ML20133G138

Nuclear Regulatory Commission Issuances for October 1996. Pages 107-228

ML20133G138
Person / Time
Issue date:	12/31/1996
From:	NRC OFFICE OF ADMINISTRATION (ADM)
To:
References
NUREG-0750, NUREG-0750-V44-N04, NUREG-750, NUREG-750-V44-N4, NUDOCS 9701150176
Download: ML20133G138 (129)

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

Single copies of this publication are available from i

National Technical information Service Springfield, VA 22161 i

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1 NUREG-0750 Vol. 44, No. 4 l Pages 107-228 l

I

! INUCLEAR REGULATORY

! COMMISSION ISSUANCES 1

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! October 1996 I

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I This report includes the issuances received during the specified period from the Commission (CLI), the Atomic Safety and Licensing 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 herein are not to be deemed a part of those opinions or have any independent legal significance.

U.S. NUCLEAR REGULATORY COMMISSION Prepared by the Division of Freedom of information and Publications Services Office of Administration U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 (301-415-6844)

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

Greta J. Dicus l Nils J. Diaz Edward McGaffigan, Jr.

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B. Paul Cotter, Jr., Chief Administrative Judge, Atomic Safety and Ucensing Board Panel j l

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CONTENTS Issuances of the Nuclear Regulatory Commission LOUISIANA ENERGY SERVICES (Claiborne Enrichment Center)

Docket 70-3070-ML ORDER, CLI-96-8, October 2,1996 . . . . . . . . . . . . . . . . ..... 107 U.S. ENRICilMENT CORPORATION (Paducah. Kentucky, and Piketon, Ohio)

Dockets 70-7001, 70-7002 '

MEMORANDUM AND ORDER, CLI-96-10, October 18, 1996 ... . I14 YANKEE ATOMIC ELECTRIC COMPANY (Yankee Nuclear Power Station)

Docket 50-029-DCOM (Decommissioning Plan)

ORDER, CLI-96-9, October 18,1996 .. . . ... . . .. .. . 112 i

Issuances of the Atomic Safety and Licensing Boards l I

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

Docket 50 219-OLA (ASLBP No. 96-717-02-OLA)

MEMORANDUM AND ORDER, LBP-96-23, October 25,1996 .. 143 J

JAMES L. SHELTON (Order Prohibiting Involvement in NRC-Licensed Activities (Effective immediately))

Docket IA 95-055 (ASLBP No. 96-712-01-EA)

MEMORANDUM AND ORDER, LBP-96-19, October 1,1996 121 I

JUAN GUZMAN (Order Prohibiting Unescorted Access or Involvement in NRC-Licensed Activities)

Docket IA 96-020 (ASLBP No. 96-715-03-EA)

MEMORA>'DUM AND ORDER, LBP-96-20, October 16,1996 .. . 128 iii

NORTHERN STATES POWER COMPANY (Independent Spent Fuel Storage Installation)

Docket 72-18-ISFSI (ASLBP No. 97-720-01-ISFSI)

MEMORANDUM AND ORDER, LBP-96-22, October 24,1996 138 1

TESTCO, INC. '

(Order Imposing Civil Monetary Penalty; General License)

Docket 150-00032-EA (ASLBP No. 96-719-04-EA)(EA 95-101)

MEMORANDUM AND ORDER, LBP-96-19, October 1,1996 . . 121 WASHINGTON PUBLIC POWER SUPPLY SYSTEM (WPPSS Nuclear Project No. 3)

Docket 50-508-OL (ASLBP No. 83-486-01-OL)

MEMORANDUM AND ORDER, LBP-96-21, October 16,1996 . 134 Issuances of Directors' Decisions ALL NUCLEAR POWER PLANTS All Dockets (All Licenses)

DIRECTOR'S DECISION UNDER 10 C.F.R. 5 2.206, DD-96-12, September 26,1996 . . . . 169 CLEVELAND ELECTRIC ILLUMINATING COMPANY (Perry Nuclear Power Plant, Unit 1; Davis-Besse Nuclear Power Station, Unit 1)

Docket 50-440-A,50-346-A DIRECTOR'S DECISION UNDER 10 C.F.R. 6 2.206, DD-96-15. October 17,1996 . . .. . . 2G4 DUKE POWER COMPANY, et al.

(Catawba Nuclear Station, Units I and 2)

Docket Nos. 50-413,50-414 (License Nos. NFP-35, NPF-52)

DIRECTOR'S DECISION UNDER 10 C.F.R. s 2.206, DD-96-14 October 10,1996 . .. . . . 187 FLORF)A POWER CORPORATION (Cr ,stal River Nuclear Generating Plant, Unit 3)

D,cket 50-302

'JIRECTOR'S DECISIOc UNDER 10 C.F.R. s 2.206, DD-96-13, October 7,1996 . . 180 iv

NORTHEAST NUCLEAR ENERGY COMPANY (Millstone Nuclear Power Station, Unit 1)

Docket 50-245 (License No. DPR-21)

DIRECTOR'S DECISION UNDER 10 C.F.R. 5 2.206, DD-96-16, October 31,1996 ... . . . ... . 214 NORTHEAST NUCLEAR ENERGY COMPANY j (Millstone Nuclear Power Station, Unit 1)

Docket 50-245 (License No. DPR-21)  !

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

, DD-96-17, October 31,1996 . . ... . . . .., 221 I

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4 Cite as 44 NRC 107 (1996) CLl-96-8 l

! I i UNITED STATES OF AMERICA 1 3- NUCLEAR REGULATORY COMMISSION ,

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COMMISSIONERS:

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 (Claiborne Enrichment Center) October 2,1996 The Commission considers a petition for review of an Atomic Safety and Licensing Board Partial Initial Decision, LBP-96-7,43 NRC 142 (1996). The Petitioner requested review of only that section in the decision that resolved all contentions on emergency planning in favor of the Applicant. The Commis-sion grants the petition for review in part and denies the petition in part. The

- Commission grants the petition only on a single issue: whether the Applicant's emergency plan clearly describes the intended role and training of the Appli- .

cant's onsite fire brigade. Finding that the Applicant has adequately clarified )

the role of the onsite fire brigade, the Commission finds no need to remand this .

question to the Board. The Commission orders that appropriate revisions be l made to the Safety Analysis Report (SAR) and Safety Evaluation Report (SER) 3 to reflect the clarified understanding of the onsite fire brigade's role.

EMERGENCY PLANNING: PREDICTIVE FINDINGS Established NRC practice permits the licensing board, where appropriate, both to refer minor safety matters to the NRC Staff for posthearing resolution, and to make predictive findings on emergency planning that will be subject to posthearing verification. But only those matters not material to the basic l I

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findings necessary for issuance of a license may be referred to the NRC Staff for posthearing resolution - e.g., minor procedural or verification questions.

ORDER

'Ihe Commission has before it a petition for review of the Atomic Safety and Licensing Board's Partial Initial Decision, LBP-96-7,43 NRC 142 (1996), i filed by the Intervenor, Citizens Against Nuclear Trash (CANT). CANT seeks Commission review of the portion of the Board's decision resolving all con-tentions on emergency planning in favor of the Applicant. The NRC Staff and the Applicant, Louisiana Energy Services (LES), oppose CANT's petition for review.

We deny the petition except for a single issue: Did the Licensing Board err ,

when, after raising a question whether the Applicant's emergency plan clearly l

describes the intended role and training of the Applicant's onsite fire brigade, l it left the question for posthearing resolution by the NRC Staff?' We hold that the Board should not have left the fire brigade question undecided, but find that ,

, any ambiguity about the intended role and training of the onsite brigade now I has been resolved by the Applicant in its answer to CANT's petition for review.

We direct the Applicant to amend its emergency plan accordingly. No further review or relief is necessary.

With respect to emergency planning it is " established NRC practice that, where appropriate, the Licensing Board may refer minor safety matters not pertinent to its basic findings to the NRC Staff for posthearing resolution, and may make predictive findings regarding emergency planning that are subject to posthcaring verification." Commonwealth of Massachusetts v. NRC, 924 F.2d 311, 331 (D.C. Cir.1991), cert. denied, 502 U.S. 899 (internal quotation and citation omitted). But only matters not material to the basic findings necessary for issuance of a license may be referred to the NRC Staff for posthearing resolution - e.g., minor procedural or verification questions. See Consolidated

. Edison Co. of New York (Indian Point, Unit 2), CLI-74-23,7 AEC 947,951-52 (1974). Accord, Philadelphia Electric Co. (Limerick Generating Station, Units I and 2), ALAB-836,23 NRC 479,494 (1986). "[T]he 'posthearing' approach should be employed sparingly and only in clear cases." Indian Point,7 AEC at 952.

3 Here, the Board found (apparently sua sponte) that testimony by the Appli-cant's expert, Peter G. LeRoy, " appears to contradict" the written description I

CANT raises several other issues in its penuon, largely related to comphance with Regulatory Guides (as oppowd to comphance with regulauons themselves). None of these issues nrets the standards for review set out in 10 C F R. I: 786(b) 108 1

of the role of the onsite fire brigade contained in two documents, the Safety Analysis Report (SAR) and Safety Evaluation Report (SER). See 43 NRC at 161. The SAR and SER describe the onsite brigade as a " supplement" to the local fire department, while the expert viewed the onsite brigade as principally responsible for some types of fires, with the local fire department a mere "back-up" Id. The Board referred this " ambiguity" to the NRC Staff and directed that, if necessary, the emergency plan, SAR, and SER be amended to reflect the actual intended role of the onsite brigade. The Board also directed the Staff to i

" ensure that the size and training of the brigade are sufficient to meet such a i differing role." Id.

By referring the role of the onsite fire brigade to the NRC Staff, the Board implicitly treated it as a minor matter. On the other hand, the Board characterized the fire brigade's role as "important" and stated that it was l

" troubled" by the ambiguity introduced by the expert's testimony. 43 NRC at 161. The issue is "important," according to the Board, "because the intended role of the onsite fire brigade may affect the number of fire brigade members needed and the kind of training the brigade should receive." Id. (emphasis added).

The fire brigade's role also appears material to the Board's basic emergency planning findings. The Board stated that LES must demonstrate that its emergency plan meets the requirements of NRC regulations. 43 NRC at 145.

The Board went on to find that the Applicant's emergency plan complies with '

the regulatory requirements to provide a "brief description of the responsibilities of licensee personnel should an accident occur" and "a brief description of . .

the training that the licensee will provide workers on how to respond to an emergency." See 43 NRC at 156-58 (citing 10 C.F.R. 69 40.31(j)(vii) and (x),

70.22(i)(3)(vii) and (x)). This finding provided, in part, the underpmning for the Board's ultimate conclusion that "the CEC [Claiborne Enrichment Center]

emergency plan complies with the Commission's emergency plan regulations."

43 NRC at 165.

In these circumstances, the Board itself ought to have resolved any question about the fire brigade's role as part of its review of the CEC emergency plan.

Under our case law, which as we explained above reserves the posthearing remedial device for " minor" matters, the Board should not have referred an "important" issue material to licensing to the NRC Staff for later resolution outside the adjudicatory process.

With the case in its current posture, however, we need not remand the fire brigade issue to the Board. The Applicant's answer to CANT's petition for review now has clarified any ambiguity in the intended role and training of the onsite brigade:

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1(a)i11.3.1.1.2. Similarly, SAR Section 4 4.4 provides that the fire brigade will be trained l to respond to fires and contain fire damage and that the local fire department is available "if assistance is needed."

These staternents are entirely consistent with Mr. LeRoy's testimony that in the event of a particular scenario involving a storage yardfire, for which there is httle likehhood that off-site fire fighting capabihty will be required,"the off-site fire fighting capabihty will be relied upon as a backup to on-site fire fighting capabilities." Leroy at 19 fol. Tr. 40. The onsite brigade, being present at the site, would provide the "first response effort" but would not replace local fire fighters who would fight a fire (if not already extinguished) upon their arrival. As the SER plainly states. "[t]he [onaite fire fighting] brigade members are trained and equipped to respond to fire emergencies and contain fire damage unul offsite help from a neighboring fire department arrives." SER at 4-33.

l l See Answer of Applicant LES in Opposition to Intervenor's Petition for Review (May 31,1996) at 6.

We hold the Applicant bound by this description of the onsite brigade's role

- which we understand to describe a "first response" but ultimately secondary role for the brigade except in instances where (as in some storage yard fires) {

it is able to extinguish the fire prior to arrival of the local fire department. We l l direct the Applicant to amend its emergency plan and its SAR to unambiguously l reflect this understanding. Similarly, we direct the NRC Staff to revise its SER l

to include an accurate description of the onsite fire brigade's clarified role. l That leaves only the question whether the emergency plan, incorporating the clarified role of the onsite fire brigade, satisfies NRC requirements. We find l it does. Our rules require but a "brief description" of the " responsibilities" of the Licensee's emergency personnel and of its " training" program. See 10 C.F.R. 6 70.22(i)(3)(vii) and (x). In this Decision we already have ordered revisions in emergency planning documents to clarify the onsite brigade's responsibilities. And an expert witness, Mr. LeRoy, has provided testimony affirming the capability and training of the onsite brigade in its clarified role.

See, e.g., Tr.173 and pp. 28-29 fol. Tr. 40. In addition, the Licensing Board has approved the emergency planning documents' description of training as a general matter. See 43 NRC at 158, Our inspection of those documents confirms the adequacy of the existing "brief description" of training, even as applied to the onsite brigade's clarified role.2 2 of course,if me have overloded any record evidence in resolvmg this pennon for renew, CAKr is free to caH it to our attennon in a pention for reconsideranon. See 10 C F R. I 2.771. h was the L.icensing Doard on its own.

rather than CANT. that first idennfied the apparent ambiguity in the record on the onsite bngade's role. See 43 NRC at 161.

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We accordingly grant the petition for review in part and deny it in part and j l direct that the emergency plan, the SAR, and the SER be amended in accordance - i i with this opinion.

IT IS SO ORDERED.

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For the Commission i

l JOHN C. HOYLE Secretary of the Commission I

Dated at Rockville, Maryland, this 2d day of October 1996.

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Cite as 44 NRC 112 (1996) CLI-96-9 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION COMMISSIONERS:

(

r Shirley Ann Jackson, Chairman Kenneth C. Rogers

! Greta J. Dicus l

Nils J. Diaz l Edward McGaffigan, Jr.

l In the Matter of Docket No. 50-029-DCOM i (Decommissioning Plan)

YANKEE ATOMIC ELECTRIC COMPANY (Yankee Nuclear Power Station) October 18,1996 The Atomic Safety and Licensing Board issued a decision and order granting Yankee Atomic Electric Company's Motion for Summary Disposition in this decommissioning proceeding. LBP-96-18,44 NRC 86 (1996). The Intervenors (Citizens Awareness Network and New England Coalition on Nuclear Pollution) filed with the Commission a Petition for Review of LBP-96-18 and also sought to stay the effectiveness of LBP-96-18 pending Commission consideration of their Petition for Review. The Commission concludes that the Intervenors' Petition for Review raises no substantial questions calling for Commission review of the Board's grant of summary disposition, and therefore denies the Petition for Review and dismisses the Stay Motion as moot. However, the Commission imposes an administrative stay to permit a reviewing court to consider in an orderly way any request for judicial stay that the Intervenors may file.

ORDER On September 27,1996, the Atomic Safety and Licensing Board issued a de-cision and order granting Yankee Atomic Electric Company's ("YAEC") Motion l

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for Summary Disposition in this decommissioning proceeding. LBP-96-18,44 i NRC 86 (1996). He Citizens Awareness Network and New England Coalition on Nuclear Pollution ("Intervenors") filed a Petition for Review of LBP-96-18 and also sought to stay the effectiveness of that order pending Commission con-sideration of their Petition for Review. The Commission on October 2,1996, issued a " housekeeping stay" to permit consideration of the Petition for Review

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, and the Stay Motion. On October 9,1996, the Commission entered a second housekeeping stay, preventing YAEC from undertaking proposed " minor" de-commissioning activities. Both the NRC Staff and YAEC subsequently filed i answers opposing Intervenors' petition and stay motion.

De Commission now concludes that the Petition for Review raises no substantial questions calling for Commission review of the Board's grant of summary disposition. See 10 C.F.R. 6 2.786(b)(4). We therefore deny the j Petition for Review and dismiss the Stay Motion as moot.

. We understand that Intervenors may seek judicial review of this final Com-3 mission action and, in the process, may seek a judicial stay preventing resump-

. tion of decommissioning activities by YAEC, To perrait a reviewing court to 4 consider such a stay request in an orderly way, we will adopt a two-stage admin-l istrative stay.' First, the stay will remain in effect until seven (7) calendar days j after the issuance date of this Order. Second, if the Intervenors file a petition for review and a motion for a judicial stay with an appropriate United States Court of Appeals within that time, the administrative stay will automatically be extended for an additional fourteen (14) calendar days or until the court of i

appeals acts on the request for a judicial stay, whichever comes first. Cf long i Island Lighting Co. (Shoreham Nuclear Power Station, Unit 1), CLI-91-8, 33 NRC 461,471-72 (1991).

It is so ORDERED.

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, For the Commission

{ JOHN C. HOYLE Secretary of the Commission Dated at Rockville, Maryland, this 18th day of October 1996.

'This stay temporanly keeps in effect the housekeeping stays issued by the Commusion on october 2 and 9, 1996 113 i

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i Cite as 44 NRC 114 (1996) CLI-%10 l UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION COMMISSIONERS:

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

In the Matter of Docket Nos. 70-7001 70-7002 U.S. ENRICHMENT CORPORATION (Paducah, Kentucky, and Piketon, Ohio) October 18,1996 l

The Commission considers four petitions for review of an initial Director's  ;

decision approving certificates for compliance for the U.S. Enrichment Corpo-ration's gaseous diffusion plants in Piketon, Ohio, and Paducah, Kentucky. For failure to meet the requirements of 10 C.F.R. Part 76, the Commission denies two petitions for review, and refers these petitions to the NRC Staff for review and response. On the ground that no " good cause" was shown, the Commission j denies a request for an extension of the time period for seeking Commission review of, and submitting comments on, the Director's decision. The Commis-sion also denies a request that any interested party be permitted to file a petition for review; only those parties that participated in the initial comment stage may petition for review of the Director's decision under Part 70.

RULES OF PRACTICE: PETITION FOR REVIEW UNDER PART 76 To be eligible to petition for review of a Director's decision on the certifica-tion of a gaseous diffusion plant, an interested party must have either submitted written comments in response to a prior Federal Register notice, or provided oral comments at an NRC meeting held on the application or compliance plan.

10 C.F.R. 5 76.62(c).

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4 i l 1 i i RULES OF PRACTICE: PETITION FOR REVIEW UNDER PART 76 4

i Part 76 contemplates a Commission decision on petitions for review of l certification decisions within a relatively short (60-day) time period. See 10

, C.F.R. 5 76.62(c). Extending the Part 76 petition deadline in the absence of a i strong reason is not compatible with the contemplated review period. I 1

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MEMORANDUM AND ORDER i l

j I. BACKGROUND i

i On September 19, 1996, the NRC published in the federal Register (61 1 fid. Reg. 49,360-63) notice of the certification decision of the Director, Office of Nuclear Material Safety and Safeguards (Director), for the U.S. Enrichment Corporation (USEC) to operate the two gaseous diffusion plants (GDPs) located at Paducah, Kentucky, and at Piketon, Ohio. The NRC also issued a Finding of i No Significant Impact (FONSI) concerning the agency's approval of the com-

! pliance plan prepared by the U.S. Department of Energy (DOE) and submitted I by USEC.

t USEC or any person whose interest may be affected, and who had submitted

_ written comments in response to the prior Federal Register Notice on the i application or compliance plan under 10 C.F.R.176.37, or provided oral i

comments at an NRC meeting held on the application or compliance plan

j. under 10 C.F.R. 9 76.39, were eligible to file a petition with the Commission i requesting review of the Director's decision within 15 days after publication of the Director's decision.10 C.F.R. 6 76.62(c).8 i The NRC received four petitions for review of the Director's decision. This

] Memorandum and Order addresses only certain threshold procedural matters

that are raised by those petitions.

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II. PETITIONS FOR REVIEW i The four petitions and related NRC actions to date are as follows:

[ 1. By letter dated September 30,1996, Vina K. Colley of McDermott, Ohio,

who serves as President of PRESS, Portsmouth-Piketon Residents for i

I Nonce of receipt of the apphcanon had appeared in the Federal Regater (60 Fed Reg. 49.026) on september i 21.1995, allowing for a 45-day pubhc comrnent period on the apphcanon and nonemg public meetings to solicit i pubhc input on the cemficanon. A second nonce appeared in the federal Regurer (60 fed. Reg 57.253) on i November 14.1995. providmg for a 45-day pubhc comment pened on the comphance plan. Public meeungs were

{ held on November 28.1995,in Portsmouth. ohio, and on December 5.1995, in Paducah, Kentucky l

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Environmental Safety and Security, petitioned for Commission review.

l Her petition was docketed at the NRC on October 4,1996. Ms. Colley l had spoken at the NRC's public meeting in Portsmouth, Ohio, on .

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November 28,1995, regarding the application and compliance plan.

On October 4,1996, the Office of the Secretary served a copy of her petition on USEC and persons who had piovided written comments on the application or compliance plan during the comment period or had provided oral comments at a meeting held on the application and compliance plan. The Office of the Secretary invited those served to file comments on Ms. Colley's petition by October 15, 1996.

2. By letter dated October 2,1996, two individuals, Mark Donham and Kristi Hanson, of Brookport Illinois, also petitioned for review. Mr.

l Donham participated in the public meeting in Paducah, Kentucky, on December 5,1995, and he and Ms. Hanson joined in earlier' written j comments. Their petition was docketed at the NRC on October 8,1996.

l' On October 9,1996, the Office of the Secretary served the petition on

! the service list, and invited those served to comment on the petition by l October 21,1996.

I 3. By letter dated September 28,1996, Neilly Buckalew, submitted a pc-l tition for review in' the capacity of Executive Director, Kwanitewk, ,

' NATIVE Resource / Network, Meriden, New Hampshire. This letter was docketed at the NRC on October 9,1996. NRC records indicate that nei-ther Neilly Buckalew nor anyone identified as representing Kwanitewk, NATIVE Resource / Network filed written comments on the certification application or compliance plan during the comment period or made oral comments at the public meetings.

4. By letter dated October 3,1996, Diana Salisbury, of Sardinia, Ohio, peti-tioned for Commission review on behalf of the Sycamore Environmental Awareness Group. This correspondence was docketed at the NRC on

- October 7,1996. By letter dated October 4,1996, docketed at the NRC on October 9,1996, Ms. Salisbury submitted an amendment to her letter of October 3,1996. NRC records indicate that neither Ms. Salisbury nor i anyone identified as representing the Sycamore Environmental _ Aware-ness Group filed written comments on the certification application or compliance plan during the comment period or made oral comments at l the public meetings. 1 Hl. THRESHOLD PROCEDURAL MATTERS l

l The petitions for review raise certain procedural matters that will be addressed I as threshold matters. These matters are as follows:

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1. Petitioners' Eligibility to Petition for Review As noted above, the Commission's regulations provide that USEC or any )

person whose interest may be affected, and who had submitted written comments )

in response to the prior Federal Register notice on the application or compliance i plan under 10 C.F.R. 9 76.37, or provided oral comments at an NRC meeting l held on the application or compliance plan under section 76.39, is eligible to file l a petition to the Commission requesting review of the Director's certification.

Two of the petitions are being rejected for failure to meet the conditions of eligibility for the filing of a petition for review.

First, since neither Neilly Buckalew nor anyone identified as representing l

Kwanitewk, NATIVE Resource / Network, Meriden, Hew Hampshire, filed writ. i ten comments on the certification application or compliance plan during the l comment period or made oral comments at the public meetings, they are not 1 eligible to seek Commission review pursuant to the plain terms of 10 C.F.R.

976.62(c). Second, since neither Ms. Salisbury nor anyone identified as repre-senting the Sycamore Environmental Awareness Group filed written comments i on the certification application or compliance plan during the comment period )

or made oral comments at the public meetings, they are not eligible to seek Commission review pursuant to the terms of section 76.62(c).

The correspondence from these parties setting forth their petitions for review will be referred to the NRC Staff for review and for appropriate response. The referral to the NRC Staff does not alter the determination that these petitions  !

are not before the Commission for review of the Director's decision. l

2. Extension of the Comment Period In her letter dated September 30, 1996, Ms. Colley also petitions for an extension of the 15-day period for petitioning for Commission review of the l Director's decision. She asks that the Commission afford no less than an additional 30-day period for filing a petition and comments on the Director's certification decision. She alleges that the 15-day period is insufficient for citizens to obtain, review, and understand the necessary materials. She contends that making materials available at the NRC and at the two GDPs does not allow for full participation by citizens and taxpayers. In their letter dated October 2, 1996, Petitioners Donham and Hanson state that they join in the request of other parties for an extension of the 15-day period for requesting review.

The requests for an extension of the petition deadline are being denied.

Commission rules allow for time extensions only for " good cause." See 10 C.F.R. 6 76.74(b). Here, Petitioners have not established good cause for creation of an additional period for seeking Commission review and for filing further comments. Petitioners do not identify any particular documents that require 117 i

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l additional time for review and comment. In addition, Ms. Colley focuses in large part only on the potential for further review by other citizens and taxpayers across the nation; she gives no specific reason why she or others eligible to file petitions need additional time. Commission rules contemplate a Commission decision on petitions for review within a relatively short (60-day) time period.

See section 76.62(c). Extending the petition deadline in the absence of a strong ,

reason is not compatible with the contemplated review period. l

3. Expansion of the Right to Seek Review Ms. Colley requests that the Commission permit comments by any interested person of the United States. In its rules, however, the Commission did provide a period for general public comment on the application and compliance plan submitted by USEC. Thus, Ms. Colley appears to object to the NRC rule that makes early participaticn a condition for filing petitions seeking Commission review of the Director's decision.

This procedural requirement, in section 76.62(c), was established through notice-and-comment rulemaking. Ms. Colley's objection to the requirement and request for its alteration will not be entertained as part of the Commission I review of the Director's decision, which necessarily focuses on technical and I environmental considerations peculiar to the Piketon and Paducah facilities. He l Commission has established a process for entertaining a petition for rulemaking (10 C.F.R. 6 2.802), i.e., to issue, amend, or rescind any regulation, that Petitioner may wish to pursue.

Oth:r matters raised by the petitions, including, for example, the various substantive challenges to the Director's certification decision and Ms. Colley's I request for national public hearings on continued operation of the GDPs, are reserved for later Commission decision.2 lbr the foregoing reasons, and pursuant to my authority under 10 C.F.R.

5 76.72(b), it is hereby ORDERED that:

1. The petition for review dated September 28,1996, from Neilly Buckalew, submitted in the capacity of Executive Director, Kwanitewk, NATIVE Re-source / Network, Meriden, Hew Hampshire, is rejected and referred to the NRC Staff for review and appropriate response;

2. The petition for review dated October 3,1996, and its amendment dated October 4,1996, by Diana Salisbury, of Sardinia, Ohio, on behalf of the 2

The Comnussion has begua recemns regensive comments to the pennons, includmg a response frorn UsEC to the Colley permon. Any issue r;used in the respimses and not addressed in this order is reserved for later Comnussion determmanon.

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d Sycamore Environmental Awareness Group, are rejected and referred to the NRC Staff for review and appropriate response; i 3. The request by Petitioners Colley, Donham, and Hanson for an additional period for seeking review and submitting comment on the Director's decision is denied; and,

4. The request by Petitioner Colley for expansion of the right to petition for Commission review of the Director's decision to any interested person is denied.

For the Commission JOIIN C. HOYLE Secretary of the Commission Dated at Rockville, Maryland, this 18th day of October 1996.

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! Atomic Safety and Licensing l Boards issuances l l

ATOMIC SAFETY AND UCENSING BOARD PANEL B. Paul Cotter, Jr.,* Chlet Administrative Judge l James P. Gleason,

• Deputy Chlet Administrative Judge (Executive) l Frederick J. Shon,* Deputy Chief Administrative Judge (Technical) i i

4 i j Members

! Dr. George C. Anderson Dr. Richard F. Foster Dr. Kenneth A. McCollom l Charles Bechhoefer* Dr. David L Hetnck Marshall E. Miller l Peter B. Bloch* Ernest E. Hill Thomas S. Moore a

{ G. Paul Bollwed lil* Dr. Frank F. Hooper Dr. Peter A. Moms

! Dr. A. Dixon Callihan Elizabeth B. Johnson Thomas D. Murphy

• Dr. James H. Carpenter Dr. Charles N. Kelber* Dr. Richard R. Partzek l Dr. Richard F. Coie* Dr. Jerry R. Kline* Dr. Harry Rein Dr. Thomas E. Elleman Dr. Peter S. Lam

• Lester S. Rubenstein Dr. George A. Ferguson Dr. James C. Lamb til Dr. David R. Schink Dr. Harry Foreman Dr. Emmeth A. Luebke Dr. George F. Tdey 1

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• Permanent panelmembers I

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Cite as 44 NRC 121 (1996) LBP-96-19 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD Before Administrative Judges:

Charles Bechhoefer, Chairman Dr. Frank F. Hooper Dr. Charles N. Kelber ,

l in the Matters of JAMES L SHELTON Docket No. IA 95-055 (Order Prohibiting involvement (ASLBP No. 96-712-01 EA) in NRC-Licensed Activities (Effective immediately))

TESTCO, INC. Docket No. 150-00032-EA (Order imposing Civil Monetary (ASLBP No. 96-719-04-EA) l Penalty; General License) (EA 95-101) ,

October 1,1996 l The Atomic Safety and Licensing Board approves a settlement agreement in a consolidated enforcement proceeding.

MEMORANDUM AND ORDER (Approving Settlement Agreement)

%ese two enforcement proceedings involve, respectively, an immediately effective enforcement order seeking to bar Mr. James L. Shelton (a radiographer) from participating in certain NRC-licensed activities for a period of 3 years (measured from October 31, 1995) and a proposed civil penalty of $5000.00 against the firm of which Mr. Shelton serves as President. Atomic Safety and Licensing Boards, consisting of the same Administrative Judges, were established for each proceeding. Those Boards issued Notices of Hearing for 121

each proceeding. 61 fid. Reg. 2848 (Jan. 29,1996) (Shelton proceeding);

61 Fed. Reg. 43,268 (Aug. 21,1996) (Testco proceeding). The proceedings have been consolidated. Prehearing Conference Order, dated August 15,1996 (unpublished); see also 61 Fed. Reg. 43,268.

On September 17, 1996, the NRC Staff advised the Atomic Safety and Licensing Boards that it had reached a settlement with both Testco and Mr.

Shelton. Under the agreement, Mr. Shelton (1) is prohibited from engaging in certain licensed activities until October 31, 1996; (2) must submit certain forms and pay certain fees prict to conducting such licensed activities during the period November 1,1996 through December 31, 1998; (3) until October 31,1998, must provide certain notifications to NRC prior to conducting those licensed activities; and (4) must pay a civil penalty of $1000 in two installments j due no later than October 31,1996. A copy of the agreement is attached hereto. I Pursuant to 10 C.F.R.12.203, settlement agreements such as have been j agreed to here are subject to Licensing Board approval,"according due weight to the position of the [NRC] staff." By motion dated September 17,1996 (delivery of which to one of the Board members was delayed until the week of September 23-27,1996), the Stalf moved that we approve the agreement, which itself recites the Staff's position that the agreement "best serves the interests of the public and the parties," as well as the Atomic Energy Act and NRC requirements, and ,

that we terminate the proceedings. l Absent any contrary information, and according due weight to the Staff's position, we hereby appprove the Settlement Agreement submitted to us and terminate the proceedings.

Pursuant to 10 C.F.R. 5 2.764, this Order is effective immediately but is subject to Commission review under 10 C.F.R. % 2.786.

It is so Ordered.

TIIE ATOMIC SAFETY AND LICENSING BOARDS Charles Bechhoefer, Chairman ADMINISTRATIVE JUDGE Dr. Frank F. Hooper (by CB)

ADMINISTRATIVE JUDGE Dr. Charles N. Kelber ADMINISTRATIVE JUDGE  !

Rockville, Maryland ,

October 1,1996 l I

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE OFFICE OF ENFORCEMENT j l

In the Matter of Docket No.150-00032 (General License (10 C.F.R. 5150.20))

(EA 95-101 and IA 95-055)

TESTCO, INC., and JAMES L SHELTON (Greensboro, North Carolina)

SETTLEMENT AGREEMENT On October 31,1995, the NRC issued a written Notice of Violation and Pro-posed imposition of Civil Penalty - $5,000 (Notice) to Testco, Inc. (Licensee or TESTCO), and an Order Prohibiting Involvement in NRC-Licensed Activi-ties (Effective immediately) to Mr. James Shelton. The Notice and the Order stated the provisions of the NRC's requirements the Licensee had violated and the amount of the civil penalty proposed for the violation.

As a result of the Licensee's failure to adequately respond to the Notice, the NRC ir, sued on March 19, 1996, an Order Imposing Civil Monetary Penalty

- $5,000 By a letter dated July 20, 1996, the Licensee requested a hearing concerning this matter before the Atomic Safety and Licensing Board, and the Board subsequently granted the request.

In telephone discussions on September 5 and 9,1996, between Mr. James Shelton, President of TESTCO, and Mr. James Lieberman, Director, Office of Enforcement Mr. Shelton indicated that TESTCO desires to settle this matter l without further litigation, as noted below. The NRC Staff concludes that this Settlement Agreement best serves the interests of the public and the parties, and the purposes of the Atomic Energy Act of 1954, as amended, and the NRC's requirements.

'Therefore, pursuant to section 81, subsections (b) and (o) of section 161, and section 234 of the Atomic Energy Act of 1954, as amended (42 U.S.C. il 2111, 2201(b),2201(o), and 2282), and 10 C.F.R. 6 2.203, the October 31,1995,and March 19,1996 Orders are hereby modified as follows:

1. Mr. Shelton is prohibited from engaging in licensed activities in areas under NRC jurisdiction until October 31, 1996. For purposes of this Settlement Agreement (Settlement), areas under NRC jurisdiction are 123

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areas in non-Agreement States, offshore waters, or any areas under l exclusive Federal jurisdiction.

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2. Prior to conducting licensed activities in NRC jurisdiction after Novem-ber 1,1996, the Licensee is required to submit an NRC Form 241 that covers the remainder of calendar year 1996 (i.e., until C cember 31, 1996). The Licensee is also required to submit an NRC For.n 241 prior j to conducting licensed activities in calendar year 1997 and calendar l year 1998. These submittals would be in accordance with 10 C.F.R.

I150.20(b); however, the Licensee should be aware that if it performs work in areas under NRC jurisdiction for more than 180 days in any cal-endar year, the Licensee is required to apply for a specific NRC license.

Fees are required to be submitted upon each filing of NRC Form 241 and before commencing work. However, a separate fee is not required

for the weekly notification under paragraph 3 below.

3. Until October 31,1998, following submittals of the yearly NRC Ibrm 241 under paragraph 2 above, Mr. James Shelton, on behalf of Testco, Inc., shall notify NRC Region II, by 9.00 a.m. EST Monday (or Tuesday, i if Monday is a Federal holiday) of each week, whether the Licensee plans to perform radiography work in areas under NRC jurisdiction.

Notification shall be made to the Chief, Materials Licensing / Inspection,

Branch 1, by facsimile at (404) 331-7437 using the attached form, and receipt shall be verified by calling (404) 331-5624.

A. If radiography work is planned, the Licensee shall provide the location of the field sites under NRC jurisdiction where the work is planned that week, as well as the specific date(s) and time (s).

. Inasmuch as the Licensee is required to submit to the NRC written notification on a weekly basis, the provisions of 10 C.F.R. )

i 9150.20(b)(1) requiring that additional NRC Form 241s be filed for j the remainder of each calendar year prior to engaging in licensed i activities are waived; the Licensee is not required to comply with the three day notification requirement as long as it is making the

weekly notifications to NRC Region II.

B. If unplarmed radiography work arises after the weekly notification, the new work cannot be performed unless the NRC has been provided a 24-hour written notification. Telephone notification is

not acceptable.

C. Notification is required to include work on Federal property in Agreement States, unless the Licensee has a written statement from the Federal agency where work is planned that the area is not under i

exclusive Federal jurisdiction.

4. The Licensee agrees to pay a civil penalty of $1,000. %e Licensee shall

, pay $500 within two weeks of the date of this Settlement and $500 no e

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later than October 31,1996. If the.$1,000 Penalty is not paid in full by October 31,1996, TESTCO agrees to pay the full penalty described in the October 31, 1995 Notice ($5,000) by November 30,1996, and waives its right for a hearing concerning the civil penalty imposed by the March 19,1996 Order.

5. The Licensee and Mr. Shelton agree to withdraw their respective requests for hearing in Docket Nos. EA 95-101 and IA 95-055 (now consolidated before an Atomic Safety and Licen. sing Boant) in consideration of the modification of the October 31, 1995 and March 19, 1996 Orders, as provided under paragraphs i through 4 above.

6. If this Settlement is violated, the October 31,1995 and the March 19, 1996 Orders shall be reinstated, and Mr. Shelton and the Licensee agree not to contest the reinstatement of these Orders.

7. The Staff, Mr. Shelton, and 'ESTCO shall jointly move the Atomic Safety and Licensing Board designated in the above-captioned proceed.

s ings for orders approving this Settlement and terminating the proceed-ings.

James Shelton, as an Individual 09/13/96 TESTCO, INC.

,ames Shelton, President 09/13/96 U.S. NUCLEAR REGULATORY COMMISSION James Lieberman, Director 09/16/96 Office of Enforcement 125 I

Fax To: Chief, Materials Licensing / Inspection, Branch i From: James Shelton, President, Testco

Subject:

Notification of Work in Areas Under NRC Jurisdiction For the Week of.J.J

1. Is radiography work planned in non. Agreement States or offshore waters?

(Yes/No)  ;

A. If the answer to Question 1 is yes, skip to 3.

B If the answer to Question I is no, and the work planned is not on a 1 Ibderal property, skip to 6.

C. If the answer to Question I is no, and the work planned is on a Federal property, go to 2.

2. Is there a written statement from the Federal agency stating that the area is not under exclusive lideral jurisdict. ion? (Yes/No)

A. If the answer to Question 2 is no, proceed to 3.

B. If the answer to Question 2 is yes, skip to 6.

3. Date and Time 4. Name and Phone 5. Work Location Address of Planned Work Number of Firm (Street Address, City, and State)

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6. I, THE UNDERSIGNED, llEREBY CERTIFY TilAT:

A. All the information in this form is true and complete.

i B. I have read and understand the provisions of the general license in 10 C.F.R. i 150.20, and understand that I am required to comply with these provisions as well as all byproduct, scurce, or special nuclear material which I possess and use in areas under NRC jurisdiction under the general license for which this form is filed with the Nuclear Regulatory Commission.

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l C. I understand that activities, including storage, conducted in areas  !

under NRC jurisdiction under the general license in 10 C.F.R. l i 150.20 are limited to 180 days in a calendar year.

D. I understand that I may be inspected by the NRC at the above listed work site locations and at the licensee home office address for activities performed in areas under NRC jurisdiction. I am also aware that I am responsible for any fees associated with any inspections.

E. I understand that conduct of any activities not described above, including conduct of activities on dates or locations different from l those described above or without NRC authorization, rnay subject me to enforcement action, including civil or criminal penalties.

! Certifying Officer, Name and Title Signatu e

, Date 127

Cite as 44 NRC 128 (1996) LBP-96-20 l

UNITED STATES OF AMERICA )

NUCLEAR REGULATORY COMMISSION l

l ATOMIC SAFETY AND LICENSING BOARD I 1

Before Administrative Judges:

G. Paul Bollwerk, Ill, Chairman Dr. Charles N. Kelber Dr. David R. Schink in the Matter of Docket No. IA 96-020 I (ASLBP No. 96-715-03-EA)

JUAN GUZMAN (Order Prohibiting Unescorted Access or involvement in NHC-Licensed Activities) October 16,1996 MEMORANDUM AND ORDER (Approving Settlement Agreement and Dismissing Proceeding)

In ajoint motion filed October 4,1996, Petitioners Juan and Laurene Guzman and the NRC Staff ask the Licensing Board to approve an attached settlement agreement and dismiss this proceeding. Finding their settlement accord is consistent with the public interest, we approve the agreement and terminate this case.

At issue in this proceeding is an April 19,1996 Staff enforcement order issued in connection with Mr. Guzman's activities while employed as a contractor employee performing piping insulation work at Baltimore Gas and Electric Company's (BG&E) Calvert Cliffs Nuclear Power Plant, Units I and 2. The immediately effective order precludes Mr. Guzman for a period of 5 years from (1) any involvement in NRC-licensed activities; and (2) obtaining unescorted access to an NRC-licensed facility. The order further provides this 5-year period began on October 18, 1994, the date on which BG&E revoked Mr. Guzman's 128

unescorted access authorization and removed him from the protected area at the

Calvert Cliffs facility for purported misrepresentations regarding his immigration status at that time. As the basis for its order, the Staff relies on Mr. Guzman's j alleged attempts to falsify background information regarding himself, including providing a fraudulent " green card" and social security card and denying that an arrest record obtained by submitting his fingerprints to the Federal Bureau of Investigation belonged to him. See 61 Fed. Reg. 18,630, 18,630-31 (1996).

i In a one-paragraph letter dated April 29,1996, Mr. Guzman and his spouse, Laurene, requested a hearing in accordance with 10 C.F.R. 6 2.202 to contest the Staff's April 1996 order. In its May 31,1996 initial prehearing order the Board sought to convene an early July 1996 prehearing conference, but subsequently granted a series of postponements to provide the Guzmans with additional time to find an attorney.' Their efforts to obtain counsel, however, ultimately were unsuccessful. Accordingly, on August 28, 1996, the Board conducted a prehearing conference during which Mr. Guzman (aided by a United States Department of State-certified Spanish interpreter 2) and Mrs. Guzman appeared pro se.

At the prehearing conference, the Board heard presentations on the pending issues of the Staff's challenge to Mrs. Guzman's standing and the efficacy of the Staff's immediate effectiveness determination.) See Tr. at 9-64. The Board l

also considered the admissibility of certain " central litigation issues" proposed by the parties. We concludea, among other things, that we would permit the enforcement order to be challenged on the ground the 5-year prohibition term is excessive when compared to other, similar cases. See Tr. at 68-70; see also Radiation Oncology Center at Marlton (Marlton, New Jersey), LDP-95-25,42

NRC 237, 238-39 (1995). We also decided we wished to receive additional submissions addressing the question of permitting litigation on the Guzman-proposed issue whether Mr. Guzman's status as a Mexican immigrant was a l

factor affecting the severity of the imposed prohibition. See Tr. at 70-73. Finally, the Board and the participants discussed future scheduling for the proceeding, '

which resulted in a directive that a 60-day discovery period would begin immediately. See Tr. at 74-83. See also Board Order (Memorializing Filing ,

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Because the Guzmans appeared to be in sonw rinanci.kl distress. see. e g. Reply to NRC staff Response Dated July 10.1996 (Aug 2.1996) at 1, and based on our behef that in this enforcement proceedmg the overall

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efficiency of the ad iudicatory process would be malenally aided if the Guzmans had counsel, the Board provided the Guzmans with informanon on organizations that could assist them in obt.umng fier or reduced-cost legal services See Board Memorandum and order (scheduhng Preheanng Conference) (Aug 12.1996) at 3 a 2 (unpubhshed); Board Memorandum and order (second Preheanng order)(June 21.1996) at 4 n.1 (unpubhshed).

2 The terms and conditions govermng the use of that meerpreter were specified in an anachment to an August

26. 1996 Board inuance See Board Mernorandum (Use of spanish Interpretco (Aug 26. 1996) attach. I

(unpubbshed). see afw Tr at 3-6.

3 Because we approve the set 6 ament reached by the parucipants. we need not resolve these issues.

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Dates and Initiation of Discovery and Requesting Settlement Status Report) 1 (Aug. 30,1996) at 1-2 (unpublished).

Following the August 28 prehearing conference, the Guzmans and the Staff initiated settlement discussions. To permit negotiations to continue, on Septem-ber 9,1996, the Guzmans and the Staff asked that we hold the proceeding, including the discovery and issue briefing schedules, in abeyance through the end of September. We granted this request, as well as a September 25, 1996 motion to continue the schedule suspension through mid-October. 'Ihereafter, the participants filed the joint settlement motion now before us.

Under the terms of the October 4,1996 settlement agreement, the Staff agrees to modify the April 1996 enforcement order to reduce from 5 to 3 years the term of the prohibition on Mr. Guzman having any involvement in NRC- i licensed activities or seeking / obtaining unescorted access to any NRC-licensed I facility. Therefore, as revised, this prohibition would be in place until October 17, 1997. In addition, the settlement agreement provides that for a subsequent 2-year period (i.e., October 17,1997, through October 16,1999), if Mr. Guzman seeks employment with any person whose operations he knows, or reasonably l should know, involve NRC-licensed or regulated activity, prior to being hired he must provide that person with a copy of the April 1996 order and the settlement I agreement. In turn, the Guzmans agree to withdraw their hearing request.

Pursuant to subsections (b) and (o) of section 161 of the Atomic Energy Act of 1954,42 U.S.C. 6 220l(b), (o), and 10 C.F.R. 9 2.203, we have reviewed the participants' joint settlement agreement to determine whether approval of the agreement and termination of this proceeding is in the public interest. Based l

on that review, and according due weight to the position of the Staff, we have '

concluded both actions are consonant with the public interest. We thus grant the participants' joint motion to approve the settlement agreement and dismiss this proceeding.

Ihr the foregoing reasons, it is, this sixteenth day of October 1996, OR-DERED that:

1. The October 4,1996 joint motion of Juan and Laurene Guzman and the Staff is granted and we approve their October 4,1996 " Joint Settlement Agree-ment," which is attached to and incorporated by reference in this Memorandum and Order.

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2. 'Ihis proceeding is dismissed.

THE ATOMIC SAFETY AND LICENSING BOARD G. Paul Bollwerk III, Chairman ADMINISTRATIVE JUDGE Charles N. Kelber ADMINISTRATIVE JUDGE David R. Schink ADMINISTRATIVE JUDGE Rockville, Maryland October 16,1996 131 1

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ATTACIIMENT i

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD

' In the Matter of Docket No. lA 96-020 (ASLBP No. 96 715-03-EA) i JUAN GUZMAN (Order Prohibiting Unescorted Access or Involvement in 1

NRC-Licensed Activities) l JOINT SETTLEMENT AGREEMENT

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On April 19, 1996, the staff of the Nuclear Regulatory Commission (Staff) i issued an Order Prohibiting Unescorted Access or Involvement in NRC-licensed Activities (Effective Immediately) to Juan Guzman. 61 Rd. Reg.18,630. On April 29,1996, Juan Guzman along with his spouse, Laurene Guzman, requested a hearing on the April 19,1996 order.' In response to Mr. and Mrs. Guzman's hearing request, an Atomic Safety and Licensing Board was established on May 20,1996. 61 Rd. Reg. 26,549. j After discussions between the Staff and the Guzmans, both the Staff and the Guzmans agree that it is in their respective interests and in the public interest to settle this proceeding without further litigation, and agree to the following terms and conditions:

1. Juan and Laurene Guzman agree to withdraw their request for a hearing, dated April 29,1996.

2. The NRC Staff agrees to the modification of the Order Prohibiting Unescorted Access or involvement in NRC-licensed Activities (Effective immediately), dated April 19,1996, as set forth in Paragraphs 3 and 4, below.

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Mrs Guzman's right to participate in the pro:eedmg was challenged by the Staff, and the issue of her status is pendmg before the Alonus safety and ticensing Board.

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3. Juan Guzman agrees that from October 18, 1994, the date of his termination of unescorted access, until October 17,1997, he is prohibited

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from seeking or obtaining unescorted access at any NRC-licensed facility and may not be involved in any NRC-licensed activities. Ibr the purposes of this agreement, the term, " licensed activities" includes any and all activities which a licensee must or is permitted to perform in order to conduct activities authorized by its NRC-issued license, including those necessary to achieve compliance with all regulatory requirements

! imposed by the Commission.

4. Juan Guzman agrees that for two years following the three year prohi-bition, (that is, from October 17,1997 to October 16, 1999), should he seek employment with any person (meaning an iudividual, a business, or other entity) whose operations he knows or reasonably should know in-volve any NRC-licensed or regulated activity, Mr. Guzman will provide i a copy of the April 19, 1996 order and this agreement to that person j prior to being hired, so that the person is aware of the Order in deciding ,

whether to hire him. l S. By signing this agreement, Mr. Guzman acknowledges his obligation, under federal statute and the Commission's regulations, to provide in- l j formation to the NRC, an NRC licensee, or a contractor of an NRC licensee that is complete and accurate in all material respects. Mr. Guz-man agrees that he will comply with all applicable NRC requirements.

4 6. Mr. Guzman acknowledges that he has read and fully understands the 1

terms of this settlement agreement.

7. The Staff and Juan Guzman shall jointly move the Atomic Safety and

Licensing Board designated in the above-captioned proceeding for an order approving this agreement and terminating this proceeding. Laurene Guzman shall file a notice of withdrawal of her hearing request at the

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same time the motion of the Staff and Mr. Guzman is filed. The terms of this agreement shall become effective upon approval of the Atomic j Safety and Licensing Board.

Juan Guzman Marian L Zobier Counsel for NRC Staff Laurene Guzman Dated this 4th day of October 1996 4

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Cite as 44 NRC 134 (1996) LBP 96 21 i

UNITED STATES OF AMERICA l NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD

Before Administrative Judges

l l Charles Bechhoefer, Chairman l Dr. Richard F. Foster Frederick J. Shon j

in the Matter of Docket No. 50-508-OL l (ASLBP No. 83-486-01-OL)

WASHINGTON PUBLIC POWER SUPPLY SYSTEM (WPPSS Nuclear Project No. 3) October 16,1996  ;

1 The Atomic Safety and Licensing Board grants the Applicant's motion to withdraw its operating license application and to terminate the proceeding.

i NEPA: AGENCY RESPONSIBILITIES He NRC cannot delegate to a local group the responsibility under the Na-tional Environmental Policy Act (NEPA) to prepare an environmental assess-ment (EA). He EA must be prepared by NRC, not a local agency, although in preparing an EA the StalT may take into account site uses proposed by a local agency.

NEPA: ENVIRONMENTAL ASSESS 51ENT, Termination of an operating license application gives rise to a need, pursuant to 10 C.F.R. 5 51.21, for an EA to consider the impacts of the termination.

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NEPA: SCOPE OF ENVIRONMENTAL ANALYSIS Because a construction permit termination would appear to have impacts that encompass operating license termination impacts, one EA would appear to suffice for both actions.

MEMORANDUM AND ORDER  !

(Withdrawal of Application)

L BACKGROUND l

1 This proceeding concerns the application for a reactor operating license for WPPSS Nuclear Project No. 3 filed by Washington Public Power Supply System

(" Applicant"). On August 16,1996, the Applicant filed a Motion for Withdrawal of Application, requesting the issuance of an order authorizing the withdrawal of the Operating License (OL) application and terminating the proceeding.

Attached to this motion was a request to the NRC Staff, dated August 8,1996, to terminate the underlying construction permit (CP).1 On September 5,1996, the NRC Staff filed a response indicating that it had no objection to our granting the motion. None of the other parties responded

- indeed, counsel for the Licensing Board Panel inquired by telephone of the one remaining Intervenor and was apprised that the Intervenor did not intend to respond to the Applicant's motion or to participate in the termination activities.

The State of Washington, participating as an Interested State, also was advised about this license termination, but it did not respond.

IL SITE DEVELOPMENT PROPOSAL

'Ihe Applicant states that it plans to transfer ownership of the entire site (which includes the previously terminated WNP-5 project) to a new interlocal agency, ,

known as the Satsop Adaptive Redevelopment Program ("SRP"), authorized by I a recent change in Washington state law. It states that the WNP-3 project will  ;

- not be completed as a nuclear power plant but that SRP will adapt and use the structures for economic development purposes. The SRP also will have authority

' Earlier, on July 12. 1983, the Apphcant noti 6ed the Aronue Safety and t.icensing Board that construction of l

<tw WNP 3 project would be deferred inde6nitely In a letter dmed May 17.1994 (updated rebruary 15. 1995).

it subsequently advised thas the Apphcant's Board of Directors voted to formally ternunate the project. The Apphcant's Board also voted at that time (1) to maintain the construction pernut (CP)in effect. (2) to consmue a

the deferred status of the oL apphcation, and (3) to preserve the project in accordance with the NRC's "Pohey Statement on Deferred Plants"(52 red keg 38.077 (1987)L

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for site restoration. As a result, the Applicant asserts that there is no basis or need for us to impose conditions on the withdrawal of the OL application or the termination of this proceeding, citing Duke Power Co. (Perkins Nuclear Station, Units 1,2, and 3), LBP-82-81,16 NRC 1128 (1982).

For its part, the Staff indicates that, prior to terminating the CP, it plans to meet with the Applicant and interested state and local agencies and to conduct a Staff site inspection. It will prepare an Environmental Assessment (EA) on the  ;

CP termination based on the meetings and any documentation it may require of i the Applicant pursuant to 10 C.F.R. 5 51.41. It pledges to " assure compliance j with all appropriate safety and environmental requirements" in the context of the CP termination request.

IIL BOARD ANALYSIS The Applicant's solution for treatment of the site -in effect, a delegation of authority to a local agency - would be sufficient only for the type of condition dealt with in the cited Perkins case, where the only issues involved were  !

whether the withdrawal should be with or without prejudice, or reimbursement I oflitigation expenses to the intervening groups. In this case, we cannot delegate to a local group the responsibilities under the National Environmental Policy I Act (NEPA) imposed upon this agency. Moreover, termination of an operating license application gives rise to a need, pursuant to 10 C.F.R. 5 51.21, for an environmental assessment (EA) to consider the impacts of the termination.

Consumers Power Co. (Midland Plant, Units 1 and 2), LBP-86-33,24 NRC 474 (1986); id., LBP-86-39,24 NRC 834 (1986). i An additional consideration here is that the CP termination, although tech-nically a different action than the OL termination before us, would appear to have impacts that would encompass the OL termination impacts. Thus, one EA would appear to suffice for both actions, and the action proposed by the Staff to prepare an EA on the CP termination appears reasonable. He EA must be prepared by NRC, not a local agency, although in preparing an EA the Staff may take into account site uses proposed by a local agency.

Normally, both parties and the Licensing Board would have an opportunity to review the Staff's EA.10 C.F.R. 5 51.lGt(b). We could, therefore, withhold any determination on the Applicant's withdrawal request until the Staff's EA is submitted to us for approval. Midland. LBP-86-39, supra. He parties, however, have expressed no interest in reviewing the termination impacts - indeed, the sole remaining Intervenor expressly declined to do so, and the State, although advised of the opportunity for comment, has not expressed any interest. Further, the Staff is charged with preparing an adequate EA on the CP termination, and from the steps it described it is taking (NRC Staff Response at 2 n.1), we see 136

no likely default in NEPA responsibilities by NRC. That being so, we decline to defer our action on the OL termination request before us pending our review of the EA.

IV. ORDER Accordingly, it is, this 16th day of October 1996, ORDERED:

1. The Applicant's motion for withdrawal of its OL application is hereby l granted; '

2. This proceeding is terminated.

3. Pursuant to 10 C.F.R. 9 2.764, this Order is effective immediately but is i subject to review by the Commission under 10 C.F.R. 6 2.786.

Ti1E ATOMIC SAFLTY AND LICENSING BOARD Charles Bechhoefer, Chairman ADMINISTRATIVE JUDGE Dr. Richard F. Foster (by CB)

ADMINISTRATIVE JUDGE l Frederick J. Shon ADMINISTRATIVE JUDGE Rockville, Maryland October 16,1996 137

Cite as 44 NRC 138 (1996) LBP-96-22 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD l Before Administrative Judges:

Charles Bechhoefer, Chairman Thomas D. Murphy Frederick J. Shon in the Matter of Docket No. 72-18-ISFSI (ASLBP No. 97 720-01-ISFSI)

NORTHERN STATES POWER COMPANY (Independent Spent Fuel Storage installation) October 24,1996 In a proceeding in which a license for an independent dry cask spent fuel storage installation is being sought, the Atomic Safety and Licensing Board describes standards for intervention and establishes dates for amending petitions and for the initial prehearing conference.

RULES OF PRACTICE: PARTICIPATION BY AN INTERESTED STATE OR LOCAL GOVERNNIENT State agencies may choose to participate either as a party under 10 C.F.R.

5 2.714 or as an interested state under 10 C.F.R. 5 2.715(c). To participate under 10 C.F.R. 5 2.714, a sta:e agency must satisfy the same standards as an individual petitioner.

RULES OF PRACTICE: INTERVENTION To participate under 10 C.F.R. 5 2.714, a petitioner must establish its standing, must indicate the aspects of the proceeding in which it seeks to participate, and must proffer at least one acceptable contention.

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RULES OF PRACTICE: STANDING In determining whether a petitioner has the requisite standing, the Commis-sion uses contemporaneous judicial concepts of standing. Under those standards, the petitioner must demonstrate (1) that it has suffered or will likely suffer "in-jury in fact" from the proposed licensing action; (2) that the injury is arguably within the zones of interest sought to be protected by the statute being enforced; and (3) that the injury is redressable by a favorable decision in the proceeding. I 1

l RULES OF PRACTICE: STANDING (GROUP) )

A group may demonstrate that it has suffered or will likely suffer injury in fact either through organizational injury or injury to a member that it represents.

RULES OF PRACTICE: STANDING (INDIAN TRIBES)

Indian Tribes have been permitted to intervene as an entity, without demon- l strating that a particular tribe member has an interest and wishes to be repre-sented by the tribe. They alw have participated in the more routine manner of l

identifying a tribe member who has individual standing but wishes tribe repre-sentation.

MEMORANDUM AND ORDER (Schedules for Further Filings and for Prehearing Conference)

This proceeding involves the application of Northern States Power Company (NSP or Applicant) for a license under 10 C.F.R. Part 72 to possess spent fuel and other radioactive materials associated with spent fuel storage in an offsite independent spent fuel storage installation (ISFSI) in Goodhue County, Minnesota. The license, if granted, would authorize the Applicant to store spent fuel in a dry storage cask system.

Pending before this Atomic Safety and Licensing Board are requests for a hearing and petitions for leave to intervene filed by seven entities (listed chronologically):

1. State of Minnesota Department of Public Service (Petition dated Septem-ber 25,1996);

2. State of Minnesota Environmental Quality Board (Petition dated October 14, 1996);

3. Prairie Island Indian Community (Petition dated October 15, 1996);

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4. He Prairie Island Coalition (Petition dated October 16, 1996);

5. City of Red Wing (Petition dated October 16, 1996);

6. City of Lake City, Minnesota (Petition dated October 17,1996); and

7. Florence Township (Petition dated October 17, 1996).

Both the Staff and Applicanti have filed responses to the Minnesota De-partment of Public Service petition (which was submitted earlier than the other petitions, each of which was timely filed). Both point out that state agencies may choose to participate either as a party under 10 C.F.R. 9 2.714 or as an interested state under 10 C.F.R. 6 2.715(c). To participate under section 2.714, the Com- i mission has long held that a state agency must satisfy the same standards as an individual petitioner. Nuclear fuel Services (West Valley Reprocessing Plant),

ALAB 263,1 NRC 208,216 n.14 (1975); Vermont Yankee Nuclear Power Corp.

(Vermont Yankee Nuclear Power Station), LBP-87-7,25 NRC 116,118 (1987).

NSP and the Staff each point out that the Public Service petition does not satisfy the requirements for participation pursuant to section 2.714 but that the Department of Public Service could qualify as an interested state under section 2.715(c) and could participate under that authority, as long as at least one petitioner is admitted as a party under section 2.714. See Niagara Afohawk Power Corp. (Nine Mile Point Nuclear Station, Unit 2), LBP-83-45,18 NRC 213 (1983). We agree.

t To, participate as a party under section 2.714, a petitioner must establish its standing, must indicate the aspects of the proceeding in which it seeks to participate, and must proffer at least one acceptable contention. The standing requirement stems from section 189a of the Atomic Energy Act, as amentled, 42 U.S.C. s 2239(a), and 10 C.F.R. s 2.714(a)(1), which provide that any person "whose interest may be affected" may seek to intervene and/or request a hearing.

" Person" is defined to include, inter alia. "public or private institution, group, government agency, , , any State or any political subdivision of, or any 1

political entity within a State." 10 C.F.R. 6 2.4. l In determining whether a petitioner has the requisite standing, the Commis-sion utilizes contemporaneous judicial concepts of standing. Sacramento Afu-nicipal Utility District (Rancho Seco Nuclear Generating Station), CLI-92-2,35 NRC 47,56 (1992); Aferropolitan Edison Co. (Three Mile Island Nuclear Sia-tion, Unit 1), CL1-83-25,18 NRC 327,332 (1983). Under those standards, the petitioner must demonstrate (1) that it has suffered or will likely suffer " injury in fact" from the proposed licensing action; (2) that the injury is arguably within the zones of interest sought to be protected by the statute being enforced; and I

NSP's answer was late hied. NSP states that. because of the wording of the federal Regater notice mitiating this proceeding. its counsel did not receive timely nonce of the Department of Pubhc Service petitmn. NsP nu,ves for us to accept its late.hled answer, pomiing out that given the early hhng of the Pubhc service petition, there wdl be no delay in the proceedmg Good cause havmg been shown. we accept NsP's late-hled answet.

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i (3) that the injury is redressable by a favorable decision in the proceeding. Pub- l lic Service Co. of New Hampshire (Seabrook Station, Unit 1), CLI-91-14, 34 l NRC 261,266-67 (1991); Pacific Gas and Electric Co. (Diablo Canyon Nuclear Power Plant, Units 1 and 2), LBP-92-27,36 NRC 196,199 (1992).

] A group may demonstrate that it has suffered or will likely suffer injury in fact either through organizational injury or injury to a member that it represents.

i More than a general statement is required - the means by which injury may be suffered must be demonstrated. Thus, for representational standing, a group

must identify at least one of its members by name and address and demonstrate how that member may be affected (such as by activities on or near the site) and show (preferably by affidavit) that the group is authorized to request a hearing on behalf of the member. Houston Lighting and Power Co. (South Texas Project, Units 1 and 2), ALAB-549,9 NRC 644,646-47 (1979).

Indian tribes, however, have been permitted to intervene as an entity, without demonstrating that a particular tribe member has an interest and wishes to be represented by the tribe. Sequoyah Fuels Corp. and General Atomics (Gore, Oklahoma Site), LBP-9419,40 NRC 9,13-14 (1994). They also have participated in the more routine manner of identifying a tribe member who has individual standing but wishes tribe representation. Umcrco Minerals Corp.,

LBP-94-18,39 NRC 369 (1994). For this proceeding, the Prairie Island Indian j'

Community should supplement its petition (as provided below) with an affidavit, cither (1) from a tribe member with an individual interest who wishes to be represented by the tribe, setting forth a description of how he or she is affected, j such as by residence a certain distance from the facility and how activities

( bearing upon the ISFS! could affect that individual; or (2) from a tribe official

, stating that the tribe wishes to participate as an entity and be represented by the a

tribe attorneys of record, and how the tribe as an entity is affected.

To participate as a party, a petitioner must also submit at least one accept-able contention, conforming to requirements set forth in 10 C.F.R. 5 2.714(b)(2).

Contentions need to be filed at least 15 days before the first prehearing confer-

ence (10 C.F.R. 9 2.714(b)(1)) or by such other date as may be specified by the Board (10 C.F.R.12.711). Petitions may be amended without leave of the Board until that same date (10 C.F.R.12.714(a)(3)).

In this proceeding, the first prehearing conference is hereby scheduled for December 17-19, 1996, in St. Paul, Minnesota, at a time and location to be announced. Members of the public are invited to attend this conference but may not otherwise participate. Petitioners may amend their petitions and submit contentions by Monday, November 25,1996 (service date). Responses to the contentions should be delivered to the Board members no later than close of business on Tuesday, December 10,1996.

During the course of the proceeding, in accordance with 10 C.F.R.12.715(a),

the Licensing Board will entertain written and oral limited appearance statements 4

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i of their positions on the issues from persons who are not parties or petitioners.

These stater L's do not constitute testimony or evidence in this proceeding but

, may help the i>oard and/or parties in their deliberations on the boundaries of the issues to be considered. Oral statements will not be heard at the December 17-4 19 prehearing conference but will be heard at later sessions of the proceeding.

Written statements may be submitted at any time. Written statements, or requests for oral statements, should be submitted to the Secretary, U.S. Nuclear

, Regulatory Commission, Washington, DC 20555, Attn: Docketing and Service Branch. A copy of such a statement or request should also be served on the Chairman of this Atomic Safety and Licensing Board, T3 F23, U.S. Nuclear Regulatory Commission, Washington, DC 20555.

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IT IS SO ORDERED. l J

FOR THE ATOMIC SAIEIT 4

AND LICENSING BOARD l

Charles Bechhoefer, Chairman ADMINISTRATIVE JUDGE i

Rockville, Maryland October 24,1996 l 1

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4 Cite as 44 NRC 143 (19%) LBP-96-23

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' UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION 1

4 ATOMIC SAFETY AND LICENSING BOARD Before Administrative Judges:

G. Paul Bollwerk,111, Chairman Dr. Charles N. Kelber Dr. Peter S. Lam in the Matter of Docket No. 50-219-OLA (ASLBP No. 96-717-02-OLA) i l j GENERAL PUBLIC UTILITIES NUCLEAR CORPORATION

! (Oyster Creek Nuclear Generating 2

Station) October 25,1996 l In this proceeding concerning citizen group challenges to a proposed technical ,

specification change regarding heavy load handling over the Oyster Creek I j Nuclear Generating Station spent fuel pool, the Licensing Board rules (1)

Petitioners Nuclear Information Resource Service (NIRS) and the Oyster Creek i Nuclear Watch (OCNW) have established representational standing as of right; i l

(2) Petitioner Citizens Awareness Network has failed to show either that it is

entitled to standing as of right or that it should be given discretionary standing, but nonetheless will be permitted to participate as an amicus curiae; and (3)

Petitioners NIRS and OCNW have put forth an admissible legal contention regarding validity of the proposed technical specification revision under the agency's " defense-in-depth" policy.

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ATOhllC ENERGY Acr: IIEARING RIGIIT OPERATING LICENSE A51ENDNIENTS: TECHNICAL SPECIFICATION CHANGES A technical specification is a license condition, and a licensee request to change that condition constitutes a request to amend the license that creates ad-judicatory hearing rights under Atomic Energy Act i 189a,42 U.S.C. 5 2239(a).

See Cleveland Electric illuminating Co. (Perry Nuclear Power Plant, Unit 1),

CLI-93-21,38 NRC 87,91 n.6,93 (1993).

ATONilC ENERGY ACT: STANDING TO INTERVENE (REPRESENTATIONAL)

RULES OF PRACTICE: STANDING TO INTERVENE (REPRESENTATIONAL)

To have standing to participate as of right in a proceeding regarding an agency licensing action, a petitioner must demonstrate that (1) it has suffered or will suffer a distinct and palpable injury that constitutes injury in fact within the zone of interests arguably protected by the governing statute; (2) the injury is fairly traceable to the challenged action; and (3) the injury is likely to be redressed by a favorable decision. In addition, when an organization seeks to intervene on behalf of its members, that entity must show it has an individual member who i can fulfill all the necessary elements and who has authorized the organization to represent his or her interests. See Yankee Atomic Electric Co. (Yankee Nuclear Power Station), CLI-96-1,43 NRC 1,6 (1996).

RULES OF PRACTICE: STANDING TO INTERVENE (CONSTRUCTION OF PETITION) l J

In making a standing determination, a presiding officer is to " construe the

[ intervention] petition in favor of the petitioner." Georgia Institute ofTechnology (Georgia Tech Research Reactor, Atlanta, Georgia), CLI-95-12,42 NRC 111, i15 (1995).

ATO511C ENERGY ACT: STANDING TO INTERVENE (INJURY IN FACT)

RULES OF PRACTICE: STANDING TO INTERVENE (INJURY IN FACT)

Relative to a threshold standing determination, even minor radiological exposures resulting from a proposed licensee activity can be enough to create i 144 i

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the requisite injury in fact. See Yankee Atomic Electric Co. (Yankee Nuclear Power Station), LBP-96-2,43 NRC 61,70, aff'd, CLI-96-7,43 NRC 235,246-48 (1996).

RULES OF PRACTICE: STANDING TO INTERVENE (AUTilORIZATION)

If individuals relied upon to establish representational standing for an organi-zation fail to indicate they are members of that organization, their proximity to the facility cannot be used as a basis for representational standing. See Florida Power and Light Co. (Turkey Point Nuclear Generating Plant, Units 3 and 4),

ALAB-952,33 NRC 521,530-31 (representational standing not present when individual relied on for standing is not organization member, but only represen-tative of another organization), aff'd, CLI-91-13,34 NRC 185 (1991).

ATO511C ENERGY ACT: STANDING TO INTERVENE (INJURY IN FACT)

RULES OF PRACTICE: STANDING TO INTERVENE (INJURY IN FACT)

Concern that " bad precedent" may be set in proceeding that could impact the petitioner's ability to contest similar matters in another proceeding is

" generalized grievance" that is "too academic" to provide the requisite injury in fact needed for standing as of right. See Ohio Edison Co. (Perry Nuclear Power Plant, Unit 1), LBP-91-38, 34 NRC 229, 248-49 (1991), aff'd as to another ruling, CLI-92-l1, 36 NRC 47 (1992), petition for review dismissed, City of Cleveland v. NRC, 68 F.3d 1361 (D.C. Cir.1995).

RULES OF PRACTICE: INTERVENTION (DISCRETIONARY)

Under the six-factor test for discretionary intervention, a primary considera-tion is the first factor of assistance in developing a sound record. See Portland General Electric Co. (Pebble Springs Nuclear Plant, Units 1 and 2), CLI-76-27, 4 NRC 610,617 (1976).

RULES OF PRACTICE: AS11CUS CURIAE Although requests for amicus curiae participation do not often arise in the context of Licensing Board hearings - in which factual questions generally predominate - because an amicus customarily does not present witnesses or cross-examine other parties' witnesses, this happenstance "does not perforce 145

i preclude the granting of leave in appropriate circumstances to file briefs or

memoranda amicus curiae (or to present oral argument) on issues of law or fact j that still remain for Licensing Board consideration." Public Service Co. of New Hampshire (Scabrook Station, Units I and 2), ALAB-862, 25 NRC 144,150 4

(1987). Thus, in the context of a proceeding in which a legal issue predominates, permitting a petitioner that lacks standing to file an amicus pleading addressing that issue is entirely appropriate.

RULES OF PRACTICE: CONTENTIONS (SCOPE; SPECIFICITY AND BASIS)

Particularly in the context of dealing with pro se petitioners, a finding regard-1 ing a contention's specificity should include consideration of the contention's bases. See Public Service Co. ofNew Hampshire (Seabrook Station, Units 1 and 2), ALAB-899,28 NRC 93,97 (1988)(both contention and stated bases should be considered when question arises regarding admissibility of contention).

RULES OF PRACTICE: CONTENTIONS (POSSIBLE FAILURE TO CO51 PLY WITII REGULATORY REQUIRE 51ENT)

If clear regulatory constraint mandates that a licensee take (or not take) a

particular action, to gain the admission of a contention founded on the premise l the licensee will not follow that requirement, a petitioner must make some t particularized demonstration that there is a reasonable basis to believe the licensee would act contrary to the explicit terms of that regulatory requirement.

I RULES OF PRACTICE: DISCOVERY (SU5151ARY DISPOSITION);

SU5151ARY DISPOSITION (DISCOVERY)

In responding to a surc'iary disposition motion, a party can assert, with appropriate supporting affidavits, that it needs discovery to answer the dispositive

. motion. See Public Service Co. of New Hampshire (Seabrook Station, Units I and 2), CL1-92-8,35 NRC 145,152 (1992).

MEMORANDUM AND ORDER (Ruling on Intervention Petition)

In a Federal Register notice published May 8,1996, the NRC Staff announced (1) a proposed "no significant hazards consideration" finding regarding an April  ;

15, 1996 request by Licensee General Public Utilities Nuclear Corporation l l

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(GPUN) to revise Technical Specification 5.3.1.B for the Oyster Creek Nuclear l Generating Station (OCNGS); and (2) an opportunity for a hearing on that l

GPUN license amendment application. See 61 fid. Reg. 20,842, 20,842-43, 20,848 (1996). Acting on the latter offering, on June 6,1996, pro s Petitioners Nuclear Information and Resource Service (NIRS), Oyster Creek Nuclear Watch (OCNW), and the Citizens Awareness Network (CAN) filed i a timely hearing request and petition to intervene seeking to challenge the l

proposed technical specification change. In response, both the Licensee and the Staff have challenged the sufficiency of the Petitioners

• hearing request, l asserting they lack standing and have not presented an admissible contention. I For the reasons set forth below, we find (1) Petitioners NIRS and OCNW have established their standing as of right; (2) petitioner CAN has failed to establish it is entitled to standing as of right or to show it should be afforded discretionary standing, but will be permitted to participate as an amicus curiae; and (3)

Petitioners NIRS and OCNW have submitted a litigable contention. Accordingly, we grant the intervention petition as it relates to NIRS and OCNW and admit them as parties to this proceeding. In addition, because the admitted contention involves a legal question, we establish a schedule for summary disposition filings to resolve that issue.

L llACKGROUND A. Technical Specification 5.3.1.11 and the GPUN Spent Fuel OIT Load ,

Program In its present form, under the headings of "AUXtUARY EGUIPAfENT' and

" Fuel Storage," OCNGS Technical Specification 5.3.1.B states that "[Iloads greater than [the] weight of one fuel assembly shall not be moved over stored irradiated fuel in the spent fuel storage facility." NRC Staff Response in Oppo-sition to Request for Hearing and Petition to Intervene of [NIRS/OCNW/CAN)

(June 26,1996) unnumbered attach. 2 (OCNGS Technical Specification p. 5.3-1 (Apr.10,1995))[ hereinafter Staff Hearing Request Response].' The amendment proposed by GPUN would take this provision, make it the first of two subparts, and provide for additional language so that the subparts would read:

1. Loads greater than the weight of one fuel assernbly shall not be rnoved over stored irradiated fuel in the spent fuel storage facility. except as noted in 5.3.1 A2.

I lu the Board's initial preheanng order, to snake it easier to locate and reference record doeunrnts, we asked that for all 6hngs the participants provide N neparate alpha or nurnene desigrunon for each appended docunrnt te g , Exhibit 1; Attachrnent A) , , See Board Menwrandum and order Onitial Prehearing Order) Oune 18.

199t0 at 4 (unpubhshed) [hereinalter Board initial Orderl We espect the parties to comply wuh this requirernens for any addiuonal hhngs in this proceeding 147

(,

2. He shield plug and associated htting hardware may be rreved over irradiated fuel assemblies that are in a dry shicided canister within the transfer cask in the cask drop protecuon systern.

Id. unnumbered attach.1 (Letter from Michael B. Roche, Vice President and Director, OCNGS, to NRC Document Control Dest (Apr. 15,1996) at unnumbefed p. 6 (proposed revised OCNGS Technical Specification p. 5.3-1)).

In its contemplated "no significant huards consideration" finding,2 the Staff explains that this proposed change is designed to " facilitate the offload of spent fuel to the Oyster Creek Independent Spent Fuel Storage Installation (ISFSI)."2 61 lied. Reg. at 20,848. As described in more detail to the Licensing Board in a background presentation made by the Licensee during an August 7,1996 prehearing conference, see Tr. at 19-37,4 the amendment request concerns a single step in the Licensee's overall plan for moving the spent fuel currently in the OCNGS spent fuel pool into dry cask storage at the facility ISFSI to await ultimate disposal.

6 The NUllOMS dfy canister storage system to be used at OCNGS has three main components: a 14-ton dry shielded canister (DSC); a 60-ton onsite transfer cask (TC); and a horizontr.1 storage module (IISM).5 The DSC is a stainless steel cylindrical vessel that can hold up to fifty-two spent fuel assemblies, each of which weighs 800 pounds. The TC, a steel and lead-lined cylinder, holds a DSC as the DSC is being loaded with spent fuel assemblies in the OCNGS spent fuel pool and then transported on a trailer between the a reactor building, where the spent fuel pool is located, and an IISM. The 11SMs for the OCNGS ISFSI are located just beside the plant in a separate, secured area.

An llSM is a reinfosted concrete unit consisting of a base mat, four walls, and a roof. Each of the ten liSMs currently at the OCNGS ISFSI holds a single, loaded DSC. A hydraulic rw pushes a loaded DSC from the TC into an 2

in accordance wuh section 189a(l)( A) of the Atonue Fnergy Act (AEA). 42 U S C. 8 2239(a)(1)(A). and to C 0 R.18 50 91 92. if adopted, the Staff's *no sigtuheant hazards consideration" hnding would pernut the Staff to issue the GPUN-requested technical specihcation change while this adjuscatury procceeng is peneng As far as the Board is aware, the Staff has not yet male that fin &ng 3

As dehned in the aget#s regulations, an independent spent fuel storage installatmn. or (SISI. as "a comples designed and consuucted for the intenm storage of spent nuclear fuel and other rahoact>ve matenals associated with spent fuel storage " 10 C F R.172.3 Under 10 C F R. Part 72. Subpart K, an agency-adopted general heense permits a reactor heensee to store spent fuel at a reactor-site ISFSI so long as the heensee uses a cask storage system approved by the agency "See mLm letter from Ann P. Hodgdon. NRC Staff Counsel, to the t icensmg Board (Aug 5.1996) (hereinafter Hodgdon txtter) unnumbered stach I. at 9.14 to -9 (oCNGS lhnal Safety Analysis Report (FSAR)) Update (Update 7 Dec.1992)); ed. unnumbered attach. 2. enct 3. at I-4 to -16 (U.i Nuclear Regulatory Comnussion.

Office of Nuclear Material Safety and Safeguards. Safety Evaluation Repon of tVectra Technologies. Inc 1 Safety Analysis Report for the Standar& zed NUHoMS Honionaal Matular Storage System for Irradmied Nuclear hiel (Dec 1994)).

8 The NJHoMd system is among the agency. approved cask storage systems. See 10 C F R. I 72 214 (Certificate No.10Nt 148 I

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IISM horizontally through an opening in the llSM. Inside the llSM, the DSC sits above the base mat on a steel frame support structure. Once the DSC is inside the llSM, the IISM opening is sealed with a reinforced concrete and steel door. Thereaften. spent fuel decay heat cooling occurs by means of a natural convection air flow system.

To get the fuel assemblics in the spent fuel pool into a DSC for transfer to an llSM, the Licensee first moves an empty DSC onto the ground floor of the reactor building and lifts the DSC up the equipment hatch opening approximately 1

100 feet to the third floor refueling deck. An empty TC is then placed at the foot of the equipment hatch opening on the ground floor. The DSC is lowered back down the equipment hatch opening into the TC, and this combined DSCRC assembly is raised back up to the refueling deck. The DSC and the annulus between the DSC and the TC then are filled with water, and the DSCRC assembly is lowered into the spent fuel pool.

i To prevent serious damage to the spent fuel pool during this last process, the Licensee has developed a cask drop protection system (CDPS). This system, which was permanently installed in the early 1970s, consists of a tapered 1

cylindrical stainless steel structure that has been attached to the sides of one

, corner of the OCNGS spent fuel pool. This cylinder, which also is filled with water, is intended to guide the DSCHC assembly and, if necessary, restrain a q falling DSCffC assembly as it is placed into the pool to await the insertion

, of the fuel assemblies into the DSC. Also, to help provide a cushion, a 2%.

inch-thick aluminum alloy base plate is attached to the bottom of each TC. If a DSCRC assembly were dropped, this base plate is intended to act as a piston y and attenuate any forces generated by water displacement and guide cylinder wall impacts.

He CDpS guide cylinder itself consists of two parts, a lower dashpot cylinder and an upper guide cylinder. The bottom and sides of the lower dashpot cylinder

have energy absorption capability to prevent damage to the spent fuel pool j bottom and walls f:om any DSC/TC assembly impacts with the guide cylinder.

I The upper guide portion of the CDPS guide cylinder has a hinged gate that can be opened to permit fuel assemblies to be loaded into the DSCRC assembly as it sits in the lower dashpot cylinder, thereby allowing both the DSCRC assembly and the fuel assemblies to remain under water in the fuel pool during the entire loading process. The CDPS also has a 1 inch-thick stainless steel top plate t

cover extending over the guide cylinder, with a hole for inserting the DSCRC assembly into the guide cylinder that is some 10 inches wider than the diameter of a DSCRC assembly with its base plate attached.

After the DSC is loaded with spent fuel assemblies, the shield plug is set

, on top of the DSC to close it. The shield plug is a 4-ton metallic disc about

514 feet in diameter and 8 inches thick. The shield plug is lowered by crane onto the loaded DSC inside the CDPS while attached to a 3-ton yoke by four i 149

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cables connected to four eyebolts imbedded in the shield plug top. The DSCfrC assembly is then removed from the CDPS by crane and the DSC is scaled on the l top with additional protective layers. The water is removed from the DSCfrC i assembly, inert gas is inserted, the TC is sealed, and the DSCfrC assembly is i taken from the reactor building and transported by trailer to the ISFSI, where j the scaled DSC is placed horizontally into an HSM, as described above. ,

The particular change in Technical Specification 5.3.1.B proposed by GPUN i would permit the shield plug - which weighs considerably more than the single I fuel assembly that now defines the load limit permitted to be moved over spent fuel- to be placed over the spent fuel assemblies in the DSC while the plug is being lowered into place.

B. NIRS/OUNW/CAN Intervention Petition and Contention l

In contesting this GPUN license amendment,' Petitioners NIRS and OCNW asserted in their June 6,1996 hearing request and intervention petition that they had fulfilled the requirements for both intervention as of right because of the proxirnity of their members to the facility, while CAN declared its standing was based on the potential injury it, New England-based membership would suffer from any " bad precedent" that might come from this proceeding. All three Petitioners argued they met the standards governing discretionary intervention as well. They further declared the " aspects" of the proposed technical specification about which they are concerned are the possibility of (1) a significant increase in accident probabilities; (2) an accident not previously identified in the Licensee's Safety Analysis Report for OCNGS; and (3) a significant reduction in operating ,

boiling water reactor (BWR) safety margins. They maintained these concerns are based on (1) NRC Bulletin 96-02, " Movement of Heavy Loads Over Spent i Fuel, Over Fuel in the Reactor Core, or Over Safety-Related Equipment"(Apr.

I1,1996); (2) NRC Information Notice 96-26, "Recent Problems with Overhead Cranes" (Apr. 30,1996); (3) a May 8,1996 NRC Daily Event Report (DER) about a 5000-pound transportation cask that was dropped on the fuel handling floor at Indian Point Unit 2 while being lifted by a crane; and (4) a December 30,1994 Preliminary Notice of Event or Unusual Occurrence (PNO II-94-055),

regarding the drop of a 350-pound core shroud head bolt over the spent fuel pool j at Georgia Power Company's Edwin Hatch Unit I that caused a 3-inch gash in the fuel pool liner and an accompanying 2000-gallon water leak that lowered the pool level by 2 inches. See [NIRS/OCNW/CAN] Request for a Hearing and

' A technical specincanon is a bcense condmon, and a heensee request to change that condiuon constitures a request to unrnd the hcense that creates adi udicatory heanng nghts under AEA secuan 189a. 42 U.S C i 2139ta).

See Cleveland Elermc illummarmt Ca (Perry Nuclear Power Plant. Unir I), Ct.1-93-21,38 NRC 87,98 n 6. 93 (1991) 150 l

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1 Petition to Intervene on [GPUN] License Amendment Request for [OCNGS] l (June 6,1996) at unnumbered pp. 2-8 [ hereinafter Intervention Petition].  :

i Both the Licensee and the Staff answered the Petitioners' hearing request.

Licensee GPUN asserted each petitioner had failed to establish its standing to intervene either as a matter of right or discretion. See GPUN's Answer Opposing Request for Hearing and Petition for Intervention of [NIRS/OCNW/CANJ (June 21,1996) at 9-18 [ hereinafter GPUN Answer]. The Staff took the position that while NIRS and OCNW had established some of their members lived or

had activities in proximity to the facility, these Petitioners had failed to show J those members would suffer any injury as a result of the proposed amendment.

The Staff also asserted that CAN had failed to establish its standing as of right and that all three Petitioners had failed to show they should be afforded discretionary standing. Purther, on the matter of the aspects of the proceeding, the Staff declared the Petitioners' aspects were not related to the subject matter of the proposed amendment, criticizing in particular the relevance of the four documents referenced by the Petitioners in support of their hearing request.

See Staff Hearing Request Response at 5-13. The Licensee, on the other hand, declared it would address the Petitioners' aspects when responding to

the Petitioners' specific contentions. See GPUN Answer at 18.

, Acting pursuant to a Board directive, on July 18, 1996, the Petitioners  ;

i filed a supplemental intervention petition in which they set forth the following 1 contention: 1 1 '

The GPUN apphcation fails to provide defense-in depth agamst the nsks of a heavy load drop

onto irradiated fuel and fails to satisfy NRC regullory guidance as provided in NUREG-

0612 " Control of IIcavy Loads At Nuclear Power Plants" pertaming to defense-in depth nsk j management to assure that a heavy load drop does not impact or encroach on irradiated fuel.  !

i Supplemental Petition of [NIRS/OCNW/CAN) (July 18,1996) at 2. As the l bases for this contention, the Petitioners made several assertions that can be summarized as follows:

A. Under 10 C F R. I 50 36(cKl). GPUN is legally required to estabhsh and maintain i safety hnuts governmg activities potennally affecting fuel rod claddmg and fuel I pool liner integnty. Technical Specification 5 31.11 is designed to establish the speci6ed safety hmits.  !

j lt As is estabhshed by a June 16. 1995 DER tReportable Event No. 28954) and a l libruary 6,1987 Licensee Event Report (LER) (LER No. 86-016-01), there are j potentially degraded fuel assembbes in the OCNGS spent fuel pool. Because there

is no assurarice that such assembhes will not be placed in a DSC, the proposed

] Technical Specification change would intrmluce an unanalped threat in the event of a shield plug drop.

C. The NRC's fundamental regulatory defense in depth principle is implo.nented in j NUREG 0612 " Control of Heavy Loads at Nuclear Power Plants," which is the 1 equivalent of a regulatory guide. Because OCNGS does not employ a single failure i 1

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, l proof crane for shield plug movement, consistent with NUREG-0612 guidehnes as I described in enclosure 1 to NRC Gerwnc Letter 85-11 Oune 28,1985), GPUN l must rely on analyr.ed safe load paths and restricted load hrruts for snovement of heavy loads "to assure, to the extent practical" that heavy loads are not camed over j or near irradiated fuel Although GPUN claims in its safety evaluation regardmg l the proposed technical specification change that a shield plug drop accident is l not credible because of GPUN adrninistrative controls (e g., rail stops), operator l training, and inspections concerning dry storage related spent fuel movements, this

[ does not adequately address human error of mechanical / electrical failure issues.

Rather, the most etTective way to avoid such failures is to restrict both human.

l directed activity ad prohibit tlw movement of heavy loads as is done with current Technical Specification 5 3.I.B. As such, consistent with the agency's NUREG-0612 defense-in-depth guidance, the existing provision cannot be revised as the Licensee has requested.

See id. at 2 6.'

In its July 29,1996 answer to the Petitioners' supplemental petition, GPUN declared both their contention and the bases put forth in support of that contention are too vague and fail to establish a genuine dispute as to a material issue of fact or law, According to GPUN, the Petitioners' reliance on NUREG-0612 is misplaced because they fail to recognize that document's admonition to assure heavy loads are not carried over spent fuel "to the extent practical."

I ha a July IN reply to the June 1996 GPUN and Staff answers to their itunalinservenuon peution among other ihmgs, the Peutioners asacrted in connectmn with d e GPUN answer thas (1) nutmthstanding GPUN's muertson that trane capacity enceeds the weight of the shield plug and hfting yoke, because that contined weight is manv tmra the weight of a fuel assembly - the hnunng weight under alw esasung technical spec 6 canon - a drop on a fully loaded DSC could cause sigm6 cant damage; (2) given that docunrnts. such as NUREG/CR-4982, " Severe Accidents in Spent itel Pools in Support of Generic Safety issue 82" Ouly 1987), estabbsh the consequences of an accident involvmg a breach of the sperd fuel pool liner and a rapid coohng water drain down are senous, the reduction in safery insgms involved in the tech $ucal speci6 canon change does involve a threat of palpable injury and a nsk to public health and safety;(3) a Nwember 27,1992 report of a substantial safety hazard 61ed  ;

under 10 C F ll Part 21 for the Susquehanna Stearn Liectric Stanon mdicates that fuel pool cochng capahihty j loss fror, a dra n down can cause the failure of other safety-related reactor operations equipment. (4) the increase I in human-directed activity and load weight involved in the spene fuel off. load acovity constitutes an nerease in the nsk of human error asul nwcharucal and/or electrical failure of load beanng eympment thas jeopardues the pubhc health and safety;(5) the damage associated with a shseld plug drop about wluch they a concerned is (a) ,

damage to spent fuelin the DSC with the pmential for recnucahty, and tb) damage to tie spent fuel pool hner with potential drain down affecting oder fuel in storage racks; and (6) dw firs! hne of defense for enticahey prevennon strategy at the OCNGS spent fuel pool is the human-directed mechanical acovity and weight hnut restncuons l imposed m the current Techmcal Speci6 cation 5 318 See Pensioners' Reply to NRC Staff and [GPUNI Answer i Opposing Request for Heanng and Pention for intervenuon of [NIRS/oCNW/CAN) Ouly 18,1996) at 2-9 Concernmg the Statf's Imswer, the Petmoners declared (I) dwar reference to NRC Bulletm 96 02, %enent of Heavy toads Over Spent luel. Over fuel in the Reactor Cnre, or over Safety-Related lituipnwnt? gnes appropnate background documentauon; (2) their rebance on NRC Infornmuon Nonce 96-26, "Recent Problems with Overhead Cranes? is appropriate because the increased nsks associated with activines over or near irradiac., fuel arising from potential crane equipment detenoranon or madequate crane eqmpment combined with inappropriate Licensee seuvities as authned in that document are relevant at oCNGS. one of the oldest Anrrican operaung reactors; (3) their rehance on the Hatch "boli drop" prehnunary nonce is apprepnate because of their concern about the possibihty of a similar fuel pool kner tear associated with a shield plug drop accident, regardless of whether the bolt weighs less than a shield plug; and (4) their rehance on the Indian Point 2 DER regardmg the drop of a 50topound netal transport container on the fuel handhng floor is appropnate because it underscores their concern that heavy load accidents can happen See ad at 9-12.

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This qualifier, GPUN asserted, nullifies the Petitioners' apparent position that  !

pennitting any load heavier than a fuel assembly to be carried over spent fuel

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will violate " defense-in-depth" principles. According to the Licensee, the only i time the shield plug is over spent fuel is when it is lowered onto the top of the l loaded DSC, a step that cannot be avoided if the spent fuel is to be properly shielded as is required by other NRC regulatory requirements. Thus, consistent with NUREG-0612, GPUN has acted to limit the movement of heavy loads over spent fuel "to the extent practical." See GPUN's Answer to Supplemental Petition of lNIRS/OCNW/CANj (July 29,1996) at 5-9.

This being the case, the Licensee assened the focus must be on the actions l

it has taken to assure heavy load lifts satisfy the preventative measures outlined in NUREG-0612, which include the use of safe load travel paths, mechanical stops to prevent crane travel outside the analyzed load paths, and use of detailed operating procedures and training. According to GPUN, its steps in these areas J have not been contested by the Petitioners. See id. at 8. l As to the Petitioners' concern about the movement of heavy loads over l degraded fuel, GPUN declared that the 1987 LER and the 1995 DER relied upon by the Petitioners provide no support for their general assertion there are )

an " undetermined" number of degraded fuel assemblies that may be loaded {

into the DSC. According to GPUN, the LER and the DER, in fact, establish  ;

only that a specific number of fuel elements - forty-seven - were damaged as a result of a specific problem with fuel pellet / clad interaction and cr., was damaged as a result of structural failure during movement. GPUN further stated l that damaged fuel assemblies have no relevance to this proceeding because the 1 certificate of compliance issued by the NRC for the NUllOMS storage system precludes damaged or unchanneled fuel assemblies from being loaded into the l DSC. See id. at 9-11. ' l Finally, regarding a possible fuel pool liner breach from a shield plug drop, GPUN asserted this concern does not deserve further scrutiny because the Petitioners have not identified the failure mechanism that would make such a drop possible or the scenario under which such a drop would impinge on the fuel pool liner. See id. at 12-15.

In its response to the supplemental petition, the Staff maintained the Peti-tioners' contention lacks specificity as to the alleged failures in the GPUN ap-plication. The Petitioners' reference to 10 C.F.R. 6 50.36(c)(1) as it sets " safety limits" is misplaced, according to the Staff, because the technical specyication in question is a " design feature," not a " safety limit." The Staff asserted the ap-propriate regulatory reference is to section 50.36(c)(4). According to the Staff, this provision covers " design features" in technical specifications, which are those features of the facility such as construction materials and geometric ar-rangements that, if altered or modified, would have a significant effect on safety and are not covered under section 50.36(c)(1)-(3) as they relate to " safety limits" 153

like limiting safety system settings, limiting control settings, limiting conditions for operation, and surveillance requirements. See NRC Staff Response to Peti-tioners' Supplemental Petition (July 31,1996) at 7.

The Staff also asserted the Petitioners

• reliance upon NUREG-0612 as providing " regulatory guidance" is misplaced because that document is not a regulation or a Staff regulatory guide. The Staff further declared the Petitioners' reliance on NUREG-0612 as a basis for contending there can be no change in the load limit set in current Technical Specification 5.3.1.B is misdirected because that NUREG does not prohibit the movement of heavy loads, but deals only with the control of movement of such loads. The Staff also responded to the Petitioners

• alleged concern about degraded fuel by reference to the NUHOMS

, certificate of compliance that precludes using a DSC to store fuel with known or suspected gross cladding breaches. Finally, the Staff declared the Licensee's CDPS makes any shield plug drop on the pool liner a matter of speculation.

See id. at 8-12.

On August 7,1996, the Board conducted a prehearing conference during which NIRS, GPUN, and the Staff had an opportunity to address further the questions of NIRS standing and the admissibility of the Petitioners' joint contention." As part of his presentation, the representative for Petitioner NIRS read into the record a statement in support of the Petitioners' contention that addressed a number of the GPUN and Staff objections. See Tr. at 66-76. Among other things, this NIRS statement made reference to three additional documents:

an April 30,1986 Staff memorandum on budget cut impacts that is asserted to provide a factual basis for the unpredictable nature of human error; the July 19, +

1996 Oyster Creek Performance Review in which the Staff finds there have been

" avoidable personnel errors" at the facility, particularly in the areas of operations and maintenance; and a July 20,1995 GPUN reply to a 1995 NRC inspection report (No. 50-219/95-09 , in which the Licensee concurs in a self-identified technical specification violation involving a failure to follow a requirement to have a licensed senior reactor operator or a senior reactor operator limited to fuel handling supervise core alterations. According to NIRS, these documents show that "the issue of human error provides support for the contention that it is indeed not practical to modify and reduce a current technical specification designed to preclude human error and/or mechanical failure from dropping a heavy load aThe August 7 preheanng confercuce was nonced in early July See Board order (scheduhng Films Deadhir for Supplemental Intervennon Petmons and Responses and for Preheanng Conference) (July 3,19%) at 2 (unpubbshed) The Board, however, was informed for the first tmr al the preheanng conference that tir designatej representatives of oCNW and CAN would mW attend the August 7 prnceeding. See Tr at 7 8. ter alm Board Memorandum tIbrwarding Documents for DocLeung and Requesung Settlenrnt status Report)(Aug.14.19%)

attachs 1-2 (unpubbshed) Actmg on the monon of the Licensee, the Board ruled that while it would not disnuss oCNW and CAN for their failure to parncipate in the conference. the NIRS representanve would sme be pernuited to make anv presentanon on the issue of oCNW's or CAN's standing in mtervene. See Tr at 9-14 154 i

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onto irradiated fuel without undermining the Defense-In-Depth Philosophy as

! established in NUREG-0612." Tr. at 70-71.

! During the prehearing conference, NIRS also sought to counter the GPUN and l ~ Staff responses to the Petitioners' supplemental petition. Besides declaring that a shield plug drop accident was a credihte event that constituted an unanalyzed

condition, NIRS asserted GPUN had not answered the Petitioners' concerns about degraded fuel assemblies because it had not provided infortnation about how the utility plans to screen the fuel for deteriorated bundles or about the consequences for criticality and shielding if such fuel bundles are involved in ,

3 a heavy load drop accident. NIRS did state the Petitioners were willing to I

concede a spent fuel pool drain down resulting from liner damage from a shield l l plug drop was an unlikely event, but asserted the GPUN safety evaluation for l the requested amendment still was insufficient because it did not adequately l l address the consequences during a shield plug lift of either a power loss to the crane drive motor or a seismic event _ See Tr. at 72-76.

In response to NIRS's expressed concern about the lack of any GPUN analysis of the consequences of a shield plug drop onto the fuel assemblies in a DSC, see Tr. at 82, GPUN made reference to analyses it had made of several " worst case" scenarios relative to a possible shield plug drop. Although maintaining that the possibility of such a drop was incredible, GPUN noted that it had analyzed the potential for recriticality if, by whatever means, all fifty two fuel assemblies in a DSC were damaged so that all the fuel is crushed t@cther in the worst possible configuration in the bottom of the canister, thereby  ;

maximizing the potential for recriticality. GPUN concluded that even under this i scenario, the potential of recriticality was very low (0.957). See Tr. at 85 '

86. In addition, GPUN analyzed the possible radiological consequences that j could result from a shield plug drop given the geometrical configuration of the canister opening and the size and shape of the shield plug. GPUN detennined that the maximum damage would accrue if the plug landed vetiically on the cask mouth, impacting sixteen of the fifty-two fuel bundles in a fully loaded cask, with a resulting potential maximum release of 6.25 millirem at the facility site boundary. See Tr. at 92-94.

These analyses, which had not been given to the Petitioners, subsequently became the subject of unsuccessful settlement negotiations. See Petitioners Communication to the Honorable G. Paul Bollwerk, Esq., Dr. Peter Lam, and Dr.

Charles Kelber Regarding Settlement with GPUN (Aug. 16, 1996). Ultimately, these analyses came into the Petitioners' hands as a result of Staff action to obtain them. See Letter from Ernest L Blake, GPUN Counsel, to the Licensing Board at I (Aug. 27, 1996). Thereafter, in a September 9,1996 pleading, commenting on the analyses, the Petitioners asserted that the expressed premise in the recriticality analysis that a drop accident would not damage the TC containing the DSC lacked justification and that the radiological consequences 155

analysis failed to address the question of occupational doses to facility workers.

See Petitioners Status Report to the lionorable G. Paul Bollwerk, !!I, Dr Peter Lam, and Dr. Charles Kelber Regarding GPUN Letter of August 23,1996 (Sept.

9,1996) at 1-2 [ hereinafter Petitioners Status Reportl.

In a September !I reply, the Licensee asserted the undamaged cask assump-tion for its nonmechanistic criticality analysis clearly was justified given the i 4-inch-thick steel walls on the cask. As to the Petitioners' assertions regard- l ing occupational doses, the Licensee labeled these complaints meritiess both because they did not account for GPUN's comprehensive worker radiation pro-tection program and because occupational exposures were not any part of the j relief the Petitioners sought in their contention or the supporting bases. See Letter from Ernest L. Blake, Licensee Counsel, to the Licensing Board at 1-2 (Sept. II,1996). The Staff likewise criticized the Petitioners' filing as an at.

tempt to raise new issues without addressing the " late-filing" factors in 10 C.F.R.

12.714(a)(1). See NRC Staff Response to Petitioners' Status Report (Sept. I1, 1996) at 2-3.

l II. ANALYSIS A. Petitioners' Standing

1. Standing as of Right As is generally the case with intervention petitions, our consideration of the Petitioners

• hearing request begins with the question of their standing as of right. To have standing to participate as of right in a proceeding regarding an agency licensing action, a petitioner must demonstrate that (1) it has suffered or will suffer a distinct and palpable injury that constitutes injury in fact within the zone of interests arguably protected by the governing statute; (2) the injury is fairly traceable to the challenged action; and (3) the injury is likely to be redressed by a favorable decision. In addition, when, as here, an organization such as NIRS, OCNW, or CAN seeks to intervene on behalf of its members, see Intervention Petition at unnumbered p. 2, that entity must show it has an individual member who can fulfill all the necessary elements and who has authorized the organization to represent his or her interests. See Yankee Atomic Electric Co. (Yankee Nuclear Power Station), CLI 96-1,43 NRC 1,6 (1996).

In this instance, Petitioners NIRS and OCNW seek to establish their standing as of right under a different theory from that used by Petitioner CAN. NIRS and OCNW assert several of their members live, work, or engage in recreational activities sufficiently close to OCNGS to provide standing as of right. In contrast, CAN declares that although its members reside many hundreds of miles from OCNGS, the concerns of CAN members about the possible movement 156 l

of large loads over the spent fuel pools of the Yankee Nuclear Power Station in northwestern Massachusetts and, in particular, the Vermont Yankee Nuclear

, Power Station in southern Vermont are sufficient to provide CAN with standing.

We address these theories separately,

a. NIRS/0CNW Standing Petitioners NIRS and OCNW have supplied an affidavit from one individual who is a member of both organizations. He asserts he lives within the OCNGS ingestion pathway zone, which generally is within a 50-mile radius of a facil-ity; that his work for OCNW, including trips to the OCNW post office box, frequently takes him within the OCNGS plume exposure emergency planning zone (EPZ), which generally is within a 10-mile radius of a facility; that his work for the local Izaak Walton League chapter, including work on conser-vation projects within 1 mile of the facility, frequently takes him within the EPZ; and that he engages in recreational activities on a bay within the EPZ,'

OCNW also relies on three other affidavits: one from a member who lives in a housing development wherein the facility emergency plan causes residents to drive toward the plant, which is within W mile; and two from individuals who, while declaring they live within the EPZ, fail to state they are OCNW members.H' Petitioners NIRS and OCNW maintain that this information, along with these affiants' assertions that a heavy load drop onto the irradiated fuel would result in offsite releases of radioactivity and that they are concerned about the health and safety consequences of such an accident involving the fuel transfer canisters, establish the requisite injury in fact to provide each organization with representational standing. See Intervention Peti' ion at 2-3.

Both the Licensee and the Staff declare that any agency precedent regarding i

a " proximity" presumption for standing in licensing cases in which there is a

" clear potential for offsite consequences" is inapplicable in the context of this narrow license amendment dealing with load handling. Instead, they assert the Petitioners must make a showing there is some distinct and palpable injury that has or will arise from the particular amendment at issue. NIRS and OCNW have

'See Intervennon Petition unnumberrxl attach. l (aftidava of Wilham decamp, Jr ). Although the respeenve 14nule and 54 mile radius designatnici set forth m the agency's genene eraergency planmng guidance are often utilued to describe a facihty's plune exposure EPZ and mgesuon pathwny zone the actual shape of these energency planmng areas depends on the characterisucs of the parueular asic. See U.S Nuclear Regulatory Comnussion/Itderal Emergency Managenent Agency, NUREG-0654/ FEMA-REP l. "Cntena for Preparanos and Evaluanon of Radiological Energency Response Plans and Preparedness in Support of Nuclear Power Plants," at iI (rev. I Nov 1980). None of the participants has provuled us wuh a desenption of the actual paranzters of ,he oCNGS ingesnon pathway rone or plurne exposure FPZ. For present purposes, therefore, we assune the genene radius designanons are appheable.

I 3"See tener from Jean Burnett to Secretary of the Comnusuon attach dune 5,1996); tetter from Shirley R.

Schmidt to Secretary of the Conutussion anach Oune 5,19%)[ hereinafter Schmidt tstler). tetter from Mana Szczech to Secretary of the Comnussion attach Oune 7,19%) [ hereinafter Szczech letter).

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failed to do this, both GPUN and the Staff state, because with the procedural and mechanical protections GPUN will utilize in moving and lowering the shield plug over the spent fuel in the DSC, the Petitioners have not i.hown there is a credible accident sequence that would result in a shield plug drop or that such a sequence will have offsite consequences. See GPUN Answer at Il-15; Staff l Ilearing Request Response at 7-8.

, In making a standing determination, we are to " construe the petition in favor

! of the petitioner." Georgia Institute of Technology (Georgia Tech Research l Reactor, Atlanta, Georgia), CL195-12,42 NRC 111, !!5 (1995). Bearing this directive in mind, we conclude there is sufficient information on the record before us to establish a reasonable basis for the assertion of Petitioners NIRS and OCNW that a shield plug drop accident can occur and that such an accident can have offsite radiological consequences that may impact the Atomic Energy Act-protected health and safety interests of their members.

I Petitioners NIRS and OCNW have provided a number of documents regard-ing load drop accidents at nuclear facilities. See, e.g., Intervention Petition

unnumbered attach. 8 (NRC Information Notice 96-26 (Apr. 30,1996); id. un-j numbered attach. 9 (Headquarters Daily Report (May 8,1996)); id. unnumbered j attach 10 (NRC Preliminary Notification of Event or Occurrence PNO-II-94-055 l (Dec. 30,1994)). These documents indicate that, for a variety of reasons in-

ciuding mechanical failure and human error, nuclear facility load drop accidents

. do happen that result in damage, sometimes substantial, to facility equipment.

Given this information, we are unable to conclude that the possibility of a shield plug drop accident is so inherently " incredible" or " irrational" that it provides na reasonable basis upon which the Petitioners can establish their standing to challenge the requested amendment.

As for the consequences of such an accident, while again asserting it is based

on a very low probability event, the Licensee has done an analysis of a " worst case" shield plug drop that indicates there could be sorne off-site consequences to such an occurrence, albeit in a range well below the public exposure limits established in 10 C.F.R. Part 100. Relative to a threshold standing determination,

however, even minor radiological exposures resulting from a proposed licensee

activity can be enough to create the requisite injury in fact. See Yankee Atomic Electric Co. (Yankee Nuclear Power Station), LBP-96-2,43 NRC 61,70, aff'd, i CLI-96-7, 43 NRC 235, 24648 (1996). In this instance, we consider the postulated exposures are sufficient to support the Petitioners' standing claims.

Finally, based on the information supplied by two of their affiants, we find

- NIRS and OCNW have established there are reasonable grounds to conclude j these radiological offsite consequences could impact organization members, i

thereby providing standing for NIRS and OCNW. Of the two individuals who i

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are NIRS and/or OCNW members," the one, who is an OCNW member, hves '

within 1/2mile of the facility, while the other, who is a member of both OCNW I and NIRS, has organization-related and recreational activities that regularly bring him within the facility's 10-mile EPZ, sometimes as close as a mile (or less)  ;

from the facility. We find this showing of residence and regular activities near i l the facility, in conjunction with the evidence of possible offsite consequences l

[ from a shield plug drop accident, sufficient to provide these individual members, I and therefore the organizations that represent them, with standing to contest GPUN's proposed technical specification change.n

b. CAN Standing l While OCNW and NIRS ground their standing as of right on the traditional

" proximity" theory, CAN uses a more unconventional approach. As was noted above, CAN's standing assertion is rooted in its concern the precedent that may be set in this proceeding could impact its ability to contest similar amendment requests made by utilities operating nuclear power plants in the Massachusetts / Vermont area that is CAN's operational base. The affidavit from CAN's member makes it clear that her residence and activities are in that New England area, which is some 200 miles from the Oyster Creek facility.

CAN's " bad precedent" argument previously has been rejected as a basis for standing as of right. In Ohio Edison Co. (Perry Nuclear Power Plant, Unit I),

LDP-91-38, 34 NRC 229, 248 49 (1991), aff'd as to another nding, CLI l 1, 36 NRC 47 (l992), petition for review dismissed, City of Cleveland v. NRC, ,

l 68 F.3d 1361 (D.C. Cir.1995), the Licensing Board found an almost identical assertion was the sort of " generalized grievance" that was "too academic" to )

provide the requisite injury in fact for standing as of right. We agree with that '

analysis, and adopt it here to reject CAN's argument regarding its standing as of right.

H Because the other two individuals have failed to indicate they are rnenters of either orgamzanon. see Schnudi Letter attach.; Szcacch Letter ariach., their proumiry to the facihty cannot be used by NIRS or oCNW as a basis for representanonal standing See florida A=cr and t.eght Ca (Turkey Point Nuclear Generaung Plant Umts 3 and 4), ALAB-952,33 NRC 521,53431 (representanonal standing not present when individual rehed on for standmg is mW orgamzauon member, but only representauve of another orgamzanon), ag'd, CLI-91 13,34 NRC j 185 (1991) 1 U Because we have before us specific evidence of possible offsite consequences in the vicimry of the facihry from a shield plug drop incidenr. we nees not reach the issue of whether any general presumpoon regardmg possible consequences and proxinaty to the facility is appropriate. Compare Virimia Electric and An,er Ca (North Anna Nuclear Power Sianon. Limis 1 and 2). ALAB-522,9 NRC 54,56 (1979) l 159 1

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2. Discretionary Standing CAN also claims that if we find it lacks standing as of right, it nonetheless should be granted discretionary standing under the governing factors the Com-mission first established in the Pebble Springs proceeding. As outlined in that ,

decision, the factors we must consider are:  !

l (a) Weighing in favor of allowing intervention -

l (1) The extent to which the petitioner's participation rnay reasonably be ex-pected to assist in developing a sound record.

(2) The nature and extent of the petnioner's property, financial, or other interest in the proceeding.

(3) De possible effect of any order which snay be entered in the proceeding on the petitioner's interest.

]

(b) Weighing against allowing intervention -

(4) De availabihty of other tneans whereby petitioner's interest will be pro-tected.

(5) The extent to which '.ne petitioner's interest will be represented by existing parties.

] (6) The extent to which petitioner's participation will inappropriately broaden or delay the proceeding.

Portland General Electric Co. (Pebble Springs Nuclear Plant, Units 1 and 2),

CLI 76-27,4 NRC 610,616 (1976).

i As the Commission has made clear, see id. at 617, the primary consideration concerning discretionary intervention is the first factor - assistance in develop-ing a sound record. In Perry, LBP-91-38,34 NRC at 250, the Licensing Board found this factor strongly supported discretionary intervention because the party in question, having previously litigated related issues before the Commission and in federal court, was well-versed in the legal and factual issues involved in j that proceeding. We cannot say the same for Petitioner CAN here. Appearing in this proceeding pro se and apparently without the assistance of any techni-cal experts, CAN has not demonstrated any special experience or expertise it will bring to this proceeding in terms of developing a sound record. We thus conclude this important factor fails to support CAN's discretionary intervention.

Concerning factors two and three, like the Perry case, see id., we find these '

weigh in favor of discretionary intervention. Although insufficient to establish

" injury in fact," CAN's interest in stopping the proposed license amendment likewise is within the " zone of interests" relevant to this proceeding. At the same time, while too speculative to support standing as of right, its concerns about prejudice to its interest are not totally untoward in that the issue before i us, as we explain below, is a legal matter that, depending on the breadth of any Commission rulings, could have implications for any future " heavy load lifting" proceedings.

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l Also as in Perry, see id., factors four and five to a degree weigh against CAN l discretionary intervention. Based on the record before us, it seems apparent the l interest of OCNW and NIRS, who already have been found to have standing, is very much like that of CAN, albeit more concrete. Up to this point, NIRS (and I

to a lesser degree OCNW) has defended those interests vigorously. Regarding l the availability of other means to protect that interest, it may well be, depending l on the rulings in this case, that CAN would have some opportunity to contest a similar amendment request relative to Yankee Rowe or Vernmnt Yankee. As  !

with Perry, however, these negative considerations are counterbalanced by the fact that, as we outline below, the issue before us appears to be one of law, so l that additional CAN participation is not likely to broaden or delay the proceedmg '

significantly. See id. at 250-51.

Considering all these factors, particularly CAN's lack of any specific showing '

about how its participation can reasonably be expected to r.ssist in developing a j sound record, we conclude that the balance does not weigh in favor of permitting ,

CAN to become a discretionary intervenor. As such, we deny its intervention l request in toto. Nonetheless, in light of CAN's apparent concern over this matter, we provide CAN with an opportunity, if CAN wishes to use it, to appear as amicus curiae and file a pleading providing the Board with its views on the legal issue we admit for litigation in this proceeding, as detailed below. See PuMic Service Co. of New Hampshire (Seabrook Station, Units 1 and 2), ALAB-862, 25 NRC 144,150 (1987)."

15. Petitioners' Contention flaving determined which of the Petitioners has standing to be a party to this adjudication, we next turn to the matter of what, if any, issues there are for litigation. Certainly, the question of the admissibility of a petitioner's proffered l l

U As the Appeal Board noted in Seahrrmk. 25 NRC at 150, she ngency's rules of practice explicitly pernut amieus cunae participuuon only in the context of appellare proceedings. As the Appeal Board also observed, inwever.

Itus hkely reflects the fact that respiests for such parucipanon Jo not often anse in the context of L.scensing Board heanngs ~ in which factual quesuons generally predonunate - because an anucus customanly does not present witnesses or cross-examine other parties' witnesses. This happenstance, the Appeal Board concluded "does not perforce preclude the granung of leaw in apprornale circunutances to rile bnefs or menwranda amicus cunae j (or to present oral argumeno on issues of law or fact that sull remain for t.icensing Board consideranon." IJ In the context of this proceethng, in which (as we conclude below) a legal issue predominates, consistent with this Appeal Board guidance we rind permitting CAN to file an anucus pleadmg addressing that issue is entirely appropnate if we later conclude this case requires an evidenuary heanng, we can then reassess the scope and means of CAN's parucipanon So that the Boaid and the other parues will know its status, on or before fridas Amember & 1996. CAN should file a pleading mthcanng whether it intends to parucipate as an anucus cunae. In deciding whether to parucipate as an amicus. CAN may wish to consnier to what degree its parucipation m this proceedmg may make it the target of issue preclusion cl.ums u c., res judicata or collateral estoppel)if a similar technical specification change is requested at one of the New England facilines about wluch it is concerned. See Perrs L.BP.9138,14 NRC at 25I n 68.

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l contentions is of equal import "because contentions play a vital role in agency ,

licensing adjudications by framing the issues for consideration." Yanice Afomic j Electric Co. (Yankee Nuclear Power Station), LBP-96-15,44 NRC 8,21 (1996). 4 In this instance, as was described above, the Petitioners have put forth one l J contention with several bases. Both the Licensee and the Staff have challenged l 1

the contentica as lacking tre necessary specificity under 10 C.F.R.12.714(b)(2) ,

as well as hiling to have a supporting basis that, as is required by section '

2.174(b)(2)(ii) and Commission precedent, see Yankee Rowe, CLI-96-7,43 NRC l at 248-49, contains information sufficient "to show that a genuine dispute exists j with the applicant on a material issue of fact or law."

On the question of specificity, the assertion of the Licensee and the Staff

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j that the Petitioners' cortention, in and of itself, lacks the requisite specificity

] hai some merit. Nonehless, and particularly in the context of dealing with 4

pro se Petitioners, a finding regarding a contention's specificity should include ,

consideration of the contention's bases. See Public Service Co. of New Namp-shire (Seabrook Station, Units 1 and 2), ALAB-899, 28 NRC 93, 97 (1988)

(both contention and stated bases should be considered when question arises I regarding admissibility of contention)-

As we have summarized them abow:, however, Bases A and B arguably provide little help in this regard. Because the focus of that contention, as it was crafted by the Petitioners, is on the agency's " defense-in-depth" principle as emhodied in NUREG-0612," the relationship between those two bases and the contention is not readily apparent.'s When the language of the contention is j considered in conjunction with Bash C, however, the requisite specificity clearly i 4 is present, i

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j "NUREG-0612 is a 1980 document that was intended to provide "the resuhs of the NRC staff's review of the 4 handhog of heavy loads and include.s tta: NRC stMs recomnendanons on actums that should be taken to assure

, safe handhng of neavy loads." U S Nudear Regulauwy Commisuon of6cc of Nuclear Reactor Regulauan, j NUREG 0612. " Control of Heavy tmia at Nuclear Power Plants." at ni Ouly 19tto). In setimg forth guidehnes for handhng heavy loads. NUREG-0612 clearly does so in the context of ca;Tytng out the regulatory plulosophy of "defensean depth." See id at 5-1 to .2. The " defense-in-depth" pnnenple as the agency pohey under uluch regulated enuues are required to safeguard the pubhe heakh and safety "through nu,luple intermeshmg and i overlappmg protecuans " Fermons knAre kirur Amer Corp (Vermont ' Yankee Nuclear Power Station). CLI-5 74-40 8 AEC 809,813 (1974).

15 As we have outhned it above, Basis A auer1s that consistem with 10 C F R. 5 50 36(exl), activines potennally affecung fuel tud claddmg and fuel poal kner miegrny are subject to safety knuts and that the cusung technical specincanon is designed to estabbsh the specahed safety hnuts by prulubitmg the movement of any load greater i ihan the weight of one fuel assembly over or near irradiated fuel. on its face. tlus basis appears to provide no j support or otherwise bear a relationdup to the Pennoners' contenuon. The same is true of Basis B To whatever

{ the degree the purported problem with degraded fuel nught support a challenge to the t.icensee's amendment request. it bears no apparent relationslap to the NUREG.0bl2 " defense indepth" concere that is the focus of the I contennon As such, it is arguable that if Bases A and B ment any consideranon, it is only as separate contemions.

Nevertheless, for the reasons set kwth below. whether as separate contennons or as bases for the Feuucners' stated contencon we hnd these concerns inadequate to provide an adnussible issue l

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Looking then to the question of the adequacy of the bases put forth in support of their contention, even if we consider Bases A and B as having an appropriate relationship to the Petitioners' stated contention, we find them inadequate to provide an admissible contention. Basis A suffers from two llaws. First, it is footed in the misapprehension that Technical Specification 5.3.1.B is a " safety 4

limit" as that term is defined in 10 C.F.R. 9 50.36(c)(1). As both the Licensee and the Staff correctly point out, this technical specification is in fact a " design j feature" under section 50.36(c)(4). Even more telling, however, is the fact that, whether Technical Specification 5.3.1.B is a " safety limit" or a " design feature,"

nothing we are aware of in connection with section 50.36 precludes a change i in the provisions of such a technical specification if the Licensee can make the appropriate showing. As such, that regulation, and so Basis A, is irrelevant

to the Petitioners' contention that the requested change somehow violates NRC 4 " defense-in-depth" principles.

As we have noted above, to establish their Basis B concern as an appropriate

, foundation for the admission of their contention, the Petitioners rely on certain Licensee documents they declare show there are at least forty-seven fuel

. assemblies in the OCNGS fuel pool with cladding failure. This is significant, i they argue, because a shield plug drop accident involving a DSC containing such degraded fuel elements is unanalyzed in terms of possible recriticality. Further, l they discount the representations of the Licensee and the Staff that loading i i such degraded fuel assemblics into a DSC would violate the generic certificate )

of compliance under which GPUN is permitted to use the NUHOMS storage  !

system on the basis they have not been provided with documentation explaining )

how the Licensee will screen irradiated fuel assemblies for defects. See Tr. at 74-75.

Even assuming the mere declaration that a particular concern is "unanalyzed" is sufficient to provide a basis for a contention, but see rankee Rowe, LBP-96-2,

43 NRC at 75-76 (contention must not only allege decommissioning plan content deficiency, but show that purported deficiency has health and safety significance

for decommissioning process), it is apparent from the materials before us that

the Petitioners' recriticality concern has indeed been analyzed. The Licensee's l recriticality study, w hich assumes all the fuel from a fully-loaded DSC is crushed

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together, clearly envelopes this concern. Therefore, relative to any purported

! lack of an analysis, there is no material factual dispute that warrants further inquiry.

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l'In their september 9 hhng. the Pennoners acknowledge the results of the tjcensee's recrtucahry analysis " appear j techmcally correct /' but then declare they have a new concern regardmg the staiement in the analysis that the impact of a shield plug drop would not be sufficient to breach the rigid structural matenal of the TC. Pennoners i Stata: Report at 12. As the staff correctly pomted out. if the Petmoners want to raise new concerns hke this (or their addmonal claim about worker esposurest they must address the late hhng standards in 10 C F R. I 2 7 t Aat IConsmued)

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Concerning the purported lack of documentation explaining the Licensee's fuel assembly screening process, as the Licensee and the Staff noted, the certificate of compliance governing the use of the NUHOMS dry storage system makes it clear that only those fuel assemblies that are " intact" with "no known or suspected gross cladding breaches" are eligible for storage in a DSC. Hodgdon Letter unnumbered attach. 2, encl. 2, at A-10 (U.S. Nuclear Regulatory Commission, Certificate of Compliance for Dry Cask Fuel Storage  :

Casks, Certificate No.1004 (Jan. 23,1995) (Table 1 lb)) [ hereinafter Certificate l of Compliance No.10N]; see id. at A-5 (Section 1.2.1 Fuel Specification Limit / Specification). Moreover, the certificate of compliance provides that

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a these fuel specifications "must be met by every individual fuel assembly to be stored" in NUHOMS casks, id. at A-10 n.(1); see id. at A-5 (Section 1.2.1 Fuel Specification Applicability); that it must be " verified and documented" that each fuel assembly to be loaded into a DSC meets these specifications, id.

at A-5 (Section 1.2.1 Fuel Specification Action); and that immediately before insertion of a spent fuel assembly into a DSC,"the identity of each fuel assembly ,

shall be independently verified and documented," id. at A-6 (Section 1.2.1 Fuel Specification Surveillance).

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l These requirements, which are conditions of the certificate of compliance, see id. at A-1 (Section 1.0 Introduction) make it apparent that in order to meet these regulatory specifications established pursuant to 10 C.F.R.16 72.212,72.236(a) to govern the use of the NUHOMS cask system, GPUN must not load degraded fuel assemblies into a DSC. Because clear regulatory constraints mandate GPUN i must not load such spent fuel, to gain the admission of a contention founded on the premise GPUN will not follow these requirements, the Petitioners must make some particularized demonstration that there is a reasonable basis to believe

. GPUN would act contrary to their explicit terms. Having failed to make such a showing," the Petitioners' degraded fuel assembly concern is inadequate to establish a material factual dispute that warrants further inquiry.'"

, Because they have made no attempt to address these standards. we need give no further consideration to their added concerns U

As we noted previously, see apra p 154. at the August 7 preheanng conference Petitioner NIRS provided several ad.huonal oCNGS.related documents desenhmg (1) a November 1994 self-idenoned and corrected techmcal specincation violanon m which a reactor core alteranon was made without the required supervision of an appropnate semor reactor operator and (2) a July 19% Staff performance review in which GPUN is crmcized for the cononued occurrence of " avoidable" operanon and nuuntenance "penonnel errors." Ahhough these documents suggest that the Ucensee's operanon is not error free, they do not provide informanon that is sufnciently specine to estabhsh the need for rurther inquiry on the factual quesuon of the Ucensee's abihty properly to screen fuel assembhes as it is reqmred to do under the NUHoMS cem6cale of comphance.

is Ahhough not directly relaied to B. isis B (or apparently either of the other proffered bases). hkewise msufficient a

to provide grounds for an admissible contenuon are the NIRS. expressed concerns about possible problems with i kud drop dunng crane power loss and seisnue events See supra p.155. As the Ucensee indicated. the former d

clum is based on a poorly drafted sentence in the GPUN safety evaluanon regarding the proposed techmcal specihcation change that falls in make it clear that an installed protective device in fact addresses the problem c.f IConuwJ) 1 164

In considering Basis C, we reach a different result. As our summary of that l

basis indicates, and as was explained to us during the prehearing conference, j with this concern the Petitioners seek to establish the " single fuel assembly" l weight limitation in existing Technical Specification 5.3.1.B reflects an agency j judgment about the particular measures that are necessary for compliance with j the purported regulatory guidance in NUREG-0612 as it is asserted to implement  ;

the " defense-in-depth" principle. According to the Petitioners, this weight limitation is a vital control meant to remove the potential that human error or any mechanical / electrical failure could cause damage to irradiated fuel. See i Tr. at 68. Because of the importance of this limitation, the Petitioners assert, l this technical specification cannot be changed.

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The Licensee and the Staff have countered with arguments suggesting that the Petitioners' interpretation of the significance and meaning of NUREG-0612 is misplaced. We find, however, that several factors provide sufficient reason to conclude Basis C establishes a material disputed issue of law that should be considered further.

The CDPS apparently has been in place for some time, see supra p.149, indicating that the Licensee (and the Staff) had some notion GPUN at some point could be in a position to place an object heavier than a fuel assembly over fuel assemblies being packaged for removal and storage. Nonetheless, the existing technical specification with its specific " fuel assembly" weight limitation seemingly was adopted for OCNGS after NUREG-0612 was issued with its "to the extent practical" language. See U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation NUREG-0612," Control of Heavy Loads at Nuclear Power Plants " at 3-9 (Table 3.2-1),5-2 (July 1980). Further, while the Staff and GPUN have asserted that NUREG-0612 is simply " guidance" that contains no regulatory mandate, as we pointed out during the prehearing conference, there are any number of references to NUREG-0612 " requirements" in the Licensee and agency documents pmvided to us. See Tr. at 99-101; see also, e.g., Certificate of Compliance No.1004, at 2 ("The [NUHOMS) TC is designed and fabricated as a lifting device to meet NUREG-0612 and ANSI N14.6 requirements.").

This, we conclude, raises a legitimate question about the regulatory signiti-cance of that document and its "to the extent practical" language. When com-bined with the Petitioners' challenge to the exact meaning of the NUREG-0612 power loss, while the latter does not account for the fact that the crane involved is scinnucally quahned. See Tr.

at 87-ML The Peutioners' prewnt showing regarding these matters f.uls to estabhsh the requisite material ractual issue in thspute that warrants further inqmry

As we have noted. see .rupra p.15 l. the pennoners have subnutted several documents they assert estabhsh there is a sigmhcant problem wHh human error at oCNGS They do so, however, not in an enempt to support a cl.um that such human error rasses questions about the adequacy of GPUN's kiad handling trainmg and procedures, but rather as support for their general assertion that it is Not pracucal" to change the cusung techmcal specihcanon without undermuung the defense-in. depth pnnciple embodied m NUREG-Or>l2 See Tr at 70 71 165

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"to the extent practical" terminology as it relates to the requested technical spec-ification change, we find there is sufficient information to pose a matter of legal-interpretation that merits further scrutiny. As such, we admit the Petitioners' i contention as it is supported (and explicated) by Basis C.

III. SCIIEDULE i l

Section 2.714(b) of title 10 of the Code of Federal Regulations declares that a contention, such as the Petitioners' that poses a legal question "must be decided on the basis of briefs or oral argument according to a schedule determined by the Commission or the presiding officer." Notwithstanding the Licensce's suggestion that admission of the Petitioners

• contention should be followed by discovery, see Tr. at i16, from all appearances the legal issue the ,

Petitioners have framed is one that could be resolved on summary disposition j without discovery.au Because the ultimate burden on this issue rests with GPUN, see 10 C.F.R. 6 2.732, we establish the following schedule for further filings:2' GPUN Summary Disposition Motion 22 fiiday, November 15, 1996 l Staff / Petitioners / Amicus Curiae Responses liiday, December 6,1996 to GPUN Summary Disposition Motion and/or Petitioners' Cross-Motion for Sum-mary Disposition GPUN Reply to Petitioners / Amicus Curiac Friday, December 20,1996 Responses and/or Petitioners' Cross-Mo-tion for Summary Disposition, and Peti-tioners Reply to Staff Response For all further pleadings in this proceeding, in addition to serving conforming paper copies on all parties, the amicus curiae (if CAN chooses to participate in this role), the Board members, and the Office of the Secretary, a courtesy copy of each filing shall be sent to all other parties, the amicus curiae, the Board members, and the Office of the Secretary by facsimile transmission, E-mail 20 Conustent with cusung agency practice, in responang to any GPUN (or siaf0 summary espouuon nwuon.

Pentioners NIRS and oCNW can assert, with any appropnate supporung affidavits, that they need discovery to answer that dispositive monon Sir PubIrc Scrwre Co of New Hampshsrc (seabrook Stauon, Umts I and 2),

CLI-92-8, .15 NRC i44, I52 (1992).

2 The noard will advise the parties at a later date d si intends to hold an oral argument regarding their summary disposinon tihngs 22 1f the Staff wishes it may file a disposinve monon on this date as well If the Staff does so, the Pennoner and amicus cunae responses should encompass both the GPUN and Staff dispouuve monons and the Staff is pernuned to file a reply 40 any such responses in accordance with the schedule.

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transmission, or other means that will ensure receipt by 4:30 p.m. Eastern Time on the date of filing.

Substantive summary disposition-related pleadings other than those autho-rized in the schedule above are not permitted without preapproval of the Board.

Board preapprovat must be sought in wnting at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before filing the pleading. The preapproval request must indicate whether the other parties to the proceeding oppose or support the request.23 IV. CONCLUSION Petitioners NIRS and OCNW have shown that (I) at least one of their members who has authorized NIRS or OCNW to represent his or her interests lives, works, or engages in recreational activities near OCNGS; and (2) there is some reasonable basis to believe that, as a consequence of a shield plug drop incident, those individuals' proximity to the facility can result in injury to their health and safety interests as those interests are protected by the Atomic Energy Act. Petitioners NIRS and OCNW thus have established their standing as of right to be parties to this proceeding. In contrast, the interest of Petitioner CAN and its proffered member (who lives well away from OCNGS) in avoiding adverse precedent from this case is too generalized and academic to provide CAN with standing as of right. Further, CAN has failed to demonstrate it should be granted discretionary standing. Therefore, CAN's intervention request is denied, ,s although it can (if it wishes) participate in the initial summary disposition stage of this proceeding as an amicus curiae.

We also conclude that the Petitioners' joint contention, as supported by Basis C as summarized above, see supra pp.151-52, is sufficient under 10 C.F.R.

5 2.714(b)(2)(iii) to establish a genuine material issue of law. As such, we admit their contention and establish a schedule for further litigation on its merits.

Ihr the foregoing reasons, it is, this twenty-fifth day of October 1996, ORDERED that:

1. Relative to the contention set forth in their July 18,1996 supplemental intervention petition, as that contention is supported by Basis C as summarized above, the June 6,1996 hearing request and petition to intervene of Petitioners NIRS, OCNW, and CAN is granted as to NIRS and OCNW and is denied as to CAN.

2. Litigation on this contention will commence immediately in conformance with the schedule and procedures specified in section III above.

33 our previous direcuves concermng the (mung and coment of motmns for extensmn of tune remam applicable.

See Board Imtial ordet at 4 167

3. In accordance with the terms specified in sections 11 and 111 above, CAN is grunted permission to participate as an amicus curiae relative to the contention admitted in this proceeding.

4, in accordance with the provisions of 10 C.F.R. 6 2.714a(a), as it rules upon an intervention petition, this memorandum and order may be appealed to the Commission within 10 days after it is served.

Ti!E ATOMIC SAFETY AND LICENSING BOARD 24 G. Paul Dollwerk, III. Chairman ADMINISTRATIVE JUDGE Charles N. Kelber ADMINISTRATIVE JUDGE Peter S. Lam ADMINISTRATIVE JUDGE Rockville, Maryland October 25,1996 1

24 Copies of tlus Menmranduni and Order have been sent this date to counsel fu GPUN and the representanves for NIRS and CAN by facsinule transnussion; to the representative for OCNw by miernet E nuul transnussion; and to Staff counsel by E-Mal transnussion through the agency's wide area networt 168

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! Decisions l Under l 10 CFR 2.206 1

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Cite as 44 NRC 169 (1996) DD-96-12' UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION i l William T. Russell, Director 1 i

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in the Matter of All Dockets j (All Licenses) i ALL NUCLEAR POWER PLANTS September 26,1996 4 By petition dated March 5,1996, Petitioner Charles Morris requested that the operating licenses of all nuclear power plants be immediately suspended, and remain suspended due to what Petitioner saw as a need to correct repeated errors in the plants' undervoltage relay setpoints and electrical distribution system j designs. Petitioner provided a number of reasons to support his request, j In a Director's Decision dated September 26,1996, the Director of Nuclear

] Reactor Regulation denied the relief sought by Petitioner, concluding that no j substantial health and safety issues had been raised by Petitioner to warrant '

, the action requested, as the NRC Staff had adequately addressed Petitioner's a concerns. With regard to the request for immediate suspension, the Director l concluded that licensees had to a large degree also already addressed the issues l raised by Petitioner.

i DIRECTOR'S DECISION UNDER 10 C.F.R. 5 2.206 l I. INTRODUCTION i

- On March 5,1996, Mr. Charles Morris (Petitioner) filed a petition with the Executive Director for Operations pursuant to section 2.206 of Title 10 of the 1

Code of Federal Regulations (10 C.F.R. 9 2.206). The Petitioner requested that l

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• Because of unusual circumstances, this three w's Decision was nm pubbshed in the Sepember 19961sswnces.

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i the operating licenses of all nuclear power plants be suspended within 90 days and remain suspended until such time as those plants have (1) discovered the

( reason for what the Petitioner asserts are repeated errors in the undervoltage l relay (UVR) setpoints (SPs) and electrical distribution system (EDS) designs l and (2) provided convincing evidence that these deficiencies have finally been corrected. Since the Petitioner had requested action within 90 days, the request was treated as a request for immediate relief. The Petitioner also requested that the aforementioned evidence be reviewed by a competent third party, in addition to the Nuclear Regulatory Commission (NRC) Staff, and that if the NRC concludes that plants may safely operate with UVRs that cannot be properly set

! for long periods of time, the NRC should reach these conclusions by wa'y of a l public meeting.

l On April 17, 1996, the Petitioner was informed that the request for the suspension of all nuclear power plant licenses within 90 days for the purposes of remedying repeated errors in UVR SPs and EDS designs was denied because licensees have, to a large degree, already addressed the issues that the Petitioner had raised. Also, the Petitioner was informed that the request was being evaluated pursuant to section 2.206 of the NRC's regulations and that a decision, as provided by section 2.206, would be made on the request within a reasonable time.

On the basis of my review of the issues raised by the Petitioner as discussed below, I have concluded that no substantial health and safety issues have been raised that would require the initiation of the action requested by the Petitioner.

II. DISCUSSION in his petition, the Petitioner stated his concern that the " enduring and widespread nature of the electrical distribution system (EDS) and undervoltage relay (UVR) setpoint (SP) errors (e.g., incorrect UVR and thermal overload setpoints) was recognized by neither the licensees nor the NRC staff," and was not included in NRC Information Notice (lN) 93-99, "Undervoltage Relay and Thermal Overload Setpoint Problems."

IN 93-99 did, in fact, inform all holders of operating licenses or construction permits of the widespread nature of the SP errors by listing approximately forty licensees with incorrectly set UVRs or thermal overload (TOL) protective r

devices. The identification o these problems was not inadvertent, but was the result of concerted NRC Staff attention to these issues. As was indicated to the Petitioner in an April 17,1996 letter acknowledging receipt of his March 5,1996 section 2.206 petition, the Petitioner himself recognized that Electrical Distribution System Functional inspections (EDSFIs) were highlighting these issues and that licensees were conduct;ng self-initiated design-basis reviews 170 l

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(possibly in anticipation of pending EDSFIs) to identify problems and were undertaldng corrective actions.

In his March 5,1996 petition, the Petitioner listed seven specific reasons that he believed caused repeated EDS and UVR deficiencies. The following is a description of each concern accompanied by the NRC Staff's response:

1. The Petitioner stated that NRC Branch Technical Position PSB-1,

" Adequacy of Station Electric Distribution System Voltages," contained in NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," which requires a degraded voltage relay with a long delay and a loss-of-power relay with a short delay, is inadequate because it does not recognize the complexity of the matter. Except for the arbitrary time delays associated with the UVRs, no recognition has been made of voltage dynamics and time dependence. Signal bandwidths, responses of tap-changing transformers, and UVR time delays have been overlooked and should be considered.

RESPONSE: NRC Branch Technical Position PSB-1 does not recom-mend that licensees arbitrarily select time delays for UVRs. On the contrary, PSB-1 states that "the selection of undervoltage and time delay setpoints shall be determined from an analysis of the voltage requirements of the Class IE loads at all onsite system distributions levels." Further, it states that " Tap settings selected should be based on an analysis of the voltage at the terminals of the Class 1E loads. The analyses performed to de.

termine minimum operating voltages should typically consider maximum unit steady state and transient loads . " Additionally, "the first time delay should be of a duration that established the existence of a sustained degraded voltage condition (i.e., something longer than a motor starting transient)" and "the second time delay should be of a limited duration such that the permanently connected Class IE loads will not be damaged."

Therefore, the Staff concludes that NRC Branch Technical Position PSB-1 is adequate as it addresses those topics that the Petitioner believes are neglected by the Branch Technical Position.

2. The Petitioner asserted that UVR tolerances are statistical in nature and not, as the Staff and design engineers often regard them, limits to the errors in the relay SPs. 'Ihis is a significant problem that may not be solved if previous approaches are utilized and decision analysis is not applied to study the consequences of attempting to prevent the occasional loss of the most vulnerable safety load at the expense of transferring a complete division to another power source with attendant problems.

RESPONSE: Regulatory Guide 1.105, " Instrument Setpoints for Safe-ty-Related Systems," states that ISA-S67.(M-1982, "Setpoints for Nuclear Safety-Related Instrumentation Used in Nuclear Power Plants," establishes NRC Staff guidance for ensuring that instrument SPs in safety-related l

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systems are initially within and remain within the technical specification i limits. Section 4.3.1 of ISA-S67.04 states that instrument accuracies l (uncertainties, errors, or tolerances) may be combined in one of five ways: algebraically, square root of the sum of the squares, statistically, probabilistically, or combinations of the first four. Justification is to be j provided for the method used.

Regulatory Guide 1.105 expands upon this point:

Paragraph 4.3 of the standard specifies the methods for combining uncertainties in deterinining a trip setpomt and its allowable values. Typically. the NRC staff has j accepted 95% as a probatuhty haut for errors. That is, of the observed distnbution I of values for a panicular error component in the empirical data base,95% of the I data points will be bounded by the value selected. If the data base follows a normal I distnbution, this corresponds to an error distribution approximately equal to a "two I sigma ** value. I Although the use of"two sigma" values (values equal to twice the stan- I dard deviations of the errors) does not completely ensure that the measured parameter will not exceed the safety analysis limit without accompanying protective action, the probability of all the individual errors occurring si-multaneously at their extreme, nonconservative, random values is very low.

Therefore, the regulatory guide and the industry standard together support a credible, statistical approach for establishing SPs that considers such things as sample size of error values, random versus nonrandom errors, and inde- ,

pendence of errors.  !

The preparatory training for EDSFI team members also did not overlook the statistical nature of the UVR tolerances. In section 4.8.2 of the EDSFI training textbook, a discussion of instrumentatiori SP problems was provided with a sample application of ISA-S67.04 to degraded voltage relays. This methodology was also discussed in the course itself. Using this knowledge, EDFSIs were conducted and findings were written covering improper degraded voltage relay SPs. As a result, licensees then followed this action with event notification and other activities as described in Information Notice 93-99.

Additionally, in response to a request from Region III pertaining to an unanalyzed degraded voltage concern at Perry Nuclear Power Plant, the Electrical Engineering Branch (EELB) of NRR in an April 13,1992 memo provided inspectors in NRC regional offices with guidance for establishing an adequate SP for the degraded voltage relays by way of reference to section 4.8.2 of the EDSFI training course manual and ReFulatory Guide 1.105. Furthermore, the Staff informed all holders of operating licenses about a statistical approach for establishment of UVR SPs when IN 91-29, " Deficiencies Identified During Electrical Distribution Functional 172

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J i Inspections," made reference to ISA-S67.04-1982 for useful guidance in

! determination of SPs.

l The Staff therefore has regarded the UVR SP determinations as statistical

! in nature.

3. The Petitioner stated that although General Design Criterion (GDC) 17 " Electric power systems," requires all EDS to be testable, only parts are tested because plants cannot conveniently be placed in a condition where actual loads can be placed on the EDS and measured.

RESPONSE: The Staff has always been aware that in certain situations it is not practical or safe to test each and every component in the exact way it is used. GDC 18, " Inspection and testing of electrical power systems,"

states that " systems shall be designed with a capability to test periodically

... the operability of the systems as a whole and, under conditions as close to design as practical. . ." Regulatory Guide 1.118, " Periodic Testing of Electric Power and Protection Systems," Revision 2, endorses IEEE Std 338-1977, " Criteria for the Periodic Testing of Nuclear Power Generating Station Safety Systems," which states that "the test program of each system shall be designed to provide for minimum interference with related operational channels, systems, or equipment." It further states that "wherever possible, tests shall be accomplished under actual or simulated operating conditions, including sequence of operations, for example, diesel load sequencing," but also l l

where it is not practicable to initiate the protectne action. the system shall be designed such that . . Designs . . shall be jusufied on the basis that there is no practical system design that would pennit operation of the actuated equipment without adversely affecting the safety or operabihty of the plant, and that the probabihty of failure of actuated equipment not tested dunng plant operation is acceptably low, and that the actuated equipment can be routinely tested when the plant is shut down.

It is the Staff's goal to have all components of the EDS periodically tested in a manner that is both reasonable and practical. Various practical l test methods such as the use of minillow paths, overlap testing, simulated loads, etc., have been found acceptable by the Staff.

NRC Temporary Instruction 2515/107 (which provided guidance for per-forming EDSFIs) required the EDSFI teams to " verify that the surveillance and test procedures are adequate to demonstrate the functionality of the equipment or system being tested or the design assumptions being veri-fied "

Therefore, as shown above, testing of the EDS is evaluated in terms of satisfying NRC requirements (GDC-l and GDC-18) utilizing the guid-ance provided by Regulatory Guide 1.118 for a reasonable and practical 173 l

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i approach (in lieu of testing each system as a whole), and tests are properly implemented in the manner described above.

4. The Petitioner pointed out that load nameplate ratings are used in voltage analyses even when common knowledge shows that most loads are operated at a fraction of their ratings. Furthermore, worst-case ambient temperatures are used to select motor protection time delays even though few loads, if any, see those conditions except during a loss-of-coolant accident when the motor protection is bypassed. Additionally, UVR output delays are treated as known quantities, when the protection of loads by time delays and inverse time overcurrent relays is a crude mitigating approach. As a related matter, the Petitioner objects to the inconsistent use of significant figures to represent EDS and UVR SP parameters.

RESPONSE: The aforementioned temporary instruction (TI) for the EDSFis stated that the inspectors should verify that values for mechanical loads used for electrical calculations are based on actual system operating points during both normal and accident conditions. The Staff expects licensees to perform accurate, conservative, and bounding calculations involving worst-case estimates for parameters such as ambient temperatures and loads. The licensees' analyses are reviewed by the Staff utilizing engineering judgment and applicable industry guidance to ensure that reasonable, yet adequately safe, solutions are provided.

It is true that, occasionally, d.:sians proposed by licensees do involve basic approaches (such as inverse time delay relays) and that some cal-culations performed by licensees involve the use of ultraprecise numerical values. What the Staff does~ require is that the designs utilized by licensees meet applicable NRC regulations and that adequate protection of public health and safety is ensured. i The Staff, therefore, concludes that component characteristics are treated and utilized properly in calculations that support EDS and UVR designs.

5. The Petitioner believed that when licensees have discovered that UVR SPs are set too low, th,: typical response has been to raise the SPs.

This, in turn, reduces the safety advantage of providing UVRs for the EDS due to more frequent and unnecessary UVR actuations accompanied by possib!c undesirable power systems transfers.

RESPONSE: In a letter dated August 8,1979, addressed to all power reactor licensees regarding the adequacy of station electric distribution systems voltages, the Staff stated that:

Protection of ufety loads from undervoltage conditions must be designed to provide the required protection without causing voltages in excess of maximum voltage ratmgs of safety loads and unhout causing spurious separations of safety buses from offsite power.

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Moreover,

[v]oltage4ime settmgs for undervoltage relays shall be selected so as to avoid spurious separation of safety buses from offsite power durmg plant startup, normal operation and shutdown due to startup and/or operatwn of electric loads.

NRC Branch Technical Position PSB-1 states that:

imporper [ sic] voltage protection logic can itself cause adverse effects on the Class IE systems and equipment such as . . spurious separation of Class IE systerra from offsite power due to normal motor starting transients.

Additionally, in IN 95-37, " Inadequate Offsite Power System Voltages During Design-Basis Events," the Staff informed power reactor licensees that although raising UVR SPs ensure that adequate voltages exist at equipment input terminals, the higher SPs also increase the potential for separation from the offsite power system during design-basis events over the range of normally anticipated offsite grid voltages.

l In a more specific example, a February 23,1995 Staff safety evaluation I

of the degraded voltage design for the Edwin 1. Hatch Nuclear Plant determined that a combination of automatic and manual actions was an acceptable alternative approach to meet the branch technical position in lieu of raising the degraded voltage SPs which could lead to unwanted plant trips. Dat safety evaluation and the above Staff guidance provide evidence that the Staff has considered avoidance of spurious bus trips as one objective to be considered when selecting an adequate SP for UVRs.

The Staff, therefore, has repeatedly and in detail both considered the detrimental effects of raising the UVR SPs and communicated its concerns to licensees.

6. The Petitioner stated that in IN 95-05, "Undervoltage Protection Relay Settings Out of Tolerance Due to Test Equipment Harmonics," the Staff discovered that peak reading voltmeters calibrated for root mean square (RMS) are affected by the proportions of harmonics in the AC bus voltages and in the calibrators used to set the UVRs. Additionally, the harmonics affect the UVR responses by changing their SPs when the harmonic content of the bus voltage changes.

RESPONSE: IN 95-05 discusses three occurrences, reported by li-censees, where harmonics in the output voltage of the power supplies used during testing and calibration of UVRs resulted in the relay SPs being out of tolerance. The SP errors were also affected by the use of digital voltmeters which do not respond to the hannonic content of the test input voltage as do the UVRs. He purpose of the IN was to inform all operating power plant licensees that harmonics in the voltage inputs (test source voltage or normal 175

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bus voltage) to the UVRs impact the actual operating points of those relays, as the Petitioner believes, and to instruct the licensees to take appropriate action (i.e., install filters, adjust SPs, select proper test equipment, etc.) to ensure that UVR SPs are adequate.

The Staff, therefore, has addressed this concern ud brought it to the attention of licensees who are taking appropriate action as discussed above.

7. The Petitioner concluded that impedances and inrush currents to motors and other loads are not known to the precision with which the Staff and the licensecs' engineers have been trying to set UVRs. Both groups must recognize that their task may be impossible and that their attempts to do so have increased the risk of a nuclear accident.

RESPONSE: Branch Technical Position PSB-1 states that voltage anal-yses (including effects of impedances and inrush currents) should be per-formed with analytical techniques and assumptions verified by actual mea-surement. It also states that, in general, test results should not be more than 3% lower than the analytical results. This level of precision has been determined to be acceptable based on engineering judgment.

Furthermore, as stated in the response to the Petitioner's fourth concern, even though licensees propose solutions involving different equipment and unique, precise calculations (which should be supported by actual test data as mentioned above), Staff reviews are conducted utilizing both guidance from Branch Technical Position PSB-1 and engineering judgment to ensure that all applicable regulations are met and that adequate protection of public health and safety is ensured. This approach provides reasonable assurance that the level of risk of a nuclear accident is not increased and remains acceptable.

Choosing an SP above an analytical limit based on minimum voltage re-quirements and below nominal voltage ranges while accounting for instru-mentation errors and analytical inaccuracies is often a challenge that leads licensees to use more precise equipment and more precise calculations. It is concerns such as these that have led the Staff to consider alternative ap- j proaches to its position on degraded voltage protection on a plant-specific basis as noted above in the Staff's response to the Petitioner's fifth concern.

Therefore, although the Staff has concluded that the task is not impossi-ble, it has recognized alternative approaches that address degraded voltage concerns without increasing the risk of an accident.

To continue the discussion, identification of probicms with UVRs and EDSs was not inadvertent. The NRC Staff had undertaken more global measures to ensure that concerns such as those raised by the Petitioner were addressed satisfactorily. Because previous NRC inspection teams had observed that the required functional capabilities of certain safety-related systems (including EDSs) were compromised due to a lack of proper engineering support and 176 1

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the introduction of various design deficiencies, EDFSIs were scheduled to be conducted for all operating plants beginning with pilot inspections in 1989.

NRC TI 2515/107 was issued on October 19,1990, to be made part of the NRC j Inspection Manual. That TI stated that calculations to establish protective relay l SPs had not been initially performed or were not updated to reflect SP changes j and plant modifications. These failures constituted some of the deficiencies that I had been encountered by previous inspection teams. The TI stated, with regard to those concerns voiced by the Petitioner, that the forthcoming inspections should verify:

. That ratings and SPs have been correctly chosen and controlled for j protective and control relays and circuit breakers to ensure proper coordination, protection, required automatic action, and annunciation.

. The adequacy of the load study, voltage profiles, voltage drop calcu-lations, motor starting study, load shedding, engineered safety features (ESF) bus load sequencing and overload trip settings for ESF loads in.

ciuding consideration of steady-state and accident-transient loads and consideration of acceleration of the loads during degraded voltage con.

ditions that may occur during various modes of plant operation and ac-cident mitigation scenarios.

. The adequacy of short-circuit calculations, design of protective relay logic and relay setting calculations, grounding calculations and schemes, and protective device coordination studies.

• That SPs for overcurrent protective relays are correctly chosen (1) to ensure proper breaker coordination between different voltage levels; (2) to prevent exceeding the vendor-specified thermal limits on motors, containment electrical penetrations and cable in;ulation systems; (3) to allow starting of electrical equipment under degraded voltage conditions; and (4) to provide adequate pretrip alarms, when applicable.

• The adequacy of SPs and time delays for other protective relays for attributes such as undervoltage, underfrequency, reverse power, ground faults, differential current, thermal overload and phase synchronization to assure functionality of the EDS.

• That mechanical loads, such as pump horsepower, correspond to actual system operating points during normal and accident conditions and have been correctly translated to electrical loads and incorporated in the electrical load list as appropriate.

. That surveillance and test procedures are adequate to demonstrate the functionality of the equipment or system being tested or the design assumptions being verified.

NRC inspectors (including NRC contractors) assigned to the EDSFI teams attended a week-long course (held in September and December 1990) to enhance their knowledge of EDSs, the TI, and related requirements. Using the guidance 177

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provided by the T1 and the EDSFI training course, the EDSF1 teams then conducted inspections of the EDSs through early 1994 at most operating nuclear power plants. As a result, numerous deficiencies were identified and documented in plant-specific EDSFI inspection reports, and corrective actions were taken.

-those corrective actions were subsequently evaluated, found acceptable by the Staff, and documented in followup inspection reports. Many of these deficiencies and corrective actions were listed in IN 93-99 and include incorrect UVR relay and thermal overload SPs caused by design errors, as well as other points raised by the Petitioner, in summary, as stated in my April 17, 1996 letter, I believe that the NRC Staff recognized the existence of repeated errors and widespread EDS design deficiencies, including those associated with UVR SPs, took appropriate actions (conducted EDSFis, identified deficiencies, required corrective actions) based on those observations, and made all licensees aware of typical design deficiencies encountered during EDSFis and licensees' self-initiated efforts by issuing ins such as IN 91-29, " Deficiencies identified During Electrical Distribution System Functional Inspections," its supplements, and IN 93-99. Additionally, the Staff  !

has continued to inform power reactor licensees of other design deficiencies ,

when they are encountered (e.g., IN 95-37 which discusses UVR SPs in l relationship to inadequate offsite power system voltages during design-basis  !

events) and will continue to do so in the future when necessary. Such action by the Staff is appropriate to address repeated errors in UVR SPs and EDS designs and to provide reasonable assurance of adequate prr.:ection of public health and safety.

l 111. CONCLUSION

'Ihe institution of proceedings pursuant to section 2.206 is appropriate only if substantial health and safety issues have been raised. See Consolidated Edison Co. of New York (Indian Point, Units 1,2, and 3), CL1-75-8,2 NRC 173,175 (1975), and Washington Public Power Supply System (WPPSS Nuclear Project l

No. 2), DD-84-7,19 NRC 899,924 (1984). This is the standard that has been applied to the concerns raised by the Petitioner to determine whether the action requested by the Petitioner, or enforcement arJion, is warranted. l l

l On the basis of the preceding assessment, I have concluded that no substantial health and safety issues have been raised by the Petitioner that would warrant the action requested by the Petitioner. I further conclude that the Petitioner's i concerns have been adequately addressed by the Staff and that there is no need j for a third-party review. Additionally, with regard to plants with UVRs that I cannot be properly set, the Staff has shown in plant-specific evaluations, such j as described above, that other alternative designs are acceptable.

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3 He Petitioner's request for action pursuant to section 2.206 is denied. As provided for in 10 C.F.R. 5 2.206(c), a copy of the decision will be filed with the Secretary of the Commission for the Commission's review. He decision will constitute the final action of the Commission 25 days after issuance unless ,

the Commission, on its own motion, institutes review of the decision in that '

time.  :

i FOR TiiE NUCLEAR REGITLATORY COMMISSION 3

i William T. Russell, Director Office of Nuclear Reactor Regulation j Dated at Rockville, Maryland, j this 26th day of September 1996.

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Cite as 44 NRC 180 (1996) DD-96-13 i

UNITED STATES OF AMERICA '

NUCLEAR REGULATORY COMMISSION .

OFFICE OF NUCLEAR REACTOR REGULATION Frank J. Miraglia, Acting Director in the Matter of Docket No. 50-302 i

FLORIDA POWER CORPORATION (Crystal River Nuclear Generating Plant, Unit 3) October 7,1996 De Acting Director grants a petition filed by Mr. Louis D. Putney, Esq., on behalf of Barry L. Dennett, to the extent that it requested the NRC to determine

~t he validity of alleged security deficiencies at Crystal River Nuclear Generating l

Plant, Unit 3 (CR3). Most of the allegations were not substantiated. The Acting Director denies the petition to the extent that it requested the Acting Director to institute a proceeding to suspend or revoke the operating license of CR3, pursuant to 10 C.F.R. 9 2.202, upon confirmation of the validity of the allegations. He Acting Director determines that with respect to the Petitioner's I substantiated concerns and other security concerns identified by the NRC Staff, the Licensee took appropriate action to correct the deficiencies and no further action is warranted.

SECURITY PLAN: DRILLS Dere is no regulatory requirement to report the results of drills to the NRC unless certain safeguards system weaknesses are discovered during the drills that  ;

could allow unauthorized or undetected access to protected or vital areas of the l reactor. See 10 C.F.R. 69 73.55 and 73.71.

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1 DIRECTOR'S DECISION UNDER 10 C.F.R. f 2.206 I. INTRODUCTION -

On March 28, 1996, Louis D. Putney, Esq., on behalf of Barry L. Bennett (Petitioner), filed a petition pursuant to section 2.206 of Title 10 of the Code of l

Federal Regulations (10 C.F.R. 5 2.206) and alleged a number of security deti- '

ciencies at Florida Power Corporation's (the Licensee's) Crystal River Nuclear Generating Plant, Unit 3 (CR3). The Petitioner requested that the U.S. Nuclear Regulatory Commission (NRC or the Staff) investigate the security deficiencies at CR3 and, upon determination of their validity, institute a proceeding to sus-pend or revoke the operating license of CR3, pursuant to 10 C.F.R. 5 2.202, until such time as these concerns are corrected. The petition was referred to the Office of Nuclear Reactor Regulat;on (NRR) for action in accordance with section 2.206.

In a letter dated April 24, 1996, to the Petitioner, the Director of NRR acknowledged receipt of the petition and informed the Petitioner that his request l was being treated as a petition under section 2.206. He April 24th letter also '

informed the Petitioner that as provided by section 2.206, action will be taken on his request within a reasonable time. Receipt of the petition was noticed in the Federal Register (61 Fed. Reg. 31,562 (1996)). De Staff has completed its l review of the issues and has reached its conclusions, which are discussed herein. l II. BACKGROUND The Petitioner alleged security deficiencies at the CR3 plant and stated that they render the nuclear security program at CR3 ineffective. As the basis of his request, the Petitioner described examples of the security concerns, which involved the following four areas: compliance with licensing requirements and maintaining an effective security program; a pattern of lax security and failure to report security breaches; a practice of using only one guard to monitor seseral protected zones or entrances to the protected area; and a reduction of security force personnel.

He NRC Staff has reviewed the petition and the results of this review are discussed below.

A special inspection was conducted during the periods of March 18-22 and April 3-5,1996, and is documented in NRC Inspection Report (lR) 50-302/96-

02. This IR contains safeguards informatiori as defined by 10 C.F.R. 5 73.21 l and its disclosure to unauthorized individuals is prohibited by section 147 of the l Atomic Energy Act of 1954, as amended, and therefore is not available for public review. However, the IR summary does not contain safeguards information and, 181

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therefore, is available for public review at the Commission's Public Document

, Room,2120 L Street, NW, Washington, DC, and at the local public document room located at Coastal Region Library,8619 W. Crystal Street, Crystal River, IL 32629.

4 III. DISCUSSION He Petitioner alleged that CR3's compliance with one of its licensing requirements, that is, maintaining a security program that would be effective l

against terrorist attack, is inadequate. Specifically, the Petitioner alleges that l an operational security response effectiveness drill conducted in 1995 was 1 unsuccessful and the results were not formally documented and reported to the  !

4 NRC. Further, the Petitioner claims that the deficiencies revealed by the drill j have never been corrected, and thus the plant remains susceptible to terrorist attack.

4 Two types of security drills have been conducted at CR3: an Operational Safeguards Response Evalustion (OSRE) by the NRC and a Security Organi.

zation Response Exercise (SORX) by the Licensee. The NRC Staff conducted an OSRE on February 15-18, 1994, and its results are documented in a letter i to the Licensee dated August i1,1994. The Licensee conducted SORX drills during May and June 1995. The Staff contacted Louis D. Putney, the attorney for the Petitioner, to clarify whether the Petitioner's concern is related to the Licensee's SORX or the NRC's Operational Safeguards Response Evaluation.

Mr. Putney confirmed that the issue is related to the Licensce's SORX drill.

In the course of the March 18-22 and April 3-5,1996 inspection, the inspector reviewed documentation, and interviewed Licensee representatives to determine whether the Licensee was meeting commitments specified in the Training and Qualification Plan (T&QP).

He inspector verified during these two inspection periods that the security force was being trained in accordance with the provisions outlined in the T&QP by reviewing 1995 records for ten randomly selected security force members employed in the position of either response team member, alarm station operator / analyst, or access control officer. All members of the security force were appropriately equipped. He records reviewed indicate that the tasks, weapon requalification scores, and physical fitness requiremems were documented sati sfactorily. Interviews with security officers in various positions verified that they were knowledgeable of their duties and responsibilities. The i inspector concluded that the Licensee, at the time of these inspections, was meeting the commitments specified in the Licensee's T&QP.

He inspector reviewed the Licensee's documentation for SORX drills, which were conducted during May and June 1995. The Licensee used attendance sheets 182

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to document each participant's attendance and performance. All participants for the seven SORX drills were documented as performing satisfactorily. In addi-tion, these attendance sheets were signed and dated by the instructor / assessor, who on several occasions was the Petitioner. The Licensee stated that the drills were successful, and inspection of the Licensee's records and interviews with its employees did not show otherwise. Upon further discussion with Licensee representatives, the inspector learned that the Licensee documented the 1993 and 1994 drills on Form TDP-307 and the 1995 drills on the attendance sheets as discussed above. Based on review of the documentation, interviews of the  ;

Licensee representatives and security officers, and direct observations, the in-spector concluded that there were no discovered vulnerabilities in the Licensee's safeguards system or violations of licensed requirements during the Licensee's SORX drills and that the Licensee's training and qualification program meets the requirements in the T&QP.

The NRC inspector verified that the 1995 SORX drill results were not reported j to the NRC, as alleged by the Petitioner. However, there is no regulatory l requirement to report the results of drills unless certain safeguards system l

i weaknesses are discovered during the drills that could allow unauthorized or i undetected access to protected or vital areas of the reactor. If the above weaknesses are discovered they are required to be compensated, corrected, l and reported or documented in accordance with NRC regulations: 10 C.F.R.

l 66 73.55 and 73.71. No such vulnerabilities in the 1995 SORX drills were l identified. The Staff did not find violations of regulatory requirements in the conduct or documentation of the 1995 drills, and the Petitioner's concerns are i not substantiated.

The Petitioner states that "there is a general laxity of security" and "a pattern of failure to report security breaches" at Crystal River. As the basis for these claims, the Petitioner cites three separate incidents that occurred in 1995 for which security reports were not filed: (1) a guard was found asleep at a compensatory post, (2) a security lieutenant took his badge off site, and (3) a guard was found reading a book instead of watching three security zones as l assigned.

Pursuant to 10 C.F.R.173.71, licensees are required to report certain safe-l guards events to the NRC within I hour of discovery, and other events must be recorded within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in the Safeguards Event Logs that are maintained by each Licensee. During the weeks of March 18-22 and April 3-5,1996, the inspector reviewed the Licensee's Safeguards Event Logs for the period January 1995 to March 1996 to verify that the criteria specified in section 73.71 were being met. The inspector verified that the three safeguards events identified by the Petitioner were documented in Security incident Reports and logged in the Licensee's Safeguards Event Log as required by section 73.71. The inspec-tor also determined that these three events were not 1-hour-reportable events 183 l

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pursuant to section 73.71 and Appendix G of Part 73. All of the three events identified by the Petitioner were properly logged and compensated for in accor-dance with section 73.71. Therefore, the Staff substantiated that these incidents occurred, but did not substantiate the Petitioner's claim of " failure to report security breaches."

During the March and April inspections, the inspector identified four viola-tions of regulatory requirements relating to failure to adhere to the Licensee's Physical Security Plan but unrelated to the specific issues raised by the Peti-tioner. By letter dated May 1,1996, the Staff issued a Notice of Violation citing these violations.

Hree of these violations are related to operability of the vehicle barrier gate, protected area lighting, and storage of safeguards material. In response, on May 31,1996, the Licensee submitted its corrective action plan to ensure that such violations would not recur.

The fourth violation related to certain compensatory measures that the Licensee implemented as part of its security upgrade. Specifically, the violation cited that the Licensee's compensatory actions decreased the effectiveness of the alarm stations and did not meet the provisions specified in 10 C.F.R. I 50.54(p).

The NRC Staff, in a letter dated March 29, 1996, informed the Licensee to cease the compensatory measures. In a subsequent meeting with the NRC on April 2,1996, the Licensee informed the NRC of the actions that it would take to maintain compliance with regulatory requirements. During the inspection of April 3-5,1996, the NRC Staff verified that the Licensee was adhering to its commitments. Although this violation was serious, the NRC Staff believes that the timely actions implemented by the Licensee to correct these deficiencies were satisfactory and that no further action by the NRC is warranted. Further, the Staff concludes that neither the incidents identified by the Petitioner with respect to security personnel's performance, nor the violations identified by the Staff constitute "a general laxity of security."

The Petitioner states that the Licensee's current practice of using only one guard to monitor several protected zones or entrances to the protected area does not provide adequate security. The Licensee has committed to monitoring multiple protected zones or entrances in its NRC-approved Physical Security Plan (hereinafter referred to as the Plan) which describes compensatory measures that must be implemented when equipment or other resources are not in service.  :

During the weeks of March 18-22 and April 3-5, 1996, the inspector reviewed j I

the Licensee's security program at CR3 with respect to guard monitoring of protected zones and found it to be in compliance with the Plan. Additionally, the inspector reviewed the established compensatory posts and determined that they were in accordance with the Licensee's Plan and also with the recommended l NRC guidance developed in NUREG-1045, " Guidance on the Application l l

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1 of Compensatory Safeguards Measures for Power Reactor Licensees," dated January 1984. 1 On the basis of its inspection, the Staff finds that the Licensee's current practice of monitoring multiple protected zones or entrances to the protected area is consistent with the Plan and provides adequate security. Therefore, the  !

Petitioner's concern regarding the adequacy of having one guard monitor several l

protected zones or entrances to the protected area was not substantiated. I The Petitioner states that the Licensee intends to reduce its security force at CR3, and on that basis, the Petitioner raises a concern that the reduction in the security force would compromise security at the plant. In a discussion with Licensee representatives on April 4,1996, the inspector confirmed that the Licensee intends to implement cost-saving measures that would employ new technology and result in a slight reduction in the number of security officers. The mere reduction in force does not indicate that plant security will be compromised. The Licensee must ensure that, notwithstanding its cost-saving measures, its plan and security staffing will meet NRC requirements and are adequate to protect public health and safety. The number of security officers the Licensee intends to utilize is required to, and will, meet the current commitments specified in the Licensee's Plan. If the Licensee decides to change the Plan commitments, it must identify the changes and submit them to NRC in accordance with NRC regulations. Therefore, the Staff finds that the Petitioner's concern regarding personnel reduction and its consequent effect on plant security is not substantiated.

IV. CONCLUSION The Petitioner's allegations have been partly substantiated. However, the NRC Staff concludes that these concerns do not warrant suspension or revocation of Florida Power's license to operate CR3. With respect to violations identified, the NRC is satisfied that the Licensee has taken appropriate action to correct the deficiencies. No further action based on concerns raised by the Petitioner is warranted. See Consolidated Edison Co. of New York (Indian Point, Units 1, 2, and 3), CLI-75-8,2 NRC 173,175 (1975); Washington Public Power Supply System (WPPSS Nuclear Project No. 2), DD-84-7,19 NRC 899,924 (1984).

Therefore, any further action on the issues addressed in this Director's Decision is not warranted and the Petitioner's request for suspension or revocation pursuant to section 2.202 is denied. As provided in 10 C.F.R. Q 2.206(c), a copy of this Director's Decision will be filed with the Secretary of the Commission for the Commission's review.

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As provided by this regulation, the Decision will constitute the final action of the Commission 25 days after issuance, unless the Commisson, on its own motion, institutes a review of the Decision within that time.

FOR THE NUCLEAR REGULATORY COMMISSION Frank J. Miraglia, Acting Director Office of Nuclear Reactor Regulation Dated at Rockville, Maryland, this 7th day of October 1996. ,

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! Cite as 44 NRC 187 (1996) DD-96-14 i l

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION-William T. Russell, Director

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s j in the Matter of Docket Nos. 50 413 4 50-414 t

(Ucense Nos. NFP-35 NPF-52) 1 j DUKE POWER COMPANY, et al.

j (Catawba Nuclear Station, Units 1 and 2) October 10,1996

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By a petition dated Ibbruary 13,1996, Charles Morris (Petitioner) requested that the U.S. Nuclear Regulatory Commission (NRC) suspend the operating

] licenses for the Catawba Nuclear Station (Catawba) and ten other unidentified licensees due to these plants lacking circuit breaker coordination. On May 1, 1996. Petitioner submitted an addendum to his retition, listing a number of cases i involving nine other nuclear power plants for which lack of protective device i coordination had been identified.

In a Director's Decision dated October 10, 1996, the Acting Director of

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Nuclear Reactor Regulation denied the relief sought by Petitioner. With regard to Catawba's lack of circuit breaker coordination, the Director concluded that i the Licensee had documented adequate technical justification for the lack of j such coordination. With regard to the other plants mentioned in the petition and addendum, the Director concluded that those cases had already been addressed I by way of the NRC's inspection report item closecut process.

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

1, INTRODUCTION I On February 13,1996, Mr. Charles Morris of Middletown, Magland, filed a i petition with the U.S. Nuclear Regulatory Commission (NRC) pursuant to Title j 10 of the Code of Federal Regulations, section 2.206 (10 C.F.R. 6 2.206). In the .

petition, the Petitioner requested the NRC to suspend the operating licenses for

! the Catawba Nuclear-Station and "some ten other licensees with uncoordinated l breakers" (not specifically identified in his initial petition) until the lack of j circuit breaker coordination has been remedied. Mr. Morris also requested that j

enforcement conferences be held on these cases and that Catawba be defueled.

l Mr. Morris also asked that the NRC take enforcement action against Catawba

{ for operating with a "known safety deficiency of which they did not inform the j +

NRC.".This aspect will be addressed separately as stated in the April 2,1996

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letter to Mr. Morris. On May 1,1996, Mr. Morris submitted an addendum to his

petition, providing a list of fourteen cases involving nine other nuclear power i plants for which lack of protective device coordination had been identified as a i concern by electrical distribution system functional inspection (EDSFI) teams;

, see Section 11 for information.

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II. DISCUSSION During an EDSFI conducted by the NRC Staff from January 13 to February I

- 14,1992, at the Catawba Nuclear Station, circuit breaker coordination deficien-l cies were ideritified for the 600-Vac essential motor control centers (MCCs) and

} the 125-Vdc system. His circuit breaker coordination issue was addressed in

! EDSFI inspection Report 50-413, 414/92-01, dated March 18,1992, as a de-

viation from a written commitment. Section 5.3.1 of the Institute of Electrical i and Electronics Engineers (IEEE) Standard 308-1974, "lEEE Standard Criteria

, for Class IE Power Systems for Nuclear Power Generating Stations," stipulates that protective devices shall be provided to limit the degradation of Class IE f- power systems. The Catawba Final Safety Analysis Report (FSAR) states that the system meets the requirements of this standard. He FSAR also states that j the protective devices on the 600-Vac essential auxiliary power (EPE) system l are set to achieve a selective tripping scheme so that a minimal amount of equip-ment is isolated for an adverse condition such as a fault.

4 Contrary to this IEEE standard, however, the Licensee's protective devices may not limit the degradation of the 125-Vdc vital instrumentation and control

(I&C) power system distribution center and other main feeder circuit breakers.

j An analysis performed by the Licensee showed that coordination did not exist 7

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for fault currents from 3500 amperes up to the maximum fault current of 9500 amperes. A fault on the battery charger feeder cable could cause both the charger and the battery to be isolated from the remainder of the distribution system and loads.

In addition, the outgoing feeder breakers for the 600-Vac essential MCCs have thermal elements and the incoming MCC breakers have instantaneous elements.

He incoming breaker (supply breaker) and the feeder breakers at each of the l 600-Vac MCCs were not coordinated for the maximum expected short-circuit current. A fault on any of the MCC outgoing feeders could cause the MCC incoming breakers to trip, resulting in a loss of the MCC.

l Enclosed with the letter dated April 16, 1992, Duke Power Company (the  !

Licensee) provided a response to this deviation that stated that the 125-Vdc l

vital I&C power (EPL) system primarily uses molded-case circuit breakers in the 125.Vdc distribution centers and power panclboards for protection. He

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battery, main, and tie breakers are equipped only with adjustable magnetic trip l units. He battery charger breaker is a thermal magnetic type with an adjustable l magnetic trip setting. He rest of the breakers are of a nonadjustable thermal magnetic type.

He Licensee's response concluded that this design was acceptable for the l following reasons:

1. He EPL system is not a shared system between the two Catawba units; I thus, a postulated fault in the EPL system of one unit will not affect the opposite unit.

2. He EPL system for each unit is composed of two completely redundant and separate trains, each consisting of two load channels for a total of four load channels per unit. A postulated fault would, at worst, disable two load channels of the same train, yet the redundant train would remain unaffected. ,

3. Selected loads such as the diesel load sequencer, essential switchgear and l load center controls, and auxiliary feedwater pump turbine controls are not only fed by the EPL system, but are auctioneered with the 125-Vdc diesel auxiliary power (EPQ) system. As a result, if the EPL system was unable to feed these loads, the EPQ system would supply them without interruption. Further, a fault on the EPL system will not affect the EPQ system or vice versa.

The Licensee's response further stares that the incoming 600-Vac breakers were incorporated in the design to provide a means of local isolation for the 600-Vac Class IE MCCs. He Licensee deemed acceptable the use of circuit breakers having a continuous rating equal to the MCC incoming rating and their instantaneous trip settings at maximum,10 times their continuous rating.

In the response to the deviation, the Licensee committed to perform a detailed study to identify acceptable methods to achieve improved protective device 189

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coordination within the EPL system and to evaluate the feasibility of climinating the incoming 600-Vac MCC breakers. The Licensee committed to either update the FSAR to justify the deviation from the IEEE Standard 308-1974 or to modify

, the system to meet this IEEE standard. Subsequent to completing the detailed

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study and evaluating the feasibility of making system modifications, the Licensee proposed modifying the FSAR.

Deterministic Analysis

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t To review and evaluate the lack of circuit breaker coordination in the l

Catawba EPL and EPE circuits, the Staff requested the Licensee to provide additional information. De Licensee's response of March 2,1994, addressed fault types, fault locations, breakers that are coordinated and breakers that are j not coordinated, the impact of the upstream breaker opening, and the safety significance of the loss of a train. The Staff also requested additionalinformation 1

regarding the 2-kV-rated interlocking armored cabling; the operating history of faults; the measures provided to detect, locate, and correct faults; and related criteria and practices incorporated to ensure continued system functional performance. He Licensee's responses to these requests were encicsed in its letter to the NRC of May 17, 1996.

1 125-Vdc Vital EPL System The EPL system is an ungrounded system and therefore can remain opera-tional for a single postulated fault of either positive-to-ground or negative-to-ground. In order to render the system inoperable, postulated faults would have to be either a simultaneous positive-to-ground and negative-to-ground fault or a double-line (positive-to-negative) fault. The former type of fault requires that two failures occur, which is beyond the design basis for the plant. He occur-rence of a single line-to-ground fault will not affect the functional capability of the power system. However, upon the occurrence of such a fault, a ground fault detector will alert the control room operator by way of an annunciator and a computer alarm. A program that seeks to maintain a dark control room annunciator board promptly addresses ground faults, he latter type of fault is thought to be unlikely in view of a study performed with information obtained from the Nuclear Plant Reliability Database System (NPRDS) and the Catawba probabilistic risk assessment (PRA). He Licensee analyzed failures at Catawba since 1985 and all U.S. plants since 1990. Bree reported cases were found in which a double-line fault occurred on a direct-current system. One case that occurred at Catawba involved a shorted lamp holder and was attributed to im-proper installation during maintenance. The two other cases occurred at nuclear 190

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chargers; in both of these other cases, the plant status was not affected. No cases were reported that involved double-line faults attributed to cable faults.

in addition, no faults of the types that could challenge the EPL system were identified in the NPRDS.

He Licensee's circuit breaker coordination analysis for the EPL system postulates faults at selected locations within the system. The analysis was performed in accordance with the guidelines of IEEE Standard 946-1993,IEEE Recommended Practice for the Design of DC Auxiliary Power Systems for Generating Stations," and included EPL system load groups A and D for both i

units. Rese two load groups for both units were analyzed since the 125-Vdc vital batteries associated with them are capable of producing the highest fault I current. The coordination analysis postulates faults at nine locations within I I

each of the four EPL load groups. These locations are as follows: (1) battery ,

charger output; (2) auctioneering diode assembly input; (3) inverter input; I (4) auctioneered distribution center bus; (5) load end of 4160Nac essential l

switchgear control power feeder breaker and first termination point of associated 1

feeder cable; (6) load end of 600Nac essential load center control power feeder breaker and first termination point of associated feeder cable; (7) load end of diesel generator load sequencer control power feeder breaker and first )

termination point of associated feeder cable; (8) power panelboard bus; and (9) load end of the largest breaker used in a power panclboard and the first i termination point of the associated feeder cable. These fault locations were chosen to represent a broad cross-section of possible fault locations. At these locations, calculated fault currents for the two A load groups (one A load group per unit) and the two B load groups are very similar, as may be expected since l l the two units are very similar. The analysis results also show that for faults at locations (2) and (4), the breakers are fully coordinated, while for faults

at locations (5), (6), (7), and (9), the breakers are partially coordinated. For

postulated faults at locations (1), (3), and (8), the breakers are not coordinated.

In the analysis, full breaker coordination is considered to exist if the breaker nearest the fault clears without operating (opening) any upstream breakers, or if

the consequences of operating an upstream breaker are no more severe than those associated with operating the breaker nearest the fault. Partial coordination is considered to exist if some of the upstream breakers, except the battery breaker or the load center incoming breaker, could operate before the breaker nearest the fault clears. For those cases in which either the battery compartment breaker

, or the load center breaker could operate before the breaker nearest the fault

, operates, coordination is considered not to exist. If an upstream breaker, such I as the load center incoming breaker, operates before the breaker nearest the fault opens, one of the four EPL system load centers would be lost.

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l The EPL circuit breaker coordination analysis neglects cable faults and credits cable resistances in the fault current calculations. He cabling used in the system is 2-kV-rated interlocking armored cable. This cabling has the same construction as nonarmored cable, except that a steel armor covering is applied around the entire outer circumference. This interlocked steel outer covering protects the cable from damage or degradation during loading, unloading, transporting, installation, and while in service at the plant. The cabling was purchased with an insulation system rated at 2000.Vac. The cable conductors were high-potential tested underwater and spark tested at the factory with values required by standards for 2-kV cable. The low voltage of the EPL system does not produce internal iomzation or corona that would cause an internal flashover or failure between conductors within the armored cable. Further, the cable insulation system has a greater thickness than the insulation system of standard 600-Vac-rated cable and therefore provides higher dielectric capability, enhanced physical protection, and added margin for aging considerations.

In addition, the Licensee had an interlocked armored cable fault test per-formed at the High Power Laboratory of the Westinghouse Electric Corpora-tion. This test did not result in any additional shorts between conductors within the multiconductor cable. Similar interlocking armored cabling is used at the Oconee Nuclear Station, which has an inservice cable monitoring program. For this program, six cable samples were installed inside one of the containment 1 buildings. At 5-year intervals, a 5-foot segment is removed from each cable sample for testing. This testing measures, documents, and trends the mechan- ,

ical and electrical properties of the cable. Past test results from this program l collectively show that cable samples are in good physical condition after 20 years in a reactor building environment. The installed interlocking armored ca- )

bling at Catawba is identical or superior to the cable that is installed at Oconee.

A similar monitoring program to evaluate and trend cable problems has been in place at Catawba since January 1995. He purpose of this program is to evaluate and record problems or malfunctions of plant cables and, if an adverse trend develops, take corrective actions to address the problem. Deficiencies that would be reported as a result of this program include short circuits, insulation damage, and problems with cable terminations and splices. Since cabling of the same basic specifications and ratings is used in both safety and nonsafety ap-plications at Catawba, all plant cabling is included in the scope of this trending program. Data on failures or problems with cables are collected at the end of each quarter; since January 1995 there has only been one failure.

Neither of the Catawba units has ever experienced a single line-to-ground fault that caused the EPL system to become inoperable. As noted previously, this result is due in part to the ungrounded system design. A complete review of the EPL system work order history revealed that five ground faults have been experienced in the last 5 years. Each of these faults resulted in an 192 l

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alarm both locally and in the control room and was caused by solenoid valve problems. %ree cases involved failed solenoid valve components, and the  ;

other two cases involved water intrusion into solenoids, which was subsequently corrected. Because of the intermittent nature and high resistance of these i

faults, it sometimes took an extensive ameunt of time to specifically locate and I correct the ground fault. However, none of these faults caused the EPL system  ;

to become functionally inoperable. The Licensee has implemented additional I measures to aggressively locate and correct ground faults that may occur in the future. Rese measures include the procurement of an advanced ground-locating device that will allow ground faults of a high-resistance nature to be located .

more readily. The EPL system work order history scarch also revealed that only l one ground fault detector has failed during the last 5 years. Because the original l

ground detector was no longer available from the manufacturer, a substitute part had to be located and an evaluation performed to verify its acceptability for use in the application. As a result, it took longer than normal to restore the unit to service. However, the EPL system is checked weekly in accordance with an administrative procedure for ground faults by way of another method that is independent of the ground detector system. Thus, in the unlikely event of a ground fault detector failure, a ground would very likely be detected by way of the independent alternate means before a fault-related problem developed.

To ensure continued functional performance of the EPL system, the following additional criteria and practices are in place at Catawba. Only a minimal amount of cable splicing is permitted, and no cable splicing is allowed in raceways. Safety-related cables routed underground are installed in conduit or cable trenches, and are not directly buried in the earth. Cable ampacities used for cables are based on 70% of the standard industry ampacity ratings. Further, for the EPL system, higher rated voltage (2000 Vac versus 125 Vac) cable is used with the steel interlocking armorjacket to provide additional physical protection.

Although the EPL system analysis described above demonstrates that full circuit breaker coordination does not exist for all postulated faults, this fact has no significance for the operational capabilities of the system because the faults that result in lack of breaker coordination are limited. These faults are limited in both type (doubled-sided, solid, low-resistance ones) and location (postulating such faults at many locations does not result in a lack of breaker coordination).

Monitoring by ground fault detectors further limits such faults since this activity minimizes the potential for bigger problems, such as positive-to-negative faults.

In the event that such a fault does result in the loss of an EPL load distribution center, an independent and redundant EPL load distribution center is provided to l supply safety-related loads. Further, should a fault-induced transient occur as a result of the loss of one of the two plant transient-inducing EPL load distribution centers, the plant can be safely shut down using only the loads powered from either one of the two EPQ system auctioneered distribution centers. In addition, 193 l

the safety significance of the loss of one EPL load group is analyzed in the Catawba FSAR. This analysis includes the loss of an EPL load group as a result

of any postulated cause. Thus, the loss of an EPL load group as a result of any cam (faults or any other cause) is within the licensing basis (i.e., analyzed in the FSAR) for Catawba Units I and 2.

l 600-Yac EPE System The Licensee also provided additional information on the lack of breaker coordination in the EPE system. This additional information included the analysis performed for the EPE system, fault locations, identification of the breakers that are coordinated and those that are not, the impact of upstream breakers opening, the significance of taking out an EPE train, and measures taken to prevent degrading the installed equipment during modification and maintenance work activities.

The fault current analysis for the EPE system was performed in accordance with the guidelines in IEEE Standard 141-1986, "IEEE Recommended Practice i for Electric Power Distribution for Industrial Plants." For each 600-Vac essential MCC, all load breakers and cables were reviewed to determine which circuit can produce the highest fault current. For each MCC, a coordination evaluation was performed for the worst-case feeder (load) breaker and the incoming (supply) breaker. In this analysis, the feeder breaker fault is modeled at the load or at J the first cable termination outside the MCC. For the fault current analysis, the

! normal load current for all nonfaulted feeder breaker loads is added to the feeder e

breaker fault current to establish the total current experienced by the incoming j

breaker during the fault. Also, in this analysis, the feeder breaker fault current i

, is obtained by adding the fault contribution from the incoming breaker and the I

fault contribution from the large motor loads connected to the bus. The fault currents were determined for both the normal and accident cases. The normal-

/J operation case produces the highest postulated fault current and, as such, is used throughout the analysis. The postulated faults in the analysis are three-phase, bolted faults, and all fault currents and load currents are based on the highest bus voltage for the normal operating case.

Fault locations for the Unit i Train A and B EPE MCC circuits were established. The Unit 2 Train A and B circuits are similar. Based on the i unlikely occurrence of bus faults and/or breaker faults at Catawba, faults were not postulated on the output of the feeder breaker. In addition, because of the 2-kV-rated interlocked armor cable protection and the fact that no faults have 4

occurred on any such cable in service at any of the Duke Power nuclear plants,  ;

faults were not postulated along the routes of the cable. Further, the fault current l calculations credit cable impedances and postulate faults at the input terminals ,

of the load or at the first cable termination after the cable leaves the MCCs. The

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2-kV-rated interlocking armored cabling esed in the EPE system is the same as that used in the EPL system. Thus, the cable analysis informatian previously mentioned for the EPL system is applicable to the EPE system.

He Unit i EPE system includes eleven MCCs. Analysis shows that for ten of these MCCs, the incoming breakers are coordinated for the worst-case postulated fault at the first cable termination outside the MCC. The remaining 4

MCC is provided with two incoming breakers, which can be powered from either a Unit I or a Unit 2 load center. The two incoming breakers supplying this MCC are not fully coordinated for a fault at the worst-case load, which is I a control room ventilation system air-handling unit. This unit is connected with i a 250-MCM cable that is 100 feet long. The other loads powered by this MCC i

are fed from smaller breakers and cables with lower maximum fault current and i

thus are coordinated with the incoming breakers.

The two incoming breakers for the one MCC are mechanically interlocked i such that one breaker is always locked in the open position. If the incoming breaker in service to this MCC trips to clear a fault, power is lost to some Train A control room ventilation system and nuclear service water system loads. An important function associated with these systems is maintaining pressurization of the control room. If this MCC is deenergized under nonaccident conditions, control room pressurization decreases until the operators manually transfer the system to Train B. This result is not viewed any differently than the result of losing the pressurizing fan alone and has little impact. If the MCC is deenergized under accident conditions, the design is such that pressurization is reestablished automatically from Train B, and this situation has little impact. l To ensure continued fault-free functional operation of the EPE system, modifications and maintenance work are controlled by station procedures. The Catawba inspection and maintenance procedure for MCC breakers addresses much of the work related to the EPE MCCs. His procedure, along with other station procedures, provides strict controls on any changes from the normal system configuration, such as placement of grounding jumpers or test alignments. These types of configuration changes are documented on a circuit alteration / restoration log sheet attached to the procedure. Before the work can be closed out and the equipment reenergized, the proper steps in the restoratian section of the procedure must be completed and verified 1 by an independent technician. Typical restoration activities performed at the completion of maintenance work on EPE MCC feeders include removing all test equipment and verifying that the MCC compartment is wired according to l

the latest wiring diagram. If required, motor phase rotation testing would also be j performed. If the feeder breaker has been removed or replaced, a thermography test of the energized breaker will be conducted. Additional specified functional l

j verification requirements, such as verifying proper full-speed operation and normal pressure and flow parameters, may be performed, depending on the l

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type of equipment involved with the work. In addition, the test requirements section of the inspection and mainyenance procedure for MCC breakers specifies that megger testing of the load is to be performed if a fault is suspected The procedure signoff sheet includes a section for recording such meggu moings.

The Licensee's March 2,1994 analysis indicated that selected circuit breakers associated with certain EPE MCCs are not coordinated for postulated faults.

Ilowever, the technical significance of this fact is low, which is due, in part, to such faults being limited in both type (bolted low-impedance faults) and location (postulating such faults in many EPE system locations does not result in lack of breaker coordination). Assurance that such faults are limited is further established by the positive test results obtained for the interlocking armored cabling and the strict adherence to maintenance procedures. In addition, an analysis of the loads powered by each of the eleven 600-Vac EPE system MCCs indicates that loss of power to any one of these MCCs because of a fault or for any other reason would not directly result in a reactor transient. Further, Trains A and B of the EPE system are redundant and, as such, loss of functions from any MCC is backed up by the redundant MCC of the other train. Finally, each MCC is provided with a control room alarm for loss of power to facilitate restoration of equipment in a timely manner by operator actions.

Probabilistic Risk Assessment To further supplement the deterministic engineering analysis results, the Staft requested the Licensee to consider using PRA techniques to better understand the likelihood and impact of the lack of breaker coordination in the Catawba EPL and EPE systems. The Licensee responded in the attachments to a letter dated December 29, 1994, by addressing EPL and EPE system uncoordinated breakers within a PRA framework. Ibliowing the review of the submitted PRA l information, the Staff requested by letter dated April 30,1996, that the Licensee specifically address the uncoordinated breaker issue including the (1) initiating event (IE) frequency; (2) conditional impact of the IE on plant operation; (3) ability to recover from an uncoordinated breaker event; and (4) recovery by way of the standby shutdown facility (SSF). The Licensee provided this additional PRA information in the enclosures to a letter dated May 17, 1996.

The paragraphs below discuss the PRA and the lack of breaker coordination in the EPL and EPE systems.

125 Vdc EPL System in the Catawba PRA, the Licensee identified a " Loss of Vital Instrumentation and Control" as an initiator-coded T14. With uncoordinated breakers, some line-196

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l to-line electrical faults in the 125-Vde feeders could cause both the loss of a vital I&C power distribution center (T14 initiator) and a subsequent turbine trip  :

and reactor trip.

l In Calculation CNC-1535.00-00-0007 enclosed in its December 29, 1994 letter, the Licensee established the frequency of the T14 initiating event at 5E-02 per year. His value had also been used in the Catawba PRA, which supported the Licensee's individual plant examination (IPE). The IE frequency had been based on the operational experience of one event in 20 reactor-years of operation at the combined Catawba and McGuire units (four units) from 1987 to 1991.

He event involved manual tripping of a 125.Vdc vital I&C power distribution center at the McGuire station in 1987. In response to this event, the NRC issued Information Notice 88-45, " Problems in Protective Relay and Circuit Breaker Coordination." Because no other Tl4 IE occurred since that time frame, the actual IE frequency would be lower.

In order to establish the fraction of the Tl4 initiator event frequency that could be associated with breaker miscoordination, the Licensee performed an NPRDS search for all de line-to-line faults. He data search included all U.S. nuclear plants from 1990 (Catawba since 1985) to the present. The NPRDS search identified only one such fau't at Catawba and three faults at all U.S. plants. In recognition of the fact that the results of NPRDS searches are dependent on j the search commands, the Staff requested the Oak Ridge National Laboratory i (ORNL) to perform a similar search. ORNL obtained the same results as did the  !

Licensee for the Duke Power plants. However, ORNL found a slightly higher j rate for the other U.S. plants. In no case did cable failure (s) result in a line.to-

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line fault or a plant trip. '

In order to estimate (bound) the contribution of a cable fault to the T14 initiator event frequency, the Licensee assumed that one cable fault occurred out of a combined 46 years of reactor operation at the Catawba and the McGuire units. This assumption resulted in a cable fault frequency of 2E-02 per unit-year. Catawba Unit I has about 18,500 cables and about 30 feeders per 125-Vdc vital distribution center. From these data, cable faults causing loss of a single distribution center have an IE frequency of 3E-05 per year ((2E-02)(30)/18,500 = 3E-05 per year). A second (somewhat higher) estimate was obtained by using the IEEE Standard 500-1984, "lEEE Guide to the Collection and Presentation of Electrical, Electronic, Sensing Component, and Mechanical Equipment Reliability Data for Nuclear Power Generating Stations," which specifies a composite cable failure rate of 7.54E-06 per hour per plant for power, control, and signal cables combined. Line-to-line cable failure rate is a small fraction of this rate. With this cable failure rate, the failure rate of a single distribution center is IE-04 per year ((7.54E-06)(8760)(30)/18,500 = IE-04 per year).

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'Ihe Catawba PRA used a generic value for bus fault probability of 2E-03 per 4 year, where the term bus fault includes distribution center or panel faults, cable faults, and terminal faults. Although this IE is only 4% of the T14 initiator frequency, it is obviously higher than the probability figures derived from plant operational experience and IEEE 500-1984 data (i.e., the cable fault contribution j was 5% of the bus fault probability using IEEE data, and 1.5% using operational experience). On the basis of this rationale, the Staff concluded that the cable fault contribution was bounded by the distribution center fault probability used in the Catawba PRA.

J Unit I has six 125-Vdc load distribution centers: lEDA,IEDB,IEDC, l 1EDD, IEDE, and IEDF. The Licensee evaluated the plant response on loss of power for each of the Unit I distribution centers. The Unit 2 system is similar to Unit 1, and the evaluation for Unit 1 is applicable to Unit 2.

The Licensee's evaluation indicates that a loss of power at IEDB or IEDC 4

would result in a loss of a vital I&C power 120-Vac inverter, one solid-state

protection system (SSPS) channel, one nuclear instrumentation channel, and a process protection channel. A loss of power at IEDA or IEDD would result in similar channel losses, plus a loss of power to process control for associated j pressurizer power-operated relief valves (PORVs), to control solenoids for certain main steam isolation valves, and to control solenoids for attendant main 4 feedwater control valves. However, except for the loss of the PORVs, a loss of any of these four distribution centers would not significantly impact the plant's accident mitigation capability. Loss of one channel of the SSPS, process protection channels, main steam isolation valves, and main feedwater control j valves would not preclude mitigation unless there were additional faults.

t Distribution center IEDE or IEDF provides control power for safety equip-ment. The Licensee's breaker coordination analysis indicates that the other four distribution centers lack full coordination. Distribution center IEDE is powered

by two power supplies that are auctioneered. One of these auctioneered power 1

supplies is from 1 EDA, and the other is from one of the trains of the 125-Vdc i EPQ system. Similarly, IEDF is powered by two power supplies that are auc-tioneered. One of these auctioneered power supplies is from lEDD and the other is from the other train of the 125-Vdc EPQ system. Thus, even though distribution centers IEDE and IEDF may be fed from uncoordinated distribu-tion centers lEDA and lEDD, respectively, in the event of loss of IEDA or lEDD, the distribution centers lEDE or IEDF will continue to be powered by the alternate power source. Further, a loss of power at lEDE or IEDF would not result in a plant transient and thus would not result in an immediate need for mitigating systems, although the resulting loss of control power to equip-j ment would require resolution within the specified time period of the applicable Technical Specifications Action Statement.

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In addition to redundant mitigation capability, Catawba is provided with a manually activated SSF. The SSF is an independent structure with its own ac and de power supplies, instrumentation, and reactor coolant makeup pump. Upon loss of normal ac or de power, the SSF can be used to remove core decay heat and provide reactor coolant pump seal protection if the event leads to the loss of all plant-side safety systems. The SSF reduces the contribution of the Tl4 initiators by more than an order of magnitude, resulting in a total contribution of 6.7E-08 per reactor-year, or less than 0.1% to the total core damage frequency (CDF).

Using a T14 IE frequency of SE-02 per year, the Licensee derived a total CDF  !

of 7.76E-05 per year in the Catawba IPE. Applying information from the IEEE i standard for cable fault frequency to the four distribution centers lacking full l coordination, which is a subset of the T14 initiator, reveals that the contribution to the total CDF from the loss of a 125-Vdc load distribution center is less i than IE-09 per reactor-year. The Licensee also performed a sensitivity study by changing the T14 IE frequency from SE-02 per year to 1.0 per year. The total CDF changed by 1.55% (i.e., the total CDF changed from 7.76E-05 per year to 7.88E-05 per year). The sensitivity study indicates that any increase in the CDF from a lack of breaker coordination would be small.

600-Vac EPE System As previously mentioned in this report, the Licensee's breaker coordination study indicates that out of eleven MCCs in the EPE system, only one MCC, IEMXG, is uncoordinated. This calculation, however, excluded all cable faults from the 600-Vac EPE system MCCs to the first cable termination on the basis that the occurrence of severe cable faults was of low probability. The Licensee states that no severe catste faults have been reported in its seven nuclear plants, which have a combined operational experience of 120 reactor-years. On the basis of the IEEE Standard 500-1984 data of 4.8 failures per million hours per plant for power cables, the Licensee calculated that a typical plant with 18,500 cables had a probability of a cable failure of 2.3E-06 per year per cable, and the probability of an MCC loss as a result of cable failure is 7E-05 per year for a typical MCC with 30 feeders.

In the Catawba PRA, loss of a 600-Vac MCC is addressed through its plant response characteristics (mission time) because the loss of an MCC does not cause a reactor transient. The Catawba PRA study identified a probability ofloss of a 600-Vac MCC as 1.5E-(M for a 24-hour mission time, and the contribution of cable faults to this mission time as SE-07. Therefore, the Catawba PRA indicates that cable faults did not have any significant impact on the overall MCC failure probability calculated in the PRA.

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l The Licensee's study revealed that a loss of any of the eleven 600-Vac EPE system MCCs would not directly lead to a reactor trip. In a review of the 600-Vac EPE system MCC loads, the Staff arrived at the same conclusion. Although such an MCC loss would not result in a reactor transient, it would render one train of safety systems inoperable and would require entry into applicable lim- '

iting conditions of operation defined in the Technical Specifications. However, l a loss of any MCC would only affect one train, and the redundant train would l

1 be available for accident mitigation. l The Licensee did not provide an analysis of the effect of SSF availability on the CDF from the loss of a 600-Vac MCC. The SSF response for the 600-Vac EPE system is expected to be similar to that previously explained herein for the EPL system.

In Calculation CNC-1535.00-00-0007, enclosed with the Licensee's letter of December 29, 1994, the Licensee indicated that on the basis of the Catawba PRA, the MCC IEMXG had a failure probability of 1.4E-04 for a 24-hour l mission time. Within this MCC, only one breaker feeding a control room l air-handling unit lacked coordination with its upstream breaker. With this uncoordinated breaker, the MCC failure rate would increase by lE-06 for a l 24-hour mission time, or the impact would be approximately two orders of l magnitude less than the total MCC failure probability. He Licensce's sensitivity  !

study provided in Calculation CNC-1535.00-00-0007 indicates that even if the i

failure rate of the uncoordinated MCC IEMXG were increased by an order of magnitude from IE-06 to IE-05, the resulting failure probability for the MCC IEMXG would increase by only 7.1%. l On the basis of these considerations, the Staff concluded that the lack of breaker coordination in the EPE system has a negligible impact on the MCC failure probability as calculated in the Catawba IPE.

Full circuit breaker caordination is a desirable design feature for ac and de power distribution systems in a nuclear plant since it assists in minimizing equipment losses if electrical faults occur. The Staff has reviewed the Licensee's submittals addressing the lack of full circuit breaker coordination within the 125-Vdc EPL and 600 Vac EPE systems. De Licensee's circuit breaker coordination analysis shows that the Catawba EPL and EPE systems lack full breaker coordination. However, the faults that must occur to cause a lack of breaker coordination in these systems are limited by type and location. Such faults have a low probability of occmTence because the interlocking armored

, cabling is unlikely to develop such faults. Further, ongoing measures, such as ground fault detection, incorporating design criteria and practices, and strict adherence to modification and maintenance procedures, tend to minimize the likelihood of the occurrence of faults within the EPL and EPE systems that would result in miscoordinated breaker <. Plant operational experience and IEEE Standard 500-1984 data indicate that line-to-line faults are of low probability.

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The probability of a line-to-line fault is 2E-02 per year and the probability of loss of a 125-Vdc distribution center is IE-04 per year. In the 600-Vac EPE MCCs, the Licensee has never experienced any severe cable fault in 120 reactor-years of operation of the seven Duke Power nuclear plants. He IEEE Standard 500-1984 data indicate a probability of a cable failure of 4.2E-02 per year end a corresponding probability of a loss of an MCC resulting from cable failure of 7E-05 per year. Rese results further support assumptions used in the Licensee's breaker coordination analysis. However, in the unlikely event that such faults should occur in an EPL or EPE system train, a redur. dant and separate train is provided to perform the safety function.

He Catawba SSF reduces the impact on CDF of a loss of either one of two i 125-Vdc distribution centers by more than an order of magnitude. Similar results l

would be expected for the 600-Vac EPE MCCs. In addition, a calculation by j the Licensee indicates that increasing the T14 IE frequency from SE-02 per year to 1.0 per year would increase the total CDF by 1.55% from 7.76E-M per year l to 7.88E-05 per year. A similar calculation for the 600-Vac MCCs indicates, that with lack of breaker coordination, the failure probability of the worst-case MCC would rise from 1.4E-04 per 24-hour mission time by IE-06 per 24-hour mission time. The Licensee's sensitivity study indicates that when the failure rate of the worst-case uncoordinated MCC was increased from lE-06 to IE-05, l

the resulting failure probability of the MCC would increase by 7.1%. Thus, the lack of circuit breaker coordination in the Catawba 125-Vdc EPL and 600-Vac EPE systems has a negligible impact on the CDF.

On the basis of this information, the Staff concludes that the Licensee has I documented adequate technical justification for the lack of breaker coordination in the Catawba 125-Vdc EPL and the 600-Vac EPE systems. Accordingly, the Staff concludes that there is no basis to suspend the Catawba operating licenses.

The Staff will pursue separately the requirement for the Licensee to bring the FSAR into conformance with the as-built plant.

l Lack of Protective Desice Coordination at Other Nuclear Plants As previously indicated in the Introduction section of this Decision, the Petitioner submitted an addendum to his petition on May 1,1996. His addendum included a list of fourteen cases, involving nine other nuclear power plants, in which lack of protective device coordination was identified as a concern by EDSFI teams. These fourteen cases were addressed by way of the NRC's inspection report item closcout process. As documented in the publicly available closcout inspection reports, these cases were resolved by (1) additional calculations and analyses showing that protective device coordination exists, and/or (2) plant hardware modifications such as replacement circuit breakers or fuses. The following list identifies each of these fourteen cases by an EDSFI 201

inspection followup item (IF1) number and the publicly available inspection teport in which the lack of protective device coordination issue was closed out.

Closcout i EDSFI IFI Report Insp. Report l Plant Name Number Date Report Date

1. Oyster Creek 219/92-80-11 7/9/92 94-01 3/10/94

2. Nine Mile Point 1 220/91-80-07 1/10/92 94-20 11/4/94

3. Nine Mile Point i 220/91-80-07 A 1/10/92 94-20 11/4/94

4. Nine Mile Point 1 220/9180-07B 1/10/92 94-20 11/4/94

5. Nine Mile Point 1 220/9180-07C 1/10/92 94-20 11/4/94

6. Dresden 237/91-201-05 9/20/91 92-21 10/8/92

7. Quad Cities 254/91-011-09A 6/24/91 94-26 12/5/94

8. Quad Cities 254/91-0119B 6/24/91 94-26 12/5/94

9. Quad Cities 254/91-0Il-9C 6/24/91 94-26 12/5/94 ,

10. Hatch 321/91-202-07 8/22/91 93-19 11/2/93 '

i 1. McGuire 369/91-09-01 2/19/91 94-20 10/12/94

12. Ibrt Calhoun 285/91-01-03 5/20/91 92-30 12/31/92

13. WNP2 397/92-01-20 5/5N2 93-16 6/4/93

14. Beaver Valley 2 412/91-80-02 4/1/92 93-27 1/24/94 Ill. CONCLUSION l The institution of proceedings in response to a request pursuant to section 2.206 is appropriate only when substantial health and safety issues have been raised. See Consolidated Edison Co. of New York (Indian Point, Units 1, 2, and 3). CL1-75-8, 2 NRC 173,176 (1975), and Washington Public Power j Supply System (WPPSS Nuclear Project No. 2), DD-84-7,19 NRC 899,923 i (1984). This standard has been applied to the concerns raised by the Petitioner j to determine if the action he requested is warranted, and the NRC Staff finds ]

=

no basis for taking such actions. Rather, as previously explained herein, the NRC Staff believes that the Petitioner has not raised any substantial health and f safety issues. Accordingly, the Petitioner's request for action pursuant to section  ;

, 2.206, as specifically stated in his letter of February 13,1996, and supplemented t by a letter dated May 1,1996, is denied. 1 A copy of this Director's Decision will be filed with the Secretary of '

s the Commission for the Commission's review in accordance with 10 C.F.R. 5 2.206(c). This Decision will become the final action of the Com. mission 25 202

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

FOR THE NUCLEAR REGULATORY COMMISSION I

Rank J. Miraglia, Acting Director Office of Nuclear Reactor l Regulation 1

Dated at Rockville, Maryland, this 10th day of October 1996.

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l l Cite as 44 NRC 204 (1996) DD-96-15 l

UNITED STATES OF AMERICA l NUCLEAR REGULATORY COMMISSION l

i l OFFICE OF NUCLEAR REACTOR REGULATION Frank J. Miraglia, Acting Director l

l In the Matter of Docket Nos. 50-440-A l 50-346-A i

! CLEVELAND ELECTRIC ILLUMINATING COMPANY  ;

(Perry Nuclear Power Plant, l Unit 1; Davis-Besse Nuclear i

!' Power Station, Unit 1) October 17,1996 l

l In a petition, dated January 23,1996, and supplemented by letters dated May l 31, and August 13,1996, the City of Cleveland, Ohio, which owns and operates j Cleveland Public Power, requested the Executive Director for Operations to take

( enforcement action against the Cleveland Electric Illuminating Company for i allegedly violating the antitrust license conditions applicable to its nuclear units. l The petition, which raised four specific issues, was referred to the Director, Office of Nuclear Reactor Regulation, for review pursuant to 10 C.F.R. 5 2.206.

In a Director's Decision issued on October 17, 1996, the Acting Director of Nuclear Reactor Regulation determined that no NRC proceeding should be instituted and no further regulatory action by the NRC is required regarding the issues raised by Petition;r. The Acting Director concluded that the matters raised l were either effectively resolved by the Federal Energy Regulatory Commission (FERC) or are pending before FERC and are within its jurisdiction to decide; j and the Petitioner otherwise failed to show it had been harmed.

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

1. INTRODUCTION The City of Cleveland, Ohio, which owns and operates Cleveland Public Power (CPP or the City), in a petition, dated January 23,1996, requested the Executive Director for Operations of the U.S. Nuclear Regulatory Commission (NRC or the Commission) to take enforcement action against the Cleveland Electric Illuminating Company (CEI) for allegedly violating the antitrust license conditions applicable to its nuclear units. The petition was referred to the Director, Office of Nuclear Reactor Regulation, for review.

CPP requested that NRC, on an expedited basis, (1) declare that CEI is obligated to provide the wheeling and interconnection services specified in the petition; (2) issue a Notice of Violation related to that obligation; (3) impose a requirement by order directing CEI to reply in writing and admit or deny violation of that obligation and setting fonh the steps it is taking to comply with the antitrust license conditions; (4) impose a requirement by order directing CEI to comply with the portions of the antitrust license conditions at issue and directing CEI to withdraw from the Federal Energy Regulatory Commission (FERC) portions of its filings in Docket No. ER93-471-000, as specified in the petition, which are contrary to CEI's obligations under the antitrust license conditions, including withdrawal of the deviation charge from rate schedules and withdrawal of that portion of the " Operating Agreement" that provides Toledo Edison highest priority treatment; and (5) impose civil monetary penalties for CEI's violations of the license conditions.

Four specific violations of the antitrust license conditions are alleged in the i City's section 2.206 petition. The first allegation is that CEI has violated i l License Condition No. 3, concerning wheeling service, by refusing to provide l 40 MW of firm wheeling service from Ohio Power Company to CPP to provide l electrical service to Medical Center Company (Medco), a former CEI retail j customer. The second allegation is that CEI has violated License Condition l l Nos. 6 and 11,5 which concern the sale of emergency power, by contracting in I the 1987 "Centerior Dispatch Operating Agreement" to provide Toledo Edison l Company emergency power on a preferential basis. The third allegation is i that CEI has violated License Condition No. 2, concerning the offering of interconnections upon reasonable terms and conditions, by failing to offer CPP a fourth interconnection point. The fourth allegation is that CEI has violated License Condition No. 2 hy imposing unreasonable deviation charges I

License Condinon No 11, wh:4 concerns wholesale power and coordinauon seruces. is rnennoned m the introductory poruon ohbe peuuon, but no argurnent is provided to support the claim nor is this condinon otherwise nwnuoned in any substantne discussmn an the peuuan l 205 I

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i for unscheduled power delivered over existing interconnections in excess of the amount scheduled for delivery.

CEI responded to the City of Cleveland's petition in a letter dated May 6, l 1996, stating that the allegations should be dismissed not only because they lack merit but also because they relate to matters currently under FERC consideration.

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l IL BACKGROUND On the basis of the record developed during the antitrust hearings of Davis-Besse and Perry, an NRC Atomic Safety and Licensing Board found, in a decision dated January 6,1977, that CEI and the other applicants engaged in activity that was inconsistent with the antitrust laws, Toledo Edison Co. (Davis.

Besse Nuclear Power Station, Units 1,2, and 3), LBP-77-1,5 NRC 133 (1977),

I aff'd with modifications, ALAB-560,10 NRC 265 (1979). The Board also found that because the municipal system of Cleveland was isolated electrically from l utilities other than CEI, and was able to obtain only emergency power from CEI, it was essential, in order for CPP to remain a viable competitor, that Cleveland I have power wheeled to it over CEI's transmission system. The Board noted that CPP was unable to obtain wheeling service because CEI would not agree to third-party wheeling on any terms. 'Ihe Board concluded that failure to exercise its authority under the Atomic Energy Act to issue license conditions would result in a continuation of this anticompetitive conduct. CEI, as an applicant.

was ordered to implement the following license condition (No. 3):

Appheants shall engage in whechng for and at the request of other entities [any electne genen. tion and/or distnbution system or municipahty or cooperative with a statutory nght or privilege to engage in either of these functions] in the CCCT [ Combined CAPCO Territories):

(a) of electric energy from dehvery points of apphcants to the ennty(ies), and, 1

1 (b) of power generated by or available to the other entity, as a resuk of its ownership or entitlements (includes but is not hmited to power made available to an entity pursuant to an exchange agreement) in generating facihties, to delivery points of Applicants designated by the other entity.

Such whechng services shall be available with respect to any unund capacity on the transmission lines of Applicants, the use of which will not jeopardize Apphcants' system. In the event Apphcants must reduce wheehng services to other entities due to lack of capacity, such reduction shall not be effected until reductions of at least 5% have been made in transmission capacity allocations to other Applicants in these proceedings and thereafter shall be made in proportion to reductions imposed upon other Applicants to this proceedmg.

Applicants shall make reasonable provisions for dwlmed transmission requirements of other entities in the CCCT in planning future transmission either indnidually or within the CAPCO l grouping. By " disclosed" is meant the giving of reasonable advance notification of future l requirements by entities utilinng whechng services to be made available by Apphcants.

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Ten other antitrust license conditions were added to the Davis Besse and Perry licenses covering the sale of wholesale power; the offering of interconnections; the sale of economy energy, maintenance power, and emergency power; access to ownership shares in the nuclear units; the sharing of reserves; and the provision of coordination services. NRC ordered that these conditions be implemented in a manner consistent with the provisions of the itderal Power Act. ALAB-560, supra,10 NRC at 295-99.

Since the late 1970s, CPP, the City of Cleveland's municipal power system,

, has sought greater access to the CEI transmission grid. CPP has its own distri-bution system and generates a portion of its own power supply requirements. To l seek out the most cost-efficient source of power supply, CPP needs meaningful

)

I access to transmission facilities serving the local area, which are owned by CEI. '

III. DISCUSSION j I

CPP alleges four specific violations of the antitrust license conditions. The first allegation is that CEI violated License Condition No. 3 by refusing to provide firm wheeling service to CPP. This allegation is the result of one disputed transaction, CEI's refusal to wheel 40 MW from Ohio Power Company to CPP l

to service Medco, currently a CEI retail customer. CPP claims that Medeo I has decided to become a native load customer of CPP and that there is no credible basis upon which to contend that the transaction at issue constitutes retail wheeling. CPP claims that there was no request for CEI to provide retail wheeling services, and the requested 40-MW wholesale purchase from Ohio Power is to serve CPP's native load. CPP alleges tnat CEI is attempting to delay the loss of a significant retail customer.

CEI responds to the allegation by stating that the written contract between j CPP and Medeo reflects a direct pass-through of CPP payments to Ohio Power. l CEI further claims that CPP is acting as a strawman to facilitate retail wheeling of power from Ohio Power to Medco. CEI contends that the transactions are shams designed to circumvent prohibitions in the Federal Power Act il212(g) and 212(h), against retail wheeling. Section 212(g) prohibits issuing orders under the Ibderal Power Act that are inconsistent with any state law that governs the retail marketing areas of electric utilities. Section 212(h) prohibits mandatory retail wheeling and sham wholesale transactions.

Two FERC proceedings are in progress concerning CEl's refusal to transmit the Ohio Power purchase: a CEI petition filed November 2,1995, requesting a ruling that CEI is not required to provide the requested service under the Federal Power Act il 211 or 212 (Docket No. EL96-9-000), and a CPP complaint filed November 29,1995, concerning CEI's refusal to transmit the Ohio Power purchase (Docket No. EL96-21-000).

207

On July 31, 1996, FERC issued an order in connection with the wheeling transaction raised in the City of Cleveland's 2.206 petition. FERC decided in favor of the City and found that CEI is obligated under the existing transmission service agreement to provide the requested transmission service and that the service did not violate the Federal Power Act. Since the transmission will be over CEl's lines to Cleveland and the sale to Medco will be over Cleveland's 138-kV line, FERC found that this case did not involve the transmission of electric energy by CEI directly to an ultimate consumer, that is, there was no

" sham" transaction. '

In a letter to the NRC dated August 8,1996, counsel for CEI stated that, based on the FERC decision, a signed service agreement reserving 40 MW of firm transmission service for the requested period September 1 through December 31, 1996, has been forwarded to the City of Cleveland. In a letter to the NRC dated August 13,1996, CPP's counsel urged the imposition of sanctions, even in light of the FERC decision, stating that "CEl's expressed willingness (August 8 letter) to comply now with its wheeling obligations does not excuse the Company's unwarranted refusal to wheel absent a directive from a federal agency." Counsel for CEI responded in an August 21,1996 letter that "CEI sought declaratory ruling on the appropriateness of this request promptly enough to obtain a determination without impacting the September I service date." CEI l

agreed to a subsequent CPP request after the FERC order and transmission service began on August 17, 1996. CEI's counsel further stated that as a result. CEI's actwns have not resulted in any loss of transtnission services to the City of 3 Cleveland. In essence. the City of Cleveland is asking for the imposition of penalties solely because CEI exercised appropnate legal procedures to determine the propnety of the service request. Such appropriate process cannot and should not be the basis for any sanctions.

In a letter to the NRC dated Se ptember 23,1996, counsel for CEI forwarded an opinion of the Ohio Supreme Court holding that the Public Utility Com-mission of Ohio (PUCO) has junsdiction to consider CEI's complaint that the Medco transaction violated the Ohio Certified Territory Act and directing PUCO to do so. The September 23,1996 letter also forwarded CEI's request for re-hearing of the FERC decision in the Medeo transaction, stating that while CEI continues to exercise its legal rights to determine the legality of the transaction, CEI would continue to honor the service agreement that it executed after the FERC decision.

'Ihe FERC order directing CEI to provide the requested transm'.ssion service effectively resolves the first issue in the 2.206 petition. Sanctions are not warranted when a licensee pursues legal procedures to resolve a disputed request for transmission service. For this reason, I am denying CPP's section 2.206 l request for an enforcement action against CEI on this first issue.

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The second issue raised by CPP alleges that CEI violated License Condition No. 6 by contracting with Toledo Edison Company to provide emergency power on a preferential basis 2 CPP objects to language in the 1987 Centerior Dispatch Operating Agreement that states that CEI and Toledo Edison (collectively

" Operating Companies") "will assign highest priority to provide each other emergency power. An Operating Company will terminate an existing emergency supply to an outside utility in order to honor a request for emergency power from an Operating Company." There is also similar priority language concerning sales of short-term power CPP has also brought this issue before FERC.

CEI's response to the second issue states that the operation of Toledo Edison and CEI as an integrated system under Centerior necessarily requires them to provide power to each other as an internal system. CEI further states that this is not an act of anticompetitive discrimination but the workings of an integrated system required by the Securities and Exchange Commission. CEI claims that CPP is treated no differently from any other outside entity and has suffered absolutely no injury from the provisions and asserts that CPP has never been denied short-term or emergency power. CEI states that it has sold and will continue to sell ernergency power to CPP on an as-needed basis and has never refused to provide emergency service when it had it available on its system.

CEI further stated that it was not aware of any instance in which short-term or emergency power was provided to CPP under terms less favorable than those to other utilities outside the Centerior system. CEI concluded that it has honored both the letter and the spirit of License Condition No. 6.5 As to the second issue, CPP has not shown that it has been harmed or could be harmed by the language in the Centerior Dispatch Operating Agreement.

Under the agreement, Toledo Edison and CEI are affiliated in that they are part of an integrated Centerior system. CPP has not shown that it has been treated differently than other outside (n?tiaffiliated) utilities, or that it has been denied access to emergency or short term power. In any event, CPP has brought its concems about the operating agreement before the FERC. For these reasons, no action by the NRC is warranted, and I am denying CPP's section 2.206 request for enforcement action against CEI on this second issue.

The third issue raised by CPP alleges that CEI has violated License Condition No. 2 by failing to offer CPP a fourth interconnection point. License Condition i No. 2 requires that CEI (and the other applicants) shall offer interconnections on reasonable terms and conditions at the request of any other local electric l

2 Speci6cany. Licenw Conditan No 6 requires CEI so sell ernergency power to requestmg enuties upon terms and comhtions no less favorable than those Apphcants make available: (a) to each other pursuant to the Central Area Power Coordmation Group (CAPCo) agreements or pursuant to bilateral contract; or (b) to n m-Appbcant enstlies outs Je the Combined CAPCo Company Terrisones.

'See vote 2. above 209

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l entities.d CPP states that a fourth interconnection point is needed to provide l reliable service to the west side of Cleveland. CPP states that the current transfer capability limit is expected to be exceeded within 2 years. CEI previously committed to permit a fourth interconnection in a letter dated September 19, l 1985, from CEI's chairman to the Mayor of Cleveland, which acknowledged l the requests for the third and fourth interconnections; and in exchange for Cleveland's agreement not to oppose the CEI merger with Toledo Edison, 1 CEI committed to concur in CPP's request for FERC approval of the two interconnections. CPP alleges that CEI has refused CPP's request for installation of a fourth interconnection.

A CPP complaint was filed with FERC in April 1993. On June 9,1995, FERC issued an order directing CEI to provide a fourth interconnection and to file with,FERC the proposed charges for the interconnection. The decision by FERC found that the letter of September 19,1985, a 1985 contract between CEl, Toledo Edison, and American Municipal Power-Ohio, and the license conditions 1 all supported the issuance of the order requiring the fourth interconnection.

CEI responded to the third issue by stating that it has complied with License Condition No. 2 by installing and maintaining three prior interconnections, sufficient to meet all of CPP's current needs, and by working toward the installation of a fourth interconnection. CEI claims it has not refused the '

fourth interconnection but instead has expended significant effort to establish i reasonable terms for the interconnection and to ensure that it is compatible in I terms of safety and reliability with CEI's system. CEI has filed suit in the Ohio Court of Common Pleas to require CPP to comply with engineering and utility I industry standards in its construction projects. CEI further claims that CPP ,

admitted in a separate lawsuit that its system does not meet applicable codes j t

and standards. On July 7,1995, CEI sought a rehearing on the FERC order to proceed with the fourth interconnection. CEI states that the rehearing was sought on the FERC order for two reasons: (1) CEI believes that the order should not

, have been issued without findings that the interconnection was warranted under l

sections 202(b) and 210 of the Federal Power Act and (2) CEI has indicated that a number of technical issues and safety and reliability concerns need to be l resolved before the interconnection can be installed.

The issue of whether CEI is required to provide a fourth interconnection was resolved with the FERC order of June 9,1995, directing CEI tn proceed with the interconnection (71 FERC 161,324). The unresolved technical, safety, and l reliability issues raised in CEl's appeal of the FERC orda will be resolved in 4

specifnally, license Condmon No. 2 requires CEI to offer meerconnections upon reasonabic terms and conditions at the request of any other electric ennues in its sernce area, with due regard for any necessary and apphcable safety procedures.

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the FERC rehearing process. For these reasons, I am denying CPP's section I 2.2% request for enforcement action against CEI on this third issue.

l The fourth and final allegation raised by CPP is that CEI has violated License ,

Condition No. 25 by imposing unreasonable deviation charges for ur wcled l l power delivered in excess of the amount CPP had scheduled fo deli ery.

l CPP states that in March 1993, CEI unilaterally filed with FERC rgosed l amendments to the 1975 Interconnection Agreement. One amendment added a

! requirement that CPP pay a deviation charge of $75 per kW-month for the I

maximum number of kilowatts of power delivered by CEI in any hour in j excess of the amount scheduled by CPP for that hour. Another amendment '

covers overscheduling of power supplies by CPP and allows CEI to retain the excess energy for its own use while paying CPP a rate equal to half of CEI's fuel cost for that excess power. CPP alleges that the deviation charges are l discriminatory and represent an anticompetitive restriction on CPP's right to j obtain interconnections on reasonable terms. CPP claims that these provisions l apply to all deviations above and below zero, no matter how insignificant.

l CPP alleges that the failure to utilize a deadband approach with no charges for small deviations from scheduled power to recognize the impossibility of zero deviations, is contrary to standard industry practice. CPP states that the l deviation charges are anticompetitive in that CPP is the only utility against l which the deviation charges would be imposed and also the only utility in direct competition with CEI.

CEI's response to the fourth issue states that this allegation distorts the meaning of License Condition No. 2, which relates to the installation of interconnections upon reasonable terms and conditions, not incentives that CEI proposes to FERC to encourage CPP to minimize unscheduled power deliveries from CEI.

l A FERC administrative law judge (AU) issued an initial decision on the

! issue of the deviation charges on November 28, 1994. CPP's arguments opposing CEI's compensation proposal (of half ofits then-current fuel charge for deviations below that scheduled) were rejected by the AU. Tne AU's decision also upheld the imposition of a deviation charge for power supplied in excess of that scheduled by CPP, but reduced the amount from $75 per kW-month to

$25 per kW-month. The decision also rejected CPP's proposed 6% deadband, finding "no reason appears why any deadband should be adopted for the purposes of this decision."

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S Sec note 4, above.

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The issues raised by CPP in this fourth allegation are primarily tariff-related issues and fall clearly under the jurisdiction of FERC.6 The final FERC decision in this matter will resolve the issues, and any excess amounts paid by CPP will be refunded with interest in accordance with FERC regulations. For these reasons, I am denying CPP's section 2.206 request for an enforcement action against CEI on this fourth issue.

i IV. CONCLUSION 4

I have concluded that FERC's order requiring CEI to provide the requested wheeling transmission service in the Medeo transaction effectively resolves the first issue raised in CPP's section 2.206 petition and request for action by NRC.

In regard to the second issue concerning CEI's contracting with Toledo Edi.

son Company to provide emergency power on a preferential basis, CPP has not shown that it had been harmed or could be harmed as a result of the language in the Centerior Dispatch Operating Agreement. Nor has CPP shown that it j has been treated differently than any other outside (nonaffiliated) utilities. This matter is also the subject of a FERC proceeding. I am therefore denying CPP's section 2.206 request for enforcement action against CEI on this second issue. I have concluded with respect to the third issue concerning CEl's alleged refusal to offer a fourth interconnection that the FERC order of June 9,1995, effectively resolves this issue by ordering CEI to provide the fourth interconnection, and that I the unresolved issues raised in CEI's appeal of the FERC order will be resolved I in the rehearing process. I have concluded that the fourth issue raised concerning devia ion charges for unscheduled power deliveries is primarily a tariff.related issue and falls clearly under the jurisdiction of FERC. The initial decision by the ALJ in this case addressed each of the concerns raised in this fourth issue.

The final FERC decision in this matter will resolve these issues, and any ex-cess amounts paid by CPP will be refunded with interest in accordance with FERC regulations. I have concluded that no enforcement action is warranted for this fourth issue. As a result of the foregoing, I have determined that no NRC l

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• As indicated in Flonda femer and Lght Co (st 1.ucie Nuclear Power Plant. Urus 2). DD 81 15,14 NRC 589 (1981), issues of terms used in heense conditions raised before rERC "will not institute a requested proceeding where the petinoner's basis for relief rests on resolution of an issue that is pending before another agency and that is pecuharly wittun the cornpetence of that agency to decide " The staff connnues to ernploy the concept of

" watchful deference" when an issue is before l'ERC. See Flunda P<=er and ught Co. (se t.ucie Nuclear Power Plant. Unit 2). DD.9510. 41 NRC 361 (1995).

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1 proceeding should be instituted and no further regulatory action by the NRC is required.

FOR THE NUCLEAR 1 REGULATORY COMMISSION l

Frank 1 Miraglia, Acting Director Office of Nuclear Reactor Regulation Dated at Rockville, Maryland, this 17th day of October 1996.

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Cite as 44 NRC 214 (1996) DD-96-16 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION l

Ashok C. Thadani, Acting Director in the Matter of Docket No. 50-245 (License No. DPR-21)

NORTrtEAST NUCLEAR ENERGY COMPANY (Millstone Nuclear Power Station, Unit 1) October 31,1996 The Acting Director of Nuclear Reactor Regulation has denied a petition by

( Anthony J. Ross that enforcement action be taken against Northeast Utilities and certain managers for violations involving the gas turbine battery, harassment and intimidation, and falsification of nuclear documents. Following his assessment

of the petition, the Acting Director concluded that appropriate enforcement I

action had already been taken for certain of the Petitioner's concerns while l other concerns were not substantiated so that additional enforcement action was not warranted and the petition should be denied.

Technical issue discussed: maintenance and surveillance.

l DIRECTOR'S DECISION UNDER 10 C.F.R. $ 2.206 l I. INTRODUCTION On January 2,1995, Mr. Anthony J. Ross (Petitioner) filed a petition with the Executive Director for Operations of the Nuclear Regulatory Commission (NRC) pursuant to section 2.2% of Title 10 of the Code of Federal Regulations (10 C.F.R. 5 2.206). In the petition, the Petitioner raised concerns regarding (1) employee harassment and intimidation by Northeast Utilities (NU); (2) the falsification of nuclear documents concerning the gas turbine battery; (3) failure 214

to enter a Technical Specification Limiting Condition for Operation (LCO) after a failed surveillance; and (4) his belief that numerous violations have occurred in 1992 and 1993 regarding the gas turbine battery. Because of these problems, the Petitioner alleges that the gas :urbine is still inoperable. In addition, the Petitioner asserts that these problems have not been handled appropriately by the NRC and NU, and that NU and the NRC are engaged in an apparent " coverup" of problems with surveillances of the gas turbine battery.

The Petitioner requested that the NRC (1) assess a Severity Level II violation and a Severity Level III violation against his department manager and his first-

)

line supervisor for their apparent violations of 10 C.F.R. 9 50.7; (2) institute sanctions against the Petitioner's first line supervisor, NU, and the Millstone Unit 1 organization for engaging in deliberate misconduct in violation of 10 C.F.R. 6 50.5; arid (3) remove the Petitioner's first-line supervisor fram his position until a " satisfactory solution to the falsifying of nuclear documents" by this individual can be achieved. l On February 23, 1995, I informed the Petitioner that the petition had been teferred to me pursuant to section 2.206 of the Commission's regulations. I also informed the Petitioner that the NRC would take appropriate action within a reasonable time regarding the specific concerns raised in the petition. I also stated that the Petitioner's allegations that the NRC has not been appropriately handling certain violations and is engaged in a " coverup" of the problems related to the gas turbine battery had been referred to the Office of the Inspector General (OIG). Therefore, this Director's Decision does not address that issue. On the basis of a review of the remaining issues raised by the Petitioner, as discussed i below, I have concluded that no substantial health and safety issues have been raised that would require the initiation of additional formal enforcement action. I II. DISCUSSION A. Ilackground The Petitioner alleges that during an annual surveillance of the gas turbine battery on September 20,1994, he identified that some of the intercell bolted connections of the gas turbine battery were greater that 65 micro-ohms, which was greater than the acceptance criteria specified in Procedure SP 779.5, " Gas Turbine Battery Annual Inspection." The Petitioner alleges that although he notified the Operations Department shift supervisor and his first-line supervisor, his first line supervisor signed the surveillance as "yes," refetTing ta the "accep-tance criteria met," when clearly the requirements were not met as specified by Procedure SP 779.5. The Petitioner alleges further that, when the Operations Department was notified by him of the failed surveillance, the Millstone Unit i organization willfully failed to enter a 4-day LCO as required by the Technical 215

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Specifications, in order to keep the unit on-line to produce revenues. In addition, {

the Petitioner asserts that about a week after this incident, he received copies l

of the 1992 and 1993 annual gas turbine battery surveillances that indicated a j number of problems and violations that have not been handled appropriately by l

NU and the NRC, and that the gas turbine is still inoperable due to these prob- '

lems. Finally, the Petitioner alleges that he has been subjected to harassment

)

and intimidation by his first-line supervisor and department manager for raising these concerns. l B. Petitioner's Concern Regarding Falsification of Nuclear Documents During an inspection held September 27 through November 15,1994, as documented in Inspection Report (IR) 50-245/94-31,50-336/94-30,50-423/94- l 28 (IR 94-31), dated December 16,1994, and an inspection held May 15 through I June 23,1995, as documented in IR 50-245/95-22, 50-336/95-22, 50-423/95- I 22 (IR 95-22), dated July 21, 1995, the NRC reviewed gas turbine battery maintenance and surveillance activities at Millstone Unit 1. The inspection  ;

! determined that on September 20, 1994, the date the Petitioner alleges the gas turbine battery failed the surveillance, the Licensee for Millstone Unit I (Northeast Nuclear Energy Company - NNECO) performed the annual surveillance of the gas turbine battery as specified by Procedure SP 779.5. This j annual preventive maintenance identified three intercell connection resistance i readings that did not meet the surveillance acceptance criterion in that the resistance readings were greater than the accepted values. The electricians i notified the shift supervisor and the maintenance foreman of the unsatisfactory readings and documented the results in the surveillance procedure, The NRC reviewed the completed surveillance and noted that the " acceptance criteria met" block was checked "yes," indicating satisfactory surveillance results; however, the resistance readings for the three intercel; connections

] were documented as unsatisfactory. The inspection therefore confirmed that the classification of this surveillance as acceptable was incorrect and, as a result, it bypassed NNECO's administrative control procedures for system operability,' and procedural review and approval However, on the basis of interviews and a review of the completed surveillance procedure, the NRC determined that the first-line supervisor documented the high resistance readings on the cover page of the surveillance, discussed the issue with the Electrical Engineering Department to determine if the high resistance readings affected operability of the battery and, on the basis of the discussion with Engineering, 3

If the classificanon of the surveillance had been deternuned to be "unsausfactory" r* acceptance entena block" checked *no"), a dciernunation of operabihty would be perforned and the related Techmcal Specificauon tro would be entered,if the gas turtune battery was inoperable.

216

l l . determined that Engineering had previously reviewed the effect of the high resistance readings and had found the battery operable. Therefore, the first-line supervisor concluded that the battery was acceptable as is.2 Further, the inspection confirmed that the Licensee's previous operability evaluation was l acceptable and that the gas turbine battery was operable. As discussed below, the NRC took enforcement action regarding a number of procedural violations associated with the gas turbine battery surveillance. 'Iherefore, based on the above, the NRC has concluded that the first-line supervisor did not willfully falsify documents.

C. Petitioner's Concern Regarding Failure to Enter Technical Specification LCO l

The inspection determined that the classification of the resistance readings as " unsatisfactory" (" acceptance criteria block" checked "no") would have ensured that a determination of operability would have been performed by the Licensee and the related Technical Specification LCO would have been entered I if appropriate. However, since the first-line supervisor documented the high  ;

resistance readings, discussed the readings with Engineering, and on the basis '

of the discussion, determined that the battery was acceptable, the Licensee did not willfully fail to enter the LCO in that the Licensee determined that the previous operability determination was valid and, therefore, that the surveillance procedure criteria had been met.

In response to the NRC IR tesults, the Millstone Unit i Director issued a memorandum to Millstone Unit 1 personnel to reinforce the expectation that if an acceptance criterion is not met, the "no" block must be checked. The Unit Director stated that he held managers and supervisors personally accountable for ensuring that their personnel understood the message in the memorandum.

In addition, NNECO held several management team meetings to ensure a full appreciation of the type of performance citaracteristics that can lead to procedural violations and to reinforce the Licensee's expectation concerning the

" acceptance criterion met" block. NNECO also revised the acceptance criterion within Procedure SP 779.5 for the three connections that have the intercell connection cables with higher resistance because of the cable length. In addition, the official plant record was corrected for the annual battery surveillance that was incorrectly marked as meeting its acceptance criterion. In a subsequent inspection report, IR 50-245/95-31, 50-336/95-31, 50-423/95-31 (IR 95-31),

dated September 19,1995, the NRC reviewed the Licensee's corrective actions 2

Ahhough the rirst-hne supervisor was technically correct that the gas turbine battery was operable, the deternunanon of battery operabihty ed not follow the Licensee's adrrunistranve controls as &scussed above 217 i

in the above areas. The NRC Staff found the Licensee's corrective actions to be timely and thorough.

In summary, on the basis of the above information, the Staff found that the Petitioner's first-line supervisor did incorrectly mark the acceptance criterion met block "yes;" however, he annotated the high resistance readings on the cover page of the surveillance and marked the block "yes" based on his determination that Engineering had previously reviewed the issue and determined the battery to be operable. Further, the Staff found that since the Licensee determined that this was previously reviewed by Engineering and found acceptable, the ,

Licensee erroneously did not follow its administrative control procedures for l

determining operability and entering of appropriate LCOs. Therefore, the NRC l determined that (1) the Petitioner's first-line supervisor did not willfully falsify ,

nuclear documents or deliberately violate NRC regulations or the Millstone Unit l 1 operating license; (2) neither he, Northeast Utilities, nor the Millstone Unit I organization violated the provisions of section 50.5; (3) the requested removal of the first-line supervisor is not warranted based on these concerns; and (4) the Licensee's corrective actions were acceptable. As discussed below, the NRC took enforcement action regarding a number of procedural violations associated 1 with the gas turbine battery surveillance. I l

l D. Additional Concerns Regarding Inoperability of the Emergency Gas Turbine l The Petitioner provides a number of examples of what he alleges demonstrate inadequate procedural compliance by the Licensee regarding gas turbine battery surveillances which indicate that the gas turbine is inoperable due to battery problems3 In IR 94-31, the NRC determined that during implementation of Procedure SP 779.5, there were a number of examples (including the examples l the Petitioner provided)in which the Procedure SP 779.5 was not followed, nor was the job stopped and the procedure revised to correct the identified errors.

Ihr example, the procedure included a caution statement following step 6.19 that required the generation of a plant information report (PIR) and subsequent determination of operability if the battery acceptance criteria are not met. The PIR was not generated until this issue was questioned by the NRC. Step 6.17 of the procedure requires that if any resistance reading was greater than 65 micro-ohms, then the terminals and straps rnust be cleaned. The Licensee did not c! an the terminal and strap connections. Step 6.22 requires that the readings

aken during the surveillance be compared with previous battery surveillance 3

The Petitioner asserted that these problems have not been handled by the NRC and NU and that NU and the NRC are engaged in an apparent " coverup" of problems As explained above. the " coverup" issue has been referred to the o!G 218

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readings to determine if there is any deterioration of the battery system. The j Licensee did not perform this review and evaluate the battery for deterio. ration l until the NRC raised the issue. The NRC determined that these examples l

in which the procedure steps were not implemented constituted a violation of '

Technical Specification 6.8.1 and Procedure SP 779.5 and issued a Notice of Violation to the Licensee (categorizing this as a Severity Level IV Violation, l Violation 50-245/94-31-02). Further, the NRC noted in IR 94-31 that neither the recognition of the procedure errors during two prior implementations of this annual surveillance procedure (1992 and 1993),4 nor the biennial procedure review completed on December 8,1993, resulted in revisions to preclude the problems encountered during the 1994 surveillance. As discussed above, in IR 95-31, the NRC reviewed the Licensee's corrective actions for this violation and found them acceptable.

In IR 94-31, the NRC concluded that the previous operability evaluation of the gas turbine battery was acceptable and, therefore, that the gas turbine battery was operable at that time due to the previous evaluation. The violation cited in the Notice of Violation included the issues the Petitioner raised, specifically that NNECO failed to perform an operability determination and subsequently did not enter the Technical Specification LCO for the gas turbine While the NRC Staff did not take the actions the Petitioner requested, the Staff did take enforcement action based on its findings. Therefore, since the NRC found the Licensee's determination of operability acceptable and the NRC took enforcement action for the related violation described above, the NRC has concluded that additional enforcement action is not warranted.

E. Petitioner's Allegations Regarding Ilarassment and Intimidation With regard to the Petitioner's assertion of harassment and intimidation, the Petitioner alleges that (1) on October 7,1994, he was given a memorandum concerning absenteeism; (2) on October 27, 1994, he was unjustly chastised by his first-line supervisor and department manager about absenteeism; and (3) on December 14, 1994, he was given a memorandum that threatened him.

The Petitioner further alleges that he believes these actions by his supervision illust ate that NU management harasses, intimidates, and retaliates against individuals who raise safety concerns with outside agencies.

As indicated in a letter to the Petitioner dated November 28,1995, from the NRC Deputy Executive Director for Nuclear Reactor Regulation, Regional Operations and Research, the Petitioner has raised several complaints since 1993 with the NRC or the Department of Labor (DOL) concerning harassment, a

'The NRC noted sinnlar examples m which the procedure was nm followed or correcied dunng the annual survedlance in 1992 and 1991 219

intimidation, or discrimination by individuals at NU because the Petitioner raised safety concerns to NU or the NRC. As explained in the letter, the NRC conducted investigations into some of the harassment and intimidation allegations that the Petitioner had raised. The NRC did not substantiate that the Petitioner suffered discrimination for raising safety concerns. Further, of the complaints of harassment and intimidation that the Petitioner raised that were investigated by the DOL, none have been substantiated.

The Staff has, in addition, reviewed the Petitioner's remaining allegations of harassment and intimidation, including those in the petition, and has concluded that they do not present sufficient information warranting further investigatory effort. Accordingly, absent a finding of discrimination by the Secretary of Labor or an Administrative Law Judge on any pending complaints, or significant new evidence from the Petitioner that would support the allegations that NU has harassed, intimidated, or discriminated against him, the NRC Staff plans no 4 further followup of the harassment and intimidation complaints. Based on the above, no further action is warranted.

III. CONCLUSION i On the basis of the above assessment, I have concluded that some of the Petitioner's concerns were substantiated and resulted in appropriate enforcement action. Other concerns were not substantiated. Therefore, no additional enforcement action is being taken in this matter.

The Petitioner's request for action pursuant to section 2.206 is denied. As provided in 10 C.F.R. s 2.206(c), a copy of this Decision will be filed with the Secretary of the Commission for the Commission's review. This Decision will constitute the final action of the Commission 25 days after issuance unless the a Commission, on its own motion, institutes review of the Decision in that time.

l FOR THE NUCLEAR REGULATORY COMMISSION i

Ashok C. Thadani, Acting Director Office of Nuclear Reactor Regulation Dated at Rockville, Maryland, this 31st day of October 1996.

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j . Cite as 44 NRC 221 (1996) DD-96-17 j

UNITED STATES OF AMERICA

NUCLEAR REGULATORY COMMISSION

$ OFFICE OF NUCLEAR REACTOR REGULATION i

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Ashok C. Thadani, Acting Director 1

In the Matter of Docket No. 50-245 (License No. DPR-21)

NORTHEAST NUCLEAR ENERGY j

COMPANY (Millstone Nuclear Power Station, Unit 1) October 31,1996

'Ihe Acting Director of Nuclear Reactor Regulation has denied a petition by Anthony J. Ross that the NRC require Northeast Utilities to review all existing work orders for the past 10 or 12 years to ensure that Quality Assurance motor and connection work does not have certain deficiencies and take enforcement action against NU and its managers, based upon the Petitioner's assertions l

of intimidation and harassment and inadequate work control and procedure i compliance. Ibliowing his review, the Acting Director has determined that none of the technical issues raised by the Petitioner reflect a lack of procedural I compliance or warrant additional action by the Staff, and that the Petitioner's assertion of harassment and intimidation does not warrant any action.

Technical issue discussed: quality assurance.

DIRECTOR'S DECISION UNDER 10 C.F.R. 5 2.206 I. INTRODUCTION On December 30,1994, Mr. Anthony J. Ross (Petitioner) filed a petition with the Executive Director for Operations of the Nuclear Regulatory Commission (NRC) pursuant to section 2.206 of Title 10 of the Code of FederalRegulations (10 C.F.R. 5 2.206). In the petition, the Petitioner asserted that (1) inadequate 221

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l work control and procedure compliance exist at Millstone Unit 1, as evidenced by the use of standard commercial-grade lugs in a gas turbine fuel forwarding pump and motor that are quality assurance (QA)' subsystems of the emergency gas turbine generator and which had apparently been crimped using diagonal pliers; improper Raychem splices, cable bend radius, and connections in the connection boxes of major safety-related QA equipment; and non-QA lugs in-stalled, and improperly performed crimping, in fire protection quality assurance (FPQA) emergency lights, and (2) he had been subjected to ridicule by the gas turbine system engineer for raising concerns regarding the lugs on the gas tur-bine fuel forwarding pump and motor and that the system engineer willfully, l violated 10 C.F.R. 66 50.5 and 50.7. l

%c Petitioner requested that the NRC (1) require Northeast Utilities (NU) to review all existing work orders for the past 10 or 12 years, with NRC oversight.

l to ensure that QA motor and connection work does not have certain deficiencies; (2) assess a Severity Level I violation against NU and its managers for apparent l violations of section 50.7 and a Severity Level III violation against the gas turbine system engineer at Millstone for his apparent violation of section 50.7 and NU's " Code of Conduct and Ethics"; and (3) institute sanctions against the system engineer and NU and its managers for engaging in deliberate misconduct in violation of section 50.5.

, By letter dated R:bruary 23,1995, the NRC informed the Petitioner that the petition had been referred to the Office of Nuclear Reactor Regulation pursuant to section 2.206 of the Commission's regulations. The NRC also informed the Petitioner that the Staff would take appropriate action within a reasonable time

regarding the specific concerns raised in the petition. On the basis of a review of the issues raised by the Petitioner as discussed below, I have concluded that the actions sought by the Petitioner are not warranted.

l II. DISCUSSION A. Inadequate Work Control and Procedural Compliance Issues The issues raised by the Petitioner regarding the improper crimping and use of commercial-grade lugs in the gas turbine fuel forwarding pump and 4 motor, improper Raychem splices, cable bend radius, and connection issues, and improper crimping and use of non-QA lugs in emergency lighting have been addressed in correspondence between the NRC and NNECO, and have been the subject of evaluations by NNECO and an NRC inspection. Specifically, by 3

Quahty Assurance compnses those quahty assurance actions related to the physical charactenstics of a matenal, structure. component. or system ahich proude a means to control the quahty of the matenat structure, component.

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letters dated December 5 and 28,1994, and February 14, 1995, and during a phone conversation on December 15, 1994, the NRC raised these issues and l

requested NNECO to submit written responses. By letters dated March 6 and April 26,1995, NNECO responded to these requests and submitted information l regarding its evaluation of these issues. On May 15 through June 21, 1995, the NRC conducted a special safety inspection, which focused on these and other maintenance issues. The inspection findings are contained in Inspection

Report (IR) 50-245/95-22,50-336/95-22,50-423/95-22 (IR 95-22), dated July '

21,1995. Finally, NNECO provided further information regarding these issues in its August 31, 1995 response to the petition. A broad summary of the j resolution of these issues is set forth below. I

1. Gas Thrbine Fuel Forwarding Pump and Motor issues

'Ihe Petitioner asserts that the Licensee inadequately controls work and l procedural compliance at Millstone, as evidenced by the use of standard l commercial-grade lugs (instead of QA lugs) in a gas turbine fuel forwarding pump and motor that are QA subsystems of the emergency gas turbine generator i and which the Petitioner asserts had been crimped with diagonal pliers (instead of the proper crimping tool). In its response to the petition, dated August 31, 1995, NNECO stated that, when the supervisor examined the lugs in question, he concluded that although the lugs were somewhat discolored as a result of age, and may have had an indented crimp, they appeared to the supervisor to be the type of lug that had been installed in the 1971-1972 time frame, when no procedures were in place with respect to the type of lug required or the method of crimping. NNECO further stated that these lugs are considered acceptable where they have already been installed (i.e., meet original electrical standards); however, when maintenance is performed requiring relugging, the lugs are upgraded and installed in accordance with current procedures.

NNECO further stated that the fact that the lugs in question were commercial grade and may have been crimped with diagonal pliers is not indicative of a work control or procedural compliance problem. The lugs appeared to the NNECO supervisor to be tbc type of lug that had been installed at or near the time of initial plant startup m accordance with the appropriate electrical standards that existed at that time. Moreover, once the concern was raised about the proper type and crimping of the lugs by the Petitioner, NNECO took prompt action by initiating a work order to replace all the lugs.

The NRC Staff discussed the issue of defective lugs with the maintenance department manager and the worker who replaced the lugs during the special safety inspection. Neither individual could remember the work in detail but stated that to ensure reliabihty, the lugs were replaced.

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5 Based on NNECO's conclusion that (1) the lugs in question had been installed in the 1971-1972 time frame when no procedures were in place with respect to the type of lug required or the method of crimping, (2) these lugs are considered j acceptable where installed, and based on NNECO's prompt action to initiate a work order and replace all the lugs, the NRC concludes that this issue does not indicate an inadequate work control or procedural compliance problem.

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2. Improper Raychem Splices, Cable Bend Radius, and Connection issues

! The Petitioner asserts that the Licensee is inadequately controlling work

, and procedural compliance at Millstone, as evidenced by improper Raychem i

splices, cable bend radius, and connections in the connection boxes of major i safety related QA equipment (low-pressure coolant injection (LPCI) and core spray (CS) pumps). In its letter dated April 26, 1995, NNECO informed the

, NRC that an operability determination had been completed on the issue of the j Raychem splice installation, and whether Raychem splice bend radii on the

LPCI and CS pumps were less than the recommended limits (five times the

! Raychem diameter). The operability determination concluded that the motor f splices were operable and that an immediate inspection to verify bend radii was not warranted. In addition, NNECO stated that 50% of the Raychem splices on l the LPCI and CS pump motors had been inspected at that time with no problems l

{ identified. In its followup letter dated August 31,1995, NNECO stated that a l visual inspection of all the LPCI and CS pump motors had been completed and )

j none of the connections exceeded the minimum bend radius. Further, NNECO j did not identify any discrepancies in the connection boxes for the LPCI and CS I l pump motors. NNECO's evaluations validated the determination that the splices l are operable.2 j As a result of its evaluation of NNECO's response and supporting documen-tation and its independent verification of two of the pump motors in question, the l NRC found NNECO's response acceptable and that no further NRC review was I needed Therefore, the NRC Staff concludes that the Raychem splices, cable i bend radius, and the connections in the connection boxes of major safety-related 1 equipment (LPCI and CS motors) are acceptable. ,

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tn a&huon. NNECO (1) performed a review of all the work orders for the current Raychem sphce installation and ven6ed that the procedure spec 6ed that a rmmmum hend radius of nye times the Raychem diameter not be exceeded. (2) venhed that the traimng the electncians receive on Raychem sphces escusses the reqmrement of not excee&ng nse nmes the nummum bend ra&us, and (h requested that Raychem detemune what the consequences of excee&ng the nummum bend ra&us would be The resuks of the Raychem tesung showed shat even if one or d

more splices exceeded the nummum bend rahus, a hghter bend radms was acceptable.

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3. Emergency Lighting issue The Petitioner asserts that the Licensee does not adequately control work and procedure compliance at Millstone, as evidenced by non-QA lugs and improperly performed crimping in FPQA emergency lights. He NRC Staff requested NNECO to review the use of improper lugs for emergency lighting at Millstone Unit 1. Specifically, the NRC requested NNECO to review the concern that all four lugs on emergency light unit (ELU) 1-ELU-21 had Thomas and Betts lugs (non-QA) rather than the required QA AMP lugs, and all four lugs were not crimped properly. In addition, the NRC Staff asked NNECO to review the concern that one lug on the emergency light 1-ELU-29 was a Thomas and Betts lug and that three of the four lugs were not properly crimped.

NNECO responded that a review of the revision history for Procedure MP ,

790.2, " Emergency Light Inspection," determined that the procedure made no I reference to a specific lug prior to April 1993. NNECO stated that because the safety classification of these ELUs is FPQA, the lugs utilized in the ELUs must be FPQA. NNECO noted that Thomas and Betts lugs are only stocked as FPQA.

NNECO stated further that an evaluation was performed to determine the consequences of Thomas and Betts lugs in lieu of AMP lugs and to determine if all lug crimps on 1-ELU-21 and -29 were adequate. Additionally, NNECO's evaluation verified the ability of I-ELU-21 and -29 to perform their design function. NNECO has determined that the lug manufacturer is not a critical issue as long as the lug is compatible with the battery terminal and the wire used. In this case, the Thomas and Betts lug is similar to the AMP lug, and both lugs are compatible with the battery terminals and wire used. A compatibility study has been completed and documented in a Replacement item Evaluation (RIE).

NNECO performed a review of previous ELU surveillances to determine whether a degraded condition had been observed for the battery terminal lugs in these ELUs; this review did not reveal any degraded conditions. The Millstone Unit i Engineering Department inspected the crimping of the battery terminations, and the eight crimps were found to be adequate. Although all battery termination lugs are insulated on these ELUs, one splice on I-ELU-29 appeared to be crimped by a die for noninsulated lugs. However, this crimp did not affect operability of the ELU since a high-resistance connection was not present, and the insulation was not damaged. Satisfactory completion of a battery discharge test confirmed the adequacy of the crimps. Nonetheless, the lug that appeared to be crimped by a die for noninsulated lugs on I-ELU-29 has been replaced.

During its special inspection, the NRC Staff reviewed the concern about emergency lighting lugs and NNECO's process for lug replacement. The NRC 225

I Staff verified that specific lugs were not called for in earlier versions of the lug replacement procedure and, therefore, as long as the lug was compatible and classified as FPQA, it could be used. Since nomas and Betts lugs are stocked as FPQA and are compatible, they could have been used in ELUs. In addition, since AMP lugs are stocked as non-QA, the plant staff would have had to fill out Form SF 486, " Upgrading FPQA Parts," to justify the upgrade of the lugs to FPQA standards.

ne NRC Staff reviewed an example of a lug changeout with an AMP lug and verified that Form SF 486 was included in the package to properly document the upgrade.

He NRC Staff reviewed the RIE form that documented the acceptability of Thomas and Betts lugs as an alternative for AMP lugs. The RIE indicated that the Thomas and Betts lugs are acceptable as an alternative item and that they will not degrade or compromise the original design basis. The NRC Staff found the RIE to be properly documented and adequate. The NRC Staff reviewed procedure MP 790.2, which was revised on April 12, 1995, and now requires that AMP lugs be used or an equivalent as evaluated and indicated by an RIE.

Since u.. RIE has been completed documenting Thomas and Betts lugs as an alternative, they are acceptable. The NRC Staff found the procedure adequate and also verified that the one questionable lug on 1-ELU-29 was replaced.

The NRC Staff concluded that the lugs on I-ELU-21 and -29 were adequately designed and qualified and that the ELUs were fully operable.

Based on NRC's findings that (1) the use of standard commercial-grade lugs in a gas turbine fuel forwarding pump and motor that are QA subsystems of the emergency gas turbine generator and which had apparently been crimped with >

diagonal pliers does not constitute an inadequate work control or procedural compliance problem; (2) the Raychem splices, cable bend radius, and the connections in the connection boxes of major safety-related equipment (LPCI and CS motors) are operable; and (3) the lugs on I ELU-21 and -29 were adequately designed and qualified and the ELUs were fully operable, the NRC Staff has determined that the Licensee adequately controls work and procedure compliance within these areas at Millstone. Therefore, the Petitioner's request to require NU to review all existing work orders for the past 10 or 12 years, with NRC oversight, to ensure that QA motor and connection work does not have certain deficiencies, is not warranted.

B. liarassment and Intimidation Issue The Petitioner alleges that he was ridiculed by the gas turbine system engineer for raising safety concerns regarding the lugs on the gas turbine fuel forwarding pump and motor and that the system engineer willfully violated sections 50.5 j l

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, and 50.7. In addition, the Petitioner alleges that NU and its managers violated 4

sections 50.5 and 50.7 and NU's " Code of Conduct and Ethics."

As indicated in a letter to the Petitioner dated November 28,1995, from the Deputy Executive Director for Nuclear Reactor Regulation, Regional Operations

and Research, the Petitioner has raised several complaints since 1993 with the NRC or the Department of Labor (DOL) concerning harassment, intimidation, or discrimination by individuals at NU because the Petitioner raised safcty 3

concerns to NU or the NRC. As explained in the letter, the NRC conducted ,

investigations into some of the harassment and intimidation allegations that

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l the Petitioner had raised. He NRC did not substantiate that the Petitioner '

suffered discrimination for raising safety concems. Further, of the complaints of harassment and intimidation that the Petitioner raised that were investigated i by the DOL, none have been substantiated.

The Staff has, in addition, reviewed the Petitioner's remaining allegations of harassment and intimidation, including those in the petition, and has concluded

that they do not present sufficient information warranting further investigatory

effort. Accordingly, absent a finding of discrimination by the Secretary of Labor j or an Administrative Law Judge on any pending complaints, or significant new l 1 evidence from the Petitioner that would support the allegations that NU has y harassed, intimidated, or discriminated against him, the NRC Staff plans no d

further followup of the harassment and intimidation complaints. Based on the above, no further action is warranted.

IIL CONCLUSION l The Licensee evaluated the technical issues and provided the results to the Staff for review. The Staff also conducted inspections to independently determine if the Licensee's conclusions and corrective actions were acceptable.

, As explained atieve, none of the technical issues reflect a lack of procedural compliance or warrant additional action by the Staff. Also, as explained above, the Petitioner's assertion of harassment and intimidation does not warrant any 3Clion.

On the basis of the above assessment, I have concluded that no issues have been raised regarding Millstone Unit I that would require initiation of enforcement action. Therefore, no enforcement action is being taken in this matter.

The Petitioner's request for action pursuant to section 2.206 is denied. As l provided in 10 C.F.R. 6 2.206(c), a copy of this Decision will he filed with the j Secretary of the Commission for the Commission's review. This Decision will o

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

FOR THE NUCLEAR REGULATORY COMMISSION Ashok C. 'Ihadani, Acting Director Office of Nuclear Reactor Regulation Dated at Rockville, Maryland, this 31st day of October 1996.

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