ML20056D912

Nuclear Regulatory Commission Issuances for May 1993.Pages 355-418

ML20056D912
Person / Time
Issue date:	08/31/1993
From:	NRC OFFICE OF ADMINISTRATION (ADM)
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References
NUREG-0750, NUREG-0750-V37-N05, NUREG-750, NUREG-750-V37-N5, NUDOCS 9308190001
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I l NUREG-0750 I Vol. 37, No. 5 Pages 355-418 I l l NUCLEAR REGULATORY COMMISSION ISSUANCES i May 1993 i - ) 1 4 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 Denials of Petitions for Ruiomaking (DPRM). The summaries and headnotes preceding the opinions reported herein are not to be deemed a part of those opinions or have any independent a legal significance. 1 i U.S. NUCLEAR REGULATORY COMMISSION i Prepared by the Division of Freedom of Information and Publications Services Office of Administration i U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 (301/492-8925) w m-- a

COMMISSIONERS tvan Selin, Chairman Kenneth C. Rogers James R. Curtiss Forrest J. Remick E. Gail de Planque B. Paul Cotter, Jr Chief Administrative Judge, Atomic Safety and Licensirg Board Panel -

., ~. CONTENTS t Issuance of the Nuclear Regulatory Commission i SACRAMENTO MUNICIPAL UTILITY DISTRICT (Rancho Seco Nuclear Generating Station) Docket 50-312-DCOM (Decommissioning Plan) MEMORANDUM AND ORDER, CL1-93-12, May 26,1993........ 355 j 1! Issuances of Directors' Decisions [ INTERSTATE NUCLEAR SERVICE CORPORATION i (Indian Orchard, Massachusetts) Docket 030-04632 -[ DIRECTOR'S DECISION UNDER 10 C.F.R. 6 2.206, DD-93-9, May 7, 1993..................................... 365 i NIAGARA MOHAWK POWER CORPORATION (Nine Mile Point Nuclear Station, Unit 1) Docket 50-220 DIRECIDR'S DECISION UNDER 10 C.F.R. 62.206, DD-93-10, May 9, 1993................................... 381 i TEXAS UTILITIES ELECTRIC COMPANY, et al. and ALL i NUCLEAR POWER PLANTS WITil THERMO-LAG FIRE BARRIEP,S (Comanche Peak Steam Electric Station, Units 1 and 2) Dockets 50-445, 50-446 FINAL DIRECTOR'S DECISION UNDER 10 C.F.R. f 2.206, DD-93-11, May 23. 1993.................................. 402 - 1 ~; I >v i l I 3 iii ' 1 ? ~' -l

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Cite as 37 NRC 355 (1993) CL1-93-12 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION COMMISSIONERS: Ivan Selin, Chairman Kenneth C. Rogers James R. Curtiss Forrest J. Remick E. Gail de Planque in the Matter of Docket No. 50-312-DCOM 7 (Decommissioning Plan) 4 SACRAMENTO MUNICIPAL UTILITY DISTRICT (Rancho Seco Nuclear Gen Jating Station) May 26,1993 + The Commission denies Sacramento Municipal Utility District's motion for reconsideration of CLI-93-3, in which the Commission granted the Environ-mental and Resources Conservation Organization discretionary intervention, ad-mitted one contention, and permitted amendment of another in a pmceeding to consider a pmposed order approving a decommissioning plan for, and authoriz-ing decommissioning of, the Rancho Seco Nuclear Generating Station. RULES OF PRACTICE: STANDING TO INTERVENE The Commission has the authority to grant intervention, as a matter of discretion, pursuant to the Commission's authority to hold hearings and to permit participation in its proceedings. f 355 h I

RULES OF PRACTICE: MOTIONS FOR RECONSIDERATION (RAISING MATTERS FOR TIIE FIRST TIME) "Ihe Commission rejects an argument raised for the first time in a motion for reconsideration as a basis for reconsideration of its admission of a contention. RULES OF PRACTICE: CONTENTIONS (LATE-FILING REQUIREMENTS) 'Ihe provisions of 10 C.F.R. 62.714(b)(2)(iii) with respect to the filing of contentions based on differences between data and conclusions in the Staff's environmental review documents and the licensce's or appliumt's environmental report are nci intended to add or remove from consideration any factor to be balanced in the determination of the admissibility of a late-filed contention under t~ 10 C.F.R. 6 2.714(a). RULES OF PRACTICE: CONTENTIONS (LATE-FILING REQUIREMENTS) Late-filed contentions, including those filed on subsequendy issued NRC environmental review documents, are subject to the late-filed criteria set out in 10 C.F.R. 5 2.714(a)(1)(i)-(v). RULES OF PRACTICE: CONTENTIONS (LATE-FILING REQUIREMENTS) Even without a showing that the Staff's environmental review documents significantly differ from the applicant's environmental report, a petitioner may be able to meet late-filed contention requirements, e.g., by presenting significant new evidence not previously available. l l RULES OF PRACTICE: CONTENTIONS (LATE-FILING l REQUIREMENTS) A showing that the Staff's environmental review documents significantly differ from the applicant's environmental report, although ordinarily sufficient to show good cause for lateness, is not in itself sufficient to make an environmental l contention admissible, because the petitioner must still meet the other criteria in 10 C.F.R. I2.714(a). 356 l l l l ........... ~ -. -

RULES OF PRACTICE: CONTENTIONS A generic -avironmental impact statement on decommissioning cannot be categorized as an " applicant's document *' for purposes of 10 C.F.R. I 2.714(b)(2)(iii). MEMORANDUM AND ORDER Sacramento Municipal Utility District (SMUD or Licensee) has filed a motion for reconsideration of the Commission's Memorandum and Order, CLI-93-3, which granted Environmental and Resources Conservation Organization (ECO) j discretionary intervention, admitted one contention, and permitted an amendment of another. 37 NRC 135 (1993). This proceeding involves ECO's challenge to i the Nuclear Regulatory Commission (NRC) Staff's proposed order approving a decommissioning plan for, and authorizing decommissioning of, the Rancho Seco Nuclear Generating Station (Rancho Seco). Ibr the reasons set forth below, we deny SMUD's motion for reconsideration. He Licensee premises its motion for reconsideration on essentially four bases.' First, the Licensee argues that the Commission should reconsider its determination to grant discretionary intervention because ECO has not met the standards required for discretionary intervention set out in Commission jurisprudence. Second, the Licensee argues that the Commission improperly l admitted the contention regarding loss of offsite power (LOOP) because the l l LOOP issue does not raise a material issue of law or fact and is unrelated to any ECO interest, and because ECO has not demonstrated a significant ability to contribute to the record on this issue. Wird, the Licensee maintains that to allow ECO the opportunity to file an amended contention regarding SMUD's decommissioning funding plan is patently unfair and prejudicial to the Licensee. Iburth, the Licensee argues that allowing the adjudication to remain open for the specific purpose of allowing contentions to be filed on the Staff's environmental I review documents is inconsistent with Commission regulttions. The Staff supports the Licensee's motion with respect to the first three noted objections for essentially the same reasons presented by the Licensee.2 De Staff does not seek reconsideration of the fourth matter because the Staff believes that the relief provided by the Commission is consistent with Commission regulations. ECO opposes the Licensee's motion.5 l l t 3 lacmanc's Manon fw RecensidaaGan. Mad 10,1993. 2 NRC staf!'s suppan of ticensee's Manan fw F- "d-ean, March 26.1993. 3 IlCo's Answa in oppaaldan to tioenses's Manan for Recamdcration. Mad 26.1993. 357

o -I Analysis ' Rr the reasons set out in more detail below, SMUD has failed to identify any error or abuse of discretion by the Commission in deciding CL1-93-3. Although SMUD asserts that its interests are compelling and that our decision is highly prejudicial, SMUD has not articulated any specific harm tiet it is suffering as a result of our order. l 1. Discrrtionary Intervention in CL1-93-3, we granted ECO discretionary intervention because ECO pre-sented several difficult questions which, if resolved in its favor, would support l standing. We also found that ECO had submitted one viable contention in addition, we stated that the Commission is presently reviev.ing the process for review and approval of decommissioning plans, including the timing and scope of public participation in the decommissioning pmcess. CLI-93-3,37 NRC at 141. He decision to grant ECO intervention, without resolving the question of standing as of right, rested on our discretionary authority to hold hearings and to permit participation in our proceedings. The Licensee does not challenge this I authority. Nonetheless, the Licensee asserts that we ignored our own standards and precedents in granting ECO discretionary intervention in this instance. We disagree. Cases cited by the Licensee, including Portland General Electric Co. (Pebble Springs Nuclear Plant, Units 1 and 2), CLI-76-27, 4 NRC 610 (1976) (hereinafter Pebble Springs) are consistent with our decision to grant ECO discretionary intervention. As we stated in Pebble Springs, it was expected that the practice of granting discretionary intervention should develop "not through precedent, but through attention to the concrete facts of particular situations." 4 NRC at 617. Although ECO did not establish a clear case for standing, ECO averred certain arguments that would support its standing. The standing issue posed questions of first impression in the context of a Staff decommissioning order. Resolution of those questions might have little, if any, future application in view of the Commission's current examination of the pmcess for approving ' decommissioning plans. Thus, in this instance we have determined that it is in the public interest to resolve the parucular matters raised by the Petitioner rather { than to expend any further resources resolving the difficult questions regarding ECO's standing as of right. In reaching our decision, we also took into account whether ECO had raised potentially litigable matters in this proceeding. De Commission admitted i one aspect of ECO's environmental contention regarding the probability of a IDOP and has permined ECO to amend its contention regarding the funding 358 i i h -f -i {

plan. Although both the Staff and Licensee argue that ECO is not capabic of ~ making a substantial contribution on either of these matters, such a conclusion 'is premature. De questions of whether a genuine issue of material fact remains regarding the IDOP contention and whether ECO has submitted an admissible amended funding plan contention are before the Licensing Board. If ECO establishes litigable matters with respect to either of these contentions, it will clearly be in the public interest to resolve these matters, which involve the adequacy of - the Licensee's discussion of credible accidents in the Environmental Report, and to determine whether SMUD has provided adequate financial assurances for funding decommissioning. On the other hand, if neither contention survives further scrutiny under the applicable provisions of 10 C.F.R. 62.714(b)(2) or $ 2.749, then such matters will be dispensed with promptly. In sum, we gave due consideration to the Pebble Springs criteria and the particular circumstances of this case in formulating our order. Rus, ' we decline to reconsider our determination to grant ECO intenention as a matter of discretion. 2. Admissibility of the Erwironmental Contention ECO's environmental contention alleged, in part, that SMUD's Environmen-tal Report is inadequate because SMUD's discussion of radiological impacts appears to rely men ly on general NRC regulations, guidance, and reports. As an example, ECO alleged that SMUD's discussion of the probability of a LOOP is inadequate. In CLI-93-3, we admitted ECO's contention that theit is no refer. ence to a particularized study to allow independent verification of the conclusion that the probability of a IDOP is less than once in 20 cars. CL1-93-3,37 NRC 3 at 146.' SMUD argues that this maner was not raised on appeal by ECO.* SMUD is incorrect. On appeal, ECO argued that the Licensing Board erred by not considering ECO's specific examples of alleged deficiencies in SMUD's - Environmental Report. Although ECO did not restate the specifics of ECO's objections to SMUD's discussion of the LOOP, ECO referred in its appellate brief to the pages of ECO's supplemental petition that, according to ECO, the Licensing Board failed to consider.5 The cited portions of ECO's supplemental

• sMUD Waion at s.

Sin support of its argumers that de licerning Board impropedy denied the admissibility cf Eco's erwironmmtal ournantias.100 staiad that "lico spens out in great desa0 the various equiremares for en envirmmamal report. Eco (Amendmcas and] suppiammt [m Perition for teeve to Irservene and Roguest for IIsaring, June 29,1992, at] 16-28. And 100 discuased in gre detaD the various ways [in which] the errr;nvanantal repart dad sue =ws thans requirements.... l'aving previously speDed oct those duties.. there is nothing further far u ' have skuse? ECX) Brief in support of Appeal fran IJtP-92-23. Sepaarnbar s.1992. si 3431. 359 b_

petition included ECO's allegation that there is no reference to a particularized study in SMUD's Environmental Report to allow independent verification of SMUD's conclusion regarding the probability of a LDOP. In its motion for reconsideration, SMUD presents a new argument regarding the IJDOP - that the probability of a LOOP is immaterial and, thus. ECO has not raised a valid contention.' This is not the argument that SMUD provided in its answer to ECO's supplement when SMUD addressed ECO's reference to the probability of a LOOP.7 Had SMUD made these arguments to the Board, it might have defeated admission of this contention. Nevertheless, SMUD is free to pursue this new line of argument before the Licensing Board in resolving any remaining matter regarding this contention We decline to reconsider our a determination to admit ECO's environmental contention with respect to the probability of a LOOP. 3. Amending the Funding Plan Contention In CLI-93-3, we permined ECO to amend its contention challenging the adequacy of SMUD's proposed funding plan because there was sufficient confusion in the proceedings below as to whether ECO could raise the objections to the funding plan in a proceeding other than the instant one. Both Staff and Licensee insist that ECO was responsible for the confusion and that it was clear that any such objections had to be raised in this proceeding. We disagree. We found that the confusion apparently stemmed from the handling of an earlier SMUD request for, and Staff consideration of, an exemption from complying with the funding requirements in 10 C.F.R. 650.75(e)(1)(ii). See CL1-93-3, 37 NRC at 148-49. Due to a subsequent change in Commission rules in July 1992, an exemption was no longer necessary and, thus, was no longer being considered by Staff.' Nevertheless, the Licensing Board still concluded that ECO's comments regarding the furkling plan would be considered when Staff determined whether to grant SMUD's exemption request, and, for this reason, the Licensing Board concluded that the " crux of ECO's concern... has been fulfilled." i BP-92-23,36 NRC 120,137 (1992). I 'sMUD's Meuen at 8 9. The staff's argumcras supposting sMUD's snaian also pane objecnons nised for the fast ume en neounsidersnan. r 71acmsee's Answar to BCo*s Amendmma and s9plemera to Peunan for teen so tmerwne and Request far 5 llearing. My 8.1992, et el 21 8 Ahhaugh we hsw already admiued the origins! contenban as 35Mdad in Ctj-93 3, we leave for the Isoensmg Board to desarmme if the further amendmans to the comannan is admissible and to deurndacif a geuine issus of mezzial! set remams agardmg em pmbebDity of a IDoP.The thusmg Board should also daarmirs if EoD's ammdad " nises maners that were not dependent en the analysis or or prchahDity d a tDoP. To the catsra that BCO raises issues that could have been mised before because they are oss dependers en the new t informanan prended regarding the probabihty of a tJ00P, ECo must meet the criscris forlawf. led commuons 'QJ-93-3. 37 NRC at 149 (cinng Prahearmg Carderance Transcript at 140). See Fmal Rule, Decommissiamns j Fundmg far Premmunely shut Down Power Reaciers,57 Fed. Reg. 30,3s3 Quly 9,1992). 360 i i I r i .i i i

5 t i In its motion for reconsideration. SMUD maintains that Staff made it clear that there would not be anothar proceeding in which ECO could challenge j the funding plan and such contentions must be raised here. In support of j this argument, SMUD quotes the NRC Staff's response to ECO's contention regarding the funding plan: Chapter 5 and Appendia C of SMUD's decorrnaissioning plan pmvide SMUD's funding plan. %c NRC staffis reviewing that proposal. Approval of the fundang plan will be in the j Order or by separare appromf. ECO provides no basis to contest the adequacy of SMUD's i funding plan in this promeding, and iu adequacy, therefore, cannot be a cornention berein.38 i Staff's mention of a " separate approval" does little more than raise further confusion as to whether there would be a separate proceeding in which ECO could challenge the funding plan. Moreover. ECO indicated its intent to ( challenge the adequacy of the funding plan." Thus, we decline to reconsider. our determination to permit ECO to amend its funding plan contention. 3 SMUD also argues that if ECO is permitted to amend its funding plan contention, the amendment should be limited to matters related to the exemption request. We leave to the Licensing Board to determine whether ECO has presented issues that go beyond the scope of the adequacy of the funding plan i and, if new matters are raised, whether such new matters meet the late-filed { criteria listed in 10 C.F.R. 0 2.714(a)(1)(i)-(v). t 4. Staying issuance of the Decommissioning Order In CLi-93-3. we ordered the Staff to withhold issuance of the decommission-ing order until after the Licensing Board has completed both its review of the admissibility of any amended or late-filed contentions and has held a hearing j or otherwise resolved any litigable matter that may be raised. 37 NRC at 152. We provided for a pre-effectiveness hearing based on the totality of the circum-stances presented in this instance. Because the Staff's environmental review documents had not been issued, we recognized that contentions might be filed by ECO pufsuant to 10 C.F.R. f 2.714(b)(2)(iii) after the Staff had completed its environmental review. Thus, we permitted a reasonable opportunity for consid-3'sMULYs Maim at 11 12 (emphasis added)(quamg NRC siaN Response to EOo's supplemmt to ha Petition fer tmve to bnervens and Request far Hearing. July 10.1991 at 27). p 13.Sae. e.g., Preheanng Csoferous Transcrip at 141 Quly H.1992) fcounsel far Eco staind that h found "that abe j L 'smaning plan rundang arungczners that the stafr originally proposed to appmwe is fun of sanuadictions. --M and ensuing redach ical and erwironmental naks due to the pennd of time of fundang; therafat. 6 [Ose fundmg plan] should bc syncied? Mueowar. caansel for Eco stated at the prohonnng cor:ferersas that *1 am j nat bere trymg to attack as everyene has recogmand.the * . fundmg plan insc1f. het in a separats peceeding... - {W]hus k is still e sepersw proceeding, they [stsK and sMUD) are entsctrag ene for nor heving ramed [1%XTs shaDenge so the funding plan) in the amucat of this pmonading n)na" 1d at 150. t 361 F 'f l i -i

r i I cration of such contentions consistent with our grant of intervention as a matter of discretion and our determination to permit a prior hearing on litigable mat-ters. De Licensee has not shown any specific support for its argument that our determination has caused it considerable hardship. Ecrefore, we decline to reconsider this portion of our onler. In the alternative, SMUD argues that (1) the Commission should direct j the NRC Staff to conclude its environmental review promptly and (2) the Commission should clarify that ECO must show that the new Staff documents on which ECO submits late-filed contentions contain data and conclusions that differ I significantly from the data rr.;onclusions in SMUD's environmental report and j the NRC's Generic E.svironmental Impact Statement on Decommissioning of .i Nuclear Pacilities (NUREG-0586) (hereinafter GEIS). With respect to SMUD's first request, we note that CL1-93-3 was not intended to delay issuance of the Staff's environmental review documents. Rus, when the Staff completes its review, the Staff's assessment or other environmental review documents should be issued promptly. With respect to SMUD's second request, the Commission emphasized in CL1-93-3 that 10 CE.R. 62.714(b)(2)(iii) of our regulations expressly provides for [ the filing of supplemental or amended contentions if the Staff's environmental review documents contain data or conclusions that differ significantly from the data or conclusions in the Licensee's environmental documents. See CL1-93-3, j 37 NRC at 153-54. We also stated that any such contentions are subject to the late-filed criteria listed in section 2314(a)(1)(i)-(v). Apparently, SMUD believes that the provision in section 2.714(b)(2)(iii), referring to contentions that may be filed on the Staff's environmental review documents, adds a sixth factor to.. the existing five factors that must be considered when determining whether a late-filed contention is admissible pursuant to section 2.714(a)(1)(i)-(v). SMUD argues that in addition to meeting the five factors listed in section 2.714(a)(1)(i)- (v), ECO must also show that the Staff's documents significantly differ fro 1. both applicant's environmental documents and the GEIS. Although information regarding the difference between the applicant's environmental report and the l Staff's environmental review documents is relevant to the " good-cause" factor, section 2.714(b)(2)(iii) neither adds nor removes from consideration any factor to be balanced in the determination of the admissibility of a late-filed contention. pursuant to section 2314(a). t ne Commission promulgated changes to section 2314, effective in Septem-ber 1989." One signi6 cant change to this section is that the. Commission now requires the petitioner to provide sufficient information with its proffered j i + Ubal Rule, Rules of Prsc6ce for Domeanc thensing Noendmgs - Needurst r.langes an die lleanna Neess. 54 Fed. Reg. s3.168 (Aug 11,1989), ag*4 sub am (Jnion <(Concerned scannsar v. NRC. 920 F.2d 50 (D C. ' Cir.1940). [ I 362 h t L-

c contention "to show that a genuine dispute exists with t!e applicant on a ma-terial issue of law or fact." 10 C.F.R. f 2.714(b)(2)(iiii As a general matter, in making this showing the petitioner must refer %c specific portions of the application that the petitioner disputes or the uns for the petitioncr's belief that required information is omitted. On environmental matters this showing must include a reference to the specific portion of the applicant's environmental report that the petitioner believes inadequate, liowever, the Commission ex-pressly recognizes that if data and conclusions in Staff environmental review documents significandy ditfer from the data and conclusions on a material issue in the applicant's envimnmental report, the petitioner could raise a genuine issue by referring to those portions of the Staff's environmental documents that the petitioner disputes. See 10 C.F.R. $ 2.714(b)(2)(iii). In pmmulgating this change we did not alter the criteria to be considered when determining the admissibility of a late-filed contention, but merely emphasized that prior agency case law makes it clear that any late-fi t d contention, including those filed on subsequent NRC cnvimnmental review documents, are subject to the late-filed criteria set out in section 2.714(a)(1)(i)-(v)." We did not mean to imply that a showing that the Staff's envimnmental review documents significandy differ from the applicant's environmental report is always necessary to raise a good contention. Even without such a showing, a petitioner may still be able to meet the late-filed contention requirements of section 2.714(a), e.g., by presenting significant new evidence not previously availabic. Likewise, a showing that the Staff's environmental review documents significantly differ from the applicant's environmental report, although ordinarily sufficient to show good cause for lateness, is not by itself sufficient to make an envimnmental contention admissible, because the petitioner must still meet the other criteria in section 2.714(a).' Further, we decline to require ECO to also show how the data and conclu-sions in Staff's environmental documents differ significantly from the data and conclusions in the GEIS for purposes of filing a contention. The GEIS cannot be categorized as an " applicant's document" under 10 C.F.R.12.714(b)(2)(iii). Conclusion For the reasons stated above, SMUD's motion for reconsideration of CL1-93-3 is denied. p 54 Fed. Reg. at 33.171 363

m_.. 3 t It is so ORDERED. - l t i Ibr the Commission l .I JOIIN C. HOYLE - I t Assistant Secretary of the '. l' Commission I Dated at Rockville, Maryland,. this 26th day of May 19?3. ? 7 I r ~ i ' i i i h r k i } a t t I y i k r 364- ' t

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l i i l l l l Directors' i Decisions l Under l 10 CFR 2.206 i I l 1 i l 1 l l 4 I l l 1 1 4

? Cite as 37 NRC 365 (1993) DD-93-9 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION i OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEG'JARDS Robert M. Bernero, Director In the Matter of Docket No. 030-04632 INTERSTATE NUCLEAR SERVICE t CORPORATION (Indian Orchard, Massachusetts) May 7,1993 The Director of the Office of Nuclear Material Safety and Safeguards grants in part and denics in part a petition filed by Gloria M. Mitchell and Linda Hammons, on behalf of the Indian Orchard Citizens Council (10CC), which requested action with regard to the Interstate Nuclear Service Corporation (INS) at Indian Orchard,. Massachusetts. ' Petitioners made ten requests and four demands. Petinoners asserted as bases for their requests and demands that the residents of the Indian Orchard neighborhood of Spring 5eid, Massachusetts, live in close proximity to INS and have expressed great concern about health issues related to the operation of INS, especially since the publication of an article in the Springfield Sunday Republican on June 7,1992, concerning radiation levels 7 at the INS perimeter fence, onsite waste storage by INS, and INS discharges to the city sewer system. The requests of Petitioners that were granted are that the NRC: (1) partic-ipate in a public meeting in Indian Orchard to respond to the concerns of the i neighborhood residents; (2) hold an unannounced inspection of INS; (3) provide to the Petitioners a copy of the NRC regulations under which INS operates; (4) check adjoining Park Department land, including Dimmock Pond, for contami-nation and illegal dumping of waste material; (5) determine what INS has done with waste material not shipped; (6) provide to the Petidoners the docket number. for INS; (7) identify a Public Document Room (PDR) for INS and its location; and (8) describe the type of monitoring done, who does it, and how frequently. Requests that were denied are that the NRC: -(1) check homes in tre area for radioactive contamination; and (2) check Loon Pond for contamination and for i 365 i I i A i

i i possible illegal dumping of waste material. Petitioners

• demands that were de-nied are that- (1) radiation readings outside the INS fence perimeters be 'T)"

at all times; (2) *0" nuclear waste byproducts from INS be allowed to enter Springfield's water / sewer system; and (3) under no circumstances should INS be allowed to store nuclear waste on its property. Petitioners' demand that INS stop using residential streets, specifically Nagle and Nichols Streets, to go to and from its plant was mooted by the voluntary actions of INS. DIRECTOR'S DECISION UNDER 10 C.F.R. f 2.206 I. INTRODUCTION r By letter dated Jtme 29, 1992, addressed to the Chairman of the Nuclear Regulatory Commission (NRC or the Commission), Gloria M. Mitchell and Linda llammons, on behalf of the Indian Orchard Citizens Council (IOCC), requested that NRC take action with respect to Interstate Nuclear Service Corporation (INS or the Licensee)in Indian Orchard, Massachusetts. 'Ihe IOCC requested an NRC response or action on ten matters or requests and made four " demands" concerning the Licensee's activities. Petitioners request that the NRC: (1) participate in a public hearing in Indian Orchard to respond to the concerns of neighborhood residents; (2) hold a surprise inspection of INS; (3) check homes in the area for radioactive contamination; (4) provide to the Petitioners a copy of the NRC regulations under which INS operates; (5) check adjoining Park Department land, including Dimmock Pond, for contamination and illegal dumping of waste material; (6) check loon Pora contamination and for possible illegal dumping of waste material; (7) determine what INS has done with waste material not shipped; (8) provide to the Petitioners *.he docket number for INS: (9) identify a Public Document Room (PDR) for INS and its location; and (10) describe the type of monitoring done. who does it, and how frequently. Petitioners funher " demand" on behalf of neighborhood residents that: (1) radiation readings outside the INS fence pcrimeters be "0" at all times; (2) "0" nuclear waste byproducts be allowed to enter Springfield's water / sewer system; (3) INS stop using residential streets, specifically Nagle and Nichols Streets, to go to and from its plant; and (4) under no circumstances should INS be allowed to store nuclear waste on its property.. Petitioners assert as bases for their lequests and demands that the residents = of the Indian Orchard neighborhood of Springfield, Massachusetts, live in riose proximity to INS and have expressed great concern over possible health issues, especially since publication of an article la the Springfield Sunday r.epublican 366 3 f 7 l

i on June 7,1992. The article reported that:. (1) radiation readings outside the . INS perimeter fence, near a waste-filled truck, were 12 to 15 times normal background radiation levels experienced in everyday life; (2) rJ1 INS waste will be stored un site beginning January 1,1993; (3) in 1989, INS waste stored was twice the volume shipped; (4) the corporate health physics rnanager of INS, j Michael Bovino, stated that waste is removed twice a year, but NRC records indicate that it is removed only once a year and not at all in 1990; (5) a person ? standing at the INS fence for two days in early May would have receive 4 a higher radiation dose than a person standing at Vermont Yankee's fence for a year because of tighter regumions for nuclear power plams; and (6) there have been allegations that INS discharges radioactive water into the city sewer system. The NRC Staff provided a partial response to 10CC by letter dated July 21,. 1992. By letter dated August 25,1992, the NRC Staff formally acknowledged receipt of the Petition and informed Petitioners that their Petition weajd be i treated as a request under 10 C.F.R. 52.206 and a decision would be issued within a reasonable amount of time. By letter dated August 25,1992, the Staff also informed INS of the Petition and invited INS to provide information for the Staff's consideration. INS responded to the Petition on August 31,1992. I have completed my evaluation of the matters raised by Petit 5ners and have determined that, for the reasons stated below, the Petition shall be granted in par

• c.,d denied in part. De Petition is granted insofar as the NRC Staff:

partici e !in a public meeting on the evening of July 23,1992, at the American Legion Post, Number 277, in Indian Orchard and responded to the concerns of the neighborhood residents; conducted an unannounced inspection of INS on July 8 and 9,1992; provided lOCC with copies of pertinent portions of NRC's j regulations; checked adjoining Park Department land, including Dimmock Pond, for contamination; reviewed INS's waste storage program; pmvided lOCC a j description of INS's radiation monitoring program; identified the location of the Public Document Room (PDR) for the INS license; and provided the docket number for the INS license. The Petition is denied with respat to the remaining requests to check homes in the area for radioactive contamination, and to check Loon Pond for contamination and possible illegal dumping of waste material. l De Petition is also denied with respect to three of IOCC's demands. The fourth i demand was mooted by the Licensee's voluntary actions. II. BACKGROUND INS is a subsidiary of UniFirst Corporation whose headquarters are located - in Springfield, Massachusetts. INS operates thirteen facilities, each of which is separately licensed by the NRC or an Agreement State. An Agreement State 367 f i I .t i

i e is one with which the NRC, or previously the Atomic Energy. Commission, has entered into an agreement under subsection 274b of the Atomic Energy Act of 1954, as amended, for the state to assume the regulatory authority and l responsibility that would otherwise be discharged by the NRC with respect to protection of public health and safety associated with the possession and use of certain categories of radioactive materials. The Commonwealth of Massachusetts is not an Agreement State and, therefore, the regulatory authority over the facility that is the subject of this Petition resides with the NRC.

l

'i One of INS's thirteen facilities is located in Indian Orchard, a community of Springfield, Massachusetts. INS at Indian Orchard holds NRC License No. 2(M)3529-01 and is authorized to possess various byproduct, source, and spaial nuclear materials in the form of contaminated material and associated decon-taminated wa*e for the collection, laundering, and decontamination of contam-inated clothing and other launderable nonapparel items. More specifically, INS j is authorized to possess the following maximum amounts of NRC-licensed ma-. terials: 0.93 terabecquerels (2.5 curies) of any byproduct material with atomic numbers 1-83; 370 megabecquerels (10 millicuries) of any bypwet material with atonic numbers 84-102; 10 kilograms of any source material, and special nuclear material with a total quantity not to exceed 0.25 kilograr,i of uranium enriched in uranium-235 or 0.020 kilogram of plutonium. INS it also autho-i rized to possess any byproduct material in individual sources not acceding 37 megabecquerels (1 millicurie) per source or 185 megabecquerels (5 mi!!icuries) i total activity for use as standards to calibrate radiation detection and measuring instruments. *Ihe license also authorizes the transport of licensed materials in accordance with 10 C.F.R. Part 71 of the Commission's regulations. l Use of licensed material is limited to the INS facility at 295 Parker Street, - Indian Orchant, Massachusetts. INS is not authorized to laander contaminated items at temporary jobsites or at a customer's facility, except as specifically authorized by the customer's license. INS is also not authorized to package or possess radioactive wastes, except those generated by the laundering activities conducted at its Indian Orchard facility. License No. 20-03529-01 was originally issued on April 15,1958, was last .i renewed on May 26,1988, and is due to expire on May 31,1993. l f III. DISCUSSION A. The NRC Staff has examined Petitioners

• concerns based on the article in the June 7, lo92 issue of the Springfield Sunday Republican. The Staff's evaluation of each of the six concerns in the article and referenced by Petitioners i

is discussed below: s 368 1

l 1. Radiation Readings Outside the INS Perimeter Fence, Near a Waste-Filled Truck, Were 12 to 15 Times Normal Background Radiation Levels Experienced in Everyday L(fe Current NRC regulations require NRC licensees to demonstrate that radiation. levels outside of the licensee's controlled area (e.g., INS's fenceline) shall not .i be greater than 20 microsieverts (2 millirems) in any I hour or 1 millisievert (100 millirems) in any 7 consecutive days.10 C.F.R. 620.105(b). Average i radiation exposure to a member of the general public from external radiation is approximately 1 millisievett (100 millirems) in 1 year. A radiation level 10-15 times background at the INS fence (or approximately 2 microsieverts (0.2 millirem) per hour) from a truck temporarily parked at INS for as long as a week and used to pick up radioactive waste would meet the current NRC hourly - and weekly standards. Beginning in January 1994, section 20.105(b) will be superseded by new requirem:.nts under 10 C.F.R. 520.1301(a). 56 Fed. Reg. 23,360 (May 21, 1991). Under the new requirements, NRC licensees must demonstrate that no individual member of the public would be exposed to more than 1 millisievert (100 millirems) of radiation above background from the licensee's activities in one year. "Ihe measurement of conformance to the new NRC requirements i must take into consideration changes in the radiation levels and the occupancy time of the maximally exposed individual member (s) of the public for the year. 1 Rr instance, in order for INS to exceed the new standard due to radiation from its waste-pickup truck, INS would have to make three radioactive waste shipments per year and the same individual members of the public would have to stand continuously at the fenceline throughout these periods. NRC inspectors j have confirmed that during the period 1989-1992, INS made no more than two rt % active waste shipments per year. Accordingly, the transient radiation level of 10-15 times txickground, or 2 microsievents (0.2 millirem) per hour, for as long as a week, would comply not only with current requirements, but also with the more restrictive new NRC requirements. INS's current environmental measurements of radiation involve weekly radi-ation surveys, the results of which have been within NRC limits. R)r transient radiation levels such as created by temporary parking of INS's waste-pickup truck, it normally would be difficult to estimate precisely the yearly radiation exposure at the fenceline based on measurements made once a week, if not for the additional surveys required by the U.S. Department of 'Ransportation (DOT).10 C.F.R. 6173.441(b). Prior to shipping its radioactive waste off site, INS is required by the DOT to perform radiation measurements at the driver's compartment, at all sides, top, and bottom of the vehicle, and at 2 meters away l from all lateral surfaces of the vehicle. The results of theac surveys are all within DOT limits. i 369 i e i

~ P b Even though not obligated by current NRC requirements to do so, the Licensee deployed thermoluminescent dosimeters (TLDs) in 1992 along the i fence of its property to measure environmental radiation levels. "Ihe use of H.Ds will improve the measurement of the annual radiation exposure at the fenceline because the devices will be continuously present, and should more definitively demonstrate whether INS has complied with NRC requirements. INS's TLD measurements for the last 6 months of 1992 are in compliance with NRC requirements. Based on the above, I conclude that Petitioners have not raised a substantial health or safety concern. 2. AllINS Waste Will Be Stored on Site Beginning January 1,1993 i The Low 12 Vel Radioactive Waste Policy Amendmerts Act of 1985 i (LLRWPAA) requires states to develop disposal capacity for their low-level ra-dioactive waste (LLW) by January 1,1993. States have several options: they may develop their own disposal facility; they may join with other states in com-pacts that will then develop disposal capacity for the member states; or they i may contract for disposal with states or compacts that have a disposal facility. Currently, Massachusetts does not have a disposal facility and is not a member of any compact. However, under an agreement between Massachusetts and the Southeast Low level Radioactive Waste Compact Commission (SLLRWCC), Massachusetts waste generators will be able to use the Barnwell, South Carolina waste disposal site until July 1994. INS intends to ship its waste to Barnwell I until July 1994, at the same frequency as in the past. See Section IILA.4, below. A number of other states are in the same situation as Massachusetts, i.e., they neither have a disposal site nor belong to a compact that has access to a disposal site. Beginning in July 1994, when Barnwell is scheduled to close its doors to states that do not belong to the SLLRWCC, the NRC recognizes l* that waste generators in these states may have no other choice but to store their LLW. Indeed, a few states (Michigan, Maine, New 11ampshire, Rhode Island, and the District of Columbia) have no disposal option at this time. Although the NRC encourages permanent disposal of LLW, and views storage as an option [ of last resort, the NRC understands that onsite interim storage may be necessary t in certain cases. Many waste ' generators also store LLW for short periods to - ~{ permit decay of very short-lived radionuclides, or to accumulate enough to i ship efficiently. In order that both short-term and long-term storage may be d accomplished safely, the NRC has developed regulations and guidance for LLW storage. Current requirements for LLW storage a; pear in 10 C.F.R. parts 20, i 30,40,50, and 70. Various guidance documents have also been published, for example, Information Notice (IN) 90-09, " Extended Interim Storage of low-level Radioactive Waste by Fuel Cycle and Materials Licensecs" and IN 89-13 " Alternative Waste Management Procedures in Case of Denial of Access to I.cw-l 370 i

level Waste Disposal Sites." Finally, in addition to the storage requirements and ' + guidance that NRC provides, NRC fuel cycle and materials licensees, including INS, are subject to regular inspections and to license reviews to assess safety and d:termine that licenses meet applicable requirements, including those related to waste storage. In 1992, INS completed a new onsite storage facility for radioactive waste. It is located underground, adjacent to the health physics laboratory, and accessible only from inside the building. He new storage facility replaces 112 storage of. waste in smilers in the parking lot next to one of the Licensee's buildings. The storage area is constructed of concrete and steel and includes a fire suppression system, a liner / collection rystern around the exterior walls and floors to direct any potential releases to a sump for collection and subsequent sampling,'and an air sampling system. Waste that is placed in this facility is already packaged for shipment. De facility is monitored on a daily basis for airborne contamination, removable contamination, and radiation levels. De NRC Staff concludes that the use of INS's new radioactive waste storage facility will increase the protection of the public health and safety because INS will be better able to monitor radiation . emissions from the waste and, if necessary, contain radioxtive releases. In July 1994, INS may have to hold its radioactive waste on site when Barnwell is scheduled to cease accepting out-of-compact waste. He new radioactive waste storage facility at INS has sufficient capacity to hold approximately 5 years of waste. At this time, the NRC Staff concludes that there is no health or safety problem related to the January 1; 1993 deadline date published in the Springfield Sunday Republican. Based on the above, I conclude that Petitioners have not raised a i substantial health or safety concern. 3. In 1989, INS Waste Stored Was Twice the 1*olume Shipped As requested by Petitioners, NRC inspectors conducted an unannounced inspection on July 8 and 9,1992. A review of INS's radioactive shipping . manifests showed that during the 4 years from 1989 to 1992, INS shipped i a total volume of.6,455.7 cubic feet of radiortive waste for final disposal at a commercial low-level radioactive waste disposal site. His averages to approximately 1,613.9 cubic feet of waste generated per year by INS during this period. Due to the limited shipping capacity of tie waste shipment truck, INS needs to make 1% waste shipments per year in order to dispose of the yearly amount of waste it generates. To maximize the use of its waste shipment truck, INS has been making two shipments every other year, and one shipment in the alternate years. Under this shipping schedule, no radioactive waste generated at INS is held for onsite storage for more than 2 years. For the Star 1989, NRC inspectors noted that INS shipped a total volume of 2,125.3 cubic feet i .f 371 I I

f r of radioactive waste, which is more than the amount of waste generated for t that year but not twice as much. INS has not exceeded the 2-year limit for onsite radioactive waste storage in the INS license. Accordingly, I conclude that Petitioners have not raised a substantial health or safety concern. ' 4. INS Stated That n'aste is Remowd Twice a Year But NRC Records indicate That it is Remond Only Once a Year and Not at A!!in 1990 -j As discussed above, NRC inspectors noted in their inspection report that INS + made two shipments in 1989, one shipment in 1990, two shipments in 1991, and one shipment to the time of the inspection in 1992. On the average, INS needed 'I to make approximately 1 W shipments per year, resulting in two shipments every other year. Accordingly, I conckdc that Petitioners have not raised a substantial health or safety concern. S. A Person Standing at the INS Fencefor 2 Days in Early May Would liare Received a liigher Radiation Dose Than a Person Standing at Vermont YanLee's Fencefor a Year Because of Tighter Regulationsfor l Nuclear Power Plants It appears that the Springfield Sunday Republican article concerns the direct radiation levels at the INS fence due to the presence of the INS waste-pickup truck compared to the annual air dose at Vermont Yankee's fence due to its gaseous effluents. Petitioners are correct that NRC's exposure limits for individual members of the general public near materials facilities such as INS are different from those for individual members of the general public near nuclear power reactors. NRC materials licensees must comply, beginning on January 1,1994, with the I millisievert (100 millirems) per year WC limit to the maximally exposed member of the general public. 10 CJ 6 20.1301. In addition, materials licensecs, such as INS, must comply wit'..a ALARA requirement which states, "[t]he licensee shall use, to the extent practicable, procedures and engineering controls based upon sound radiation protection principles to achieve occupational a doses and doses to members of the public that are as low as is reasonable. achievable (ALARA)." 10 CER. 5 20.1101(b). i Nuclear power reactors are required to comply with requirements in 10 Cf.R. Part 20 as well as with technical specification requirements to meet tie criteria in 10 CF.R. Part 50, Appendix 1. Nuc! car power reactors and fuel cycle facilities are also required to meet the Environmental Protection Agency's (EPA) Uranium Puel Cycle Standard of 0.25 millisievert (25 millirems) per year. Sec 40 CE.R. Part 190. The annual limit of 0.25 millisievert (25 millirems) for a maximally i 372 i i

f t exposed individual was derived by EPA based on ALARA considerations. For gaseous effluents, the NRC Part 50, Appendix I criterion mentioned in the newspaper article and recited by the Petitioners is 0.05 millisievert (5 millirems) per year to a maximally exposed member of the general public. Direct radiation exposure to a maximally exposed metaber of the general public at the fenceline is not specifically addressed in Appendix I. However, nuclear power reactors ( are required to meet the 0.25 millisievert (25 millirems) per year EPA limit i that includes exposures from direct mdistion exposure as well as from gaseous j cffluents.10 C.F.R. 6 20.105(c). i Although there are differences in the regulatory limits for nuclear power reactors and for materials facilities, the differences are based on whether ALARA has been incorporated into the limits for a cenain category oflicensees t (i.e., nuclear power reactors and fuel cycle facilities), or must be considered in addition to the limits (i.e., materials facilities). Rese limits are all significantly below any observable health effects that could affect the public. NRC inspectors have found that the radiation levels at the INS fenceline are well within NRC limits. See Section III.A.1, above. Moreover, INS has moved the location of f f its laundry and waste-pickup trucks to reduce radiation levels at those fenceline locations described in the Springfield Sunday Republican article, in keeping with ALARA. Accordingly, I have concluded that Petitioners have not raised a substantial health or safety concern. 6. There Hart Been Allegations That INS Discharges Radioactive Water into the City Sewer System He Commission's regulations allow the discharge of liquids, containing very .j low levels of radioactive materials, into the sanitary sewer.10 C.F.R. 6 20.303. Licensees are required to monitor and control any such discharges and to make available the documentation of such discharges for NRC inspection.10 C.F.R. p 620.401. Water used by INS for nuclear laundry purposes is first filtered to remove as much of the radioactive materials from the water as possible. This water then goes into holding tanks where the water is sampled for radioactivity i levels and compared to NRC. authorized limits before release into the sanitary sewer. The July 8-9, 1992 unannounced NRC inspection of INS found no violation of NRC limits concerning releases to sanitary sewers. In addition, NRC inspectors took a water sample from INS's wastewater holding tank and, f by independent measurements, found the radioactivity levels in the water to be within NRC limits. Accordingly, I conclude that Petitioners have not raised a substantial health or safety concern. 373 t k I l l I

B. The NRC Staff has evaluated Petitioners' ten requests for responses or actions by the NRC. That evaluation and my disposition of each of the ten I requests are discussed below. Petitioners requested that the NRC: 1. Participate in a Public Hearing in Indian Orchard to Respond to the Concerns of Neighborhood Residents In response to this request, representatives of NRC and the Commonwealth of Massachusetts attended a public meeting on the evening of July 23,1992, at a local American Legion post The meeting was attended by approximately. 75 people and lasted about 2% hours. The meeting was moderated by Mrs. Linda llammons of the IOCC. At this meeting, NRC Staff discussed with the - attendees the results of the NRC inspection on July 8-9, 1992, and answered all health and safety concerns directly with members of IOCC. Therefore, this request has, in effect, been granted. l 2. Hold a Surprise inspection ofINS Although the NRC Staff had conducted an unannounced inspection at INS in December 1991, the NRC Staff conducted another unannounced inspection on July 8-9, 1992, to review recent events and to provide a current basis for i the discussions scheduled at the July 23,1992 public meeting. A representative from the Department of Public IIcalth of the Commonwealth of Massachuseus accompanied the NRC inspectors on July 8,1992. Copies of the NRC Inspection Report for the July 8-9,1992 inspection utre sent to IOCC before the July 23, 1992 meeting, for discussion at that meeting. In addition, extra copies of the NRC Inspection Report were made available to all attendees at the leginning of the public meeting. This request has, therefore, been granted.. 3. Check Homes in the Areafor Radioactive Contamination - NRC does not normally monitor private houses for radioactive contamination. Based on radiation surveys and soil sample measurements taken by NRC inspectors along the Licensee's fenceline, and a review of INS survey records, the Staff does not have any technical basis to conclude that local homes could have been contaminated due to loss of radiological control at INS. 'Ihis information was made available, through the inspection report, to the 10CC. Ilowever, the Commonwealth of Massachusetts' personnel have taken radiation readings in the local area with several neighbors in auendance. No radiation levels above normal background were found. Petitioners have presented no substantial health or safety concern. Accordingly, this request is denied. 374 P

r i 4 Provide to the Petitioners a Copy of the NRC Regulations Under 1Vhich INS Operates By letter dated July 21,1992, Richard Cooper,11, Director of the Division of Radiation Safety and Safeguards, NRC Region I, provided copics of the NRC regulations under which INS operates to Gloria Mitchell, President of IOCC. l This request has, therefore, been granted. A S. Check Adjoining Park Department isnd, Including Dimmock Pond,for Contamination and !!!egal Dumping of lYaste Afaterial During Oc July 8-9,1992 inspection, NRC inspectors took direct radiation l readings around the Dimmock Ibnd area, including die unimproved road be-tween tie Ibnd and the Licensee's property and trails along the Parker Avenue side of the Pond. No readings above normal background were detected during these surveys. 'Ihe inspectors also took two water samples from Dimmock Pond to cleck for radioactivity. In addition, the NRC inspectors obtained a sediment sample, consisting of a composite sample taken from Dimmock Pond near the Licensce's property. Finally, the Commonwealth of Massachusetts also took a water sample and a sediment sample from Dimmock Ibnd. Analyses of all these samples have identified no detectable radiation levels or radioactive materials above normal background. Based on the results of the above measurements, [ review of INS's radioactive waste storage and shipping records, and other in-spection results, the Staff has no information that could demonstrate that there .- i has been illegal dumping of waste material by INS. The request to conduct surveys, therefore, has been granted. 6. Check Loon Pondfor Contamination andfor Possible filegal Dumping of \\Yaste Afaterial During the July 8-9, 1992 inspection, NRC inspectors did not obtain any evidence that supported the allegation that there may have been illegal dumping of radioactive waste material in IAon Ibad. Further, since no radioactive contamination was found in Dimmock Pond, which is adjacent to the INS property, the Staff concluded that sampling Loon Pond, physically separated from the INS property by Parker Street and railroad tracks, and several hundred yards away, would le neither necessary nor reasonable. I conclude that 3 Petitioners have presented no substantial health or safety concern. Tierefore, 7 this request is denied. i 5 375 i I

I i 7. Determine What INS Has Done with Waste MaterialNot Shipped INS is required by its license to safely store its radioactive waste at NRC-authorized locations. Prior to May 1992, INS stored its radioactive waste mside trailers located next to one of its buildings. In May 1992, INS began using a newly constmeted storage facility for radioactive waste. His onsite storage facility has been described earlier. See Section III.A.2, above. At this time, INS is using the new storage facility only for short-term storage of radioactive waste, in compliance with its license. INS's NRC license does not currently permit the storage of any radioactive waste at INS for more than 2 years. INS wauld have to submit an application for amendment of its license, as discussed in accordance with Information Notice 90-09, " Extended Interim Storage of IAw-level Radioactive Waste by Riel Cycle and Matenals Licensees," if INS wished to store its radioactive waste for a period longer than 2 years. De Petitioners' icquest, therefore, has been granted. 8. Provide to the Petitioners the DocLet Numberfor INS By letter dated August 25, 1992, Robert M. Bernero, Director Office of r Nuclear Material Safety and Safeguards, provided the docket number for the t INS license to Gloria Mitchell and Linda Hammons of IOCC. His request, therefore, has been granted. 9. Identify a Public Docurnent Room (PDR)for INS and Its Location By letter dated July 21, 1992, Richard Cooper, II, Director, Division of Radiation Safety and Safeguards, NRC Region I, provided this information to Gloria Mitchell of IOCC. He location of the NRC Public Document Room for INS is at the NRC Region 1 office at 475 Allendale Road, King of Prussia, - Pennsylvania,19406. His request, therefore, has been granted.

10. Describe the Monitoring Done, Who Does it, and How Frequently INS is required to perform radiation surveys as are necessary to comply with l

10 C.F.R. Pan 20 and to evaluate the extent of radiation hazards that are or may be present.10 C.F.R. 6 20.201(b). De licensee must also maintain records of these surveys.10 C.F.R. 5 20.401. The particular types of radiation surveys that INS performs to satisfy NRC requirerents were approved by the NRC during the licensing process and are described in the July 8-9,1992 NRC inspection report, a copy of which was sent to Gloria Mitchell of IOCC before the July 23,1992 public meeting. In addition, extra copies of the inspection report were 376

i made available to all attendees at the start of the public meeting. This request, therefore, has been granted. 5 C. I have considered Petitioners' four " demands" on behalf of neighborhood residents, and deny three demands, the fourth having been mooted by INS's voluntary actions, for the reasons stated below. Petitioners demand that: 1, Radiation Readings Outside the INS Fence Perimeters Be 0" at A!! i ~ Times l As noted above, the average annual background external radiation to a [ member of the general public is about I millisievert (100 millirems) per year. Therefore, it is not possible to achieve a radiation reading outside the INS fence perimeters of"0" at all times. Nonetheless, members of the general public ought not to be exposed to any more radiation above background from NRC-licensed activities than is absolutely necessary, regardless of whether the radiation level is within NRC limits.10 C.F.R. 6 20.l(c). This is the NRC's ALARA policy. In keeping with the ALARA policy, INS is reviewing the staging of transient waste and laundry shipping trucks to reduce the potential of an.' fenceline radiation exposure. The NRC will continue to monitor INS's ALARA program through inspection and licensing actions. Petitioners have not raised a substantial health or safety concern. Accordingly, this demand is denied. 2. "0" Nuclear Waste Byproducts Be Allowed to Enter Springfield's Water / Sewer System -i NRC regulations require licensees to monitor and document their releases-into the sanitary sewer.10 C.F.R. 5 20.401. Licensees are limited in terms of both the concentration and quantity of radioactive materials that can be disposed via the sanitary sewer.10 C.F.R. f 20.303. The levels of radioactivity permitted l to be put into the sanitary sewer are considered by NRC not to present any threat to the public health or safety. NRC inspectors did not find any sanitary sewer ~ releases to date by INS in excess of NRC limits. In addition, the NRC Staff authorized INS, by license amendment dated October 8,1992, to use a.new liquid waste treatment system wiach should improve the Licensee's capability to filter out radioactive materials from its laundry wastewater before disposal into the sanitary sewer. Moreover, in the new 10 C.F.R. 5 20.2003(a)(1), which will become effective on January 1,1994, the type of radioactive materials that can be disposed into the sanitary sewer is clarified to further restrict the type of materials allowed in water. Current technology is not capable of filtering out all radioxtive materials from wastewater before it is discharged into the sanitary 377.

t P f sewer. To require "0" rufcases would go beyond the bounds of the ALARA [ policy and technical feasibility. Petitioners have raised no substantial health or safety concern. Accordingly, this demand is denied. I 3. INS Stop Using Residential Streets, Specifically Nogle and Nichols Streets, to Go to andfrom its Plant All NRC licensees who transport licensed material outside the confines of their plant or other places of use must comply with appropriate DOT -{ requirements in 49 C.F.R. Parts 170-189. 10 C.F.R. 6 71.5. In the most recent inspection of INS, NRC inspectors did not find any violation of DOT requirements. Although there are no DOT restrictions on the use of residential l i streets, INS has voluntarily submitted a plan to IOCC to use an alternate route that does not include resMential streets. IOCC las accepted INS's plan. Accordingly, this demand has been satisfied by the Licensee's voluntary actions f and is moot. 4. Under No Circumstances Should INS Be Allowed to Store Nuclear l ' Waste on its Property 'Ihe NRC Staff recognizes the concerns of the local community with regard to - the long-term storage of radioactive waste on a licensce's property. Should INS wish to store its radioactive waste for a longer period than what is currently allowed under its license, it must submit a license amendment application to the NRC. NRC Information Notice No. 90-09 provides guidance to fuel cycle and materials licensees on information needed in license amendment requests to authorize extended interim storage of low-level radioactive waste (LLW) at licensed operations. As stated in this information notice, NRC does not consider storage as a substitute for disposal. Ilowever, NRC will consider extended interim storage of LLW at the Licensee's site only if disposal is i not,a viable option and the waste can be stored safety. Information Notice No. 90-09 provides the ir. formation that the licensee must submit to the NRC in order for NRC to make a health and safety determination. Ibr'a facility such as INS to continue to operate, a certain amount of radioactive waste will necessarily be generated. Also, INS storage activities are covered by NRC's regulatory (including inspection) program for storage, as described earlier in Section llI.A.2. The NRC will continue to monitor the Licensce's activities to i ensure that public health and safety will not be compromised.. In view of the above, and the Licensee's compliance with NRC's regulatory limits, Petitioners have raised no substantial health or safety concern. Accordingly, this demand is denied. [ t 378 l t

1 IV. CONCLUSION 1 i he institution of proceedings pursuant to 10 C.F.R. 62.202 is appropriate only where substantial health and safety issues have been raised. Sec Consol-idated Edison Co. of New York (Indian Point, Units 1, 2, and 3), CL1-75-8. 2 NRC 173,175-76 (1975); Washington Public Power Supply System (WPPSS Nuclear Project No. 2), DD-84-7,19 NRC 899,923 (1984). This is the standard that I have applied to determine whether the actions requested by Petitioners are warranted. F ne Staff has carefully considered the ten " requests" and four " demands" of Petitioners. In addition, the Staff has evaluated the bases for Ittitioners' i requests and demands. Ibr the reasons discussed above, there are no substantial public health and safety concerns warranting NRC action concerning the "four l demands" of Petitioners. Accordingly, three of the Petitioners

• demands are de-

[ nied and one demand was mooted by the voluntary action of the Licensec. Eight of the Petitioners' requests were granted insofar as NRC Staff: participated in 5 a public meeting on the evening of July 23,1992, at a local American Legion l hall and responded to the concerns of the neighborhood residents; conducted an unannounced inspection of INS on July 8 and 9,1992; provided IOCC with. copies of pertinent portions of NRC's regulations; checked adjoining Park De- - j partment land, including Dimmock Pond, for contamination; reviewed INS's waste storage program; provided IOCC a description of INS's radiation mon-itoring program; identified the location of the Public Document Room (PDR) for the INS license; and provided the docket number for the INS license. The Petition is denied with respect to IOCC's requests to check homes in the area for radioactive contamination and to check Loon Pond for contamination and possible illegd dumping of waste material, because Petitioners failed to raise a substantial health or safety concern. As provided by 10 C.F.R. 6 2.206(c), a copy of this Decision will be filed with the Secretary of the Commission i for the Commission's review. The Decision will become the final action of the I 379 L E

t 3 Commission twenty-five (25) days after issuance unless the Commission on its _j own motion institutes review of the Decision within that time. } i FOR Tile NUCLEAR lj REGULA7 DRY COMMISSION i' Robert M. Bernero, Director Office of Nuclear Material Safety and Safeguards Dated at Rockville, Maryland, this 7th day of May 1993. i i i a ? t ~t h r i I 1 0 e i I 380 I I I [ f + r i

I l Cite as 37 NRb 381 (1993) DD-93-10 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION Thomas E. Murley, Director in the Matter of Docket No. 50-220 NIAGARA MOHAWK POWER CORPORATION (Nine Mlle Point Nuclear Station, Unit 1) May 9,1993 The Director of the Office of Nuclear Reactor Regulation grants in part and denies in part a petition filed with the Nuclear Regulatory Commission (NRC) Staff by Ben L. Ridings (Petitioner) on October 27, 1992, requesting that the NRC issue an immediately effective order directing Niagara Mohawk Power Corporation (NMPC) to cease power operation of Nine Mile Point Nuclear Unit 1 (NMP-1) and place the reactor in a cold-shutdown condition. The Petitioner sought relief based on allegations that (1) NMPC is operating NMP-1 in violation of the requirements for availability of emergency core cooling system (ECCS) high-pressure coolant injection (HPCl), (2) NMPC has failed to provide the mandatory emergency backup power to the HPCI system at NMP-1, and (3) 45% of the containment isolation valves have administrative i deficiencies. Petitioner also alleged that NMPC, NMPC's quality assurance group, and the NRC had reviewed these safety concerns and, contrary to any practical justification, had remained silent. Petitioner concluded that the NRC and NMPC's quality assurance organization have failed to remain independent, and therefore called for a congressional investigation into these matters. PLANT DESIGN: GENERAL DESIGN CRITERIA The General Design Criteria (GDC) do not apply to plants with construction permits issued prior to May 21,1971, such as NMP-1. Such plants comply with the intent of the GDC, and do not need exemptions imm the GDC; they were 381 s

evaluated on a plant-specific basis, determined to be safe, and licensed by the Commission. PLANT DESIGN: ECCS Ibr a plant with an Emergency Core Cooling System comprised of redundant. [ safety-grade core spray and automatic depressurization systems designed to accommodate the range of loss-of-coolant accidents from the smallest up to the largest line break, a safety-grade feedwater system operating in an IIPCI i mode is not needed to satisfy ECCS requirements. i PRICE-ANDERSON ACT (INSURANCE) i The Price-Anderson Act requires each utility to provide $200 million in liability insurance for public liability claims that might arise from a nuclear accident at its site. In addition, all commercial nuclear power plant licensees must participate in an industry public liability self-insurance plan which subjects each licensee to a potential liability of $63 million for each commercial nuclear - power plant that it operates for claims that might arise from a single nuclear accident at any commercial nuclear power plant licensed by the NRC. NMPC has obtained and is maintaining the appropriate amount of liability insurance. -j i I TECliNICAL ISSUES DISCUSSED 'Ihe following technical issues are discussed: Com spray system; Automatic depressurization system; Feedwater system in high-pressure coolant injection 2 mode: Non-safety-related methods for coolant injection; Inservice testing pro-i' gram applied to feedwater system and containment isolation valves; Appendix J testing of containment isolation valves; Consistency and completeness of tech-nical specifications and updated Final Safety Analysis Report regarding listing of containment isolation valves, valve stroke time requirements, and valve actu-ation signals; and Licensee self-assessment program - Regulatory Compliance Group, Quality First Program. 'i DIRECTOR'S DECISION UNDER 10 C.F.R. f 2.206 [ -i Introduction On October 27, 1992, Mr. Ben L. Ridings (Petitioner) filed a Petition for consideration in accordance with 10 C.F.R. 62.206 with the Nuclear Regulatory 382 b r i L w t

R Commission (NRC or Commission). The Petitioner requested that the Commis-sion take direct review of the Petition. However, the Commission declined to take direct review and referred the Petition to the Director, Office of Nuclear Reactor Regulation (NRR), for consideration. 'Ihe Petitioner sequested that the NRC issue an immediately effective order directing Niagara Mohawk Power Corporation (NMPC) to cease power operation of Nine Mile Point Nuclear Station Unit 1 (NMP-1) and place the reactor in a cold-shutdown condition. The Petition also asked the Commission to hold a public hearing before authorizing resumption of plant operadon. As bases for these requests, the Petitioner asserted that (1) NMPC is operating NMP-1 in violation of the requirements for availability of an emergency core cooling system (ECCS) high-pressure coolant injection (HPCI) system including the failure to provide the mandatory emergency backup power to the 11PCs system; (2) 45% of the containment isolation valves have administrative deficiencies; and (3) NMPC, NMPC's quality assurance group, and the NRC have reviewed these safety concerns and, contrary to any practical justification, have remained silent 'lhe Petition was placed in the Public Document Room and a copy of the Pention was sent to NMPC in a letter of November 19, 1992, for NMPC's review and comraents regarding the issues raised in the Petition. In a letter of December 21,1992. NMPC commented on the issues raised in the Petition. In a leuer of December 4,1992, I acknowledged receipt of the Petition, informed the Petitioner that the Commission had declined to take direct review of the Petition, denied Pedtioner's request for immediate action, and told the Pentioner that a final decision on the Pention would be issued within a reasonable time. My December 4,1992 letter to the Petitioner also requested that the Petitioner give the NRC some specific information that was not fully legible or not provided in the Petition, in response to my request for specific information, the Petitioner submitted a document titled *Information Requested by Office of Nuclear Reactor Regu-lation" as an attachment to a letter received by tne NRC Office of the Execudve Director for Operations on January 5,1993. In his response, tic Petitioner also asserted that the NMP-1 facility will not meet the leakage limits of 10 C.F.R. Part 50, Appendix J, when the leakage rates of Category A containment isola-tion valves are added to the leakage total for the NMP-1 containment building. (As defmed by ASME Code 9 XI, Category A valves are those for which seat leakage is limited to a specific maximum amount in the closed position for fulfillment of their function.) In addition, the Petidoner contends that NMPC's asserted failures to comply with tre requirements of 10 C.F.R. Part 50 precludes NMFC from operating NMP-1 with limited liability.' A copy of the Petitioner's response was sent to NMPC in a letter of January 11,1993, for NMPC review and comments regarding the issues raised in the response. In a letter of February 383 .l 1 i

9,1993, NMPC commented on the issues raised in the Petitioner's response. A copy of the Petitioner's response was also placed in the Public Document - Room. I have now completed my evaluation of the Petition and the Petitioner's response ("Information Requested by Office of Nuclear Reactor Regulation"). De Petitioner's request for correction of the NMP-1 Technical Specification (TS) to correctly list the NMP-1 containment isolation valves, their initiating signals, and their stroke times is granted However, for the reasons given in the discussion below, the Petitioner's request for other actions is denied. Discussion The NRC Staff's evaluation of the Petitioner's assertions follows. 1. NMP-1 Does Not Meet NRC Requirementsfor an ECCS HPCI System De Petitioner asserted that NMP-1 does not meet NRC requirements for an ECCS HPCI system for the following reasons: (a) NMP-1 fails to meet General Design Criterion (GDC) 33, " Reactor coolant makeup"; GDC 35, " Emergency core cooling"; GDC 36, "In-spection of emergency core cooling system"; and GDC 37," Testing of emergency core cooling system," of Appendix A," General De-sign Criteria for Nuclear Power Plants," to 10 C.F.R. Iwt 50 because NMP-1 does not have an ECCS HPCI system to provide abundant emergency core cooling in the event of a small-break loss-of-coolant accident (LOCA). Petitioner also asserted that the feedwater system operating in its HPCI mode is not an acceptable alternative system because it does not have a backup electric powe supply from an onsite emergency diesel generator. (b) Of the forty-seven valves in the feedwater injection flow path, forty-four are not included in the NMP-1 inservice testing program for pumps and valves. The NRC Staff's review of these issues and conclusions is based on the original design and licensing basis of NMP-1, as follows. NMP-1 is a General Electric boiling-water reactor with a Mark I containment. After appropriate review and evaluation by the staff of the U.S. Atomic Energy Commission (AEC), predecessor regulatory agency to the NRC, tte NMP-1 Consuuction Permit was issued to NMPC on April 21,1%5. On March 24,1969, the AEC staffissued a irport to the Advisory Committee on Reactor Safeguards in which the AEC staff stated I 384 d 1 i

k We recognize diat the NMP facility was not designed in accordana with the current set d the Cornmission's general design criteria. Ilowever, as discussed in our evaluation, the interers features and capatslity provide a basis for reasonatile assurana that the facility design meets the intera of tie criteria. He NMP-1 Provisional Operating License was issued to NMPC on August 22,1969. He " Technical Supplement to Petition for Conversion from Pro-visional Operating License to Full-Term Operating License," dated July 1972, gave information related to the extent to which NMP-1 conforms to the GDC. He NRC did not require NMPC to design NMP-1 in accordance with the GDC because NMP-1 was evaluated on a plant-specific basis, determined to be safe, and licensed by the Commission. The NRC Staff also notes that NMP-1 received a construction permit on April 21,1965, a date that preceded the issuance of the GDCs in Appendix A to 10 C.F.R. Part 50. (The GDCs were issued on May 21,1971.) In a September 18, 1992 Staff Requirements Memorandum (SRM) to the NRC Executive Director for Operations, the Commission set forth its position that the NRC Staff will not apply the GDCs to plants with construction permits issued before May 21, 1971. The SRM continued: At the time of promulgation of Appendix A to 10 CI'R Part 50. the Commission stressed that the GDC were not new requiremeras and were promulgated to more clearly aniculate the licensing requirements and pract',ce in effect at that time. While compliance with the intern of the GDC is imponant, each plant licensed before the GDC were formally a&pted was evalualcd on a plant specific basis, determined to be safe, and liansed by tte CommissiorL Ibrthermore, current regulatory processes are sufficient to ensure that plants continue to be safe and comply with the irnent of the GDC. llackfsning the GDC would provide little or no safety benefit while requiring an extensive carnmitment of resources Plaras with construction permits issued prior to May 21,1971, do not need exemptions from tie GDC. Herefore, GDC 33,35,36, and 37 do not apply to NMP-1. He AEC published its acceptance criteria for emergency core cooling systems for light-water power reactors on' January 4,1974 (39 Fed. Reg.1003). His then-new regulation added Appendix K to 10 C.F.R. Part 50 which specifies analytical techniques to be employed for the evaluation of ECCS acceptability. NMP-1 was originally licensed to the Interim Acceptance Criteria of 10 C.F.R. t 50.46, which were effective while the AEC was promulgating this regulation. He AEC Safety Evaluation Report of Decen.ber 27,1974, concluded that the NMP-1 ECCS satisfies the requirements of section 50.46 and Appendix K to Part 50 as finally promulgated. Dat conclusion was reached without selying on or taking credit for the feedwater system operating in its HPCI mode. Moreover, NMP-1 meets the intent of the GDC by providing redundant methods for reliably cooling the reactor core (and meeting the requirements of section 50.46) under 385

postulated accident conditions. The provisional operating license was convened to a full-term operating license on Decemter 26,1974 The following is a summary of the NRC Staff's analysis of how the NMP-1 ECCS satis 5es NRC requirements. The NMP-1 ECCS includes the core spray system (CSS), consisting of two seprate and independent loops, and an automatic depressurization system (ADS). He CSS and ADS are described in UFSAR llVll.A and V.B.5.0, respectively. Each CSS loop consists of two 100% pump combinations (i.e., two core spray pumps and two core spray topping pumps). He maximum discharge pressure of each pump combination is approximately 350 psig. The four core spray pumps and four core spray topping pumps get electric power from offsite sources or from the onsite emergency diesel generators. De logic for the ADS is powered by ac emergency power supplies, and the six electrically actuated relief valves get electric power from the de emergency power supplies (station batteries in parallel with battery chargers). He CSS is a safety-related system that is designed to accommodate the range of loss-of-coolant accidents from the smallest up to the largest line break. For large breaks, the CSS can maintain the peak cladding tempemture within the acceptance criteria of section 50.46 without assistance from the ADS because the i reactor depressurizes sufficiently fast for the CSS to achieve rated flow before the criteria of section 50.46 are exceeded. For small breaks, i.e., breaks below about 0.30 square foot, the ADS is provided and it will operate to depressurize the reactor to permit water injection by the CSS before the criteria of section 50.46 are exceeded. The criteria of section 50.46 are not exceeded, assuming a single failure that disables one of the two available CSS loops and without taking credit for operation of the feedwater system in the HPCI mode. In addition to the CSS and ADS, NMP-1 also has and utilizes the feedwater system operating in an HPCI mode and two control rod drive pumps operating in the coolant injection mode to inject water into the reactor at reactor operating pressure in the event of a small-break LOCA. Successful operation of these systems is desirable since their proper operation may negate the need to unnecessarily actuate the ADS valves. However, the NMP-1 LOCA safety analyses do ret rely on wate'r injection by either the control rod drive pumps or the feedwater system operating in the HPCI mode to satisfy the requirements of section 50.46. The foregoing conclusion has been reaffirmed in the General Electric Com-pany's (GE's) IDCA analysis for each subsequent fuel cycle. Each analysis was performed to demonstrate compliance with the requirements.of section 50.46 without taking credit for the feedwater sys:cm operating in the HPCI mode. He analysis for the current fuel cycle was prepared in response to the require-ments of NMP-1 TS 6.9.1.f," Reporting Requirements, Core Operatmg Limits Report." l 386 1 1 s i

Therefore, the NRC Staff concludes that the Petitioner's assertion that the NMP-1 HPCI system (feedwater system operating in the HPCI mode) must meet GDC 33,35,36, and 37 and that the HPCI system must be part of the ECCS and be supplied with backup electrical power from an onsite emergency diesel generator is incorrect. NMP-1 does not have and does not need an ECCS HPCI system because the existing NMP-1 ECCS satisfies the requirements of section 50.46 and Pan 50, Appendix K, without reliance on the feedwater system operating in the HPCI mode. Non-Safety-Related Methodsfor Coolant injection in addition to the CSS and ADS, NMP-1 has two control rod drive pumps which can be operated in the coolant injection mode and with a feedwater system which can be operated in an HPCI mode to inject coolant at reactor operating pressure. Each control rod drive pump is rated at 85 gpm at a head of 3760 feet. Operation of the control rod drive pumps in the coolant injection mode is described in section X.C of the UFSAR. Operability of the control rod drive pumps in the coolant injection mode is required by TS 3.1.6," Control Rod Drive Pump Coolant Injection." Electric power for the control rod drive pumps comes from either offsite sources or from the onsite emergency diesel generators. Operation of the feedwater systera in the HPCI mode is described in section Vill of the UFSAR. Operability of the HPCI system is requin:d by TS 3.1.8, "High Pressure Coolant Injection." The HPCI system utilizes the two condensate storage tanks, the main condenser hotwell, two condensate pumps, condensate demineralizers, two feedwater booster pumps, feedwater heaters, two motor-driven feedwater pumps, an integrated control system, and all associated piping. and valves. The HPCI system is capable of delivering 6840 gpm into the reactor vessel at reactor pressure when using two trains of feedwater pumps. The HPCI - system gets electric power from normal offsite sources by either of the two 115-kV lines, but not from the onsite emergency diesel generators. NMP-1 also has the capability of automatically realigning the HPCI system to receive electric power from a dedicated generator at the Bennetts Bridge Hydro Station in the event of a loss of power to both II5-kV offsite lines. Although this hydrogenerator is not equivalent to an onsite emergency diesel generator, it is a highly reliable source of backup power. Operation of the control rod drive pumps in the coolant injection mode and the feedwater system in the HPCI mode is described in the UFSAR. The contral rod drive pumps and the HPCI system are required to be operable by the NMP-1 TS. The control rod drive pumps and the HPCI system are required to be operable by the TS to provide coolant injection without unnecessarily actuating the ADS valves. However, the NMP-1 safety analyses do not rely on operation of the control rod drive pumps in the coolant injection mode or on operation of 387

i i i the feedwater system in de 11PCI mode to provide emergency core cooling or to meet the acceptance criteria of section 50.46. NMP-1 was designed and constructed, and began operation (Provisioral Operating License issued on August 22, 1969), before May 21,1971, when j the GDCs were issued. He final emergency core cooling acceptance criteria of section 50.46 were satisfied by the NMP-1 ECCS (one out of two loops of_ the CSS operating in conjunction with the ADS) without reliance on either the control rod drive pumps operating in the coolant injection mode or tte feedwater system operating in the HPCI mode. Therefore, neither the control rod drive pumps nor the feedwater system operating in the HPCI mode is required to meet section 50.46 criteria for EOCS equipment. Applicability ofIST Program to Feedwter System With regard to the Petitioner's concern regarding the failure to include forty-four of the forty-seven valves in the feedwater injection flow path in the NMP-1 inservice testing (IST) program for pumps and valves, as discussed above, the NMP-1 safety analyses do not rely on feedwater system operation in the HPCI mode to provide emergency core cooling or to satisfy the criteria of section 50.46. Furthermore, the feedwater system is not otherwise required to be a l safety-related system. Ibr nuclear power facilities whose construction permits were issued before January 1,1971 (as was the case for NMP-1), paragraph (f) i of 10 C.F.R. 0 50.55a requires the IST programs for those facilities to include, to the extent practical, IST requirements for pumps and valves classified as ASME [ Code Class 1,2, and 3. However, the NMP-1 feedwater system is not a safety-related system and is not classified as an ASME Code Class I,2, or 3 system. .t Therefore, the Commission's regulations do not require these valves to be part of tie NMP-1 IST program. However, the feedwater isolation valves (31-0IR,31-02R,31-07, and 31-08) function as part of the reactor coolant system pressure l boundary and are, therefore, included in the NMP-1 IST program for pumps and valves. Rese valves are also containment isolation valves and, as such, are included in TS Table 3.2.7. IIPCI System-Conclusion in summary, the HPCI system is not required to meet GDC 33,35,36, and f 37, it is not required to be part of the ECCS with backup electric power from an onsite emergency diesel generator, nor is its operation required to satisfy the emergency core cooling requirements of section 50.46. The existing ECCS g satisfies the emergency core cooling requirements of section 50.46 without reliance on the non-safety-related feodwater systern operating in the F.*CI mode. [ -[ 388 i t -h

l I ne Petitioner does not raise any new issues regarding the design or operation of t!e feedwater system operating in the HPCI mode. Accordingly, I find that the Petition contains no basis to order a shutdown of NMP-1 or to institute such a proceeding as requested by the Petitioner; therefore, dds portion of the Petition is denied. 2. Forty-Fire Percent of the Containment Isolation Valves liare Administrathe Deficiencies; the NMP-1 Facility WillNot Meet the Exakage Limits of JO C.F.R. Part 50, Appendix J, When the Leakage Rates of Category A Containment isolation Yahts Are Added to the Leakage Totalfor the NMP-1 Containment Building De Petitioner asserted that 45% of the NMP-1 containment isolation valves have administrative deficiencies. Attachment 5 to the Petition listed eighteen notes in which the Petitioner identified specific deSciencies associated with the containment isolation valves. The asserted deficiencies included: (1) failure to list certain containment isolation valves in die TS or UFS AP, f I tables that list the containment isolation valves; (2) failure to test the containment isolation valves in accordance with the requirements of 10 CF.R. Part 50, Appendix J; (3) failure to test the containment isolation valves in accordance with the requirements of the NMP-1 IST program; (4) inconsistencies in valve stroke time requirements between the TS tables and the UFSAR; (5) inconsistencies in valve actuation signals as specified in the TS tables, the UFSAR tables, and on plant drawings. 1 Primary reactor containments are required to meet the containment leakage test requirements given in Appendix J of Part 50. The purpose of containment leakage tests performed in accordance with the requirements of Appendix J are to ensure that (1) leakage through the primary reactor containment and systems and components penetrating primary containment do not exceed allowable leakage rate values as specified in the plant's technical specifications or associated bases; and (2) periodic surveillance of reactor containment penetrations and isolation valves is performed so that proper maintenance and repairs are made during l de service life of the cont 2inment, and systems and componerrs perenating primary containment. De maximum allowable leakage rate (La) for the NMP-1 pnmary containment is 1.5 weight percent of the contained air per 24 hours at a test pressure of 35 psig. Section III.C.3 of Appendix J further limits the combined leakage for all penetrations and valves subject to Types B and C Tests (as defined in sections 11.0 and II.H of Appendix J) to less than 0.6012. Type C Tests are intended to measure containment isolation valve leakage rates. 389 t h t k 4

Containment isolation valves are provided on lines penetrating the drywell i and pressure suppression chamber to ensure integrity of the containment when required during emergency and postaccident periods. Containment isolation an accident, are automatically controlled by the reactor protection system. valves, which must be closed to ensure containment integrity immediately after The NRC Staff has reviewed the deficiencies identified in Attaciunent 5 to i the Petition. Each of the notes listed in Attachusent 5 to the Petition and I the NRC Staff's contsponding specific findings are discussed in Attachment 1 to this Director's Decision (not published). Sevemi of the Petitionefs notes included comments regarding compliance with GDC 55, " Reactor coolant pressure boundary penetrating containment"; GDC 56, " Primary containment r ? isolation"; and GDC 57, " Closed system isolation valves," of Appendix A to 10 C.F.R. Ibn 50. These comments are not individually addressed in the NRC i Staff findings since, as previously noted, the NRC Staff has concluded that the - j GDCs of Appendix A to 10 C.F.R. Part 50 are not applicable to NMP-1. Re NMP-1 containment isolation valves are listed in two tables in the NMP-I operating license TS and in three tables in the NMP-1 UFSAR. NMP-1 TS nble 3.2.7 and NMP-1 UFSAR Table VI-3a listing containment isolation valves are i titled " Reactor Coolant System Isolation Valves." NMP-1 TS Table 3.3.4 listing l containment isolation valves is titled Pnmary Containment Isolation Valves 1 Lines Entering Free Space of the Containment." NMP-1 UFSAR Able VI. 3b listing containment isolation valves is titled " Primary Containment Isolation and Blocking Valves Lines Entering Free Space of the Containment." NMP- { l UFSAR hble VI-3c listing containment isolation valves is titled " Primary i' Containment Isolation and Blocking Valves Lines with a Closed Loop Inside Containment Vessels." He NRC Staff had previously identified, through its inspection p40gmm. j administrative deficiencies in the TS and UFSAR listings of the containment isolation valves similar to those identified in the Petition. An evaluation of NMP-1 compliance with the requirements of Appendix J to Pan 50 was sent to NMPC in a letter and attached safety evaluation of May 6,1988. De NRC Staff letter and the attached safety evaluation stated that several changes were required to the NMP-1 TS and requested that NMPC submit a license amendment to revise the NMP-1 TS. In a lener of November 20, 1990, NMPC submitted a proposed license amendment to update the containment isolation valve tables and to bring the TS l into conformance with the requirements of Part 50, Appendix J, and the NRC Staff's safety evaluation of May 6,1988. NRC Staff and NMPC representatives discussed the contents of the November 20,1990 submittal in a meeting held on March 5,1991. Following this meeting, NMPC representatives requested i that the NRC Staff suspend review of the November 20,1990 submittal, since 390 f l t

NMPC would be revising and resubmining the proposed TS based on comments from the March 5,1991 meeting. In a letter of February 7,1992, NMPC submitted a proposed license amendment that superseded the November 20,1990 submittal and incorporated the comments from the March 5,1991 meeting between NMPC and NRC StafT. De NRC Staff reviewed the February 7,1992 submittal and issued a request for additional information (RAI) to NMPC on November 30.1992. NMPC responded to this RAI in a letter of January 29,1993. De NRC Staff conducted an onsite inspection of the containment isolation valve issue during the period February 1-5, 1993. De purpose of the onsite inspection was to obtain more informadon about the containment isolation valve issue. The findings of the onsite inspection are summarized below. The detailed results of that inspection are reported in combined Inspection Report No. 50-220/93-01 and 50-410/93-01, dated March 23,1993. Completeness of15 and UFSAR Tables of Containment isolation Valves During the February 1-5,1993 onsite inspection, the NRC Staffindependently developed a list of containment isolation valves using plant drawings. In order to compare this list with TS 'Pables 3.2.7 and 3.3.4 of the February 7,1992 license amendment request, the NRC Staff needed to understand the criteria used by NMPC in the development of the tables. NMPC stated that the TS tables were developed to list any containment isolation valves that received an automatic isolation signal from the reactor protection system (RPS). On the basis of a comparison using this criterion, the NRC Staff concluded that the two lists were consistent, with two exceptions. Specifically, the proposed TS tables did not include valves 63-04 and 63-05 (postaccident sampling system return isolation valves) identified on Drawing F-45089-C, Sheet 8. Revision 3, as containment isolation valves. Follow.ag discussions with the NRC Staff, NMPC changed tih criterion for listing valves as containment isolation valves in the TS tables. ne revised criterion included only those isolation valves closest to the containment. On the basis of this change, the following revisions were made in a February 18,1993 supplement to the February 7,1992 submittal: Valves 63-04 and 63-05 were not included in the TS table because they do not serve as containment isolation valves. The NRC Staff verified that, while these valves receive automatic isolation signals from the reactor protection system, they are located in a branch line outside of containment isolation valves 63.1-01 and 63.1-02 that also trceive automatic isolation signals from the RPS. Valves 63.1-01 and 63.1-02 are included in TS Table 3.3.4. 391

Valves 05-02 and 05-03R (cmergency cooling high point vent to main steam); 39-11R, 39-12R, 39-13R, and 39-14R (emergency cooling steam line drain to main steam); and 05-0IR,05-04R,05-11, and 05-12 (emergency cooling high point vent line), were deleted from TS bble 3.2.7. The NRC Staff verified that containment isolation valves 39-03,39-04, 39-05, 39-06, 39-07R, 39-08R, 39-09R. and 39-10R are located between the subject valves and the reactor coolant system and are included in TS Table 3.2.7. Valves 80-114 and 80-115 (containment spray discharge to waste disposal system) were deleted from TS Table 3.3.4. De NRC Staff verified that isolation valve 80-118 provides the containment isolation function for the subject genetration. NMPC updated TS Table 33.4 to include valve 80-118 in a February 18, 1993 supplement to the February 7,1992 submittal. On the basis of this review, the NRC Staff agreed that the proposed change was appropriate, since the revision states that the valves located closest to the containment are to be considered the containment isolation valves rather than valves in branch lines that are outboard of valves closer to the containment. His revised criterion serves to minimize extensions of the containment and is, thereby, consistent with the intent of GDC 55, 56, and 57, even though the i GDCs are not applicable to NMP-1, ne NRC Staff also determined that the proposed TS tables did not include six normally closed manual isolation valves. NMPC stated that these valves i had not originally been included because they were normally closed, manually operated valves that do not receive an automatic isolation signal frcxn the RPS. Ilowever, NMPC committed to include four of these manual valves (72-479,72-3 480,114-114, and 114-116) in TS Table 3.3.4 so that all containment isolation - valves will be listed in the TS tables. NMPC included these valves in TS Table 33.4 in the February 18,1993 supplement to the February'7,1992 submittal. The other two valves (110-165 and 110-166) were not included in the TS tables since this line has been capped and the penetration will be tested as part of Type B penetration testing. Therefore, these two valves are no longer classified as containment isolation valves. In addition, the NRC Staff independently reviewed the technical data in the TS tables. With the exception of three items described below, all entries were independently verified to be correct. (1) On proposed TS page 148, the bracket indicating that the listed ini-tiating signal was indicated as being applicable to all four penetra-tions (drywell supply, suppression chamber supply, drywell return, and suppression chamber return) of the 11/0 #12 sampling system 2 was incorrectly drawn. De bracket erroneously indicated that the i'l-t 392 r

1 l tiating signal was applicable to the self-actuating check valves when, ) in fact, the initiating signal was applicable only to the de solenoid valves in the drywell supply and suppression chamber supply lines.~ j (2) On proposed TS page 148, note (1) was incorrectly applied to four places on the #11 H/0 sampling entries (drywell supply, suppmssion 2 chamber supply, drywell return, and suppression chamber return). Note (1) states: "Dese valves do not have to be vented during the-Type 'A' test. However, Type C leakage from these valves is added. l to the Type A test results." Since these lines are required to be vented during Type A tests, this note should not apply to these valves. (3) On proposed TS page 148a, note (1) was also incorrectly applied to. + the containment atmosphere monitoring supply line entry since this line is required to be vented during Type A tests. l These administrative deficiencies were discussed with NMPC and were corrected in the February 18,1993 supplement to the February 7,1992 submittal. ] De NRC Staff concluded that all other technical data entries in the TS tables j were correct. De NRC Staff verified consistency between the pertinent elementary RPS l wiring drawings and the valve isolation actuation signals listed in the February 7,1992 license amendment request. The NRC Staff reviewed a sample of re-l cent test data to determine if these valves responded properly to dreir actuation signals. Specifically, test results from the most recent performance of Proce-dure NI-ST-R2,'" Loss of Coolant Accident and Emergency Diesel Generator Simulated Automatic Initiation Test" (July 9-11, 1992), were reviewed. This test inserted low-low reactor water level and high drywell pressure signals (the most common actuation signals for containment isolation valves) and verified -. that the specified isolation valves closed. Review of the test results revealed that all valves listed in the pmcedum responded properly. The NRC Staff also verified that NMPC had similar test procedures in place j for all containment isolation valves and that these procedures were being used .[ to verify proper isolation valve response to other actuation signals. Containment Leakage Rate Testing l Appendix J of Part 50 establishes the NRC requirements for containment .} leakage testing. Appendix J requires performance of three types of containment leakage tests (Type A Tests Type B Tests, and Type C Tests)..Rese three types of tests are explained in sections ll.F. II.G, and II.H of Appendix J. Type A Tests are tests intended to measure the primary reactor containment overall integrated leakage rate (1) after the containment has been completed and is ready for operation and (2) at periodic intervals dereafter. i 393 i e u e i

1 i . Type B Tests are tests intended to detect local leaks and to measure leakage across each pressure-containing or leakage-limiting boundary for the following _ primary reactor containment penetrations: (1) containment penetradons whose design incorporates resilient seals, gaskets, or sealant compounds, piping penetrations fitted with expan-l sion bellows, and electrical penetrations fitted with flexible metal seal 6 ll assemblies; (2) airlock door seals, including door operating mechanism penetrations .[ which are part of the containment pressure boundary; (3) doors with resilient seals or gaskets, except for scal-welded doors; and (4) components other than those listed in 1,2, or 3 (above) that must meet the acceptance criteria in section III.B3 of Appendix J (combined 1 leakage rate for all penetrations and valves subject to Type B and C j Tests shall be less than 0.60 La). Type C Tests are tests intended to measure containment isolation valve i leakage rates. l i Leakage Rate Testing of Containment isolation Vales r De NRC Staff reviewed the most recent local leakage rate test (LLRT) results associated with Procedure N1-TSP-201-550. " Local Leak Rate Tbst - Summary (Type B and C Tbsts). His procedure is used to track the combined leakage rate for all penetrations subject to Type B and C Tests following a Type A Test to verify that the measured combined leakage rate is less than the Appendix J allowable leakage rate of 0.60 la and that the leakage rate limits of TS 433.f(1)(b)(i) and (ii) and 433.f(1)(c) are not exceeded. The NRC Staff } determined that the leakage rate totals were consistent with the requirements of l Appendix J and the TS as of January 29, 1993. De NRC Staff verified that the leakage rates from all primary containment isolation valves requiring Type C testing were included. Independent calculations of the total Type C leakage i i rates, based on the test data in the procedure, confirmed the accuracy of the - value determined by NMPC. =l Step 9.8 of the procedure indicated that the leakage rate limit of TS 433J(1)(b)(i) applies to the sum of the leakage rates from testable penetra-tions and the isolation valves listed in the 13 tables. Six normally closed man-1 ual isolation valves (72-479,72-480,114-114,114-116,110-165, and 110-166) were not included in the TS tables in the February 7,1992 license amendment request. However, leakage rate values for these valves were properly included in the calculation for combined Type C leakage rates. This inconsistency was corrected by adding four of these manual isolation valves to TS Table 33.4 in the February 18,1993 supplement to the February 7,1992 submittal Re other j 394 r l

two valves were not included in tic TS tables and will be deleted from Type C testing since they are no longer classified as containment isolation valves; this penetration has been capped and will be tested as part of Type B testing. He NRC Staff reviewed Drawing F-45089-C, Sheets 8 through 10, and verified that test procedures have been identified for all of the containment isolation valves requiring Type C testing per Appendix 3 of Part 50. Leakage Rate Testing of Water-Scaled Containment holation Yalves X Section III.C.3 of Appendix J to Part 50 states that Icakage from containment isolation valves that are scaled with fluid (water, for NMP-1) from a seal system may be excluded when determining the combined leakage rate for all [. penetrations and valves subject to Types B and C Tests, provided that Such valves have been demonstrated to have fluid leakage ntes that do not exceed those specified in the TS or associated bases, and The installed isolation vdve seal-water system fluid inventory is sufficient to ensure the scaling function for at least 30 days at a pressure of 1.10 Pa. l The February 7,1992 license amendment request, which was supplemented j l by the February 18,1993 submittal and anproved by License Amendment No. .i l 140 issued on April 12,1993, specifies in the TS that the maximum allowable l ws:ter leakage mte from water-scaled valves shall be limited to 0.5 gpm per nominal inch of valve diameter up to a maximum of 5 gpm. Dese water leakage rate limits are consistent with the requirements of paragraph 4.2.2.3(e) of the ASME Operations and Maintenance Standards Part 10 (OM-10) of the 1989 l Edition of section XI of the ASME Code, which was incorporated by reference in paragraph (b) of 10 C.F.R. 6 50.55a, effective September 8,1992 (57 Fed. j Reg. 34,666). The NRC Staff reviewed the most recent leakage rate test results of valves designated in the TS as being subject to water-scal testing and determined t' tat ' ' i all such valves met their applicabic leak test requirements. The NRC Staff concluded that, based on the provisions of section 111.C.3 of Appendix J, the ~ leakage rates from these water-scaled valves may be properly excluded when determining the combined leakage rate for all penetrations and valves subject to. { i Types B and C Tests. ne NRC Staff reviewed note (6) of TS Table 3.3.4 in the February 7,1992 i J proposed license amendment. Note (6) states that the following valves have l-a water-scal capability and that no Appendix J or IST leakage rate testing is I required: Valves 63.1-01,63.1-02,05-05, and 05-07 are properly excluded from I Appendix J and IST leakage rate testing since these valves have no atmospheric leak path. 395

e Valves 80-15, 80-16, 80-17, 80-18, 80-19, 80-35, 80-36, 80-37, 80-38,80-39,80-65,80-66,80-67, and 80-68 have adequate water seals that did not require water leak rate tests according to the NRC Staff's safety evaluation of May 6,1988. Derefore, the NRC StaiT concluded that these water-scaled valves are [ properly excluded from Appendix J and IST leakage rate testing. l Inservice Testing of Containment isolation Valves ne NRC Staff reviewed Revision 3 of the Second 10-Year Inservice Testing Program Plan for NMP-1 and verified that the plan included the independently developed list of containment isolation valves and appropriate exercising and stroke-time ~ test requirements (for power-operated valves). he NRC Staff re-viewed the following two surveillance tests which implement the IST require-ments: N1-ST-04, " Reactor Coolant System isolation Valves Operability Test," performed November 16-18, 1992; N1-ST-05," Primary Containment Isolation Valves Operability Test," performed on November 7,1992. This review revealed that all the isolation valves listed in the procedures had. been exercised and, if required, stroke-time tested. 'Ihe procedures specified stroke time limits and the measured results were consistent with the IST program and, if specified, with the TS limits. On the basis of these reviews of IST l data, the NRC Staff concluded that all containment isolation va'. /es listed in the procedures have been properly exercised and stroke-time tested as part of the Licensee's ongoing IST program. De NRC Staff also verified that test procedures are in place for all required IST testing of containment isolation valves. UFSAR Update l in its January 29,1993 letter, NMPC committed to update the UFSAR and correct deficiencies therein by June 30,1993. De NRC Staff will verify, as part of its routine reviews of UFSAR updates, that UFSAR *Ihbles VI-3a, VI-3b, and VI-3c have been corrected. r t Containmentisolation Valves-Conclusion In summary, the NRC Staff concluded that (1) the containment isolation l valve deficiencies identified by the Petitioner were administrative in nature; (2) l notwithstanding the administrative deficiencies, the operability of the contain-l 396 [ f 5 9

? h i ment isolation valves was being maintained in accordance with the requirements of the 'I3 and IST program and the valves were being properly tested in ac-cordance with all applicable regulatory requirements; (3) the leakage rates of water-scaled valves were properly excluded when determining the combined leakage rate for all penetrations and valves subject to Types B and C Tests; (4) the Licensee has committed to update the UFSAR by June 30,1993 to cor-rect the identified deficiencies; and (5) License Amendment No.14(. W sed on April 12,1993, to the Nine Mile Point Nuclear Station Unit 1 Facility Operating License DPR-63 corrected the administrative deficiencies related to the contain-ment isolation valve listings in the TS. Therefore, to the extent that t'ne Petitioner f sought correction of the TS tables to correcdy list the NMP-1 containment iso- ~ lation valves, their initiating signals, and their stroke times, this relief has been granted. As stated above NMPC has committed to correct the UFSAR tables by June 30,1993. Action to require earlier change to the UFSAR tables is not i needed in light of the NRC Staff's confirmation of valve operability during an onsite inspection conducted February 1-5, 1993. Petitioner's request for other actions based on containment isolation valve deficiencies is denied. 3. NMPC, NMPC's Quality Assurance Group, and the NRC Han ~l Reviewed These Safety Concerns and, Contrary to Any Practical L Justification, Haw Remained Silent The Petitioner was employed at NMP-1 as a contractor from November 13, f 1989, to Januar~ 18,1990. During that employment, the Petitioner expressed l several concerns to NMPC regarding the design and operation of the NMP-1 feedwater system in its HPCI mode and what he believed were various inconsistencies in the listings of the containment isolation valves in the TS, in the UFSAR, and on the plant drawings. The NRC Staff has reviewed NMPC records regarding the processing of the Petitioner's concerns by the NMPC Regulatory Compliance Group and by the NMPC Quality First Program. A summary of NMPC's consideration of the Petitioner's concerns follows. l Review of Concerns by NMPC Regulatory Compliance Group

• Ihe Petitioner initially submitted his concerns regarding the design and opemtion of the feedwater system in its HPCI mode and what he believed were various inconsistencies in the listings of the containment isolation valves in the TS, in the UFSAR, and on the plant drawings to the NMPC Regulatory Compliance Group during January 1990. In a letter dated July 31,1990, to NMPC, the Petitioner subsequently also submined these concerns to the NMPC i

397 i .} S i I l i I, s

~. f I Quality First Program (Q1P). ne NRC Staff review of NMPC records disclosed that the concerns the Petitioner submitted to the NMPC Regulatory Compliance Group and to the NMPC Q1P covered the same topics he submitted to the NRC l [ in his 10 C.F.R. 52.206 Petition dated October 27,1992, and evaluated herein by the NRC Staff. NMPC evaluated the Petitioner's concerns regarding the feedwater system operating in the IU2Cl mode during February 1990 and determined that the l NMP-1 accident analyses do not rely on the HPCI system for mitigation of any accidents. NMPC's conclusion regarding operation of the feedwater system in 1 d a co stent i e cl 20 t b CS th Director's Decision. After reviewing NMPC records, the NRC Staff concluded that NMPC had properly reviewed the Petitioner's concerns regarding the HPCI system. De NRC Staff reviewed NMPC records which showed that, in January l 1990, the Petitioner communicated to the NMPC Regulatory Compliance Group 'l what he believed wem various inconsistencies in the listings of containment j isolation valves in the TS, in the UFSAR, and on the plant drawings. The Petitioner also expressed concerns about the performance of IST and leak tests ? according to the requirements of Appendix J. NMPC reviewed the Petitioner's concerns between January and July 1990. NMPC determined that some of the l Petitioner's concerns had been previously reviewed and found acceptable in 7 NRC Staff-approved safety evaluations and that some of his concerns had been resolved by issuance of NRC Staff-approved schedular exemptions. NMPC r also referred the Petitioner's list of concerns to the NMPC Licensing Group to ensure that applicable concerns would be addressed by including them in the license amendment then in preparation with the purpose of resolving deficiencies identified in the NRC Staff safety evaluation of May 6,1988. After l reviewing NMPC records, the NRC Staff concluded that the NMPC Regulatory Compliance Group had processed the Petitioner's concerns in an appropriate manner. Review of Concerns by HMPC Quality First Program De Petition stated that following a perceived period of ina tion by NMPC, the Petitioner notified the NMPC Q1P of his concerns regarding (1) opration of the feedwater system in the HPCI mode and (2) the containment isolation valves. - The Q1P is an NMPC program designed to give its employees a confidential .l forum for reporting potential problems that affect quality or safety on the job. QlP is directed by the NMPC Quality Assurance Department and applies to the a receipt, control, investigation, resolution, feedback to the originator, and repons - a 398 r 1i t l i

to NMPC management of any concerns identified. QlP is not governed by NRC regulatory requirements, except as related to protected activities by employees. Although NMPC employces are encouraged to report potential problems to the NMPC QlP during their employment and upon termination of employment, 1 NMPC representatives stated that NMPC personnel had searched the QlP files and found no record of the Petitioner contacting the QlP prior to receipt of a letter from the Petitioner, dated July 31, 1990. NMPC informed the NRC Staff that QlP records were not considered plant records unless a valid quality concern was deterrr.med to exist. Derefore, it is possible that no records exist because previous contacts may have been made but had been treated as having no basis. The NRC Staff reviewed a copy of the letter NMPC received from the Pedtioner, dated July 31,1990, in which concerns regarding operation of the feedwater system in the IU'Cl mode and the containment isolation valves were outlined. These concerns repeated the ones made previously by the Petitioner f. to the NMPC Regulatory Compliance Group. I De NRC NMP-1 resident inspectors were informed by a Q1P representative on August 6,1990, that the July 31,1990 letter had been received. According to records reviewed by the NRC Staff, NMPC had reviewed the Petitioner's concerns between August and November 1990. These records showed that NMPC closed out these concerns on November 28,1990, after contacting the Petitioner and obtaining his agreement for closure. NMPC again determined j that the concerns regarding operation of the feedwater system in the HPCI .j mode had no basis since the NMP-1 safety analyses do not rely on operation of the feedwater system in the HPCI mode to satisfy the emergency core cooling requirements of section 50.46. The NMPC Licensing Group received { the concerns regarding the containment isolation valves for consideration in the { proposed license amendment development. The NRC Staff concluded that the l NMPC QlP organization processed the Petitioner's concerns appropriately. ( As noted in the discussion of_ operation of the feedwater system in the HPCI mode,information regarding the design features of the NMP-1 feedwater system, including operation in the HPCI mode, has been readily available in l the public records, and the NRC Staff was well aware of this information over the life of the NMP-1 plant. The NRC Staff concerns about NMP-1 compliance with the requirements of Part 50, Appendix J, have been a matter of public record since May 6,1988, when the NRC Staff issued its letter with j its attached safety evaluation. As noted above, One NRC Staff concluded that. j NMPC's Regulatory Compliance Group and QlP representatives handled the Petitioner's concerns in an appropriate manner. Therefore, I have concluded that the Petitioner's assertion that NMPC, NMPC's quality assurance group, and the NRC have known of these safety concerns and have remained silent has i 399 1 i I l l ---_____-_________.-_______L-_.-_

f no basis. Accordingly, the Petitioner's request for enforcement action against ~ NMP-1 on this part of the Petition is denied. Although I have denied this portion of the Petition, a copy of the Petition has been referred to the NRC Office of the inspector General for whatever review and action the Inspector General deems appropriate. J l 4 Insurance t De Petition asserts that NMPC is not insured to operate NMP-1.in the manner described in the Petition. In order to operate a commercial nuclear - .j power plant within the United States with " limited liability," an NRC licensee ~ f must have and maintain financial protection (e.g., liability insurance). He Price-Anderson Act requires NMPC to provide $200 million in liability insurance for j public liability claims that might arise from a nuclear accident at the NMP-1 site. In addition, NMPC (along with all other commercial nuclear power plant licensees) must participate in an industry self-insurance plan which subjects it l to a potential liability of $63 million for each commercial nuclear power plant that it operates, for public liability claims that might arise from a single nuclear accident at NMP-1 or any other commerc!al nuclear power plant licensed by the NRC, his liability insurance cannot be purchased from the nuclear liability insurance pools unless the pools are satisfied that a licensee is operating its [ commercial nuclear power plant in accordance with NRC regulations. Contrary. + to the assertions in the Petition, NMPC has obtained and is maintaining the j appropriate amount of liability insurance. t Conclusion 1 ne Petitioner requested that the NRC issue an immediately effective order directing NMPC to cease power operation of NMP-1 and to place the reactor in a cold-shutdown condition pending full compliance with NRC regulations. %c { Petition also asked the Commission to hold a public hearing before authorizing j resumption of plant operation. On April'12,1993, the NRC Staff issued License Amendment No.140 i to the NMP-1 Facihty Operating License DRP-63. This license amendment corrects the NMP-1 TS tables that list the containment isolation valves,' their initiating signals, and their stroke times. To the extent the Petitioner sought j such corrections, this relief has been granted. Further, NMPC has committed j to update the UFSAR, by June 30, 1993, to list the containment isolation valves correctly. De NRC Staff will verify this commitment as part of its i routine reviews of UFSAR updates. De NRC Staff views these changes as l administrative corrections since the NRC Staff has concluded that all relevant - ' ia 400 1 W h 1 - i 4 1

valves were appropriately tested. With regard to the other requests made by the Petitioner, an immediate shutdown of NMP-1 and the institution of a public hearing before authorizing resumption of plant operation, as requested by the Petitioner, is appropriate only where substantial health and safety issues have been raised. Sec Consolidated Edison Co. ofNew York (Indian Point, Units 1,2, and 3), CL1-75-8, 2 NRC 173,175 (1975), and Washington Public Power Supply System (WPPSS Nuclear Project No. 2), DD-84-7,19 NRC 899,923 (1984). Ibr the reasons discussed above, I find no basis for taking such actions. Rather, on the basis of the review efforts by the NRC Staff, I conclude that no substantial health and safety issues have been raised by the Petitioner. Accordingly, the Petitioner's remaining requests for action pursuant to section 2.206 are denied. A copy of this Decision will be placed in the Commission's Public Document Room, Gelman Building,2120 L Street, NW, Washington, DC 20555, and at the Local Public Document Roorn, Reference and Documents Depanment, Penfield Library, State University of New York, Oswego, NY 13126. A copy of this Decision will also be filed with the Secretary for the l_ Commission's review as stated in 10 C.F.R. 62.206(c) of the Commission's j regulations. FOR THE NUCLEAR REGULATORY COMT 'ISSION Thomas E. Murley, Director Office of Nuclear Reactor Regulation Dated at Rockville, Maryland, this 9th day of May 1993. f l [The attachments have been omitted from this publication but can be found in the NRC Public Document Room,2120 L Street, NW, Washington, DC.] t 1 I l lO 401 .0 j l l

Cite as 37 NRC 402 (1993) DD-93-11 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION Dr. Thomas E. Murley, Director in the Matter of TEXAS UTILITIES ELECTRIC Docket Nos. 50-445 COMPANY, et at 50-446 (Comancho Peak Steam Electric Station, Units 1 and 2) CAROLINA POWER AND LIGHT Docket Nos. 50-324 COMPANY 50-325 (Brunswick Steam Electric Plant, Units 1 and 2) CAROLINA POWER AND LIGHT Docket No. 50-400 ] COMPANY, et al. (Shearon Harris Nuclear Power Plant) DETROIT EDISON COMPANY, et al Docket No. 50-341 (Enrico Ferm! Atomic Power Plant,. Unit 2) 4 WASHINGTON PUBLIC POWER Docket No. 50-397 SUPPLY SYSTEM ' I (WPPSS Nuclear Project No. 2) GULF STATES UTILITIES Docket No. 50-458 COMPANY l (River Bend Station, Unit 1) l 1 - l 1 402-I m__:.____.______________1________________1 _______________________2__

r.. ARKANSAS POWER AND LIGHT Docket No. 50-368 COMPANY (Arkansas Nuclear Two) DUQUESNE LIGHT COMPANY, et al. Docket Nos. 50 334 (Beaver Valley Power Station, 50-412 Units 1 and 2) COMMONWEALTH EDISON COMPANY Docket Nos. 50-456 (Braidwood Nuclear Power Station, 50-457 Units 1 and 2) TENNESSEE VAL EY AUTHORITY Docket Nos. 50-259 (Browns Ferry Nuclear Plant, 50-260 Units 1,2, and 3) 50-296 COMMONWEALTH EDISON COMPANY Docket Nos. 50-454 (Byron Nuclear Power Station, 50-455 Units 1 and 2) UNION ELECTR!C COMPANY Docket No. 50-483 (Callaway Plant, Unit 1) ILLINOIS POWER COMPANY, et af. Docket No. 50-461 (Clinton Power Station, Unit 1) INDIANA AND MICHlGAN Docket Nos. 50-315 POWER COMPANY 50-316 (Donald C. Cook Nuclear Plant, Units 1 and 2) NEBRASKA PUBLIC POWER DISTRICT Docket No. 50-298 (Cooper Station, Unit 1) FLORIDA POWER CORPORATION Docket No. 50-302 . (Crystal River Unit No. 3 Nuclear Generating Plant) TOLEDO EDISON COMPANY Docket No. 50-346 (Davis-Besse Nuclear Power Station, Unit 1) l 403 l I l lL

P PACIFIC GAS AND ELECTRIC Docket Hos. 50-275 COMPANY 50-323 (Diablo Canyon Nuclear Pov :t Plant, 7 Units 1 and 2) IOWA ELECTRIC LIGHT AND Docket No. 50-331 POWER COMPANY (Duane Arnold Energy Center) ENTERGY OPERATIONS,INC. Docket No. 50-416 (Grand Gulf Nuclear Station, Unit 1) CONNECTICUT YANKEE ATOMIC Docket No. 50-213 POWER COMPANY l (Haddam Neck Plant) GEORGIA POWER COMPANY Docket Nos. 50-321 (Edwin L Hatch Nuclear Power 50-366 Plant, Units 1 and 2) CONSOLIDATED EDISON COMPANY OF Docket No. 50-247 NEW YORK (Indian Point, Unit 2) COMMONWEALTH EDISON COMPANY Docket Hos. 50-373 (La Salle County Station, Units 1 50-374 and 2) PHILADELPHIA ELECTRIC COMPANY Docket Nos. 50-352 (Limerick Generating Station, Units 1 50-353 I and 2) MAINE YANKEE ATOMIC Docket No. 50-309 POWER COMPANY (Maine Yankee Atomic Power Station) DUKE POWER COMPANY, et al Docket Nos. 50-369 (William B. McGuire Nuclear 50-370 l Station, Units 1 and 2) f 404

NORTHEAST NUCLEAR ENERGY Docket Nos. 50-245 COMPANY 50-336 (Millstone, Units 1,2, and 3) 50-423 + NORTHERN STATES POWER Docket No. 50-263 COMPANY (Monticello Nuclear Generating Plant, Unit 15 s NIAGARA MOHAWK POWER Docket Nos. 50-220 CORPORATION 50-410 (Nine Mlle Point Nuclear Station, Units 1 and 2) VIRGIN!A ELECTRIC AND POWER Docket Nos. 50-338 COMPANY 50-339 (North Anna Pcuer Station, Units 1 and 2) GENERAL PUBUC UTILITIES Docket No. 50-219 NUCLEAR CORPORATION (Oyster Creek Nuclear Generating Station) CONSUMERS POWER COMPANY Docket No. 50-255 (Palisades Nuclear Power Facility) ARIZONA PUBLIC SERVICE Docket Nos. 50-528 COMPANY, et al 50-529 (Palo Verde Nuclear Generating 50-530 Station, Units 1,2, and 3) PHILADEl.PHIA ELECTRIC COMPANY, et al Docket Nos. 50-277 L (Peach Bottom Atomic Power Station, 50-278 Units 2 and 3) l CLEVELAND ELECTRIC ILLUMINATING Docket No. 50-440 COMPANY, et al. (Perry Nuclear Power Plant, Unit 1) 405 r l i T

.._...-e. t 'I 'I r NORTHERN STATES POWER Docket Nos. 50-282 COMPANY 50-306 - (Prairie Island Nuclear Generating l Plant, Units 1 and 2) .[ SOUTHERN CAUFORNIA EDISON Docket Nos. 50-361 COMPANY, et al. 50-362 (San Onofre Nuclear Generating [ Station, Units 2 and 3) i TENNESSEE VALLEY AUTHORITY Docket Nos. 50-327 l . (Sequoyah Nuclear Plant Units 1 50-328 { and 2) HOUSTON UGHTING AND - Docket Nos. 50-498 f POWER ' COMPANY, et al 50-499 . (South Texas Project, Units 1 and 2) FLORIDA POWER AND UGHT l COMPANY Docket Nos. 50-335 (St. Lucie Nuclear Power '50-389 Plant, Units 1 and 2) SOUTH CAROUNA ELECTRIC AND Docket No. 50-395 .j GAS COMPANY, et al 'l (Virgil C. Summer Nuclear Station, l Unit 1) l VIRGINTA ELECTRIC AND POWER Docket Nos. 50-280 .I COMPANY 50-281 f (Surry Nuclear Power Station, + Units 1 and 2) l l PENNSYLVANIA POWER AND Docket Nos. 50-387 UGHT COMPANY 50-388 (Susquehanna Steam Electric Station, Units 1 and 2) l d I 406 !l i 'I f 1 l 1 fl ] w

GENERAL PUBUC UT1UTIES Docket No. 50-289 NUCLEAR CORPORATION (Three Mile island Nuclear t Station, Unit 1) PORTLAND GENERAL ELECTRIC Docket No. 50-344 COMPANY, et af. l (Trojan Nuclear Plant, Unit 1) t FLORIDA POWER AND LIGHT Docket Nos. 50-250 COMPANY 50-251 -i (Turkey Point Nuc ear Generating Plant, Units 3 and 4) VERMONT YANKEE NUCLEAR Docket No. 50-271 POWER CORPORADON (Vermont Yankee Nuclear Power Station) GEORGIA POWER COMPANY, et af. Docket Hos. 50-424 (Vogtle Electric Generating Station, 50-425 Units 1 and 2) i ENTERGY OPERATIONS,INC. Docket No. 50-382 (Waterford Steam Electric Station, j Unit 3) TENNESSEE VALLEY AUTHORITY Docket Nos. 50-390 (Watts Bar, Units 1 and 2) 50-391 WOLF CREEK NUCLEAR OPERATING Docket No. 50-482 CORPORATION (Wolf Creek Generating Station, Unit 1) COMMONWEALTH EDISON COMPANY Docket Nos. 50-295 (Zion Station, Units 1 and 2) 50-304 I May 23,1993 On February 1,1993, the Director of the Office of Nuclear Reactor Regulation 'i issued a Parual Director's Decision Under 10 C.F.R. 5 2.206 (DD-93-3, 37 i 407 l a 1 I i

I NRC 113 (1993)) responding to a petition dated July 21,1992, submitted by the Nuclear Information and Resource Service and others (Petitioners) which requested that the U.S. Nuc! car Regulatory Commission (NRC) take enforcement actions in Ught of nermo-Lag fire barrier test failures. Fire barriers are generally regtured at operating commercial nuclear power plants by the NRC's regulations or facility license conditions. Petitioners submitted additional filings on August 12, 1992, September 3,1992, and December 15, 1992. All issues raised by Petitioners except those raiscd in the December 15,1992 supplement were addressed in DD-93-3, ne remaining issues were to be addressed in a Final Director's Decision. New concerns regarding Hermo-Lag material raised in the December 15, 1992 filing can be summanzed as the existence and creation of voids in Thermo-Lag material, staples in the material, and possible erroneous information given to utilities concerning the weight of the material. Petitioners also restated assertions made in their earlier filings regarding alleged inconsistencies in tests conducted j to determine the toxicity of nermo-lag material when it is burned. i Petitioners requested that the NRC immediately suspend the operating li-censes and construction permits of nuclear facilities that use Hermo-Lag or, alternatively, that the NRC order each licensee to remove and replace its Hermo-Lag during its next refueling outage or before beginning operation. l On May 23,1993, the Director issued a Final Director's Decision Under l 10 C.F.R. 52.206 which addressed Petitioners

• remaining issues and denied the

~ relief sought by the Petitioners. De Director concluded that no substantial health and safety issues had been raised by the Petitioners. FINAL DIRECTOR'S DECISION UNDER r i 10 C.F.R. s 2.206 I. INTRODUCTION By a petition dated July 21, 1992, the Nuclear Information and Resource 4 Service (NIRS), Alliance for Affordable Energy, and Citizens Organized to Protect Our Parish (the Petitioners), requested that the U.S. Nuclear Regulatory Commission (NRC) take enforcement action regarding the Gulf States Utilities' (sometimes referred to as GSU) River Bend Station, demanding that its operating license be suspended until GSU can demonstrate, thmugh independent testing, that it meets the NRC's fire protection regulations (Appendix R to Part 50 of - Title 10 of the Code of Federal Regulations (10 C.F.R. Part 50)). In addition, the Petitioners demanded that the NRC Staff immediately issue Generic Letter (GL) 92-XX, the draft of which was circulated for public comment on February 11, 1992, and close any nuclear power plant for which the Licensee cannot 408 - l j

b f prove, through independent testing, that it mecis the NRC's fire protection regulations until it does meet them. By an addendum to the petition dated August 12,1992, the Petitioners requested immediate action related to the Comanche Peak, Shearon Harris, Fermi-2, Ginna, WNP-2, and Robinson nuclear facilities. Joining in filing the addendum were a number of other organizations: Citizens for Fair Utility Regulation, Don't Waste New York, Citizens Against Radioactive Dumping, Coalition for Alternatives to Shearon liarris, Conservation Council of North Carolina, Safe Energy Coalition of Michigan, Steve Langdon, Essex County Citizens Against Fermi-2, Natural Guard, and Northwest Environmental Advocates.8 ne petition and addendum were submitted under the provisions of 10 C.F.R. 5 2.206 of the NRC's regulations. Notice of receipt of these filings i was published in the FederalRegister on August 26,1992 (57 Fed. Reg. 38,702). in their filings the Petidoners alleged a number of deficiencies concerning l Thermo-Lag 330-1 material (Thermo-Lag). manufactured by hermal Science, Inc. (TSI), including failure of Hermo-Lag fire barriers during 1-hour and 3-hour fire endurance tests, deficiencies in procedures for installation, noncon-formance with NRC regulations for quality assurance and qualification tests, the combustibility of the material, ampacity miscalculations, lack of seismic tests, the failure to pass hose stream tests, the high toxicity of substances emined from the burning material, and the declaration by at least one utility, GSU, that the fire inrrier was inoperable at its River Bend Station. He Petitioners also alleged that a fire watch cannot substitute for an effective fire barrier indefinitely and that the NRC Staff had not adequately analyzed the use of fire watches. On the basis of these allegations, the Petitioners requested emergency en-forcement action to immediately suspend the operating licenses for River Bend Station, Comanche Peak Unit 1, Shearon llarris, Fermi-2, Ginna, WNP-2, and Robinson, pending a demonstration that these facilities meet NRC fire protec-l tion requirements. De Petitioners also requested that the NRC issue (1) a stop-work order regarding the installation of Thermo-Lag at Comanche Peak Unit 2'. and (2) a generic letter by September 5,1992, that would require licensees to submit information to the NRC demonstating compliance with fire protection requirements. Where facilities could not demonstrate compliance, the Petition-l crs requested immediate suspension of the operating licenses for such facilities until such time as compliance with NRC fire protection requirements could be i shown. He Petitioners' submittals were referred to the Office of Nuclear Re-actor Regulation for preparation of a response. In a letter of August 19, 1992, the Director, Office of Nuclear Reactor Regulation, denied the Petitioners' request for emergency relief, The HEC Staff concluded that the immediate suspension of the operating licenses for River .t I Rererence to Ptniamnars heremabr shall also include these organiradans. 409 i i ~$ - l t ? E

t l -- 1 p L 2 Bend Station, Comanche Peak Unit 1, Shearon liarris, Fermi-2, WNP-2, Ginna, and Robinson was not warranted. De NRC Staff also determined that a stop-work crder or the suspension of the construction permit for Comanche Peak Unit 2 was not warranted and concluded that the generic letter would be issued l in accordance with the NRC Staff's action plan regarding Hermo-Lag issues - and that it was not necessary to accelerate the issuance of the generic letter. On September 3,1992, the Petitioners filed an " appeal" with the Commission in sesponse to the NRC Staff's denial of August 19,1992, of the request for emergency enforcement action. In the " appeal," k Petitioners removed the Ginna and Robinson plants from their request and added Brunswick Units 1 and

2. He Petitioners again aleged that Thermo-Lag is an inadequate fire barrier, that compensatory measures do not substitute for regulatory compliance, and i

that fire watches are inadequate substitutes for fire barriers. In a letter of November 9,1992, the Secretary of the Commission informed the Petitioners that their " appeal" request had been referred to the hTC Staff i for appropriate consMeration in conjunction with its review of the Petitioners

• carlier filings.

He NRC Staff evaluated the issues raised in the above-referenced Petitioners' [ l submittals and issued a Partial Director's Decision on Febmary 1,1993 (58 Fed. Reg. 7595). De Partial Director's Decision concluded that the Petitioners had not presented information sufficient to constitute a basis to i - issue a stop-work order suspending installation of Thermo-Lag in, or to suspend the construction permit for, Comanche Peak Unit 2; - immediately suspend the operating licenses for Comanche Peak Unit 1, Shearon Harris, Fermi-2, WNP-2, Brunswick Units 1 and 2, and River Bend Station;- - have issued GL 92-XX before September 5,1992. In a letter of December 15, 1992.2 NIRS filed another petition pursuant to section 2.206 raising additional issues regarding Thermo-Lag. This filing alleges that it contains new information regarding deficiencies with Thermo-Lag q material, restates the concerns raised by NIRS and others in earlier submittals, l and requests that the NRC immediately suspend the operating licenses and constmetion permits of nuclear facilities that use Hermo-Lag as a fire barrier material or, alternatively, that the NRC order each licensee to remove and replace its Thermo-Lag during its next refueling outage or before beginning operation. Notice of receipt of the December 15,1992 petition was published in the federal. Register on Febmary 16,1993 (58 Fed. Reg. 8637). 3 De NRC Staff reviewed the December 15,1992 petition for any new issues and information that were not addressed in the Partial Director's Decision of 3 2 h is prawned that NIRs b sdn saing on behalf of all the Peunaners. I 410 -} P ~$ 1 3

{

t t February 1,1993. New concerns regarding Hermo-Lag material that have been I raised and were not previously addressed can be summarized as the existence of and creation of voids, staples in the material, and possible erroneous information given to utilities concerning the weight of the material. In a letter of February 4,1993, I denied the December 15,1992 request for emergency relief, ne NRC Staff concluded that immediately suspending the operating licenses or construction permits of all nuclear power plants that use Rctmo-Lag fire barrier material until the %ctmo-Lag is removed or replaced was not warranted. I also stated that the NRC would treat the December 15, 1992 submittal as a supplement to the earlier filings by the Petitioners and would l address the new concerns in a Final Director's Decision to be issued within a i reasonable time. Upon consideration of the new concerns and information given in the December 15, 1992 petition, I have determined that the Petitioners have not presented sufficient information that would constitute a basis to immediately suspend the operating licenses or construction permits of all nuclear power plants i that use Thermo-Lag fire barrier material, until the hermo-Lag is removed or replaced, or alternatively, to order each licensee to remove and replace its Hermo-Lag during its next refueling outage, or before beginning operation. IL DISCUSSION ne specific issues raised by the Petitioners in the petition of December 15, 1992, are summarized below, together with the NRC Staff's evaluation.' A. Voids and Staples De Petitioners state that they recently learned that hermo-Lag contains l' voids, which the Petitioners characterize as "in layman's terms, areas where there is essentially a front and back to the material but no or little middle." According to the Petitioners, where voids exist, the material could burn through very rapidly, negating the material's effectiveness as a fire barrier. De Petitioners also state that they believe some voids may be created by." bending" Dermo-Lag material around electrical conduits, and that it is common practice to bend .i - the insterial, causing it to crack, and then to staple the material together and cover it with another layer of Dermo-Lag. He Petitioners contend not only that this practice creates voids, but also that the staples may sene as a " heat - sink" causing combustion and " speedy failure" of the material. 3 A general hastancal sunm.ary of concerns and issues relating so Thcrrno-tag is car.tained in the Panial Danctor's Decisuus of Febnary 1.1993, and win not be repeated here. i 411 .i h J. i t N ? er

i 1. Analysis of Concerns The concerns raised by the Petitioners appear to be based at least in part l' on reports of observations made during cutting operations at Comanche Peak Steam Electric Station (CPSES). The existence of delaminations was discov-cred at Comanche Peak Unit 2. A " delamination" refers to the separation of the component layers of a Hermo-Lag panel with a resulting air gap between the component layers. This phenomenon was the subject of communications i between the NRC and TSI' referred to by the Petitioners in their December l 15,1992 petition. Although the Petitioners do not specifically question delam-7 inations, the NRC Staff nonetheless has analyzed the condition and believes that there is reasonable assurance that the existence of a delamination not de-l tected during fabrication or installation would not have a significant effect on j the performance of properly installed Thermo-Lag material. The Staff bases its judgment on the fact that the process through which Thermo-Lag material i performs its function (partial intumescence and char layer formation) and which has been observed by the NRC Staff on many occasions, results in a significant expansion of the material as the char layer forms and any air gap between layers would be essentially eliminated. More importantly, the total amount of material {' specified to be installed is still present. With respect to " voids" as characterized by the Petitioners, i.e., an area of a Thermo-Lag panel "whefe there is essentially a front and back" to the panel _j but "no or liule middle," the NRC Staff has inspected %ermo-Lag materials purchased for its own tests and Hermo-Lag used in other tests and has found that small air pockets entrapped inside Dermo-Lag material are inherent in all prefabricated panels and preshaped conduit sections, not simply those found at -l Comanche Peak. The " voids" observed by the NRC Staff mnged in size from less than %-inch diameter to about the size and shape of a lima bean. They are formed during the course of manufacturing; they do not result from bending ' i the material to form conduit sections? On the basis of the successful fire tests 1 s

• In communuations betwoen the NRC and Tsi, the air gap betwem delsminsied campanent layers of Berne Lag was refened to at one pois as a void." llowever, comrary to the Petauners' ese of the term "vad." an air gap associated with s delamination das not indicate that the cais6ng total smauru of rnaturial is less than wtat was arsended.

3While the process of farming conduit sec6ans at the factory fawn beding hermo-lag panels dos not produce voids d this nanus, bendmg can produce legitudmal erads or fissures along the outer layer of the secnana, i De manufacturcr*s procedures require that crads in cmduit sections be repaired by force filhng the cracks with %ermo tag 330-1 Trowel orade material umil the repair are carnplete. De crack fermanan and repair are pan of the nornal manufaciunns process for preformod Thermo-Lag conduit sations and is an inherent candinen for emduit sec6ana. To the entent the penneners suggest that instaDauon of Thernetag includa

• bending" of material around electncal coruluits, resuhing in crada. this is not a :

'ed installadan pricdce. Rather. preformnd %nrnetag conduit sec6ans are used to erwlase caduha at the sinn. Also, to the enters that the ' Paineers suggest that an overlay d Thermo-tag material snay create a vasd in the nature d an air gap associsied with a delamination, such an air gap. as d6~=W above, is not unacceptable. In addaian, oweday configurations have bem the subject of successful fare acceptance tests. 412 -l l, i l

performed for Comanche Peak 2, where test specimens consisted of hundreds of square feet of prefabricated %crmo-Lag materials that were not known to be any different from other Ecrmo-Lag materials with respect to inherent void characteristics, the NRC Staff believes that these voids will not cause premature failure of properly designed and installed fire barriers.* He assertion that the use of staples may serve as a " heat sink" causing combustion and " speedy failure" of the material is not supported by the Petitioners. It is true that the discretionary use of staples to repair cracks and enhance the structural integrity of conduit sections has been an ongoing practice in the manufacture of 1-hour and 3-hour Hermo-Lag prefabricated items. Furthermore, Texas Utilities Electric (TU) installation procedure QCP-CV-107 " Application of Fire Barrier and Fire Proofing Material," at CPSES, specifies the use of W-inch, Arrow or Bostich T-50 staples to secure stress skin to prefabricated panels. Contrary to the Petitioners' assertion, however, there is no evidence fmm TU tests of Thermo-Lag material containing staples that would suggest that the existence of staples could cause nermo Lag to fail to perform properly, in addition, before being tested at TU, test specimen panel seams were through-stitched with stainless steel wire. De amount of metal I available as a " heat sink" in such stitching is greater than in the staples used in practice; however, test results did not indicate any adverse effects from the stitching, Herefore, on the basis of the information provided above, the NRC Staff f concludes that the Petitioners' assertions that voids will negate Thermo-Lag material's effectiveness as a fire barrier and that staples may serve as a " heat sink" causing combustion and " speedy failure" are without foundation. 2. Compensatory hieasures De information required by Generic Letter 92-08, "nermo-Lag 330-1 Fire Barriers," issued December 17, 1992, will show whether Hermo-Lag fire barriers have been properly qualified and installed, and are representative of tested configurations.' Without such confirmation, Hermo-Lag fire barriers 'It should be noiad that the emisience of voids in rue bamer materials is nat unique to Therm *Iag. O;her matenals used in twild fue barners, such as concrcie blocks, bricks, and celbng tacs may also carusin voids. 7 Generic Leuer 92-0R required heensees using Therma Lag 3341 barriers to simie whesher or not 0) the.. licensee has quahned the Thermo4ag 3304 fue barriers by conduenna fue endurance tests in auerdance with the NRC's - ; ~ and guidance or Econsmg --- '

(2) the fue barrier contaguranans instauod in J

the plant repnses the materials, workmanship, and methods of assembly. dimensions, and conriguranens of ') the quahricanon test assembly cualiguranons; and Q) the Ecermee has evaluated any devianons frorn the tested j cariguranons. These informananal m.-.;. apply to all 1-hour and 34 aur Therma tag 3301 meterials and u bamer systems assembled by any method, sud as by assembling pretormed panels and conduit shaps, as well j as sway, trowel, and bnah m appbcanan assembly, should a leermee answer any si the three items (above) in the neganve, the beensee is to describe all correenve actions needed and include a schedule fw completing such scnans and shall dnacribe all compensatory rnessures taken in accordance with their techrucal specincanons er administranve controls. 413 l idz___________.____1___

l I t are to be considered inoperable, a status that requires licensees to implement compensatory measures until operability of the barrier can be demonstrated. In their responses to NRC Bulletin 92-01, Supplement 1. "Pailure of Thermo-Lag 330-1 Fire Barrier System to Perform Its Specified Fire Endurance Function," all 7 licensees of operating plants named in the petition of December 15,1992, have + affirmed that compensatory measures are in place consistent with their technical specifications or license conditions for an inoperable fire barrier. The NRC Staff has evaluated compensatory measures, such as fire watches, and has found that they continue to adequately protect the public health and safety when barriers are inoperable. Herefore, the NRC Staff has concluded that there is no immediate threat to the public heahh and safety, given that compensatory measures have, in fact, been instituted. Accordingly, the suspension of the operating licenses for those reactors listed in the December 15, 1992 petition is'not warranted, Licensees will be permitted to cease compensatory measures when they have affirmed that fire barriers have been properly qualified and installed and there { is reasonable assurance that such barriers will perform their intended function. l B. Installed Weight of Thermo-Lag De Petitioners state that TS1, the manufacturer of Thermo-Lag, may have 3 given utilities erroneous information about the weight of Hermo-Lag. Specif-ically, the Petitioners state that the "as installed actual weight" of Thermo-Lag may vary from 92.5 pounds per cubic foot (Ib/ft') to as much as 140 lb/ft', rather than the 78.5 lb/ft' dry density figure given by TSL The Petitioners contend that this difference is important because there often is very little room for error in calculating the weight loads for cable trPys and conduits supported by hangars and other supports. The Petitioners give no explanation of how they derived their density figures; the Petitioners' figures appear to be based, however, on the use of nominal thicknesses of preformed Thermo-Lag sections to calculate the volume of a 8 section or panel. The use of nominal-thickness dimensions instead of the actual-thickness dimensions, which may in fact be larger, may result in a calculated density that is greater than the actual density. His would occur, for example, when a panel is manufactured and finished slightly thicker than the nominal thickness, and is then weighed. If such weight is divided by the assumed volume derived from nominal dimensions, the resulting calculated density will i be greater than the density determined if the weight was divided by the actual greater volume. sgg %. as used here means the minimum rmished thkkness Ested by the manuraaurer for a standard pann! a section. 414 l l i i i

4 In any event, the Staff reviewed the adequacy of the weight and density values that TSI provides in its product literature. In discussions with the manufacturer, j the NRC Staff learned that all Thermo-Lag products, which are manufactured ~ in standard sizes, are weighed before shipment to ensure that the weight of t} the product falls within the allowed limits of TSI's quality control procedures? i Products that fall outside specified weight tolerances are rejected. ne NRC Staff was also informed by TSI that the density of Dermo-Lag { products is approximately 78 lb/ft' with some normal manufacturing tolerance i and that there has not been any change in the density of Dermo-Lag during the time the products have been manufactureel. Upon curing, there is no difference in density among Hermo-Lag products. Based on a review of TSI quality control procedures, the Staff does not consider the manufacturing density variations to be significant. 7 As detailed later in this discussion, the NRC Staff conducted an inspection of TSI's quality assurance program in December 1991, by reviewing selected criteria from 10 C.F.R. Part 50, Appendix B, including handling, control, a identification, storage and shipping of materials, control of measuring and test equipment, and control of purchased materials, and concluded that the program was adequate. nus there is assurance that the TSI weight and density specifications are being met. In summary, the NRC Staff has determined that the actual weights of Dermo-l Lag panels and sections are controlled, and has verified that an adequate quality assurance program exists. Accordingly, there is reasonable assurance ,[ that accurate weight and density specifications are being provided by the manufacturer to licensees. Derefore, in ' view of the foregoing, the NRC Staff .) has determined that the suspension of the operating licenses or construction permits for those reactors listed in the December 15, 1992 petition is not warranted based on the. Petitioners' allegation of TSI providing erroneous information about the weight or density of Hermo-Lag. -i C. Inconsistencies in Toxicity Test Results In the December 15,1992 petition, the Petitioners restate assertions made i in prior filings that tests conducted by Southwest Research Institute (SwRI)' indicate that burning Thermo. Lag can release highly toxic gases, specifically, hydrogen cyanide and carbon monoxide.: The Petitioners assert that if a fire were to occur, fire watch personnel could be overcome by these toxic gases and -{ addinon, en NRC inspection dTs1 in Doonmbar 1991 examined sucords dat indicated that Tsl had a kmssiandats 'N staft vented the existece of a weight control a9sxarication by revicwing TS! qualizy contml procedures, in practice of weishang hrme tag products. Funbermore. Teams thilnies Doctric has verited that Thermo-tag products k has received were wnhin specif ed weight mlerances, by actuaDy weighing such pmducts a receip j inspections. 'g 415 i I Iy r f;

be unable to perform their functions; further, their health and safety could be subject to " inappropriate risk." he Petitioners argue that "[i]f the staff's tests have shown that the burning of Hermo-Lag does not produce toxic fumes," there are limited possible scenarios to explain the differences in the results of the SwRI and Staff's tests that raise concerns sufficieritly troublesome to immediately shut plants down. In particular, the Petitioners are concerned that inconsistent test results may reflect poor quality assurance, or may ref!cet a change in hermo-Lag's composition that has not been tested. Prior to conducting its own tests, the NRC Staff reviewed the SwRI test report referenced by the Petitioners. The Staff review resulted in a numle of unanswered questions regarding the development of the SwRI test results, specifically, questions regarding the protocol, test procedures, and quality controls used in the conduct of the test. Rather than expending resources to resolve these unanswered questions, and to independently address issues concerning the toxicity of Hermo-Lag, in part raised by the SwRI test report, the NRC Staff decided to conduct an independent toxicological evaluation of the combustion products of Hermo-Lag fire barrier material. He NRC, in conjunction with the National Institute of Standards and Technology (NIST), determined that the products of combustion of Thermo-Lag do include hydrogen cyanide and carbon monoxide; however, Hermo-Lag combustion products also were determined to be comparable in toxicity to the burning of Douglas fir lumber or flexible plyurethane foam, and no more toxic than the products of combustion of other materials, such as cable insulation, installed in the plant.28 Rus, when circumstances require taking into account toxic releases during a plant fire, the presence of Thermo-Lag would introduce no unique considerations. Because the NRC Staff's tests clearly showed the release of some toxic substances, to the extent that the Petitioners are assuming the SwRI and NRC Staff's test results are inconsistent because they erroneously believe "the staff's tests have shown that the burning of Hermo-Lag does not produce toxic fumes," their assumption is incorrect. Moreover, a comparison of the SwRI test results to I the NRC Staff's test results is inappropriate where it has not been established that the two tests were conducted in an identical fashion, and thus such a comparison provides little or no basis to conclude that there is a quality assurance problem. or that there have been changes in the composition of Hermo-Lag.22 - Indeed, other information suggests that the Petitioners

• two hypotheses are incorrect.

10 The sesuhs of the NIsT tests were providad in NisT Report of Test IR 3951,"roticok,gical Evaluation of the Combustian Products rrom a Therma! Barrier Eterial Decomposed IJnder Flaming and Nonfiaming Condaians." dated April 29,1992. 28 in ract de NRC staft's cannpansan of the suRI test repon accompanying NIRs's sapenbar 3.1992 rding, with the Nlsr resi supon in& cates that there were at 1 cast several ust parammers or conditions that &rfamd hetween the two seats. 416 i .a l 1

~. i i h First, regarding the possibility of changes in composition, six Thermo-ug t samples of different form and vintage were submitted to NIST and, at the request of the NRC Staff, the samples were subjected to a detailed chemical analysis by NIST to characterize the chemical composition of the materials. He results of the analysis revealed the same elemental profile and similar composition and behavior for all six samples, which provides some indication that there have been no changes in the formulation of Thermo-l.ag over the years. Second, questions related to TSI's quality assurance (QA) program have been addressed in an NRC Staff inspection of TSI's QA program on December 16-20, 1991. i. De results are provided in NRC Inspection Report 99901226/91-01. De NRC inspectors verified the implementation of the QA program by reviewing l selected criteria from 10 C.F.R. Part 50, Appendix B, including nonconforming f materials; identification and control of materials; handling, storage, and shipping of materials; control of measuring and test equipment; and control of purchased materials. On the basis of the observations, the NRC inspectors concluded that TSI's QA program was adequate with the exception of two nonconformances (which were subsequently corrected) that would have had no bearing on the uniformity of the composition of Dermo-Lag. Thus, in the NRC Staff's view, there are no tuses for the Petitioners

• concerns that there may be poor quality i

assurance or there may have been a change in Thermo-Lag's composition that has not been teste4. On the basis of the NIST test results, the NRC Staff concluded in the Partial Director's Decision that the thermal decomposition of Thermo-Lag under actual fire conditions would not introduce concerns regarding either the composition or quantity of toxic materials produced, above and beyond the usual concerns regarding the toxicity of a fire that burns typicalin-plant combustibles. Accordingly, the NRC Staff noted that the toxicity levels evaluated did not suggest that precautions above and beyond those that would normally be taken f during an in-plant fire should be considered. Rus, the NRC Staff concluded that fire-watch personnel can perform their functions of finding incipient fires and notifying appropriate response personnel without sacrificing personal safety.. Because the National Institute of Standards and Technology is the recognized authority in establishing the standards by which to conduct and in conducting toxicological tests, the NRC Staff has the highest confidence in the results of the i NIST tests. Thus, the NRC Staff continues to believe that fire-watch personnel can perform their functions safely. -l ~! IIL CONCLUSION i De Petitioners request that the NRC order the immediate suspension of the operating licenses or construction permits of all nuclear power plants that i !o -I 417 l 1 ~ i i I I

use Dermo-Lag material as a fire barrier, until the Thermo-Lag is removed and replaced. Alternatively, the Petitioners request that the NRC order cach reactor licensee to remove and replace its Dermo-Lag during its next refueling outage, or before beginning operation. Rese requests presumably are based on the allegations discussed above, in addition to those addressed in the Partial Director's Decision, where I found that no substantial health and safety issues had been raised. With regard to the requests made by the Petitioners, the institution of proceedings pursuant to 10 C.F.R. 5 2.206 to shut down certain facilities using j nermo-Lag fire barrier material is appropriate only where substantial health and [ safety issues have been raised. See Consolidated Edison Co. ofNew York (Indian Point. Units I,2, and 3), CLI-75-8,2 NRC 173,175 (1975), and Washington Public Power Supp!y System (WPPSS Nuclear Project No. 2), DD-84-7,19 NRC 899, 923 (1984). With respect to the issues discussed in this Final Director's l Decision, I find no basis for taking such actions. Rather, on the basis of the review efforts by the NRC Staff, I conclude that no substantial health and safety j issues have been raised by the Petitioners. Accordingly, the Petitioners' requests l for action pursuant to section 2.206 are denied. A copy of this Decision will be placul in the Commission's Public Document Room, Gehnan Building, 2120 L Street. NW, Washington, DC 20555, and in the Local Public Document Room for the named facilities. A copy of this Decision wi!! also be filed with the Secretary of the Com-i } mission for the Commission's review as provided in 10 C.F.R. f 2.206(c) of the [ Commission's regulations. r FOR TIIE NUCLEAR REGULATORY COMMISSION Thomas E. Murley, Director Office of Nuclear Reactor .i Regulation i Dated at Rockville, Maryland, 1 this 23d day of May 1993. t i 418 ~ 5 t}}