Gleason,* Deputy Chief Administrative Judge (Executive)

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

Members Dr. George C. Anderson Dr. Richard F. Foster Dr. Kenneth A. McCollom Charles Bechhoefer* Dr. David L. Hetrick Marshall E. Miller Peter B. Bloch* Emest E. Hill Thomas S. Moore

G. Paul Bollwerk lil* Dr. Frank F. Hooper Dr. Peter A. Morris Dr. A. Dixon Callihan Dr. Charles N. Kelber* Thomas D. Murphy * ,

Dr. James H. Carpenter Dr. Jerry R. Kline* Dr. Richard R. Parizek Dr. Richard F. Cole

Dr. Peter S. Lam

Dr. Harry Rein Dr. Thomas E. Elleman Dr. James C. Lamb Ill Lester S. Rubenstein Dr. George A. Ferguson Dr. Linda W. Little Dr. David R. Schink Dr. Harry Foreman Dr Emmeth A.Luebke Dr. George F.Tidey

Permanent panelmembers

Cite as 45 NRC 441 (1997) LBP 9711 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD PANEL -

i Before Administrative Judges:

Peter B. Bloch, Presiding Officer

Dr. Peter S. Lam, Special Assistant ;

In the Matter of Docket No. 55-20726 SP (ASLBP No. 97 727 01-SP R) i (Re: Senior Reactor Operator License)

RALPH L TETRICK (Denial of Application for Reactor Operator License) June 25,1997 he Presiding Officer in this Subpart L proceeding, having requested further information i this remand proceeding, affirmed his earlier determination that Mr. Tetrick had incorrectly answered the remanded question on his Senior Reactor Operator's examination. Plant procedures involved in this question were interpreted to require an understanding of the root cause of the incident described in the question.

RULES OF PRACTICE: MOTION FOR RECONSIDERATION; REMAND Re Presiding Officer expressed confidence that in deciding this case the Commission will be aware that motions for reconsideration are frequently filed before presiding officers, both at the end of cases and after interim orders. Puerto Rico Electric Power Authority (North Coast Nuclear Plant, Unit 1), ALAB-648, 14 NRC 34,37-38 (1981).

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MEMORANDUM AND ORDER (Determination of Remand Question)

Memorandum

. 'Ihe purpose of this Memorandum is to determine the question remanded to me by the Commission, in light of the additional evidence provided to the Commission on appeal and then to me in response to questions asked of the

parties.

L PROCEDURAL HISTORY On May 20,1997, the Commission issued CL197-5,45 NRC 355 (1997),

concerning an appeal of my initial decision, LBP 97 2,45 NRC 51, $3 (1997).

In that decision, the Commission charged me with redetermining the correctness of Mr, Tetrick's answer to Question 63 on his examination, in light of a letter of May 1.1997, from R.J. Hovey, Vice President of the Turkey Point Plant (Hovey letter).' The Hovey letter was submitted by the NRC Staff to the Commission-

- as an attachment to a Staff brief filed on May 2,1997.

On May 27,1997, I issued an unpublished Memorandum and Order in which I asked the parties a series of questions designed to elicit information helpful in determining this remand. In response, the parties filed: (1) Memoranda from l Ralph L. Tetrick, with attachments (including plant procedures, a letter from l- R.J. Hovey of May 1,1997, and a Memorandum from Brian J. Stamp, undated) dated June 6,1997 (Tetrick Answers); and (2) "NRC Staff's Response to the Presiding Officer's Memorandum and Order (Questions Relevant to Remand), ,

June 13,- 1997 (Staff Answers) and " Supplemental Affidavit of Brian Hughes and Thomas A. Peebles, June 13,1997 (Staff Supplemental Affidavit).

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, IL QUESTION 63 Examination Question 63, which is the subject of this remand, stated as follows:

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~3 Unless there 6s a showing of"compelbng cause." matters raised for the Arst time on appeal generally will not be 4

conssdered, especially when they involve factual matters that could have been raised before the presiding offtcer.

Puerre Are Bertric Power AurAarirr (North Coast Nuclear Plant. Unit 1). A1.AB-648,14 NRC 34,37.18 (1981).

In accordance with the Comnumon's direenons in dus remanded case, the parues' Ehngs before the Commission

- are considered to be a part of the decisional record.

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~ Plant {.rnditto0Ls:

. Preparanons are being madefor refuehng operations.

- The refuehng cavity is filled with the transfer tube gate nahe open

~~ Alarm annunciators H In. SFP LO LEVEL and G=95, Ch7 aft Suhlt HI LEVEL ore in clarm.

Which ONE of the following is the required IMAIEDIATE ACTION in response to these crmdaions?

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. a. Veryy alarms by checksng containment samp level recorder and spent fuel level sndication.

, b. Sound the containment evacuation alarm.

$ c, initiase crmtainmeru ventilation isolation.

d ' Initiate control room vent.lation isolation.

III. 'THE INITIAL DECISION .

In my initial decision, LBP-97 2, I decided, based on the record then before

me, that

- The Staff has persuaded me that when two concurrent annunciators sound, indicating that there is an off normal event that could cause harmful radiation within the containnwnt, that

, 14 operator should take the sequired IMMEDIATE ACTION. Given the important safety IJ ' problem that is being indicated by two different annunciators, there is not the time to verify

, that each of the annunciators is working properly. That they sound together is enough

. conoboration to act immediately to prevent injury to the heakh of plant employees.

45 NRC at $5. Thus, I concluded that the correct response to this question was "b" rather than "a," which was Mr. Tetrick's answer.

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IV. ADDITIONAL INFORMATION

'A. . Applicable Plant Procedures

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' Mr.-Tetrick has demonstrated, in his memorandum of June 6,1997,' that 3-k ONOP2 033.2 - Refueling Cavity Seal Failure is not the only plant procedure that requires an immediate action. The phrase "immediate action" also occurs 1 in 3-ARP3 097.CR - Control Room Annunciator Response and in 3-ONOP-

033.1 ' Spent Fuel Pool (SFP) Cooling System Malfunction.

3 ONoP stands for "off normal operaung procedure *

-3 ARP stands for

annunciator response procedure

and also is referred to as

annunciator response guidehnes."

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B. Important " Note" Contained in Procedure in the attachmer.ts filed with me by Mr.Tetrick, on page 7 of 3-ARP-097.CR, there is a box that sets forth a Feneral principle that the indicated actions are "a guide for operators in responding to single annunciafors." Note that they are "a guide." Note also that they apply to single annunciators and not to multiple annunciators, where understanding the pattern or the root cause becomes more important and where " applicable off-normal and emergency procedures" come into play. The relevant section of 3-ARP-097,CR, called NOTES, states:

1. The annunciator panel attachments indicate apprornate operator action for Control Room panel annunciators The acuans hsted are intended to be a guide for operaton in respondmg to smgle annunciators and not intended to be a substitute for good judgment bawd on thorough understandmg of plant conditions and equipmen*

2. hiany off normal plant conditions will result in several annunciators hghting alnost simultaneously. In such a case, operators are expected to respond to the root cause of the problem and maintain the unit m a safe condition LAW Im accordance with) apphcable off-j normal and emergency procedures. His a tion may not necessanly correspond to that of j the attachments.

l C. Staff Argument The Staff has discussed extensively the root cause of the signals postulated to be present in Question 63. It bases its answer to the question on this understanding of root cause. It states (Staff Supplemental Affidavit at 9-11):

We have carefully considered htr Tetnck's answer to this question. In our view, it rehects a fundanwntal nusunderstandmg of the importance t.nd signincance of an ONOP, in contrast to a nuclear facihty's many othen plant procedures. Further, hir, Tetnck's answer ignores the signincance of the spectnc plant conditions desenbed in the stem of Question 63, which nuist be considered in an SRO apphcant's selection of the proper answer to ilus question.

Question 63 exphetti) posited the following speenhc plant conditions:

Plant conditions:

- Preparations are being made for refueling operations.

- The refuchng cavity is hlied with the transfer tube gate valve open.

- Alarm annunciators H l/1, SFP LO LEVEL and G-9/5, CNTh1T SUh1P 111 LEVEL are in alarm Under these plant conditions, w here these two mutually supportive and conhrmatory annun-ciators (spent fuel pool low level and containment sump high level) are soundmg together, a competent apphcant for a senior reactor operator hcense should have recognized, unequivo-cal:y, that the operator is required to sound the contamment evacuation alarm, in accordance with 3 ONOP-033.2. We note that although htr.Tetrick's July 1996 submittal did not discuss tlus ONOP, in his h!ings before the Presidmg Ofhcer in September and December 1996 he 444 I

agreed the two annunciators specified in Quewuon 63 are

mutually supportive and sufficient to enter 3 ONOP 033.2 " REFUELING CAVITY SEAL FAILURE."

. , Queshovi 63 does not constitute an abs'ract qmtion of only theoretical interest. Rather, the question seeks to test apphcants on their fu9damenth! co ripetence to respond to actual plant condiuons, specifted thewin Question 63 descithes a potential refuehng cavity seal failure, during refueling operationt Tiw initial plaru con &uoL provided in the stem of the question state that "the refuelaig cavity is filled with the transfer tube gate valve open." This condition rneans that the Spnt Wel Pool is connected tthrough the transfer tubel to the refuehng cavtty in the Contamment Buildmg Another imtial condition states

" Alarm annunesators 11 1/l, SFP LO LEVEL and G 9/5, CNTMT SUMP HI LEVEL are in alarm." The concurrent soundmg of these two alarms would indicate that the water level has decreased in the Spent Fuel Pool and has increased in the Contaanment Buildmg sump.

Because the Spent Fuct Pool is connected to the Refuchng Cavity (inside the Containment Building) through the transfer canal, the actuation of these two alarms at the same tune would conftrm leakage from the Refucimg Cavity to the Contamment Building sump. This leakage would most probably be due to the refuchng cavity seal leaking or faihng. Under the conditions described in Question 63, prompt notificanon to plant personnel of the nature of the emergency by soundmg the contamment evacuation alarm is the only appropnate IMMEDIAni ACTION.

, Question 63 is based upon a real-hfe incident that occurred at the Haddam Neck plant, where a refuehng cavity seal failure resulted in a substantial dramage of the water in the refueling cavity within a matter of nunutes - an event which could have potentially resuhed in lethal radiation doses to plant personnel. This event led to the issuance of IE Bulletm 84 03 on August 24,1984. At the time of the event, the iefuehng cavity was filled in preparation for refuelmg and, fortuitously, the transfer tube gate valve (which connects the spent fuel pool to the refuchng cavity) was closed. The Staff evaluated tlus event as Genenc Issue 82, and deternuned that it has signiicant safety irnphcauons for all water-cooled nuclear power plants in the United States - and each such facahry. includmg Turkey Point, was required to address this problent See NUREG/CR-4515. "Closcout of IE Bulieun 84 03: Refuelmg Cavity Water Scal"(June 1990)(portions of which are provided as Attachment I hereto).

It should be further noted that Quesuon 63 posits a situation in which "the refuelmg cavity is hited with the transfer tube gate valve open"- unhke the event at fladdam Neck, where the gate was closed. While sigmfacant radiation dosos may have been avoided at fladdam Neck due to the transfer tube Fate being closed, a different result might have occurred at Turkey Point, under the condiuons stated in Question 63, if the plant operators decided, like Mr.

Tetnck, to verify alarms before takmg the required "lMMEDIATE ACTION" of soundmg the containment evacuation alarm.

V. ANALYSIS AND CONCLUSIONS I am persuaded by the Staff that I should uphold my initial determinatior An operator must act on an understanding of the root cause of an event, trusting the plant's instruments to deduce what is happening, 'Ibrkey Point does have procedures for " responding to single annunciators," Note from 3-ARP-097.CR, discussed above at p. 444. As also discussed above, at p. 444, these procedures l

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specifically state that they are "not intended to be a substitute for goodjudgment based on thorough understanding of plant conditions and equipment."d I asked several questions in my order of May 27. Among those questions were the following:

What precisely would he (Mr. Tetrickl do dunng these 20 seconds Ithat he says he would use to venfy the vahdity of instrument readings]? What evidence might he find that would persuade him not to take the required IMMEDIATE ACTION after he took steps to venfy the alarm?

He answers to these questions were very important because they would show whether there was any legitimate reason to hesitate in taking the immediate action required by the ONOP. For example, is there some instrument reading that could be easily taken and that would give an operator confidence that the instruments were wrong? If so, then the decision to check further could be based on an understanding of what was happening in the reactor and not based solely on a mechanical reading of a tangential provision that relates to single annunciators. However, Mr. Tetrick did not respond directly to my question. In -

particular, he gave no indication of any instrument reading or set of readings that would persuade him not to take the required immediate action in the 3-ONOP-033.2. Tctrick Answers, bottom of p.1 (responding to Question #2).

I conclude that Mr. Tetrick should have acted from an understanding of the root cause of the event portrayed in Question 63. Ilad he done so, then only answer "b." would be correct. His failure to understand that failed to mitigate the risks described by Staff and quoted at p. 445, above.

I am unpersuaded by Mr. Tetrick's attempt to rely on the hrkey Point training program and " management expectations." See Tetrick Answers at 1, second paragraph from the bottom. He is responsible for knowing the correct, safe action to take in response to plant conditions. The NRC cannot be expected to certify an operator based on his reliance on an incorrect response allegedly taught to him. NRC licenses only those operators who demonstrate that they will respond correctly and safely to plant conditions.

I am not convinced by the letter from RJ. Hovey of Florida Power and Light to Mr. Stuart A. Richards of the NRC. (Tetrick Reply, unnumbered Attachment.) Mr. Hovey states, in one key sentence, "If the question is interpteted to be asking for an immediate action for the receipt of an annunciator, response (a) is correct." I do not interpret the question as Mr. Hovey suggests.

There is not one annunciator, but two. What is called for by the question is an understanding of plant conditions and how to respond to two consistent, d

Procedure loNoP.0D 1 requires an "imrnediate accon" conustiis of-. "venfy annunciated alarm is vahd."

However, with the simultaneous indicanons postulated in Quesuon 63, the two alarms venfy the vahdity of one another Thus, there is no funher need to venfy these alarms.

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simultaneous annunciators. htoreover, the Annunciator Response Procedure (ARP) contains a note that makes it clear that it cannot be mechanically applied under these circumstances. (Scr Note 5 of 3-ARP-097, CR, above.)

Similarly, I am not persuaded by the memorandum of Brian J. Stamp, Acting Operations Supervisor, because I consider his understanding of Question 63 to be the same as that of Mr.Tetrick and thus incorrect. (Tetrick Reply, unnumbered Attachment).

I conclude, after considering all the information before me, that Mr. Tetrick answered Question 63 incorrectly.

VI. PROCEDURAL IMPLICATIONS In this remand, I have addressed information filed by Mr. Tetrick that was not filed in a timely manner prior to my Initial Decision. I would note that the Staff's appeal also seems to be based on new information. I am confident that in deciding this case the Commission will be aware that motions for reconsideration are frequently filed before presiding officers, both at the end of cases and after interim orders. It is important for the efficiency of licensing procedures that there be a clear principle that requires parties to file information prior to the decisions of judges rather than waiting for an opinion before adding new infomiation to the record.

Order For all the foregoing reasons and upon consideration of the entire record in this matter, it is, this 25th day of June 1997, ORDERED that:

In response to CLI 97-5,45 NRC 355 (1997), the Presiding Officer reaffirms his determination that the response of Ralph L, Tetrick to Question 63 of his Examination to be a Senior Reactor Operator (SRO) was incorrect.

Peter B. Bloch, Presiding Officer ADMINISTRATIVE JUDGE Rockville, Maryland 447

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Cite as 45 NRC 449 (1997) DD-9712 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION I

l OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS Carl J. Paperiello, Director in the Matter of Docket No. 040-8948 (License No. SMB 1507)

SHIELDALLOY METALLURGICAL CORPORATION (Newfield, New Jersey) June 6,1997 i

i By a letter dated July 22,1996, Mr Sherwood Bauman (Petitioner) requested

j. that the following actions be taken with regard to NRC Licensee Shieldalloy Metallurgical Corporation (SMC): (1) that the previous site Licensee have its license reinstated such that it and SMC become co-responsible for the remediation and decommissioning of the SMC site; (2) that all NRC or State of Ohio parties involved in wrongdoing related to this issue be dismissed from employment and criminally charged where appropriate; (3) that the NRC terminate its development of an environmental impact statement (EIS) for the SMC site; (4) in place of the EIS, the NRC order SMC and its predecessor to submit a decommissioning plan limited to remediation of licensed material; and (5) that the Ohio Environmental Protection Agency and Department of Health should evaluate all unlicensed slag found at the SMC site. He request was considered as a petition submitted pursuant to 10 C.F.R. 5 2.206.

In a Director's Decision issued on June 6,1997, the Director of Nuclear -

Material Safety and Safeguards denied the relief sought by Petitioner. The Director concluded that it would be inappropriate to reinstate the previous Licensee's license for the SMC site, as SMC was the current Licensee and therefore responsible for decommissioning the site. For similar reasons, the Director denied Petitioner's request to order SMC and the previous Licensee to submit a decommissioning plan. With regard to Petitioner's allegations of wrongdoing, with respect to any such activity by NRC employees the allegation was referred to the NRC Office of the Inspectc. General. He Director also 449

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. concluded that the current EIS was properly evaluating all slag at the SMC site, contrary to Petitioner's claim that the scope of the EIS exceeded NRC authority.

- Finally, the Director concluded that Petitioner's request for action by State of

- Ohio agencies was properly addressed by those agencies and not the NRC, DIRECTOR'S DECISION UNDER 10 C.F.R. 6 2.206 I.' INTRODUCTION By letter dated July 22, 1996, addressed to the U.S Nuclear Regulatory Commission (NRC) and Ohio Department of Health (ODH), Sherwood Bauman, Chairperson of the organization "Save Wills Creek Water Resources Committee" (Petitioner), requested certain actions concerning NRC Licensee Shieldalloy Metallurgical Corporation (Shieldalloy) and former NRC Licensee Foote Mineral (now Cyprus Foote Mineral Company (CFM)). NRC is treating the request as a

- petition under 10 C.F.R. 5 2.206 of the Commission's regulations. He Petitioner ,

i requested that the following actions be taken: j (1) . NRC should reinstate Foote Mmeral's original license so that Shieldalloy and CFM become co-responsih!c licensees conceming the proper remediauon and decomnussioning of the Shieldalloy site;

-- (2) Any and all parties involved in any wrongdoing, as alleged in the Peutioner's letter, should be terminated from employment, and, where appropnate, ennunal charges pursued, (3) - NRC sin >uld terminate the development of the environmental impact statement tEIS) for the Shieldalloy site; (4) . In place of the EIS, Shieldalloy and CFM should be jointly ordered to submit a decommissioning plan, for heensed material, that includes only a plan to remediate licensed matenal, includmg gradmg and evaluation of all various assorted options.

- One opuon considered should be offsite disposal at a licensed disposal facility; and

($) The Ohio Environmental Protection Agency (OEPA) and Ohio Department of Health tODH) should evaluate all unlicensed slag found at the Shieldalloy site.

- NRC acknowledged receipt of the petition in a letter to the Petitioner dated October 11,1996. The petition was also noticed in the Federal Register on April 10,1997 (62 Fed. Reg,' 17,650),'

I Normally, the Petitioner (by lener) and the pubhc (through a F.deral Reguser retice) are nonfied at approumauly the same nme. In tids case, because of an adnumstranve onussion. the Fedend Assurer nouce was not pubhshed untd Apnl1997/

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De Petitioner also sent an undated letter to Presidetit Clinton at appruximately the same time as his July 22, 1996 letter to NRC. The White House referred that letter to NRC for response. All of the substantive issues raised in that letter are addressed in this Director's Decision.

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11. IIACKGROUND Plant Illstory l l

Shieldalloy owns and operates a plant that produces ferroalloys, near the l city of Cambridge, Ohio. Cambridge is in eastern Ohio, approximately 130 km (80 miles) east of Columbus, Ohio, on Interstate 70. The facility is between Cambridge and Byesville, Ohio, and within the valley of Wills Creek, the major stream in the area.

Rrroalloys are mixtures (alloys) of iron and one or more other elements (e.g.,

vanadium, titanium, and niobium) that are typically used in steel production or other alloy manufacturing processes. The principal alloy produced today at the Shieldalloy plant is a 60% vanadium /40% iron alloy. Shieldalloy sells its product to steel manufacturing companies, which then add it to batches of steel to produce vanadium alloy steels with a fraction of 1% concentration of vanadium. Vanadium imparts increased strength and hardness to steel.

l, Facility operations began in the early 1950s under the ownership of Vanadium l Corporation of America (VCA). Foote Mineral Corapany merged with VCA in 1967, in 1987, Shieldalloy purchased the facility from Foote Mineral Company and has continued alloy production at the site since then. The plant has produced a variety of alloys for the steel industry over the years.

He production of metal alloys has resuhed in waste byproducts, the principal one being slag, a hard, rock like residue. During alloy production, almost all radionuclides contained in the incoming ores were incorporated into the waste slag. Since the inception of operations until the late 1980s, the facility disposed of most of its waste slags and other wastes on site. At the present time, the

- facility's waste is largely contained in the East and West slag piles (named for their onsite locations). Together they contain approximately 250,000 cubic meters (seven million cubic feet) of slag and cover approximately 5.7 hectares (14 acres) of land. The slag itself contains both radioactive materials, such as uranium and thorium isotopes (including their daughter products: such as radium and radon), and nonradioactive metals such as vanadium, chromium, arsenic, copper, and zinc. He slag produced today is largely recycled in steel 2

fan.g4rer producu are atortue species (or p38hd.s) formed by the ra&oacuve decay or another auchde, wruch is caHed the " parent." For example. when U decays. Th* is produced. This thermm isotope also decays and produces a.hhoonal ' daughters " Radium and radon are daughter products in the uranium decay chain.

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manufacturing as a flux, i.e., a material that removes impurities. Shicidalloy halted onsite slag disposal in the late 1980s.

Several types of radioactive slag are contained in the East and West piles. In the early years of plant operation, ferrocolumbium (now known as ferroniobium) ores were used for alloy production. These ores contained licensable quantities of source material (i.e., uranium (U) and/or thorium (Th) in concentrations greater than 0.05'7c). See 10 C.F.R. Il 40.4, e 13(a). The slag from processing these cres contains elevated concentrations of U '8 and Th and their daughter products, and emits gamma radiction that is easily detected.

Two other types of slag at the site, ferrovanadium and Graina18, are also radioactive, but neither was produced under the original license that exp;ted in 1975. Radioactive ferrovanadium slag is believed to have resulted from the plant using vanadium concentrates as feed material for alloy production. These concentrates probably resulted from ores processed in another facility to remove the uranium for use in weapons and/or nuclear fuel production. He radioactive daughter products of the uranium, such as Th*, and valuable elements, such as vanadium, remained in the byproduct material. Only small amounts of the parent radionuclides, US and Um, and much less than would be expected in material that had not been processed to remove these radionuclides, are present.

Unlike the ferroniobium slag produced under the original license, the ra-dioactivity of the ferrovanadium and Grainal# slags is difficult to detect. Some radionuclides in the ferroniobium slag are strong emitters of gamma radiation, which can easily be detected with hand held instruments. The ferrovanadium and Grainal* slag radiation is principally emitted as alpha particles from Th*,

which are much more difficult to detect with field instruments. The significan!

radioactivity in these slags was not discovered until after they were produced.

The license issued to Shieldalloy in 1987 is for " uranium and tho.;un . . as a contaminant in slag from previous alloy furnace operations."

I Doth the radioactive materials and metals contained in the onsite East and West slag piles could have potentially adverse effects on human health and the environment. In fact, some metals have leached into streams and sediments next to the slag piles. Little or no migration of radioactive materials has taken place to date. Because of the potential effects of the slag on the environment and human health, both the State of Ohio and NRC plan to oversee remediation and cleanup of contamination at the site.

Ferrovanadium slag containing small amounts of radioactive contamination and possibly other slag with radioactive elements have been used in some residential and commercial properties in the Cambridge, Ohio area. The slag was sold or given away by the company for use as construction and driveway fill material before 1987. The short-term hazard from this slag is negligible. He long-term hazard is small and principally derives from unlikely scenarios such as a family growing crops adjacent to their driveway for their consumption as 452

. _ . . - . _ _ ~ - . , . .. _ ,_ _ _ _ . _ _- . _ _ _ _ _._ _ _ _.__

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food. hiost calculated doses are a fraction of background radiation. CFht has a G separate program under way to identify these properties, evaluate any long-term hazards, and perform any necessary remediation. Several properties have been 4

identified for remediation, and CFht is taking steps to remow the material and safely store it elsewhere. Although the homeowners possess the slaF, CFht is carrying out measurer to ensure that the offsite slag is addressed, although

~ CFht is no longer a licensee, NRC and the State of Ohio are overseeing CFht's

i evaluation and remediation of these offsite prcperties, and have met with the j public in the area to discuss the issue.

NRC Regulatory Program Related to Decommissioning the Shieldalloy Facility VCA and its successor, Foote hiineral Company, held a license to possess -

source material from 1953 to 1975. At that time. Foote hiineral allowed the license to expire and did not request its renewal, although it continued to possess source material. in 1987, Shieldalloy obtained an NRC license (ShiB-1507) for the possession of the source material at the facility. When a licensee is no longer performing the principal activities for which the license was issued (in this case, metal alloy production from radioactive ores), NRC regulations require that the site be decommissioned and the license terminated, See 10 C.F.R. I 40.42. 'Thus, NRC's regulatory program for the Shieldalloy site is directed toward these goals, in 1987 and 1990, Shieldalloy submitted decommissioning plans to NRC

. proposing in-situ disposal of the slag piles. Subsequent to the development of these plans, however, NRC determined that an environmental impact statement

- (EIS) would need to be prepared, in accctdance with NRC regulations contained in 10 C.F.R. Part 51, which implements the National Environmental Policy Act of 1969 (NEPA). In order to evaluate Shieldalloy's proposal for onsite disposal of the slag piles, it is necessary to assess impacts on the environment, through

_ preparation of an EIS. The EIS examines onsite disposal alternatives, as well as other alternatives including offsite disposal of the slag.

Under the Atomic Energy Act of '1954 (AEA), NRC is responsible for regulating the safe use of certain radioactive materials (source, byproduct, and special nuclear radioactive materials) to ersure that public health and safety are protected from the effects of radiation. Under NEPA, NRC is obligated to take a range of environmental impacts into account in its decisionmaking process on

~ decommissioning attematives. The environmental costs of an action are to be

- weighed against its benefits. As described above, NRC considers the regulatory decision on decommissioning of the Shieldalloy facility to be a major federal action that may significantly affect the quality of the human environment. For that reason, and pursuant to NRC regulations in Part 51 implementing NEPA,-

NRC is preparing an EIS. The scope of the EIS includes both radiological and 453

l nonradiological impacts of the proposed action and alternatives to it including impacts on land use, air quality, noise, and transportation, in addition to the radiological impacts to the public that NRC regulates under the AEA.

When the EIS is completed (espected to be in late 1997). Shieldalloy will be required, under NRC regulations in section 40,42, to submit a revised de. l commissioning plan consistent with the findings of the EIS.1hus, Shieldalloy's ;

previous submittals of decommissionig plans will be superseded by the newest -

one. -

i Ohio's Regulatory Program for Remediation of the Shieldalloy Site The Ohio Environmental Protection Agency also has a program to oversee remediation of the Shieldalloy facility, consistent with its implementation of the Comprehensive Environmental Re:ponse, Compensat!sn, and Liabilities Act (CERCLA). Ohio's effort covers contamination at the facility and on property nest to the site (mostly wetlands and stream sediments). Vanadium compounds and other waste have migrated into soils and sediments, both on site and off site, and into a stream that runs through the property. The State has entered into a proposed Preliminary injunction Consent Order (Consent Order) to require Shieldalloy and CFM to carry out a remediation plan described in i Ohio's Decision Document.8 The proposed Consent Order would also requ.

! Shieldalloy and CFM to pay civil penalties to Ohio. Public comrnents were ,

i received by the State of Ohio on the proposed Consent Crder, and it is expected

! to be rnade final in the near future.

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) Relationship lietween State of Ohio and NRC Programs for Shieldalloy

) Remediation of the Shieldalloy site involves various potential polluta* reg *:

} lated under overlapping laws. Ohio is responsible for overseeing the remediation i

of contamination under the CERCLA process. Pursuant to NEPA, NRC is re.

sponsible for considering the impacts on the environment from all contamination

] at the facility in evaluating various alternatives for site remediation. Under the ;

j AEA, and once the EIS is completed and a decommissioning program approved, i NRC is also responsible for ensuring that the site is properly decommissioned,

in this case meaning that radiological contamination is reduced to safe levels, ,

! and if onsite disposal is approved. shat appropriate institutional ccntrols for long. :

! term land use and monitoring are established and implemented. NRC and the l State of Ohio have been coordinating their individual efforts e ensure that a y .

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3 i 0tuo 1.nvironmrmal Protecuon Agency theimon Docunum for the stueklalloy Metauurgical Corporaud S te, i Candedge, ohio LPA, April 1. len. i i

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I I coordinated approach to site remediation is required of Shieldalloy and Cf'M by f 4 the Federal and State governments.

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[ 111. DISCUSSION

NRC Staff has examined the Petitioner's requests in his Petition of July 22, ;

1996, as follows: 5 4

j (O The NRC should reinuate Tht's enginal heense so that $hield.dluy and CT'M tv- f come co responsible licenwes with regard to the proper remediation and decond j missionita of the $hicidalloy site.

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The Petitioner argues that Foote Mineral should now be made a co responsible i

licensee along with Shieldalloy because Foote Mineral allowed the license to j expire and it was not appropriately retired by NRC. De Petitioner states that !

NRC did not investigate the Licensee's claims that no materials of licensable -l j concern were remaining on site when the license expired. .

i' In a September 9,1975 letter, the NRC notified Ibote Mineral Company that t

' its Source Material License (SMB 850) had expired on August 31.1975, FMC is submitted a " Certificate of Disposition of Materials. AEC 314" to the NRC on l September 15.1975, and the NRC retired the license on October 14,1975. A .I i- site visit was not conducted by NRC Staff to verify disposal of the licensed material. As NRC stated to'the Petitioner in a January 19, 1995 letter from i

, NRC's Region !!! cffice, j Ahhough the licenw 'nord is unclear,it appears to NRC staff that Fume Mineral may have mistakenly assumed that thonuin and utenium in the slag were no longer considered source - t matenal because their concentrations were generally less than 0.05% tiy weight. The NRC ,

settred the beense based on the cornpleted AIC)l4 form, which indicated that "No malertals ('

have twen procured tiy the hcensee."4 Retirement of expired licenses without conducting an onsite inspection was f

accepted NRC practice in 1975, although the policy has since changed to j- require onsite inspection to verify that sites of this type have been properly-

decontaminated. There is no evidence that Foote Mineret Company personnel 4,

committed any wrongdoing in this matter, ,

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The January 19.199$ letter from NRC's Region ni ofrice to the Pet 6uoner semed that arc rerm 314 ladicawd i l *

that *all semaining source mmenal(e 3. eses) had been tramferred and so longer esisted on the Camhndt e mie Sime the hcense was seured and beenned opernoons ceased, the NRC Ad not in pect dunng the pened of october

- 1975 until early 1987? AEC Form 314 sia.es that "No mawnah have been procured by the beensee.' as noted -

J nhove In any case, there appears to have been conivuon by the thenwe over what connutuisd source matenal. ~

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and the IAensee appears to have nusinkenly assumed the the slag was not covered by the ensiing beense NRC's 4 : totmng of the beense was based on the informanon in Al.C.)l4 that no beenuble matenal was on site.

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l With respect to issuinF a license to CFht, NRC's li:ensing authonty is contained in the AllA, and specific licensing provisions have been incorporated !

into NRC's reFulati ons in 10 C.F.R. Part 40. Ste 10 C.F.R. 6 40.l(a). ne !

regulations generally require that, where applicable, a possessor of radioactise ,

materials chain an NRC license. Shieldalloy is thv owner and ponessor of l the slag piles, and controls them in accordance with NRC license ShtB 1507.

l Thus, NRC regulates the radioactive materials on the Shieldalloy site through '

its Licensee, Shicidalloy hietallurgical Corporation.

l ne State of Ohio, however, has entered into a propostd Consent Order l with Shieldalloy and CFhi that would require those companies to irnplement I

ternediation activities at the site. Rus, the Petitioner's request that CFht be made co-responsible for remediation of the site is satisfied in part by that Consent Order CFhi's responsibility, however, is defined and required by the proposed Consent Order with the State of Ohio and not by NRC license as the Petitioner had requested. NRC is satisfied that this approach is adequately protecting public health and safety.

Ibr the above reasons, the Petitioner's request that CFht be made a co-responsible Licensee for remediation of the site is denied.

(2) Any and all parties involved in any wrongdoing, as alleged in the Petiuoner's letter.

Should tw terminated rrom employrnent, and where appropnaie cnmmal charges pursued As a general matter, NRC takes enforcement action against individuals who engage in deliberate misconduct involving NRC regulated activities. Ilowever, the Petitioner has not provided any specific information to support a charge of deliberate mis:onduct by any individual, As noted earlier in this response, ibote hiineral did provide incorrect information more than 20 years ago to support NRC's retiring of the license, it appears that they mistakenly assumed that the uranium and thorium in the slag were no longer considered source material and thus did not require a license. Here is no evidence of deliberate misconduct by Foote hiineral Company in this matter.

As a separate matter, Petitioner's assertion; of wrongdoing by NRC employ.

ecs (i.e., collusion with Ohio agencies regarding jurisdiction of offsite slag so as to avoid " legal problems"), have been referred to the NRC Office of the Inspecter Gentral (010),

(3) The NRC should terminate the devek pnrnt of the emironmental impact statement (EISI for the Shieldalloy site.

De Petitioner requesti d,at the cts. tent SIS being developed for this facility be terminated, as federal law, acuording to the Wtitioner, does not allow NRC to evaluate waste streams diat fall outside clits jurisdictional control. According to 156

l the Petitioner, the EIS is evaluating both

licensed" and " unlicensed" slag, w hich I exceeds NRC's authority. The Petitioner also argues that NRC consideration of i

" unlicensed" rnaterials will result in inadequate protection of the public from l

" licensed" materials.

De Petitioner is correc' that NRC is evaluating all of the onsite slag as part of the EIS, including nonradioactive slag containing metals such as vanadium. De Petitioner is in error, however, in stating that federal law does not allow NRC to evaluate theme wastes. The requirements to assess environmental impacts of major federal actions affecting the environment under NEPA are quite broad and extend beyond NRC's usual licensing authority under the AEA. Environmental impacts that are to be assest.ed under NEPA include impacts on local schools, traffie, and noise that result from different alternatives for remediating the site. The environmental impacts that are required to be evaluated also include those resulting from onsite chemicals (including vanadium and other metals contained in the slag and their possible migration into groundwater), in addition to radioactive materials. NRC's draft EIS issued for public comment (NUREG.

1543, July 1996) contains a comprehensive discussion of all environmental .

impacts, not just those from radioactive materials. Thus, contrary to the Petitioner's assertion, federal law in this case requires NRC to consider a broad ranFe of environmental impacts and, therefore, all of the slag at the facility. l Whether the slag is " licensed" or " unlicensed" is not a factor in determining the l scope of the EIS.

De Petitioner also states, as a reason for this request, that the radiation doses to members of the public would be well above 600 millitem/ year (mrem /yr) from licensed materials, and higher than those cuculated for " licensed" and

" unlicensed" waste when included together. The Petitioner is incorrect. In the draft EIS (NUREG 1543 July 1996), NRC has modeled the slag piles as they currently exist, and used conservative modeling assumptions to help ;

ensure that actual releases, if any, will be bounded by the EIS calculations.

These calculations of radiation doses to members of the public are based on the actual slag piles, and are not affected by any arbitrary divisions of the material into, for example, " licensed" and " unlicensed" slag. Each pile has 1 certain concentrations of radionuclides and chemicals, and each is modeled in l the EIS Releases from both piles are used to evaluate potential impacts on human health, in the draft EIS analysis, NRC has calculated a maximum dose of 6 mrem /yr for an offsite individual. The annual cancer mortality risk for this dose is approximately 3 x 104 NRC has also calculated a radiation dose of 42 mrem /yr to an onsite residential farmer, when both piles are capped with clay. The annual cancer mortality risk for this dose is approximately 2 x 105,

(~Ihe residential farmer scenario assumes failure of institutional controls, such as fences and deed restrictions. Ren, the hypothetical farmer that establishes a i residence and farm on site is assumed to drink water obtained from a well that 457

is drilled adjacent to the piles, and eat crops grown on site that are irrigated with groundwater from the well.)

in summary, as explained above, NRC is approprit.tely evaluating the envi-tonmental impacts of all slag at the Shieldalloy site. 'Iherefore, this request is denied.

(4) in place of Itw EIS, Stueldalloy and CI'M should te jointly ordered to submn a decomnussionmg plan for hcenwd matenal that includes only a plan to tenwd ste hcensed matenal, including gradmg and evaluauon of all vanous assorted opuons.

One opuon considered should be offsne dnposal at a hcensed disposal facihty.

As noted above, Shieldalloy, as the NRC Licensee, is responsible for ra-dialogical decommissioning of the site. Therefore, this request is denied for the same reasons as the request to require that CFM obtain an NRC license.

Furthermore, as noted above, the option of offsite disposal of the slag is being considered, albeit pursuant to the EIS and not the Petitioner's suggested joint decommissioning plan Finally, the Staff has previously noted, in response to the first request, that the State of Ohio has made Cl%1 responsible for certain aspects of remediation by means of a proposed Consent Order.

(5) The Ohio Ermronnwnial protecuon Agency and Otuo Depannwnt of fleahh should esaluate all unheensed slag found at sie Shieldalloy site.

This request can only be implemented by the State of Ohio and is, therefore, l not properly addressed here. 'The Petitioner did contact the Ohio Department of Ilealth regarding his request. As noted earlier, however, the State of Ohio has entered into a proposed Consent Order with CI%1 and Shieldalloy, and has been conducting its own review of all of the materials at the site in accordance with CERCLA.

IV. CONCLUSION For the reasons discussed above, the Petitioner's requests for action pursuant to section 2.206 are denied, A coly of this Decision wdl be placed in the Commission's Public Document Room at 2120 L Street, NW, Washington, DC 20555, and at the local public document room for the Shieldalloy facility in the Guernsey County Public Library. De Director's Decision will 31so be made available on the NRO Electronic Bulletin Board at 1 800-952-9676. A copy of this Decision will be tiied with the Secretary for the Commission's review, in accordance with rection 2.206 458

As provided by this regulation, the Decision will constitute the final action 4

of the Commission 25 days after issuance, unless the Commission on its own motion institutes a review of the Decision within that time.

FOR Tile NUCLEAR REGULATORY COMMISSION ;

2 Carl J. Paperiello. Director l Office of Nuclear Material Safety

and Safeguards j 4

Dated at Rockville, Maryland, this 6th day of June 1997.

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Cite as 45 NRC 460 (1997) DD 9713 i J

l UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION l

OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS Carl J. Paperiello, Director in the Matter of Docket No. 03016055 ADVANCED MEDICAL SYSTEMS,INC.

(Cleveland, Ohio) June 13,1997

%c Director of the Office of Nuclear Material Safety and Safeguards (NMSS) denies a petition filed with the Nuclear Regulatory Commission (NRC or Commission) by letter dated March 3,1993, by William B. Schatz Esq.. on behalf of the Northeast Ohio Regional Sewer District (District or Petitioner),

requesting that actions be taken regarding Advanced Medical Systems, Inc. (the Licensee), ne petition was partially Franted, as explained in the Decision. De Director denies the remaining requests of the petition on the basis of analysis of the technical issues and the Commission's authority to grant the requested ,

relief, set forth in the Decision, which analysis showed that the Commission did not have such authority and that no technical basis warranted Franting the '

petition.

JURISDICTION No statute or reFulation grants the Commission authority to require a licensee to pay, in effect, compensatory damages to private individuals. Tankee Atomic Electric Co. (Yankee Nuclear Power Station), CLI 96-7, 43 NRC 235, 269

.(1996). A court of competent jurisdiction, and not the NRC, is the proper

- forum for such an individual to seek compensatory damages from a licensee.

TECHNICAL ISSUE DISCUSSED

l. De following technical issue is discussed: Contamination of sewer line.

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i DIRECTOR'S DECISION UNDER 10 C.F.R. 6 2.206 j I, INTHODUCTION :

By letter dated March 3,1993, addressed to Mr. James Taylor, former Exect live Director for Operations. U.S. Nuclear Regulatory Commission (NRC),

4 William D. Schatt. Esq., on behalf of the. Northeast Ohio Regional Sewer 3 District (District), requested that NRC take action .with respect to Advanced l Medical Systems, Inc. (AMS), of Cleveland, Ohio, an NRC Licensee, ne !

District requested, pursuant to 10 C.F.R.12.206, that NRC: (1) modify AMS _

i License No. 3419089-01 to require that AMS assume all costs resulting from the offsite release of cobalt-60 that has been deposited at the District's Southerly ,

Wastewater Treatment Center (S%TC); and (2) order AMS to decontaminate ;

the sewer connecting its London Road facility with the public sewer at London l

Road, and continue downstream with such decontamination to the extent that i sampling indicates is necessary,

! he District alleges the following bases for its request: (1) cobalt 60 has

! been discovered in the ash piles resulting from the incineration of sewage sludge at the District's SWTC; (2) AMS is the only Licensee in the District's service j area authorized to process cobalt.60 in a loose metallic form consistent with -

J the form present in the ash; (3) AMS is the only entity (except for the former 1

owner of the London Road facility) that has reported discharging cobalt 60 to the ,

sanitary sewer system leading to the SWTC;(4) NRC documents present ample i evidence of cobalt 60 contamination at the London Road facility, including numerous drains inside the building; ($) there are excessive exposure rates in the j- sewer connecting the building to the public sewer system;(6) this sewer line has

. been classified as a restricted atea, which effectively denies the District access '

to the manhole for sampling industrial discharfes; and (7) the AMS lamdon j Road facility is the source of the cobalt 60 at the S%"10, i i By letter dated April 2,1993, the Director, Office of Nuclear Material Safety and Safeguards, NRC, formally acknowledged receipt of the petition and i- informed the District that its request was being treated pursuant to section 2.206 of the Commission's regulations. A notice of the receipt of the lectition was i published in the federal Register on Tuesday, April 13, 1993 (58 Fed. Reg.

- 19,282). Staff sent a copy of the letter dated April 2,1993, with a copy of the ;

petition, to AMS.

By letters dated September 13,1994, October 13,1994, and April 29,19%,

the District filed supp'ements to its March 3,1993 petition. De District's September 1994 supplement requested that NRC commence enforcement actions 4 against AMS for violations of 10 C.F.R. Il 20.401(c)(3) and 20.303(a), based on

. assertions that the disposal records maintained by AMS are grossly inaccurate, in 461

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violation of section 20.401(c)(3), and that AhtS discharged material to the sewer that was not readily soluble in or dispersible in water, in violation of section 20.303(at in addition, the Septemt er 1994 supplement requested that th: hiarch 3,1993 petition be granted immediately insofar as it requested that Ah15 be held responsible for all costs arising from contamination of the District's treatment plant and that Ah15 be required to decontaminate the sewer downstream from the London Road facility, in its October 1994 supplement, the District requested that NRC commence an enforcement action against AhtS for violation of 10 C.F.R. 6 20.2003, based on the assertion that AhtS had recently discharged cobalt 60 to the sewer that was not soluble or readily dispersible biological material, in violation of that provision. In its April 1996 supplement, the District requested NRC action on a license requiring AhtS to safely and reasonably decontaminate the London Road interceptor (the sewer), or, if NRC's position is that such action has already been ordered, NRC action requiring Ah15 to actually complete the decontamination.

Since receipt of the hiarch 3,1993 petition NRC has amended AhtS' license such that one of the District's requests has already been partially Franted, as set forth below. I have completed my evaluation of the remaining matters raised by the D. strict and have determined that, for the reasons ttated below, the other requests in the petition and its supplements should be denied.

II. IIACKGROUND NRC issued License No. 34 19089-01 to AhiS on November 2,1979.

Picker Corporation had previously owned and operated the licensed operation, facilities, and equipment since 1959. liom 1979 to mid 1991, the AhtS license authorized the possession of 150.000 curies (5550 terabecquerels) of cobalt 60 in solid fonn for the purpose of manufacturing sealed sources for distribution to authorized recipients for use in teletherapy units (used at medical facilities for treatment of medical conditions). He AhiS license currently limits possession to 150.000 curies (5550 terabecquerels) as solid metal and 135,000 curies (4995 terabecquerels) in sealed sources, for use in installing and servicing teletherapy units, and training; the current license does not authorite manufacture of sealed sources for distribution. he license also authorires possession of 40,000 curies (1480 terabecquereh) of cesium 137 in scaled sources, and 4(M0 kilograms of plated depleted uranium shieldmg, incident to teletherapy and industrial radiography installation, maintenance, and service. He facility that houses the licensed material is located on London Road in Cleveland, Ohio.

. The District is responsible for operating three wastewater treatment facilities in and around the Cleveland, Ohio metropolitan area. He District's SWir has been operating since 1927 to remove grit and debris from wastewater that the 462

l District services. His process involves incineration of sludge, transport of the residual ash in a slurry to settlement and evaporation ponds, and eventual transfer of the dried ash to landfills. De S%T also incinerates sludge generated at other facilities, including the District's Easterly Plant, which services the area where AMS is located.

In Aril 1991, NRC identified cobalt 60 at the S%"IC in ash piles coincidental to an aerial radiation survey of an unrelated site. In September 1991 and March 1992, at the request of NRC, Oak Ridge Institute for Science and Education (ORISE) performed surveys at the SWTC to determine the extent of the cobalt-60 contamination at the facility, he results of the ORISH surveys are reported

-in " Radiological Characteritation Survey for Selected Outdoor Areas Northeast Ohio Regional Sewer District, Southerly Wastewater plant, Cleveland, Ohio,"

Final Report, August 1992 (hereafter referred to as *ORISE report"). ne ORISE report indicated that there were elevated direct radiation readings that were caused by cobalt 60 contamination, with elevated concentrations in soil and sediment samples. Based on this ORISE report and information collected and examined by NRC Staff, NRC estimated that a total activity of 414 millicuries (15.3 gigabecquerels) of cobalt 60 existed at the 5%7C in 1992.

Since the District needs to transfer the dried ash frorn the evaporation ponds to

- continue operations, NRC approved the site remediation strategy for ash removal.

and had ORISE perform an independent survey to evaluate the radiological status of the remediated area. The District performed a radiological characterization of the facility to better determine the amount of cobs.lt-60 that is actually present on the S%TC site; the District's consultant estimated the quantity of cobalt 60 in the North Fill Area, as of 1993, to be about 443 millicuries (16.4 gigabecquerels).

As discussed below, NRC has evaluated the District's concerns and bases for its requests for NRC action, Although NRC has amended AMS' license to require remediation of the interceptor Sewer line operated by the District in the vicinity of the connecting line from the AMS facility, which partially grants one of the District's requests, the District's remaining requests are denied for the reasons discussed below, -

Ill.- DISCUSSION A. - Timing and Source of Contamination identified at the SWTC in 1991, cobalt 60 was discovered in the North Fill Area.* De Staff's review of the history of the SW"!E revealed that, after renovation of the incinerators

'signi6 cant newls of cobah40 sequmag reme&auon wm ducovmd in the North Fill Area. in the ensims in-Place Ash secuan of the south nli Area. and in the northern secuan of the south Fill Area (kly the Ncwth Fill Area contanunanon can be dated with any degree of certamty, although AMs secords inecate thai 1989 was the last year AMs encharged cobah40 Arestly into the sanitary neuer syssent 463

between 1975 and 1978, the incinerators came back on line in Nosember 1978, and the cunent ponds were put into use for the first time. %e ponds were then cleaned for the first time from December 1982 to h1 arch 1983. The District removed the ash from the evaporation ponds and placed it in the Nonh Fill Area, which was then landscaped. %is was the only time the Nonh Fill Area was used for ash disposal. Accordingly, the cobalt 60 entered the District's system and was deposited at 5%TC betwecn late 1978 (when the ponds were first used) and December 1982 (when the ponds were first cleaned and the ash placed in the North Fill Area). See Memorandarn for Carl J. Paperiello, former Deputy RcFi onal Administrator, NRC Regiou !!!, from Loren J. Hueter, Radiation Specialist, Division of Nuclear hiaterial Safety, NRC Region 111, on the subject of " Report on Trip to General Chemical Corporation (Non licensee),

$000 Warner Road, Cleveland, Ohio, and to Northeast Ohio Regional Sewer District 6000 Canal Road, Cleveland, Ohio" (Docket No. 030-18276; License No. 3417726-02) dated June 13, 1991, The Staff's conclusion as to when cobalt-60 contamination entered the sanitary newer system is supported by the District's letter, dated September 13,1994, w hich stated that the earliest possible date that the cobalt 60 could have been discharged into the sanitary sewer was not more than a week or two before the opening of lagoons in October 1978.

In an attempt to determine all possible contributors of cobalt 60 contamina.

tion to the SWTC, NRC conducted a file review of all licenses issued since 1975, active and terminated, for activities at facilities in the rip code areas serviced by the District. NRC contacted existing and previous licensees for additional information. De U.S. Department of Energy was also contacted to determine if any of its operations in the Cleveland metropolitan area could have contributed to the cobalt 60 contamination at the SWTC. Although other cobalt-60 users were found in the NRC's file search, it was concluded that no facihty, other than AhtS' facility at 1020 London Road, Cleveland, Ohio, was authorir.ed to possess the quantities of unsealed cobalt 60 that could have contributed to the levels of cohalt 60 contamination found at the SWTC, hiemorandum from Roy Caniano, Chief, Materials inspection Branch, Division of Radiological Safety &

Safeguards (DRSS), Region 111, to William L. Axelson, Director, DRSS, dated November 7.1994 (hereafter "Caniano Memo").

Given the information as to the timing of the disposals into the sewer system that caused the cobalt 60 contamination at the S%TC, the Staff included Picker, which previously used the facility under NRC license, in its review and inspection, although the District did not seek action against Picker. Current and former Picker employees, as noted in Inspection Report No,03016055/93003 (Section 3.C), issued November 7,1994, stated that liquid radioactive waste was toutinely discharged from the London Road facility. hey stated, however, that the 1 curie (37 gigabecquerels) per year annual gross quantity disposal limit (10 C.F.R.120.303) was never exceeded during their respective tenures.

MA

!!ased on the information gathered during the inspection, it is highly likely that Picker Corporation discharged cobalt-60 into the sanitary sewerage sptem every year that it operated the London Road facihty, including the 1978 and 1979 time period of interest. As for AMS, its records indicate that a total of 209 millieuries (7.73 gigabecquerels) of unsealed cobalt 60 was disposed of into the sanitary sewerage system during the period 1980 to 1989. Caniano Memo at 3. AMS records indicate that 1989 was the last year that cobalt.60 was discharged directly into the sanitary sewerage sptem. NRC Inspection Report No. 03016055/93003 (DRSS) at 7, issued November 7,1994. AMS records also specifically list releases during the 1980 1982 time frame, inspection Report No. 030-16055/93002 at 17, issued August 2,1993. The information gathered by the Staff indicates, therefore, that cobalt 60 was likely released from the London Road facility during the 1979 1982 period of interest by both Picker ;

and AMS.

AMS has recorded discharging cobalt-60 to the sanitary sewer system that eventually leads to S%TC, as described above. AMS records indicate, howeser, that it had been discharging cobalt 60 in accordance with the quantities and concentrations authorized by the then app:lcable regulations and license. NRC's inspection and review of records have not revealed any documentation at AMS or other evidence that would indicate discharges in excess of authorized limits.

11 Request for NRC Action to Require AMS to Assume the Cost Resulting from Offsite Release of Cobalt 60 he Staff has carefully considered the action the District has requested and the bases stated by the District for its request, in addition, the Staff has evaluated the results of its inspections and all available information related to the District's requests. None of the available information, individaally or taken together, demonstrates that AMS violated NRC regulatory limits or other requirements related to the discharge of cobalt 60 into the sanitary sewer system.

In a proceedmg involving the decommissioning of the Yankee Nuclear Power Station near Rowe, Massachusetts, the Commission stated that it had no authority to grant an intersenor's request for compensation sitnitar to the District's.

l'anAce Atomic Eicctric Co. (Yankee Nuclear Power Station), CL196-7,43 NRC 235 (1996). In the l'anAce proceeding, the licensee had initiated substantial decommissioninF of its facility through a " Component Removal Project"(CRP) under a new Commission policy interpreting the decommissioning rule (10 C.F.R. l$0.82) and had removed and disposed of many nadioactive components through the CRP. De intervenors succeeded in challenFi ng the Commission policy, which had allowed the licensee to initiate the CRP without an opportunity for a hearing. CAN v. NRC, 59 F.3d 294 (1st Cir.1995). As relief for the failure to offer an opportunity for a hearing, and based on their assertion that 465

the CRP had caused workers and the public to receive doses far above those as low as ressonably achievable, the intervenors requested the Commission to acquire the licensee to es'ablish a fund for the treatment of cancers caused by the doses resulting from the CRP. limArc, CLI.96 7. 43 NRC at 268. In tejecting the intervenors' arruments, the Commission held that "no statute or regulation grants the Comtnission authority to require the Licensee to pay (in effect) compensatory damages to private individuals." 1d. at 269, "Ihe District's request for compensation from AMS for costs resulting from offsite releases of cobalt 60 fiom the London Road facility is not mateiially different from the l'antee intervenors' request for compensation. No statute authorites the NRC to require any licensee to pay suct sompensatory damages, especially in a case in which 'he releases that resulted in the third party's damages were within applicable NRC limits. j

'Ihe District, in addition to filing its petition with NRC, instituted a court l action against AMS and other defendants for tort remedies, including property damaFe and remediation costs, resulting frorn the discharge of cobalt 60 into the District's system. The action, which was pending before the United States District Court for the Northern District of Ohio, Eastern Division (Case No.1:94 ,

f CV 2555), has been rettled. Letter dated January 2,1997, from LX English, Esq., Northeast Ohio Regional Sewer District, to J. Madera, Division of Nuclear Material Saferuards NRC. A court of competent jerisdiction, and not NRC, is the proper forum for the District to seek compensatory damages from AMS.

Accordingly, the District's request for NRC action to require AMS to assume the costs resulting from the release of cobalt.60 is denied.

C. Request to Require AMS to Decontaminate the Sewer Connecting its London Road Facility with the Public Sewer at lamdon Road and Continue Dowmtream to the Exteti AMS/NRC Sampling Indicates is Necessary By letter dated April 29,1996, the District supplemented its oriF inal petition with a request that AMS be required to " safely and reasonably" decontaminate ,

the London Road interceptor, in addition, the District requested that NRC take action to have AMS complete the decontamination of the interceptor if NRC believed that it had already ordered AMS to take action to decontaminate the interceptor. The indicated sewer connection that was identified as having excessive exposure rates is on AMS preperty. NRC did issue a Notice of Violation (NOV) for AMS' violation of 10 C.F.R. $ 20.105, in that the exposure rates in the accessible sewer line on the AMS facility were excessive for an unrestricted area. NOV issued to AMS, License No. 3419089 01, dated May 5,1988, resulting from a special ssfety inspection conducted on April 13,1988 (NRC Inspection Report No. 16055/88001 (DRSS)). However, the 466

manhole controlling access to the sewer connection was designated a restricted area; the sewer cover on the AMS property was secured with a lock and bar; and the sewer connection area was panially decontaminated, reducing the contamination and exposure rate levels. Letter from T.J. liebert, Chairman, Radioisotope Committee AMS, to R.E. Burgin, Senior Radiation Specialist, NRC Region 111, dated May 23,1988. "Ihese facts were confirmed by Oak Ridge Associated Universitics, contracted by NRC to perform a radiological survey to determine the then-cunent conditions at the AMS facility. Srr Oak Ridge As ocirded Universities Report, " Radiation Survey of the Advanced Medical Systems, Inc., London Road Facility, Cleveland, Ohio," Final Report, at 20 (April 1989). The exposure rates are no longer considereu execssive as a result of the decontamination performed by AMS and the designation of the manhole as a restricted area. Moreover, in 1995, AMS permanently scaled the lateral from the old manhole to the sewer line. AMS also remosed most of the oriFinal foundation underdrain system and replaced it with a new, clean system. AMS

!s currently required to test the groundwater pumped from the new foundation underdrain system, to ensure compliance with section 20.2003.

'Ihe NRC has taken action by issuing Amendment No. 32 to AMS' license, dated March 17, 1995, in which the NRC, through Condition 19.F. required AMS to remediate the London Road interceptor in the vicinity of the abandoned l

lateral, as desenbed in an AMS letter proposing action to temediate contaminated l piping. See " Action Plan for the L(mdon Road Facility" at 2 (Jan. 27,1996).

License Condition 19 required that remediation of the interceptor be completed within 90 days (i.e., by June 15, 1995). In Amendment No. 35 to AMS' license, dated June 16, 1995, NRC required AMS to mitiate remediation activities no later than July 8,1995, and to notify NRC no later than July 14,1995, to confirm initiation of the remediation of the interceptor. Amendment No. 35, however, deleted the June 15,1995 date for completion of remediation of the interceptor imposed by Amendment No. 32.

By a letter dated July 12,1995, AMS informed NRC that it would not stan the remediation of the interceptor until July 29, 1995, and did not provide an estimated completion date for the remediation, as AMS funhet infonned NRC that it needed the District's approval to access the interceptor. Letter frorn R.

Meschter, Radiation Safety Officer (RSO), AMS, to J. Caldwell, NRC, dated July 12,1995. By a letter dated July 19,1995, AMS informed NRC that, for the same reasons F i ven in the July 12,1995 letter, h would not initiate remediation until August II,1995. Letter from R. Mr.chter, RSO, AMS, to J Caldwell, NRC, dated July 19, 1995. At that ume, AMS and the District still had not agreed on arrangements for entry and evaluation of the interceptor, in a letter dated January 2,1997, from L.K. English, Esq., Northeast Ohio Regional Sewer District, to J. Madera, NRC, the District forwarded a copy of a settlement agreement between the District and AMS regarding their coun 467

litigation. He seitlement indicates that AhtS agreed, inter nIia, to pay the District a fixed sum, and the District agreed to allow reconnection of the AhtS facility to the London Road interceptor after AhtS' taking certain actions pertaining to conditions of the facility, and to design and construction of the connection. %e part of the agreement concerning reconnection provides an alternative to use the present manhole located in London Road, prmided that the plans include decontamination of the interceptor, at AhtS' cost, before such use. He aFreement specifies conditions and procedures under which AhiS may plan to use the present manhole in the interceptor, in a meetinF with NRC and AhtS on libruary 10,1997, AhtS indicated that it was its intention to reconnect.

Official Transcript of Proceedings: "Public hiceting with Advanced hiedical Systems, Inc.," at 50 51 (itb.10,1997). Ah15 stated that it will probably take from 9 months to a year and a half for reconnection to actually happen.

Id. at $1. In summary, insofar as Amendments Na, 32 and 35 require AhtS to remediate the sewer connecting its London Road facility with the public sewer, this request of the District has been partially granted. Although access to the interceptor is now controlled, License Condition 19.F requires AhtS to remediate the interceptor. %e Staff intends to pursue this matter in the near future, it is the Staff's intent that the access concerns be resolved promptly, so that remediation may begin and be completed as soon as practical.

D. Other lasues Raised in Supplements to l'etition fly letters dated September 13. 1994, and October 13, 1994, the District supplemented its original petition with a request that NRC commence an appropriate enforcement action against AhtS for the snaintenance of grossly inaccurate records of disposal of radioactive material from 1978 to 1993, in violation of section 20.401(b)(3) (in effect through December 31,1993). He District also asserted that AhtS had disposed of cobalt 40 that was not "readily soluble or dispersible in water," in violation of section 20.303 (in elfect through December 31, 1993), and had more recently discharged cobalt 60 which was not "icadily soluble or dispersible biological material," in violation of section 20.2003 (in effect on January 1,1994, and thereafter).

He Staff has conducted numerous recent inspections at the London Road facility to address the District's concerns over cobalt 40 discharFes into the sanitary sewerage system. On h1 arch 15,1995, NRC issued a Notice of Violation to Ah15 for failures to: (1) evaluate the quantity of cobalt 40 released to the sewer system resulting from facility floods and certain decontamination activities; and (2) remove nonsuspendible solids by the use of a cloth filter, as required by AMS' license conditions. He background relating to unmonitored releases resulting from facility floods and certain decontamination activities is set forth below, 468 l

I

l i

ne information as to when the unmonitored releases occuned came imm cunent and former hcLer and AhtS employees and identified several occasions in the late 1960s and the mid- to late 1980s when the basement was flooded, tesulting in backflow into the sewer system. He available infonnation indicated that not all of these occurrences were evaluated to identify the amount of radioac.

tivity ti.at may have been released. Inspection Report No. 030-16055/93003, at 1619. Ilased on the extensive infonnation prosided by the interviewees, the Staf f concluded that it was unlikely that the cumulative total quantity of cobalt 40 released during these unmonitored releases exceeded a few hundred millicuries.

id.

l As to the filtering of the wastewater pumped from holding tanks in the Waste lloid Up Tank room, the information gathered from the interviewees strongly indicated that the filter was not always in place from the mid 1970s through the mid 1980s, thus raising the potential for cobalt 40 pellets to have been discharged through this route into the sewer system. Id. at 14.

ne NRC has already taken enforcement action for the failures to: (1) evaluate and report cer ain releases into the sewer system as a result of facility floods or decontamination activities that likely included cobalt 40; and (2) ensure '

that wastewater in the holdup tanks was passed through filters that should have captured any cobalt 40 pellets before the release of the water to the sewer system. %c Staff does not believe that further enforcement action for the matters identified in the Septernber 1994 supplement is warranted.

Regarding the October 1994 supplement's request for enforcement action l for violation of section 20.2003, the Staff has not found evidence, based on NRC interviews and review of records, that AhtS intentionally disposed of cobalt 40 into the facility's drains leading to the District's sanitary sewerage system since hiny 1989. He AhlS records contain no discharge log entries after this date. Furthermore, Ah1S has not Fenerated liquid radioactive waste from manufacturing operations in several years, and has no plans to do so in the future, because of termination of source manufacturing operations. See Inspection Report No. 030-16055/93002. Ilowever, bc.th the District and the Staff performed sampling (post January 1,1994, the effective date of revision of 10 C.F.R. Part 20) that identified cobalt 40 at the point of discharge of the sanitary sewerage piping from the London Road facility into the District's sewer line. See the District's supplement to its petition, dated October 13,1994, and l Inspection Report No. 030-16055/94003, issued on December 6,1994. He presence of the cobalt 40 appears to be a result of plate-out of cobalt 40 onto l the walls of the piping leading from the London Road facility. ne Staff had characterized the results of its sampling as indicating an apparent violation of section 20.2003. /d.

De sampling perfonned by the District and subseque;t sampling performed by the Staff in early 1995 indicated that some or all the cobalt 60 detected might 469 l

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be " soluble." as that term is defined in NRC Information Notice No. 94-07, dated January 28,1994. The uncertainty as to the solubility of the cobalt.60 prompted ;

the Staff to begin preparations for a solubility analysis of the sarnple taken on August 17. 1994. In accordance with Region ill policy, those sutnples had been transfened back to the District, on whose property the samples had been taken.11ecause of further analyses the District Pad performed on the sampler.,

the samples no longer existed in their original form; therefore, further solubility analyses could not be performed. Ibrther representative samples of the water at this point in the waste stream could not be taken because of the District's plugging of the pipe, in view of the inability of the Staff to determine that the cobalt.60 in the sampled water was, in fact, insoluble, there was an insufficient basis to cite Ah15 for a violation of section 20.2003. Furthermore, there is not now a significant potential for discharge of cobalt.60 from the London Road facility to the District's system because: (1) old piping connecting the facility to the District's lines has been plugged;(2) the District has not permitted AhtS to connect new clean piping installed by AhtS to the District's lines; and (3)

AhtS collects and treats all water used on the site and holds it in tanks before it is determined not to contain insoluble cobalt.60.

"the Staff believes that the vast majority of cobalt 60 inventory and activity discharged into the District's sanitary sewerage system was dispersible, it can be expected that a small amount of readily dispersible material would plate out onto the sewer system pipes over the long history of cobalt.60 discharges by Picker and AhtS. Staff concludes that the fact that a r. mall amount of cobalt-60 built up ove. time in sewer pipes leading from the Ahi$ facility, by itself, does not support the District's assertion that a discharge in violation of section 20.303 or 20.2003 occurred.

IV. CONCLUSION Ibr the reasons discussed above, no basis exists for taking any action, in additien to the action described above, in response to the requests in the petition and its supplements. Accordingly, no further action pursuant to section 2.206 is

$>cing taken in this matter.

As provided by 10 C.F.R. 5 2.206(c), a copy of this Decision will be filed with the Secretary of the Comrnission for the Commission's review, The De-cision will become the final action of the Commission twenty five (25) days after i

470

issuance unless the Commission on its own motion institutes review of the Decision withir. that time.

1OR 111E NUCLEAR REGULATORY COhth11SSION Carl J. Paperiello Director Office of Nuclear hinterial Safety and Safeguards Dated at Rockville, htaryland, this 13th day of June 1997.

471

Cne at 45 NRC 472 (1997) DD 9714 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION Samuel J. Collins, Director in the Matter of - Docket No. 50 219 GENERAL PUSLIC UTILITIES NUCLEAR CORPORATION (Oyster Creek Nuclear Generating Station) June 16,1997 By a petition dated April 1.1997. Berkeley Township Environmental Com-mission (Petitioners) requested that the NRC direct Oyster Creek Nuclear Gener.

ating Station (OCNGS o- Licensee) to shut down its operations during a planned transfer of fuel from wet to dry storage. The request was considered as a petition submitted pursuant to 10 C.F.R. 6 2.20&

In a Director's Decision issued on June 16, 1997, the Director of Nuclear Reactor Regulation dismissed Petitioners' request as premature. The Director concluded that bccause OCNGS would first have to submit a request for a license aruendment to perform the action in question, which it had not yet done and on which the IYitioners would heve an opportunity to comment, there was no basis for the Commission to take the requested action at this time.

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

1. INTRODUCTION By a petition subn'itted i pursuant to 10 C.F.R. 6 2.2% and dated April 1, 1997 (petition). Berkeley Township Environmental Commission (Petitioners) requested that the U.S. Nuclear Regulatory Commission (NRC) take action with regard to Oyster Creek Nuclear Generating Station (OCNGS) operated by GPU Nuclear Corporation (GPU or Licensee). The Petitioners requested that the NRC 472

?

direct the Licensee to shut down OCNGS during an upcoming planned transfer of fuel from wet to dry storage.

lhe Petitioners bated their request on the following assertions: (1) the load transfer path for the 100-ton fuel transfer casks passes over the reactor's containtnent mechanism and other safety related equipmenu (2) NRC Ilulletin 96-02, dated April 11, 1996, states that a dropped cask could damage both isolation condensers and the torus, creating the possibihty of an unisolable leak which in industry jargon describes a situation perilously close to a nuclear meltdown;(3) the operating record of GPU demonstrates it is capable of human

. enor, including dropping heavy loads; (4) 11erkeley Township could not be successfully evacuated in the event of a serious nuclear accident at OCNGS; and (5) the safer, simpler alternative of turning off the reactor while lifting 10(hton loads over the containment can be easily implemented.

For the reasons stated below, I have dismissed the Petitionere request as prernature.

II. DISCUSSION The Petitioners have requested that the NRC take action against the Licensee on a matter involving the potential transfer of spent fuel during plant operation.

Ilowever, this is an activity for which the Licensee has not yet requer.ted authoritation from the Commission. At a public meeting on libruary 29.1996, the NRC informed GPU that it would have to obtain a license amendment to move fuel from wet to dry storage, using the facility's existing crane, while the reactor is operating at power. The Staff had reviewed the Licensee's safety evaluation of its crane, including the crane upgrades, and concluded that all safety concerns had been addressed and resolved and that the planned movement of spent fuel to the dry storage facility during plant operation would be safe and

- in accordance with all license requirements, llowever, the NRC also determined that because the possibility of an unreviewed safety question existed before GPU made modifications to uPFrade its reactor building crane, GPU would have to submit a request for a license amendment for the proposed cask movement. If GPU submits such an amendment request to the NRC, pursuant to 10 C.F.R. 6 50.91.' it will be published in the Federal Register for public corr. ment, and an opportunity for a public hearing will be provided. *lhe Petitioners and other interested members of the public then would have the opportunity to express their concerns about the amendment. As noted above, the Licensee cannot cuan 50 91 spectries the Conumssion psuedures to be followed when n receives an applicauon requesting an anunchnrnt to an operaung hcense. including pacedures for consulung the state an wtuch the recahty as kwated and pnxedures rur hourying the pubbt or the license anendnent and the opportunsty rur a hearr.g 473 M

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transfer the fuel while operating with its current crane configuration without being issued a license amendment.8 111. CONCLUSION l

'Ihe NRC Staff has reviewed the Petitioners' request that GPU shut down its reactor during its transfer of fuel from wet to dry storage. 'Ihe Licensee does not now have ,* request before the Commission to amend its license to allow such a transfer. As a result, before any Commission action could even be contemplated, the Licensee would have to make such a request purs'iant to NRC regulations, with the aforementioned opportunities for public particiration in the resolution of any such reque t. Ibr this reason, the petition is dismissed as premature.

A copy of this Director's Decision will be filed with the Secretary of the Commission for the Commission to review as stated in 10 C.F.R. 5 2.206tc).

This Decision will become the final action of the Commission 2$ days after issuance, unless the Commission, on its own motion, institutes a review of the Decision within that time.

FOR Tile NUCLLAR REGULATORY COhthilSSION Samuel J. Collins, Director Office of Nuclear Reactor ,

Regulation Dated at Rockville, blaryland, this 16th day of June 19 7. ,

l 3

The I.Jcensee s ceremly considering vertous opuuns far nioving tte spes fuel frorn wet to dry storage, such as requesting a hcense anendnwm based on already corrg&ted upgrades to the fracts buikhng crane, transfernns tlw spem fuel ahee the reacaw 6s shut down, and furttwr upgradang the reactor bulkthg crane la neel the critena fm a single-failure-proof crane te which case an anwndnwns to transfer fuel from wet to dry starage enay not be required. The Comnu-non has not required became anwndarms for facihees handhng heavy ka6 thas ernphy a crane nwenns ow brecahemions and design cniena in NURIG-OSM.

single-fiulure-Pront Danes tw Nuclear Power Plants.* However. NRC technxal staff will evaluate any opuan selected to ensure that all safety concerns are adequately adJressed and docurnemed.

l 474 f

i,

i 3 i i

Cite as 45 NRC 475 l1W7) DD-9715

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS i

.i

l. Malcolm R. Knapp, Acting Director [

in the Matter of WISCONSIN ELECTRIC POWER Docket Nos. 50 266 i COMPANY 50-301

. (Point teach Nuclear Plant, 72 5 2

Units 1 and 2)

CONSUMERS POWER COMPANY Docket Nos. 50 255 i (Psileados Nuclear Plant) ~ 72 7 ENTERGY OPERATIONS, INC. Docket Nos. 50 313 (Arkansas Nuclear One, Units 1 50 364 i

and 2) 72 13 SIERRA NUCLEAR CORPORATION Docket No. 721007 4 June 18,1997 By a petition filed on October 18, 1996, the organitations Don't Waste 1 i Michigan and Lake Michigan lideration requested, pursuant to 10 C.F.R. 1 6 2.206, that the NRC prohibit the loading of Ventilated Storage Casks until . -

an independent, third. party revie of the design has been perfortned to address their concerns and the certificate of compliance, safety analysis report, and safety evaluation report for the casks have been amended to contain operating

. controls and limits to prevent hazardous conditions. The Director of the Office l of Nuclear Material Safety and Safeguards, in the following Decision, denies the Petitioners' request.

j. e 475 i 4

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DIRECTOR'S DECISION UNDER 10 C.F.R. 5 2.206

1. INTRODUCTION !

On October 18, 1996, Don't Waste hiichigan and the Lake Michigan Fed. l eration (Petitioners) filed a petition pursuant to section 2.206 of Title 10 of the Code of federal Regulations (10 C.F.R. 5 2.206) requesting that the U.S.

Nuclear Regulatory Commission take the following action:

Protubit loadmg of Venulated Storage Casks (VSC 24s) until the certificate of comphance (COC), the safety analysis repon (S AR), and the safety evaluation report t$CR) are anended following an independent. durd-party review of the VSC-24 design. to address concerns raised by the Peuuoners' engmcenrig consuhant. Dr Rudolf flausler.

- Die petition has been referred to me pursuant to section 2.206. By letter dated December 10, 996, to Dr. Mary Sinclair and Ms. Eleanor Roemer, on :

behalf of the Petitioners, NRC acknowledged receipt of the petition and provided the NRC Staff's determitation that the petition did not require immediate action by the NRC. Notice of receipt was published in the federal Register on January 13, 1997 (62 Fed. Reg.1783).

On the basis of the NRC Staff's evaluation of the issues and for the reasons given below, I have determined tnat the Petitioners' request should be denied.

II. BACKGROUND On May 28.19%, a hydrogen gas ignition occurred during the welding of

'the shield lid after spent fuel had been loaded into a VSC 24 at the Point Beach Nuclear Plant. The hydrogen was formed by a chemical reaction between a ,

zine based coating (Carbo Zinc. II) and the borated water in the spent fuel pool. On June 3,1996, the NRC issued confirmatory action letters (CALs) to those licensees using or planning to use VSC 24s for dry storage of spent nuclear fuel, i.e., Licensees for Point Beach Nuclear Plant, Palisades Nuclear Generating Plant, and Arkansks Nuclear One (ANO). The CAL issued to the Licensee for s ANO was supplemented on June 21,1996, and the CALs issued to the Licensees for Point Beach and Palisades were supplemented on June 27. 1996. " Die CALs, ns supplemented, documented the Licensees' commitments not to load or unload a VSC-24 without resolution of material compatibility issues identified in a forthcoming generic communication and subsequent NRC confirmation of corrective actions taken by the Licensees. The generic communication was issued on July 5,1996, in the form of NRC Bulletin 96-M, " Chemical, Galvanic, or Other Reactions in Spent Fuel Storage and Transportation Casks" NRC 476 i

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Bulletin 96-04 notified addressees about the potential for adverse chemical, galvanic, or other reactions among the materials of a spent fuel storage or transportation cask, its contents, and the environments the cask may encounter during use. %e actions requested in Bulletin 9644 included reviewing the cask materials for potential adverse reactions, evaluating the short term and lonF-term effects of any identified reactions, and determining the adequacy of cask operating procedures to minimlic the consequences of any identified teactions.

%e NRC Staff has acknowledged that the event demonstrated that the cask vendor's (Sierra Nuclear Corporation) SAR for the VSC 24 and related NRC review, as documented in the NRC Staff's SER, did not adequately address the use of a zine based coating and its reaction with the acidic water in spent fuel pools.

In response to Bulletin 96-04 and to subsequent NRC Staff inquiries, the Licensees for ANO, Point Beach, and Palisades submitted to the NRC evalu-ations of possible material interactions and the effects of such interactions on cask performance and operation. The Licensees also submitted inforr.ation on g the operating controls and limits that were irnplemented to prevent hazardous !

conditions that may result from adverse material interactions. %c operating controls and limits included controls for the environments that the casks en-counter during use, requirements for inspections and environmental sampling, and additional precautions for various cask operations.

De NRC Staff evaluated the responses submitted by the Licensee for ANO.

As documented in the Staff's safety evaluation dated December 3,1996, the Staff determined that the Licensee's submittals prosided the necessary level of confidence that the VSC-24 can be used to safely store spent fuel over the 20-year permd of the certificate. The Staff also determined that the operating controls and limits proposed by the Licensee are acceptable and satisfy regulatory requirements. By a separate letter, also dated December 3,1996, the Staffinformed the Licensee for ANO that its corrective actions had been verified by inspections performed bv the NRC Staff. Shortly thereafter, the Licensee initiated cask loading activities.

De NRC Staff also evaluated the responses submitted by the Licensees for Point Beach and Palisades, As documented in the Staff's safety evaluations dated, respectively, April 8,1997, and June 12, 1997, the Staff determined that Ge Licensees

cvaluations and proposed operating controls and limits are acceptable and satisfy regulatory requirements. However, the CALs placed on Point Beach and Palisades still remain in place until an NRC inspection is performed to verify that the Licensees' corrective actions are properly implemented.

477

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Ill. DISCUSSION i t

The petition requests an NRC order to users of VSC 245 not to lotd additional !

casks until: (1) the COC, SAR, and SER are amended to contain operating !

. controls and limits to prevent hazardous conditions; (2) an independent third. ;

party review team has examined the safety issues raised by the Petitioners; [

(3) the potential impacts of all material aspects of the casks have been fully assessed; (4) there is experimental verification of temperature calculations and

{

heat transfer assessments and other design assumptions; and (5) the safety of the material coatings on components and structures has been justified.

Item 1: Prohibit Leading of VSC.24s Pending Amendment i of Documents !

As noted in the NRC letter to the Petitioners on December 10,1996, the !

Petitioners' request to amend the COC, SAR, and SER is similar to a request -

made by the Citizen's Utility Board (CUB, in a petition dated September 30, ,

1996 'Ihe NRC Staff denied the CUB petition on April 17, 1997, for the reasons that are identical to the reasons stated here in denying the first part of the Petitioners' request.

The circumstances set forth above made clear that, following the event at-Point Beach, the NRC Staff recognized Bat additional evaluation of potential material interactions was warranted for all spent fuel transportation and s'torage . !

casks. In regard to the VSC.24, the event and subsequent NRC inspections made it apparent that actual changes in the operating procedures or the design of the cask would be necessary. CALs were issued to confirm Licensees' commitments to refrain from loading VSC-24s pending completion of the NRC Staff's review of the responses to Bulletin 96 04 and verification of the associated corrective actions. As discussed, the CALs established a process by which the NRC Staff :

could obtain confidence that operating controls and limits to address potential hazardous conditions are developed and implemented by each licensee using VSC-24s.

In particular, the CAL process ensures that Licensees will incorporate the 4 necessary operating controls and limits into revised plant procedures. Moreover,

- under existing NRC requirements, the Licensee must adequately implement l those revised procedures. For this reasor, no changes to the COC or SAR

_ i are needed to ensure that enforceable operating controls and lim 4s are in place

- to address potential hazardous conditions during the loading or unloading of a cask. Further, as previously indicated, the NRC Staff has documented the process, information, and results of its review of the Licensees' responses to Bulletin 96-04 for use of the VSC 24 at ANO, Point Beach, and Palisades in safety evaluations available for public review, j 478 ,

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Although the actions taken as part of the CAL process provide adequate assurance that technical and regulatory compliance iscues raised by the event at Point !!each will be resolved before a licensee loads or unloads a VSC-24, the NRC Staff agrees with the Petitioners that it would be beneficial if the SAR and other licensing-basis documents accurately describe the identified chemical reaction and the associated operating controls and limits. The NRC Staff is currently reviewing a proposed amendment to the SAR and COC for the VSC 24 design and will ensure that the information related to the identified chemical reaction and associated operating controls is adequately addressed in the appropriate licensing basis documents. In addition, the NRC Staff is processing a petition for rulemaking, PRM 72-3, that may lead to additional updating ofindependent spent fuel storage installation SARs and the inclusion of infonnation on operating controls and limits implemented as a result of the event at Point Beach. Ilowever, the previously discussed controls to be implemented by the Licensees and verified by the Staff as part of the CAL process, and the enforceability of those controls under existing NRC requirements, make it unnecessary to require revision of the specific licensing documents cited by the Petitioners as a precondition for resuming cask onerations at the facilities using VSC 24s. Therefore, there would be no regulatory basis for granting the first part of the petition to require amendment of the COC, SAR, or SER before further loading of VSC 24s.

Item 2: Prohibit Loading of VSC 24s Pending Independent, Third Party Review Petitioners request the NRC to prohibit loading of VSC 24s ,mtil the COC, SAR, and SER are amended following an independent, third party review to address concems raised by the Petitioners. The NRC Staff performed a review of the VSC 24 design prior to certification in 1993. As a result of the review, the Staff determined that the design and operation of the cask system is in compliance with 10 C.F.R. Part 72. 'Ihe Staff also concluded, with a high ,

degree of asa.irance, that the VSC-24 will safely store spent fuel over the 20-year period of the certificate. Notwithstanding the Staff's review and determination in 1993, the Petitioners are claiming that a new, independent review is needed before further VSC 24s are loaded.

While the event at Point Beach revealed the need for additional evaluation by licensees and NRC of potential material interactions in the VSC-24 (and other transportation and storage casks). the actions already taken, in the Staff's judgment, provide an adequate response. In particular Bulletin 96-04 was issued to request additional information from licensees using the VSC 24 on material interactions and compatibility in the VSC 24 and on the corrective actions implemented. The NRC Staff then received and reviewed the responses 479

submitted by the Licensees for ANO, Point Beach, and Palisades. The Staff's reviews (as well as the Licensecs') have been exhaustive and were performed by an interdisciplinary team af engineers knowledgeable in materials, corrosion, metallurFy, chemistry, structural engineering, heat transfer, nuclear engineering, and other technical fields needed to perform the reviev.. The results of the Staff's reviews, including the necessary corrective actions, are documented and justified in the Staff's December 3,1996. April 8,1997, and June 12, 1997 safety evaluations. These corrective actions include: cleanliness checks before placing the cask in the spent fuel pool, venting and monitoring of the air space beneath the VSC 24 shield lid during welding or cutting activities, discontinuing welding or cutting should the hydreFen concentration exceed 0.4% by volume (10% of the minimum amount necessvy for a combustible concentration), and sampling the boron concentration in the spent fuel pool and multiassembly scaled basket (MSB) water. While the Staff agreed that the corrective actions were necessary to prevent hazardous conditions during the loading and un%ading of VSC 24s, the information submitted by the Petitioners does not raise any new issues or provide any reason for the Staff to question its conclusion that the VSC 24 will safely store spent fuel over the 20-year period of the certificate.

In reaching this conclusion, the NRC Staff evaluated the specific concerns raised by the Petitioners related to the design of the VSC 24. He Staff believes that these concerns have already been addressed by the recent evaluations submitted in response to Bulletin 96-04, by information submitted to NRC to support the certification of the VEC-24 design in 1993, or by other information submitted in support of NRC review and inspection activities. Each of the Petitioners' specific concerns is addressed below.

(i) The Petitioners claim that the cask design allows for fuel elements

?o be in contact with the zine primer, creating a galvanic couple that will accelerate the corrosion of the zinc. The NRC Staff considered galvanic effects between the Zircaloy fuel rods and the Carbo Zinc 11 coating. The Staff agrees that a galvanic effect would increase the corrosion rate of the zinc, with a corresponding increase in the hydrogen gas generation rate; as the zine in the Carbo Zine 11 coating is polarized to a more active potential. However, in the VSC-24 design, several factors reduce the amount of rinc polarization such that there would not be a significant increase in hydroge,i generation. One factor is the contact resistances between the stainless steel fuel assembly end-fittings and the Zircaloy fuel sods and between the end. fittings and the Carbo Zine 11 paint.

Another factor is the geometry of the VSC-24 and the fuel assemblies. The fuel assemblies are placed in fuel storage sleeves with a clearance of approximately 0.1 inch to 0.5 inch between the sides of the fuel assembly and the sleeves. This clearance and the physical design of the fuel assemblies create shielding between the fuel rod surfaces and the Carbo Zine 11 coating. This shielding effectively reduces the galvanic action between the Zircaloy fuel rods and the Carbo Zinc 480

1 I II coating. ne Zircaloy fuel rods could contact the Carbo Zine coated sleeves if the fuel assembly is not centered in the storage sleeves or if the fuel rods are bowed. Ilowever, the shielding effect and small Carbo Zinc /Zircaloy contact krea would still prevent si Fnificant Falvanic action. Ilydrogen concentration measurements made at Point Beach and the hydrogen monitoring performed at ANO during lo ding of a VSC 24 in December 1996 (NRC Inspection Report Nc.s. 50-313/96 25 and 7213/96-02) Support the conclusion that significant falvanic action between the Zircaloy and zine coating, and hence, increased

! hydroFen generation, is not occurring in the VSC 24. In addition, even if there was an increase in hydrogen generation because of the galvanic action, the j Staff has determined that the controls implemented by the Licensees for ANO and Point Beach would prevent accumulation of a combustible concentration of hydrogen and its ignition. He Staff will also review and verify the adequacy of the controls implemented by the Licensee for Palisades.

(ii) De Petitioners claim that there were numerous discrelancies in the responses to Bulletin 96-N. As noted, the NRC Staff completed its review

, of responses for ANO, Point Beach, and Palisades. He Staff found tnese I responses to be acceptable and found no discrepancies of concern. There were minor differences in the operating controls implemented at the three facilities.

liowever, the Staff reviewed these controls and concluded that all three sets of l controls are adequate to preclude hazardous conditions during cask operation.

(iii) ne Petitioners claim that the epoxy coating applied to the exterior of the multiassembly scaled basket (MSB) could not withstand the temperatures l developed during long-term storare. Technical data on the type of epoxy coating i used on the MSB were provided by the Licensees in their responses to Bulletin 96-N e The data show that the epoxy is temperature-resic: nt up to 350*F.

The SAR for the VSC-24 (which the Staff reviewed and accepted prior to 3

certification in 1993) shows that under nonnal or off normal storage conditions, the temperature of the htSB exterior will not exceed 300'F, for the maximum allowable heat load of 24 kW and, therefore, will act degrade the epoxy.

(iv) The Petitior.ers claim that the low-ten,rrature specification in the COC for moving the VSC 24 htSB was not properly translated to the MSB shell material compositions. Low temperature embrittlement of the MSB shell material was evaluated by the NRC Staff during its safety review before certification of the VSC 24. De composition of the MSB shell material (SA516, Grade 70 carbon steel) is specified in the American Society for Mechanical Engineers, Boiler & Pressure Vessel Code, Section 11. SA-516, " Specification for Pressure Vessel Plates, Carbon Steel, for Moderate and Lower Temperature Service." The impact testing requirements for the MSB material are found in American Society for Testing and Materials Specification A370 (ASTM A370), " Methods and Definitions for Mechanical Testing of Steel Products."

As specified in the COC, SER, and SAR, each MSB shell material must be 481 s

T l

shown, during fabrication, by Charpy test per ASTM A370, to have 15 ft lbs -

l of ebsorbed enerFy at -50 F. Further, movement of the h1SB must occur only at ambient temperatures of 0*F or above to avoid potential brittle fracture of the htSB material.8 The NRC Staff considers the 50 F temperature difference to provide sufficient margin because it places the htSP material at a temperature that is significantly above the temperature where brittle tracture could occur. It should also be noted that the temperature of the htSB shell itself would actually be substantially higher than the ambient temperature (e.g.,20 F for 25 year-old fuel), thus providing an even higher margin. In addition, it is highly unlikely that any hiSB movement activity would take place at temperatures below 0 F.

(v) The Petitioners claim that zine-steel interaction at 800 F to 1000 F and possible steel embrittlement over a 20-year period were not ;onsidered. Zine-steel interaction at the 800*F to 1000 F temperature range was not considered and is not a concern because, as documented in the VSC-24 SAR, temperatures in the htSB will not reach 800 F during storage. hiaximum temperaturec would be 688'F under normal conditions and 708'F under off-normal conditions, for the maximum allowable heat load of 24 kW. Furthermore, over the storage period, the temperatures within the htSB will continue to decrease as the heat load decreases due to the decay of the spent fuel.

, (vi) The Petitioners claim that the effect of molten zine on Zircaloy has not been verified experimentally. 'Ihe NRC Staff evaluated the durability and behavior of the zine coating under the range of storage temperatures. 'ihe presence of molten zinc is not expected under the storage temperatures and conditions; thus the effect of molten zine on Zircaloy is not a concern. However, as documented in the Staff's safety evaluations for ANO (dated December 3, 1996), Point Beach (dated April 8,1997), and Palisades (dated June 12, 1997),

the Staff did evaluate the potential interaction between zine vapor and Zircaloy and the effect of this interaction. Based on the information provided in the responses to Bulletin oC04, the Staff concluded that the potential interaction between zine vapor and inrealoy presented no immediate or long term safety concern for the spent fuel stored in th- VSC-24.

(vii) The Petitioners clai: , L:0A acuum-drying process does not seem to have been experimentally venM ', .uur drying is a well-established, widely used method for removing moisture from spent fuel storage and tramportation casks. The process used for the VSC-24 is a common process, which the NRC Staff evaluated arJ determined to be acceptable duril g the safety review before I

At Pahsades, the lacenset has adnumstrauvely set a reunimum ambient temperature of 10*F for movmg the hrst

, (ma MsBs (CM5B-01 through On to be loaded because the shcH material for these MSBs does not have 15 ft-lbs of absorbed energy at -50*F, Rather these MSDS have 15 ft-lbs of absorbed energy at - 40*F. Thus to retam the SOT tenverature marpm. the Ucensee has restre ' movement of these four MSBs to an ambient temperature of 10*F or abow 1hc NRc Staff has reviewed anf groved the ticensee's adnumstranve knut, as documrnted m NRC safety evaluanon dated septembet 26.19h 482

certification in 1993. In the Staff's judgment, experimental testing to verify n

, well established process is unnecessary.

(viii) The Petitioners claim that the thermal analyses for the VSC 24 have not been experimentally verified. De thermal analyses for the VSC 24 contained conservative key assumptions, including a total heat generation of 1 kW per assembly (a total of 24 kW per cask). This assumption is conservative because it is highly unlikely that each assembly loaded in the cask will generate I kW of heat. In addition, the assembly and total cask heat loads will continually decrease over time as the spent fuel decays. In light of the conservatisms in the thermal analyses, the Staff does not see the need for requiring experimental verification of the VSC 24 thermal analyses. Nevertheless, the COC requires that a thermal test be performed on the first VSC-24 to be loaded. The pumose of the test is to measure the heat removal performance of the VSC-24 system.

The Licensee for Palisades performed such a test and summarized its results in a letter to NRC dated June 10, 1993. He temperatures measured during the test were lower than the predicted temperatures. The results thus indicate that the VSC 24 performs its intended heat removal function. The thermal test at Palisades was performed with a 12 kW heat load. To date, no VSC-24s have been loaded with greater than 12-kW heat hud. As required by the COC, the thermal test must be performed for the first cask to use any higher heat loads, up to 24 kW, ne NRC Staff believes, based on the foregoing, that an independent, third-party review is not warranted by the Petitioners' specific concerns However, NRC myiew activities relating to the VSC-24 will nonetheless continue, in nmia.W. NRC inspection activities at the facilities operated by the Licensees,

& " ^. 4 vendor, and the VSC 24 fabricators may lead to additional reviews M M WC-24. In addition, the Staff is currently reviewing a proposed

, anie. ,ent, submitted by the VSC 24 vendor, to the SAR and COC for the VSC-24 design. This review will be performed in accordance with the Staff's

. " Standard Review Plan for D,y Cask Storage Systems" (NUREG-1536) to 1 ensure the thoroughness, quality, and consistency of the review. Where relevant, recent operational, technical, and safety issues related to the VSC 24 design will be considered by the Staff in this review.2

, in addition, it is my judgment that the NRC Staff is fully capable of fulfilling the responsibility for reviewing, approving, and certifying dry cask storage systems to be used under 10 C.F.R. Part 72 which, by law, belongs to the NRC, in conducting its review, the NRC Staff must have reasonable assur e that the cask system will safely store spent fuel over the period of the certificate.

2 Recent concems relanns to the MSB closure welds. as documented in NRC Inspecaon Report No. 7210W97 2% daed April 15,1997, may result in further evaluauons or the v5C-24 deugn and if necessary, appropnate regulatory menon to ensure continued safe use of the v5C-24 483

i 5

Further, the Staff will assign the necessary resources and expertise to perform such reviews. When the NRC Staff lacks either the resources or expertise to perform all or portions of the review in-house, the NRC may, and does, supplement its own ranks by using outside specialists.

item 3: Prohibit Loading of VSC 24s Pending Assessment of 4

Cask Materials i

Petitioners request the NRC to prohibit loading of VSC-24s until the potential -

impacts of all material aspects of the casks have been fully assessed. As previously stated,Bulletin 96-04 was issued to request information on material interactions and compatibility in spent fuel storage and transportation casks. In j response to this request, the Licensees for ANO, Point Beach, and Palisades submitted evaluations on possible material interactions in the VSC-24 and the effects of such interactions on cask perfonnance and operation. He only significant material interaction identified was between the zine based coating j and the borated spent fuel pool water, As previously discussed, the operating controls and limits put in place by the Licensees provide an adequate level of confidence to prevent the adverse effects of this interaction (generation and

, possible ignition of hydrogen gas and possible depletion of boron in the water).

i ne Staff reviewed these evaluations and, based on the information provided, j concluded that none of the identified material interactions would adversely affect

the VSC 24's ability to safely store spent fuel over the 20-year period of the l certificate. He results of the Staff's reviews are oocumented in the Staff's 1

December 3,1996, April 8,1997, and June 12, 1997 safety evaluations for ANO, Point Beach, and Palisades, respectively.

' Item 4: Prohibit Loading of VSC 24s Pending Experimental Verification of Thermal and Other Design Assumptions Petitioners request the NRC to prohibit loading of VSC-24s until there is experimental verification of temperature calculations and heat transfer assess-ments and other design assumptions. He thermal and other engineering and

design analyses for the VSC-24 contained conservative key assumptions which

!- are discussed in the SAR and SER. In additica, the acceptance criteria for these

analyses have margins of safety that the Staff considers to be sufficient. In light of the conservatisms and safety margins'in the thermal and other analyses, the Staff does not see the need for requiring experimental verification of the thermal and other design assumptions used in evaluating the VSC 24, 4

e l

4 3

l item 5: Prohibit Loading of VSC 24s Pending Assessment of l Material Coatings

- Petitioners request the NRC to prohibit loading of VSC 24s until the safety

- of the material coatings on components and structures has been justified. As discussed above, material interactions within the VSC 24 and their effect on

' cask operations and performance were evaluated by the Licensees in response to Bulletin 96-04 and reviewed by the Staff. Specifically, the Licensees evaluated, and the Staff reviewed, the use of the zine-based coating, its reaction with borated water and other cask environments, and the effect of the reaction or reaction products on cask operations and on the performance of the various cask components and structures. De Staff concluded that use of existing VSC-24s with the zine-ba,cd coating is acceptable in light of the operating controls and limits for preventing hazardous conditions that must be properly implemented

. by Licensees during cask loading and unloading. Based on the information provided, the Staff also concluded that neither the coating itself, nor its reaction with borated water or other cask environments, would have an adverse effect on the performance of the cask components or structures during the period of spent fuel storage.

IV. CONCLUSION The Petitioners requested that the NRC prohibit loading of VSC 24s until the COC, SAR, and SER are amended to contain operating controls and limits to prevent hazardous conditions. After reviewing each of the Petitioners' ,

claims, I conclude that, for the reasons discussed above, no adequate basis

- exists for Franting the Petitioners' request to prohibit Licensees' use of. the VSC-24 for dry cask storage of spent nuclear fuel at Palisades, Point Beach, or ANO pending: (1) revision of the SAR, SER, and COC for the VSC.

24 to contain operating ' controls and limits to prevent hazardous conditions; (2) an independent third party review to examine the safety issues raised by the Petitioners; and (3) experimental verification of temperature calculations and heat transfer assessments and other design assumptions.' 1%rthermore, I conclude that the Petitioners

other two requests, an assessment of potential

- impacts of VSC 24 material aspects and a safety justification of material coatings

!' on components and structures, have already been fulfilled through the Staff's p - review of the Licensecs' responses to Bulletin 96-04.

l- A copy of this Decision will be filed with the Secretary of the Commission -

l for the Commission to review in accordance with 10 C.F.R. 6 2.206(c).

I i

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As provided by this regulation, this Decision 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 within that time.

FOR THE NUCLEAR REGULATORY COMMISSION Malcolm R. Knapp, Acting Director Office of Nuclear Material Safety and Safeguards Dated at Rockville, Maryland, this 18th day of June 1997.

486 '

Cite as 45 NRC 487 (1997) DD-9716 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION Sarr vel J. Collins, Director j

- In the Matter of Docket No. 50-160 GEORGIA INSTITUTE OF TECHNOLOGY (Georgia Tech Research Reactor, Atlanta, Georgia) June 27,1997 He Director of the Office of Nuclear Reactor Regulation (NRR) denies a

- petition filed with the Nuclear Regulatory Commission (NRC or Commission) by letter dated October 23,1994, by Ms. Pamela Blockey-O' Brian (Petitioner),

requesting that a:tions be taken regarding the Georgia Tech Research Reactor (GTRR) operated by the Georgia Institute of Technology (the Licensee). The petition was deferred pending a decision by the Atomic Safety and Licensing Board (ASLB) on Georgia Tech's license renewal application, in which issues substantially similar to the Petitioner's were raised. He petition is denied based on the Director's analysis of the technical issues, set forth in the Decision, which analysis showed no technical basis warranting granting the petition.

DISPOSITION OF ISSUES RAISED VIA PETITION PURSUANT TO 10 C.F.R. 5 2.206 THAT ARE ALSO UNDER CONSIDERATION IN A PROCEEDING BEFORE A PRESIDING OFFICER Re Commission ordinarily expects the Staff to deny a petition filed pursuant

= to 10 C.F.R. 5 2.206 that raises the same issues that are being considered in a pending adjudication on the basis of the pendency of the identical matters in a proceeding involving the same licensee or facility. Georgia Power Co. (Hatch Nuclear Plant, Units I and 2; Vogtle Electric Generating Plant, Units 1 and 2), CLI 9315, 38 NRC 1, 2 3 (1993); see General Public Utilities Nuclear Corp. (Three Mile Island Nuclear Station, Units 1 and 2; Oyster Creek Nuclear 487 i

e e

4

Generating Station), CLI.85-4,21 NRC 561,563-65 (1985); Pacific Gas and Electric Co. (Diablo Canyon Nuclear Power Plant, Units I and 2), CLI 81-6, 13 NRC 443,446 (1981). (This general rule is not intended to bar a petitioner from seekinF immediate enforcement action from the Staff in circumstances in which the presiding officer is not empowered to grant such relief. Vogt/c, 38 NRC at 3.) 'Ihe same result can be achieved by the Staff defening consideration of issues raised in a petition filed pursuant to 10 C.F.R. 5 2.206 that are being considered in a pending proceeding involving the same licensee and facility.

TECilNICAL ISSUES DISCUSSED

'lhe following technical issues are discussed: hianagement of the GTRR; Security.

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

1. INTRODUCTION On October 23, 1994, his. Pamela Blockey-O'Brien (the Petitioner) filed a petition with the U.S. Nuclear Regulatory Commission (NRC) Staff pursuant to 10 C.F.R. 5 2.206. This petition requested that the NRC Staff revoke the license for the Georgia Tech Research Reactor (GTRR), shut down this research reactor and its support facilities, and remove all radioactive material and contamination off site to a government created " National Sacrifice [A]rea" such as the Savannah River or Oak Ridge facilities. In addition, the Petitioner requested that the NRC Staff withdraw all license authority nationwide involving the discharging or dumping of any quantity of radioactive material into all the sewers or waters in the United States or oceans of the world, and withdraw all licenses to all nuclear facilities, including nuclear power plants (NPPs), that operate under "as low as reasonably achievable" (ALARA) principles. Finally, the Petitioner requested that the NRC Staff prohibit the transportation of radioactive material by mail and modify every license issued to transponers of radioactive materials and builders of NPPs to require these parties to put, in 2 foot-high letters, on everything they transport or build, the words " DANGER RADIOACTIVE" and, in smaller letters, "there is no safe level of radiation, any exposure can effect health."

As bases for the request to shut down and decontaminate GeorF ia Tech Research Reactor, the Petitioner asserted that (1) a water flume comes out of the ground "destabilizing the reactor and the ground in some way"; (2)

"[r]adiation levels in soil and vegetation climb markedly in GA EPD [ Georgia 488

Environmental Protection Division] documents" around the reactor; (3) there is no record of air monitoring ever havinF been done; (4) heavy rainfall causes water to back up in the sewer and drainage lines causing flooding of the reactor parking lot and campus, as well as causing sinkholes, " puff ups" on campus ground, and welded-shut manhole covers to be blown off; (5) radioactive contaminants have been routinely discharged into the sanitary sewer from the reactor's wastewater holding tank and contamination spread by backup of the sewage system; (6) should the reactor be further destabilized, the reactor and the tank holding cobalt 60 could " break apart," causing radioactive contaminants to " drain into groundwater /down sewers /into the runoff ditch"; (7) the reactor is in an earthquake zone; (8) there is absolutely no reason to keep the reactor operating; (9) security at the reactor is extremely lax; and (10) in case of an accident or terrorist attack, evacuation of the campus and downtown Atlanta would be impossible, especially during the 1996 Olympics.

In a Partial Director's Decision Under 10 C.F.R 6 2.206, dated July 31,1995 (DD-95-15), the Acting Director, Office of Nuclear Reactor Regulation (NRR),

for the reasons stated in that decision, denied the Petitioner's requests except for the request that the NRC Staff revoke the license of the GTRR, shut down this research reactor and its support facilities, and remove all radioactive material and contamination off site to a government-created " National Sacrifice [A]rea" such as the Savannah River or Oak Ridge facilities, insofar as that request rested on bases numbers (8) and (9), and that portion of basis (10) that deals with potential terrorist attacks, as set forth above. See DD-95-15,42 NRC 20,40 n.37 (1995).

(The portion of basis (10) that relates to evacuation and emergency planning also is discussed in DD-95-15,42 NRC at 40-43.)

Basis (8) includes concerns that substantial management deficiencies persist.

Basis (9) involves concerns about security. Basis (10) includes concerns about evacuation in case of a terrorist attack. Since these concerns were related to issues in an ongoing license renewal proceeding before an Atomic Safety and Licensing Board (ASLB), they were not addressed in DD-95-15. The Commission ordinarily expects the Staff to deny a petition filed pursuant to sectitn 2.206 that raises th same issues that are being considered in a pending adjudication on the basis of the pendency of the identical matters in a proceeding involving the same licensee or facility, Georgia Power Co. (Hatch Nuclear Plant, Umts 1 and 2; Vogtle Electric Generating Plant, Units 1 and 2), CL1 15,38 NRC 1,2-3 (1993); see General Public Utilities Nuclear Corp. (hree Mile Island Nuclear Station, Units 1 and 2; Oyster Creek Nuclear Generating Station), CL1-85-4,21 NRC 561,563-65 (1985); Pacific Gas and Electric Co.

(Diablo Canyon Nuclear Power Plant, Units 1 and 2), CLI-816,13 NRC 443, 446 (1981). (This general rule is not intended to bar a petitioner from seeking immediate enforcement action from the Staff in circumstances in which the presiding officer is not empowered to grant such relief. Vogtle,38 NRC at 3.)

489

The same result can be achieved by the Staff deferring consideration of issues raised in a petition filed pursuant to section 2.206 that are being considered in a pending pracceding involving the same licensee and facility, as was done with regard to Petitioner's concern regarding the management of the GTRR.

The NRC Staff received additional letters dated Nosember 12 and December 4, 1994, and Rbtuary 21, February 23, h1 arch 6, March 28, April 19, hiay 18.

June 27, and July 18,1995, from the Petitioner and also considered these letters in DD 95-15.

His Final Director's Decision addresses the management concerns in issue (8) above and security concerns in issues (9) and (10) above for the request to shut down and decontaminate the GTRR in the section 2.206 petition of October 23,1994. The NRC Staff received additional letters from the Petitioner dated -

August 18, August 21, August 28, August 31, September 17, and October 27, 1995; and January 10, January 27, h1 arch 14, and hiay 24, 1996. All letters related to this petition were considered in this Final Director's Decision and have been placed in the Public Document Room and docketed under the GTRR

+

Docket Number (50-160). For the reasons set forth below, the Petitioner's remaining request is denied.

II. DISCUSSION A. htanagement of the GTRR Petitioner stated that "[t]here is no reason to keep the [GTRR] operating,"

and asserted that substantial management deficiencies persist. As stated above, DD-95-15 did not address the management issue since it had been admitted in a proceeding on the renewal of the license for the GTRR.

He history of the license renewal proceeding is set forth in the ASLB's initial Decision in that proceeding. LBP 97-7, 45 NRC 265, 268 70 (1997).

A copy of that decision was sent to the Petitioner, In the initial Decision, the ASLB concluded, in part, that:

1. The Apphcant's performance m the post restart penod, although not entirely sat <

isfactory, has substantially improved since the shutdown of the reactor in 1988. fttther, Georgia Tech's perfonnance in the post-restart penod does not support GANE's assemon that management of the GTRR is inadequate and that the hcense renewal application should therefore be denied Nor has GANE met its burden of demonstrating thnt " substantial man.

agement denciencies persist"

2. , We conetude that G ANE has not demonstrated " management impropneties or poor 'mtegnty' [thatl relate directly to the proposed licensing action," or that "the GTRR as presently organized and staffed [ fads tol provide reasonable assurance of candor and withngness to follow NRC regulations." Moreover, the evidence supports 6ndmgs that "the facihty's current management encourages a safety-conscious attitude, and provides an environment in which employees feel they can freely voice safety concems," and there is 490

" reasonable assurance that the GTRR facihty can be safely operated" m that "the GTRR's current spanagenn'nt in] esther is unfit [nlor structured unacceptably."

3. The Apphcant's managernent of the Georgia Tech Research Reactor comphes with all apphcable regulatory requirements, and proudes reasonable assurance that its management of the GTRR facthry, upon the renewal of License No. R 97. will not be ininucal to the common defense and secunty or to the heahh and safety of the pubhc.

Id. at 312-13 (citations omitted).

De ASLD's initial Decision considered all the evidence submitted on the record during the proceeding. The Petitioner did not submit any information to the NRC in support of its petition that was significantly different from the evidence considered by the ASLB in the license renewal proceeding on the management issue.

Since the ASLB proceeding record closed in June 1996, four additional NRC inspections of the GTRR facility have been conducted (NRC Inspection Reports No. 50-160/96-02,50-160/96-03, $0-160/96-04, and 50-160/96-05 which were sent to the Petitioner). Three of the inspections found no violations; the violations that were found and documented in NRC Inspection Report No. 50-160/96-02, do not provide a basis for changing the NRC Staff's conclusion with regard to Georgia Tech's management of the facility.

The NRC Staff's inspection findings subsequent to the close of the ASLB record do not provide a basis for concluding that substantial management deficiencies have arisen with regard to the GTRR since the record in the license renewal proceeding closed. The Petitioner does not otherwise provide any information that would be a basis for the NRC Staff to conclude at this time that the management and organization of the Georgia Tech Research Reactor fails to comply with the Atomic Energy Act and NRC regulations. Although the Petitioner in very broad terms opposes operation of the facility, the application makes clear that its intended purpose is in keeping with lawful uses authorized in the Atomic Energy Act of 1954, as amended. He proposed operation has been found to acceptably comply with all applicable NRC regulatory requirements.

Based on the foregoing, the NRC Staff concludes that no information has been provided on this issue to warrant the action requested by the Petitioner, B. Security issues Petitioner raised two issues regarding security, asserting that (1) security at the GTRR is extremely lax and (2)in case of accident or terrorist attack, evacuation of the campus and downtown Atlanta would be impossible, especially during the 1996 Olympics. These two issues are discussed below.

Georgia Tech has implemented a security plan for the research reactor that is consistent with the applicable requirements of 10 C.F.R. Part 73, " Physical 491

Protection of Plants and Materials." This has been confirmed through the relatively recent NRC safeguards and security related inspection activities in NRC Inspection Reports No. 50-160/95-02, 50160/95-N, 50-160/95-05, 50-160/96-01, 50-160/96-03, and 50-160/96-N. (Inspection Reports No. 50-160/95 02,50-160/95-04, and 50-160/96-01 were admitted into evidence in the license renewal proceeding.)

Inspection Report No. 50-160/95-02 identified a violation for a failure to submit material status reports in a timely manner. Otherwise the inspection found that the safeguards and security activities were acceptable, On October 26,1995, a television news media crew entered the Neely Nuclear Research Center, which houses the GTRR, and explored and filmed portions of the center. In response, the NRC conducted an inspection ot' the GTRR from October 3 to November 3,1995, as documented in NRC Inspection Report No.

50-160/95-04, which states:

This Special announced safeguards inspection was conducted to review the circumstances surroundmg an umnvited tour of portions of the Neely Nuclear Research Center by a television news media crew which occurred, apparently.on the morning of October 26,1995.

, Neither the licensee nor the inspector could 6nd any evidence of a secunty breach of the protected area. One hcensee employee was idenu6ed who had seen parts of the video made by the telesision crew supposedly on October 26.1995; according to that employee, the video shows two secunty doors being challenged by the television crew which remained locked. Tius employee stated that the video shows the crew tounng intenor and extenor areas of the Center wiuch are open to the pubhc or students and staff. On Novemter 10, l the inspector viewed the television showing of the video taken dunng this esent and could fmd no indication that the television crew had unauthonzed access to the protected / radiation controlled area. . No violanons or devuitions were identihed.

In view of these inspection 6ndings, the television media crew's tour is not a basis for granting the Petitioner's request.

The ASLB discussed these events in the context of the contention regarding management deficiencies, and made findings of fact consistent with this conclu-sion. LBP 97-7,45 NRC at 296-98. It stated:

Upon review of the evidence of this event, we agree with the Is]taff , that the IN Television film crew's intrusion into the reactor complex does not reflect inadequate management by the [alpplicant. To the contrary, the secunty plan appears to have worked as intended, in comphance with applicable regulatory requirements. Further, as observed by the [s)taff, the [alpphcant's subsequent decision to upgrade its secunty measures beyond the requirements of the secunty plan may be viewed as demonstratmg good managenaljudgment.

Thus, this matter does not provide grounds for denying or conditioning the hcense renewal application.

Id. at 298 (citation omitted).

492

inspection Report No. 50-160/95-05 refers to the inspection conducted De-cember 5 7,1995:

The special mspection addressed the facihty's reactor status, physical inventory determi-nations, and other actmties associated with n aintaining a natenal control and accounting program within regulatory requirements, the bcensed possession hmit, and authorized uses of special nuclear material. Within the scope of the inspection, no non comphance issues were ident 6ed. The mspector determined that the heensee had implemented adequate controls for special nuclear material ($NM), and that accurate SNM accounting records were being nuintamed. ;

Inspection Report No. 50-160/96-01 refers to the inspection c. acted on January 17 and 18,24 and 25,29 and 30, and February 5 7,9.15-18 .tnd March 15,1996. This inspection examined security provisions for fuel processing and shipment offsite. As an additional precaution in regards to security during the Olympic Games, the Licensee had determined to remove all GTRR fuel from the f cility prior to the Gar.ics and not to replace it until after the Games.

The inspection found that in addition to meeting regulatory requirements the Licensee provided additional measures (e.g., a guard was assigned to various observed activities).

Inspection Report No. 50-160/96-03 refers to the inspection conducted on June 17,18, and 27, 'nd July 3, 5, and i t,1996. This inspection included onsite and offsite review of security preparations for the Olympic Games.

' The inspection concluded: "The controls implemented by the licensee and the precautions taken are adequate to protect licensee personnel and the public."

The inspection documented in Inspection Report No. 50-160/96-04 was conducted on July 17 and 29,1996. This inspection reviewed the preparation for the Sammer Olympic Games and found that:

[Tlhe university had taken additional safeguards measures to control access to the Campus and to the Research Control Area. The licensee had taken additional safeguards nwasures to contrc? access to the Neely Nuclear Research Center (NNRC). The additional security l measures taken as a result of the 1996 Olympic Games were reviewed and/or observed by the inspectors. On July 17 and 29.1996, the inspectors visited the Neely Nuclear Research Center, met with the Director of the Center, toured the facthry and venfied contmued comphance with the Physical Secunty Plan (PSP). The inspectors were granted unfettered access to the Research Control Area as well as to the Center and emergency access during the Olympics was assured treause the inspectors and selected management of Region 11 had been provided with special picture badges to facihtate NRC response. The presence of mihtary pohee, Campus pohce and additional State and liederal law enforcement officers in the immediate vicinity of the Center was observed by the inspectors. The f access controls, barners, assessment capabihties, communication capabihties and detection equipment required by the NRC were in place. Additional etterior lights had been installed by the heensee to assist patrolhng of6cers. Additional fencing around the Center was also noted by the inspectors. . . The inspector concluded that the hcensee was meeting NRC requirernents and had effecuvely imposed proactive secunty measures.

493 r

With reFard to the contention on the physical security of the site during the 1996 Summer Olympic Games held in Atlanta, Georgia, the ASLB decision observed that the Apphcant, respondmg to several Comnussion inquiries relative to secunty at the Olympic Games, deternuned to remove all nuclear fuel from the site pnor to the Olympic Games and not to replace it until after the Games. The Commission accordmgly remanded the secunty contention to us for appropriate action . . and we issued a Parual Initial Decision dumissing the contenuon as moot.

LDP-97-7,45 NRC at 270. See LDP-95-19,42 NRC 191 (1995).

In summary, the physical security plan was verified to provide acceptable procedures for event response and access control, t.nd the security preparations for the Olympics were acceptable. Observations of the facility and activities confirmed the use of sceurity-related equipment and controls as required by the physical security plan and consistent with the special nuclear material that is present at the facility. He Petitioner asserted that security at the research reactor 3

was lax; however, access is controlled and monitored as required. Further, this evaluation confitmed the continued acceptability of the security provisions to i

deal with potential terrorists attacks. The findings do not provide a basis for changing the conclusion reached in DD 9515 on the adequacy of emergency plans for the facility. DD-95-15,42 NRC at 40-43. The NRC Staff has found no reason to conclude that the security at the reactor is not acceptable. De Petitioner provided no facts to conclude otherwise.

111. CONCLUSION With regard to the requests made by the Petitioner discussed herein, the NRC Staff finds no basis for taking such actions. Accordingly, the Petitioner's requests for action, pursuant to section 2.206 on the Georgia Tech Research Reactor, are denied.

A copy of this Decision will be filed with the Secretary for the Commission as provided by 10 C.F.R. 6 2.206(c) of the Commission's regulations. As provided by this regulation, the Decision will constitute the final action of the Commission 494

25 days after issuance unless the Commission, on its own motion, institutes review of the Decision in that time.

FOR THE NUCLEAR REGULATORY COMMISSION f

Frank J. Miraglia, Acting Director Office of Nuclear Reactor Regulation Dated at Rockville, Maryland, t

- :his 27th day of June 1997, J

495

ML20210L101

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