Text

2011/02/03-Applicant's Consolidated Memorandum in Opposition to New York State'S Motion for Summary Disposition of Contention NYS-35/36 and in Support of Its Cross-Motion for Summary Disposition
	ML110460187
Person / Time
Site:	Indian Point
Issue date:	02/03/2011
From:	Bessette P M, Dennis W C, O'Neill M J, Sutton K M; Entergy Nuclear Operations, Morgan, Morgan, Lewis & Bockius, LLP
To:	; Atomic Safety and Licensing Board Panel
	SECY RAS
References
	50-247-LR, 50-286-LR, ASLBP 07-858-03-LR-BD01, RAS E-449, DESIGNATE 16 ATTACHMENTS, NOTE; RAS E-450 THROUGH RAS E-454
	Download: ML110460187 (176)
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD)In the Matter of ))ENTERGY NUCLEAR OPERATIONS, INC. ))(Indian Point Nuclear Generating Units 2 and 3))Docket Nos. 50-247-LR and 50-286-LR ASLBP No. 07-858-03-LR-BDOi February 3, 2011 APPLICANT'S CONSOLIDATED MEMORANDUM IN OPPOSITION TO NEW YORK STATE'S MOTION FOR

SUMMARY

DISPOSITION OF CONTENTION NYS-35/36 AND IN SUPPORT OF ITS CROSS-MOTION FOR

SUMMARY

DISPOSITION Kathryn M. Sutton, Esq.Paul M. Bessette, Esq.Martin J. O'Neill, Esq.William C. Dennis, Esq.COUNSEL FOR ENTERGY NUCLEAR OPERATIONS, INC February 3, 2011 S C c, 3ý S C:ý3 TABLE OF CONTENTS Page I. PRELIM INARY STATEM ENT ...........................................

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2 II. APPLICABLE NRC REQUIREMENTS

................................

5 A. 10 C.F.R. Part 54-Aging Management Review .............................................

5 B. 10 C.F.R. Part 51-Environmental Review and SAMA Analysis ....................

6 C. Controlling NEPA Principles Applicable to NRC SAMA Analysis ...............

10 III. STATEM EN T OF FACTS ........................................................................................

12 A. Procedural History of Contention NYS-35/36

.................................................

12 B .The Staff's FSE IS ...........................................................................................

16 IV. LEGAL STANDARD FOR

SUMMARY

DISPOSITION

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19 V .A R G U M E N T ....................................................................................................................

20 A. Entergy Is Entitled to Judgment As a Matter of Law Because the Facts of Record Show That Its Cost Estimates Are Sufficient For Purposes of a SA M A A nalysis ...............................................................................................

20 B. Entergy Also Is Entitled to Judgment As a Matter of Law Because the FSEIS Adequately Explains Why Implementation of the Cost-Beneficial SAMAs Is Not Required in Connection With IPEC License Renewal ....... 22 C. NYS's Motion for Summary Disposition of NYS-35/36 Should Be Dismissed As Moot And Because It Lacks A Valid Factual or Legal Basis ....... 25 VI. CONCLUSION

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38 CERTIFICATION OF COUNSEL UNDER 10 C.F.R. § 2.323(B) ....................

39 i LIST OF ATTACHMENTS Attachment Description Statement of Material Facts in Support of the Applicant's Cross-Motion for Summary Disposition of Contention NYS-35/36 (Feb. 3, 2011)2 Applicant's Response to New York State's Statement of Material Facts'Not in Dispute (Feb. 3, 2011)3 Excerpt from NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants, Vol. 1 (May 1996), available at ADAMS Accession Nos. ML040690705 4 Excerpt from Final Rule: Nuclear Power Plant License Renewal; Revisions, 60 Fed. Reg. 22,461 (May 8, 1995)5 Excerpt from Proposed Rule: Nuclear Power Plant License Renewal, 55 Fed. Reg. 29,043 (July 1.7, 1990)6 Excerpt from Final Rule:. Environmental Review for Renewal of Nuclear Power Plant Operating Licenses, .61 Fed. .Reg. 28,467 (June 5, 1996)7 Excerpt from NEI 05-0 1, "Severe Accident Mitigation Alternatives

(SAMA) Analysis, Guidance Document" (Rev. A, Nov. 2005), available at ADAMS Accession No. ML060530203 8 Final License Renewal Interim Staff Guidance LR-ISG-2006-03:

Staff Guidance for Preparing Severe Accident Mitigation Alternatives.Analyses (Aug. 2007), available at http://www.nrc.gov/reading-rm/doc-collections/isg/license-renewal.html 9 Notice of Availability of the Final License Renewal Interim Staff Guidance LR-ISG-2006-03:

Staff Guidance for Preparing Severe Accident Mitigation Alternatives Analyses, 72 Fed. Reg. 45,466 (Aug. 14, 2007)1.0 NL-09-151, Letter from Fred Dacimo, Entergy, to NRC, "Telephone Conference Call Regarding Met Tower Data for SAMA Analysis" (Nov.16, 2009), available at ADAMS Accession No. ML093 340049 11 NL-09-165, Letter from Fred Dacimo, Entergy, to NRC, "License Renewal Application

-SAMA Reanalysis Using Alternate Meteorological Data" (Dec. 11, 2009), available at ADAMS Accession No. ML093580089

-ii -

Attachment Description 12 NL- 10-013, Letter from Fred Dacimo, Entergy, to NRC, "License Renewal Application

-. Supplement to SAMA Reanalysis Using Alternate Meteorological Tower Data" (Jan. 14, 2010), available at ADAMS Accession No. ML 100260750 13 Nuclear Energy Institute; Denial of Petitionfor Rulemaking, 66 Fed. Reg.10,834, 10,834 (Feb. 20, 2001)14 Excerpt from NRR Office Instruction LIC-202, "Procedures for Managing Plant-Specific Backfits and 50.54(f) Information Requests" (Rev. 2, May 17, 2010), available at ADAMS Accession No. ML092010045.

15 Excerpt from NUREG- 1850, Frequently Asked Questions on License Renewal of Nuclear Power Reactors (Mar. 2006), available at http://www.nrc.gov/reading-rm/doc collections/nuregs/staff/srl850/

sri 850_faq_lr.pdf.

16 Excerpt from NUREG- 1800, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants (Rev. 1, Sept. 2005), available at ADAMS Accession No. ML052110007

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of ) Docket Nos. 50-247-LR and) 50-286-LR ENTERGY NUCLEAR OPERATIONS, INC. )(Indian Point Nuclear Generating Units 2 and 3) )_) February 3, 2011 APPLICANT'S CONSOLIDATED MEMORANDUM IN OPPOSITION TO NEW YORK STATE'S MOTION FOR

SUMMARY

DISPOSITION OF CONTENTION NYS-35/36 AND IN SUPPORT OF ITS CROSS-MOTION FOR

SUMMARY

DISPOSITION Pursuant to 10 C.F.R. § 2.1205(b), Entergy Nuclear Operations, Inc. ("Entergy")

hereby responds to New York State's ("NYS") motion for summary disposition of consolidated contention NYS-35/36.1 In-accordance with 10 C.F.R. § 2.1205(a), Entergy also moves the Atomic Safety and Licensing Board ("Board")

for summary disposition of the same contention.

NYS-35/36 raises issues related to Entergy's severe accident mitigation alternatives

("SAMA") analysis for Indian Point Nuclear Generating Units 2 and 3 ("IP2" and "IP3," collectively "Indian Point Energy Center" or "IPEC"). Specifically, it contests the adequacy of the NRC Staff s ("Staff')

discussion of certain cost-beneficial SAMAs, as documented in its Final Supplemental Environmental Impact Statement

("FSEIS).2 Entergy submits this consolidated memorandum both (1) in support of its cross-motion for summary disposition of NYS-35/36 and (2).in opposition to NYS's January 14, 2011 Motion.3 State of New York's Motion for Summary Disposition of Consolidated Contention NYS-35/36 (Jan. 14, 2011)("NYS Motion").2 NUREG-1437, Supp. 38, Generic Environmental Impact Statement for License Renewal of Nuclear Plants, Regarding Indian Point Nuclear Generating Unit Nos. 2 and 3, Final Report (Dec. 2010) ("FSEIS").

This filing contains 16 supporting attachments.

Attachment 1 contains the "Statement of Material Facts in Support of the Applicant's Cross-Motion for Summary Disposition of Contention NYS-35/36." Attachment 2 contains the "Applicant's Response to New York State's Statement of Material Facts Not in Dispute."

For the reasons that follow, Entergy requests that the Board resolve this matter in favor of Entergy and the Staff based on the current record and dismiss NYS-35/36 as a matter of law. In the alternative, if the Board does not grant Entergy's cross-motion for summary disposition, then it should deny NYS's competing Motion and resolve NYS-35/36 after a hearing on the merits.,4 I. PRELIMINARY STATEMENT As required by the National Environmental Policy Act ("NEPA"), the NRC's generic environmental impact statement

("GEIS") for license renewal documents the NRC's thorough evaluation of the probability and consequences of severe reactor accidents.'

A severe accident is defined as a beyond-design-basis accident that could result in substantial damage to the reactor core, whether or not there are serious off-site consequences.

As codifiedin 10 C.F.R. Part 51, the GEIS concludes that the probability-weighted consequences of a severe accident are small for all plants (i.e., the radiological impacts do not exceed permissible levels established in the 6 NRC's regulations).

IPEC, which takes its unremitting obligation to comply with NRC safety and environmental requirements very seriously, is no exception to this NRC determination.

Nonetheless, in accordance with the procedural dictates of NEPA, an EIS must contain a reasonably complete discussion of possible measures to mitigate environmental impacts.7 The GEIS contains no generic determination with respect to severe accident mitigation alternatives, or, SAMAs.8 Therefore, in accordance with 10 C.F.R. § 51.53(c)(3)(ii)(L), Entergy prepared a 4 Entergy respectfully submits that there is no triable issue of fact and that summary disposition of NYS-35/36 should be granted in its favor. In the event the Board disagrees, Entergy still maintains that NYS's Motion should be denied due to the fatal factual and legal deficiencies identified herein. In that scenario, it appears that the parties' and Board's only recourse would be to be resolve the contention after a hearing on the merits.5 See NUREG- 1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants, Vol. 1, at 5-114 to 5-115 (May 1996) ("GEIS"), available at ADAMS Accession Nos. ML040690705

[Att. 3]..6 See id; 10 C.F.R. Part 51, Subpart A, App. B, Table B-i (Postulated Accidents; Severe accidents).

7 Robertson

v. Methow Valley Citizens Council, 490 U.S. 332, 351-52 (1989).8 GEIS, Vol. 1 at 5-115 [Att. 3].

detailed site-specific evaluation of possible SAMAs for IPEC. Like prior NRC SAMA analyses, the IPEC SAMA analysis considers accident scenarios whose probability of occurrence is so low that the NRC has excluded them from the spectrum of design-basis accidents postulated for a plant. Entergy identified a number of SAMAs a's potentially cost-beneficial in its final SAMA analysis.The NRC Staff s December 2010 FSEIS fully evaluated and disclosed possible mitigation measures based on the Staff s independent and detailed technical review of Entergy's SAMA analyses.

Those analyses included the conceptual cost estimates for SAMA implementation that Entergy prepared following NRC-approved guidance.

The FSEIS concluded that the IPEC SAMA analysis is reasonable and sufficient for a license renewal submittal and complies with applicable NRC requirements.

It also correctly explained that the NRC lacks the regulatory authority to require, as part of a Part 54 license renewal proceeding, implementation of SAMAs (even those identified as~potentially cost-beneficial) if they are unrelated to aging management.

The Staff s conclusion is fully consistent with well-established legal principles governing a federal agency's authority and obligations under NEPA as well as the scope of NRC license renewal proceedings.

Furthermore, the U.S. Supreme Court has held that "NEPA imposes no substantive requirement that mitigation measures actually be taken." 9 While NEPA demands that a "hard look" be given to the potential measures for mitigating environmental impacts, it does not provide a legal basis for mandating the implementation of such measures.

10 Furthermore, the NRC's license renewal regulations in Part 54 are, by design, confined to aging management issues during the renewal period. As such, the Commission has made clear 9 Methow Valley, 490 U.S. at353 n.16.10 Id. at 350 ("NEPA itself does not mandate particular results, but simply prescribes the necessary process.").

that it does not intend to impose requirements on a license- renewal applicant that go beyond what is necessary to adequately manage aging effects during extended operations.

Any question regarding the adequacy of a plant's current licensing basis ("CLB"), including proposed enhancements thereto, generally must be addressed under the NRC's Part 50 backfit rule and is separate from the license renewal process.For thesereasons, Entergy and the Staff have fully met their obligations under NEPA and the NRC's NEPA-implementing regulations in Part 51. Accordingly, the Board should enter summary disposition of NYS-35/36 in favor of Entergy and the Staff. The issues or omissions alleged therein are moot in light of the Staff's issuance of the FSEIS.Regardless, NYS's Motion is not supported by the relevant facts, law, and guidance.

In particular, NYS alleges that Entergy's SAMA implementation cost estimates are "unfinished" for purposes of NEPA." But as shown below, this is not the case. And, because there is no regulatory basis for mandating implementation of the SAMAs (which would involve modifications to the IPEC CLBs) as part of this license renewal proceeding, there is no need for the additional engineering cost estimates NYS alleges are necessary.

Finally, none of the cost-beneficial SAMAs cited by NYS is necessary for adequate protection of the public health and safety or, given the findings in the GEIS, necessary to ensure that the consequences of severe accidents are small. Accordingly, NYS's Motion should be denied because it lacks an adequate factual and legal basis." NYS Motion at 17.

II. APPLICABLE NRC REQUIREMENTS The scope of a license renewal proceeding is limited to discrete safety and environmental issues, as defined by the Commission's regulations at 10 C.F.R. §§ 54.4 and 54.21(a), and 10 C.F.R. §§ 51.75(d) and 51.95(c).1 2 Key relevant requirements are discussed below.A. 10 C.F.R. Part 54-Aging Management Review Part 54, which governs license renewal, limits the scope of license renewal to aging-13 management-related matters and does not require applicants to revisit CLB issues. Part 54.requires applicants to demonstrate how current or proposed programs will effectively manage the effects of aging during a period of extended operation, such that important systems, structures, and components

("SSCs") will continue to perform their intended function in the renewal period. 14 Part 54's regulatory history clarifies the relationship between Part 54 and the backfit rule in 10 C.F.R. § 50.109. For example, the Statement of Considerations

("SOC") for the NRC's July 1990 proposed Part 54 rulemaking states: If the staff or the licensee seeks to make changes in a plant's licensing basis for reasons other than age-related'degradation, they should be pursued either in the existing operating license or the renewed license, once issued. 'Staff-initiated changes would be evaluated in accordance with the backfit rule, 10 CFR 50.109.15 12 Duke Energy Corp. (McGuire Nuclear Station, Units 1 & 2; Catawba Nuclear Station, Units 1 & 2), CLI-O1-20, 54 NRC 211, 212 (2001) (citing Fla. Power & Light Co. (Turkey Point Nuclear Power Plant, Units 3 & 4), CLI-O1-17, 54 NRC 3, 6-13 (2001)).13 See Turkey Point., CLI-01-17, 54 NRC at 7-9.14 Final Rule: 'Nuclear Power Plant License Renewal; Revisions, 60 Fed. Reg. 22,461, 22,463 (May 8, 1995) [Att.4]. See also 10 C.F.R. § 54.21(a).

For example" applicants must identify any additional actions (e.g., inspections, testing, maintenance) needed to manage aging effects during the license renewal term.15 Proposed Rule: Nuclear Power Plant License Renewal, 55 Fed. Reg. 29,043, 29,047 (July 17, 1990)'(emphasis added) [Att. 5]. Cf Duke Energy Corp., (McGuire Nuclear Station, Units 1 & 2; Catawba Nuclear Station, Units 1 & 2), CLI-02-28, 56 NRC 373, 388 n.77 (2002) (stating that the ultimate agency decision on whether to require facilities with ice condenser containments to implement any particular SAMA will fall under a Part 50 current licensing basis review).

In the May 1995 SOC for its final revisions to Part 54, the Commission reaffirmed that it "does not intend to impose requirements on a licensee that go beyond what is necessary to adequately manage aging effects."'16 B. 10 C.F.R. Part 51-Environmental Review and SAMA Analysis The NRC promulgated Part 51 to implement NEPA's requirements and, in 1996, specifically amended Part 51 to address the scope of its environmental review for license renewal applications.1 7 The NRC divided the environmental requirements for license renewal into generic and plant-specific components.

It prepared the GEIS to evaluate and document those generic impacts that are well understood based on extensive current-fleet operating experience.

The NRC Staff must prepare a plant-specific supplement to the GEIS that adopts applicable generic impact findings from the GEIS, evaluates any new and significant information, and discusses site-specific impacts. Under. Section 51.95(c)(4), the relevant decision standard is whether the adverse environmental impacts of license renewal (when compared with the environmental impacts of other energy generating alternatives) are so great that preserving the option of license renewal for energy planning decisionmakers would be unreasonable.19 As relevant here, the NRC's GEIS provides an evaluation of severe accident impacts that applies to all U.S. nuclear power plants. 2 A severe accident is defined as a beyond-design-basis accident that could result in substantial damage to the reactor core, whether or not there are 16 Final Rule: Nuclear Power Plant License Renewal; Revisions, 60 Fed. Reg. at 22,490 [Att. 4].17 See Final Rule: Environmental Review for Renewal of Nuclear Power Plant Operating Licenses, 61 Fed. Reg.28,467 (June 5, 1996), amended by 61 Fed. Reg. 66,537 (Dec. 18, 1996) [Att. 6].18 See-10 C.F.R. § 51.95(c).19 Final Rule: Environmental Review for Renewal of Nuclear Power Plant Operating Licenses, 61 Fed. Reg. at 28,485 [Att. 61. See also 10 C.F.R. § 51.95(c)(4).

20 Entergy Nuclear Generation Co. (Pilgrim Nuclear Power Station), CLI-10-1 1, slip op. at 37-38 (Mar. 26, 2010).

serious off-site consequences.

2 1 1 Based on the GEIS evaluation, 10 C.F.R. Part 51 concludes that the "lt]he probability weighted consequences of atmospheric releases, fallout:onto open bodies of water, releases to ground water, and societal and economic impacts from severe accidents are smallfor all plants." 2 2 The NRC has noted that the GEIS analyses represent adequate, plant-specific estimates of the impacts from severe accidents that would generally over-predict, rather than under-predict, environmental consequences.23 Thus, a plant-specific analysis of severe accident impacts is not required in individual license renewal proceedings.

As the Commission recently put it, "NRC SAMA analyses are not a substitute for, and do not represent, the NRC NEPA analysis of potential impacts of severe accidents." 2 4 Nonetheless, Part 51 states that if the Staff has not previously considered SAMAs for a license renewal applicant's plant in an EIS or in an environmental assessment, then the applicant must complete an evaluation of alternatives that may mitigate severe accidents.

2 5 The need for a site-specific SAMA analysis flows from the NRC's NEPA regulations (which require consideration of mitigation alternatives in initial and supplemental EISs), and the Third Circuit's 1989 Limerick decision (which requires a NEPA review of severe accident mitigation design alternatives

("SAMDAs")

at the initial operating license stage).2 6 At the time of its 1996 Part 51 rulemaking, the Commission was unable to reach a generic conclusion regarding mitigation 21 FSEIS, Vol. 1 at 5-3.22 10 C.F.R. Part 51, Subpart A, App. B, Table B-1 (Postulated Accidents; Severe accidents) (emphasis added).23 Final Rule: Environmental Review for Renewal of Nuclear Power Plant Operating Licenses, 61 Fed. Reg. at 28,480 [Att. 6].24 See Pilgrim, CLI-10-11, slip op. at.37.25 10 C.F.R. § 51.53(c)(3)(ii)(L);

see also id. Part 51, Subpart A, App. B, Table B-1.26 Final Rule: Environmental Review for Renewal of Nuclear Power Plant Operating Licenses, 61 Fed. Reg. at 28,480 [Att. 6]; see also Limerick Ecology Action v. NRC, 869 F.2d 719, 736-39 (3rd Cir. 1989) (holding that the NRC could not generically dispense with the consideration of SAMDAs, under NEPA, through a policy statement issued pursuant to its Atomic Energy Act authority).

SAMDAs apply to plants seeking their initial operating licenses.

SAMAs, in contrast, apply to plants seeking renewal of their previously-issued operating licenses.

alternatives for purposes of license renewal, because not all licensees had completed the agency's Part 50-based ongoing regulatory program related to severe accident mitigation.

2 7 By definition, a SAMA analysis considers postulated events whose probability.

of occurrence is so low that they are excluded from the spectrum of design-basis accidents postulated for a plant by the Commission's regulations.

2 8 The purpose of a SAMA analysis is to identify potential changes to a nuclear power plant, or its operations, that (1) could further reduce the already very low risk of postulated severe reactor accident scenarios, and (2) may be cost-beneficial to implement.

2 9 Whether a SAMA may be cost-beneficial to implement is based upon a weighing of the cost to implement the SAMA with the reduction in (monetized) risks to public health, occupational health, and offsite and onsite property.3 0 TheNuclear Energy Institute

("NEI") has issued a guidance document, NEI 05-01, Revision A, to assist NRC license renewal applicants in preparing SAMA analyses.3 1 The Staff has approved and recommended the use of NEI 05-01 by license renewal applicants.

3 2 SAMA analysis uses probabilistic risk assessment

("PRA") and cost-benefit analysis methods..

Broadly speaking, a SAMA analysis includes four major steps: (1) quantification of the level of risk 27 See Final Rule: Environmental Review for Renewal of Nuclear Power Plant Operating Licenses, 61 Fed. Reg.at 28,480-481 (discussing the SAMA analysis requirement)

[Att. 6].28 See FSEIS, Vol. 1 at 5-2 to 5-3, 5-11.29 See Pilgrim, CLI 11, slip op. at 3.30 See Duke Energy Corp. (McGuire Nuclear Station, Units 1 & 2; Catawba Nuclear Station, Units I & 2), CLI-02-17, 56 NRC 1, 7-8.(2002).

31 NEI 05-01, Severe Accident Mitigation Alternatives (SAMA) Analysis, Guidance Document, at i (Rev. A, Nov. 2005) ("NEI 05-01"), available at ADAMS Accession No. ML060530203

[Att. 7]. NEI 05-01 relies upon regulatory analysis techniques described inNUREGiBR-0 184, Regulatory Analysis Technical Evaluation Handbook (Jan. 1997), is a result of experience gained through past SAMA analyses, and incorporates insights gained from review of NRC evaluations of SAMA analyses and associated requests for additional information

("RAIs").32 See Final License Renewal Interim Staff Guidance LR-ISG-2006-03:

Staff Guidance for Preparing Severe Accident Mitigation Alternatives Analyses (Aug. 2007) ("LR-ISG-2006-03"), available at http://www.nrc.gov/reading-rm/doc-collections/isg/license-renewal.html

[Att. 8]. See also Notice of Availability of the Final License Renewal Interim Staff Guidance LR-ISG-2006-03:

Staff Guidance for Preparing Severe Accident Mitigation Alternatives Analyses, 72 Fed. Reg. 45,466 (Aug. i4, 2007) [Att. 9].

associated with potential reactor accidents using plant-specific PRA and other risk models;(2) examination of the major risk contributors and identification of possible ways (i.e., SAMAs)of reducing that risk; (3) estimation of the benefits and implementation costs associated with specific SAMAs; and (4) comparison of the costs and benefits of the identified SAMAs to determine whether the SAMA is cost-beneficial.

Relevant to NYS-35/36, NEI 05-01 provides the following guidance with respect to the estimation of SAMA implementation costs*As SAMA analysis focuses on establishing the economic viability of potential plant enhancement when compared to attainable benefit, often detailed cost estimates are not required to make informed decisions regarding the economic viability of a particular modification.

SAMA implementation costs may be clearly in excess of the attainable benefit estimated from a particular analysis case.For less clear cases, engineering judgment may be applied. to determine if a more detailed cost estimate is necessary to formulate a conclusion regarding the economic viability of a particular SAMA.Nonetheless, the cost of each SAMA candidate should be conceptually estimated to the point where economic viability of the proposed modification can be adequately gauged.3 3 Notably, the Commission recently cited NEI 05-01 in stating that "detailed cost estimates often are not necessary to gauge the economic viability of a particular SAMA candidate." 3 4 NEI 05-01 further states that the use of conservative (i.e., low) implementation cost estimates may be used to offset uncertainty in the cost estimates.

3 5 33 NEI 05-01, at 28 [Att. 7].34 Pilgrim, CLI-10-1 1, slip op. at 39 n.151.35 NEI 05-0 1, at 30 [Att. 7].

C. Controlling NEPA Principles Applicable to NRC SAMA Analysis The NRC's consideration of mitigation alternatives-which includes SAMAs-is governed by the NEPA rule of reason.3 6 Under NEPA's rule of reason, the Commission need only consider the range of alternatives reasonably related to the scope and goals of the proposed major federal action.3 7 Here, the proposed action is renewal of the IP2 and IP3 operating licenses.

As defined in 10 C.F.R. Part 54 and its regulatory history, the NRC's safety review for license renewal centers on managing the effects of aging resulting from operation beyond the initial license term.3 8 As defined in the GEIS and 10 C.F.R; § 51.95(c)(4), the purpose or focus of the NRC's environmental review of a license renewal application is to determine whether or not the adverse environmental impacts of license renewal are so great that preserving the option of license renewal for energy planning decisionmakers would be unreasonable.

As noted above, SAMA analyses are not a substitute for, and do not represent, the NRC NEPA analysis of potential impacts of severe accidents.

3 9 Rather, the NRC's GEIS for license renewal provides a generic evaluation of severe accident impacts and the technical basis for the NRC's conclusion that the probability-weighted consequences of severe accidents are small for all plants.4 0 SAMA analysis is a site-specific mitigation analysis, not a worst-case or a best-case 36 McGuire/Catawba, CLI-02-17, 56 NRC at 12 (citing Vt. Yankee Nuclear Power Corp. v. Natural Res. Def Council, Inc., 435 U.S. 519, 551 (1978) ("NRDC");

Citizens Against Burlington

v. Busey, 938 F.2d 190, 195 (D.C. Cir. 1991)).37 Nuclear Mgmt. Co., LLC (Monticello Nuclear Generating Plant), LBP-05-31, 62 NRC 735,753 (2005).38 See 10 C.F.R. §* 54.29; Final Rule: Nuclear Power Plant License Renewal; Revisions, 60 Fed. Reg. at 22,463[Att. 4].39 Pilgrim, CLI-10-1 1, slip op. at 37.40 Id. at 37-38 (citing GEIS, Vol. 1 at 5-12 to 5-106).

impacts analysis.41 The goal of SAMA analysis is only to determine what potential plant enhancements may be cost-beneficial in mitigating the effects of a postulated severe accident.4 2 Consistent with the U. S. Supreme Court's holding in Methow Valley, the Commission has stated that NEPA does not require actual implementation of mitigation measures-including SAMAs.'3 In Methow Valley, the Court held that NEPA requires a "reasonably complete discussion of possible mitigation measures," but "imposes no substantive requirement that mitigation measures actually be taken." 4 4 It further noted that there is a fundamental distinction between a procedural requirement that an agency discuss possible mitigation measures in sufficient detail, and a substantive requirement that the agency implement those measures.45 Thus, "because it is only procedural and not substantive in nature, NEPA does not require agencies to implement any of the mitigation measures discussed in the FEIS." 4 6 In summary, the Commission has described its obligations with respect to consideration of mitigation alternatives as follows: The purpose of addressing possible mitigation measures in an FEIS is to ensure that the agency has taken a hard look at the potential environmental impacts of a proposed action. An EIS therefore must address mitigation measures in sufficient detail to *ensure that 41 Id. at 38.42 Id. at 39.43 Id. at 38 (citing Methow Valley, 490 U.S. at 353).44 Methow Valley, 490 U.S. at 352, 353 n.16. See also, Laguna Greenbelt, Inc. v. U.S. Dep't of Transp.,.42 F.3d 517, 528 (9th Cir. 1994) ("NEPA does not require a fully developed plan that will mitigate all environmental harm before an agency can act; NEPA requires only that mitigation be discussed in sufficient detail to ensure that environmental consequences have been fully evaluated." (citationsomitted));

Nat'l Parks & Conservation Ass 'n v. US. Dep 't of Trans., 222 F.3d 677, 681 n.4 (9th Cir. 2000) ("Contrary to National Parks' assertion, a mitigation plan need not be legallyenforceable, funded or even in final form to comply with NEPA's procedural-requirements.");

Cnty. of Rockland v. FAA, 335 Fed.Appx.

52 (DC. Cir. 2009) ("NEPA does not impose 'substantive requirement that a complete mitigation plan be actually formulated and adopted' before agency can act") (quoting Methow Valley, 490 U.S. at 352).45 Methow Valley, 490 U.S. at 352-53 46 Citizens Concerned About Jet Noise, Inc. v. Dalton, 48 F.Supp.2d 582, 603 (E.D.Va. 1999) (citing Methow Valley, 490 U.S. at 353).-I1 -

environmental consequences have been fairly evaluated.

An EIS need not, however, contain a complete mitigation plan, or a detailed explanation of specific measures which, will be employed.

Indeed, a mitigation plan need not be legally enforceable, funded or even in final form to comply with NEPA's procedural requirements.

As long as the potential adverse impacts from a proposed action have been adequately disclosed, it is not improper for an EIS to describe mitigating measures in general terms and rel[y] on general processes

...47 As the foregoing suggests, the NRC has broad discretion in determining how thoroughly to analyze a particular subject, including mitigation measures 1 4 8 Where environmental impacts are minimal or speculative, an EIS may focus on broad mitigation measures.4 9 To prevail on a NEPA claim, therefore, an intervenor must demonstrate that the Staff has "unduly ignored or minimized pertinent environmental effects." 5°III. STATEMENT OF FACTS A. Procedural History of Contention NYS-35/36 Entergy originally submitted its SAMA analysis in April 2007 as part of the Environmental Report ("ER") for the Indian Point license renewal application.

5" The NRC 47 Hydro Res., Inc. (P.O. Box 777, Crownpoint, NM 87313), CLI-06-29, 64 NRC 417, 426-27 (2006) (emphasis added) (internal quotation marks and citations omitted) (finding FEIS discussion of mitigation measures to be adequate and concluding that "Intervenors

.... demand a level of detail not required by NEPA").48 S. Nuclear Operation Co. (Early Site Permit for Vogtle ESP Site), LBP-09-7 69 NRC 613, 631-32 (2009)(citing La. Energy Services, L.P. (Claiborne Enrichment Center), CLI-98-3, 47 NRC at 103 (1998); Vt. Yankee Nuclear Power Corp. (Vt. Yankee Nuclear Power Station), ALAB-919, 30 NRC 29, 44 (1989); Limerick Ecology Action, Inc., v. NRC, 869 F.2d at 739.49 Border Power Plant Working Group v. Dep't of Energy, 467 F.Supp.2d 1040, 1070 (S.D. Cal. 2006) (quoting Nat 'l Parks & Conservation Ass 'n v. US.. Dep 't of Transp., 222 F.3d 677, 681 (9th Cir. 2000)). See also Vogtle ESP Site, 69 NRC at 631-32, 692 (stating that an agency "may decline to examine issues the agency in good faith considers remote and speculative or inconsequentially small," and finding that some impact greater than SMALL to the resource at issue must be involved to trigger the requirement for a more detailed alternatives analysis than presented in the FEIS) (internal quotation marks and citations omitted).50 Duke Energy. Corp. (McGuire Nuclear Station, Units 1 & 2; Catawba Nuclear Station, Units 1 & 2), CLI-03-17, 58 NRC 419, 431.51 See generally Indian Point Energy Center, Applicant's Environmental Report, Operating License Renewal Stage § 4.21 (Apr. 2007) (Severe Accident Mitigation Alternatives);

id., Att. E (Severe Accident Mitigation Alternatives Analysis), available at http://www.nrc.gov/reactors/operating/licensing/renewal/applications/

indian-point.html#application.

documented its initial review of Entergy's SAMA in its December 2008 draft SEIS ("DSEIS").

5 2 Subsequently, in response to Staff RAIs, Entergy revised its original SAMA analysis.Specifically, in a November 2009 letter to the NRC Staff, Entergy confirmed the existence of an error in the wind direction inputs to the code used for the SAMA analysis.5 3 Consequently, Entergy committed to correct the wind direction inputs and accordingly re-analyze the SAMAs for both units, including revising the estimates of off-site population dose and offsite economic costs.5 4 Shortly thereafter, in December 2009, Entergy submitted its Revised SAMA Analysis with the corrected wind direction inputs.5 5 In March 2010, NYS submitted new contentions NYS-35 and NYS-36 as challenges to Entergy's Revised SAMA Analysis.5 6 Both contentions alleged that Entergy and the Staff have failed to comply with NRC regulations and guidance, ignored the "legal mandate'.'

imposed by the Third Circuit's Limerick decision, 5 7 and violated the Atomic Energy Act ("AEA"), NEPA, and the Administrative Procedure Act ("APA").5 8 NYS-35 alleged, in particular, thatEntergy has not provided sufficiently complete cost-benefit analyses for nine specific SAMAs already identified in the December 2009 Revised 52 NUREG-1437, Supp. 38j Generic Environmental Impact Statement for'License Renewal of Nuclear Plants, Regarding Indian Point Nuclear Generating Unit Nos. 2 and 3,. Draft Report for Comment (Dec. 2008)("DSEIS").

53 See NL-09-151, Letter from Fred Dacimo, Entergy, to NRC, "Telephone Conference Call Regarding Met Tower Data for SAMA Analysis" at 1 (Nov. 16, 2009) ("NL-09-15 1"), available at ADAMS Accession No.ML093340049

[Att. 10].4 Id.55 See NL-09-165, Letter from Fred Dacimo, Entergy, to NRC, "License Renewal Application

-SAMA Reanalysis Using Alternate Meteorological Data" (Dec. 11, 2009) ("NL-09-165" or "Revised SAMA Analysis"), available at ADAMS Accession No. ML093580089

[Att. 11].56 See State of New York's Motion for Leave to File New and Amended Contentions Concerning the December 2009 Reanalysis of Severe Accident Mitigation Alternatives (Mar. 11, 2010) ("Motion for Leave");*

State of New York's New and Amended Contentions Concerning the December 2009 Reanalysis of Severe Accident Mitigation Alternatives Reanalysis (Mar. 11, 2010) ("March 2010 SAMA Contentions").

57 See Limerick Ecology Action, Inc., v. NRC, 869 F.2dat .719.58 See March 2010'SAMA Contentions at 14-17 & 27-35.

SAMA Analysis as potentially cost-beneficial.

5 9 NYS-3 5 centered on Entergy's intention, as stated in the ER and Revised SAMA Analysis, to perform additional, future engineering project cost-benefit analysis for those SAMAs identified as potentially cost-beneficial.60 This, NYS claimed, "indefinitely postpones" the cost-benefit analyses necessary to "finally determine" which of the nine SAMAs are cost-effective, and precludes a rational decision regarding which mitigation measures, if any, are sufficiently cost-effective to warrant implementation through license conditions in the renewed operating licenses.61 In short, NYS claims that the Staff needs"final", implementation cost estimates for theseSAMAs, but only to determine which of them require implementation as a condition to license renewal.NYS-36 focused on nine other SAMAs, asserting thatEntergy's Revised SAMA Analysis showed these SAMAs, for the first time, to have substantially greater benefits in excess of their costs.62 More specifically, NYS claimed that these SAMAs now have a "margin of benefit over cost [that] is so high that there is little chance that even a more complete cost estimate will be able to eliminate the substantialbenefit." 6 3 NYS-36 further alleged that the Staff acceptance of the Revised SAMA Analysis would violate NEPA and the APA absent a rational basis for any decision to not require implementation of "clearly" cost-effective SAMAs.6 4 Both Entergy and the Staff opposed the admission of NYS-35 and NYS-36 in their entirety.

Nonetheless, in LBP-10-13, the Board admitted and consolidated the two new, 59 See id at 13, 15-17, 22-35 60 See, e.g., id. at 15, 23-25.61 Id. at 25; see also Motion for Leave at 11.62 March 2010 SAMA Contentions at 36.63 Id. at 46. See also id at 37.64 Id. at 40-41.

contentions.65 Importantly, however, the Board admitted the contentions only in part, emphasizing that the Staff need not require implementation of SAMAs as part of a license renewal proceeding.

6 6 Therefore, the*Board rejected as outside the scope of this proceeding any portion of these contentions demanding implementation of cost-beneficial SAMAs.6 7*The Board concluded, however, that the Staff has the "option" to impose, prior to license renewal, additional conditions to an applicant's CLB as a backfit underl10 C.F.R. Part 50, and that as a result of its SAMA review, the Staff could choose to impose such conditions that are necessary to protect the environment under a Part 50 backfit procedure.

6 8 Based on this reasoning, the Board found NYS-35 admissible insofar as it alleged that the Staff's Draft SEIS did not provide a rational basis for granting the license extension without mandating a CLB backfit of any SAMAs that in any "final analysis" are classified as cost-effective.69 In addition, in view of Entergy's stated intention to perform further engineering cost-benefit analyses, the Board admitted the portion of NYS-35 calling for completion of the cost-benefit analysis to determine which SAMAs are cost-beneficial to implement.

7 0 The Board also admitted NYS-36 in part, finding a triable issue of fact as to whether the Staff has fulfilled its duty under NEPA to take a hard look at the SAMAs deemed potentially cost-beneficial in Entergy's Revised SAMA Analysis.

The Board stated that the Staff must explain in its record of decision why it would authorize license renewal without requiring 65 See Entergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 and 3), LBP- 10-13, slip op. at 26-30, 34-36 (June 30, 2010).66 See id at 29, 34..67 See id.68 Id. at 4, 28-29.69 Id. at 29.70 Id. at 35.

implementation of plainly cost-beneficial SAMAs- as a condition precedent to the granting of license renewal.7'Entergy and the Staff filed petitions for interlocutory review of the Board's decision to admit NYS-35/36.

7 2 The Commission denied the petitions on the ground that they did not meet the strict criteria for interlocutory review in Section 2.341(f)(2).

7 3 However, in so ruling, the Commission clarified the scope and focus of NYS-35/36.

Specifically, the Commission stated that it does not read LBP- 10-13 to require the Staff either to impose license conditions or to undertake formal Part 50 backfit analyses for the potentially cost-beneficial SAMAs identified in contention NYS-35/36.

7 4 The Commission further emphasized that NYS-35/36 seeks further"explanation" of the SAMA analysis conclusions, including whether the current SAMA cost-benefit estimates "are sufficient for the NEPA analysis." 7 5 B. The Staffs FSEIS The Staff issued its FSEIS in December 2010. Section 5.2 of the FSEIS summarized the Staff s final review of Entergy's.

SAMA evaluation for IP2 and IP3, including its December 2009 Revised SAMA Analysis.7 6 Appendix G of the FSEIS documented the Staff sdetailed technical evaluation of Entergy's SAMA analysis.7 7 The FSEIS concluded that the methods used, and the implementation of those methods, were. sound, and that "[t]he treatment of SAMA benefits and 71 Id.72 Applicant's Petition for Interlocutory Review of LBP-10-13 (July 15, 2010); NRC Staff's Petition for Interlocutory Review of the Atomic Safety and Licensing Board's Decision Admitting New York State Contentions 35 and 36 on Severe Accident Mitigation Alternatives (LBP-10-13) (July 15, 2010).73 Entergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 and 3"), CLI-10-30, slip op. at 5-7 (Nov. 30, 2010). Although the Commission denied interlocutory review, it noted that Entergy's and the Staff's arguments "are not without force," and that portions of the Board ruling may warrant review later in the proceeding.

Id. at 6.SId. at6.75 Id.76 FSEIS, Vol. 1 at 5-4 to 5-13.77 Id., Vol. 3, App. G.

costs support the general conclusion that the SAMA evaluations performed by Entergy are reasonable and sufficient for the license renewal submittal." 7 8 The Staff reviewed Entergy's cost estimates for implementing SAMAs. Section G.5 of the FSEIS describes this review.79 As noted therein, Entergy estimated the costs of implementing the candidate SAMAs through the application of engineering judgment and the use of other licensees' estimates for similar improvements.

Table G-6 lists, among other items, the SAMA implementation cost estimates.

8 0 The Staff reviewed the bases for these cost estimates and, for certain improvements, it compared the cost estimates to estimates developed previously for similar improvements (including estimates developed as part of other licensees' SAMA for operating reactors and advanced light-water reactors).

8'As stated in Section G.5, as part of its Revised SAMA Analysis using corrected meteorological data, Entergy subjected a subset of the candidate SAMAs to more rigorous cost-estimating techniques; i.e., those SAMAs that appeared to be cost-beneficial based on the new benefit estimate and the original implementation cost estimate.8 2 For two IP2 SAMAs (IP SAMAs 17 and 40) and four IP3 SAMAs (IP3 SAMAs 17, 20, 40, and 50), the revised cost estimate resulted in the SAMA becoming non-cost-beneficial.

8 3 For each of these SAMAs, the Staff requested that Entergy provide the basis for the revised cost estimate, including a breakdown of the major cost factors.84 Entergy provided this information by letter dated January 78 Id., Vol. 1 at 5-11 (emphasis added).79 Id., Vol. 3 at G-34 to G-40.80 Id. at G-36 to G-38.81 Id. at G-34 to. G-35.82 Id. at G-39.83 Id.84 Id.

14, 2010, in which Entergy explained that it developed the revised cost estimates using its standard process for preparing conceptual-level project cost estimates.85 The Staff reviewed this additional cost information to determine the degree to which the revised cost estimates and their constituent costs comport with the nature, magnitude and complexity of each change.8 6 It determined that Entergy' s revised cost estimates were reasonable, and that they resulted in an appropriate determination that the subject candidate SAMAs are not cost-beneficial.

8 7 The Staff further concluded that all of Entergy's cost-estimates were reasonable, generally consistent with prior estimates by other licensees, and sufficient and appropriate for use in the SAMA evaluation.

8 8 Section 5.2.6 of the FSEIS also discussed the Board's admission of NYS-35/36 and directly addresses the allegations presented in the consolidated contention.

It states,* in part:[T]he NRC staff has determined that none of the potentially cost-beneficial SAMAs are related to the license renewal requirements in 10 CFR Part 54 (i.e., managing the effects of aging) (SEIS § 5.2.6)., Under the NRC's regulatory system, any potentially cost-beneficial SAMAs that do not relate to 10 CFR Part 54 requirements.would be considered, to the extent necessary or appropriate, under the agency's oversight of a facility's current operating license in accordance with 10 CFR Part 50 requirements, inasmuch as such matters would pertain not just to the period of extended operation but to operations under the current operating license term as well. Thus, there is no regulatory basis to suggest that potentially cost-beneficial SAMAs that are unrelated to Part 54 requirements must be imposed as a backfit to the CLB, as a condition for license renewal.85 Id. (citing NL-10-013, Letter from Fred Dacimo, Entergy, to NRC, "License Renewal Application

-Supplement to SAMA Reanalysis Using Alternate Meteorological Tower Data" (Jan. 14, 2010) ("NL- 10-0 13), available at ADAMS Accession No. ML100260750

[Att. 12]).86 Id. The associated modifications involved either major plant modifications or changes to safety-related SSCs.Id. For example, the modifications for which Entergy prepared more detailed cost estimates involve erecting a barrier to protect the containment liner, installing secondary side guard pipes, increasing secondary side pressure capacity, enhancing the reactor coolant system depressurization capabilities.

Id. Thus, in addition to hardware costs, the modifications would require extensive design work and safety analysis calculations, including seismic analyses, thermal analyses, and analyses for piping or penetration interferences.

Id.87 Id. at G-39 to G-40.88 Id. at 40.

Finally, the NRC staff notes that SAMAs, by definition, pertain to severe accidents-i.e., those accidents whose consequences could be severe, but whose probability of occurrence is so low that they may be excluded from the spectrum of design basis accidents

("DBAs)that have been postulated for a plant (see GEIS §§ 5.3.2, 5.3.3, 5.4);this is consistent with the conclusions reached in.§ 5.2.2 of this SEIS concerning severe accidents at IP2 and IP3. The Commission has previously concluded, as a generic matter, that the probability-weighted radiological consequences of severe accidents are SMALL.GEIS § 5.5.2; 10 CFR Part 51, App. B,-KTable B 1. As stated in§§ 5.1.1 and 5.1.2 above, no significant new information has been identified that .would remove IP2 and IP3 from these generic determinations.

Thus, there is no regulatory basis to impose any of the potentially cost-beneficial SAMAs as a condition for license renewal of IP2 and 1P3--even if those potentially cost-beneficial SAMAs are "finally" found to be cost-beneficial.

8 9 Based on these considerations, the FSEIS concluded that the Staff "has provided a detailed discussion of SAMA costs and benefits in this SEIS, which satisfies the NRC's obligation, under NEPA and related case law, to consider SAMAs in a license renewal proceeding." 9 0 IV. LEGAL STANDARD FOR

SUMMARY

DISPOSITION The NRC's standards for summary disposition are set forth in 10 C.F.R. § 2.710.91 Summary disposition is appropriate where relevant documents or affidavits show that there is no genuine issue as to any material fact and that the moving party is entitled to a decision as a matter of law.9 2 A party opposing a properly-supported summary disposition motion may not rest upon mere allegations or denials, but must state specific, significantly probative facts showing that there is a genuine issue of fact for hearing.9 3 It is not sufficient, however, for there 89 Id, Vol. 1 at 5-11 to 5-12 (emphasis added).9' Id. at 5-1l.91 Although 10 C.F.R. § 2.1205 governs the submission of motions for summary disposition in Subpart L proceedings, it directs the Board to apply the standards for summary disposition in Subpart G of Part 2.92 Pilgrim, CLI-10-1 1, slip op. at 12 (citing 10 C.F.R. § 2.710(d)(2)).

93 Id. (citing l'0 C.F.R. § 2.710(b)).

merely to be the existence of "some alleged factual dispute between the parties, for "the requirement is that there be no genuine issue of material fact." 9 4 Therefore, only disputes over facts that might affect the outcome of a proceeding would preclude summary disposition."Factual disputes that are ... unnecessary will not be counted.'9 5 V. ARGUMENT A. Entergy Is Entitled to Judgment As a Matter of Law Because the Facts of Record Show That Its Cost Estimates Are Sufficient For Purposes of a SAMA Analysis The Board admitted that portion of NYS-35/36 calling for completion of the cost-benefit analysis to determine which SAMAs are cost-beneficial to implement as a contention of omission.96 Specifically, the Board stated that the applicant must supply information that is sufficiently complete for the Commission to be able to explain its decision.9 7 Entergy has done precisely that, as reflected in the ER, the Revised SAMA Analysis and, finally, the Staff s FSEIS review. Accordingly, the Board should dismiss this aspect ofNYS-35/36 as moot.NRC regulations require applicants like Entergy to perform site-specific SAMA analyses.However, they do not mandate or prescribe a specific methodology for conducting those evaluations, including the associated cost-benefit analysis of SAMA candidates.

In this instance, Entergy followed the guidance provided in NEI 05-01, which the NRC Staff has endorsed.98 ER Section 4.21.5.4 and Attachment E to the ER (Sections E.2.3 and E.4.3) describe the manner in which Entergy prepared its original SAMA cost implementation cost estimates.

As explained therein, and consistent with NEI 05-01, Entergy developed cost estimates for 94 Id at 12 (quoting Anderson v. Liberty Lobby, 477 U.S. 242, 247-48 (1986).95 Id. (quoting Liberty Lobby, 477 U.S. at 248).96 Indian Point, LBP-10-13, slip op. at 29.97 Id. at 35.98 See Pilgrim, CLI 11, slip op. at 4 n. 11 (citing NEI 05-01 as NRC-endorsed guidance for performing SAMA analyses).

implementing each SAMA candidate to the extent necessary to allow it to make an informed judgment about the economic viability of the proposed improvement.

9 9 Entergy followed this same cost-estimating process in its Revised SAMA Analysis.t 0 0 For many SAMAs, Entergy used the same implementation cost estimates reported in Tables E.2-2 and E.4-2 of the April 2007 ER.'0' In some cases, however, it was necessary to perform more detailed, plant-specific cost estimates to confirm whether certain SAMAs were, in fact, potentially cost-beneficial.

1 0 2 Tables 4 through 7 of the Revised SAMA Analysis reflect the results of Entergy's updated cost-benefit analyses.'

0 3*The Staff reviewed Entergy's original and revised SAMA implementation cost estimates.

As discussed above, the Staff reviewed the bases for Entergy's cost estimates and also compared the cost estimates to estimates developed previously for similar mitigation measures, including estimates developed as part of other licensees' SAMA analyses.

Where Entergy's original implementation cost estimates appeared high, or revised cost estimates contained its Revised SAMA analysis resulted in certain SAMAs becoming non-cost-beneficial, the Staff requested and obtained from Entergy additional information and justification for the values.'0 4 Based on its review, the Staff found that the cost estimates provided by Entergy are reasonable and 99 See ER, Att. E at E.2-3 to E.2-4, E.4-3 to E.4-4.100 Revised SAMA Analysis at 8-9 [Att. 11]101 See id. at 7, 10-28 (Tables 4 and 5).102 See id at 7-8.103 'See id. at 10-28, 30-3 1.104 FSEIS, Vol. 3, App. G at G-39. For example, the Staff questioned a seemingly high cost estimate ($800,000) for changing the pressurizer PORV block valves from normally closed to normally open in conjunction with IP2 SAMA 53. Entergy provided justification for the cost estimate in a February 2008 RAI response.

As noted above, the Staff also requested that Entergy provide the bases for certain revised cost estimates (including an explicit breakdown of the cost estimate in terms of the major cost factors).

Entergy provided this additional information by letter dated January 14, 2010. See NL-10-013

[Att. 12].

sufficient to support the SAMA evaluation.

1 0 5 *Thus Entergy has supplied "sufficient facts" to allow the Staff to explain its conclusions and to satisfy its "hard look" obligations under*NEPA.106 Insofar as NYS argues otherwise, it misinterprets both the facts of record and controlling legal precedent.

B. Entergy Also Is Entitled to Judgment As a Matter of Law Because the FSEIS Adequately Explains Why Implementation of the Cost-Beneficial SAMAs Is Not Required in Connection With IPEC License Renewal The second issue raised in NYS-35/36 also warrants dismissal as a matter of law. In admitting NYS-35/36, the Board held that, to comply with NEPA, the Staff must explain any.decision not to require implementation of any cost-beneficial SAMAs prior to license renewal. 1 0 7 The Board also described this aspect of the contention as one of omission, because it alleged that the SEIS does not provide a rational basis for granting the license extension without mandating a CLB backfit as a prerequisite for the extension.

1 0 8 But the Board also stated that "the NRC Staff does not have to require implementation, and an intervenor such as New York cannot demand implementation from the NRC Staff as part of a license renewal proceeding."'1 0 9 The Staff has revised Section 5.2.6 of the FSEIS to provide the further explanation alleged to be missing."l 0 Therefore, this aspect of the cOntention is moot and should be dismissed as a matter of law. Section 5.2.6 provides a clear, cogent, and legally correct explanation-and certainly one that is adequate for purposes of NEPA-as to why the Staff has not required implementation of any of the SAMAs identified in the FSEIS as cost-beneficial.

105 FSEIS, Vol. 3, App. G at G-40.106 Indian Point, LBP-1O-13, slip op. at 29.107 Id. at 35.108 See id. at 29-30."09 Id. at 29 (emphasis added)."10 FSEIS, Vol. 1 at 5-11 to 5-12.

The FSEIS explains that the Staff lacks a regulatory basis to require implementation of those SAMAs because they are notfrelated to the aging management requirements of Part 54. it further explains that any additional consideration of those SAMAs for possible implementation would take place under the NRC's Part 50 regulatory framework, inasmuch as the identified plant modifications would affect operations under the current operating license term. 1 1 2 Contrary to NYS's claim, the Staff's explanation is fully consistent with settled legal principles governing the scope of NRC license renewal proceedings.

The NRC's Part 54 regulations are confined to the small number of aging management issues determined by the Commission to be relevant to protecting the public health and safety during the renewal term. 113 The Commission chose to leave all other issues to be addressed by the agency's existing regulatory oversight processes-i.e., outside the deliberately-narrow scope of its license renewal process. 114 Furthermore, the existence or availability of the backfit rule does not alter this fact. The backfit rule is an AEA-derived requirement that, by its terms, centers on protection of the public health and safety. It creates a process, separate and distinct from license renewal, by which the NRC may require safety-based measures after initial licensing.

In the absence of an "adequate protection" or "compliance" exception, 115 the NRC may order a facility backfit only if it is cost-... Id. at,5-11.112 Id 113 See 10 C.F.R. § 54.29; Turkey Point, CLI-01-17, 54 NRC at 7-8.114 See id.115 See 10 C.F.R. § 50.109(a)(4)(i)-(iii).

.justified and achieves "a substantial increase in the overall protection of the public health and safety. or the common defense and security."'"16 The federal courts consistently have interpreted NEPA as a procedural statute that requires disclosure and analysis of environmental impacts, not one that imposes substantive obligations for the protection of the environment.

117 With respect to mitigation analysis, "NEPA not only does not require agencies to discuss any particular mitigation plans that they might put in place, it does not require agencies--or third parties-to effect any.,,l"'

Given that SAMA analysis is a form of NEPA mitigation analysis, there is no requirement that cost-beneficial SAMAs be implemented by the NRC or its licensees."1 9 In this case, the FSEIS fully explains the Staff's SAMA analysis conclusions, including its basis for not requiring implementation of cost-beneficial SAMAs. Therefore, the Staff has met its NEPA obligations:

Here, in a generic EIS the NRC has conducted a thorough NEPA evaluation of the probability and consequences of severe reactor plant accidents, and in plant-specific EISs the NRC staff has discussed at 116 Id. § 50.109(a)(3).

SUMMARY

DISPOSITION OF CONTENTION NYS-35/36 Entergy Nuclear Operations, Inc. ("Entergy")

submits this statement of undisputed material facts in support of its Cross-Motion for Surmmary Disposition of New York State ("NYS") Consolidated Contention 35/36 ("NYS-35/36").'

1. Severe accident mitigation alternatives

("SAMAs"), by definition, pertain to severe accidents; i.e., those accidents whose consequences could be severe, but whose probability of occurrence is so low that they may be excluded from the spectrum of design basis accidents

("DBAs) that have been postulated for a plant. See NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants, Vol. 1, §§ 5.3.2, 5.3.3,.5.4 (May 1996) ("GEIS"), available at ADAMS Accession Nos. ML040690705.

2. Based on the GEIS evaluation of severe accident impacts, 10 C.F.R. Part 51 concludes that the "[t]he probability weighted consequences of atmospheric releases, fallout onto open bodies of water, releases to ground water, and societal and economic impacts from severe accidents are small for all plants." 10 C.F.R. Part 51, Subpart A, App. B, Table B-1 (Postulated Accidents; Severe accidents).

3. GEIS analyses represent adequate, plant-specific estimates of the impacts from severe accidents that would generally over-predict, rather than under-predict, environmental See Applicant's Consolidated Memorandum in Opposition to New York State's Motion for Summary Disposition of Contention NYS-35/36 and in Support of Its Cross-Motion for Summary Disposition (Feb. 3, 2011).

consequences.

See Final Rule: Environmental Review for Renewal of Nuclear Power Plant Operating Licenses, 61 Fed. Reg. 28,467, 28,480 (Juhe 5, 1996), amended by 61 Fed. Reg.66,537 (Dec. 18, 1996) [Att. 6]..4. 10 C.F.R. Part 51 states that if the Staff has not previously considered SAMAs for a license renewal applicant's plant in an EIS or in an environmental assessment, then the applicant must complete an evaluation of alternatives to mitigate severe accidents.

10 C.F.R.§ 51.53(c)(3)(ii)(L);

see also 10 C.F.R. Part 51, Subpart A, App. B, Table B-1.5. The purpose of a SAMA analysis is to identify potential changes to a nuclear power plant, or its operations, that (1) could further reduce the already very low risk of postulated severe reactor accident scenarios, and (2) are cost-beneficial to implement.

It is not a substitute for, and does not represent, the NRC NEPA analysis of potential impacts of severe accidents.

See Final Rule: Environmental Review for Renewal of Nuclear Power Plant Operating Licenses, 61 Fed. Reg. at 28,479-480

[Att. 6]; Entergy Nuclear Generation Co.(Pilgrim.Nuclear Power Station), CLI 11, slip op. at 3, 37-38 (Mar. 26, 2010); See NUREG-1437, Supp. 38, Generic Environmental Impact Statement for License Renewal of Nuclear Plants, Regarding Indian Point Nuclear Generating Unit Nos. 2 and 3, Final Report, Vol. 1 at 5-2 to 5-3, 5-11 (Dec. 2010) ("FSEIS").

6. Entergy submitted its SAMA analysis for Indian Point Units 2 and 3 ("IP2" and"IP3") in April 2007 as part of the Environmental Report ("ER") for the Indian Point Energy Center("IPEC")

license renewal application

("LRA").'

See ER Sec. 4.21 & Att. E (Severe Accident Mitigation Alternatives Analysis), available at http://www.nrc.gov/reactors/operating/

licensing/renewal/applications/indian-point.html#application.

.7. The NRC Staff documented its initial review of Entergy's SAMA in its December 2008 draft Supplemental Environmental Impact Statement

("SEIS") for the IPEC LRA. See NUREG-1437, Supp. 38, Generic Environmental Impact Statement for License Renewal of Nuclear Plants, Regarding Indian Point Nuclear Generating Unit Nos. 2 and 3, Draft Report for Comment (Dec. 2008) ("DSEIS").

8. On December 11, 2009, Entergy submitted a Revised SAMA Analysis that include corrected wind direction inputs. See NL-09-165, Letter from Fred Dacimo, Entergy, to NRC, "License Renewal Application

-SAMA Reanalysis Using Alternate Meteorological Data"

(Dec. 11, 2009) ("NL-09-165" or "Revised SAMA Analysis"), available at ADAMS Accession No. ML093580089

[Att. 11].9. The Nuclear Energy Institute

("NEI") has issued a guidance document, NEI 05-01, Revision A, to assist NRC license renewal applicants in preparing SAMA analyses.

See NEI 05-01, Severe Accident Mitigation Alternatives (SAMA) Analysis, Guidance Document, at i (Rev. A, Nov. 2005), available at ADAMS Accession No. ML060530203

[Att. 7].10. The Staff has approved and recommended the use of NET 05-01 by license.renewal applicants.

See Final License Renewal Interim Staff Guidance LR-ISG-2006-03:

Staff Guidance forPreparing Severe Accident Mitigation Alternatives Analyses (Aug. 2007) ("LR-ISG-2006-03"), available at http://www.nrc.gov/reading-rm/doc-collections/isg/license-renewal.html

[Att. 8 & Att. 9].11. Consistent with the NRC-approved guidance NEI 05-01, Entergy developed cost* estimates for implementing each SAMA candidate to the extent necessary to allow it to make an informed judgment about the economic viability of the proposed improvement.

See ER, Att. E at E.2-73 to E.2-4 & E.4-3 to E.4-4. Entergy followed this same cost-estimating process in its Revised SAMA Analysis.

Revised SAMA Analysis at 8-9 [Att. 11].12. The NRC Staff issued the FSEIS in December 2010. Section 5.2 of the FSEIS.summarized the Staff's final review of Entergy's SAMA evaluation for IP2 and IP3, including its December 2009 Revised SAMA Analysis.

FSEIS, Vol. 1 at 5-4 to 5-13. Appendix G of the FSEIS further documented the Staff's technical evaluation of Entergy's SAMA analysis.

Id.,'Vol. 3, App. G. The FSEIS concluded that "[t]he treatment of SAMA benefits and costs support the general conclusion that the SAMA evaluations performed by. Entergy are reasonable and sufficient for the license renewal submittal." Id., Vol. 1 at 5-11.13. Section G.5 of the FSEIS described the Staff's review of Entergy's cost estimates for implementing SAMAs. FSEIS, Vol. 3, App. G at G-34 to G-40. Section.G.5 explained that Entergy estimated the costs of implementing the candidate SAMAs through the application of engineering judgment and the use of other licensees' estimates for similar improvements.

Table G-6 of the FSEIS listed, among other items, the SAMA implementation cost estimates.

Id at G-36 to G-38. The Staff reviewed the bases for these cost estimates and, for certain improvements, it compared the cost estimates to estimates developed previously for similar improvements

(including estimates developed as part of other licensees' SAMA for operating reactors and advanced light-water reactors).

Id. at G-34 to G-35.14. Section G.5 also stated that, as part of its Revised SAMA Analysis using corrected meteorological data, Entergy subjected a subset of the candidate SAMAs to more.rigorous cost-estimating techniques; i.e., those SAMAs that appeared to be cost-beneficial based on the new benefit estimate and the original implementation cost estimate.

FSEIS, Vol. 3, App.G at G-39. At the request of the NRC Staff, Entergy provided additional information concerning the implementation cost estimates for this subset.of the candidate SAMAs by letter dated January 14, 2010. See NL-10-013, Letter from Fred Dacimo, Entergy, to NRC, "License Renewal Application

-Supplement to SAMA Reanalysis Using Alternate Meteorological Tower Data" (Jan. 14, 2010) ("NL-10-013), available at ADAMS Accession No. ML100260750

[Att. 12].15. The Staff reviewed this additional cost information to determine the degree to which the revised cost estimates and their constituent costs comport with the nature, magnitude and complexity of each change. FSEIS, Vol. 3, App. G at G-39. The Staff determined that..Entergy's revised cost estimates were reasonable, and that they resulted in an appropriate.determination that the subject candidate SAMAs are not cost-beneficial.

Id. at G-39 to G-40.16. The Staff concluded that Entergy's implementation cost estimates for all SAMA candidates are reasonable, generally consistent with prior estimates by other licensees, and sufficient and appropriate for use. in the SAMA evaluation.

FSEIS, Vol. 3, App. G at G-40.17. Section 5.2.6 of the FSEIS summarized the Staff's conclusions with respect to.*Entergy's SAMA analyses for IP2 and IP3, including the methods used and the implementation of those methods. FSEIS, Vol. 1 at 5-11 to 5-12. The Staff concluded that the SAMA evaluations performed by Entergy for iP2 and IP3 "are reasonable and sufficient for the license renewal submittal." FSEIS, Vol. 1 at 5-11.-18. The Staff also concluded that there is no regulatory basis to impose any of the potentially cost-beneficial SAMAs as a condition for license renewal of IP2 and IP3. Section 5.2.6 of the FSEIS explains the basis for this conclusion as follows: Moreover, the NRC staff has determined that none of the potentially cost-beneficial SAMAs are related to the license renewal requirements in 10 CFR Part 54 (i.e., managing the effects of aging) (SEIS § 5.2.6).

Under the NRC's regulatory system, any potentially cost-beneficial SAMAs that do not relate to 10 CFR Part 54 requirements would be considered, to the extent necessary or appropriate, under the agency's oversight of a facility's current operating license in accordance with 10 CFR Part 50 requirements, inasmuch as such matters would pertain not just to the period of extended operation but to operations under the current operating license term as well. Thus, there is no regulatory basis to suggest that potentially cost-beneficial SAMAs that are unrelated to Part 54 requirements must be imposed as a backfit to the CLB, as a condition for license renewal.FSEIS, Vol. 1 at 5-11.19. Section 5.2.6 of the FSEIS further explained that no significant new information has been identified that would remove IP2 and IP3 from the generic determinations made in the GEIS and. 10 C.F.R. Part 51 that severe accidents are the probability of occurrence of severe accidents is so low that they may be excluded from the spectrum of DBAs that have been postulated for a plant (GEIS §§ 5.3.2, 5.3.3, 5.4), and that the probability-weighted radiological consequences of severe.accidents are small for all plants. FSEIS, Vol. 1 at 5-11 to 5-12.20. NYS-35/36 does not challenge any of Entergy's current SAMA implementation cost estimates (whether for cost-beneficial or non-cost-beneficial SAMAs) by alleging that they are inaccurate (e~g., too large) or based on unacceptable methods or assumptions.

21. NYS-35/36 does not allege that any new or additional SAMAs should have been.identified beyond those already identified in the Revised SAMA Analysis and the FSEIS.22. As noted in Section 5.2.6 of the FSEIS, in LBP-10-13, the Board stated that "the NRC Staff does not have to require implementation

[of potentially cost-beneficial SAMAs], and an intervenor such as New York cannot demand implementation from the .NRC Staff as part of a license renewal proceeding." See Entergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 and 3), LBP-10-13, slip op. at 29 (June 30, 2010).

Respectfully submitted, William C. Dennis, Esq.Entergy Nuclear Operations, Inc.440 Hamilton Avenue White Plains, NY 10601 Phone: (914) 272-3202 Fax: (914) 272-3205 E-mail: wdennis@entergy.com Kathryn M. Sutton, Esq.Paul M. Bessette, Esq.Martin J. O'Neill, Esq.MORGAN, LEWIS & BOCKIUS LLP 1111 Pennsylvania Avenue, N.W.Washington, D.C. 20004 Phone: (202) 739-5738 E-mail: ksutton@morganlewis.com E-mail: pbessette@morganlewis.com E-mail: martin.oneill@morganlewis.com COUNSEL FOR ENTERGY NUCLEAR OPERATIONS, INC.Dated in Washington, D.C.this 3rd day of February 2011 DB1/66506255.1 Applicant's Consolidated Memorandum in Opposition to New York State's Motion for Summary Disposition of Contention NYS-35/36 and in Support of Its Cross-Motion for Summary Disposition (February 3, 2011)NYS-35/36:

ATTACHMENT 2

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD)In the Matter of ) Docket Nos. 50-247-LR and 50-286 LR)ENTERGY NUCLEAR OPERATIONS, INC. ) ASLBP No. 07-858-03-LR-BDO1 (Indian Point Nuclear Generating Units 2 and 3 ).) February 3, 2011 APPLICANT'S RESPONSE TO NEW YORK STATE's STATEMENT OF MATERIAL FACTS NOT IN DISPUTE A. Preliminary Statement Pursuant to 10 C.F.R. §§ 2.1205(b) and 2,710(b), Entergy Nuclear Operations, Inc.("Entergy" or "Applicant")

respectfully submits this response to the State of New York's Statement of Material Facts Not in Dispute that it filed in support of its January 14, 2011 Motion for Summary Disposition of Consolidated Contention NYS-35/36.

1 Entergy sets forth its statement-by-statement response to New York State's ("NYS") assertions in correspondingly-numbered paragraphs.

As reflected below, Entergy disputes numerous statements presented by NYS as undisputed statements of material fact. Nonetheless, Entergy respectfully submits that there is no triable issue of fact and that summary disposition of NYS-35/36 should be granted in its favor for the reasons set forth in Entergy's accompanying Consolidated Opposition and Cross-Motion.

2 Like the NYS Motion, NYS's Statement of Material Facts Not in Dispute is based on (1).See The State of New York's Statement of Undisputed Material Facts in Support of Its Motion for Summary Disposition of Consolidated Contention NYS-35/36 (Jan. 13, 2011) ("NYS Motion").2 See Applicant's Consolidated Memorandum in Opposition to New York State's Motion for Summary Disposition of Contention NYS-35/36 and in Support of Its Cross-Motion for Summary Disposition (Feb. 3, 2011) ("Consolidated Opposition and Cross-Motion").

misrepresentations and/or misinterpretations of the facts of record and the governing law, and (2)facts that are not relevant or material to the resolution of NYS-35/36, as admitted by the Board.B. Enterizv's Specific Responses to NYS's Purported Undisputed Material Facts 1. *The Final Supplemental Environmental Impact Statement Regarding Indian Point Nuclear Generating Stations Unit 2 and Unit 3 ("FSEIS,)

identifies 13 potentially cost beneficial severe accident mitigation alternatives

("SAMAs")

for Indian Point Unit 2 ("IP2") and 9 potentially cost-beneficial SAMAs for Indian Point Unit 3 ("1P3"). FSEIS at 5-9 to 5-10. These totals include the two unnumbered SAMAs involving a safety valve gagging device for steam generator tube rupture (SGTR) events. FSEIS at 5-10, G-48 (each for IP2 and IP3).Response:

Undisputed as a description of the contents of the referenced portions of the FSEIS.2. The FSEIS notes that the cost-benefitanalyses have not been completed for these 22 SAMAs and that further analysis is required.

FSEIS at 5-11.Response:

Disputed.

NYS incorrectly describes the contents of the FSEIS. The FSEIS does not state that Entergy'.s cost-benefit analyses are incomplete or that "further analysis" is "required." Rather, page 5-11 of the FSEIS states, in relevant part: The NRC staff reviewed Entergy's analysis, as revised, and concludes that the methods used, and the implementation of those methods, were sound. The treatment of SAMA benefits and costs support the general conclusion that the SAMA evaluations performed by Entergy are reasonable and sufficient for the license renewal submittal

....Based on its review of the SAMA analysis, as revised, the staff concurs with Entergy's identification of areas in which risk can be further reduced in a cost-beneficial manner through the implementation of all *or a subset of potentially cost-beneficial SAMAs. Given the potential for cost-beneficial risk reduction, the staff considers that further evaluation of these SAMAs by Entergy is appropriate.

However, none of the potentially cost-beneficial SAMAs relate to adequately managing the effects of aging during the period of extended operation.

Therefore, they need not be implemented as part of IP2 and IP3 license renewal pursuant to 10. CFR Part 54.Under the NRC's regulatory system, any potentially cost-beneficial SAMAs that do not relate to 10 CFR Part 54 requirements would be considered, to the extent necessary or appropriate, under the agency's oversight of a facility's current operating license in accordance with 10 CFR Part 50 requirements, inasmuch as such matters would pertain not just to the period of extended operation but to operations under the current operating license term as well. Thus, there is no regulatory basis to suggest that' potentially cost-beneficial SAMAs that are unrelated to Part 54 requirements must be imposed as a backfit to the CLB, as a condition for license renewal.FSEIS, Vol. 1 at 5-11. As the quoted FSEIS passage indicates, neither the. NRC Staff nor Entergy has stated that Entergy's SAMA analyses, including its present cost estimates, are incomplete for purposes ofthe submission required by NEPA or 10 C.F.R. Part 51.3. in the original NYS Contention 35, New York State identified 9 potentially cost-beneficial SAMAs for which final cost analyses had not been completed.

These 9 SAMAS are: IP2 009 Create a reactor cavity flooding system;IP2 021. Install additional pressure or leak monitoring instrumentation for interfacing system loss of coolant accidents (ISLOCAs);

IP2 022 Add redundant and diverse limit switches to each containment isolation valve;.IP2 062 Provide a hard-wired connection to a safety injection (SI) pump from the alternate safe shutdown system (ASSS) power supply;IP2 053 Keep both pressurizer PORV block valves open;IP3 007 Create a reactor cavity flooding system;IP3 018 Route the discharge from the main steam safety valves through a structure where awater spray would condense the steam and remove most of the fission products (cost beneficial in TI-SGTR sensitivity in Section [8]);IP3 019 Install additional pressure or leak monitoring instrumentation for ISLOCAs; IP3 053 Install an excess flow valve to reduce the risk associated with hydrogen explosions.

Entergy December 2009 SAMA Reanalysis, December 11, 2009 Letter from Fred Dacimo to U.S. Nuclear Regulatory Commission,

Subject:

License Renewal Application

-SAMA Reanalysis Using Alternate Meteorological Tower Data, NL-09-165, Attachment 1, at 10-28 (Tables 4 & 5), ML093580089

("December 2009 SAMA Reanalysis")

Response:

Disputed to the extent that Paragraph 3 states that Entergy has not completed "final cost analyses" for any of the SAMAs identified in its Environmental Report and the NRC Staff s FSEIS, including the nine SAMAs identified in Paragraph

3. See Response to Paragraph 2 above. Undisputed to the extent that the nine potentially cost-beneficial SAMAs listed in Paragraph 3 are the. nine potentially cost-beneficial SAMAs identified in NYS-35.4. With regard to the 9 SAMAs identified in ¶ 3, supra, Entergy indicated that "the above potentially cost beneficial SAMAs have been submitted for engineering projectcost-benefit analysis." December 2009 SAMA Reanalysis, Attachment 1, at 31-32.Response:

Undisputed to the .extent that Paragraph 4 quotes a statement contained on page 32 of Entergy's December 2009 Revised SAMA Analysis.

See NL-09-165, Letter from Fred Dacimo, Entergy, to NRC, "License Renewal Application

-SAMA Reanalysis Using, Alternate Meteorological Data" at 31-32 (Dec. 11, 2009) ("NL-09-165), available at ADAMS Accession No. ML093580089

("Revised SAMA Analysis")

[Att. 11]. Disputed insofar as the quoted statement

("the above potentially cost beneficial SAMAs have been submitted for engineering project cost-benefit analysis")

actually applies to all SAMAs identified as potentially cost-beneficial in Entergy's December 2009 Revised SAMA Analysis, not just to the nine SAMAs identified by NYS in Paragraph 3 above.

5. Entergy has not submitted a completed cost-benefit analysis for any of these 9 SAMAs to the NRC Staff.Response:

Disputed.

See Responses to Paragraphs 2 and 3 above.6. The FSEIS accepts the incomplete cost-benefit analyses as "sound" and concludes that the "evaluations'performed by Entergy are reasonable and sufficient for the license renewal submittal." FSEIS at 5-11.Response:

Disputed to the extent that Paragraph 6 states or implies that the cost-benefit analyses contained in Entergy's SAMA analyses for IP2 and IP3 are "incomplete." See Response to Paragraph 2 above. Undisputed to the extent that Paragraph 6 partially quotes text from the FSEIS, the relevant portion of which states in full: The NRC staff reviewed Entergy's analysis, as revised, and concludes that the methods used, and the implementation of those methods, were sound. The treatment of SAMA benefits and costs support the general conclusion that the SAMA evaluations performed by Entergy are reasonable and sufficient for the license renewal submittal.

FSEIS, Vol. 1 at 5-11.7. In the original NYS Contention 36, New York State identified 9 additional potentially cost-beneficial SAMAs where the benefits were substantially greater than the cost and contended that either the SAMA should be required to be implemented or a rational basis should be given for why it should not be implemented.

The following table accurately summarizes the findings in the December 2009 SAMA Reanalysis regarding these 9 substantially cost-effective SAMAs for IP2 and IP3:

Original Original Baseline New Baseline Baseline New Baseline Benefit with Ienefit with SAMA Numter and Description Benefit Blenefit Uncertainty' Uncertainly Old Cost New Cost T2 $420,459 $1,357,046

$885,176 $2,856,939

$494,000 $938,000 SAMA 028: Provide a portable dieel-drivcn battery charger.WP2 $984.503 S2,350.530 12,072,638

$4,948.485

$1,656,00)0

$16,0M0 SAteIA 044: llse.r.re water system as backup for steam generator inventory.

IP2 $1,722,733 S5,591,781

$3,626,807

$11,772,170 S200,000 1200,000 SAMA 054: Install flood alarm in the 480VAC switchgear room.1P2 $387,821 $1.27.5,337

$816.,181

$2684,920

$216,000 S216,000 SAMA 060: Proyide added protection against flood propagation from stairwell 4 into the 480VAC switehgear room.tP2 S853,187 $2,734,991

$1,796,183

$5,799,982

$192,000 $1982000 SAMA 06 1: Provide added protection against flood propagation from the deluge room into the 480V switchgear room, 1I>2 $1.722,733 S.59t.781

$3,626,807

$11,772,170

$560,000 $560,000 SAMA 065: Upgrade the ASSS to allow timely restoration of seal injection and cooling-IP3

$4,073,152

$1,847,657

$5,903,118

$1.288,000

$1,288,000 SAMA 055: Provide hardwired connection wt one St or RHR pump Irm the Appendix R bus (MCC 312A)-IP3 $1,365,046 54,359,371

$1,978,321

$6,317,929

$560,000 1560,000 SAMA 061 Upgrade the A3S3 to allow timely restoration of seal injection and cooling.193 S 1,365,N46 S4,359,371

$1,978,328

$6,317,929

$196,800 1196,80K SAMA 062: Install flood alarm in the 480VAC switchgear room.Response:

Disputed because "substantially greater than" and "substantially cost-effective" are undefined and ambiguous.

Disputed to the extent that Tables 4 and 5 of Entergy's December 2009 Revised SAMA Analysis dd not contain entries for "Original Baseline Benefit,""Original Baseline Benefit with Uncertainty," and "Old Cost." Undisputed to the extent that the numerical values reported in the third, fifth,- and seventh columns of the Table above ("New Baseline Benefit," "New Baseline Benefit with Uncertainty," and "New Cost," respectively) coincide with the values listed in Tables 4 and 5 of the Revised SAMA Analysis (in the columns titled *"Benefit," "Benefit with Uncertainty," and "Estimated Cost," respectively).

8. The Staff has not made a decision on the merits on whether any of the cost-effective SAMAs identified in ¶ 7, supra, should be added as license conditions for IP2 or IP3 and Entergy has not committed to adopting any of these SAMAs.

Response:

Disputed.

The FSEIS explicitly states: "[N]one of the potentially cost-beneficial SAMAs relate to adequately managing the effects of aging during the period of extended operation.

Therefore, they need not be implemented as part of 1P2 and IP3 license renewalpursuant to 10 CFR Part 54." FSEIS, Vol. 1 at 5-11 (emphasis added). The FSEIS further explains that the NRC lacks regulatory authority to impose any of the potentially cost-beneficial SAMAs as a condition for license reneWal of IP2 and IP3. See id. at 5-11 to 5-12.Also disputed to the extent that Paragraph 8 suggests that Entergy has a,legal obligation to"commit to adopt" any SAMA as part of the license renewal process.9. The FSEIS does not includeany additional cost-benefit analyses beyond those provided by Entergy in its December 2009 SAMA Reanalysis and early submittals.

Response:

Disputed to the extent that, at the request of the NRC Staff, Entergy provided additional information concerning the implementation cost estimates forcertain SAMAs by letter dated January 14, 2010. See NL-10-013, Letter from Fred Dacimo, Entergy, to NRC, "License Renewal Application

-Supplement to SAMA Reanalysis Using Alternate Meteorological Tower Data" (Jan. 14, 2010) ("NL-10-013), available at ADAMS Accession No.ML100260750

[Att. 12]. This information is discussed on pages G-39 to G-40 of Appendix G to the FSEIS (Vol. 3). Also disputed to the extent Paragraph 9 implies that Entergy's SAMA analyses for IP2 and IP3 or the FSEIS contain incomplete cost-benefit analyses.

See Entergy's Responses to Paragraph 2 and 3 above.10. The FSEIS concludes that completed cost-benefit analyses for the 18 above-identified potentially cost-effective SAMAs and a decision on the merits as to whether any of these potentially cost-effective SAMAs should be included as license conditions for the proposed extended operating licenses for IP2 or IP3 need not be made prior to the Atomic Safety and Licensing Board reaching a decision on the proposed license renewal applications for IP2 and IP3. FEIS at 5-11 to 5-12.Response:

Disputed for the same reasons set forth in Entergy's Responses to Paragraph 2 and 8 above. In short, the FSEIS does not state that Entergy's cost-benefit analyses, including its associated SAMA implementation cost estimates for the 18 potentially cost-beneficial SAMAs identified in NYS-35/36, are incomplete.

In addition, the FSEIS explicitly states that, because none of the potentially cost-beneficial SAMAs relate to adequately managing the effects of aging during the period of extended operation, "they need not be implemented as part of IP2 and IP3 license renewal pursuant to 10 CFR Part 54." FSEIS, Vol. 1 at 5-11 11. The entire basis for NRC Staff s decision to not require completion of the cost analyses for all potentially cost-effective SAMAs and to not require implementation of any of the potentially cost-effective SAMAs is contained in the FSEIS ¶ 5.2.6. FSEIS at pp. 5-11 to 5-12.Response:

Disputed because the phrase "entire basis" is ambiguous.

Also disputed because the basis for the Staff s conclusion in the FSEIS is supported by and consistent with long-standing judicial and/or Commission interpretations of NEPA and the NRC's license renewal regulations (i.e., 10 C.F.R. Parts 54 and 51 and their associated regulatory histories).

Also disputed to the extent that Paragraph 11 suggests that Entergy's "cost analyses for all potentially cost-effective SAMAs" are not complete for purposes of the submission required by NEPA or 10 C.F.R. Part 51. See Entergy's Responses to Paragraphs 2 and 3 above.12. The Indian Point reactors were never subjected to a SAMA analysis as part of the NEPA review during the proceedings for their initial operating licenses during the 1970s. FSEIS at 5-4, § 5.2.

Response:

Undisputed insofar as the cited page of the FSEIS states that "SAMAs have not been previously considered for IP2 and IP3." FSEIS, Vol. 1 at 5-4.13. According to NRC Staff and Entergy, as of 2000 approximately 16,971,000, people live within 50 miles of IP2 and IP3. FSEIS at 2-124.Response:

Undisputed to the extent that the FSEIS states: "Approximately 16,791,654 people live within 50 mi (80 km) of IP2 and IP3." FSEIS, Vol. 1 at 2-124.Disputed as the relevance or materiality of Paragraph 13 to the resolution of NYS-35/36, as admitted bythe Board.14. According to NRC Staff and Entergy, as of 2000 approximately 1,113,000 people live within 20 miles of IP2 and IP3. FSEIS at 2-124.Response:

Undisputed to the extent that the FSEIS states: "According to the 2000 census, approximately 1,1 13,089 people lived within 20 mi (32 km) of IP2 and IP3...." FSEIS, Vol. 1 at 2-124. Disputed as the relevance or materiality of Paragraph 14 to the resolution of NYS-35/36, as admitted by the Board.15. According to NRC Staff, IP2 and IP3 each are located in a high-population area.FSEIS at 2-124.Response:

Undisputed to the extent that the FSEIS states that the population density within 20 mi (32 km) of IP2 and IP3 is 886 persons per sq mi (332 persons per 6 sq km), andthat"[t]his density translates to the least sparse Category 4 (greater than or equal to 120 persons per square mile Within 20 mi)." FSEIS, VoL 1 at 2-124. Disputed as the relevance or materiality of Paragraph 15 to the resolution of NYS-35/36, as admitted by the Board.16. Entergy and NRC Staff project that by 2035 approximately 19,228,000 people will be within 50 miles of IP2 and IP3. FSEIS at G-20.

Response:

Undisputed to the extent that the FSEIS states that Entergy used a 2035 projected population value.of 19,228,712 in its SAMA analysis.

FSEIS, Vol. 3,_App. G at G-25, G-28, Disputed as the relevance or materiality of Paragraph 16 to the resolution of NYS-35/36, as admitted by the Board.17. Of all the power reactors in the United States, the Indian Point reactors have the highest surrounding population both within a 50-mile radius and a 10-mile radius. NYS Contention 35, ¶¶ 10, 12; NYS Contention 36, ¶ 19; AEC, Population Distribution Around Nuclear Power Plant Sites, Figure 2: Typical Site Population Distribution (5-50 Miles) (April 17, 1973); FEMA, Nuclear Facilities

& Population Density Within 10 Miles (June 2005).Response:

Disputed as the relevance or materiality of Paragraph 17 to the resolution of admitted contention NYS-35/36.

With respect to the contents of the cited documents, the documents speak for themselves.

William C. Dennis, Esq.Entergy Nuclear Operations, Inc.440 Hamilton Avenue White Plains, NY 10601 Phone: (914) 272-3202 Fax: (914) 272-3205 E-mail: wdennis@entergy.com Respectfully submitted, Kathryn M. Sutton, Esq.Paul M. Bessette, Esq.Martin J. O'Neill, Esq.MORGAN, LEWIS & BOCKIUS LLP 1111 Pennsylvania Avenue, N.W.Washington, D.C. 20004 Phone: (202) 739-5738 E-mail: ksutton@morganlewis.com E-mail: pbessette@morganlewis.com E-mail: martin.oneill@morganlewis.com COUNSEL FOR ENTERGY NUCLEAR OPERATIONS, INC.Dated in Washington, D.C.this 3rd day of February 2011 Applicant's Consolidated Memorandum in Opposition to New York State's Motion for Summary Disposition of Contention NYS-35/36 and in Support of Its Cross-Motion for Summary Disposition (February 3, 2011)NYS-35/36:

ATTACHMENT 3

NUREG-1437 Vol. 1 Generic Environmental Impact Statement for License Renewal of Nuclear Plants Main Report Final Report Manuscript Completed:

April 1996 Date Published:

May 1996 Division of Regulatory Applications Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 ENVIRONMENTAL IMPACTS OF ACCIDENTS site-specific consideration of severe accident mitigation for license renewal will only identify procedural and programmatic improvements (and perhaps minor hardware changes) as being cost-beneficial in reducing severe accident risk or consequence.

Therefore, a site-specific consideration of alternatives to mitigate severe accidents shall be performed for license renewal unless. such a consideration has, already been included in a previous EIS or related supplement.

Staff evaluations of alternatives to mitigate severe accidents have already been completed and included in an EIS or supplement for Limerick, Comanche 'Peak, and Watts Bar; therefore, severe accident mitigation need not be reassessed for these plants for license renewal.5.5

SUMMARY

AND CONCLUSIONS The foregoing discussions have dealt with.the environmental impacts of accidents during operation after license renewal. The primary assumption for this evaluation is that the frequency (or likelihood of occurrence) of an accident at a given plant would not increase during the plant lifetime (inclusive of the license renewal period) because regulatory controls ensure the plant's licensing basis is maintained and improved, where warranted.

However, it was recognized that the changing environment around the plant is not subject to regulatory controls and introduces the potential for changing risk.Estimation of future 9evere accident consequences and risk was based upon existing risk and consequence analyses found in FES for recently licensed plants because these include severe accident analyses and constitute a representative set of plants and sites for the United States.5.5.1 Impacts from Design-Basis Accidents The envi ronmental impacts of postulated accidents were evaluated for the license renewal period in GEIS Chapter 5. All plants have had a previous evaluation of the environmental impacts of design-basis accidents.

In addition, the licensee will be required to maintain acceptable design and performance criteria throughout the renewal period. Therefore, the calculated releases from design-basis accidents would not be expected to change. Since the consequences of these events are evaluated for the hypothetical maximally exposed.individual at the time of licensing, changes in the plant environment will not affect these evaluations.

Therefore, the staff concludes that the environmental impacts of design-basis accidents are of small significance for all plants. Because the environmental impacts of design basis accidents are of small significance and because additional measures to reduce, such impacts, would be costly, the staff concludes that no mitigation measures beyond those implemented during the current term license would be warranted.

This is a Category 1 issue.5.5..2 Impacts from Severe Accidents 5.5.2-1 Atmospheric Releases The evaluation of health and dose effects caused by atmospheric releases used a prediction process to identify those plant sites that are bounded by existing analyses.Existing analyses represent only a subset of operating plants. A particular portion of this subset, specifically those plants having severe accident analyses in their respective FESs, was used in this evaluation.

El (which is a function of population and wind direction), in conjunction with the FES severe accident analyses, was then used to develop a means to predict NUREG-1437, Vol. 1 5-114 NUREG-1437, Vol. 1 5-114 ENVIRONMENTAL IMPACTS OF ACCIDENTS consequences for all plants. Average values and 95 percent UCB values were estimated.

Table 5.6 provides the results of this prediction process.Results indicate that the predicted effects of a severe accident during MYR at the 74 sites of nuclear power plants in the United States are not expected to exceed a small fraction of that risk to which. the population is already exposed. In addition, the dose to individuals was also predicted.

Results indicate that the highest average individual, dose would be 3 x 10-4 rem/RY.This dose compares to an average of 3 x 10'1 rem/person/year for all other causes, including radon. Therefore, the probability-weighted consequences from atmospheric releases associated with severe accidents is judged to be of small significance for all plants.5.5.2.2 Fallout onto Open Bodies of Water The results of comparative analyses for the drinking-water pathway concluded .that Great Lakes sites have the same order-of-magnitude risk that was calculated in the Fermi 2 FES, which is only a small fraction of the risk from atmospheric pathway releases.

River sites with potentially greater risk than in the Fermi FES are amenable to interdiction, which can significantly reduce risk. In the case of the aquatic food pathway, interdicted jopulation exposures are less than or essentially the same as atmospheric pathway releases.

For both the drinking water and aquatic food pathways, the probability-weighted consequences from fallout due to severe accidents is of small significance.

5.5.2.3 Releases from Groundwater The comparative analyses for this pathway were done by first segregating all sites into six general categories as called out in the NRC LPGS (NUREG-0440) and then estimating if the risk consequences calculated in existing analyses (including the LPGS) bounds the risks for all other plants within each category.Of the six categories, three are judged to be bound by existing analyses.

These categories are Great Lake sites, estuaries, and dry sites.For the other categories, estimates were made of the degree to which groundwater releases could exceed existing analyses.

For all six categories, the staff concluded that the risk to the population was either a small fraction of that for atmospheric releases or, in a few cases,. comparable to that from atmospheric releases.

Therefore, the probability-weighted consequences from groundwater releases due to severe accidents is judged to be of small significance for all plants.5.5.2.4 Societal and Economic Risks The expected costs resulting from a severe accident at nuclear power plants during their renewal periods have been predicted from evaluations presented in 27 FESs.Estimates of the extent of land contamination have also been presented.

In both cases, the conditional impacts are judged to be of small significance for all plants.5.5.2.5 SAMDAs The staff concluded that the generic analysis summarized above applies to all plants and that the probability-weighted consequences of atmospheric releases, fallout onto open bodies of water, releases to ground water, and societal and economic impacts of severe accidents are of small significance for all plants. However, not all 5-15N-EG147,Vol 5-115 NUREG-1437, Vol. I ENVIRONMENTAL IMPACTS OF ACCIDENTS plants have performed a site-specific analysis of measures that could mitigate severe accidents.

Consequently, severe accidents .are a Category 2 issue for plants that have not performed a site-specific consideration of severe accident mitigation and submitted that analysis for Commission review.5.6 ENDNOTES 1. While a dose as low as 10 rem may cause such observable physiological changes as chromosomal aberrations, these changes are not classified as, clinical injury.2. Also referred to as the Rogovin report.3. Grand Gulf, Sequoyah, Surry, Peach.Bottom, and Zion.4. The FitzPatrick and Nine Mile Point units are located closely enough to assume that they are located on the same site. A similar observation can be made for the Hope Creek and Salem units.5. Because the hypothetical sites were to be modeled as either PWRs or BWRs, those using population data of actual PWR sites utilized updated WASH-1400 source terms taken from the Byron FES (NUREG-0848), while those using population data for BWRs utilized updated WASH-1400 source terms taken from the Clinton FES (NUREG-0854).

5.7 REFERENCES

Acharya, S., and R. M. Blond, Transcript of Testimony Before the Atomic Safety and Licensing Board in the*Matter of Consolidated Edison of New York (Indian Point, Unit 2) and Power Authority of the State of New.York (Indian Point, Unit 3) (Docket Nos. 50-247-SP and 50-286-SP), Sections III.C.A and III.C.B, n.d.BEIR-I, The Effects on Populations of Exposure to Low Levels of Ionizing Radiations, National Research Council, Advisory Committee on the Biological Effects of Ionizing Radiation, National Academy of Sciences, Washington, D.C., 1972.BEIR-III, The Effects on Populations of Exposure to Low Levels of Ionizing Radiation:

1980, National Research Council, Advisory Committee on the Biological Effects of Ionizing Radiation, National Academy of Sciences, Washington, D.C., 1980.BEIR-V, Health Effects of Exposure to Low Levels of Ionizing Radiation, National* Research Council, Advisory Committee on the Biological Effects of Ionizing Radiation, National Academy of Sciences, Washington, D.C., 1990.Bertini, H. W., Descriptions of Selected Accidents That Have Occurred at Nuclear Reactor Facilities, ORNL/NSIC-176, Oak Ridge National Laboratory, Oak Ridge, Tennessee, April 1980.Codell, R. B., "Potential Contamination of Surface Water Supplies by Atmospheric Releases from Nuclear Plants," Health Physics 49(5), 713-30, November 1985.ConEd (Consolidated Edison Company), Indian Point Nuclear Power Plant, Unit 2 Final Safety Analysis Report, June 1982.Crick, M. J., and G. S. Linsley, An Assessment of the Radiological Impact of the Windscale Reactor Fire, National.Radiological Protection Board, Chilton, United Kingdom, 1983.DOE/ER-0332, M. Goldman et al., Health and Environmental Consequences of the Chernobyl Nuclear Power Accident, U.S. Department of Energy, Washington, D.C., 1987.NUREG.1437, Vol. 1 5-116 NUREG-14-37, Vol. 1 5-116 Applicant's Consolidated Memorandum in Opposition to New York State's Motion for Summary Disposition of Contention NYS-35/36 and in Support of Its Cross-Motion for Summary Disposition (February 3, 2011)NYS-35/36:

ATTACHMENT 4

Federal Register / Vol. 60, No. 88 / Monday, May 8, 1995 / Rules and Regulations

.22461 require the approval of OMB under 44 U.S.C. 3501, et seq.Comments A general description of the statutory basis for this final rule was set forth in the interim rule published on September 16, 1994, (59 FR 4753.0). The interim rule provided 60 days for comments.

No comments were received during the interim rule comment period of September 16 through November 15, 1994. This final rule provides that in determining net proceeds for shorn wool or mohair, effective for 1993 and subsequent marketing years, marketing charges for commissions, coring, or grading shall not be deducted.

This rule provides atithorized representatives of USDA and CCC access to the premises of buyers and sellers of wool and mohair in order to inspect their records for authenticity:

This provision had been accidentally omitted when the wool regulations and mohair regulations were combined in 1991. This final rule also clarifies the definition of nonmarketing charges to make it consistent with the calculation of net proceeds and net proceeds for payment purposes.Section 1468.18(d) was inadvertently omitted from the interim rule. This provision was accidently omitted when the mohair regulations and the wool regulations were combined in 1991 (56 FR 40233, August 14, 1991). This final rule, in part, merely reinstates the omitted provision.

List of Subjects in 7 CFR Part 1466 Grant program-agriculture, Livestock, Mohair, Reporting and recordkeeping, Wool.Accordingly, the interim rule amending 7 CFR part 1468 published on September 16, 1994, (59 FR 47530) is adopted as final with the following changes: PART 1468-WOOL AND MOHAIR 1. The authority citation for 7 CFR part 1468 continues to read as follows: Authority:

7 U.S.C. 1781-1787; 15 U.S.C.714b and 714c.2. In § 1468.3 the definition of"Nonmarketing charges" is revised to read as follows:§ 1468.3 Definitions.

Nonmarketing charges means charges paid by or for the account of the producer that are not directly related to improving the marketability of the shorn wool or mohair, such as, but not limited to, storage bags, advances, interest on advances, shearing, and association dues, and are not deducted from the producer's gross proceeds to determine net proceeds for payment purposes and are deducted from gross proceeds to determine net proceeds.ft f * * *3. Section 1468.18 is amended by adding paragraph (d) to read as follows:§ 1468.18 Maintenance and Inspection of records.(d) At all times during regular business hours, authorized representatives of CCC or USDA shall have access to the premises of the applicant, of the marketing agency, and of the person who furnished evidence to an applicant for use in connection with the application, in order to inspect,.examine, and make copies of the books, records, and accounts, and other written data as specified in paragraphs (a), (b), and (c) of this section.Signed at Washington, DC, on May 1, 1995.Bruce R. Weber, Acting Executive Vice President, Commodity Credit Corporation, IFR Doc. 95-11180 Filed 5-5-95; 8:45 aml BILLING CODE 3410-05--M NUCLEAR REGULATORY COMMISSION 10 CFR Parts 2,51, and 54 RIN 3150-AF05 Nuclear Power Plant License Renewal;Revisions AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule.SUPPLEMENTARY INFORMATION:

I. Background.

II. Final Action.Ill. Principal Issues.a. Continued validity of certain findings in previous rulemaking.

b. Reaffirmation of the regulatory philosophy and approach and clarification of the two principles of license renewal.c. Systems, structures, and Components within the scope of license renewal.d. The regulatory process andaging management.

e. Reaffirmation of conclusions concerning the current licensing basis and maintaining the function of systems, structures, and components.

f. Integrated plant assessment.

g. Time-limited aging analyses and exemptions.

h. Standards for issuance of a renewed licefise and the scope of hearings.i. Regulatory and administrative controls.IV. General Comments and Responses.

V. Public Responses to Specific Questions.

VI. Availability of Documents.

VII. Finding of No Significant Environmental Impact: Availability.

VIII. Paperwork Reduction Act Statement.

IX. Regulatory Analysis.X. Regulatory Flexibility Act Certification.

XI, Non-Applicability of the Backfit Rule.I. Background The previous license renewal rule (10 CFR Part 54) was adopted by the Nuclear Regulatory Commission (NRC)on December 13, 1991 (56 FR 64943).This rule established the procedures, criteria,-and standards governing the renewal of nuclear power plant operating licenses.Since publishing the previous license renewal rule, the NRC staff has conducted various activities related to implementing this rule. These activities included:

developing a draft regulatory guide, developing a draft standard review plan for license renewal, interacting with lead plant licensees, and reviewing generic industry technical reports sponsored by the Nuclear Management and Resources Council (now part of the Nuclear Energy Institute (NEI)).In November 1992, the law firm of Shaw, Pittman, Potts, and Trowbridge submitted a paper to the NRC that presented the perspective of Northern States Power Company on the license renewal process. The paper included specific recommendations for making the license renewal process more workable.

In addition, industry representatives provided the Commission with views on several key license renewal implementation issues.In late 1992, the NRC staff conducted a senior management review and discussed key license renewal issues with the Commission, industry groups,

SUMMARY

The Nuclear Regulatory Commission (NRC) has amended its regulations to revise the requirements that an applicant must meet for obtaining the renewal of a nuclear power plant operating license. The rule also clarifies the required information that must be submitted for review so that the agency can determine whether those requirements have been met and changes the administrative requirements that a holder of a renewed license must meet. These amendments are intended to provide a more stable and predictable regulatory process for license renewal.EFFECTIVE DATE: June 7, 1995.FOR FURTHER INFORMATION CONTACT: Thomas G. Hiltz, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone:

(301) 415-1105.HeinOnline

-- 60 Fed. Reg. 22461 1995 22462 Federal Register / Vol. 60, No. 88 / Monday, May 8. 1995 / Rules and Regulations and individual licensees.

The NRC staff presented its recommendations regarding several of these key license renewal issues in two Commission policy papers: SECY-93-049." "Implementation of 10 CFR Part 54,'Requirements for Renewal of Operating Licenses for Nuclear Power Plants,'and SECY-93-113. "Additional Implementation Information for 10 CFR Part 54, 'Requirements for Renewal of Operating Licenses for Nuclear Power Plants."'In its staff requirements memorandum (SRM) of June 28, 1993, the Commission stated that it is essential to have a predictable and stable regulatory process Clearly and unequivocally defining the Commission's expectations for license renewal. This process would permit licensees to make decisions about license renewal without being influenced by a regulatory process that is perceived to be uncertain, unstable, or not clearly defined. The Commission directed the NRC staff to convene a public workshop to evaluate alternative approaches for license renewal that best take advantage of existing licensee activities and programs as a basis for concluding that aging will be addressed in an acceptable manner during the period of extended operation.

In particular, the Commission directed the NRC staff to examine the extent to which greater-reliance can be placed on the maintenance rule (10 CFR 50.65,"Requirements for Monitoring the Effectiveness of Maintenance at Nuclear.Power Plants") as a basis for concluding that the effects of aging will be effectively managed during the license renewal term.On September 30, 1993, the NRC staff conducted a public workshop in Bethesda, Maryland, that was attended by over 180 people. Attendees included nuclear utilities, industry organizations, public interest groups, architect and engineering firms, consultants and contractors, and Federal and State governments.

In December 1993, the NRC staff forwarded SECY-93-331,"License Rendwal Workshop Results and Staff Proposals for Revision to 10 CFR Part 54. 'Requirements for Renewal of Operating Licenses for Nuclear Power Plants,"'

to the Commission.

The NRC staff recommended that the Commission amend 10 CFR Part 54.In its SRM of February 3, 1994, the Commission agreed with the NRC staff's conceptual approach (explained in SECY-93-331) for performing license renewal reviews and directed the staff to proceed with rulemaking to'amend 10 CFR Part 54. The Commission believes that the license renewal process should focus on the management of the effects of aging on certain systems. structures, and components during the period of extended operation.

An objective for the amendment is to establish a more stable.and.predictable license renewal process.The amendment will identify certain systems, structures, and components' that require review in order to provide the necessary assurance that they will continue to perform their intended function for the period of extended operation.

On May 23, 1994, the NRC staff.provided the Commission with its proposed amendment to the license renewal rule in SECY-94-140,."Proposed Amendment to the Nuclear Power Plant License Renewal Rule (10 CFR Part 54)." In the SRM of June 24, 1994, the Commission approved the pdblication of the proposed rule amendment for a 90-day public comment period. In the SRM, the Commission directed the staff to (1)ensure consistency in the use of thb terms "structures, systems, and components" and "structures and components," (2) solicit comments on the ability of existing programs to detect failures in redundant structures and components before there is a loss of intended system or structure function, (3) address the need for § 54.4(a)(3) in the statements of consideration for the proposed rule, and (4) review the necessity of retaining

§ 54.4(a)(4) and include the rationale for its conclusions in the proposed rule.On September 9, 1994, (59 FR 46574)the'proposed revisions to the license renewal rule were published in the Federal Register for a 90-day public comment period. The public comment period ended on December 9, 1994. The I Throughout the Statement of Considerations, the phrases. "systems.

structures, and compon.ents" and "structures and components" are used& As a matter of clarification.

the Commission intends that the phrase. "systems.

structures, and components" applies to the matters involving the discussions of the overall renewal review, the specific license renewal scope (§ 54.4), time-limited aging analyses (§54.21(c)).

and the license renewal finding (§ 54.29). The phrase. "structures and components" applies to matters involving the Integrated plant assessment (IPA) required by § 54.21(a) because the aging management review required within the IPA should be a component and structure level review rather than a more general system level review. The phrase systems, structures, and components applies to the evaluation of time-limited aging analyses* required by § 54.21(c) because such plant-spe'cific analyses may have been carried out, for the initial* operating term, for either systems, structures, or components.

Reevaluation for the renewal term is intended to focus on the same systems, structures.

or components subject to the initial term time-limited aging analyses.

The finding required by§ 54.29 considers both the results of the integrated plant assessment and the time-limited aging analyses and. therefore, the phrase system, structures, and components is applicable to this section.Commission received 42 separate responses concerning the proposed rulemaking for license renewal. In early April 1995. after reviewing SECY-95-067, "Final Amendment to the Nuclear Power Plant License Renewal Rule (10 CFR Part 54)," the Nuclear Energy Institute and Yankee Atomic Electric Company provided additional comments.

All comments received have been considered in developing this final rule.Comments on the proposed rule came from a variety of sources. These included:

a private citizen, 3 public interest groups (Sierra Club-Atlantic.

Chapter. Public Citizen, and the Ohio Citizens for Responsible Energy Inc.). I Federal organization (Department of Energy (DOE)), 4 State organizations (Illinois Department of Nuclear Safety (Illinois), Connecticut Department of Public Utility Control (Connecticut), New Jersey Department of Environmental Protection (New Jersey).and Nevada Agency for Nuclear Projects, Nuclear Waste Project Office (Nevada)), 2 industry organizations (NEI and Nuclear Utility Group on Equipment Qualification (NUGEQ)), 2 vendor owners groups (Babcock and Wilcox (B & W) Owners Group and Westinghouse Owners Group), 2 vendors/consultants (B & W Nuclear Technologies and Westinghouse Electric Corporation), and 27 separate nuclear power plant licensees.

All 27 licensees endorsed the comments provided by NEI, and some utilities also provided additional comments.The Commission specifically solicited responses to five questions in the proposed rule. The questions and the responses to them can be found In Section V of the Supplementary Information also known as the Statement of Considerations (SOC).* Many of the letters contained similar comments, which were grouped together and are addressed on an issue basis. The NRC has responded to all of the significant points raised by the commenters.

Those comments that are applicable to a specific issue discussed in a specific section of the Supplementary Information portion of this document are discussed within that section. Comments received that are not responsive to a particular issue are addressed in Section IV. Public comments received on the proposed rule are available for inspection and copying for a fee at the Commission's Public Document Room located at 2120 L Street NW.(Lower Level).Washington, DC HeinOnline

-- 60 Fed. Reg. 22462 1995 Federal Regiter / Vol. 60. No. 88 1 Monday, May 8. 1995 1 Rules and Regulations 22463 I1. Final Action The final rule revises certain requirements contained in 10 CFR Part 54 and establishes a regulatory process that Is simpler, more stable, and more predictable than the previous license renewal rule. The final rule continues to ensure that continued operation beyond the term of the original operating license will not be inimical to the public health and safety. The more significant changes made to the previous license renewal rule are as follows: (I1 The intent of the license renewal review has been clarified to focus on the adverse effects of aging rather than identification of all aging mechanisms.

The final rule is intended to ensure that important systems, structures, and components will continue to perform their intended function in the period of extended operation.

Identification of individual aging mechanisms is not required as part of the license renewal review. The definitions of age-related degradation, age-related degradation unique to license renewal, aging mechanisms, renewal term, and effective program have been deleted.(2) The definitions of integrated plant assessment (IPA) (§ 54.3) and the IPA process (§ 54.21(a))

have been clarified to be consistent with the revised focus in item (I) on the detrimental effects of aging.(3) A new § 54.4 has been added to replace the definition of systems, structures, and components "important to license renewal" in §54.3. Section 54.4 defines those systems, structures, and components within the scope of the license renewal rule and identifiesthe important functions (intended functions) that must be maintained.

The requirement to include systems, structures, and components that have limiting conditions for operation in facility technical specifications within the scope of license renewal has been deleted.(4) In § 54.21(a).

the IPA process has been simplified.

The wording has been changed to resolve any ambiguity associated with the use of the terms systems, structures, and components (SSCs) and structures and components (SCs). A simplified methodology for determining whether a structure or component requires an aging management review for license renewal has been delineated.

Only passive, long-lived structures and components are subject to an aging management review for license renewal. Sections 54.21 (b)and (d) have been deleted, and a new§ 54.21(c) dealing with time-limited aging analyses (TLAA) and § 54.21(d)dealing with requirements for the fin&l safety analysis report (FSAR)supplement have been added. The requirement in § 54.21(c) of the previous rule to review any relief from codes and standards has been deleted, and the requirement in § 54.21(c) of the previous rule to review exemptions from regulatory requirements has been clarified and linked with the time-limited aging analyses.(5) In S 54.22, the requirement to include detailed justification for certain technical specification changes in the FSAR supplement has been modified to require that the detailed justification be included in the license renewal application.

16) In § 54.29, the standards for issuance of a renewed license have been changed to reflect the revised focus on the detrimental effects of aging concerning structures and components requiring an aging management review for license renewal and any time-limited issues (including exemptions) applicable for the renewal term. A new§ 54.30 has been added to distinguish between those issues identified during the license renewal process that require resolution during the license renewal process and those issues that require resolution during the current license term.(7) In § 54.33, requirements for continuation of the current licensing basis (CLB) and conditions of renewed licenses have been changed to delete all reference to age-related degradation unique to license renewal (ARDUTLRJ.

Section 54.33(d) of the previous rule, which requires a specific change control process, has been deleted.f8) In §54.37. additional records and recordkeeping requirements have been changed to be less prescriptive.

Sectionhas been deleted.Ill. Principal Imues a. Continued Validity of Certain Findings in Previous Rulemaking The principal purpose of this final rule is to simplify and clarify the previous license renewal rule. Unless otherwise clarified or reevaluated, either directly or indirectly, in the discussion forthis final rule, the conclusions in the SOC for the previous license renewal rule remain valid (56 FR 64943: December 13, 1991).One commenter stated that the prvious license renewal rule has been substantially modified in the proposed rule so as to constitute a "recision'*

of the previous rule.The Commission does not believe that this final rule represents a recision of the previous license renewal rule, 10 CFR Part 54. As stated in the SOC for the proposed rule, "[u]nless otherwise clarified or reevaluated, either directly or indirectly, in the discussion for this proposed rule, the conclusions in the SOC for the current license renewal rule remain valid * *" September 9, 1994 (59 FR 46576). Some of the subjects resolved in the previous Part 54 rulemaking that remain unaffected by this final rule Include the concept of the CLB, the nature of the current regulatory process, the regulatory process for assuring compliance with the CLB. form of the renewed license, the term of the renewed license, antitrust considerations, and the applicability of the provisions of the Price-Anderson Act.Furthermore, regardless of whether this final rule constitutes a recision of the previous rule, the Commission agrees with the commenter that the Administrative Procedure Act (APA)requires the Commission to provide a"reasoned analysis" for the changes to Part 54 that are being adopted in this final rule. The Commission takes issue with the commenter with regard to whether the SOC for the proposed and for the final rule adequately explain the bases for the changes. The Commission believes that this SOC provides a detailed discussion setting forth the perceived problems with the previous license renewal rule as well as a discussion of the bases for this final rule. In sum, the Commission has fulfilled its obligation under the APA to provide the bases for this rule, regardless of whether the changes that are being adopted in this final rule constitute a recision of the previous license renewal rule.b. Reaffirmation of the Regulatory Philosophy and Approach and Clarificatioh of the Two Principles of License Renewal (i) Regulatory Philosophy In developing the previous license renewal rule, the Commission concluded that issues material to the renewal of a nuclear power plant operating licEnse are to be confined to those issues that the.Commission determines are uniquely relevant to protecting the public health and safety and preserving common defense and security during the period of extended operation.

Other issues would, by definition, have a relevance to the safety and security of the public during current plant operation.

Given the Commission's ongoing obligatibn to oversee the safety and security of operating reactors, issues that are relevant to current plant operation will be addressed by the existing regulatory HeinOnline

-- 60 Fed. Reg. 22463 1995 22490 Federal Register / Vol. 60, No. 88 / Monday, May 8, 1995 / Rules and Regulations data needed, and completing and reviewing the collection of information.

Send comnments regarding this burden estimate or any other aspect of this collection of information, including suggestions for reducing this burden, to the Information and Records Management Branch (T6 F33), U.S.Nuclear Regulatory Commission, Washington, DC 20555-0001; and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0155), Office of Management and Budget, Washington, DC 20503.IX. Regulatory Analysis The NRC prepared a draft regulatory analysis of the values and impacts of the proposed rule and of a set of significant alternatives.

The draft regulatory analysis was placed in the Commission's public document room for review by interested members of the public. In addition, a summary of the findings and conclusions of the regulatory analysis were published in the Federal Register (59 FR 46591, .September 9, 1994) concurrent with the proposed rule. No comments were received on the regulatory analysis.

The regulatory analysis has been finalized and is available for inspection in the NRC Public Document Room, 2120 L Street NW (Lower Level), Washington DC. Single copies of the analysis may be obtained from Joseph J. Mate, Office of Nuclear Reactor Regulation, U.S.Nuclear Regulatory Commission, Washington DC 20555, (301) 415-1109.X. Regulatory Flexibility Act Certification As required by the Regulatory Flexibility Act of 1980, (5 U,S.C. 605 (b)), the Commission certifies that this final rule does not have a significant economic impact upon a substantial number of small entities.

The final rule sets forth the application procedures and the technical requirements for renewed operating licenses for nuclear power plants. The owners of nuclear power plants do not fall within the definition of small business entities as defined in Section 3 of the.Small Business Act (15 U.S.C. 632), the Small Business Size Standards of the Small Business Administration (13 CFR Part 121), or the Commission's Size Standards (56 FR 56671; November 6, 1991).XI. Non-Applicability of the Backfit Rule.This rule, like the previous license renewal rule, addresses the procedural and technical requirements for obtaining a renewed operating license for nuclear power plants. Although this amendment constitutes a change to an existing regulation, the NRC has determined that the backfit rule, 10 CFR 50.109, does not apply because this amendment only affects prospective applicants for license renewal. The primary impetus for the backfit rule was"regulatory stability." Once the Commission decides to issue a license, the terms and conditions for operating under that license would not be changed arbitrarily post hoc. As the Commission expressed in the preamble for 10 CFR 52, which prospectively changed the requirements for receiving'design certifications, the backfit rule-[Wias not intended to apply to every regulatory action which changes settled expectations.

Clearly, the backfit rule would not apply to a rule which imposed more stringent requirements on all future applicants for construction permits, even though such a rule might arguably have an adverse impact on a person who was-considering applying for a permit but had not done so yet. In this latter 6ase, the backfit rule protects the construction permit'holder, but not the perspective applicant, or even the present applicant.

(54 FR 15385-86; April 18.1989).Regulatory stability from a backfitting standpoint is not a relevant issue with respect to this rule. There are no licensees currently holding renewed nuclear power plant operating licenses who would be affected by this rule. No applications for license renewal have been docketed.

It is also unlikely that any license renewal applications will be submitted before this rule becomes effective.

Consequently, there are no valid licensee or applicant expectations that may be changed regarding the terms and conditions for obtaining a renewed operating license. Accordingly, this rule'does not constitute a "backfit" as defined in 10 CFR 50.109(a)(1).

Furthermore, one reason the Commission is amending 10 CFR Part.54 is because of the concerns of nuclear power plant licensees who were dissatisfied with the previous requirements in 10 CFR Part 54 and urged the Commission to modify the.rule to address their concerns.

Under this circumstance, the policy objective of the ba&kfit rule would not be served by undertaking a backfit analysis.Regulatory and technical alternatives for addressing the concerns with the previous 10 CFR Part 54 were analyzed and considered in the regulatory analysis that has been prepared for this rule. Preparation of a separate backfit statement would not provide any substantial additional benefit: Therefore, the Commission has determined that a backfit analysis pursuant to 10 CFR 50.109 need not be prepared for this rule.NEI commented that the NRC should review its determination regarding the application of backfit protection.

to license renewal. Although not clearly stated in its comments, NEI appears to argue that the protection afforded by 10 CFR 50.109 should.apply in individual license renewal proceedings when the NRC seeks to impose requirements that"go beyond what is necessary for adequately managing the effects of aging on intended functions in the period of extended operation (i.e., enhancements)." NEI stated that in such cases, the NRC should perform an analysis to demonstrate that the proposed additional requirements will result in substantial increase in overall safety and that direct and indirect costs are justified relative to the safety benefit. Furthermore, NEI believes that if there are two ormore' means of adequately managing the effects of aging, cost must be taken into, account in selecting an alternative.

The industry's desire for a special provision in the rule that would impose backfit-style requirements on the Commission's review is neither necessary nor appropriate.

The intent of the licenserenewal rule is clear-tb ensure that the effects of aging on functionality of certain systems, structures, and components are adequately managed in the period of extended operation.

The Commission does not intenrd to impose requirements on a licensee that go beyond what is necessary to adequately manage aging effects. The focus of the industry's concern appears to be on potential disagreements between the Commission and renewal applicants regarding what is or is not considered "adequate" for managing the effects of aging. The* Commission understands the industry's concern, but does not believe it appropriate or consistent with current practice to further limit (i.e., beyond the limits established by the rule) the NRC staff in its review of an application for a renewal license.Additionally, the Commission sees no justification for requiring a consideration of costs among alternative aging management programs.

The renewal process is designed such that a renewal applicant proposes the alternatives it believes manages the effects of aging for those structures and components defined by the rule. The NRC.staff has the responsibility of reviewing the applicant's proposals and determining whether they are adequate such that there is reasonable assurance that activities authorized-by the renewed license will continue to be HeinOnline

-- 60 Fed. Reg. 22490 1995 Federal, Register / Vol'. 60, No. 88 / Monday, May 8, 1995 1 Rules and Regulationw 12491 conducted in accordance with the CLB.The* Commission believes that this license renewal review must necessarily be. performed without regard to cost.List of Subjects 10 CFR Part 2 Administrative practice and procedure.

Antitrust, Byproduct material, Classified information, Environmental protection, Nuclear materials, Nuclear power plants and reactors, Penalties, Sex discrimination, Source material, Special nuclear material, Waste treatment and disposal.10 CFR Part 51 Administrative practice and procedure, Environmental impact statement, Nuclear materials, Nuclear power plants and reactors, Reporting and recordkeeping requirements.

10 CFl Part 54 Administrative practice and procedure, Aging, Effects of aging, Time-limited aging analyses, Backfitting, Classified information, Criminal penalties, Environmental protection, Nuclear power plants. and reactors, Reporting and recordkeeping requirements.

For the reasons set out in the preamble and under the authority of the Atomic Energy-Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 552 and 553, the Commission is adopting the following amendments to .10 CFR Parts 2, 51, and 54.PART 2-RULES OF PRACTICE FOR DOMESTIC LICENSING PROCEEDINGS

1. The authority citation for Part 2 is revised to read as follows: Authority:

Secs. 161,181,68 Stat, 948, 953, as amended (42 U.S.C. 2201, 2231); sec.191, as amended, Pub. L.87-615, 76 Stat. 409 (42 U.S.C. 2241); sec, 201, 88 Stat. 1242, as amended (42 U.S.C. 5841); 5 U.S.C. 552.Section 2.101 also issued under secs. 53, 62, 63, 81, 103, 104, 105, 68 Stat. 930. 932, 933, 935, 936, 937, 938, as amended (42 U.S.C.2073, 2092, 2093, 2111, 2133, 2134, 2135);sec. 114(0, Pub. L.97-425, 96 Stat. 2213, as amended (42 U.S.C. 10134(Q);

sec. 102, Pub.L.91-190, 83 Stat. 853, as amended (42 U.S.C. 4332); sec. 301, 88 Stat. 1248 (42 U.S.C. 5871). Sections 2.102, 2.103, 2.104, 2.105, 2.721 also issued under sacs. 102,103, 104. 105, 183, 189, 68 Stat. 936, 937, 938, 954, 955, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2233, 2239). Section 2.105 also issued under Pub. L. 97ý-415, 96 Stat. 2073 (42 U.S.C. 2239). Sections 2.200-2.206 also issued under secs. 161b, i, o, 182,186, 234, 68 Stat. 948-951, 955, 83 Stat. 444, as amended (42 U.S.C. 2201(b), (i), (o), 2236.2282); sec. 206, 88 Stat. 1246 (42 U.S.C.5846). Sections 2.600-2.606 also issued under sec. 102, Pub. L.91-190, 83 Stat. 853 as amended (42 U.S.C. 4332). Sections 2.700a, 2.719 also issued under 5 U.S.C. 554.Sections 2.754, 2.760. 2.770. 2.780, also issued under 5 U.S.C. 557 Section 2.764 and Table 1A of Appendix C are also issued under secs. 135, 141, Pub. L.97-425, 96 Stat.2232, 2241 (42 U.S.C. 10155, 10161). Section 2.790 also issued under sec, 103, 68 Stat. 936, as amended (42 U.S.C. 2133) and 5 U.S.C.552. Sections 2.800 and 2.808 also issued under 5 U.S.C. 553. Section 2.809 also issued under 5 U.S.C. 553 and sec. 29, Pub. L.85-256, 71 Stat. 579. as amended (42 U.S.C.2039). Subpart K also issued under sec. 189, 68 Stat. 955 (42 U.S.C. 2239); sec. 134, Pub.L.97-425. 96 Stat, 2230 (42 U.S.C. 10154).Subpart L also issued under sec. 189, 68 Stat.955 (42 U.S.C. 2239). Appendix A also issued under sec. 6, Pub, L.91-560, 84 Stat. 1473 (42 U.S.C. 2135). Appendix B also issued under sec. 10, Pub. L.99-240, 99 Stat. 1842 (42 U.S.C. 2021b et seq.].2. In § 2.758, paragraphs (b) and (e)are revised to read as follows:§ 2.758 Consideration of Commission rules and regulations in adjudicatory proceedings.(b) A party to an adjudicatory proceeding involving initial or renewal licensing subject to this subpart may petition that the application of a specified Commission rule or regulation or any provision thereof, of the type described in paragraph (a) of this section, be waived or an exception made for the particular proceeding; The sole, ground for petition for waiver or exception shall be that special.circumstances with respect to the subject matter of the particular proceeding are such that the application of the rule or regulation (or provision thereof) would not serve the purposes for which the rule or regulation was adopted. The petition shall be accompanied by an affidavit that identifies the specific aspect or aspects of the subject matter of the proceeding as to which the application of the rule or regulation (or provision thereof)would not serve the purposes for which the rule or regulation was adopted, and shall set forth with particularity the special circumstances alleged to justify the waiver or exception requested.

Any other party may file a response thereto, by counter affidavit cr otherwise.(e) Whether or not the procedure in paragraph (b) of this section is available, a party to an initial or renewal licensing proceeding may file a petition for rulemaking pursuant to § 2.802.DART 51-ENVIRONMENTAL CROTECTION REGULATIONS FOR fnOMESTIC LICENSING AND RELATED REGULATORY FUNCTIONS 3. The authority citation for Part 51 is revised to read as follows: iuthority:

Sec. 161, 68 Stat. 948, as amended, Sec. 1701, 106 Stat. 2951, 2952, 2953, (42 U.S.C. 2201, 2297f; secs. 201, as amended, 202, 88 Stat. 1242, as amended, 1244 (42 U.S.C. 5841, 5842). Subpart A also issued under National Environmental Policy Act of 1969, secs. 102,104, 105, 83 Stat. 853-854, as amended (42 U.S.C. 4332,, 4334., 4335); and Pub. L.95-604, Title I1, 92 -stat.3033-3041; and sec. 193, Pub. L. 101-575, 104 Stat. 2835, 42 U.S.C. 2243). Sections 51.20, 51.30, 51.60, 51.61, 51.80, and 51.97 also issued under secs. 135, 141, Pub. L.97-425, 96 Stat. 2232, 2241. and sec. 148. Pub.L. 100-203, 101 Stat. 1330-223 (42 U.S.C.10155. 10161, 10168). 'Section 51.22 also issued under sec. 274,73 Stat. 688. as amended by 92 Stat. 3036-3038 (42 U.S.C.2021) and under Nuclear Waste Policy Act of 1982, sec. 121, 96 Stat. 2228 (42 U.S.C.10141). Sections 51.43, 51.67, and 51.109 also under Nuclear Waste Policy Act of 1982, sec. 114(f), 96 Stat. 2216, as amended (42 U.S.C. 10134(0): 4. In § 51.22, paragraph (c)(3) is revised to read as follows:§51.22 Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review.(c)(3) Amendments to Parts 20, 30. 31, 32, 33, 34.35, 39, 40, 50, 51,54, 60, 61, 70, 71, 72, 73, 74, 81 and 100 of this chapter which relate to-(i) Procedures for filing and reviewing.

applications for licenses or construction permits or other forms of permission or for amendments to or renewals of licenses or construction permits or other forms of permission;(ii) Recordkeeping requirements; or (iii) Reporting requirements; and (iv) Actions on petitions for rulemaking relating to these*amendments.

5. Part 54 is revised to read as follows.PART 54-REQUIREMENTS FOR RENEWAL OF OPERATING LICENSES FOR NUCLEAR POWER PLANTS General Provisions Sac.54.1 Purpose 54.3 Definitions.

54.4 Scope.54.5 Interpretations.

54.7, Written communications.

54.9 Information collection requirements:

OMB approval HeinOnline

-- 60 Fed. Reg. 22491 1995 Applicant's Consolidated Memorandum in Opposition to New York State's Motion for.Summary Disposition of Contention NYS-35/36 and in Support of Its Cross-Motion for Summary Disposition (February 3, 2011)NYS-35/36:

ATTACHMENT 5

Federal Register / Vol. 55, No.-I3T:/

Tuesday,.

July: 17, 1990 /.P roposed" Rules 29043.(vi).Milk chocolate:and.the.milk chocolate-component.of'other products;(vii) Airy other dairy product not otherwise.specified in this section.

Divlslon,.

Agricultural-Marketing Service Proposal No. &I-Make. such changes as may be necessary to make, alt marketing agreements and the orders conform with.any amendments, thereto. that may result-from this.hearing..

Copies of-thls- notice of'hearing and-the. orders:may be.procured from the Market Administrators or from the Hearing Clerk, Rlo0m:1083; South.Building, Uifted'States Department of: Agriculturae.

Wahington.

DC. 20250.: or.may be.inspected.there.

I Copies of the transcrilpt~otestimony taken.at will not be.available, for didtribution through.the-Hearing.Clerks'.Office.

ltyou.wish.to.

purchase a.copy, arrngements.may be made.wittbthe repr6rter a"iihe.hearing, From the time that a. hearing notice.is.

issued and until the issuance ofa final decision in aproceeding, Department employees involved in.the decisional process are prohibited from discussing, the merits of the hearing issues on an ex parte basis with any person having.an interest in the proceeding, For:this:particular proceeding,.

the-prohibition zapplies.to employees.

in. the f1llowing organizationsal uits: Office of' th 6 Secretary.

of'Agriculture Office of the Administrator, Agricultural Marketing Service Office of the General Counsel Dairy Division.

Agricultural Marketing, Service.(Waahingt.0n~office.

only]Offices of 0llth6 Market Administrators-Procedural matters are not subject to the.abo.ve prohibition andmay-be discussed at any time.Signed at Washington; DC. om July It 1890.Daniel Haley.Administratorz-

[FR Doc. 90-1062.FPiled:1-16.-MG 8:46.am)Oi.LaI0 COON 341042".NUCLEAR REGULAtORY COMMISSION.

10 CFRParW'*2s50,.and s5 RIN4151-AD04 Nuclear Powerftrnt.Llcense-Renewal AGENCy: NuclearRegtlatory-Commissions n.il&ACTINo: Iproposed rule.

SUMMARY

.TheNuclear Regulatory.

-.Commission is: proposing to issue a rule..: that would establish the requirements that an applicant for renewal'of'a.

n j. 3 ar power plant operating.

license must meetthkn information that must be submitted' to the NRC for review so that the.agency can, determine whether those: requirements have in fact been met, and the' application' procedures.

This.proposedrule.will inform nuclear power-plant licensees of necessary requirements for renewing operating licenses;.

DATES: The comment period expires.October- 15 1990. Comments received after this. date-willtbe considered ifit is practical to do-so, but' the Commission is able to assureconsideration only for.comments'recelved' on- or:before this.date. " ADDRESSES:.Coment may be sent to the.Secretary of.thoe'Cmmisiion.

U.S. -Nuclear Regulatory.Commission, Washington.

D. 2055..Attention:

Docketing.

and.SerVice Branch, or may be hand-delivered to One White Flint North, 11555:Rockville Pike, RockWille, MD 20852; between 7T3V" am and 4:15 pm'Federal workdays.'

Copies of'comments received may be.examined at.the Cbmmission's Public Document Room at 2120 L.St., NW, (Lower Level), Washington.DC, between the hours of 7:45 am and.4:15 pm Federal workdays.FOR FURThER. INFORMA-TIOt-CONTACT: George Sege, Office.ofNuclear Regulatory Research, LLS. Nuclear.Regulatory Commisslon.Washingtn., D.C.20555.

Telephone; (3011 492.-3917.

SUPPLEMENTARY INFORMATIONr

1. Introduction.

I. Background..

III. Proposed Action.IV. Principal Issues: a. Regulatory Philosophy and Approach.b. Current UcensingBa:ais..

c. Aging Management-.

d. Naturetof License.e. Latest Date for Filing Renewal Application, the Timely Renewal Doctrine, and Suffilcency of'the Renewal.Application.

f: E'arliest Date for-FilIng.Applications.

&- Renewal Term.h. Effective.Date of-Renewed LIcense.L Content of Applicatipn--Technical Information.

I. Environmental Information.

L. B'ackfit Cinsiderations.

I. Hearings.m. Report-ofthe AdvisoryCommittee on.ReactorSbfeguards.'

n. Emergency.

Planning Considerations.

o. Plant.Pjhysical Security

p. Operator-LicensinsConsiderations..

q. Financial Qualification Considerations.

r. Deconmmissioning'Conslderations.

a. Antitrust Reviwe....

.Cbmpllarne with 10,.CFPkrt 14(Y V. Questions:

VI. Availability of Documents,,.

VIL Environmental Impact.VIII. Paperworkeductio.n Act Statement.

IX. Regulatory Analysls,.

X.Regulatory Flexibility Act CertiulcatiltL Xl. Non-Applicability of Backfit Rule.I. Introduction The-Atomic Ehergy. Act .of 1954 (AEAJ limits. the duration ofmost operating licenses-for nuclear 'power plants to a maximum of'40 years, but permits their renewal. The Commission's regulations at 10 CFR 50.51 implement this:authority by permitting, renewal. However, § 50.51 provides no standards or procedures for determining renewalapplications.

The nuclear utility. industry has expressed considerable-interest In operating-.

existing.

nuclear pb.wer.plants.beyond their. Initial. term ofloperaeion-IThe.

indistry.

has.undertaken several'initiatives.In support of plant life.extension..

A Steering Committee.

on Nuclear Plant Life Extension (NLJ.LEX')

has been formedlunder the direction of the- Nuclear Managementand.Resources Council (NUMARC).

The Electric Power Research Institute (EPRI), in cooperation with the' U.S.- Department of Energy (DOE), and two utilities have sponsored research on life extension, Including.

pilot studies on two nuclear plants, Surry-1 and Monticello.

This has culminated in DOE funding.of two lead applications for renewal of the-operating licenses, for the-Yankee Roweand Monticello.

facilities; The: nuclearindustry haa urged- the NRC to develop: standarda and.procedures.

for license, renewal.sor that the utilities would know what will be.required to obtain a renewed operating license. The Industry states that-a license renewal rule is needed now.because of themeed for a significant number of plants to make decisionsin the near future as. to whether to seek.license renewal. For the- oldest nuclear power plants the expiration of their original operating licenses is approaching.

If thelo08 nuclear power plants-licensed as-of the-end of 1989-were licensed fbr'40 years-from:

the-date of their operating license, the first eight plants will have- their-licenses expire-during the years 2000 to 2009, with .another 40 licenses expiring by Utilities contend that they will require 10 to 15 years to plan- and build replacement powerplants-if the" operating licenses for existing nuclear powerplants are notvrenewed, They also contend- that-the NRC's' technical requirements for-license-renewal must'be- established' before utilities can-reasonably.

determine whether'renewal of their existing operating-licenses, Is HeinOnline

-- 55 Fed. Reg. 29043 1990 Federal Roegister

/ Vol. 55; No:- 137>/. Tuesday. July: 17, 1990 / Proposed Rules 29043.(vi). Milk chocolate-and the.milk chocolate-component.of'otherproducts;(vii) Any other dairy, product not otherwise-specified, in this section.ProposedLliy.the:JaIrý Division., Agricultural-Markeling Service Proposal No. E&-Make, such changes as may be necessary to make. all marketing agreements and the orders conform with.any amendments, thereto. that may result-from this;hearing.

Copies, oftthis- notice of[hearing and-the. orders:may-be-procured from the Market Administrators or from the Hearing Clerk. Rom.1083; Sbuth.Buildifg.

Uhited'StatesDepartment of: Agriculture.

Wahington.

DC 20Z50. or.may bedinspected.there.

Copies of the transcriipt.of.testimony taken-at thehehering willnot be.available, fio diatributiori through. the Hearin'.Clerki Office. I.lyoiiwish.to.

purchase.a.cop'y', arrangments.may be made-with'the reporter.

&'flie.hearing.

From the time that a. hearing notice.is.

issued and until the issuance of-a final decision in a proceeding, Department employees Involved in.the decisional process are prohibited from-discussing, the merits of the hearing issueson an ex parte basis with any person havingan interest in the proceeding..F.or this particular proceeding,:the prohibition applies to eripl6yeei In. the following-organizatidna'iinits:

-Office of the Secretary, of Agriculture Office of the Administrator, Agricultural Marketing Service Office Of th General Counsel Dairy Division.

Agricultural Marketing.

Service.(Washington.office.onlyj Offices of pllthe Market Administrators.

Proceduraltmatters are not subject to the, above prohibition -and may be discussed at any time.Signed~at Washington; DC. om July 11, 1990.Daniel Haley, Admninstrotor-

[FR Doc. 90-10826.Fitaci7-0.-00; 8:45.am]011L11 COOS 34102-0".NUCLEAR REGULATORY' COMMISSION.

10 CFR-PartS"-2,50.

a'n'd54-RIN4150-AD04-Nuclear Power Piunt.icoense-RenewalNuc] ar-Regula~tory-Commission.

ACTION: Proposed rul&

SUMMARY

.The.Nuclear Regulatory.

Commission ia proposing to issue-a rule..that would establish the requirements that an applicant for renewal-ofa.

njp -! -ar power plant operating.

license must meetthe, informationr that must be submltted'to the NRC for review so that the.agency can determine whether those: requirements have in fact been met and the- application procedures, This.proposed.rule.

will inform nuclear power-plant licensees of necessary requirements for renewing.

operating licenses..

DATES: The comment period expires.October-1'S 1990. Comments received after this. date-will be considered ifit is practical to do-so, but- the Commission is able to assure consideration only for-.comments, received-on- or-before this.date.ADOODRESSEMS:.Comments may be sent to the.Secretary of~te Commislon.

U.S. -Nuclear Regulatory.Commission, Washington, DC. 20555,.Attention:

Docketing.

and.SerVice Branch, or may be hand-delivered to One White Flint North, 11555-Rockville Pike, Rockville, MD 20852; between 7.30"am and 415pm Federal workdays; Copies of comments received may be .examined at.the Cbommission's Public Document Room at 2120 L.St., NW. (Lower Level), Washington, DC, between the hours of 7:45 am and.4:15 pm Federal workdays.FOR FURTHER. INEORMA-TIO*

CONTACT'George Sage. Office-of Nuclear Regulatory Research.

US, Nuclear.Regulatory Commission.-Washington, D.C.20555.

Telephone; (301) 492--391.SUPPLEMENTARY .NFQRMATiOtC-

1. Introduction.

1. Background..

I. Proposed Action.IV. Principal Issues: a. Regulatory Philosophy end Approach.b. Current LicensingqrBasis..

C. Aging Management.

d. Nature-of License.a. Latest Date for Filing Renewal Application, the Timely Renewal Doctrine, and S"uffibiency of the Renewal Application; L Earliest Date for-Filing:Applications, g. Renewal Term.h. Effectlve.Date.of-Renewed License.L Content of Application-Technical Information.

J, Environmental Information.

k. lfackfit Cbnsiderations.

i. Hearings.mn Report-ofthe Advisory-Committee on-Reactor-Safeguardst

n. Emergency.Planning Considerations.

o. PlantP~hysical Security C6tialdarations.

p. Operator-LlcensingýConslderations..

q. Financial Qualification Considerationna.

r. Deconmlissionlng iConsiderations.

a. Anti&rust Review.'L with 1U.CFt Phrt 14(1 V. Questions; VI. Availability of Documents..

VIL Environmental Impact.VIII. Paperwork~eduction Act Statement IX. Regulatory Analysis.X.Regulatory Flexibility Act Certifficafston XI. Non-Applicability of Backflit Rule.I. Introduction The-Atomic Energy- Act of 1954 (ARAJ limits- the-duration ofmost operating.

licenses-for nuclear power plants to a maximum of 40 years, but permits their renewal. The Commission's regulations at 10 CFR 50.51 implement this: authority by permitting.

renewal. However, § 50.51 provides no standards or procedures for determining renewal applications.

The nuclear utility, industry has expressed considerable-interest in operating,.

existing nuclear pb.wer.plants.beyond their. Initial, term of-operation..

The.indiustry-has.iundertaken se.veraL initiatives; in support of plant life.extension..A Steering.Committee.

on Nuclear Plant Life Extension (NUPLEX)has been formed'.under th'e direction of the-Nuclear Managementand Resources Council (NUMARC).

The Electric Power Research Institute (EPRI), in cooperation with the U.S. Department of Energy (DOE), and two utilities have-sponsored research on life extension, including:

pilot studies on two nuclear plants, Surry-1 and Mbnticello.

This has culminated in DOE fundingof two lead applications:

for renewal of the-operating licenses, for the-Yankee-Roweand Monticello.

facilities.

The:nuclearindustry.

haasurged_

the..NRC. to develop: standards and.procedures.

for license, renewal.so-that the utilities would know what will be.required' to obtain -a renewed operating license..The Industry states that-a license renewal rule is needed now.because of the-need for a-significant number ofplants to make. decisions;in the near future as.to whether to seek license renewal. For the, oldest nuclear power plants the expiration of-their original operating licenses Is approaching.

If the;108 nuclear power plants.licensed as-of the-end of 1989 were licensed, fbr-40 years- from, the-date of their operating license, the first eight plants will have, their-licensees expire-during the years 2000 to 2009, with another 40 licenses expiring by 2014.Utilities contend that they will require 10 to 15 years to-plan-and build replacement-powerplants-if the *operating licenses for existing nuclear powerplants are.notrenewed.

They also contend that the NR's- technical requirements forlicense-renewal must-be- established-before utilities-can-reasonably determine whether-renewal of their existing operating-licenses is HeinOnline

-- 55 Fed. Reg. 29043 1990 29044 Federal Register /Vol. 55, No. 137 / Tuesda;, July 17. 1990 / Proposed Rules I economically and techilcally justified.(For more information on the expiration of facility operating licenses, see Appendix A to the Regulatory Analysis for License Renewal, NUREG-1362.)

To ensure a reasoned process for considering license renewal for those who may pursue it, the NRC has determined to proceed with license renewal rulemaking now in order to establish the requirements for renewal of nuclear power plant operating licenses in a timely fashion.II. Background The NRC's research program on the degradation of nuclearpower plant systems, structures, and components (SSCs) due to aging began in the early 1980s. In 1982, the NRC staff, recognizing the potential impact of plant aging phenomena on the continued safe operation of nuclear power plants, convened a "Workshop on Plant Aging" in Bethesda, Maryland.

The purpose of the workshop was to focus attention.

on how to best proceed to identify and resolve the various technical plant aging issues relevant to life extension.

In 1985, the Division of Engineering of the Office of Nuclear Regulatory Research issued the first comprehensive program plan (NUREC-1144) for nuclear power plant aging research.

By 1986, age-related degradation became a more important priority with the recogni tion that utilities were interested in extending the life of their existing power.plants beyond the term of up to 40 years of their original operating licenses.

In response, the NRC staff developed the "Plan to Accomplish Technical Integration for Plant Aging/* Life Extension" (May 1987) and established a Technical Review Group for Aging and Life Extension (TIRGALEX).

The objectives'of TIRGALEX were to clearly define the technical safety and regulatory policy issues associated with plant aging and life extension and to develop a plan for resolving the issues in a timely, well-integrated manner. In May 1987, the TIRGALEX report was issued. It identified a broad spectrum of technical safety and regulatory policy issues.'These included identification of systems, structures, and components that are susceptible to aging .and could adversely affect safety; degradation processes; testing, surveillance, and maintenance requirements; and criteria for evaluating residual life. TIRGALEX concluded that many aging phenomena are readily managed and do not pose.major technical issues that would preclude life extension, provided that necessary compensatory measures such-as maintenance, sdrveillance, repair, and replacement are effectively implemented during the'extended operation, and fori a numbei of the measures during the existing license term as well.Simultaneously, a request for comments on establishmentof a policy statement on life extension was published In the Federal Register (51 FR 40334; November 6,1986). Comments were requested on seven major policy, technical, and procedural issues.(21 separate questions).

The first and sixth policy areas focused on the timing of regulatory action on life extension, including the need for a policy statement, and timing of resolution of ,policy, technical, and procedural issues.The earliest and latest dates for filing a life extension application and the potential term of such an extension were the subject of the second and a portion of the fourth policy area. The question of an appropriate licensing basis was the third policy issue, including the need for and role of a probabilistic risk assessment (PRA). The fourth and fifth areas focused on technical Issues regarding the nature of aging degradation, its Identification and mitigation, and the need for research.and changes to industry codes and standards.

The final policy area was the* need for procedural changes in the Commission's regulations for handling life extension requests.

A total of 58 written comments were received from the electric utility industry, public interest groups, private citizens, Independent consultants; and.government agencies.

These comments* were reviewed and a summary provided.in SECY-87-179, "Status of Staff Activities to Develop a License Renewal Policy, Regulations and Licensing Guidance and to Report on Public.Comments'.' (July 21, 1987)i* Based on these comments, the staff began to specifically identify and resolve the wide variety of policy and technical issues relevant to life extension.

In August 1988, the staff published an Advance Notice of Proposed Rulemaking (ANPRM) in the Federal Register (53 FR 32919; August 29, 1988) in which the Commission announced its intention to bypass a policy statement and go directly to preparing a proposed rule on license renewal. The ANPRM also announced the availability of NUREG-1317,-"Regulatory Options for Nuclear.Plant License Renewal," and requested comments on' the Issues discussed' First, three alternative licensing bases for.assessing the'adequacy of'alife 'extension application were presented and discussed: (a') the existing-lidensing basis for a facility, (b) supplenmezitatibn.

of the existing licensing basis with reviews in safety significant areas, or 1c)compliance with new plant standards at the time the application is. submitted, Commenters were asked to identify whether any other major, regulatory options for license renewal should be considered, and whether verification of the existing licensing basis at each plant should be required for license renewal.Second, two alterndtives for handling uncertainties in age-related degradation were described and discussed: (a)emphasize maintenance, Inspection, and reliability assurance, or (b) emphasize defense-in-depth.

The relative merit of the two alternatives was the second subject for comment. Third, the advisability of preparing a generic environmental impact statement (EIS)and whether part 51 should be amended to permit the NRC the option of preparing an environmental assessment (EA] instead of an EIS were discussed.

Finally, 12 procedural and policy issues were discussed.

Comments on the environmental, procedural, and policy issues were invited, Fifty-three written comments were received from nuclear industry groups and individual utilities, public interest groups, and Federal and State agencies in response to the ANPRM and commenting on NUREG-1317.

An overview and summary analysis of the comments are contained in NUREG/CR-5332, "Summary and Analysis of Public Comments on NUREG-1317:

Regulatory Options for Nuclear Plant License Renewal" (March 1989).Also in 1988 the NRC, in cooperation with the American Nuclear Society'(ANS), the American Society of Civil Engineers (ASCE), the American Society of Mechanical Engineers (ASME), and'the Institute of Electrical and Electronics Engineers (IEEE), sponsored an 'International Nuclear Power Plant Aging Symposium.

The symposium, which was held In Bethesda, Maryland, from August 30 to September 1, 1988. was attended by more than 550 internationally prominent nuclear scientists and engineers from 16 countries.

The symposium focused on the potential safety issues arising from progressive aging of nuclear power plants.-These Issues included aging of structures in austenitic steel; fatigue life of structural materials, aging of insulating materials, degradation of pumps and valves, reliability of safety system components, radiation and thermal embrittlement of metals,-and.

erosion-corrosion of fluid-mechanical systems. Discussion bddreised topics in the stafts report NUREG-1317, .which, had been'published inmiedlately HeinOnline

-- 55 Fed. Reg. 29044 1990 Federal Register / Vol. 55, No. 137 / Tuesday, July 17, 1990 / Proposed Rules 29045"preceding the symposium.

The proceedings of the symposium were published as NUREG/CP-0100 in March 1989.The NRC staff s views on specific license renewal issues, as evolved in early 1989, were presented to the public in an NRC panel discussion and question and answer session at the NRC's Regulatory Information Conference, held on April 18, 19, and 20, 1989. Among the issues discussed were the nature of a renewed license (renewed license versus amendment .of existing license), the need for probability risk assessment

[PRA), integration with the Individual Plant Examination (IPE) process, and compliance with'the National Environmental Policy Act (NEPA).On October 13, 1989 (54 FR 41980), the Commission announced that a workshop would be held on November 13 and 14, 1989, to focuson specific technical issues, including identification of the significant technicalissues bearing on safety, the nature and content of .standards for issuance of a renewed license, and the appropriate role and scope of deterministic and probabilistic risk assessments.

In addition, the schedule for rulemaking and alternatives for addressing compliance with NEPA were identified as issues for discussion.

General questions to focus workshop.

discussions were provided in the Federal Register notice and later*supplemented by a more detailed set of questions.

In addition, the Federal Register notice included a "Preliminary Regulatory Philosophy and Approach for License Renewal Regulation" and an"Outlihe of a Conceptual Approach to a License Renewal Rule." Written comments on the questions posed, the statement of regulatory philosophy, and: the conceptual rule outline were accepted by the agency up to December 1, 1989. Transcripts were made of the entire workshop.

Two hundred and one individuals (not including NRC staff)representing 89 organizations registered for the workshop.

A partial listing by category includes 62 individuals representing 28 electric utilities, 10 individuals representing 2 nuclear industry groups, 16 individuals representing 4 nuclear vendors, 5 individuals.representing 2 architect-engineer firms. 36 industry consultants representing 26 firms, 5 individuals representing 4.State agencies, 2 journilis.t from itrade press organizations, and I individual from a public Interest group. Comments a.provid.. during the workshop were;fiom 'ndus't representatives and individuals affiliated with the nuclear industry.

The Nuclear Management and Resources Council. (NUMARC).

Yankee Atomic Electric Company, and Northern States Power Company presented prepared comments at each session. In addition, written comments were received from 12 organizations, including substantial submissions by NUMARC, Yankee Atomic, Northern States Power, Westinghouse, the Illinois Department of Nuclear Safety, and an independent consultant.

DOE was the only Federal agency submitting written comments.

No comments were submitted by any public interest group.i11. Proposed Action The Atomic Energy Act, which permits renewal of licenses, and the license renewal rule already in effect (10 CFR 50.51) do not contain specific procedures, criteria, and standards that must be satisfied In order to renew a license, The proposed rule would.establish the procedures, criteria, and standards governing nuclear power plant license renewal. *The followingare the principal elements of the proposed rule: (1) The licensing basis for a nuclear power plant.during the renewal term will consist of the current licensing basis for that plant together with any additional considerations related .to possible degradation through aging of systems, structures, and components (SSCs) important to license renewal, necessary to ensure that the facility can continue to be operated without undue risk to the health and safety of the public; The "current licensing basis" includes all applicable NRC .requirements and licensee commitments, as defined in the rule.(2) Provisions are included requiring renewal applicants to perform and submit an integrated plant assessment, in which systems, structures, and components important to license renewal are identified and screened, to determine and describe the required age-related degradation management actions.(3) An application is required to contain specified information for NRC review, including a description of plans for. aging management.

(4) Opportunity for public hearings is provided.S{(5) Application may be made not more than 20 years before license expiration.

It must be made not less than 3 years.before license expiration for the timely renewal provision of 10 CFR 2.109 to , apply. ..! -..... ., (6) A renewal license is effective upon.its issuance..

-.(7{) A renewal term may be granted or approved as justified by the licensee, %but not for more than 20 years beyond the original license expiration.

IV. Principal Issues a. Regulatory Philosophy and Approach (i) Two Principles The regulation that the Commission proposes for license renewal is founded on two key principles.

The first principle is that, with the exception of age-related degradation, the current licensing basis for each reactor provides and maintains an acceptable levelof safety for operation during any renewal period.The second and equally important principle is that each plant's current licensing basis must be maintained during the renewal period,.in part through a program of age-related degradation management for systems, structures, and components that are important in this connection.(ii) First Principle:

Licensing Basis Retention The current licensing basis, as used above, means the Commission requirements and licensee commitments imposed on a nuclear power plant at the time of the initial license, as modified or supplemented by themany additional requirements that have been imposed on the licensee by the Commission subsequent to the initial license and by the additional commitments made by the licensee during the period of plant operation up to the filing-of a renewal application.

This principle is founded on the Commission's Initial finding of adequate protection for the initial design and construction of a nuclear power plant, as well as the Commission's continuing oversight and regulatory.actions with respect to nuclear power plants. The Commission may issue an operating license to a utility. only if it can make the findings required by 10 CFR 50.57. More specifically, the Commission must conclude that the facility will operate in compliance with the application, as amended, and the rules and regulations of the Commission.

Further, the Commission must conclude that the authorized activities can be conducted without endangering the health and safety of the public and that the issuance of an operating license will not be inimical to the common defense and security or the publichealth and safety. Thus, when the Comminission issues an initial operating-license, it has determined that the design, construction, and proposed operationof the facility satisfy the Commuission'srequirements

-and provide adequate protection of the public health and~safety and common defense and security.

-: *.... _HeinOnline

-- 55 Fed. Reg. 29045 1990 29046 Federal Register / Vol. ss. No. 137 / Tuesday. IJly 17, 1990 / Proposed Rules However. the licensing basis upon which the Commission determined that an acceptable level of safety existed does not remain fixed for the term of the operating license. Rather, the licensing basis continues to evolve during the term of operation, in part due to the continuing regulatory activities of the Commission.

These Include research.inspections, and the evaluation of operating experience.

New requirements and guidance are promulgated by the Commission which may require plant modifications on a plant-specific basis;generic and unresolved safety issues are resolved and the resolution may require that licensees evaluate and modify their designs: and additional evaluations are routinely required as the Commission Identifies areas of plant operation that require additional understanding.(iii) Review of Operating Events The Commission has a program for the review of operating events at nuclear power plants, As a requirement of the current licensing basis, and one which would continue during the renewal term. each licensee is required to notify the Commission promptly of any plant event that meets or exceeds the threshold defined in 10 CFR 50.72 and to file a written licensee event report for those events that meet or exceed the threshold defined in10 CFR 50.73. ThIs information is reviewed daily and followup efforts are carried out for events that appear to be potentially risk significant or are Judged to be a possible precursor to a more severe event.Depending on the significance, further action may be taken to notify all licensees or to Impose additional requirements.

Information on operating events is disseminated by the NRC in the form of information notices, bulletins, and other reports; by individual licensees in the form of licensee event reports. The total process offers a high degree of assurance that events that are potentially risk significant or precursors to potentially significant events are being reviewed and resolved expeditiously.(iv) Generic Safety Issues As described in SECY-89-138.

the Commission also maintains an active program for evaluating and resolving generic. Issues that may impact public health andesafety.

A generic safety issue (GSI1 involves a safety concern that may affect the design, construction, or operation of all, several, or a class of reactors or facilities.

Its resolution may have a potential for safety improvements and promulgation of new or revised requirements or guidance.

It should be noted, however, that all unresolved GSIs generally address only enhancements of safety. This conclusion was determined during the initial .evaluation of the generic concernt which assessed whether any aspect of the generic concern might have a significant impact on the protection of the public health and safety or that immediate remedial action would be warranted.

The licensing basis of individual plants includes changes that have resulted from resolution of generic Issues determined to be applicable and will Include applicable generic-Issue-derived changes in the future.A special group of 22 generic safety issues deemed to be of sufficient significance to warrant both a high-priority resolution effort and special attention In tracking was designated as unresolved safety issues (USIs). All USIa have been resolved.

Most of the USI resolutions have been implemented:

the remainder are being implemented on a schedule found satisfactory by the staff.The USI and GSI resolution process is limited to Issues that are not of such gravity that immediate action (remedy or shutdown)

Is required.Cost-benefit analyses were employed as part of the basis of resolving GSIs involving safety enhancement above the adequate-safety leveL In these tradeoffs between net safety benefit and net cost, the remaining plant operating term ordinarily enters the calculations.

However. such calculations do not have a precision sufficient to make a significant distinction between plant operating terms with and without a 20-year renewal given the fact that these decisions have been based on average plant ages in the first half of a 40-year license term. Accordingly, it is not necessary to reexamine in the license renewal context such cost-benefit calculations underlying decisions not to.backfit. Should special circumstances In connection with a particular issue as applied to a particular plant warrant reassessment.

such reassessment would be undertaken on a plant-specific basis.(v) Systematic Evaluation Program In 1977 the NRC initiated the Systematic Evaluation Program (SEP) to review the designs of older operating nuclear power plants and thereby confirm and document their safety. The reviews were organized into approximately 90 review topics (reduced by consolidations from 137 originally Identified).

The review results were documented in a series of Integrated Plant Safety Assessment Reports. As a result of these reviews with respect to some of the Issues, the licensees proposed and implemented procedural or hardware modifications or additional analyses to define corrective actions that would improve plant safety with respect to the differences from current requirements that were Identified.

The SEP effort highlighted a smaller group of 27 regulatory topics for which corrective action was generally found to be necessary for all of the initial SEP plants end for which significant safety improvements for other operating plants of the same vintage could be expected.The topics on this smaller list are referred to as the SEP "lessons learned." and the staff expects that these topics would be generally applicable to operating plants that received their construction permits in the late 1960s or early 19709.As part of the current staff effort associated with documenting the regulatory processes that contribute to the continued adequacy of the current licensing bases at operating plants, the.staff has under way a short-term effort to identify how specific "lessons , learned" from the SEP effort have been factored into the licensing bases of all operating plants or into ongoing regulatory programs.

The staff program includes identification and definition of the'lessons learned as generic safety issues and determination of the appropriate priority rankings for the resolution of these issues. The staff effort will take public comments on this issue into account.(vi) Consistency of Regulatory Philosophy The regulatory philosophy containing.the two fundamental principles is also consistent with the Commission polic*stated in the Policy Statement entitled"Severe Reactor Accidents Regarding Future Designs and Existing Plants (50 FR 23138; August 8,1985)." in this Policy Statement, the Commission concluded that existing plants pose no undue risk to public health and safety. Moreover, the Commission stated that it has ongoing nuclear safety programs.described in NUREG-1070, that include the resolution of unresolved safety and generic safety issues, the Severe Accident Research Program. operating experience and data evaluation concerning equipment failures and human error, and scrutiny by NRC inspectors to monitor the quality of plant construction, operation, and maintenance.

If new safety information were to' become available, from any source, to question the conclusion of no undue risk, then the technical Issue(s) so identified would be resolved by the NRt.under its backfit policy and other existing procedures including the possibility of generic rulemaking.

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-- 55 Fed. Reg. 29046 1990 Federal Register I Vol. 55, No. 137 / Tuesday, July 17, 1990 I Proposed Rules..29047 (vii) Probabilistic Risk Assessment Although a plant-specific probabilistic risk assessment (PRA) or plant safety assessment (PSA) will not be a requirement for the renewal of plant operating licenses, the Commission recognizes that a plant-specific probabilistic assessment can be used as an effective tool that can provide integrated Insights into the plant design and procedures and provide an additional measure of overall plant safety. The Commission understands that all plants will have completed a plant-specific PRA as part of the individual Plant Examinationprogram.

Probabilistic assessment techniques could also be used as a supplemental tool in the renewal applicant's integrated plant assessment that is to underlie the plant's age-related degradation management program as well as in monitoring, the safety implications of a plant's performance during the renewal term. As part of the monitoring function, time trends in the frequency of events or in the rate of deterioration of equipment with significant safety implications could be identified.

This monitoring process helps ensure that acceptable levels of safety are maintained during the lifetime of any reactor, including any renewal terms.(viii) Ongoing Assurance Thus, the-Commission-required changes to a plant's licensing basis provide ongoing assurance that the original Commission conclusion of adequate protection of the health and safety and common defense and security continues to remain valid throughout the remaining term of the facility's operating license..(ix) Second Principle:

Maintaining the Licensing, Basis During Renewal Term The second principle for license renewal is that the Commission must ensure that the plant-specific licensing basis is maintained during the renewal term. This principle is a necessary complement to the first principle.

As discussed above, the first principle is founded upon a generic determination that each nuclear power plant's licensing basis, if complied with, provides reasonable assurance of adequate protection throuhout the renewal term. Therefore, it follows that each nuclear power plant that complies with its license throughout the renewal term will actually provide the adequate protection thought to be provided by the licensing basis. The Commission believes that adherence to the licensing basis can and will be ensured by: (a)* 54.33(d), which states the licensing basis for the renewed license shall include the plants' current licensing basis as defined In § 54.3(a), Including those provisions addressing age-related degradation, and (b) continuing the NRC's regulatory oversight program throughout the term of a plant's renewed license. In this manner the current licensing basis will remain enforceable by the Commission throughout the term of the renewed license to the same extent as during the original licensing term.The Commission intends to continue its regulatory oversight program throughout the term of renewed licenses.This program, which is discussedbelow in greater detail inSectionb, "Current LicensingBasis," has been successful'in the past iensuringlicensee compliance with-applicable requirements and licensee-commitments.,as-well as identifying important areas of noncompliance.

The Commission believes that this oversight, if continued throughout the term of the renewed license and modified as necessary to reflect new information and experience of extendedoperation.

will also provide assurancethat licensees will comply with their plants' licensing bases during* the term of their renewed licenses.(x) Licensing-Basis Changes The principle of compliance with the licensingbasis does not preclude changesoto the4icensingbasis; as discussed above, the licensing basis changes-throughout the term of the original operating license, and will also change-throughout the term of the renewed license. However, changes to the plant's current-licensing;basis that are unrelated to age-related degradation will not be considered orproposed by the Commissionin~deterniining whether to grant the renewal application.

Such changes to the licensing basis are inconsistent with the first key principle of license renewal viz.. that the current liceuising basis for a plant is sufficient to ensure adequate protection.

The proposed rule incorporates this principle by making a generic finding in § 54.29 that the plant-specific licensing bases for all nuclear power plants are sufficient to ensure adequate protection to the public health and safety. This finding will preclude reexamination of the adequacy of a plant's current licensing basis inindividual license renewal proceedings.

To preclude, the renewal proceeding from developing into a general reconsideration of a plant's current licensing basis, the proposed rule has been carefully structured to establish a regulatory process that is precisely directed at addressing age-related degradation during the renewed license term. Sections 54.19,54.21, and 54.23, which specify the information that must be submitted in a renewal application.

require only information regarding administrative matters, age-related degradation, and environmental impact.While the applicant must submit a list of documents describing portions of the current licensing basis which are relevant to the system, structure, and component screening process, the rule does not require submission of information relating to the adequacy of a plant's-current licensing basis. Section 54.29, which defines the~standard for issuance of a renewed license, does not require a finding that theplant's current licensing basis issufficient to ensure adequate protection.

I The only situation in which a plant's current licensing basismay be changed in a license renewalproceeding is when the licensee, asserts that it is impossible or extremely impractical,for it to comply with its current liceiisingbasis due to age-related degradation.

Any changes in the current 'licensing basis proposed by the licensee to accommodate age-related degradation must be thoroughly analyzed and justified, by the licensee.The justification-must show that the licensee-proposed:changes provide adequate protection to -the public health and-safety.

Such licensee-initiated

.changes in the current licensing basis would be subjectto challenge in a hearing.If the staff or the licensee-seeks to make changes.in a plant's licensing basis for reasons other than age-related degradatlon,-they should-be pursued either in the existing-operating license or the renewed license, once issued. Staff-initiated, changes would-be-evaluated in accordance with the backfit rule, 10 CFR 50.109.b. Current Licensing Basis (i} Current Licensing Basis Explained As discussed earlier and as defined in this proposed rule, the current licensing basis (CLB) means the Commission requirements for the plant that are in effect at.the time of the renewal application.

Included are the requirements at the time that the initial license for the plant was granted together with requirements subsequently imposed. It includes the licensee's commitments for complyingwith those requirements at the-time the initial license was granted, including those documented in the operating license application or Final Safety Analysis Report (FSAR). Further, the CLB HeinOnline

-- 55 Fed. Reg. 29047 1990 Applicant's Consolidated Memorandum in Opposition to New York State's Motion for Summary Disposition of Contention NYS-35/36 and in Support of Its Cross-Motion for Summary Disposition (February 3, 2011)NYS-35/36:

ATTACHMENT 6

28467 Rules and Regulations Federal Register Vol. 61, No. 109 Wednesday, June 5, 1996 This section of the FEDERAL REGISTER contains regulatory documents having general applicability and legal effect, most of which are keyed to and codified in the Code of Federal Regulations, which is published under 50 titles pursuant to 44 U.S.C. 1510.The Code of Federal Regulations is sold by the Superintendent of Documents.

Prices of new books are listed in the first FEDERAL REGISTER issue of each week.NUCLEAR REGULATORY COMMISSION.

10 CFR Part 51 RIN 3150-"AD63 Environmental Review for Renewal of Nuclear Power Plant Operating Licenses AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule.

SUMMARY

The Nuclear Regulatory Commission (NRC) is amending its regulations regarding environmental protection regulations for domestic licensing and related regulatory functions to establish new requirements for the environmental review of applications to renew the operating licenses of nuclear power plants; The amendment defines those environmental impacts for which a.generic analysis has been performed that will be adopted inplant-specific reviews for license renewal and those environmental impacts for which plant-specific analyses are to be performed.

The amendment improves regulatory efficiency in environmental reviews for license renewal by drawing on the considerable experience of operating nuclear power reactors to generically assess many of the environmental impacts that are likely to be associated with license renewal. The amendrhent also eliminates consideration of the need for generating capacity and of utility economics from the environmental reviews because these matters are under the regulatory jurisdiction of the States and are not necessary for the NRC's understanding of the environmental consequences of a license renewal decision.The increased regulatory efficiency will result in lower costs to both the applicant in preparing a renewal application and to the NRC for reviewing plant-specific applications and better focus of review resources on significant case specific concerns.

The results should be a more focused and therefore a more effective NEPA review for each license renewal. The amendment will also provide the NRC with the flexibility to address unreviewed impacts at the site~specific stage of review and allow full consideration of the environmental impacts of license renewal.The NRC is soliciting public comment on this rule for a period of 30 days. In developing any comment specific attention should be given to the treatment of low-level waste storage and disposal impacts, the cumulative radiological effects from the uranium fuel cycle, and the effects from the disposal of high-level waste and spent fuel.DATES: Absent a determination by the NRC that the rule should be modified, based on comments received, the final rule shall be effective on August 5, 1996. The comment period expires on July 5, 1996.ADDRESSES:

Send comments to: The Secretary of the Commission, U.S.NuclearRegulatory Commission, Washington, DC 20555-0001, Attention:

Docketing and Services Branch, or hand deliver comments to the Office of the Secretary, One White Flint North, 11555 Rockville Pike, Rockville, Maryland between 7:30 a.m. and 4:15 p.m. on Federal workdays.

Copies of comments received and all documents cited in the supplementary information may be examined at the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC between the hours of 7:45 a.m. and 4:15 p.m. on Federal workdays.FOR FURTHER INFORMATION CONTACT: Donald P. Cleary, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone:

(301) 415-6263; e-mail DPC@nrc.gov.

SUPPLEMENTARY INFORMATION:

I. Introduction II. Rulemaking History III. Analysis of Public Comments A. Commenters B. Procedural Concerns 1. Public Participation and the Periodic Assessment of the Rule and GEIS 2. Economic Costs and Cost-Benefit Balancing 3. Need for Generating Capacity and Alternative Energy Sources C. Technical Concerns 1. Category and Impact Magnitude Definitions

2. Surface Water Quality 3. Aquatic Ecology 4. Groundwater Use and Quality 5. Terrestrial Ecology 6. Human Health 7. Socioeconomics

8. The Uranium Fuel Cycle and Solid Waste Management

9. Accidents 10. Decommissioning

11. Need for Generating Capacity 12. Alternatives to License Renewal 13. License Renewal Scenario 14. Environmental Justice IV. Discussion of Regulatory Requirements A. General Requirements B. The Environmental Report 1. Environmental Impacts of License Renewal 2. Consideration of Alternatives C. Supplemental Environmental Impact Statement 1. Public Scoping and Public Comments on the SEIS 2. Commission's Analysis and Preliminary Recommendation

3. Final Supplemental Environmental Impact Statement D. NEPA Review for Activities Outside NRC License Renewal Approved Scope V. Availability of Documents VI. Submittal of Comments in an Electronic Format VII. Finding of No Significant Environmental Impact Availability VIII. Paperwork Reduction Act Statement IX.Regulatory Analysis X. Regulatory Flexibility Act Certification XL Small Business Regulatory Enforcement Fairness Act XII. Backfit Analysis I. Introduction The Commission has amended its environmental protection regulations in 10 CFR part 51 to improve the efficiency of the process of environmental review for applicants seeking to renew an operating license for up to an additional 20 years. The amendments are based on the analyses conducted for and reported in NUREG-14371 "Generic Environmental Impact Statement for License Renewal of Nuclear Plants" (May 1996). The Commission's initial decision to undertake a generic assessment of the environmental impacts associated with the renewal of a nuclear power plant operating license was motivated by its beliefs that: (1) License renewal will involve nuclear power plants for which the 28480 Federal Register / Vol. 61, No. 109 / Wednesday, June 5, 1996 / Rules and Regulations assessed in an EA and found to be insignificant.

Further, the Commission has conducted EAs for seven specific licensed ISFSIs and has reached a finding of no significant environmental impact for each site. Each EA addressed the impacts of construction, use, and decommissioning.

Potential impacts that were assessed include radiological impacts, land use, terrestrial resources, water use, aquatic resources, noise, air quality, socioeconomics, radiological impacts during construction and routine operation, and radiological impacts of off-normal events and accidents.

Trends in onsite spent fuel storage capacity and the volume of spent fuel that will be generated during an additional 20 years of operation are considered in the GEIS.Spent fuel storage capacity requirements can be adequately met by ISFSIs without significant environmental impacts. The environmental impacts of onsite storage of spent fuel at all plants have been adequately assessed in the GETS for the purposes of an environmental review and agency decision on renewal of an operating license; thus, no further review within the license renewal proceeding is required.

This provision is relative to the license renewal decision and does not alter existing Commission licensing requirements specific to on-site storage of spent fuel.The environmental impacts from the transportation of fuel and waste attributable to license renewal are found to be small when they are within the range of impacts of parameters identified in Table S-4. The estimated radiological effects are within regulatory standards.

The nonradiological impacts are those from periodic shipments of fuel and waste by individual trucks or rail cars and thus would result in infrequent and localized minor contributions to traffic density.Programs designed to further reduce risk, which are already in place, provide for adequate mitigation.

Recent, ongoing efforts by the Department of Energy to study the impacts of waste transportation in the context of the multi-purpose canister (see, 60 FR, 45147, August 30, 1995) suggest that there may be unresolved issues regarding the magnitude of cumulative impacts from the use of a single rail line or truck route in the vicinity of the repository to carryall spent fuel from all plants. Accordingly, NRC declines to reach a Category 1 conclusion on this issue at this time. Table S-4 should continue to be the basis for case-by-case evaluation of transportation impacts of fuel and waste until such time as a detailed analysis of the environmental impacts of transportation to the proposed repository at Yucca Mountain becomes available.

9. Accidents Concern. Several commenters expressed concerns regarding the appropriateness of the severe accident determination in the GEIS and with the treatment of severe accident mitigation design alternatives (SAMDAs) for license renewal. A group of commenters identified areas of concern that they believe justify severe accidents being classified as a Category 3 issue. The areas included seismic risks to nuclear power plants and site-specific evacuation risks. Several commenters questioned whether the analyses of the environmental impacts of accidents were adequate to make a Category I determination for the issue of severe accidents.

The contention is that a bounding analysis would be established only if plant-specific analyses were performed for every plant, which was not the case. Instead, the GEIS analysis made use of a single generic source term'for each of the two plant types.Response.

The Commission believes that its analysis of the impacts of severe accidents is appropriate.

The GEIS provides an analysis of the consequences of severe accidents for each site in the country. The analysis adopts standard assumptions about each site for parameters such as evacuation speeds and distances traveled, and uses site-specific estimates for parameters such as population distribution and meteorological conditions.

These latter two factors were used to evaluate the exposure indices for these analyses.

The methods used result in predictions of risk that are adequate to illustrate the general magnitude and types of risks that may occur from reactor accidents.

Regarding site-evacuation risk, the radiological risk to persons as they evacuate is taken into account within the individual plant risk assessments thatform the basis for the GEIS. In addition, 10 CFR Part 50 requires that licensees maintain up-to-date emergency plans. This requirement will apply in the license renewal term as well as in the current licensing term.As was done in the GEIS analysis, the use of generic source terms (one set for PWRs and another. for BWRs) is consistent with the past practice that has been used and accepted by the NRC for individual plant Final Environmental Impact Statements (FEISs). The purpose of the source term discussion in the GEIS is to describe whether or not new information on source terms developed after the completion of the most recent FEISs indicates that the source terms used in the past under-predict environmental consequences.

The NRC has concluded that analysis of the new source term information developed over the past 10 years indicates that the expected frequency and amounts of radioactive release under severe accident conditions are less than that predicted using 'the generic source terms. A summary of the evolution of this research is provided in NUREG- 1150, "Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants" (December 1990), and its supporting documentation.

Thus, the, analyses performed for the GEIS represent adequate, plant-specific estimates of the impacts from severe accidents that would generally over-predict, rather than under-predict, environmental consequences.

Therefore, the GEIS analysis of the impacts of severe accidents for license renewal is retained and is considered applicable to all plants.Based on an evaluation of the comments, the Commission has reconsidered its previous conclusion in-the draft GEIS concerning site-specific consideration of severe accident mitigation.

The Commission has determined that a site-specific-consideration of alternatives to mitigate severe accidents will be required at the time of license renewal unless a previous consideration of such alternatives regarding plant operation has been included in a final environmental impact statement or a related supplement.

Because the third criterion required to make a Category 1 designation for an issue requires a generic consideration of mitigation, the issue of severe accidents must be reclassified as a Category 2 issue that requires a consideration of severe accident mitigation alternatives, provided this consideration has not already been completed.

The Commission's reconsideration of the, issue of severe accident mitigation for license renewal is based on the Commission's NEPA regulations that require a consideration of mitigation alternatives in its environmental impact statements (EISs) and supplements to ElSs, as well as a previous court decision that required a review of severe mitigation alternatives (referred to as SAMDAs) at the operating license stage.See, Limerick Ecology Action v. NRC, 869 F.2d 719 (3d Cir. 1989).Although the Commission has considered containment improvements for all plants pursuant to its Containment Performance Improvement (CPI) program, which identified potential containment improvements for site-specific consideration by licensees, Federal Register / Vol. 61, No. 109 / Wednesday, June 5, 1996 / Rules and Regulations 28481 and the Commission has additional ongoing regulatory programs whereby licensees search for individual plant vulnerabilities to severe accidents and consider cost-beneficial improvements, these programs have not yet been completed.

Therefore, a conclusion that severe accident mitigation has been generically considered for license renewal is premature.

The Commission believes it unlikely that any site-specific consideration of severe accident mitigation alternatives for license renewal will identify major plant design changes or modifications that will prove to be cost-beneficial for reducing severe accident frequency or consequences.

This Commission expectation regarding severe accident mitigation improvements is based on the analyses performed to date that are discussed below.The Commission's CPI program examined each of the five U.S.containment types to determine potential failure modes, potential plant improvements, and the cost-effectivenesses of such improvements.

As a result of this program, only a few containment improvements were found to be potentially beneficial and were either identified for further NRC research or for individual licensee evaluation.

In response to the Limerick decision, an NRC staff consideration of SAMDAs was specifically included in the Final Environmental Impact Statement for the Limerick 1 and 2 and Comanche Peak 1 and 2 operating license reviews, and in the Watts Bar Supplemental Final Environmental Statement for an operating license. The alternatives evaluated in these analyses included the items previously evaluated as part of the CPI Program, as well as improvements identified through other risk studies and analyses.

No physical plant modifications were found to be cost-beneficial in any of these severe accident mitigation considerations.

Only plant procedural changes were identified as being cost-beneficial.

Furthermore, the Limerick analysis was for a high-population site. Because risk is generally proportional to the population around a plant, this analysis suggests that other sites are unlikely to identify significant plant modifications that are cost-beneficial.

Additionally, each licensee is performing an individual plant examination (IPE) to look for plant vulnerabilities to internally initiated events and a separate IPE for externally initiated events (IPEEE). The licensees were requested to report their results to the Commission.

Seventy-eight IPE submittals were received and seventy-five IPEEE submittals will be received, covering all operating plants in the United States. These examinations consider potential improvements to reduce the frequency or consequences of severe accidents on a plant-specific basis and essentially constitute a broad search for severe accident mitigation alternatives.

The NRC staff is conducting a process review of each plant-specific IPE submittal and IPEEE submittal.

To date, all IPE submittals have received a preliminary review by the NRC with 46 out of 78 completed; for the IPEEE submittals, 24 of the 75 are under review. These IPEs have resulted in a number of plant procedural or programmatic improvements and some plant modifications that will further reduce the risk of severe accidents.

In conclusion, the GEIS analysis of severe accident consequences and risk is adequate, and additional plant-specific analysis of these impacts is not, required.

However, because the ongoing regulatory program related to severe accident mitigation (i.e., IPE and IPEEE)has not been completed for all plants and consideration of severe accident mitigation alternatives has not been included in an EIS or supplemental EIS related to plant operations for all plants, a site-specific consideration of severe accident mitigation alternatives is required at license renewal for those plants for which this consideration has not been performed.

The Commission expects that if these reviews identify any changes as being cost beneficial, such changes generallywould be procedural and programmatic fixes, with any hardware changes being only minor in nature and few in number.NRC staff considerations of severe accident mitigation alternatives have already been completed and included in an EIS or supplemental EIS for Limerick, Comanche Peak, and Watts Bar. Therefore, severe accident mitigation alternatives need not be reconsidered for these plants for license renewal.Based on the fact that a generic consideration of mitigation is not performed in the GEIS, a Category 1 designation for severe accidents cannot be made. Therefore, the Commission has reclassified severe accidents as a Category 2 issue, requiring only that alternatives to mitigate seyere accidents be considered for those plants that have not included such a consideration in a previous EIS or supplemental EIS. The Commission notes that upon completion of its IPE/IPEEE progiam, it may review the issue of severe accident mitigation for license renewal and consider, by separate rulemaking, reclassifying severe accidents as a Category 1 issue.The Commission does not intend to prescribe by rule the scope of an acceptable consideration of severe accident mitigation alternatives for license renewal nor does it intend to mandate consideration of alternatives identical to those evaluated previously.

In general, the Commission expects that significant efficiency can be gained by using site-specific IPE and IPEEE results in the consideration of severe accident mitigation alternatives.

The IPEs and IPEEEs are essentially site-specific PRAs that identify probabilities of core damage (Level 1 PRA) and include assessments of containment performance under severe accident conditions that identify probabilities of fission product releases (Level 2 ). As discussed in Generic Letter 88-20,"Individual Plant Examination for Severe Accident Vulnerabilities" (November 23, 1988), one of the important goals of the IPE and IPEEE was to reduce the overall probabilities of core damage and fission product releases as necessary by modifying hardware and procedures to help prevent or mitigate severe accidents.

Although Level 3 PRAs have been used in SAMDA analyses to.generate site-specific offsite dose estimates so that the cost-benefit of mitigation alternatives could be determined, the Commission does not believe that site-specific Level 3 PRAs are required to determine whether an alternative under consideration will provide sufficient benefit to justify its cost. Licensees can use other quantitative approaches for assigning site-specific risk significance to IPE results and judging whether a mitigation alternative provides a sufficient reduction in core damage frequency (CDF) or release frequency to warrant implementation.

For example, a licensee could use information provided in the GEIS analysis (exposure indices, wind frequencies, and demographics) to translate the dominant contributors to CDF and the large release frequencies from the IPE/IPEEE results into dose estimates so that a cost-benefit determination can be performed.

In some instances, a consideration of the magnitude of reduction in the site-specific CDF and release frequencies alone (i.e., no conversion to a dose estimate) may be sufficient to conclude that no significant reduction in off-site risk will be provided and, therefore, implementation of a mitigation alternative is not warranted.

The Commission will review each severe accident mitigation consideration provided by a license renewal applicant on its merits and determine whether it 28482 Federal Register / Vol. 61, No. 109 / Wednesday, June 5, 1996 / Rules and Regulations constitutes a reasonable consideration of severe accident mitigation alternatives.

10. Decommissioning Concern. Several commenters requested further clarification of the NRC's position regarding decommissioning requirements, especially whether the total impacts address returning the site to green field conditions.

Response.

The decommissioning chapter of the GEIS analyzes the impact that an additional 20 years of plant operation would have on ultimate plant decommissioning; it neither serves as the generic analysis of the environmental impacts associated with decommissioning nor establishes decommissioning requirements.

An analysis of the expected impacts from plant decommissioning was previously provided in NUREG-0586, "Final-Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities" (August 1988). The analysis in the GEIS for license renewal examines the physical requirements and attendant effects of decommissioning after a 20-year license renewal compared with decommissioning at the end of 40 years of operation and finds little difference in effects.With respect to returning a site to green field condition, the Commission defines decommissioning as the safe removal of a nuclear facility from.service, the reduction of residual contamination to a level that permits release of the property for unrestricted use, and termination of the license.Therefore, the question of restoring the land to a green field condition, which would require additional demolition and site restoration beyond addressing residual contamination and radiological effects, is outside the current scope of the decommissioning requirements.

Moreover, consistent with the Commission's conclusion that license renewal is not expected to affect future decommissioning, any requirement relative to returning a site to a green field and the attendant effects of such a requirement would also not be affected by an additional 20 years of operation.

Therefore, the issue of returning a site to pre-construction conditions is beyond the scope of license renewal review.* Concern. Several commenters expressed concern that, because a residual radioactivity rule is still not in place, the LLW estimates should be reexamined.

Response.

The NRC does have criteria in place for the release of reactor facilities to unrestricted access following decommissioning.

These include the guidance in Regulatory Guide 1.86, "Termination of Operating Licenses for Nuclear Reactors" (which provides guidance for surface contamination), dose rate limits from gamma-emitting radionuclides included in plant technical specifications, and requirements for keeping residual contamination as low as reasonably achievable (ALARA) as included in 10 CFR part 20. These criteria were used in developing NUREG-0586, the final GEIS on decommissioning of nuclear facilities, which was published in August of 1988. One conclusion from the analysis conducted for NUREG-0586 was that waste volumes from decommissioning of reactors are not highly sensitive to the radiological criteria.

A proposed rule dated August.22, 1994, would codify radiological criteria for unrestricted release of reactors and other nuclear facilities and for termination of a facility license following decommissioning.

NUREG-1496, the draft GEIS for the proposed rule on radiological criteria, included analyses of a range of radiological release criteria and confirmed the earlier conclusions that waste volumes from decommissioning of reactors are not sensitive to the residual radiological criteria within the range likely to be selected.

This range included residual dose levels comparable to the radiological criteria currently being used for reactor decommissioning.

Based on the insensitivity of the waste volume from reactor decommissioning to the radiological criteria, the Commission continues to believe, as concluded in the decommissioning section of the GEIS, that the contribution to environmental impacts of decommissioning from license renewal are small. The Commission further concludes that these impacts are not expected to change significantly as a result of the ongoing rulemaking.

Therefore, the determinations in the GEIS remain appropriate.

11. Need for Generating Capacity Concern. In addition to the major procedural concern discussed earlier about the treatment of need for generating capacity, several commenters raised concerns about the power demand projections used in the GEIS.Some commenters noted that any determination of need quickly becomes dated and, therefore, the demand for and the source of electrical power at the time of license renewal cannot be accurately predicted at this time.Moreover, they believe that the NRC's *analysis is not definitive enough to remain unchallenged for 40 years.Another commenter criticized the analysis because it focused only on energy requirements without making appropriate distinctions between energy and peak capacity requirements, plant availability, and capacity factors.Response.

The NRC has determined that a detailed consideration of the need for generating capacity is inappropriate in the context of consideration of the environmental impacts of license renewal. Thus, the NRC will limit its NEPA review of license renewal applications to the consideration of the environmental impacts of license renewal compared with those of other available generating sources. Hence, the concerns regarding demand projections used in the draft GEIS are no longer an issue and they.have been removed from the GEIS.12. Alternatives to License Renewal Concern. in addition to the procedural concern discussed earlier about the treatment of alternative energy sources as a Category I issue, several commenters expressed concerns about the comparison and analysis of alternative energy sources, as well as the economic analysis approach used in the draft GEIS. Consistent with their arguments against the Category 1 designation of alternatives, the commenters questioned the approach adopted in the GElS of comparing only single alternative energy sources to license renewal. They believe that the NRC's failure to consider a mix of alternatives ignores the potential for other alternative sources -of power that are available to different regions of the nation,, such as demand-side management, cogeneration, purchased power from Canada, biomass, natural gas, solar energy, and wind power. They also indicated that this approach neglects a utility's ability to serve its customers with a portfolio of supply that is based on load characteristics, cost, geography, and other considerations, and fails to consider the collective impact of the alternatives.

Furthermore, the possible technological advances in renewable energy sources over the next 40 years are not addressed.

One commenter argued that designating the issue of alternative energy sources as Category 1 allows a license renewal applicant not to consider the additional requirement of economic threshold analysis.

Relative to the economic analysis of the alternatives to license renewal, another commenter questioned the proposed requirement for the license renewal applicant to demonstrate that the "replacement of equivalent generating capacity by a coal-fired plant has no demonstrated cost advantage over the individual nuclear power plant license renewal."

Federal Register / Vol. 61, No. 109 / Wednesday, June 5,'1996 / Rules and Regulations 28483 According to the corifmenter, this requirement would force the applicant to perform an economic analysis of an alternative to license renewal. The commenter further argued that NEPA does not require an economic consideration.

Response.

In response to these concerns, the final rule no longer requires a cost comparison of alternative energy sources relative to license renewal. Furthermore, the alternative energy sources discussed in the final GEIS include energy conservation and energy imports as well as the other* sources discussed by the commenters.

An analysis of the environmental impacts of alternative energy sources is included in the GEIS but is not codified in 1O CFR part 51.The NRC believes that its consideration of alternatives in the GEIS is representative of the technologies available and the associated environmental impacts. With regard to'consideration of a mix of alternative sources, the Commission recognizes that combinations of various alternatives may be used to replace power generation from license renewal.13. License Renewal Scenario Concern. Several' commenters raised concerns related to the license renewal scenario evaluation methodology as implemented in the GEIS. The fundamental issues were the degree of.conservatism built into the scenario and the appropriateness of an upper bound type approach in characterizing the , refurbishment activities (and associated costs) in light of NEPA requirements to determine reasonable estimates of the environmental impacts of Federal actions.Regarding the concerns that the refurbishment schedules and scenarios developed for the GEIS were too conservative, several commenters indicated that many of the activities slated for completion during the extended refurbishment before license renewal would actually be completed by many facilities during the course of the current licensing term. The effect of having only one major outage instead of'leveling work over three or four outages could lead to an over-estimate of the refurbishment activities and costs that any particular plant would expect to see.Response.

In response to this concern, the NRC has revised the GEIS to include two license renewal program scenarios.

The first scenario refers to a "typical" license renewal program and is intended to be representative of the type of programs that many plants seeking license renewal might implement.

The second scenario retains the original ,objective of establishing an upper bound of the impacts likely to be generated at any particular plant. The typical scenario is useful for estimating impacts at plants that have been well maintained.

and have already undertaken most major refurbishment activities necessary for operation beyond the current licensing term. The conservative scenario estimates continue to be useful for estimating the maximum impacts likely to result from license renewal.The revised approach of providing two separate license renewal scenarios also alleviates the concern about the use of a bounding scenario for license renewal activities.

The NRC acknowledges that some applicants for license renewal may not be required to perform certain major refurbishment or replacement activities and, therefore, may have fewer or shorter outages.However, the two scenarios described in the GElS are neither unrealistic nor overconservative in representing the range of activities that could be expected for license renewal and the possible schedule for performing these activities.

14. Environmental Justice On February 11, 1994, the President issued Executive Order (E.Q.) 12898,"Federal Actions To Address Environmental Justice in Minority Populations and Low-Income Populations" (59 FR 7629, February 16, 1994). This order requires each Federal agency to make achieving environmental justice part of its mission by identifying and addressing, as appropriate, disproportionately high and adverse human health or environmental effects'of its programs, policies, and activities on minority and low income populations.

The Commission will endeavor to carry out the measures set forth in the executive order by integrating environmental justice into NRC's compliance with the National Environmental Policy of 1969 (NEPA), as amended. E.O. 12898 was issued after publication of the proposed rule and the receipt of comments on the proposed rule. As a result, no comments were received regarding environmental justice reviews for license renewal.Therefore, a brief discussion of this issue relative to license renewal is warranted:

As called for in Section 1-102 of E.O.12898, the EPA established a Federal interagency working group to, among other things, "* *

provide guidance to Federal agencies or criteria for identifying disproportionately high and adverse human health or environmental effects on minority populations and low-income populations

* *." The CEQ was assigned to provide this guidance to enable agencies to better comply with E.O. 12898. Until the CEQ guidance is received, the Commission intends to consider environmental justice in its evaluations of individual license renewal applications.

Greater emphasis will be placed on discussing impacts on minority and low-income populations when preparing NEPA documents such as EISs, supplemental EISs, and, where appropriate, EAs.Commission requirements regarding environmental justice reviews will be reevaluated and may be revised after receipt of the CEQ guidance.IV. Discussion of Regulatory Requirements A. General Requirements In this final rule, the regulatory requirements for performing a NEPA review for a license renewal application are similar to the NEPA review requirements for other major plant licensing actions. Consistent with the current NEPA practice for major plant licensing actions, this amendment to 10'CFR Part 51 requires the applicant to submit an environmental report that analyzes the environmental impacts associated with the proposed action, considers alternatives to the proposed action, and evaluates any alternatives for reducing adverse environmental effects. Additionally, the amendment requires the NRC staff to prepare a supplemental environmental impact statement for the proposed action, issue the statement in draft for public comment, and issue a final statement after considering public comments on the draft.The amendment deviates from NRC's current NEPA review practice in some areas. First, 'the amendment codifies certain environmental impacts associated with license renewal that wereanalyzed in NUREG-1437,"Generic Environmental Impact Statement for License Renewal at Nuclear Plants" (xxxx 1996).Accordingly, absent new and significant information, the analyses for certain impacts codified by this rulemaking need only be incorporated by reference in an applicant's environmental report for license renewal and in the Commission's (including NRC staff, adjudicatory officers, and the Commission itself) draft and final SEIS and other environmental documents developed for the proceeding.

Secondly, the amendment reflects the Commission's decision to limit its NEPA review for license renewal to a consideration of the environmehtal 28484 Federal Register / Vol. 61, No. 109 / Wednesday, June 5, 1996 / Rules and Regulations effects of the proposed action and alternatives to the proposed action.Finally, the amendment contains the decision standard that the Commission will use in determining the acceptability of the environmental impacts of individual license renewals.The Commission and the applicant will consider severe accident mitigation alternatives to reduce or mitigate environmental impacts for any plant for*which severe accident mitigation alternatives have not been previously considered in an environmental impact statement or related supplement or in an environmental assessment.

The Commission has concluded that, for license renewal, the issues of need for power and utility economics should be reserved for State and utility officials to decide. Accordingly, the NRC will not conduct an analysis of these issues in the context of license renewal or perform traditional cost-benefit balancing in license renewal NEPA reviews. Finally, in a departure from the approach presented in the proposed rule, this final rule does not codify any conclusions regarding the subject of.alternatives.

Consideration of and decisions regarding alternatives will occur at the site-specific stage. The discussion below addresses the specific regulatory requirements of this S.amendment and any conforming changes~to 10 CFR part 51 to implement the Commission's decision to eliminate cost-benefit balancing from license renewal NEPA reviews.B. The Environmental Report 1. Environmental Impacts of License Renewal Through this final rule, the NRC has amended 10 CFR 51.53 to require an* applicant for license renewal to submit an environmental report with its application.

This environmental report must contain an analysis of the environmental impacts of renewing a license, the environmental impacts of alternatives, and mitigation alternatives.

In preparing the analysis of environmental impacts contained in the environmental report, the applicant should refer to the data provided in appendix B to 10 CFR part 51, which has been added to NRC's regulations as part of this rulemaking.

The applicant is not required to provide an analysis in the environmental report of those issues identified as Category 1 issues in Table B-1 in Appendix B. For those issues identified as Category 2 in Table B-i, the applicant must provide a specified additional analysis beyond that contained in Table B-1. In this final rule, 10 CFR 51.53(c)(3)(ii) specifies the subject areas of the analysis that must be addressed for the Category 2 issues.Pursuant to 10 CFR 51.45(c), 10 CFR 51.53 (c) (2) requires the applicant to consider possible actions to mitigate the adverse impacts associated with the proposed action. This consideration is limited to designated Category 2 matters. Pursuant to 10 CFR 51.45 (d), the environmental report must include a discussion of the status of compliance with applicable Federal, State, and local environmental standards.

Also, 10 CFR 51.53(c) (2) specifically excludes from consideration in the environmental report the issues of need for power, the economic costs and benefits of the proposed action, economic costs and benefits of alternatives to the proposed action, or other issues not related to environmental effects of the proposed action and associated alternatives.

In addition, the requirements in 10 CFR 51.45 are consistent with the exclusion of economic issues in 10 CFR 51.53(c)(2).

2. Consideration of Alternatives Pursuant to 10 CFR 51.45(c), 10 CFR 51.53(c) (2) requires the applicant to consider the environmental impacts of alternatives to license renewal in -the environmental report. The treatment of alternatives in the environmental report should be limited to the environmental impacts of such alternatives.

The amended regulations do not require a discussion of the economic costs and benefits of these alternatives in the environmental report for the operating license renewal stage except.as necessary to determine whether an alternative should be included in the range of alternatives considered or whether certain mitigative actions are appropriate.

The analysis should demonstrate consideration of a reasonable set of alternatives to license renewal. In preparing the- alternatives analysis, the applicant may consider-information regarding alternatives in NUREG-1437, "Generic Environmental Impact Statement for License Renewal of Nuclear Plants" (xxxx 1996).The Commission has developed a new decision standard to be applied in environmental impact statements for license renewal as discussed in Section IV.C.2. The amended regulations for license renewal do not require applicants to apply this decision standard to the information generated in* their environmental report (although the applicant is not prohibited from doing so if it desires).

However, the NRC staff* will use the information contained in the environmental report in preparing the environmental impact statement upon which the Commission will base its final decision.3. Consideration of Mitigation Alternatives Consistent with the NRC's current NEPA practice, an applicant must include a consideration of alternatives to mitigate adverse environmental impacts in its environmental report.However, for license renewal, the Commission has generically considered mitigation for environmental issues associated with renewal and has concluded that no additional site-specific consideration of mitigation is necessary for many issues. The Commission's consideration of mitigation for each issue included identification of current activities that adequately mitigate impacts and evaluation of other mitigation techniques that might or might not be warranted, depending on such factors as the size of the impact and the cost of the technique.

The Commission has considered mitigation for all impacts designated as Category 1 in Table B-1.Therefore, a license renewal applicant need not address mitigation for issues so designated.

C. Supplemental Environmental Impact Statement This amendment also requires that the Commission prepare a supplemental environmental impact statement (SEIS), consistent with 10 CFR 51.20(b) (2). This statement will serve as the Commission's independent analysis of the environmental impacts of license renewal as well as a comparison of these impacts to the environmental impacts of alternatives.

This document willalso present the preliminary recommendation by the NRC staff regarding the proposed action.Consistent with the revisions to 10 CFR 51.45 and 51.53 discussed above in regard to the applicant's environmental report, this rulemaking revises portions of 10 CFR 51.71 and 51.95 to reflect the Commission's approach to addressing the environmental impacts of license renewal.The issues of need for power, the economic costs and benefits of the proposed action, and economic costs and benefits of alternatives to the proposed action are specifically excluded from consideration in the supplemental environmental impact statement for license renewal by 10 CFR 51.95(c), except as these costs and benefits are either essential for a determination regarding the inclusion of an alternative in the range of alternatives considered or relevant to mitigation.

The supplemental Applicant's Consolidated Memorandum in Opposition to New York State's Motion-for Summary Disposition of Contention NYS-35/36 and in Support of Its Cross-Motion for Summary Disposition (February 3, 2011)NYS-35/36:

ATTACHMENT 7

NEI 05.01 [Rev A]Nuclear Energy Institute Severe! Accident Migo Alternatives (SAMA). Analysis Guidance Document November 2005 Nuclear Energy Institute, 17761 Street N. W., Suite 400, Washington D.C. (202.739.8000)

EXECUTIVE

SUMMARY

This document provides a template for completing the severe accident mitigation alternatives (SAMA) analysis in support of license renewal. Its purpose is to identify the information that should be included in the SAMA portion of a license renewal application environmental report to reduce the necessity for Nuclear Regulatory Commission (NRC) requests for additional information (RAIs). The method described relies upon NUREG/BR-0184 regulatory analysis techniques, is a result of experience gained through past SAMA analyses, and incorporates insights gained from review of NRC evaluations of SAMA analyses and associated RAls.i NEI 05-01 (Rev A)November 2005 DCPWR This analysis case was used to evaluate plant modifications that would increase the availability of Class JE DC power (e.g., increased battery capacity or the installation of a diesel-powered generator that would effectively increase battery capacity).

Although the proposed SAMAs would not completely eliminate the potential failure, a bounding benefit was estimated by removing the battery discharge events and battery failure events. This analysis case was used to model the benefit of SAMAs'4, 5, 10, 12, and 24.7.1.2 COST OF SEVERE ACCIDENT RISK WITH SAMA IMPLEMENTED Using the risk measures from Section 7.1.1, calculate severe accident impacts in four areas: off-site exposure cost, off-site economic cost, on-site exposure cost, and on-site economic cost using the same procedure used for the baseline ease described in Section 4.As in Section 4.5, sum the severe accident impacts and combine with the external events multiplier (Section 3.1.2.4) to estimate the total, cost of severe accident risk with the SAMA implemented.

Use of the external events multiplier is inappropriate for some SAMAs. For example, SAMAs specifically related to external events that would not impact internal events (e.g., enhanced fire detections) and SAMAs related to specific internal event initiators (e.g., guard pipes for main steam line break events). Provide a discussion of SAMAs on which the external events multiplier was not applied.7.1.3 SAMA BENEFIT Subtract the total cost of severe accident risk with the SAMA implemented from the baseline cost of severe accident risk (maximum benefit from Section 4:5) to obtain the benefit.List the estimated benefit. for each SAMA candidate.

Table 11 provides a sample portion of a Phase II SAMA candidate list with estimated benefits listed.7.2 COST OF SAMA IMPLEMENTATION Perform a cost estimate for each of the Phase II SAMA candidates.

Describe the cost estimating process and list the cost estimate for each SAMA candidate.

As SAMA analysis focuses on establishing the economic viability of potential plant enhancement.

when compared to attainable benefit, often detailed cost estimates are not required to make informed decisions regarding the economic viability of a particular modification.

SAMA implementation costs may be clearly in excess of the attainable benefit estimated from a particular analysis case. For less clear cases, engineering judgment may be applied to determine if a more detailed cost estimate is necessary to formulate a conclusion regarding the economic viability of a particular SAMA. Nonetheless, the cost of each SAMA candidate should be conceptually estimated to the point where economic viability of the proposed modification can be adequately gauged.28 NET 05-01 (Rev A)November 2005 For hardware modifications, the cost of implementation may be established from existing estimates of similar modifications from previously performed SAMA and SAMDA analyses.Costs associated with implementation of a SAMA including procurement, installation, long-term maintenance, surveillance, calibration, and training should be considered.

Discuss conservatisms in the cost estimates.

For example, cost estimates may not include the cost of replacement power during extended outages required to implement the modifications.

They also may not include contingency costs associated with unforeseen implementation obstacles.

Estimates based on modifications that were implemented or estimated in the past may be presented in terms of dollar values at the time of implementation (or estimation), and not adjusted to present-day dollars. In addition, implementation costs originally developed for SAMDA analyses (i.e., during the design phase of the plant) do not capture the additional costs associated with performing design modifications to existing plants (i.e., reduced efficiency, minimizing dose, disposal of contaminated material, etc.).Table 11 provides a sample portion of a Phase II SAMA candidate list with cost estimates.

29 NEI 05-01 (Rev A)November 2005 8 SENSITIVITY ANALYSES Evaluate how changes in SAMA analysis assumptions would affect the cost-benefit analysis.Perform the following sensitivity analyses, as applicable.

Table 12 contains sample sensitivity analysis results.8.1 PLANT MODIFICATIONS Major changes to the plant, such as power uprate or. steam generator replacement, may be planned or may have occurred since the model freeze date, as described in Section 3.1 and Section 3.2. If the Level I or Level 2 PSA model used for the SAMA analysis does not address a major plant change, a sensitivity analysis may be performed to support discussion of the impact of the change on the SAMA analysis results.In this sensitivity analysis, modify the PSA model (or its results) to simulate incorporation of the plant modification and perform the Phase II analysis with the revised severe accident risk results.Sufficient margin exists in the maximum benefit estimation that the Phase I screening should not have to be repeated in the sensitivity analysis.Discuss the plant modification and how its effects were simulated in the PSA model. Provide pertinent results and discuss how they affect the conclusions of the SAMA analysis.

If SAMAs appear cost-beneficial .in the sensitivity results, discussion of conservatisms in the analysis, (e.g., conservatisms in cost estimates discussed in Section 7.2), and their impact on the results may be appropriate.

8.2 UNCERTAINTY

A discussion of CDF uncertainty, and conservatisms in the SAMA analysis that off-set uncertainty, should be included.

For example, use of conservative risk modeling to represent a particular plant change may be used to offset uncertainty in risk modeling; use of conservative implementation cost estimates may be used to offset uncertainty in cost estimates; and use of an uncertainty factor derived from the ratio of the 9 5 th percentile to the mean point estimate for internal events CDF may be used to account for CDF uncertainties.

Estimate an uncertainty factor based on this discussion and perform a sensitivity analysis using the uncertainty factor on the results. [Based on analysis to date the ratio of the 95dh percentile to the mean point estimate for typical internal events CDF values is 2 to 5 (Reference 1).]Provide pertinent results and discuss how they affect the conclusions of the SAMA analysis.

If SAMAs appear cost-beneficial in the sensitivity results, discussion of conservatisms in the analysis, (e.g., conservatisms in cost estimates discussed in Section 7.2), and their impact on the results may be appropriate.

30 Applicant's Consolidated Memorandum in Opposition to New York State's Motion for Summary Disposition of Contention NYS-35/36 and in Support of Its Cross-Motion for Summary Disposition (February 3, 2011)NYS-35/36:

ATTACHMENT 8

August 2, 2007 Mr. James H. Riley Nuclear Energy Institute 1776 I Street, NW, Suite 400 Washington, DC 20006-3708

SUBJECT:

FINAL LICENSE RENEWAL INTERIM STAFF GUIDANCE LR-ISG-2006-03:

STAFF GUIDANCE FOR PREPARING SEVERE ACCIDENT MITIGATION ALTERNATIVES ANALYSES

Dear Mr. Riley:

By letter dated August 10, 2006, we issued ourproposed License Renewal Interim Staff Guidance, LR-ISG-2006-03, for public comments.

By letter dated September 15, 2006, Dr. Jill Lipoti, State of New Jersey, Department of Environmental Protection, provided comments and requested that the staff take this opportunity to include a review of the impacts of terrorist attacks on nuclear facilities as part of the National Environmental Policy Act in the interim staff guidance.

The staff reviewed Dr. Lipoti's comments and provided responses in a letter dated July 31, 2007, which can be found in the Agencywide Document Access and Management System (ADAMS) under Accession Number ML071380307.

No changes were incorporated in the Final LR-ISG-2006-03 based on these comments.Enclosure 1 is our Final LR-ISG-2006-03.

A notice relating to this Final LR-ISG will be published in the Federal Register.

If you have any questions regarding this matter, please contact Mr. Richard L. Emch, Jr., by telephone at 301-415-1590 or by e-mail at RLE(nrc.qov, An identical letter was sent to Mr. David Lochbaum at the Union of Concerned Scientists.

Sincerely;'

IRA!Pao-Tsin Kuo, Director Division of License Renewal Office of Nuclear Reactor Regulation Project No. 690

Enclosure:

Final LR-ISG-2006-03 cc w/encl: See next page DISTRIBUTION:

See next page Adams Accession No. ML071640,133 OFFICE PM:DLR:REBB LA:DLR PM:DLR:RLRB OGC NAME REmch YEdmonds ILTran SUttal (NLO w/comments)DATE 06/15/07 06/26/07 107/3/07 07/13/07 OFFICE BC:DLR:REBB BC:DLR:RRLRB I D:DLR NAME I RFranovich RAuluck PTKuo DATE 07/23/07 07/27/07 08/2/07 OFFICIAL RECORD COPY Letter to: James H. Riley, David Lochbaum, from: Pao-Tsin Kuo. dated: Augqust 2, 2007

SUBJECT:

FINAL LICENSE RENEWAL INTERIM STAFF GUIDANCE LR-ISG-2006-03:

STAFF GUIDANCE FOR PREPARING SEVERE ACCIDENT MITIGATION ALTERNATIVES ANALYSES HARD COPY DLR RF REmch LTran E-MAIL: PUBLIC MRubin RPalla RidsNrrOd RidsNrrAdro RidsNrrAdra RidsNrrAdes RidsNrrDIr RidsNrrDra RidsNrrDnrl RidsNrrDlrRebb RidsNrrDIrReba DLR Staff FINAL LICENSE RENEWAL INTERIM STAFF GUIDANCE LR-ISG-2006-03:

STAFF GUIDANCE FOR PREPARING SEVERE ACCIDENT MITIGATION ALTERNATIVES ANALYSES Introduction A Severe Accident Mitigation Alternatives (SAMA) analysis is required as part of a license renewal application, if a SAMA analysis has not already been performed for the plant and reviewed by the NRC staff. SAMA analyses have been performed and submitted to the NRC for all applications for license renewal received by the staff thus far. Therefore, this LR-ISG is being recommended as guidance consistent with our goal to more effectively and efficiently resolve license renewal issues identified by the staff or the industry.Back~ground and Discussion After receiving extensive requests for additional information regarding the SAMA analyses, several applicants for license renewal concluded that they did not fully understand the kind of information that the NRC staff was expecting to see in SAMA analyses.The Nuclear Energy Institute (NEI) developed a generic guidance document to help clarify the NRC staffs expectations regarding the information that should be submitted in SAMA analyses.On April 8, 2005, NEI submitted NEI 05-01, "Severe Accident Mitigation Alternatives (SAMA)Analysis -Guidance Document." The NRC staff reviewed this guidance document, and by letter, dated July 12, 2005, provided comments on NEI 05-01. The NRC staffs comments were discussed during a public meeting between NEI and NRC on July 21, 2005.On February 17, 2006, NEI submitted its NEI 05-01, Revision A, dated November 2005. The NRC staff reviewed and concluded that this versionfully resolved the NRC staff's comments.

In addition, the NRC staff concluded that NEI 05-01, Revision.A, describes existing NRC regulations and facilitates complete preparation of SAMA analysis submittals.

Some applicants for license renewal have submitted SAMA analyses using the guidance provided in NEI 05-01, Revision A. The NRC staff found improved quality in the submitted SAMA.analyses and a reduction in the number of requests for additional information for those applications.

that followed the guidance provided in NEI 05-01, Revision A.Recommended Action*The staff is recommending that applicants for license renewal follow the guidance provided in NEI 05-01, Revision A, when preparing their SAMA analyses.

The staff finds that NEI 05-01, Revision A, describes existing NRC regulations, and facilitates complete preparation of SAMA analysis submittals.

Although this final LR-ISG does not convey a change in the NRC's regulations or how they are interpreted, it is being provided to facilitate complete preparation of future SAMA analysis submittals in support of applications for license renewal. The NRC staff plans to incorporate the guidance provided in NEI 05-01, Revision A, into a future update of Supplement 1 to Regulatory Guide 4.2, "Preparation of Supplemental Environmental Reports for Applications to Enclosure Renew Nuclear Power Plant Operating Licenses." This LR-ISG provides a clarification of existing guidance with no additional requirements.

For those that are interested in reviewing NEI 05-01, Revision A, the Agencywide Documents Access and Management System (ADAMS)Accession Number is ML060530203.

Applicant's Consolidated Memorandum in Opposition to New York State's Motion for Summary Disposition of Contention NYS-35/36 and in Support of Its Cross-Motion for Summary Disposition (February 3, 2011)NYS-35/36:

ATTACHMENT 9

45466 Federal Register/Vol.

72, No. 156/Tuesday, August 14, 2007/Notices clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated July 20, 2007..No significant hazards consideration comments received:

No..Dated at Rockville, Maryland, this 2nd day of August 2007.For the Nuclear Regulatory Commission.

Catherine Haney, Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation.

[FR Doc. E7-15459 Filed 8-13-07; 8:45 am]BILLING'CODE 7590-01-P NUCLEAR REGULATORY COMMISSION Notice of Availability of the Final License Renewal Interim Staff Guidance LR-ISG-2006-03:

Staff Guidance for Preparing Severe Accident Mitigation Alternatives Analyses AGENCY: Nuclear Regulatory Commission (NRC).ACTION: Notice of Availability.

SUMMARY

NRC is issuing its Final License Renewal Interim Staff Guidance LR-ISG-2006-03 for preparing severe accident mitigation alternatives (SAMA)analyses.

This LR-ISG recommends that applicants for license renewal use the Guidance Document Nuclear Energy Institute 05-01, Revision A, (ADAMS Accession No. ML060530203) when preparing their SAMA analyses.

The NRC staff issues' LR-ISGs to facilitate timely implementation of the license renewal rule and to review activities associated with a license renewal application.

The NRC staff will also incorporate the approved LR-ISG into the next revision of Supplement I to Regulatory Guide 4.2, "Preparation of Supplemental Environmental Reports for Applications to Renew Nuclear Power Plant Operating Licenses." ADDRESSES:

The NRC maintains an Agencywide Documents Access and Management System (ADAMS), which provides text and image files of NRC's public documents.

These documents may be accessed through the NRC's Public Electronic Reading Room on the Internet at http://www.nrc.gov/reading-rm/adams.html.

Persons who do not have access to ADAMS or who encounter problems in accessing the documents located in ADAMS should contact the NRC Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by e-mail at pdri@nrc.gov.

FOR FURTHER INFORMATION CONTACT: Mr..Richard L. Emch, Jr., Senior Project Manager, Office Of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone 301-415-1590 or by e-mail at rle@nrc.gov.

SUPPLEMENTARY INFORMATION:

Attachment 1 to this Federal Register notice, entitled Staff Position and Rationale for the Final License Renewal Interim Staff Guidance LR-ISG-2006-03: Staff Guidance for Preparing Severe Accident Mitigation Alternatives (SAMA) Analyses contains the NRC staff's rationale for publishing the Final LR-ISG-2006-03.

Attachment 2 to this Federal Register notice, entitled Proposed License Renewal Interim Staff Guidance LR-ISG-2006-03:

Staff Guidance for Preparing Severe Accident Mitigation Alternatives (SAMA)Analyses, contains the guidance for preparing SAMA analyses related to license renewal applications.

The NRC staff approves this LR-ISG for NRC and industry use. The NRC staff will also incorporate the approved LR-ISG into the next revision of Supplement 1 to Regulatory Guide 4.2, "Preparation of Supplemental Environmental Reports for Applications to Renew Nuclear Power Plant Operating Licenses." Dated at Rockville, Maryland, this 2nd day of August 2007.For the Nuclear Regulatory Commission.

Pao-Tsin Kuo, Director, Division of License Renewal, Office of Nuclear Reactor Regulation.

Attachment 1-Staff Position and Rationale for the Final License Renewal Interim Staff Guidance LR-ISG-2006-03: Staff Guidance for Preparing Severe Accident Mitigation Alternatives Analyses Staff Position:

The NRC staff recommends that applicants for license renewal follow the guidance provided in Nuclear Energy Institute (NEI) 05-01,"Severe Accident Mitigation Alternatives (SAMA) Analysis-Guidance Document," Revision A, when preparing their SAMA analyses.Rationale:

The NEI developed a generic Guidance Document NEI 05-01, Revision A,.to help clarify the NRC staff's expectations regarding the information that needs to be included in SAMA analyses.

The NRC staff reviewed and concluded that NEI 05-01, Revision A, describes existing NRC regulations and facilitates complete preparation of SAMA analysis submittals.

The staff finds that utilization of the guidance provided in NEI 05-01, Revision A, will result in improved quality in SAMA analyses and a reduction in the number of requests for additional information.

Attachment 2-Final License Renewal Interim Staff Guidance LR-ISG-2006-03: Staff Guidance for Preparing Severe Accident Mitigation Alternatives Analyses Introduction A severe accident mitigation alternatives (SAMA) analyses is required as part of a license renewal application, if a SAMA analysis has not already been performed for the plant and reviewed by the NRC staff. SAMA analyses have been performed and submitted to the NRC for all applications for license renewal received by the staff thus far. Therefore, this LR-ISG is being recommended as guidance consistent with our goal to more effectively and efficiently resolve license renewal issues identified by the*staff or the industry.Background and Discussion After receiving extensive requests for additional information regarding the SAMA analyses, several applicants for license renewal concluded that they did not fully understand the kind of information that the NRC staff was expecting to see in SAMA analyses.The Nuclear Energy Institute (NEI)developed a generic guidance document to help clarify the NRC staff's expectations regarding the information that should be submitted in SAMA analyses.

On April 8, 2005, NEI submitted NEI 05-01, "Severe Accident Mitigation Alternatives (SAMA)Analysis-Guidance Document." The NRC staff reviewed this guidance document, and by letter, dated July 12, 2005, provided comments on NEI 05-01. The NRC staff's comments were discussed during a public meeting between NEI and NRC on July 21, 2005.On February 17, 2006, NEI submitted its NEI 05-01, Revision A, dated , November 2005. The NRC staff reviewed and concluded that this version fully resolved the NRC staff s comments.

In addition, the NRC staff concluded that NEI 05-01, Revision A, describes existing NRC regulations, and facilitates complete preparation of SAMA analysis submittals.

Some applicants for license renewal have submitted SAMA analyses using the guidance provided in NEI 05-01, Revision A. The NRC staff found improved quality in the submitted SAMA analyses and a reduction in the'I Federal Register/Vol.

72, No. 156/Tuesday, August 14, 2007/Notices 45467 number of requests for additional information for those applications that followed the guidance provided in NEI 05-01, Revision A.Recommended Action The staff is recommending that applicants for license renewal follow the guidance provided in NEI 05-01, Revision A, when preparing their SAMA analyses.

The staff finds that NEI 05-01, Revision A, describes existing NRC regulations, and facilitates complete preparation of SAMA analysis submittals.

Although this proposed LR-ISG does not convey a change in the NRC's regulations or how they are interpreted, it is being provided to facilitate complete preparation of future SAMA analysis submittals in support of applications for license renewal. The NRC staff plans to incorporate, the guidance provided in NEI 05-01, Revision A, into a future update of Supplement 1 to Regulatory Guide 4.2,"Preparation of Supplemental Environmental Reports for Applications to Renew Nuclear Power Plant Operating Licenses." This LR-ISG provides a clarification of existing guidance with no additional requirements.

For those that are interested in reviewing NEI. 05-01, Revision A, the Agencywide Documents Access and Management System (ADAMS) Accession Number is ML060530203.

[FR Doc. E7-15926 Filed 8-13-07; 8:45 am]BILLING CODE 7590-01-P FOR FURTHER INFORMATION CONTACT: Stephen L. Sharfman, General Counsel, 202-789-6820 and stephen.sharfman@prc.gov.

SUPPLEMENTARY INFORMATION:

On August 3, 2007, the United States Postal Service filed a request seeking a recommended decision from the Postal Regulatory Commission approving a Negotiated Service Agreement (NSA) with The Bradford Group.1 The NSA is proffered as functionally equivalent to the Bookspan NSA recommended by the Commission in Docket No. MC2005-3 (baseline agreement).

[70 FR 42602.]The Request, which includes six attachments, was filed pursuant to chapter 36 of title 39, United States Code.2 The Postal Service has identified The Bradford Group, along with itself, as parties to the NSA. This identification serves as notice of intervention by The Bradford Group. It also indicates that The Bradford Group shall be considered a co-proponent, procedurally and substantially, of the Postal Service's Request during the Commission's review of the NSA. Rule 191(b) 139 CFR 3001.191(b).]

An appropriate Notice of The Bradford Group of Appearance and Filing of Testimony as Co-Proponent, August 3, 2007, has been filed.In support of the direct case, the Postal Service has filed Direct Testimony of Broderick A; Parr on Behalf of the United States Postal Service, August 3, 2007 (USPS-T-1) and library reference USPS-LR-L-1, MC2004-3 Opinion and Further Recommended Decision Analysis for The Bradford Group NSA. The Bradford Group has separately filed direct testimonies of Steve Gustafson (BG-T-1) and Wendy Ring (BG-T-2) both on behalf of The Bradford Group, August 3, 2007. The Postal Service has reviewed The Bradford Group testimony and, in accordance with rule 192(b) [39 CFR 3001.192(b)], states that such testimony may be relied upon in presentation of the Postal Service's direct case. USPS-T-1 at 3.'Request of the United States Postal Service for a Recommended Decision on Classifications and Rates to Implement a Functionally Equivalent Negotiated Service Agreement with Bradford Group, August 3, 2007 (Request).

2 Attachments A and B to the Request contain proposed changes to the Domestic Mail Classification Schedule and associated rate schedules; Attachment C is a certification required by Commission rule 193(i) specifying that the cost statements and supporting data submitted by the Postal Service, which purport .to reflect the books of the Postal Service, accurately set forth the results shown by such books; Attachment D is an index of testimony and exhibits; Attachment E is a compliance statement addressing satisfaction of various filing requirements; and Attachment F is a copy of the Negotiated Service Agreement.

The Request relies on record testimony entered in the baseline docket. This material is identified in the Postal Service's Compliance Statement, Request Attachment E.Requests that are proffered as functionally equivalent to baseline NSAs are handled expeditiously, until a final determination has been made as to their proper status. The Postal Service's Compliance Statement, Request Attachment E, is noteworthy in that it provides valuable information to facilitate rapid review of the Request to aid participants in evaluating whether or not.the procedural path suggested by the Postal Service is appropriate.

The Postal Service submitted several contemporaneous related filings with its Request. The Postal Service has filed a proposal for limitation of issues in this docket.3 Rule 196(a)(6)

[39 CFR 3001.196(a)(6)].

The proposal identifies issues that were previously decided in the baseline docket, and key issues that are unique to the instant Request.Rule 196(b) 139 CFR 3001.196(b)]

requires the Postal Service to provide written notice of its Request, either by hand delivery or by First Class Mail, to all participants of the baseline docket.This requirement provides additional time, due to an abbreviated intervention period, for the most likely participants to decide whether or not to intervene.

A copy of the Postal Service's notice was filed with the Commission on August 3, 2007.4 The Request, accompanying testimonies of witnesses Parr (USPS-T-1), Gustafson (BG-T-1), and Ring (BG-T-2), the baseline agreement, and other related material can be accessed electronically, via the Internet, on the Commission's Web site (http://www.prc.gov).

I.

Background:

Baseline Bookspan Negotiated Service Agreement, Docket No. MC2005-3 If a request predicated on a NSA is found to be functionally equivalent to a previously recommended, and currently in effect, NSA, it may be afforded accelerated review. Rule 196 [39 CFR 3001.196].

The Postal Service asserts that the NSA in the instant Request is functionally equivalent to the now in effect Bookspan NSA recommended by the Commission in Docket No. MC2005-, 3 United States Postal Service Proposal for Limitation of Issues, August 3, 2007.4 Notice of the United States Postal Service Concerning the Filing of a Request for a Recommended Decision on a Functionally Equivalent Negotiated Service Agreement, August 3, 2007.POSTAL REGULATORY COMMISSION

[Docket No. MC2007-4; Order No. 23]Negotiated Service Agreement AGENCY: Postal Regulatory Commission.

ACTION: Notice and order.

SUMMARY

This document establishes a docket for consideration of the Postal Service's request for approval of contract rates with The Bradford Group.It identifies key elements of the proposed agreement, which involves Standard Mail letters and flats rates, and addresses preliminary procedural matters.DATES: 1. August 24, 2007: Deadline for intervention and responses to limitation of issues. 2. August 28, 2007: Prehearing conference, 11 a.m. in the Commission's hearing room.ADDRESSES:

Submit comments electronically via the Commission's Filing Online system at http://www.prc.gov.

Applicant's Consolidated Memorandum in .Opposition to New York State's Motion for Summary Disposition of Contention NYS-35/36 and in Support of Its Cross-Motion for Summary Disposition (February 3, 2011)NYS-35/36:

ATTACHMENT 10 vEnhtergy Enter-qy Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249 Tel (914) 788-2055 Fred Dacimo Vice President License Renewal NL-09-151 November 16, 2009 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

SUBJECT:

REFERENCE:

Entergy Nuclear Operations Inc.Telephone Conference Call Regarding Met Tower Data for SAMA Analysis Indian Point Nuclear Generating Unit Nos. 2 & 3 Docket Nos. 50-247 and 50-286.License Nos. DPR-26 and DPR-64 1. Meeting Minutes, Telephone Conference, November 9, 2009, between Ms. Kimberly Green NRC and Entergy Staff Regarding Met Tower Data for SAMA Analyses

Dear Sir or Madam'.Entergy Nuclear Operations,

Inc is providing this letter to document a discrepancy found with the wind direction inputs to the MACCS2 code for the IPEC SAMA analysis.

Entergy found the discrepancy, during investigations to resolve a question from the NRC staff regarding wind.directions.

The method used to obtain the five-year average hourlywind directions has been determined faulty. This issue has been entered into Entergy Corrective Action Program as CR-HQN-2009-01044.

IPEC will correct the wind-direction inputs, re-analyze the SAMAs for both units and provide the results to the NRC by December 16, 2009.As requested during the telephone conference summarized in reference 1, Entergy will provide the following information on or before December 16, 2009..* The meteorological data and justification supporting its use in the SAMA analysis (e.g., if a single year is used or an average of several years),* Revised estimates of the offsite population dose and offsite economic costs,* Identification of the meteorological tower elevation from which meteorological data were obtained and the rationale for selecting the data from that tower elevation,* Revised SAMA analysis results, specifically for the analysis case discussed in response to RAI 4e, dated February 5, 2008, and The complete MACCS2 input file used for the reanalysis (in electronic format).A-ooi NL-09-151 Page 2 of 2 If you have any questions, or require additional information, please contact Mr. Robert Walpole at 914-734-6710.

Sincer FRD/dmt-

Attachment:

1. Regulatory Commitment cc: Mr. Samuel J. Collins, Regional Administrator, NRC Region I Mr. Sherwin E. Turk, NRC Office of General Counsel, Special Counsel Ms. Kimberly Green, NRC Safety Project Manager Mr. John Boska, NRR Senior Project Manager, IPEC NRC Resident Inspector's Office Mr. Paul Eddy, New York State Department of Public Service Mr. Francis J. Murray, President.

and CEO, NYSERDA ATTACHMENT 1 TO NL-09-151 Regulatory Commitment ENTERGY NUCLEAR OPERATIONS, INC.INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 & 3 DOCKET NOS. 50-247 AND 50-286 LICENSE NOS. DPR-26 AND DPR-64 NL-09-151 Attachment I Docket Nos. 50-247 & 50-286 Page 1 of I Regulatory Commitment The following table identifies an action committed to by Entergy in this document.

Any other statements in this submittal are provided for information regulatory commitments.

purposes and are not considered to be COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION"_ _I AUDIT ITEM Entergy will provide the following information on or IP2: NL-09-151 Appendix E before December 16, 2009. December 16, Section 4.21 2009 The meteorological data and justification supporting its use in the SAMA analysis (e.g., IP3: if a single year is used or an average of December 16,.several years), 2009 Revised estimates of the offsite population dose and offsite economic costs, Identification of the meteorological tower.elevation from which meteorological data were obtained and the rationale for selecting the.data from that tower elevation,* Revised SAMA analysis results, specifically for the analysis case discussed in response to RAI 4e, dated February 5, 2008, and" The complete MACCS2 input file used for the reanalysis (in electronic format). L Applicant's Consolidated Memorandum in Opposition to New York State's Motion for Summary Disposition of Contention NYS-35/36 and in Support of Its Cross-Motion for Summary Disposition (February 3, 2011)NYS-35/36:

ATTACHMENT 11

Enterqy Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB PRO. Box 249 Buchanan, NY 10511-0249 Tel (914) 788-2055 Fred Dacimo Vice President License Renewal NL-09-165 December 11, 2009'U.S. Nuclear.Regulatory C6mmission Attn: Document Control D0esk Washington, DC 20555-0001

SUBJECT:

License Renewal Application

-SAMA Reanalysis Using Alternate Meteorological Tower Data Indian Point Nuclear Generating Unit Nos. 2 & 3 Docket Nos. 50-247 and 50-286 License Nos. DPR-26 and DPR-64

REFERENCE:

1. Entergy Nuclear Operations Inc. Letter NL-09-151, "Entergy Nuclear Operations Inc. Telephone Conference Call Regarding Met Tower Data for SAMA Analysis" dated November 16, 2009

Dear Sir or Madam:

In Reference 1 above, Entergy Nuclear Operations, Inc (Entergy) committed to providing the following information on or before December 16, 2009.* The meteorological data and justification supporting.its use in the SAMA analysis (e.g., if a single: year is used or an average of several years),* Revised estimates of the offsite population dose and offsite economic costs, 0 Identification of the meteorological tower elevation from which meteorological data were obtained and the rationale for selecting the data from that tower elevation,* Revised SAMA analysis results, specifically for the analysis case discussed in response to RAI 4e, dated February 5, 2008, and* The complete MACCS2 input file used for the reanalysis (in electronic format).The purpose of this letter is to transmit the requested information.

Attachment 1 provides the SAMA reanalysis using alternate Meteorological Tower Data.There are no new commitments identified in this-submittal.

If you have any questions, or require additional information, please contact Mr. Robert Walpole at 914-734-6710.

NL-09-165 Page 2 of 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on Si FRD/dmt

Enclosure:

.License Renewal Application-SAMA Reanalysis Using Alternate Meteorological Tower Data cc: Mr. Samuel J. Collins, Regional Administrator, NRC Region I Mr. Sherwin E. Turk, NRC Office of General Counsel, Special Counsel Ms. Kimberly Green, NRC Project Manager Mr. John Boska, NRR Senior Project Manager IPEC NRC Resident Inspector's Office Mr. Paul Eddy, New York State Department of Public Service Mr. Francis J. Murray, President and CEO, NYSERDA ATTACHMENT 1 TO NL-09-165 License Renewal Application

-SAMA Reanalysis Using Alternate Meteorological Tower Data ENTERGY NUCLEAR OPERATIONS, INC.INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 & 3 DOCKET NOS."50-247 AND 50-286 LICENSE NOS. DPR-26 AND DPR-64 NL-09-165 Attachment 1 Docket Nos. 50-247 &.50-286 Page 1 of 33 INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 LICENSE RENEWAL APPLICATION SAMA Reanalysis Using Alternate Meteorological Tower Data NRC Requests from November 9. 2009 Teleconference (1) Provide meteorological data and justification supporting its use in the SAMA analysis (e.g., if a single year is used or an average of several years).(2) Provide revised estimates of the offsite population dose and offsite economic costs.(3) Provide identification of the meteorological tower elevation from which meteorological data were obtained and the rationale for selecting the data from that tower elevation.

(4) Provide revised SAMA analysis results, specifically for the analysis case discussed in response to RAI 4e, dated February 5, 2008.(5) Provide the complete MACCS2 input file used for the reanalysis (in electronic format).Response The following document provides responses to the requests listed above.(1) Section [4] describes and justifies the meteorological data used in the SAMA reanalysis.

(2) Revised estimates of the offsite population dose and offsite economic costs are provided in Tables 1 and 2.(3) Identification of the meteorological tower elevation from which meteorological data were obtained and the rationale for selecting the data from that tower elevation are provided in Section [2].(4) Revised SAMA analysis results, specifically for the analysis case discussed in response to RAI 4e, are provided in Tables 4 and 5.(5) The complete MACCS2 input files used for the SAMA reanalysis listed in Section[10] are provided in electronic format.

NL-09-165 Attachment 1 Docket Nos. 50-247 & 50-286 Page 2 of 33 TABLE OF CONTENTS[1] INTRODUCTION

...................................

..................

3[2] PREPARATION OF ANNUAL METEOROLOGICAL DATA ..................................

4[3] NON-,METEOROLOGICAL LEVEL 3.MODEL INPUTS ..............

...........................

5[41 MACCS2 ANALYSIS AND RESULTS ..............................................................

5[5] COST BENEFIT ANALYSIS RESULTS .....................................

..................

7[6] REVISED COST ESTIMATES

..........................................

I ..... 7[7] MAIN STEAM SAFETY VALVE GAGGING SAMA (UPDATED RESPONSE TO ROUND 2 RAI 6) ...................................................

29[8] TI-SGTR SENSITIVITY ANALYSIS (REVISED RESPONSE TO ROUND 2 RAI 5 ) ...................................

..............................

.....................

2 9[9] CONCLUSION

.....................................................

31[10] MACCS2 INPUT FILES ..............................................

33[11] REFERENCES, .. .. ..............................................

......33 NL-09-165 Attachment I Docket Nos. 50-247 & 50-286 Page 3 of 33 IP2 and IP3 SAMA Reanalysis

[1] Introduction The IP2 and IP3 Severe Accident Mitigation Alternative (SAMA) analyses originally described in the Environmental Report (ER) of the license renewal application, dated April 3, 2007, used site specific meteorological data (wind speed, wind direction, temperature, and accumulated precipitation) obtained from the IPEC onsite meteorological monitoring system (Reference 1).As permitted by NEI 05-01, "Severe Accident Mitigation Alternatives (SAMA) Analysis Guidance Document,"'

1 (Reference

4) five years of meteorological data (2000-2004) were averaged and used in the original SAMA analyses.

Since the SAMA analyses began in the fall of 2005, these five years were the most recent data available at the time of the original analyses.

The.five-year data included 43,848 (two leap years) consecutivehourly values of wind speed, wind direction, precipitation, and temperature recorded at the IPEC meteorological tower from January 2000 through December 2004. The results of the original SAMA analyses were reported in the ER and clarified in response to questions from the Nuclear Regulatory Commission (References 2 and 3).As described above, the original SAMA analyses used five year averages of wind speed, wind direction, precipitation, and temperature.

The averaging method for wind direction, however, was determined to be incorrect and, as a result, the averaged wind direction data was not representative of wind direction conditions in the region for the five year period (Reference 5).Therefore, the SAMAs have been reanalyzed using a single representative year of meteorological data as described below. As described further in Section [4] below, Year 2000 was selected as the representative year because, of the five years of data, it is the year that, resulted in the most conservative

.(i.e. largest) calculated population doses. Using one representative year avoids the need to average multiple years of meteorological data, including wind direction.

In accordance with NEI 05-01 recommendations, the original SAMA analyses described in the ER included multiple cases including.a baseline case with uncertainty and three sensitivity cases (use of a 3 percent discount rate, use of a longer plant life, and consideration of economic losses by tourism and business).

The sensitivity cases in the ER did not identify additional potentially cost beneficial SAMAs beyond those already identified by the baseline with uncertainty case.During their review, the Nuclear Regulatory Commission (NRC) Staff noted that incorporation of tourism and business losses could result in identification of additional cost beneficial SAMAs if it was considered the baseline case and multiplied to account for uncertainties.

Therefore, in " response to request for additional information (RAI) 4e, Entergy provided the results of a revised uncertainty analysis in which the impact of lost tourism and business was analyzed as the baseline analysis and multiplied to account for uncertainties (Reference 2). This uncertainty case resulted in the identification of two additional potentially cost beneficial SAMAs for IP2 and one additional potentially cost beneficial SAMA for IP3. Since it resulted in the largest number of potentially cost beneficial SAMAs, the RAI 4e analysis case is the most conservative case.The SAMA reanalysis described below was performed for the same most conservative case;I NEI 05-01 was endorsed by NRC in Federal Register/

Vol; 72, No. 156 / Tuesday, August 14,2007.

f, NL-09-165 Attachment 1 Docket Nos. 50-247 & 50-286 Page 4 of 33 i.e., the RAI 4e analysis case in which the impact of, lost tourism and business was analyzed as the baseline analysis and multiplied to account for uncertainties.

Following the SAMA reanalysis, an additional sensitivity case was analyzed to provide a revised response to Round 2 RAI 5 (Reference 3). This sensitivity case determined the impact of applying values derived from NUREG-1570, Risk Assessment of Severe Accident-Induced Steam Generator Tube Rupture, although final industry.consensus on the thermally-induced steam generator tube rupture (TI-SGTR) issue has not yet been reached. See.Section

[8] for a description of this sensitivity case-As a result of the SAMA reanalysis and sensitivity case using a conservatively representative, single year of meteorological data (2000), three additional SAMA candidates were found to be potentially cost beneficial for mitigating the consequences of a severe accident for IP2 and three additional SAMA candidates were'found to be potentially cost beneficial for IP3 (in addition to those previously designated as cost beneficial in Section 4.21.6 of the ER and References 2 and 3).[2] Preparation of Annual Meteorological Data The MACCS2 code accepts 8,760 consecutive hourly values (one year) of meteorological data.Each of the five years of meteorological data used in the original analysis was prepared for input into the MACCS2 code by converting values recorded at the primary meteorological tower at the IPEC site to-the units used by MACCS2, assigning an atmospheric stability class based upon..the temperature data, and using data substitution to fill in limited missing data.The primary meteorological tower at IPEC records data on an hourly basis at three elevations, *1Oin, 60m, and 122m. All available data from the 1Om elevation of the primary meteorological tower was used in both the original SAMA analysis and reanalysis because it is closest to the assumed release height of 30m and,, therefore, would be most representative of the conditions at the point of release. Both the .original SAMA analysis and reanalysis assumed a release height of 30in.because it is approximately half the height above grade level of the lP2 and IP3 containment bbiildings, as recommended by NEI 05-01 to provide adequate dispersion of the plume to the surrounding area. Data from this elevation is also currently used in calculations for the effluent release reports submitted to the NRC pursuant to 10 CFR Part 50.36a and the IPEC emergency plan.Data -substitution methods used in the SAMA reanalysis were in accordance with Environmental Protection Agency (EPA) guidance provided in Reference

6. These methods included substitution of limited missing meteorological data with data interpolated, averaged, or curve-fit from previous and subsequent hours and substitution of valid data collected from the 60m elevation.

In the MACCS2 input file for 2000, which was conservatively selected for use in-the SAMA reanalysis, the following data substitutions were made.Seventy-four hours of 10-meter wind direction data was missing for day 316 hour0.00366 days 0.0878 hours 5.224868e-4 weeks 1.20238e-4 months 14 through day 319 hour0.00369 days 0.0886 hours 5.274471e-4 weeks 1.213795e-4 months 15 (in METIOO.inp).

To maintain consistency and wind variability, data from the 60-meter sensor was substituted for the seventy four hours of missing 10-meter wind direction data.

NL-09-165 Attachment 1 Docket Nos. 50-247 & 50-286 Page 5 of 33 Eight hours of temperature data was missing for day 95 hours0.0011 days 0.0264 hours 1.570767e-4 weeks 3.61475e-5 months 2 through 9 (in METIOO.inp)..

Values for these eight hours were obtained by linear interpolation of the preceding and subsequent valid temperature values.Data for all meteorological parameters was missing on day 104 hours0.0012 days 0.0289 hours 1.719577e-4 weeks 3.9572e-5 months 10 and 11, day 104 hour0.0012 days 0.0289 hours 1.719577e-4 weeks 3.9572e-5 months 19, day 255 hours0.00295 days 0.0708 hours 4.21627e-4 weeks 9.70275e-5 months 18 and 19, and day 294 hours0.0034 days 0.0817 hours 4.861111e-4 weeks 1.11867e-4 months 10 and 11 (in METIOO.inp).

Substitute values were obtained by interpolation, curve-fitting, or averaging the preceding and subsequent valid data values as appropriate.

[3] Non-Meteorological Level 3 Model Inputs In addition to meteorological data, MACCS2 also uses input data for population, land fraction, watershed class, regional economic data, agriculture data, emergency response assumptions, and source terms. These inputs are described in Sections E. 1.5 and E.3.5 of the ER.For the regional average value of non-farm wealth (VALWNF), a value of $208,838.49/person was used in the SAMA reanalysis consistent with sensitivity case 3 in the ER. As mentioned in Section [1], the RAI 4e analysis case (which is ER sensitivity case 3 multiplied to account for uncertainty) is;the most conservative case, resulting in the largest number of potentially cost beneficial SAMAs. The reanalysis was performed for the RAI 4e analysis case in which the impact of lost tourism and business was analyzed as the baseline analysis and multiplied to account for uncertainties.

Consequently, the revised benefit results for all SAMAs include the impact of lost tourism and business, as described in the response to request for additional information (RAI) 4e (Reference 2).The other, non-meteorological data were the same as those described for the baseline case in the ER (described in Sections E.1.5 and E.3.5 of the ER). 'Since the reanalysis uses the same non-meteorological input data as the original RAI 4e analysis case, the only difference between the original RAI 4e analysis and the reanalysis is the meteorological data.[4] MACCS2 Analysis and Results As-with the original SAMA analysis, the SAMA reanalysis also used MACCS2 to estimate the mean population dose risk (PDR) and offsite economic cost risk (OECR). Preliminary results from MACCS2 using each of the five years of meteorological data (2000-2004) were compared.Since the dose and economic cost results for all of the individual years were similar, the year that resulted in the most conservative (i.e. largest) doses (year 2000) was selected as the representative year for use in the SAMA reanalysis.

This method of choosing a representative.year agrees with. the example provided in NEI 05-01. The revised estimated mean values of PDR and OECR for IP2 and IP3 using year 2000 meteorological data are presented in Table-1 for IP2 and Table 2 for IP3. Comparison of the values in Tables 1 and 2 with those in ER Tables E.1-14 and E.3-14 shows that the individual year PDR and OECR values are larger than the original ER values due to removal of wind direction biases introduced by the faulty wind direction averaging method.Table 3 provides a breakdown of the total population dose by containment failure mode, similar to information provided in -response to RAI 2a (Reference 2).

I-NL-09-165 Attachment 1 Docket Nos. 50-247 & 50-286 Page 6 of 33 Table 1 IP2 Mean PDR and OECR Using Year 2000 Meteorological Data Offsite Population Offsite Release Frequency

-aDose Economic.

Dose Risk Economic Cost (PDR) Cost Mode) (person-sv)*

.",(person-Risk (OECR)MW n ($) rem/yr) ($/yr)NCF 1.19E-05 4.75E+01 9.98E+04 5.64E-02**

1.18E+00 EARLY HIGH 6.50E-07 6.51 E+05 2.05E+11 4.23E+01 1.33E+05, EARLY MEDIUM 4.23E-07 1.94E+05 5.87E+10 8.21E+00 2.48E+04 EARLY LOW 1.11E-07 7.93E+04 6.39E+09 8.81 E-01 7.1OE+02 LATE HIGH 6.88E-07 1.63E+05 4.64E+10 1.12E+01 3.19E+04 LATE MEDIUM 3.43E-06 6.87E+04 6.06E+09 2.36E+01 2.08E+04 LATE LOW 6.43E-07 1.61 E+04 6.59E+08 1.04E+00 4.24E+02 LATE LOWLOW 5.82E-08 1.38E+04 5.62E+08 8.04E-02 3.27E+01 Total 8.74E+01 2.12E+05* lsv=l)00rem

- ý 5.64E-02 (person-rem/yr)

= 1.19E-05 (/yr) x 4.75E+01 -(person-sv) x 100 (rem/sv)Table 2 IP3 Mean PDR and OECR Using Year 2000 Meteorological Data Offsite .Population Offsite Release Frequency Population Economic Dose Risk Economic Dose ,(PDR) Cost Mode (Iyr) Cost (D)Cs (person-sv)* (person- Risk (OECR)($) rem/yr) ($/yr)NCF 6.30E-06 8.04E+01 2.95E+05 5.06E-02**

1.86E+00 EARLY HIGH 9.43E-07 5.08E+05 1.70E+11 4.79E+01 1.60E+05 EARLY MEDIUM 1.24E-06 2.OOE+05 5.55E+10 2.47E+01 6.87E+04 EARLY LOW 1.46E-07 5.21E+04 3.58E+09 -.7.59E-01 5.21E+02 LATE HIGH 4.23E-07 1.63E+05 4.61E+10 6.89E+00 1.95E+04 LATE MEDIUM 2:01E-06 6.85E+04 6.06E+09 1.37E+01 1.22E+04 LATE LOW .3.75E-07 1.61 E+04 6.58E+08 6.03E-01 2.47E+02 LATE LOWLOW 5.66E-08 1.38E+04 .5.62E+08 7.81E-02 3.18E+01 Total 9.48E+01 2.61 E+05* lsv=100rem

- 5.06E-02 (person-remfyr)

= .6.30E-06

(/yr) x 8.04E+01 (person-sv) x 100.(rem/sv)

NL-09-165 Attachment 1 Docket Nos. 50-247 & 50-286 Page 7 of 33 Table 3 Breakdown of PopulationDose by Containment Failure Mode IP2 IP3 Containment Failure Population Percent Population Percent Mode Dose Dose (person-rem/yr)

Contribution (person-rem/yr)

Contribution Intact containment 0.06 0.06% '0.05 0.05%Basemat melt-through 4.08 4.67% 2.42 2.56%Gradual overpressure 28.27 32.35% 16.78 17.70%Late hydrogen burns 3.55 4.07% 2.11 2.23%Early hydrogen burns 8.64 9.89% 3.16 3.33%In-vessel steam explosion 0.57 0.65% 0.21 0.22%Ex-vessel steam explosion 0.0027 0.00% 0.0010 .0.00%Vessel overpressure 4.10 4.69% 1.50 1.58%Containment isolation 0.0375 0.04% 0.0137 0.01%ISLOCA 6.61 7.57% 4.18 4;41%SGTR 31.46 36.00% 64.35 67.89%Total 87.4 100 94.8 100[5] Updated Cost Benefit Analysis Results The cost benefit reanalysis was performed using the MACCS2 results for year 2000. The results are reported in Table 4 for IP2 and in Table 5 for IP3. The assumptions used to determine the change in plant risk that could be realized by implementation of each of the SAMAs originally described ih Sections E.2.3 and E.4.3 of the ER and subsequent RAI responses were not altered in this reanalysis.

Therefore, the CDF reduction for each SAMA has not been repeated in the tables. The benefit values account for risk reduction in both internal and external.events (using multipliers described in Section 4.21.5.4 of the ER) and include the economic impact of lost tourism and business following a severe accident (as discussed in Section [3]). The benefit with uncertainty values account for analysis uncertainties (using multipliers described in Section 4.21.5.4 of the ER). Except as noted with "t" and as described in Section [6], the estimated cost values for SAMA candidates are the same as those reported previously (in Tables E.2-2 and E.4-2 in the ER and References 2 and 3).[6] Revised Cost Estimates As described in Sections E.2.3 and E.4.3 Qf the ER, the original SAMA implementation costs were conceptually estimated to the point where conclusions regarding the economic viability of the proposed modification could be adequately gauged. Specifically, in the original analysis, the initial cost estimate for each SAMA was obtained by applying engineering judgment to determine if the implementation cost was clearly in excess of the estimated attainable benefit or by applying an existing estimate from a previous SAMA analysis.

The engineering judgment cost estimates were conservative i.e. minimum or low cost estimates to implement the SAMA.

NL-09-165 Attachment 1 Docket Nos. 50-247 & 50-286 Page 8 of 33 A SAMA that appeared to be cost beneficial with the initial implementation cost estimate was subjected to successively more comprehensive and more precise cost estimating to determine if the SAMA was indeed potentially cost beneficial.

Only the final cost estimate for each SAMA was reported in the ER.This method of cost estimating is consistent with NEI 05-01 guidance which states the following."As SAMA analysis focuses on-establishing the economic viability of potential plant enhancement when compared to attainable benefit, often detailed cost estimates are not required to make informed decisions regarding the economic viability of a particular modification.

SAMA implementation costs may be clearly in excess of the attainable benefit estimated from a particular analysis case. For less clear cases, engineering judgment may be applied to determine if a more detailed cost estimate is necessary to formulate a conclusion regarding the economic viability of a particular SAMA.Nonetheless, the cost of each SAMA candidate should be conceptually estimated to the point where economic viability of the proposed modification can be adequately gauged.For hardware modifications, the cost of implementation may be established from existing estimates of similar modifications from previously performed SAMA and SAMDA analyses." Comparison of Tables 4 and 5 with'the tables provided in response to RAI 4e (Reference 2)shows that the benefit obtained from each of .the SAMAs (except those with' no benefit) has increased in the reanalysis.

Consistent with the approach described in NEI 05-01 and used in the original analysis, SAMAs in the reanalysis that appeared to be cost beneficial with the new benefit estimate and the old implementation cost estimate were subjected to more comprehensive and precise cost estimating techniques to determine if they are indeed potentially cost beneficial.

The cost estimates for SAMAs noted with "t" in Table 4 and Table 5 are those that were developed in more detail.For example, in the reanalysis, IP3 SAMA 040,,"Provide automatic nitrogen backup to steam generator atmospheric dump valves," was estimated to have a benefit with uncertainty of$344,225 (see Table 5). If this benefit is compared to the original cost estimate of $214,000, IP3 SAMA 040 appears cost beneficial.

A more comprehensive plant-specific cost estimate was performed to determine if IP3 SAMA 040 is indeed potentially cost beneficial.

This more comprehensive estimate concluded that a modification to provide automatic nitrogen backup to the steam generator atmospheric dump valves at IP3 would actually cost approximately

$950,000 (see Table 5). Since this value is greater than the revised benefit with uncertainty, IP3 SAMA 040 is not potentially cost beneficial.

Also, in the reanalysis 1P2 SAMA 062, "Provide a hard-wired connection to an SI pump from ASSS power supply," was estimated to have a benefit with uncertainty of $1,789,822 (See Table 4). If this benefit is compared to the original cost estimate of $722,000, IP2 SAMA 062 appears cost beneficial.

The original cost estimate was reviewed and found to have conservatively not included some of the expenses necessary to implement the modification.

Therefore, a more comprehensive cost estimate was performed to determine if this SAMA is indeed potentially cost beneficial.

This estimate concluded that a modification to provide a hard-wired connection to a safety injection pump from an alternate.safe shutdown system power NL-09-165 Attachment 1 Docket Nos. 50-247 & 50-286 Page 9 of 33 supply would actually cost approximately

$1,500,000 (See Table 4). Since the more-comprehensive cost estimate is still smaller than the revised benefit, IP2 SAMA 062 is potentially cost beneficial following the reanalysis.

Entergy's standard process for development of conceptual level project estimates was followed for the new, more comprehensive SAMA. implementation cost estimates.

The estimates capture anticipated expenses by identifying all parts of the organization that must support the proposed SAMA modification from the conceptual perspective.

Typical expenses associated with project cost estimating include calculations, drawing updates, specification updates, bid evaluations, contract issuance, design package preparation, walkdowns, planning and scheduling, estimating, procurement, configuration management, as-low-as-reasonably-achievable (ALARA), quality control and quality assurance, training, simulator changes, information technology, design basis- update, construction, multi-discipline and independent review of design concepts and calculations, 50.59 review, final safety analysis report (FSAR) update, cost control, contingency, security, procedures, post.work testing, and project management and close-out.

In addition, the project cost estimates include corporate indirect charges.

-m ----------- -m ----NL-09-165 Attachment 1 Docket Nos. 50-247 & 50-286.Page 10 of 33 Table 4 Results of Cost Benefit Analysis of IP2 SAMA Candidates Risk Reduction Benefit with 1P2 Phase II SAMA (%0) Benefit Uncertainty Estimated Cost Conclusion PDR OECR Uncertainty 00i -Create an ..independent RCP seal 1.60% 1.42% $374,757 $788,963 $1137,000 Not cost injection system with a $ ,effective dedicated diesel.002 -Create an independent RCP seal 1.49% 1.42% $350,396 $737,676 $1,000,000 Not cost injection system without a effective dedicated diesel.003 -Install an additional Not cost CCW pump. p 0.00% 0.00% $0 $0 $1,500,000 effective 004 -Enhance procedural Not cost guidance for use of 0.23% 0.00% $48,723 $102,574 $1,750,000 effective service water pumps.005 -Improve ability to cool the RHR heat~Not cost exchangers by allowing 0.34% 0.47% $105,892 $222,931 $565,000 effective manual alignment of the fire protection system...._._

_006 -Add a diesel building 0.11% 0.07% $30,496 $64202 $274,000 Not cost high temperature alarm. .11%.. effective 00.7 -Install a filtered containment vent to 16.70% 6 Not cost provide fission product 6.13% $1,725,939

$3,633,555

$5,700,000 effective scrubbing.

M-, Mm 'we o A" MY n -41m 4MI M -M NL-09-1 65 Attachment 1 Docket Nos. 50-247 & 50-286 Page 11 of 33 Table 4 Results of Cost Benefit Analysis of IP2 SAMA Candidates Risk Reduction Benefit with IP2 Phase II SAMA (%) Benefit Uncefiatityh Estimated Cost Conclusion PDR OECR 008 -Create a large concrete crucible with heat removal potential 47.03% 34.43% $6,347,528

$13,363,217

$108,000,000.

Not cost under the base mat to effective contain molten core debris.009 -Create a reactor 47.03% 34.43% $6,347,528

$13,363,217

$4,1001000t Retain cavity flooding system.010 -Create a core melt Not cost 47.03% 34.43% $6,347,528

$13,363,217

$90,000,000 source reduction system. _ _"_effective 011 $10,900,000ens t Not cost 01!- Provide a means to 17.51% 21.23% $3,091,966

$6,509,402

$10,900,000 effective inert Containment..

effective____________

012 -Use the fire protection system as a backup source for the 0.00% 0.00% $0 $0 .$565,000 effect* effective containment spray system. "_ _013 -Install a passive Not cost containment spray 0.00% 0.00% $0 $0 $2,000,000 effective system.014 -Increase the depth of the concrete base mat or use an alternative Not cost C e alternati 11.56% 4.25% $1,194,251

$2,514,214

>$5,000000 effective concrete material toefctv ensure melt-through does not occur.

M" -am 1001 low ----NL-09-165 Attachment 1 Docket Nos. 50-247 & 50-286.Page 12 of 33 Table 4 Results of Cost Benefit Analysis of IP2 SAMA Candidates Risk Reduction IP2 Phase u SAMA ( o) Benefit Benefit with Estimated Cost Conclusion PDheSOER BUncertainty PDR I OECRCnrtbu 015 -Construct a building connected to primary Not cost containment that is .effective maintained at a vacuum. _016 -Install a redundant

$58000 Not cost containment spray 0.00% 0.00% $0 $0 $5,800,000 effect sytm 1.Ieffective system.__________

017 -Erect a barrier that provides containment liner Not cost protection from ejected 10.07% 11.79% $1,742,298

$3,667,996

$5,500,000f effective core debris at high pressure.018 -Install a highly reliable steam generator shell-side heat removal Not cost system that relies on 0.46% 0.47% $73,618 $154,986 $7,400,000 effective natural circulation and stored water sources.019 -Increase secondary side pressure capacity Not cost such that a SGTR would 30.21% 39.15% $5,594,541

$11,777,981

>$100,000,0100t effective not cause the relief valves to lift.

m ." -0 M 1 = ---m -NL-09-165 Attachment 1 Docket Nos. 50-247 & 50-286 Page 13 of 33 Table 4 Results of Cost Benefit Analysis of IP2 SAMA Candidates Risk Reduction B .enefit with *IP2 Phase II SAMA R e) Benefit wentinth Estimated Cost Conclusion Benefit. Uncertainty PDR OECR 020 -Route the discharge from the main steam safety Valves through a Not cost structure where a water 2.97% 4.25% $580,766 $1,222,665

$9,700,000 Notect spray would condense the steam and remove most of the fission products., 021 -Install additional pressure or. leak Retain monitoring 11.33% 14.62% $2,093,852

$4,408,109

$3,200,000t (New)instrumentation for ISLOCAs. -022 -Add redundant and diverse limit switches to Retain e5.72% 7.55% $1,071,465

$2,255,716

$2,200,000t (New)each containment(Nw isolation valve.023 -Increase leak testing of valves in ISLOCA 5.72% 7.55% $1,071,465

$2,255,716

$7,964,000 Not cost aeffective paths.024 -Ensure all ISLOCA 11.33% 14.62% $2,093,852

$4,408,109

$9,700,000 Not cost releases are scrubbed.

1133 _4.2 effective 025 -Improve MSIV $258,310 $476,000 Not cost design. 0.57% 0.94% $122,697 $258310$47,00 effective 026 -Provide additional Not cost DC battery capacity.

0.23% 0.00% $48,723 $102,574 >$1,875,000 effective 027 -Use fuel cells Not cost, instead of lead-acid 0.23% 0.00% $48,723 $102,574 $2,000,000 effective batteries.

-m) -i -n -: --, ,-, --' -M i¸m -NL-09-165 Attachment 1 Docket Nos. 50-247 & 50-286 Page 14 of 33 Table 4 Results of Cost Benefit Analysis ofl1P2 SAMA Candidates Risk Reduction IP2 Phase II SAMA (%). enefit Benefit with Estimated Cost Conclusion

_ PDR OECR Uncertainty 028 -Provide a portable diesel-driven battery 9.3.8% 7.08% $1,357,046

$2,856,939

$938,000t Retain charger.029 -Increase/improve Not cost DC bus load shedding.

0.23% 0.00% $48,723 $102,574 $460,000t effectiVe 030 -Create AC power cross-tie capability with 0.23% 0.00% $56,813 $119,607 $1,156,000 Not cost other unit. effective 031 -Create a backup source for diesel cooling 0.23% 0.00% $40,632 $85,541 $1,700,000 Not cost (not from existing system). effective 032 -Use fire protection Ntcs system as a backup 0.23% 0.00% $40,632 $85,541 -$497,000 effect source for diesel cooling. effective 033 -Convert under-voltage AFW and reactor protective system Not cost actuation signals from 2- 0.00% 0.00% $0 $0 $1,254,000 effective out-of-4 to 3-out-of-4 logic. -___034 -Provide capability Not cost for diesel-driven, low 0.06% 0.05% $8,180 $17,221 >$632,000 effective pressure vessel makeup.035 -Provide an additional high pressure 0.34% 0.47% $73,529 $154,798 $5,000,000 Not cost injection pump with effective independent diesel.

-w-0 S-, JEW -'"- , EM &w ý m a on-, M"-NL-09-165 Attachment 1 Docket Nos. 50-247 & 50-286 Page 15 of 33 Table 4 Results of Cost Benefit Analysis of IP2 SAMA Candidates Risk Reduction Benefit with IP2 Phase II SAMA _____(%) Benefit B Estimated Cost Conclusion PDR OECR Uncertainty 036 -Create automatic swap-over to recirculation 0.46% Not cost cooling upon RWST 0.47% $138344 $291,251 >$1,000,000 effective depletion.

.037 -Provide capabilityNot cost for alternate injection via 0.06% 0.05% $8,180 $17,221 $750,000 effective diesel-driven fire pump.038 -Throttle low pres .sure injection pumps earlier in medium or large- 0.11% 0.07% $22,405 $47,169 $82,000 Not cost break LOCAs to maintain effective reactor water storage tank inventory.

039 -Replace two of three motor-driven SI pumps Not cost with diesel-powered 0.34% 0.47% $73,529 $154,798 $2,000,000 effective pumps: -040 -Create/enhance a " reactor coolant .3.20% 3.77% $572,408 $1,205,070

$2,000,000t Not cost depressurization system:. effective 041 -Install a digital feed Not cost water ugrade. 0.92% 0.47% $179,154-

$377,167 $900,000 wae pgae effective 042 -Provide automatic nitrogen backup to steam Not cost genraoratosheic 0.23% '0.00% $16,360 $34,441 $214,000 efctv generator atmospheric

'.effective dump valves.043 -Add a motor-driven 0.92%

0 79,154 $377,167 $2,000,000 Not cost feed water pump. .9 0 effective NL-09-165 Attachment 1 Docket Nos. 50-247 & 50-286 Page .16 of 33 Table 4 Results of Cost Benefit Analysis of IP2 SAMA Candidates Risk Reduction Benefit with IP2 Phase II SAMA (%) Benefit Uncertainty Estimated Cost Conclusion PDR OECR 044 -Use fire water system as backup for 14.19% 9.91% $2,350,530

$4,948,485

$1,656,000 Retain steam generator inventory..

045 -Replace current pilot operated relief valves with larger ones such that only 3.32% 1.89% $667,806 $1,405,907

$2,700,000 Not cost one is required for effective successful feed and bleed.046 -Modify emergency operating procedures for Not cost ability to align diesel 0.00% 0.00% $0 $0 $82,000 effective power to more air compressors.

047 -Add an independent 0.00% 0.00% $0 $0. $300,000 Not cost boron injection system. _ _effective 048 -Add a system of relief valves that prevent equipment damage from a 0.46% 0.47% $105,981 $223,119 $615,000 eNot cost* " effective pressure spike during an ATWS.049 -Install motor generator set trip breakers 0.23% 0.00% $32,541 $68,508 $716,000 Not cost in control room. effective

---- ----- ---NL-09-165 Attachment 1 Docket Nos. 50-247 & 50-286 Page 17 of 33 Table 4 Results of Cost Benefit Analysis of IP2 SAMA Candidates Risk Reduction Benefit with IP2 Phase II SAMA .Benefit Uncertainty Estimated Cost Conclusion PDR OECR 050 -Provide capability to remove power from the 0.23% 0.00% $32,541 $68,508 $90,000 Not cost bus powering the control effective rods. "__051- Provide digital large Not cost break LOCA protection.

effective 052 -Install secondary Not cost side guard pipes up to the 1.72% 1.89% $294,384 $619,756 $1,100,000 effective MSIVs.053 -Keep both pressurizer PORV block 3.32% 1.89% $659,715 $1,388,873

$800,000 Retain valves open.054 -Install flood alarm in the 480V switchgear 39.24% 28.77% $5,591,781

$11,772,170

$200,000 Retain room. _055 -Perform a hardware modification to allow high- Not cost head recirculation from 0.00% 0.00% $0 $0 $1,330,000 effective either RHR heat exchanger.

056 -Keep RHR heat exchanger discharge 0.23% 0.00% $48,723 $102,574 $82,000 Retain motor operated valves (MOVs) normally open.057 -Provide DC power 0.46% .0.47% $89,800 $189,052 $376,000 Not cost backup for the PORVs. effective m~mm~ ~ --- --NL-09-165 Attachment 1 Docket Nos. 50-247 & 50-286 Page 18 of 33 Table 4 Results of Cost Benefit Analysis of IP2 SAMA Candidates Risk Reduction Benefit with IP2 Phase II SAMA Benefit Uncertainty Estimated Cost Conclusion PDR OECR 058 -Provide procedural guidance to allow high- Not cost head recirculation from 0.00% 0.00% $0 $0 $82,000 either RHR heat effective exchanger.

059 -Re-install the low pressure suction trip on the AFW pumps and 0.23% 0.00% $24,450 $51,474 $318,000 Not cost enhance procedures to effective respond to loss of the normal suction path.060 -Provide added -, protection against flood propagation from 8.92% 6.60% $1,275,337

$2,684,920

$216,000 Retain stairwell 4 into the 480V switchgear room.061 -Provide added protection against flood propagation from the 19.34% 14.15% $2,754,991

$5,799,982

$192,000 Retain deluge room into the 480V switchgear room.062 -Provide a hard-wired connection to an 6.06% 4.25% $850,165 $1,789,822

$1,500,000t Retain Sl pump from ASSS (New)power supply.(

m M m M m -m m m M m m-M m m m m m m NL-09-165 Attachment 1 Docket Nos. 50-247 & 50-286 Page 19 of 33 Table 4 Results of Cost Benefit Analysis of IP2 SAMA Candidates Risk Reduction P2 Phase SAMA Benefit ett Estimated-Cost Conclusion IP2 haseII SMA (%)Beneit Uncertainty PDR OECR 063 -Provide a water-tight door for additional 0.11% 0.00% $32,452 $68,320 $324,000 Not cost protection of the RHR 0 effective.

pumps against flooding.064 -Provide backup cooling water source for Not cost theCC hat0.23% 0.00% $40,632 $85,541 $710,000 efctv the CCW heat -effective exchangers.

065 -Upgrade the ASSS to allow timely 39.24% 28.77% $5,591,781

$11,772,170

$560,000 Retain restoration of seal injection and cooling.066 -Harden the EDG building and fuel oil Not cost transfer pumps against 8.96% 6.19% $2,505,846

$5,275,465

$33,500,000t effective tornados and high winds. "_" 067 -Provide hardware connections to allowthe Not cost r wa w s them t 0.02% 0.00% $9,727 $20,477 $576,000 effective primary water system toefctv cool the charging pumps. ._....068 -Provide independent source of cooling for the 0.06% 0.01% $13,408 $28,227 $710,000 Not cost recirculation pump effective motors.I Cost estimate revised from what was previously reported.

See Section [6] for more information.

m n --m -----n- -m ----m NL-09-165 Attachment 1 Docket Nos. 50-247 & 50-286 Page 20 of 33 Table 5 Results of Cost Benefit Analysis of IP3 SAMA Candidates Risk Reduction IP3 Phase II SAMA M Benefit Uncertainty Estimated Cost Conclusion PDR OECR 001 -Create an independent RCP seal 0.74% .0.38% $236,610 *$342,913

$1,137,000 Not cost injection system with a .effective dedicated diesel.002 -Create an independent RCP seal Not cost injection system without a 0.63% 0.38% $201,222 $291626 $1,000,000 effective dedicated diesel.003 -Install an additional Not cost CWum.0.00%

0.00% $0 $0 $1,500,000 ccw pump. effective 004 -Improved ability to cool the RHR heat Not cost exchangers by allowing 0.53% 0.38% $130,575 $189,240 $565,000 effect manual alignment of the fire protection system.005- Install a filtered containment vent to Not cost 9.60% 2.68% $1,497,163

$2,169,801

$5,700,000 provide fission product *effective scrubbing.

006 -Create a large concrete crucible with heat removal potential Not cost under the base mat to 24.16% 14.94% $5,038,071

$7,301,552.

$108,000,000 effective contain molten core debris.

m ---m ----m -----* ----l NL-09-165 Attachment 1 Docket Nos. 50-247 & 50-286 Page 21 of 33 Table 5 Results of Cost Benefit Analysis of IP3 SAMA Candidates Risk Reduction Benefit with 1P3 Phase II SAMA (%) Benefit netit Estimated Cost Conclusion PDR OECRUncertainty 007 -Create a reactor' Retain 007-Create ad reactor. 24.16% 14.94% $5,038,071

$7,301,552

$4,100,000"t Rean cavity flooding system. ________ _____ ______ (New)008 -Create a core melt 24.16% 14.94% $5,038,071

$7,301,552

$90,000,000 Not cost source reduction system. 2,0effective 009 -Provide means to Not cost inert containment.

8.76% 9.20% $2,412,095

$3,495,790

$10,900,000 effective 010 -Use the fire protection system as a Not cost backup source for the 0.00% 0.00% $0 $0 $565,000 effective containment spray system.011 -Install a passive Not cost containment spray 0.00% 0.00% $0 $0 $2,000,000 effective system.012 -Increase the depth of the concrete base mat or use an alternative Not cost oncrete manaterialto 5.59% 1:53% $867,404 $1,257,107

>$5,000,000 eftive concrete material to effective ensure melt-through does not occur..013 -Construct a building connected to primary 21.73% 15.71% $4,883,602

$7,077,683

$61,000,000 Not cost containment that is 2 effective maintained at a vacuum.014 -Install a redundant Not cost containment spray 0.00% 0.00% $0 $0 $5,800,000 effective system. I I _ I _I

---m -----~ ---- ---NL-09-165 Attachment 1 Docket Nos. 50-247 & 50-286 Page 22 of 33 Table 5 Results of C6st Benefit Analysis of IP3 SAMA Candidates Risk Reduction IP3 Phase II SAMA (%) Benefit Uncerity wit__PDR OECR Uncertainty 015 -Erect a barrier that provides containment liner protection from ejected 4.32%. 4.21% $1,140,695

$1,653,182

$5,500,000t effct core debris at high .pressure.016 -Install a highly reliable steam generator shell-side heat removal Not cost system that relies on 6.27% 4.98% $1,401,717

$2,031,473

$7,400,000 effective natural circulation and stored water sources.017 -Increase secondary side pressure capacity Not cost such that an SGTR would 45.15% 53.64% $13,520,698

$19,595,215

>$100,000,000t effective not cause the relief valves to lift.018 -Route the discharge from the main steam safety valves through a Not cost structure where a water 11.08% 13.41% $3,327,028

$4,821,779

$12,000,00t effective spray would condense the steam and remove most of the fission products.019 -Install additional pressure or leak Retain monitoring 7.07% 8.43% $2,126,663

$3,082,120

$2,800,000t (New)instrumentation for ISLOCAs.

--m -m m -m -m -NL-09-165 Attachment 1 Docket Nos. 50-247 & 50-286 Page 23 of 33'Table 5 Results of Cost Benefit Analysis of IP3 SAMA Candidates Risk Reduction Benefit with 1P3 Phase II SAMA (%) Benefit Estimated Cost Conclusion PDR OECRUncertainty PDR OECR ______020 -Add redundant and diverse limit switches to 3.59% 421% $1069272 $1549670 $4000ooot Not cost each containment effective isolation valve.021 -Increase leak testing Not cost of valves in ISLOCA 3.59% 4.21% $1,069,272

$1,549,670

$10,604,000 effective paths.022 -Ensure all ISLOCA 7.07% 8.43% $2,126,663

$3,082,120

$9,700,000 Not cost releases are scrubbed.

effective 023 -Improve MSIV 0.00% 0.00% $0 $0 $476,000 Not cost design. 0.00_ effective 024 -Provide additional 0 Not cost DC battery capacity.

0.11% 0.00% $47,141 $68,320 >$1,875,000 effective 025 -Use fuel cells Not cost.instead of lead-acid 0.11% 0.00% $47,141 $68,320 $2,000,000 effective batteries.

026 -Increase/improve 0.11% 0.00% $47,141 $68,320 $460,000t Not cost DC bus load shedding.

effective 027 -Create AC power Not cost cross-tie capability with 0.11% 0.00% $70,647- $102,387 $1,156,000.

effective other unit.028 -Create a backup Not cost source for diesel cooling 0.03% 0.00% $15,318 $22,199 $1,700,000 effective (not from existing system), effective 029 -Use fire protection Not cost system as a backup 0.03% 0.00% $15,318 $22,199 $497,000 effective source for diesel cooling.

m -m m ---- -m -m -m -mI NL-09-165 Attachment 1 Docket Nos. 50-247 & 50-286 Page 24 of 33 Table 5 Results of Cost Benefit Analysis of IP3 SAMA Candidates Risk Reduction RP3 Phase ReAu Benefit with Estimated Cost Conclusion IP3_PhaseI

_ _AMA PDR BeneCitUncertainty PDR OECR 030 Provide a portable Not cost diesel-driven battery 0.95% 0.38% $213,363 $309,222 $938,000t effective charger. effective 031 -Convert under-voltage, AFW and reactor Not cost protective system 0.53% 0.38% $118,822 $172,206 $1,254,000 effective actuation signals from 2-out-of-4 to 3-out-of-4 logic.032 -Provide capability Not cost for diesel-driven, low 0.21% 0.00% $23,764 $34,441 >$632,000 effective pressure vessel makeup.033 -Provide an additional high pressure 0.42% 0.38% $118,693 $172,019 $5,000,000 -Not cost injection pump with effective independent diesel.034 -Create automatic Not cost swap-over to recirculation 1.27% 0.77% $530,551 $768,914 >$1,000,000 effective upon RWST'depletion.

035 -Provide capability Not cost for alternate injectiori via 0.21% 0.00% $23,764 $34,441 $750,000 effective diesel-driven fire pump.036 -Throttle low pressure injection pumps earlier in medium or large- 0 Not cost 1--* 0.00% 9.00% $11,753 $17,033 $82,000 break LOCAs to maintain effective reactor water storage tank inventory.

m ml ---m -NL-09-165 Attachment 1 Docket Nos. 50-247 & 50-286 Page 25 of 33 Table 5 Results of Cost Benefit Analysis of IP3 SAMA Candidates Risk Reduction.

IP3 Phase II SAMA (%) Benefit Benefit with Estimated Cost Conclusion PDR OECR Uncertainty 037 -Replace two of three motor-driven SI pumps 042% 0380 $118,693 $172,019 Not cost with diesel-powered

.0 .. $2,000,000 effective pumps.038 -Create/enhance a Ntcs reactor coolant 0.95% 0.77% $237,516 $344,225 $4,600,000 effect depressurization system. effective 039 -Install a digital feed Not cost w0.95% 0.00% $271,481 $393,450 $900,000 water upgrade. effective.

040 -Provide automatic nitrogen backup to steam Not cost generator atmospheric 0.95% 0.77% $237,516 $344,225 .$950,000t effective dump valves.041 -Add a motor-driven 0.95% 0.00% $271,481 $393,450 $2,000,000 Not cost feedwater pump. 05effective 042 -Provide hookup for portable generators to Not cost power the turbine-driven 0.11% 0.00% $47,141 $68,320 $1,072,000 effective AFW pump after station batteries are depleted.043 -Use fire water system as backup for Not cost steam generator 1.58% 1.15% $450,490 $652,885 $1,656,000 Notect enr reffective inventory.

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

m --m -m m --1l --- -NL-09-165 Attachment 1 Docket. Nos. 50-247 & 50-286 Page 26 of 33 Table 5 Results of Cost Benefit Analysis of 1P3 SAMA Candidates Risk Reduction Benefit with IP3 Phase II SAMA (0) Benefit .Uncertainty Estimated Cost Conclusion PDR OECR-044 -Replace current pilot operated relief valves with larger ones such that only Not cost larroneis rquid f t o 4.75% 4.21% $1,246,989

$1,807,230

$2,700,000 effect one is required for effective successful feed and bleed.045 -Add an independent Not cost boron injection system. 0.0% _000 $0_$0_$300,000_

effective 046 -Add a system of relief valves that prevent Not cost equipment damage from a 0.74% 0.00% $224,210 $324,943 $615,000 effective pressure spike during an ATWS.047 -Install motor Not cost generator set trip breakers 0.11% 0.00% $35,388 $51,287 $716,000 effective in control room.048 -Provide capability to remove power from the Not cost bu owrngte otrl0.11%

0.00% $35,388 $511287. $90,000*bus powering the control effective rods.049 -Provide digital large 0.00% 0.00% $0 $0 $2,036,000 Not cost break LOCA protection.

effective 050 -Install secondary Not cost side guard pipes up to the 9.07% 8.81% $2,447,095.

$3,546,515

$9,671,000t effective MSIVs.051 -Operator action: Not cost Align main feedwater for 0.11% 0.00% $23,635 $34,254 $55,000 effective secondary heat removal.

M " M-- -M M M m -M--1 M M m M M NL-09-165 Attachment 1 Docket Nos. 50-247 & 50-286 Page 27 of 33 Table 5 Results of Cost Benefit Analysis of 1P3 SAMA Candidates Risk Reduction IP3 Phase II SAMA Benefit Benefit with Estimated Cost Conclusion PDR OECR Uncertainty 052 -Open city water supply valve for 1.05% 0.77% $249,398 $361,446 $50,000 Retain alternative AFW pump suction.053 -Install an excess flow valve to reduce the .2.07% 1.51% $498,795 $722,892 $228,000 Retain risk associated with hydrogen explosions.

054 -Provide DC power 0.00% 0.00% $0 $0 $376,000 Not cost backup for the PORVs. effective 055 -Provide hard-wired connection to a Sl or RHR pump from the 18.35% 11.49% $4,073,152

$5,903,118

$1,288,000 Retain Appendix R bus (MCC 312A).056 -Install pneumatic controls and indication for Not cost the turbine-driven AFW 0.11% 0.00% $47,141 $68,320 $982,000 effective pump.057 -Provide backup cooling water source for 0.21% 0.00% $59,023 $85,541 $109,000 Not cost the CCW heat effective exchangers.

058 -Provide automatic 0.21% 0.00% $94,282 $136,640 $1,868,000 Not cost DC.power backup._ I I effective m m m IM -M-m- m m m-m m M -M M NL-09-165 Attachment 1 Docket Nos. 50-247 & 50-286 Page 28 of 33 Table 5 Results of Cost Benefit Analysis of IP3 SAMA Candidates Risk Reduction Benefit with 1P3 Phase II SAMA (%) Benefit Uncertainty Estimated Cost Conclusion PDR OECR 059 -Provide hardware connections to allow the Not cost primary water system to 0.00% 0.00% $0 $0 $576,000 effective cool the charging pumps.060 -Provide independent Not cost source of cooling for the 0.00% 0.00% $0 $0 $710,000 effective recirculation pump motors.061 -Upgrade t h-e ASSS -to allow timely 19.73% 12.26% $4,359,371

$6,317,929

$560,000 Retain restoration of seal injection and cooling.062 -Install flood alarm in the 480 VAC 19.73% 12.26% $4,359,371

$6,317,929

$196,800 Retain switchgear room.I P3 SAMA 048 -Cost as corrected in response to RAI 4e (Reference 2)Cost estimate revised from what was previously reported.

See Section [6] for more information.

NL-09-165 Attachment 1 Docket Nos. 50-247 & 50-286 Page 29 of 33[7] Main Steam Safety Valve Gagging SAMA (Updated Response to Round 2 RAI 6)The benefit associated with installing a device to gag a stuck-open main steam safety valve following a steam generator tube rupture (SGTR) was originally assessed in response to Round 2 RAI 6 (Reference 3). In that response, the estimated benefit with uncertainty assuming that this SAMA is fully successful in preventing all thermally-induced steam generator tube ruptures was almost $3 million for IP2 and over $4 million for IP3. As indicated in that response, with an estimated cost of $50,000, this additional SAMA is potentially cost beneficial and it has been submitted for engineering project cost benefit analysis for more detailed examination of viability and implementation cost. With the revised meteorological input data used for the SAMA reanalysis, the total benefit of this SAMA is now estimated to be about $13 million for IP2 and $19 million for IP3.[8] TI-SGTR Sensitivity Analysis (Revised Response to Round 2 RAI 5)In response to Round 2 RAI 5 (Reference 3), a sensitivity study was performed to determine the impact of applying values derived from NUREG-1570.

The T!-SGTR sensitivity study was performed again, as described below, to determine the impact of applying NUREG-1 570 values to the SAMA reanalysis and provide an updated response to Round 2 RAI 5.The full lists of IP2 and IP3 Phase II SAMAs were reviewed for impact. Of those, the following twenty-seven IP2 SAMAs and twenty-two.IP3 SAMAs were identified as potentially impacted by the TI-SGTR assumption.

IP2 SAMAs: 1, 6, 18, 19, 20, 25, 26, 27, 28, 29,30, 31, 32, 35, 39, 40, 42,*44, 46, 52, 54, 59, 60, 61, 62, 65, 66 IP3 SAMAs: 1, 16, 17, 18, 23, 24, 25, 26, 27, 28, 29, 30, 33, 38,.40, 42, 43, 55, 56, 58, 61, 62 Since IP2 SAMAs 28, 44, 54, 60, 61, 62 and 65 and IP3 SAMAs 55, 61 and 62 were previously determined to be potentially cost beneficial, they were not re-evaluated.

Of the remaining SAMAs, those for which the irmplementation cost outweighed the benefit by less than a factor of, five were re-evaluated.

This screening criterion was applied to facilitate the re-evaluation by limiting it to those potentially impacted SAMA candidates with a realistic possibility of becoming cost beneficial.

The appropriateness of this screening criterion is justified by the fact that only.one of the twelve SAMAs evaluated was found to be potentially cost beneficial following this conservative sensitivity analysis.

See paragraph prior to Table 6 for discussion of conservatism.

The SAMAs re-evaluated were: IP2 SAMAs: 1, 6, 25, 29, 40, 52 IP3 SAMAs: 1, 16, 18, 30, 40, 43 The baseline case (Table 5.8 of NUREG-1570) associated with moderate tube degradation was used for this sensitivity study. The full conditional induced SGTR value (0.25) shown for that case was used. The NUREG-1 570 conditional probability was applied to all high/dry sequences in the Level 2 model for each unit; in both station blackout and transient sequences.

The benefit values in this sensitivity analysis included the additional impact of the loss of tourism and business.

Tables 6 and 7 show the values for the IP2 and IP3 SAMAs evaluated in this NL-09-165* Attachment 1 Docket Nos. 50-247 & 50-286 Page 30 of 33 sensitivity analysis.

While the severe accident costs of both the baseline case and the individual SAMAs increased, the extent to which the revised TI-SGTR assumption impacted the benefit varied, based on the nature of the specific SAMA.IP3 SAMA 18 was found potentially cost beneficial as a result of this sensitivity analysis.Although the NUREG-1 570 baseline case values were used for this sensitivity analysis, the.baseline case applies to a steam generator with a moderate flaw distribution.

The IP2 and IP3 steam generators have been replaced and are being maintained in accordance with the stringent standards recommended by NEI 97-06. The 1P2 and 1P3 steam generators have only 0.19% and 0.12% of the tubes plugged, and would be classified as "pristine" in accordance with generic criteria established by Westinghouse for categorizing steam generator tube integrity.

Corrosion has not been observed in either the IP2 or IP3 steam generators.

Therefore, use of the baseline case for this sensitivity study is conservative relative to application of the NUREG-1570 results for pristine generators (Table 5.8, Case 8).Table 6 -IP2 TI-SGTR Sensitivity Results Original TI-SGTR Revised IP2 Phase II SAMA Benefit with Benefit with Estimated Cost Conclusion Uncertainty Uncertainty 001 -Create an independent RCP seal $788,963 $892,287 $1,137,000 Not cost injection system with a effective dedicated diesel. ._006 -Add a diesel building $64,202 $223,493 $274,000 Not cost high temperature alarm. effective Not cost 025 -Improve MSIV design. $258,310 $430,516 $476,000 effetive* .effective 029 -Increase/improve DC Not cost bus load shedding.

$102,574 $257,560 $460"000t effective 040 -Createlenhance a Not cost reactor coolant $1,205,070

$1,325,614

$2,000,000t effective depressurization system.052 -Install secondary side guard pipes up to the $619,756 $878,065 $1,100,000 Not cost MSI~s.effective MSIVs.II t Cost estimate revised from what was previously reported.

See Section [6] for more information.

NL-09-165 Attachment 1 Docket Nos. 50-247 & 50-286 Page 31 of 33 Table 7- IP3 TI-SGTR Sensitivity Results Original TI-SGTR Revised IP3 Phase II SAMA Benefit with Benefit with Estimated Cost Conclusion Uncertainty Uncertainty 001 -Create an independent RCP seal $342,913 $480,678 $1,137,000 Not cost injection system with a effective dedicated diesel.016 -Install a highly reliable steam generator shell-side Not cost heat removal system that $2,031,473

$2,289,783

$7,400,000 effective relies on natural circulation and stored water sources.018 -Route the discharge from the main steam safety valves through a Retain*structurewhere a water $4,821,779

$14,637,545

$12,000,000t spray would condense the (New)steam and remove most ofithefission products.030 -Provide a portable Not cost diesel-driven battery $309,222 $515,869 $938,000t effective charger.040 -Provide automatic nitrogen ýbackup to steam $950 Not cost generator atmospheric

$344,225 effective dump valves.043- Use fire water system Not cost as backup for steam $652,885 $825,091 $1,656,000 effective generator inventory.

t Cost estimate revised from what was previously reported.information.

See Section [6] for more[9] Conclusion In the SAMA reanalysis using a conservatively representative, single year of meteorological data (2000), the following additional three SAMA candidates were found to be potentially cost beneficial for mitigating the consequences of a severe accident for IP2 (in addition to those previously designated as cost beneficial in Section 4.21.6 of the ER and References 2 and 3).021 -Install additional pressure or leak monitoring instrumentation for interfacing system loss of coolant accidents (ISLOCAs)022 -Add redundant and diverse limit switches to each containment isolation valve 062 -Provide a hard-wired connection to a safety injection (SI) pump from the alternate safe shutdown system (ASSS) power supply, NL-09-165 Attachment 1 Docket Nos. 50-247 & 50-286 Page 32 of 33 In the SAMA reanalysis using a conservatively representative, single year of meteorological data, the following three SAMA candidates were found .to be potentially cost beneficial for mitigating the consequences of a severe accident for IP3 (in addition to those previously designated as cost beneficial in Section 4.21.6 of the ER and References 2 and 3).007 -Create a reactor cavity flooding system 018 -Route the discharge from the main steam safety valves through. a structure where a water spray would condense the steam and remove most of the fission products (cost beneficial in TI-SGTR sensitivity in Section [8])019 -Install additional pressure or leak monitoring instrumentation for ISLOCAs As described in the aging management review results for the integrated plant assessment presented in Sections 3.1 through 3.6.of the license renewa! application, IP2 and IP3 have programs for managing aging effects for components within the scope of license renewal (Reference 1). Since these programs are sufficient to manage the effects of aging during the license renewal period without implementation of the above SAMA candidates for IP2 and IP3, these potentially cost beneficial SAMAs need not be implemented as part of license renewal pursuant to 10 CFR Part 54. However, consistent with those SAMAs identified previously as cost beneficial, the above potentially cost beneficial SAMAs have been submitted for engineering project cost benefit analysis..Since some of the potentially cost beneficial SAMAs address the same risk contributors, implementation of an optimal subset of these SAMAs could achieve a large portion of the total risk reduction at a fraction of the cost, and render the remaining SAMAs no longer cost beneficial.

IP2 SAMAs 54, 65, and the main steam safety valve gagging SAMA have the highest priority for implementation due to their potential for significant risk reduction and relatively low implementation cost (cost estimate is less than 20% of the benefit with uncertainty).

SAMAs 9, 21, 28, 44, 53, and 56 would have Second-priority based on their potential for risk reduction and their mitigation of plant risk contributors not addressed by the highest priority SAMAs. .The remaining potentially cost beneficial SAMAs (22, 60, 61, and62) are considered lowest priority because their benefit and cost estimates are similar or because their benefit is expected to be reduced significantly if the higher priority SAMAs are implemented.

IP3 SAMAs 52, 61, 62, and the main steam safety valve.gagging SAMA have the highest priority for implementation due to their potential for significant risk reduction and relatively low implementation cost (cost estimate is less than 20% of the benefit with uncertainty).

SAMAs 7, 53, and 55 would have second priority based on their potential for risk reduction and their mitigation of plant risk contributors not addressed by the highest priority SAMAs. The remaining potentially cost beneficial SAMAs (18 and 19) are considered lowest priority because their* benefit and cost estimates are similar or because their benefit is expected to be reduced significantly if the higher priority SAMAs are implemented.

NL-09-165 Attachment 1 Docket Nos. 50-247 & 50-286.Page 33.of 33[10] MACCS2 Input Files The following MACCS2 input files, used in the analysis described above, are provided in electronic format.Filename Description siteiec.inp site input file with loss of tourism and business metiO0.inp meteorological data for year 2000 chrbiec.inp chronc input file with loss of tourism and business earbi-noE.inp early input file atmbi2ns.inp:

atmos input file for IP2 atmbi3ns.inp atmos input file for IP3[11] References

1. Entergy Letter NL-07-039,:

Indian Point Energy Center License Renewal Application, April 23, 2007 2. Entergy Letter NL-08-028, Reply to Request for Additional Information Regarding License Renewal Application

-Severe Accident Mitigation Alternatives Analysis, February 05, 2008 3. Entergy Letter NL-08-086, Supplemental Reply to Request for Additional Information Regarding License Renewal Application

-Severe Accident Mitigation Alternatives Analysis, May 22, 2008.4. NEI 05-01, Severe Accident Mitigation Alternatives (SAMA) Analysis Guidance Document [Revision A], November 2005 5. Entergy Letter NL-09-151, Entergy Nuclear Operations Inc. Telephone.

Conference Call Regarding Met Tower Data for SAMA Analysis Indian Point Nuclear Generating Unit Nos. 2 & 3, November 16, 2009 6. Procedures for Substituting Values for Missing IWS Meteorological Data for Use in Regulatory Air Quality Models, Dennis Atkinson and Russell F. Lee, July 7, 1992[May be found on the "Meteorological Guidance" page at epa.gov.]7. Severe Accident Mitigation Alternatives Analysis Applicant's Environmental Report for License Renewal, Calvert Cliffs Nuclear Power, April 1998.

Applicant's Consolidated Memorandum in Opposition to New York State's Motion for Summary Disposition of Contention NYS-35/36 and in Support of Its Cross-Motion for Summary Disposition (February 3, 2011)NYS-35/36:

ATTACHMENT 12 Entergy Indian Point Energy Center 450 Broadway.

GSB P.O. Box 249 Buchanan.

N.Y. 10511-0249 Tel (914) 788-2055 Fred Dacimo Vice President License Renewal NL-10-013 January 14, 2009 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

License Renewal Application

-Supplement to SAMA Reanalysis Using Alternate Meteorological Tower Data Indian Point Nuclear Generating Unit Nos. 2 and 3 Docket Nos. 50-247 and 50-286 License Nos. DPR-26 and DPR-64

REFERENCE:

1. Entergy Nuclear Operations Inc. Letter NL-09-165, "SAMA Using Alternate Meteorological Tower Data" dated December 11, 2009.

Dear Sir or Madam:

In Reference 1 above, Entergy Nuclear Operations, Inc (Entergy) provided results of a SAMA reanalysis using alternate meteorological data.The purpose of this letter is to provide additional,.

clarifying information related to the SAMA reanalysis to answer questions raised during a teleconference between Entergy and NRC on January 7, 2010.1) Summaries of the revised implementation cost estimates for IP2 SAMAs 17 and 40 and IP3 SAMAs 17, 20, 40, and-50 are attached.

The estimates are similar to the estimate provided in response to request for additional information 51.2) With the revised population dose and offsite economic costs resulting from the SAMA reanalysis described in the referenced letter, the total present dollar-value equivalent benefit associated with completely eliminating severe accidents caused by internal events is about $4.5 million for IP2 and $5.1 million for IP3. Use of a multiplier of 3.8 for IP2 and 5.5 for IP3 to account for NL-10-013 Docket Nos. 50-247 and 50-286 Page 2 of 2 external events increases the value to $17 million for IP2 and $28 million for IP3 and represents the dollar value benefit associated with completely eliminating the risk of severe accidents caused by all internal and external events at IP2 and IP3, respectively.

3) The revised SAMA implementation cost estimates presented in the referenced letter do not include replacement power costs.The revised SAMA implementation cost estimates do not include a markup to account for inflation.

However, the revised estimates were developed using Entergy's standard process for development of conceptual level project estimates utilizing spreadsheets containing 2009 rates for material, labor, insurance, fees, etc. Since. the original implementation estimates were developed prior to 2009, the new estimates inherently account for inflation since the time of the earlier estimates.

There are no new commitments identified in this submittal.

If you have any questions, or require additional information, please contact Mr. Robert Walpole at 914-734-6710.

Sincerely, FD/mb

Attachment:

1. License Renewal Application , Summaries of Revised SAMA Implementation Cost Estimates cc: Mr. S. J. Collins, Regional Administrator, NRC Region I Mr. J. Boska, Senior Project Manager, NRC, NRR, DORL Mr. Sherwin E. Turk, NRC Office of General Counsel, Special Counsel Ms. Kimberly Green, NRC Safety Project Manager NRC Resident Inspectors Office, Indian Point Mr. Paul Eddy, NYS Dept. of Public Service Mr. Francis J. Murray, Jr., President and CEO, NYSERDA ATTACHMENT 1 TO NL-10-013 License Renewal Application

-Summaries of Revised SAMA Implementation Cost Estimates ENTERGY NUCLEAR OPERATIONS, INC INDIAN POINT NUCLEAR GENERATING UNITS 2 AND 3 DOCKET NOS. 50-247 & 50-286 Attachment 1 NL-10-013 Docket Nos. 50-247 and 50-286 Page 1 of 6 INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 LICENSE RENEWAL APPLICATION SUPPLEMENT TO SAMA REANALYSIS IP2 SAMA 17 Implementation Cost Estimate The SAMA improvement would be to develop a new modification that would provide an additional stainless steel barrier to protect the containment liner from ejected core debris at high pressure.

This modification would require extensive design work to create the modification package. The associated design calculations include seismic analysis, loading analysis, and possible associated analysis for piping or penetration interferences.

Specific issues to be addressed for this modification include the following.

1) Change licensing basis, which may require Nuclear Regulatory Commission involvement.

2) Modify or provide new floor loading calculation, seismic analysis, and reactor building volume calculation.

3) Revise procedures and training.4) Install and test the new barrier.5) Change various documents, i.e. Technical Specification Basis, Final Safety Analysis Report, system descriptions, design, basis documents, preventative maintenance tasks.6) Change safety analysis calculations.

The following is a breakdown of the cost estimate.Task. Description Cost $Develop modification documents, including calculations 377,824 and drawings Procedure changes and training 12,188 Installation, material and labor 2,364,167 Safety related:cost increase i.e. quality assurance, 192,969 quality control, material increase, etc.Project management/

modification support 495,767 Installer mobilization, tools and training, construction 680,779 management fee, insurance, performance bond Installation contingency considering lack of design 465,426 details Indirect charges (loaders) 841,467 5,430,587 Total (rounded to 5,500,000)

Attachment 1 NL-10-013 Docket Nos. 50-247 and 50-286 Page 2 of 6 IP2 SAMA 40 Implementation Cost Estimate The SAMA improvement would be to install bypasses around the existing atmospheric dump valves to depressurize the reactor coolant system. This would allow successful low pressure emergency core cooling system injection following a small loss of coolant accident and high pressure safety injection failure. This modification would require extensive design work to create the modification package. The associated design calculations would include seismic analysis, loading analysis, setpoint calculations, environmental qualification (EQ) calculations, and possible associated analysis for piping interferences.

Specific issues to be addressed for this modification include the following.

1) Change licensing basis, which may require Nuclear Regulatory Commission involvement.

2) Modify or provide new seismic analysis, loading. analysis, setpoint calculations and EQ datasheets.

3) Revise procedures and training.4) Install and test the new hardware.5) Change various documents, Le. Technical Specification Basis, Final Safety Analysis Report, system descriptions, design basis documents, preventative maintenance tasks.6) Change simulator and associated procedures.

7) Change safety analysis calculations.

The following is a breakdown of the cost estimate.Task Description Cost $Develop modification documents, including calculations 299,250 and drawings.Procedure changes and training 25,000 Installation, material and labor 360,000 Safety related cost.increase i.e. quality assurance, 178,500 qualityicontrol, matefialrincrease,,etc Project management

/ modification support 89,250 Installer mobilization, tools and training, construction 178,500 management fee, insurance, performance bond Installation contingency considering lack of design 476,000 details Indirect charges (loaders) 321,300.1,927,800 Total (rounded to 2,000,000)

Attachment 1 NL-10-013 Docket Nos. 50-247 and 50-286 Page 3 of 6 IP3 SAMA 17 Implementation Cost Estimate The SAMA improvement would be to increase secondary side pressure capacity to reduce the frequency of steam generator tube rupture events and decrease the likelihood of relief valves lifting and subsequently failing to reseat.It was assumed based on engineering judgment that the nuclear steam supply system vendor would not be able to qualify the existing.components to a high enough pressure rating to meet the intent of the SAMA. Therefore, the steam generators, piping from the feedwater regulating valves through the steam generators and up to the main steam isolation valves, feedwater regulators and main steam isolation valves would have to be replaced with upgraded components.

The associated instrumentation would also have to be upgraded.The associated design calculations would include seismic analysis, loading analysis, setpoint calculations, EQ calculations, and possible associated analysis for piping interferences.

Specific issues to be addressed for this modification include the following.

1) Change licensing basis, which may require Nuclear Regulatory Commission involvement.

2) Modify or provide new seismic analysis, loading analysis, setpoint calculations and EQ datasheets.

3) Revise procedures and training.4) Install and test the new hardware.5) Change various documents, i.e. Technical Specification Basis, Final Safety Analysis Report, system descriptions, design basis documents, preventative maintenance tasks.6) Change simulator and associated procedures.

7) Change safety analysis calculations.

The following is a breakdown of the cost estimate.Task Description Cost $Develop modification documents, including 51,458,750 calculations and drawings Procedure:

changes and training 25,000 Installation, material and labor 330,000,000 Safety related cost increase i.e. quality assurance, 99,517,500 quality control, material increase, etc Project management

/ Modification Support 49,758,750 Installer mobilization, tools and training, 99,517,500 construction management fee, insurance, p~erformance bond Installation contingency considering lack of design 265,380,000 details Indirect charges (loaders)

.179,131,500 1,074,789,000 Total (shown as> 100, 000, 000)

Attachment 1 NL-10-013 Docket Nos. 50-247 and 50-286 Page 4 of 6 IP3 SAMA 20 Implementation Cost Estimate The SAMA improvement would be to develop a new modification that would provide redundant and diverse limit switches to each containment isolation valve. This modification would provide additional indications of valve position by monitoring the valve position with one of two different instruments, which would improve the reliability or availability of the indication.

This modification would require extensive design work to create the modification package. The associated design calculations would include seismic analysis, loading analysis, loop error calculations, setpoint calculations, EQ calculations, and possible associated analysis for safe shutdown.Specific issues to be addressed for this modification include the following.

1) Change licensing basis, which may require Nuclear Regulatory Commission involvement.

2) Modify or provide new seismic analysis, setpoint calculations, limit and torque switch settings, EQ datasheets.

3) Revise procedures and training..4) Install and test the new instrument loops.5) Change various documents, i.e. Technical Specification Basis, Final Safety Analysis Report, system descriptions, design basis, documents, preventative maintenance tasks.6) Change simulator and associated procedures.

7) Change safety analysis calculations.

The following is a breakdown of the cost estimate.Task:Description Cost $Develop modification documents, including calculations 741,191 and drawings Procedure changes and training 35,500 Installation, material and labor* 1,469,816 Safety related cost increase i.e. quality assurance, -125,980 quality control, material increase, etc.Project. management

/ modification support 302,738 Installer mobilization, tools and training, construction 270,762 management fee, insurance, performance bond Installation contingency considering lack of design 377,663 details Indirect charges (loaders)

.594,921 3,918,571 Total (rounded to 4,000,000)

Attachment 1 NL-10-013 Docket Nos. 50-247 and 50-286 Page 5 of 6 IP3 SAMA 40 Implementation Cost Estimate The SAMA improvement would be to provide automatic nitrogen backup to the steam generator atmospheric dump valves. For each of the four atmospheric dump valves, the modification would use existing nitrogen at the panel and would tie into the instrument air supply using a regulator and check valves in both the nitrogen and IA lines to prevent back feed. A local pressure gage, low pressure switch to the nitrogen line and an alarm in control room would also be required.

The associated design calculations would include setpoint calculations, EQ calculations, and possible associated analysis for piping interferences.

Specific issues to be addressed for this modification include the following.

1) Change licensing basis, which may require Nuclear Regulatory Commission involvement.

2) Modify or provide new setpoint calculations and EQ datasheets.

3) Revise procedures and training.4) Install and test the new hardware.5) Change various documents, i.e. Technical Specification Basis, Final Safety Analysis Report, system descriptions, design basis documents, preventative maintenance tasks.6) Change simulator and associated procedures.

7) Change safety analysis calculations.

The following is a breakdown of the cost estimate.Task- Description Cost $Develop modification documents, including calculations and drawings 259,500.Procedure changes and training 25,000 Installation, material and labor 95,000 Safety related, cost increase .i. e. quality assurance, quality control, material increase; etc. -0 Project management

/ modification support 49,500 Installer mobilization, tools and training, construction management fee, insurance, performance bond 99,000 Installation contingency considering lack of design details 264,000 Indirect charges (loaders) 158,400 950,400 Total (rounded to 950,000)

J Attachment 1 NL-10-013 Docket Nos. 50-247 and 50-286 Page 6 of 6 IP3 SAMA 50 Cost Estimate The SAMA improvement would be to install secondary guard pipes up to the main steam isolation valves. This modification would require extensive design work to create the modification package. The associated design calculations would include thermal analysis, seismic analysis, setpoint calculations, EQ calculations, and possible associated analysis for piping interferences.

Specific issues to be addressed for this modification include the following.

1) Change licensing basis, which may require Nuclear Regulatory Commission involvement.

2) Modifyor provide new thermal analysis, seismic analysis, setpoint calculations andEQ datasheets.

3) Revise procedures and training.4) Install and test the new hardware.5) Change various documents, i.e. Technical Specification Basis, Final Safety Analysis Report, system descriptions, design basis documents, preventative maintenance tasks.6) Change simulator and associated procedures.

7) Change safety analysis calculations.

The following is a breakdown of the cost estimate.TaskoDescrIption Cost $Develop modification documents, including calculations and drawings 1,847,750 Procedure .changes~and training 25,000 Installation, materialand labor 1,560;,000 Safety related-cost increase i.e. quality assurance, quality control,.

material increase, etc 895,500 Project. management

/ modification support 447,750 Installer mobilization, tools and training, construction .management fee, insurance, performance bond 895,500 Installation contingency considering lack of design details 2,388,000 Indirect charges (loaders) 1,611,900 9,671,400 Total (rounded to 9,671,000)

Applicant's Consolidated Memorandum in Opposition to New York State's Motion for Summary Disposition of Contention NYS-35/36 and in Support of Its Cross-Motion.for Summary Disposition.(February 3, 2011)NYS-35/36:

ATTACHMENT 13 10834 Proposed Rules Federal Register Vol. 66, No. 34 Tuesday, February 20, 2001 This section of the FEDERAL REGISTER contains notices to the public of the proposed issuance of rules and regulations.

The purpose of these notices is to give interested persons an opportunity to participate in the rule making prior to the adoption of the final rules.NUCLEAR REGULATORY COMMISSION 10 CFR Part 51[Docket No. PRM 51-7]Nuclear Energy Institute; Denial of Petition for Rulemaking AGENCY: Nuclear Regulatory Commission.

ACTION: Denial of petition for rulemaking.

SUMMARY

The Nuclear Regulatory Commission (NRC) is denying a petition for rulemaking submitted by the Nuclear Energy Institute (NEI) (PRM-51-7).

The;petitioner requested that the NRC amend its regulations to delete'the requirement to consider Severe Accident Mitigation Alternatives

'(SAMAs) as part of the environmental review to support license renewal decisions.

The NRCis denying the petition because the NRC must continue to consider SAMAs for issuance of a new or renewed operating license for a power reactor in order to meet its responsibilities under the National Environmental Policy Act (NEPA), notwithstanding the legal arguments presented in the petition.

However, the NRC staff will continue to work with stakeholders to determine if efficiencies in the conduct of SAMA analyses for environmental reviews can be realized.ADDRESSES:

Copies of the petition for rulemaking, the public comments received, and the NRC's letter of denial to the petitioner are available for public inspection or copying for a fee, at the NRC's Public Document Room, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland.

These documents are also available at the NRC's rulemaking website at http://ruleforum.llnl.gov FOR FURTHER INFORMATION CONTACT: Donald P. Cleary, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone (301) 415-3903, e-mail dpc@nrc.gov.

SUPPLEMENTARY INFORMATION:

Background

On July 14, 1999, the NRC received a petition for rulemaking submitted by the NEI, by letter dated July 13, 1999. On September 2, 1999 (64 FR 48117), the NRC published a notice of receipt of the petition (PRM-51-7).

The petitioner requested that the NRC amend its regulations to delete the requirement for the NRC to evaluate Severe Accident Mitigation Alternatives (SAMAs) as part of its National Environmental Policy Act (NEPA) review associated with license renewal. The petitioner requests that the NRC take this action to achieve consistency in the scope of its regulatory requirements for environmental, protection under NEPA, 10 CFR part 51, and its technical requirements for license renewal under the Atomic Energy Act, 10 CFR part 54.The technical requirements for renewal of operating licenses are specified in 10 CFR part 54 (60 FR 22461; May 8, 1995). This regulation focuses the license renewal review on certain types of systems,.

structures, and components that the NRC has determined require evaluation to ensure that the effects of aging will be adequately managed in the period of extended operation.

This regulation is based on two regulatory principles.

The first principle of license renewal is that, with the possible exception of the detrimental effects of aging on the functionality of.certain plant systems, structures, and components in the period of extended operation and possibly a few other issues related to safety only during extended operation, the ongoing regulatory process is adequate to ensure that the licensing bases of all currently operating plants provide and maintain an acceptable level of safety. The second principle of license renewal is that the plant-specific licensing basis must be maintained during the renewal term in the same manner and to the same extent as during the original licensing term. This principle is attained, in part, through a program of age-related degradation management for systems, structures, and components that are within the scope of license renewal. There is no requirement in 10 CFR part 54 for analysis of SAMAs.The NRC's regulations implementing NEPA appear in 10 CFR part 51. The regulations contain specific provisions related to the requirements for the environmental review of applications to renew the operating licenses of nuclear power plants. See, for example, 10 CFR 51.53(c) and Subpart A, Appendix B.The regulations were developed to improve the efficiency of the process of environmental review for applicants seeking to renew a nuclear power plant operating license for up to an additional 20 years. The regulations are based on generic analyses reported in NUREG-1437, "Generic Environmental Impact Statement for License Renewal of Nuclear Plants" (May 1996) and in part on NUREG-1437, Vol. 1, Addendum 1 (August 1999). Those environmental issues for which the NRC made generic findings that may be adopted in individual plant license renewal reviews are defined as Category 1 issues in the rule. Those environmental issues that require further site-specific review are defined as Category 2 issues in the rule. The regulations also provide for the consideration of "new and significant information" that might change a previous finding or introduce issues not previously reviewed and codified in the regulations.

With respect to the issue of environmental effects of severe accidents from license renewal, the NRC found that the probability weighted consequences are small. Specifically, the regulations state in Table B-1: "The probability-weighted consequences of atmospheric releases, fallout onto open bodies of water, releases to groundwater, and societal and economic impacts from severe accidents are small for all plants." Accordingly, the impacts of severe accidents are encoded inthe rule and are not open for review in individual license renewal actions.However, one of the criteria for a Category 1 finding is, as stated in footnote 2 of Table B-1, Part 31,"Mitigation of adverse impacts associated with the issue have been considered in the analysis, and it has been determined that additional plant-specific mitigation measures are likely not to be sufficiently beneficial to warrant implementation." At the time the final rule was promulgated in 1996, the NRC discussed the ongoing regulatory programs focused on individual plant vulnerabilities to severe accidents and cost-beneficial improvements for reducing severe HeinOnline

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66, No. 34/Tuesday, February 20, 2001/Proposed Rules 10835 accident frequency or consequences.

For each plant, an individual plant examination (IPE) to look for plant vulnerabilities to internally initiated events and a separate IPE for externally initiated events (IPEEE) was performed (61 FR 28467; June 5, 1996). The NRC believed that it would be premature to reach a generic conclusion regarding severe accident mitigation alternatives before completing these programs.Therefore, even though the Commission has reached a generic conclusion on the magnitude of severe accident impacts, the issue is nevertheless designated as a Category 2 issue because of the unresolved questions regarding mitigation, and applicants for license renewal are subject to the following requirement at 10 CFR 51.53(c)(3)(ii)(L): "If the staff has not previously considered severe accident mitigation alternatives for the applicant's plant in an environmental impact statement or in an environmental assessment, a consideration of alternatives to mitigate severe accidents must be provided." The NRC stated, "* *

that upon completion of its IPE/IPEEE program, it may review the issue of severe accident mitigation for license renewal and consider, by separate rulemaking, reclassifying severe accidents as a Category 1 issue" (61 FR 28481; June 5, 1996).The Petition The petition was submitted by the Nuclear.Energy Institute (NEI) by letter dated July 13, 1999. Its receipt was noticed in the Federal Register on September 2, 1999 (64 FR 48117), with a full description of its content. The petitioner requested the NRC "* *

to delete 10 CFR 51.53(c)(3)(ii)(L) and, thereby, eliminate the requirement for NRC to evaluate SAMAs as part of the NEPA review associated with license renewal." The rulemaking would include conforming changes to 10 CFR part 51, Appendix B and NUREC-1437.

The petitioner requests elimination of the requirement for SAMA reviews in 10 CFR part 51 on the belief that the requirement conflicts with the technical requirements for license renewal in 10 CFR part 54. The petitioner argues that actions to evaluate and address SAMAs are part of each licensee's current licensing basis and that 10 CFR part 54 is designed to separate matters related to maintaining the current licensing basis from those considered in a license renewal review. The petitioner's argument, briefly stated, is as follows.The petition makes reference to the two principles of license renewal, discussed in the Background section above. The first principle focuses the license renewal review on age-related degradation of plant systems, structures, and components.

The second principle is continuation of the current licensing basis during the renewal term, in part, through a program of age-related degradation management of systems, structures, and components that are important to license renewal. The petitioner notes that 10 CFR 54.39,"Matters not subject to a renewal review," specifically provides that deviations from the current licensing basis identified in the integrated plant assessment performed for license renewal will be corrected under the terms of the current license and are not within the scope of the license renewal review. The petitioner then states that actions to evaluate and address SAMAs are part of each licensee's current licensing basis, citing the IPE and IPEEE program to identify and evaluate plant-specific severe accident vulnerabilities and ways to mitigate those vulnerabilities.

-Concluding that SAMAs are outside of the scope of a 10 CFR part. 54 license renewal review, the petitioner then presents legal arguments for deleting SAMAs from the NEPA review. The essence of these arguments is that 10 CFR part 54 defines the scope of the proposed Federal action, and that Federal action establishes the scope of environmental consequences of license renewal that are to be reviewed under NEPA. Citing several court cases, the petitioner asserts that this approach is consistent with the "rule of reason" that generally governs environmental impact reviews under NEPA. The petitioner then states, "Thus, under the 'rule of reason,' the impacts appropriately considered under NEPA would be those that reasonably flow, from the part 54 decision-making." Next, the petitioner cites two cases to support the position that there should be no consideration of SAMAs for license renewal. In City of Aurora v. Hunt, the court ruled that a new procedure to use a specific airport runway in particular weather conditions involved "* *

no significant safety impact *

to trigger further assessment or inquiry under NEPA." 749 F.2d 1457, 1468 n. 8 (10th Cir.1984) overruled on other grounds by Village of Los Ranchos de Albuquerque

v. Marsh, 956 F.2d 970 (10th Cir. 1992).In the second court case, Upper Snake River Chapter of Trout Unlimited v.Hodel, the court ruled that the Department of Interior did not have to prepare an environmental impact statement (EIS) to adjust the flow of water from a dam to accommodate drought conditions where the range of flow change was within the contemplation of the original project.921 F.2d 232, 235 (9th Cir. 1990). The petitioner concludes from these decisions that a NEPA review of SAMAs is not required in the license renewal review because, (1) the current licensing basis is not subject to evaluation in a license renewal review, and (2) by maintaining the current licensing basis in the renewal term, there will be no change in risk of a severe accident due to license renewal.The petitioner goes on to assert that NRC's requirement to include SAMAs in NEPA license renewal reviews was based on an overly broad application of language in the-Limerick Ecology Action v. NRC, 869 F.2d 719 (3rd Cir. 1989),.decision and that the decision "* * *leaves undisturbed the proposition that the 'rule of reason' defines whether the EIS has addressed the significant aspects of the probable environmental consequences

*

that reasonably may flow from the proposed action-renewing a plant's license as that plant is currently designed and operated." Finally, citing a number of court cases, the petitioner argues that "* * *judicial precedents allow the NRC to eliminate SAMAs from consideration in license renewal proceedings based on a determination, through proper rulemaking, that severe accidents are highly unlikely." Public Comments on the Petition The NRC received letters from 11 commenters.

Ten of the comment letters supported the petition.

Nine of those letters were from nuclear utilities and the tenth was from NEI, providing supplemental information to support the arguments made in the petition.Except for one comment, Comment 1 below, all of the comments made by supporters of the petition reiterated arguments made in the petition.

Because those arguments are addressed in the NRC's reasons for denying the petition they are not addressed in the comment response below. A public interest group provided the one letter opposed to the petition; and NRC's responses to their comments are provided below.Comment 1: A utility commented that the costs of performing the SAMA reviews required by Part 51 are not justified when compared to the small potential safety benefits that result from the reviews, when the costs associated with implementing changes to realize those benefits are evaluated, and when the fact that the reviews are largely duplicative of the previously completed Individual Plant Examination (IPE) and Individual Plant Examination for HeinOnline

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66, No. 34/Tuesday, February 20, 2001 /Proposed Rules External Events (IPEEE) programs is considered.

Response:

The NRC believes that it should continue to consider SAMAs for individual license renewal applications to continue to meet its responsibilities under NEPA. That statute requires NRC to analyze the environmental impacts of its actions and consider those impacts in its decisionmaking.

In doing so, Section 102(2)(C) of NEPA implicitly requires agencies to consider measures to mitigate those impacts when preparing impact statements.

See Robertson

v. Methow Valley Citizens Council, 490 U.S. 332 (1989). NRC's obligation to consider mitigation exists whether or not mitigation is ultimately found to be cost-beneficial and whether or not mitigation ultimately will be implemented by the licensee.

Id. The NRC understands that a SAMA analysis*can be relatively expensive and is prepared to discuss ways in which SAMA analyses can be conducted efficiently while, at the same time, ensuring that NRC meets its NEPA responsibilities.

Comment 2: Granting the petition would continue the NRC's recent course of "regulatory subtraction" during which it has "methodically amputated.and dismantled its statutory authority." Further, numerous site-specific and generic challenges have precipitated"beyond design basis" events, and demonstrate that it is imperative to maintain Severe Accident Mitigation Alternatives evaluations.

Response:

The NRC has denied the petition because it believes that the legal arguments presented are insufficient to demonstrate that a license renewal NEPA review need not consider alternatives to mitigate the potential for and consequences of severe accidents.

Comment 3: Given the NRC's shrinking budget, "this type of frivolous legal action must be indexed to punitive damages." NEI "must be held accountable, and reimburse the NRC for all legal and administrative costs.associated with this malicious petition." Response:

While NRC has denied the petition, NRC does not believe that there are any aspects of the submittal that would suggest an abuse of the petition process. Accordingly, whether or not reimbursement measures are even available to the Commission, no Commission action is warranted in this regard.Reasons for Denial The Commission is denying the petition for the following reasons: 1. Scope of the License Renewal Rule The petitioner's principal argument for the elimination of SAMAs as part of the NEPA review associated with individual license renewal reviews is that the scope of license renewal establishes a basis. for deleting SAMAs from associated NEPA reviews. -In particular, the petitioner believes that because the NRC's safety review under Part 54 does not require consideration of all aspects of plant operation and administration, the agency's review of environmental impacts under NEPA should be similarly limited. -In its petition and subsequent comments, NEI identified several Federal court cases*and NRC decisions to support its position.'

The petitioner believes that the primary thrust of these cases is that no consideration of impacts is necessary where the proposed Federal action would not change the status quo. In its comments, the petitioner indicated that"[tlhe line of cases using the status quo analysis doesnot turn on maintaining the level of safety per se, but on whether the major federal action will change the operation of the facility sufficient to warrant an inquiry into the changes in environmental effect." The Commission does not find the petitioner's arguments here compelling.

By approving a license renewal application under Part 54, the Commission authorizes operation of the entire plant for an additional 20 years beyond the initial licensing term. Thus, the review of the environmental impacts of this Federal action Under the provisions of Part 51 appropriately involves the consideration of environmental impacts caused by 20 additional years of operation.

The petitioner is correct in stating that the Commission, in promulgating 10 CFR part 54, has limited its safety review under the Atomic Energy Act to certain aspects of the plant that are directly related to aging and other issues specific to the license renewal. The petitioner is also correct in pointing out that many environmental impact issues, such as SAMAs, are not addressed in the NRC's safety review Under Part 54. In fact, the vast majority of environmental impacts from license renewal required to be considered by the NRC under its NEPA review (in accordance with Part 51) are not included in the analysis conducted in fulfilling the NRC's Atomic Energy'City of Aurora v. Hunt, 749 F.2d 1457 (l0th Cir.1984)(overruled on other grounds);

Upper Snake River Chapter of Trout Unlimited

v. Hodel, 921 F.2d 232, 235 (9th Cir. 1990); Consumers Power Company, (Big Rock Point Nuclear Plant), ALAB-636, 13 NRC 312 (1982); and General Electric Company (GE Morris Operation Spent Fuel Storage Facility), LBP-82-14, 15 NRC 530 (1982).Act responsibilities under Part 54 (see;10 CFR part 51 Subpart A, Appendix B, Table B-I).However, under NEPA the NRC is charged with considering all of the environmental impacts of its actions, not just the impacts of specific technical matters that may need to be reviewed to support the action. These impacts may involve matters outside of the NRC's jurisdiction or matters within its jurisdiction that, for sound reasons, are not otherwise addressed in the NRC's safety review during the licensing process. In the case of license renewal, it is the Commission's responsibility under NEPA to consider all environmental impacts stemming from its decision to allow the continued operation of the entire plant for an additional 20 years. The fact that the NRC has determined that it is not necessary to consider a specific matter in conducting its safety review under Part 54 does not excuse it from considering the impact in meeting its NEPA obligations.

The Commission does not believe that the various cases offered by the petitioner provide convincing support for the elimination of the review of SAMAs. It would appear that the logical extension of many of the petitioner's arguments go far beyond the mere elimination of SAMAs consideration from license renewal reviews. Indeed, to the extent that license renewal involves a continuation of impacts already experienced at the site under the current operating license, the arguments made by the petitioner would appear to call for the elimination of almost the entire environmental review of impacts from operation during the license renewal term, a position clearly at odds with the Commission's approach to the matter and also, as discussed below, inconsistent with case law related to relicensing.

The Commission does not dispute that a line of cases exists under NEPA law which excuses agencies from preparing EISs (or considering certain environmental impacts) where the Federal action does not change existing environmental conditions.

See, for example, State of North Carolina v.Federal Aviation Administration, 957 F.2d 1125 (4th Cir. 1992); Cronin v.Department of Agriculture, 919 F.2d 439 (7th Cir. 1990). In most of these cases, the Federal action taken does not itself create any additional impacts to activities that are ongoing and will continue with or without the Federal action. None of these cases appears to provide firm support for the petitioner's argument that the NRC can ignore the impacts of its actions in the context of HeinOnline

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Vol 66, No. 34/Tuesday, February 20, 2001/Proposed Rules 10837 license renewal. In fact, at least one circuit court squarely addressed the issue of relicensing and concluded that there is the need to consider environmental impacts in that context.In Confederated Tribes and Bands of the Yakima Indian Nation v. Federal Energy Regulatory Commission, 746 F.2d 466 (9th Cir. 1984), the Ninth Circuit Court of Appeals considered whether the Federal Energy Regulatory Commission (FERC) was required to prepare an EIS for its relicensing decision for the Rock Island Dam. In response to the FERC's argument that there had been "no change in the status quo" and thus no EIS was necessary, the court found: Relicensing

*

is more akin to an irreversible and irretrievable commitment of a public resource than a mere continuation of the status quo. [Citation omitted] Simply because the same resource had been committed in the past does not make relicensing a phase in a continuous activity.Relicensing involves a new commitment of the resource, which in this case lasts for a forty-year period.2 The court's statements here are consistent with NRC's position and its practice in promulgating and implementing the license renewal rule.The cases offered in support of the petitioner's arguments offer no compelling reasons to alter this approach.In City of Aurora v. Hunt, 3 the Federal Aviation Administration (FAA), through a rulemaking, approved a new approach procedure for the Stapleton airport in order to reduce delays caused by the use of the existing procedure during periods of low visibility.

The City of Aurora challenged the rule claiming, among-other things, that the FAA failed to discuss the safety risks of the new procedure in its environmental assessment.

In ruling against the City's claim, the Court pointed out that the FAA was required by law to issue the new procedure only if it did not involve a change in safety risk. The FAA considered and responded to a vast number of safety concerns as part of the rulemaking process. Accordingly, the Court found that the agency's approval of the procedure, in itself, was adequate to fulfill the agency's responsibility under NEPA. In a footnote, the Court explained that "[wihile an agency may be required to consider the effects that will occur if a risk is realized, where no increase in risk is permitted, as here, no significant safety impact exists to trigger 2746 F.2d 466 at 476-477.749 F.2d 1457 (10th Cir. 1984).further assessment or inquiry under NEPA." 749 F.2d at 1468, n. 8.While certain aspects in the City of Aurora decision provide some general support for the petitioner's argument, the facts in that case do not appear to be sufficiently analogous to support the elimination of SAMAs reviews for license renewal. First of all, the Court found the FAA's decision to permit the new procedure, in essence, served as a finding of an equivalent level of flight safety and thus allowed the FAA to meet its NEPA obligations even though safety was not explicitly considered in the EA itself. Under NRC's license renewal process, NRC's review under Part 54 does not itself meet the agency's NEPA obligations.

Environmental issues such as the potential impacts of severe accidents during the license renewal term do not fall under the Part 54 review. Accordingly, unlike the FAA in City of Aurora, NRC cannot use the Part 54 process as the vehicle for meeting its NEPA responsibilities for considering SAMAs in the license renewal context in the same way that the FAA was allowed to use its procedure approval process in City of Aurora. Secondly, it should be noted that, absent the.NRC's decision to approve a license renewal application, the licensee's plant will not operate an additional 20 years.Accordingly, the NRC's action is a "but for" cause of those additional impacts and NRC has the responsibility to consider those impacts under NEPA. In City of Aurora, the FAA's rule permitted the use of a new landing procedure at the airport. While there is no explicit discussion in the decision, it appears that the current landing procedures at the airport would have continued whether or not FAA had issued the new procedure.

Accordingly, the status quo in the context of the City of Aurora decision appears to have been the continued operation of the airport, whereas the status quo in the context of license renewal is the expiration of the facility's operating license...

Similarly, the decision in Upper Snake River Chapter of Trout Unlimited v. Hodel 4 does not appear to provide strong support for the petitioner's proposal.

In that case, the court found that the reduction in river flows approved by Federal agencies was not a major Federal action within the meaning of NEPA. The court held that, in allowing the flow reductions, the Federal defendants were "simply operating the facility in the manner intended" and that they were doing"nothing new, nor more extensive, nor other than that contemplated when the 921 F.2d 232 (9th Cir. 1990).project was first operational." 921 F.2d at 235. In other words, the flow reductions were part of the normal operations originally approved by the agencies in that case. Conversely, in the license renewal context, the additional

.20 years of operation authorized by a renewed license were not considered during the initial licensing of the* facility.

Thus, the reasoning in Upper Snake River Chapter does not appear to be applicable to NRC's license renewal decisions.

The Commission believes, and has stated before, that a license renewal decision by NRC is a major Federal action that warrants the preparation of an environmental impact statement (61 FR 55637, 66541;December 18, 1996).In submitting comments on its petition, NEI identified several NRC decisions which it believes support its position.

The first, Consumers Power Company (Big Rock Nuclear Plant)ALAB 636, 13 NRC 312 (1982), involved a license amendment request to expand the Big Rock Point Nuclear Plant's spent fuel pool. As NEI indicates, the Appeal Board emphasized the limited scope of the request in rejecting claims that aspects of the plant's continued operation should also be considered in the EA. As quoted by the petitioner, the Appeal Board found that "there are no environmental changes to evaluate" with the secondary or indirect effects (e.g., the plant's continued operation) of the spent fuel pool licensing decision.13 NRC at 328. The petitioner's comments indicate that: The Appeal Board correctly noted that, by granting the license amendment request, the Commission is not also issuing approval to alter any other aspect of the plant's operation or the licensed operating term of the facility.Petition for Rulemaking (Docket No.PRM-51-7; July 13, 1999), letter from NEI to Secretary, NRC, dated November 16, 1999, at pp. 2,.3. The Commission believes that the petitioner's own statement here demonstrates the lack of support Consumers Power Company provides for its own position.

In the context of license renewal, the Commission is, in fact, approving an extension of the licensed operating term of the facility.

Accordingly, the facts in Consumers Power Company are not analogous to those presented by license renewal. While the Commission has appropriately decided through rulemaking that it may focus its safety evaluation on certain matters specified in Part 54, its overall license renewal decision applies to the operation of the entire plant. Therefore, the limited scope considered in Consumers Power HeinOnline

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10838 Federal Register/Vol.

66, No. 34/Tuesday, February 20, 2001/Proposed Rules Company is not present in the license renewal context.Finally, petitioners have also cited General Electric (Morris Operation Spent Fuel Storage Facility)

LBP-82-14, 15 NRC 530 (1982). In that case, the Atomic Safety and Licensing Board ruled that NRC did not have to issue an EIS for the license renewal of a storage facility.

However, in that case, the NRC staff did issue an environmental impact appraisal (referred to under current NRC regulations as an environmental assessment (EA)) for the action. There is no suggestion that the NRC staff was free to eliminate or ignore consideration of the impacts of the action. Rather, the Board agreed with the NRC staff that the impacts of the action were not significant enough to warrant the preparation of a full EIS and, instead, an environmental impact appraisal was sufficient.

The Commission believes that the preparation of EISs, not EAs, are appropriate in the context of license renewal. However, whether an EIS or an EA is prepared for a particular action, the Commission still is responsible for considering the environmental impacts of the action. Accordingly, this case seems to provide little support for the petitioner's position.2. Impact of the Limerick Decision The petitioner is correct in stating that the 3rd Circuit's holding in LimerickEcology Action v. NRC does not itself preclude NRC from ever eliminating SAMA reviews from its licensing actions. Specifically, the court held that the NRC could not generically dispense* with the consideration of SAMAs through a policy statement.

Instead, the NRC would need to do so through a* generic rulemaking similar to the one completed for Table S-3 (10 CFR 51.51)and upheld by the Supreme Court in Baltimore Gas and Electric v. Natural Resources Defense Council, 464 U.S. 87 (1983). Despite the limited nature of its holding, the court in the Limerick decision identified a variety of issues that NRC would have to address in*order to eliminate the consideration of SAMAs. In addition to holdirrg that the NRC could not eliminate consideration of these SAMAs through a policy statement, the court also suggested that the generic consideration of SAMAs may be difficult to accomplish given differences in individual plants. 869 F.2d at 733-739. In addition, as the petitioner has indicated, the court rejected NRC's argument that severe accidents were remote and speculative because there was no basis for this conclusion in the agency's record. Id. at 739-741.Despite the adverse ruling handed to NRC, the Limerick-decision outlines several paths the Commission could attempt to follow in order to eliminate the requirements to analyze both severe accidents and associated mitigation alternatives in individual license renewal reviews. First of all, the Commission could attempt to conclude generically through rulemaking that it has considered these matters and that further consideration in individual license renewal actions is not warranted.

In other words, the NRC would change the designation of the severe accident issue to "Category 1" for license renewal in Appendix B of 10 CFR part 51. Secondly, as discussed in Section 3 of this notice, the Commission could eliminate consideration of SAMAs for license renewal based on a finding that severe accidents, in the context of plant operation during the license renewal term, are remote and speculative.

The Commission believes that insufficient information is available to conclude generically that a SAMA analysis is not warranted for individual plant license renewal reviews. In promulgating the license renewal rule in 1996, the Commission indicated that it"may review the issue of severe accident mitigation for license renewal and consider, by separate rulemaking, reclassifying severe accidents as a Category 1 issue" (61 FR 66537; 66540;December 18, 1996). In early 1999, in anticipation of completion of the IPE and IPEEE programs, the NRC staff began considering the actions needed to fulfill the commitment made in the Federal Register notice. The IPE program has been, completed and the findings of the program are summarized in NUREG-1560, "Individual Plant Examination Program: Perspective on Reactor Safety and Plant Performance," December 1997. The IPEEE program is nearing completion.

The current target for completing the reviews of the balance of the individual submittals is January 2001. A draft insights report -will be issued for public comment in April 2001 and the final report is scheduled to be completed in October 2001.Over the past year, the staff has considered the scope of the analysis that would be required to reach generic technical conclusions supporting a rulemaking to reclassify severe accidents as a Category 1 issue. While the information developed in the IPE/IPEEE program provides a valuable starting point, considerable staff and contractor effort would be required to extend the conclusions resulting from the IPE/IPEEE reviews to draw generic conclusions regarding SAMAs. This would include the need to evaluate changes in plant design and procedures since the IPEs/IPEEEs were completed, incorporate changes in the state of knowledge regarding certain severe accident issues, and to extend the IPE/IPEEE analyses to include offsite consequences.

In addition, both benefit and cost considerations of potential plant improvements would need to be developed.

Further, there is uncertainty whether, at the conclusion of this effort, the staff would be successful in developing a sufficient technical basis to reclassify severe accidents as a Category 1 issue. Given the resources that would be required and the uncertainty in achieving a successful outcome, the staff does not believe it would be cost beneficial to pursue rulemaking at this time.In September 2000, the staff issued Supplement I to Regulatory Guide 4.2,"Preparation of Supplemental Environmental Reports for Applications to Renew Nuclear Power Plant Operating Licenses," which includes guidance on information and analysis content on SAMAs for environmental reports submitted as part of license renewal applications.

Its use is intended to ensure the completeness of the information provided, to assist the NRC staff and others in locating the information, and to shorten the review process. The staff will continue to work with stakeholders to determine if additional efficiencies in the conduct of SAMA analyses for environmental reviews can be realized.

Furthermore, if new information becomes available that indicates it is feasible to reclassify SAMAs to Category 1, the staff will notify the Commission and provide a recommendation as to a course of action.Accordingly, the Commission believes that there is an inadequate basis for a rulemaking to change severe accidents from a Category 2 to Category 1 issue at this time. Applicants should continue to refer to the guidance set out for SAMA analyses in the Statements of Consideration for the license renewal rule (61 FR 28467, 28480-28482; June 5, 1996). The NRC staff will continue to work with stakeholders to discuss the process by which SAMA reviews are done and to determine if efficiencies are possible while ensuring compliance with NRC's NEPA responsibilities to consider the environmental impacts of its licensing decisions.

3. Consideration of Remote and Speculative Impacts The Commission agrees with the petitioner that there is support in the HeinOnline

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66, No. 34/Tuesday, February 20,. 2001 /Proposed Rules 10839* case law for the proposition that NEPA does not require the consideration of remote and speculative risks.5 The court in the Limerick proceeding rejected the NRC's argument that severe accidents were remote and speculative because the court could find no basis for the conclusion in the NRC record. Id. at 739-741. The Commission is not prepared to reach the conclusion that the risks of all severe accidents in the context of license renewal are so unlikely as to warrant their elimination from consideration in our NEPA reviews. Even though there is a low probability of a severe accident, the NRC has invested considerable resources toward understanding potential severe accident sequences and alternatives for further reducing the probability of and mitigating the consequences of severe accidents, but has not yet established an agency record that severe accidents may be eliminated from NRC's NEPA reviews. In reviewing.licensing actions outside of the license renewal context, it may be possible for the NRC to conclude that certain severe accident scenarios are remote and speculative and do not warrant detailed consideration for the purposes of the NEPA review for that particular NRC action. However, for the purposes of consideration of severe accidents in the context of license renewal NEPA reviews, the NRC staff has not developed the necessary basis for concluding that such occurrences are remote and speculative, and thus inappropriate for NRC review under* NEPA. This position does not alter the conclusion that, in light of margins of safety and defense-in-depth, the likelihood of radiological offsite consequences is small.In its comments, the petitioner cited two cases which, in its view, demonstrate that NEPA's requirements are satisfied where potential impacts to the environment are remote and difficult to quantify and ongoing regulatory safeguards are in place to protect against potential risks of impacts into the future. Environmental Defense Fund v. Andrus, 619 F.2d 1368 (1oth Cir. 1980) reh'g en banc denied; and Citizens for Environmental Quality v.Lyng, 731 F. Supp. 970 (D. Colo. 1989).While these cases may provide more support for the general proposition that remote and speculative impacts need not be considered under NEPA, they do not displace the Commission's responsibility to make the threshold determination based on the NRC record 5 See, e.g., Limerick Ecology Action v. NRC, 869 F.2d at 739; San Luis Obispo Mothers for Peace v.NRC, 751 F.2d 1287, 1300-01 (D.C. Cir. 1984).that severe accidents are remote and speculative for the purpose of license renewal reviews. As discussed, the Commission is unable to reach that conclusion.

For the reasons cited in this document, the Commission denies the petition.Dated at Rockville, Maryland, this 13th day of February, 2001.For the Nuclear Regulatory Commission.

Annette L. Vietti-Cook, Secretary of the Commission.

[FR Doc. 01-4104 Filed 2-16-01; 8:45 am]BILUNG CODE 7590-1-P NUCLEAR REGULATORY COMMISSION 10 CFR Parts 73, 76, and 95[Docket No. PRM-76-1]United Plant Guard Workers of America; Denial of Petition for Rulemaking AGENCY: Nuclear Regulatory Commission.

ACTION: Denial of petition for rulemaking.

SUMMARY

The Nuclear Regulatory Commission (NRC) is denying a petition for rulemaking submitted by the United Plant Guard Workers of America (PRM-76-1). The petitioner requested that the NRC amend its regulations concerning security at thegaseous diffusion plants to address sites that have both special nuclear material security concerns and protection of classified matter concern's; to require that these facilities be able to detect, respond to, and mitigate threats of a sabotage event; and to require that the security force be armed and empowered to make arrests in limited situations.

The petitioner believes that these amendments are necessary to address the protection of classified information, equipment and materials, and special nuclear material at the gaseous diffusion plants.ADDRESSES:

Copies of the petition for rulemaking, the public comments received, and NRC's letter to the petitioner may be examined at the NRC Public Document Room, 11555 Rockville Pike, Rockville, MD. These documents also may be viewed and downloaded electronically via the rulemaking website.Documents created or received at the NRC after November 1, 1999 are also available electronically at the NRC's Public Electronic Reading Room on the Internet at http://www.nrc.gov/NRCI ADAMS/index.html.

From this site, the public can gain entry into the NRC's Agencywide Document Access and Management System (ADAMS), which provides text and image files of NRC's public documents.

For more information, contact the NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737, or by e-mail, pdr@nrc.gov.

FOR FURTHER INFORMATION CONTACT: Merri Horn, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone (301) 415-8126, e-mail mlhl@nrc.gov.

SUPPLEMENTARY INFORMATION:

The Petition On May 10, 2000 (65 FR 30018), the NRC published a notice of receipt of a petition for rulemaking filed by the United Plant Guard Workers of America.The petitioner requested that the NRC amend its regulations concerning security at the gaseous diffusion plants to address sites that have both special nuclear material security concerns and.protection of classified matter concerns;to require that these facilities be able to detect, respond to, and mitigate threats of a sabotage event; and to require that the security force be armed and empowered to make arrests in limited situations.

The petitioner believes that these amendments are necessary to address the protection of classified information, equipment and materials, and special nuclear material at the gaseous diffusion plants.First, the petitioner asserted that the regulations do not adequately address sites that have both nuclear material security concerns and classified matter concerns.

The petitioner believes that the applicable regulations were not appropriately merged in the regulations governing gaseous diffusion plants to address a site that covers the protection of classified information, equipment and materials, and special nuclear material.As an example, the petitioner stated that the Controlled Area Fence Line does provide a minimum level of protection against the unauthorized removal of special nuclear material contained in 10- and 20-ton cylinders.

However, the petitioner questioned whether the fence line adequately protects against the unauthorized removal of restricted information, equipment, and other materials or the unauthorized access to these types of materials.

The petitioner asserted that other facilities that possess Category III quantities of special nuclear material regulated by the NRC do not share the HeinOnline

-- 66 Fed. Reg. 10839 2001 Applicant's Consolidated Memorandum in Opposition to New York State's Motion for Summary Disposition of Contention NYS-35/36 and in Support of Its Cross-Motion for Summary Disposition (February 3, 2011)NYS-35/36:

ATTACHMENT 14

.U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation.

NRR OFFICE INSTRUCTION Change Notice Office Instruction No.: Office Instruction Title: Effective Date: Approved By: Date Approved: Primary Contact: LIC-202, Revision 2 Procedures for Managing Plant-Specific Backfits and 50.54(f) Information Requests May 17, 2010 Timothy McGinty May 12, 2010 Andrea Russell 301-415-8553 Andrea.Russell(10n rc.qov Responsible Organization:

NRR/DPR/PGCB Summary of Changes: This is Revision 2 of LIC-202, "Procedures for Managing Plant-Specific Backfits and 50.54(f) Information Requests," which is revised toupdate the primary contact, correct language and formatting errors, clarify the staff process for identifying, determining, resolving, and implementing plant-specific backfits, and provide circumstances where the staff may limit or prohibit discussion of the staffs position that constitutes the potential backfit with the licensee, prior to the backfit determination.

Training: Training summary materials will be provided to the NRR divisions by the implementation date.ADAMS Accession No.: ML092010045 OFFICE DPR/PGCB/PM DPR/PGCB/LA DPR/PCGB/BC DPR/D DE/D DRA/D NAME ARussell CHawes MMurphy TMcGinty PHiland MCunningham DATE 07/21/09 07/23/09 04/22/10 04/26/10 08/11/09 01/14/10 OFFICE DLR/D DORL/D DIRS/D DCI/D DSS/D OGC (NLO)NAME BHolian JGiitter FBrown MEvans WRuland GMizuno (JLubinski for)DATE 10/27/09 08/28/09 07/31/09 08/10/09 03/17/10 03/09/10 OFFICE NRR/DDECS PMDA/D NRR/DDRSP NRR/D NAME JGrobe MGiwines B1oger ELeeds (WRuland for) (TMcGinty for) (TMcGinty for)DATE 04/27/10 05/10/10 05/11/10 05/12/10 OFFICIAL RECORD COPY NRR OFFICE INSTRUCTION LIC-202, Revision 2 Procedures for Managing Plant-Specific Backfits and 50.54(f) Information Requests 1. POLICY The Backfit Rule, Title 10 of the Code of Federal Regulations (10 CFR) 50.109, governs the requirements for backfitting of nuclear power plants. It requires that the Nuclear Regulatory Commission (NRC) justify each backfit with either a backfit analysis or a documented evaluation.

The term backfit is used in these procedures to denote modification of or addition to (1) systems, structures, components, or design of a facility;(2) the design approval or manufacturing license for a facility; (3) the procedures or organization required to design, construct, or operate a facility if the modification or addition results from a new or amended provision in the Commission rules or from the imposition of a regulatory staff position interpreting the Commission rules that is either new or different from a previously applicable staff position.

To be considered a backfit, these new or different positions must be taken after certain dates specified.

in 50.109(a)(1)(i-vii).

Furthermore, a backfit is plant-specific when it involves the imposition of a position that is unique to a particular plant.1 A modification or action proposed by a licensee is not deemed to be subject to the Backfit Rule.For new nuclear power reactors using the provisions of 10 CFR Part 52, applicable backfitting and issue finality provisions are also found in that part. For nuclear power reactor combined licenses, backfitting provisions are contained in both 10 CFR 50.109 and in Subpart C of Part 52.The Backfit Rule ordinarily does not apply in renewal of a nuclear power plant operating license under 10 CFR Part 54. However, if the NRC proposes to address safety issues outside the scope of Part 54 (e.g., time-limited aging analyses and aging of long-lived passive structures, systems, and components (SSCs)), then any actions necessary to address such out-of-scope safety issues are subject to the Backfit Rule. Once a renewed license is issued, the Backfit Rule applies, with the exception of the identification of newly-identified SSCs under 10 CFR 54.37(b).Section 50.54(f) of 10 CFR enables the staff to seek information necessary to determine the need to modify, suspend, or revoke a plant license. Section 50.54(f) further states"Except for information sought to verify licensee compliance with the current licensing basis for that facility, the NRC must prepare the reason or reasons for each information request prior to issuance to ensure that the burden to be imposed on respondents is justified in view of the potential safety significance of the issue to be addressed in the requested information." NRC Management Directive (MD) 8.4, "NRC Program for Management of Plant-Specific Backfitting of Nuclear Power Plants," establishes roles and responsibilities for NRC staff implementation of 10 CFR 50.109 and 10 CFR 50.54(f).Plant-specific backfitting involves positions unique to a particular plant, whereas generic backfitting involves the imposition of the same or similar positions on more thanone plant.An exception to this general principle is when the NRC conditions approval of the licensee-initiated modification or action upon a matter not technically relevant to the determination of the licensee's proposed modification or action.

Applicant's Consolidated Memorandum in Opposition to New York State's Motion for Summary Disposition of Contention NYS-35/36 and in Support of Its Cross-Motion for Summary Disposition (February 3, 2011)NYS-35/36:

ATTACHMENT 15 NUREG- 1850 FREQUENTLY ASKED QUESTIONS ON LICENSE'RENEWAL OF NUCLEAR POWER REACTORS Final Report Manuscript Completed:

February 2006 Date Published:

March 2006 Division of License Renewal Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 4.3.8 How do I get answers to my questions that fall outside the scope of the environmental review from the NRC?There are three different ways for members of the public to receive answers to questions that fall outside the scope of the environmental review: " Public meetings -Members of the public are invited to plant-specific public meetings (see response to Question 5.1.3), where NRC staff members are available to answer any questions related to NRC-regulated activities that members of the public may have, including those that are outside the scope of the environmental review. Public Meeting" NRC website -Answers to many questions that are outside the scope of the environmental review are also found on the NRC website, www.nrc.gov.

The NRC has also developed a number of Frequently Asked Questions documents as well as informational brochures and fact-sheets to address issues that are of concern to the public. These documents, brochures, and fact sheets are also located on the NRC website..NRC environmental project manager -For plant-specific questions that are outside the scope, members of the public could contact the environmental project manager assigned by the NRC for the license-renewal review for the specific plant. The phone number for each of the NRC environmental project managers is given on the NRC website, as well as in Federal Register notices and at the public meetings.

The NRC environmental project manager can either answer questions or direct callers to the appropriate person in the agency for responding to their questions that are outside the scope of the review.4.4 Severe Accident Mitigation Alternatives (SAMAs) Review An analysis of SAMAs is included as part of the environmental review of the application for license renewal if it had not been considered Severe accidents are earlier.for the facility.

A definition of SAMAs and an explanation of those that could result why they are included in the environmental review are included in this in substantial damage section. In addition, the process used to evaluate SAMAS and the types of changes that may occur in the plant as a result of the analysis are to the reactor core, also discussed.

whether or not there are serious off-site 4.4.1 What is a Severe Accident Mitigation Alternatives consequences.(SAMAs) review?The SAMAs review is an evaluation of alternatives to mitigate severe accidents.

Severe accidents are those that could result in substantial damage to the reactor core, whether or not there are serious off-site consequences.

The NRC staff reviews and evaluates SAMAs to ensure that changes that could improve severe accident safety performance are identified and evaluated.

Potential improvements could include hardware modifications, changes to procedures, and changes to the training program.March 2006 4-31 Final FAQs for License Renewal March 2006 4-31 Final FAQs for License Renewal In some cases, SAMAs may have already been evaluated by the A Severe A ccident NRC staff in a previous environmental impact statement (EIS),.M1itigation Alternatives supplement, or environmental assessment (EA) written for a facility review is an evaluation of before the applicant applied for license renewal. In such cases, the alternatives to mitigate evaluation does not have to be repeated for that particular facility, severe accidents.

according to NRC regulations in 10 CFR 51.53. However, if the NRC staff has not previously evaluated SAMAs for an applicant's plant in an EIS, a supplement, or an EA, the license renewal applicant is required to consider alternatives to mitigate severe accidents as part of the license renewal application.

4.4.2 Why are SAMAs considered part of the environmental review?The Commission's regulations (10 CFR Part 5 1) require that license renewal applicants consider alternatives to mitigate severe accidents if the NRC staff has not previously evaluated severe accident mitigation alternatives (SAMAs) for the applicant's plant in an environmental impact statement (EIS) or related supplement or in an environmental assessment.

This requirement results, in part, from a court decision partially reversing a Commission decision upholding an Atomic Safety and Licensing Board (ASLB) determination that it need not adjudicate a contention alleging that the NRC staff must consider, under the National Environmental Policy Act (NEPA), severe accident mitigation design alternatives in an environmental impact statement for an operating license (Limerick Ecology Action v. NRC, 869 F.2d 719 [3d Cir. 1989]).. At the time that this requirement was established, licensees were identifying individual plant vulnerabilities and considering cost-beneficial improvements, e.g., in the individual plant examination (IPE) and the individual plant examination for external events (IPEEE) programs.

The Commission believes that it is unlikely that any site-specific consideration of severe accident mitigation alternatives for license renewal would identify major plant design changes or modifications proving to be cost-beneficial for reducing severe accident frequency or consequences.

However, because these programs had not yet been completed at all plants, the Commission considered a general conclusion regarding severe accident mitigation to be premature, and required a site-specific consideration of SAMAs forlicense renewal for plants lacking previously evaluated SAMAs in an EIS, related supplement, or environmental assessment.

4.4.3 What is the process for the Severe Accident Mitigation Alternatives (SAMAs) review?The evaluation of SAMAs is a four-step process, as shown in Figure 4.1. The first step is to characterize overall plant risk and the leading contributors to the risk. This typically involves the extensive use of a plant-specific probabilistic safety assessment (PSA) study. The PSA identifies the different contributors of system failures and human errors that would be required for an accident to progress either to core damage or to containment failure.The second step is to identify potential improvements that could reduce the risk. Information from the PSA, such as dominant accident sequences, equipment failures, and operator actions is used to identify plant improvements that would have the greatest impact in reducing risk. Improvements identified in other NRC and industry studies, as well as SAMAs analysis for otherplants, are also considered in this process.Final FAQs for License Renewal 4-32 March. 2006 Final FAQs for License Renewal 4-32 March.2006

Step 1. Characterize Plant Risk. Use of a plant-probabilistic safety assessment (PSA) study to identify contributors of system failures and human errors required for an accident to progress to core damage or containment failure" Source,,, of Potential SAMAs: imp'novernepts identified by other SAMAs analyses NRC and-industry studies of potential plant improvements.

Potentid improvements identified in PE and IPEEE_lRevi",., ýt lio'.latest update to the plant PSA ,Reasons-0figina'ISAIVIAs-areRemoved from Further ,Conýide'ration-Ný3t ipfdi-ýhle toplant design Alreýý 1 , niý,lemented

_oiiniiý-,, to ,,ioth er-SAMA or was combined Not a significant safety benefit Step 3. Quantify Risk Reduction P otential and Implementation Costs -Deteirm " ine the risk red .uction potential from the proposýea rn'provernents from Step 2 and then estimatehow much these',proposed improvementsýwoulVcost.

Man y-of the SAMAs will'be removed from'fUrther", consideration at.this.point because theywould c6sl ,ý-than*.the' rnaxirný6ni attainable benefit (MAB) -the dollar.value of th, 'benefit-if the risk could be-redUced.to-zero.

SAMAs -,ting rnore'tha6 MAB cannot'Ibe costben6fiýiaýi.

Detailed-16:nefit'and cost estimates are made.for ,remainingSAIVIAs Step 4. Determine Whether theAmplementation of Improvements is Justified

-Conipaiýe the costs andý.benefits of the proposed improveiiients, determine if the improvement provides a reduction in total risk and whether the risl luý.Iiý,n is associated with aging effects during the ' L-ii,)d of extended operations, to deterruinelwheýý-i improvements are justified.

Express benefits in dollars and compare costs and'bei,,_ýt!i,, ý,! pr iý,)sed improvements to identif, ber, fi,-i.,l nýpiovements.

lfbýný fit -,f -ý ý,',YA is largerthan cost tn,ýýcoriý.i 1-i-j t cost-beneficial.

Figure 4.1. General ized.Process for Identifying and Evaluating Potential Severe Accident Mitigation Alternatives (SAMAs)The third step is to quantify the risk-reduction potential and the implementation cost for each of the improvements.

The risk reduction is typically estimated using A conservative analysis that generally overestimates the risk-reduction potential by assuming that the plant improvement is highly effective in eliminating the accident sequence that the impýovementis intended to address. Implementation costs are generally underestimated by neglecting certain cost factors, such as maintenance costs or sufveillance costs associated with the plant modification.

Overestimating the risk-reduction potential and under estimating the implementation costs in this step make it more likely that a potentially useful safety improvement would be retained for further consideration in the final step.The risk-reduction potentials and the implementation cost estimates are used in the final step, which is to determine whether implementation of any of the improvements is justified.

In determining whether the improvement is justified, the NRC staff looks at three factors: whether the improvement is cost-beneficial,.in other words, whether the estimated benefit is greater than the estimate of the March 2006 4-33 Final FAQs for License Renewal implementation cost of the SAMAs; whether the improvement provides a significant reduction in total risk, in other words, whether it eliminates a sequence or containment failure mode that contributes to a large fraction of plant risk; and whether the risk reduction is associated with aging effects during the period of extended operation, which would be an improvement implemented as part of the license renewal process.4.4.4 What is the outcome of the review?The Severe Accident Mitigation Alternatives analysis results in a list of plant improvements.

that meet the criteria of"* being cost beneficial

providing a significant reduction in total risk, and" being associated with aging effects during the period of extended operation.

The outcome of the Severe Accident Mitigation Alternatives (SAMAs) analysis is a list of plant improvements that meet the criteria of being cost-beneficial, provide a significant reduction in total risk, and are associated with aging effects during the period of extended operation.

In some cases, however, the review leads to a determination that there are no specific SAMA candidates that are cost-beneficial.

This may be the case where there is a low residual level of risk and where the applicant has, in fact, already implemented many plant improvements.

In other cases, a SAMA that is potentially cost-beneficial may not relate to adequately managing the effects of aging during the period of extended operation.

Such SAMAs need not be implemented as part of the license renewal pursuant to 10 CFR Part 54.4A4.5 Does the applicant have to implement any identified changes?The SAMAs analyses that have been perfbrmed to date have fbund SAMA4s that were cost beneficial, or at least possibly cost beneficial subject tofiurther analysis; in approximately half of the plants. None of the SAMAs identified related to managing the ejcits of aging during the period of extended operation.

The only changes that must be implemented by the applicant as part of the license renewal process are those that are identified as being cost-beneficial, that provide a significant reduction in total risk, and that are related to adequately managing the effects of aging during the*period of extended operation.

However, the Severe Accident Mitigation Alternatives (SAMAs) evaluation may identify some plant enhancements that appear to be cost-beneficial but that are not related to adequately managing the effects of aging during the period of extended operation.

Such enhancements are considered as current operating issues and are further evaluated as changes that might appropriately be made under the current operating license'rather than as a license renewal issue.4.4.6 Have any changes been implemented at a plant as a result of the Severe Accident Mitigation Alternatives (SAMAs) review?The SAMAs analyses that have been performed to date have found SAMAs that were cost-beneficial, or at least possibly cost-beneficial subject to further analysis,-

in approximately half of the plants. However, none of the SAMAs identified related to managing the effects of aging during the. period of extended operation.

Therefore, they did not Final FAQs for License Renewaf 4-34.March 2006 need to be implemented as part of license renewal, pursuant to the regulations in 10 CFR Part 54. In general, the cost-beneficial SAMAs were identified for further evaluation by the licensee under the current operating license. In several cases, the applicant has decided to implement the modifications even though they were not related to license renewal.4.5 Storage and Disposal of Spent Nuclear Fuel Although the storage and disposal of high-level waste are not within the scope of environmental issues pertaining to license renewal, questions about these topics are asked frequently during public meetings and other opportunities for public comment. In the interest of providing a full picture of the issues associated with nuclear power facilities, this section provides information about the Nuclear Waste Policy Act and the status of Yucca Mountain as a repository for spent nuclear fuel from commercial reactors and the storage of spent fuel at nuclear power facilities.

4.5.1 What is the Nuclear Waste Policy Act?The Nuclear Waste Policy Act establishes the Federal government's responsibility to provide a place for the permanent disposal of high-level radioactive waste and spent nuclear fuel and the generators'(commercial nuclear power facilities')

responsibility to.bear the costs of permanent disposal.

The Act authorizes and requires the U.S. Department of Energy (DOE) to locate and build a permanent repository and an interim storage facility and develop a transportation system to safely link nuclear plants to the repository and interim storage facility.

The Act was signed into law by President Reagan on January 7, 1983. The Act obligated DOE to begin disposal of high-level radioactive waste from commercial nuclear facilities by January 31, 1998. To date, an application for licensing these facilities has not been submitted to.the NRC.4.5.2 What is the status of Yucca Mountain?For over two decades, research has been conducted to determine whether a site near Yucca Mountain, Nevada, is suitable for safely isolating highly radioactive nuclear waste. Four major steps -site characterization, site approval, licensing review, and construction

-have to be completed before operation of the proposed high-level waste repository could occur. The first step, site characterization, including excavation of exploratory tunnels and testing of groundwater, has been completed.

The result of the characterization study was documented by the development of an environmental Yucca Mountain, Nevada impact statement (EIS). A draft EIS was March 2006 4-35 Final FAQs for License Renewal March 2006 4-35 Final FAQs for License Renewal Applicant's Consolidated Memorandum in Opposition to New York State's Motion for Summary Disposition of Contention NYS-35/36 and in Support of Its Cross-Motion for Summary Disposition (February 3, 2011)NYS-35/36:

ATTACHMENT 16 NUREG-1800, Rev. 1 Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants Manuscript Completed:

September 2005 Date Published:

September 2005 Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 4.1 IDENTIFICATION OF TIME-LIMITED AGING ANALYSES Review Responsibilities Primary- Branch responsible for the TLAA issues Secondary-Other branches responsible for engineering, as appropriate 4.1.1 Areas of Review This review plan section addresses the identification of time-limited aging analyses (TLAAs).The technical review of TLAAs is addressed in section 4.2 through 4.7. As explained in more detail below, the list of TLAAs are certain plant-specific safety analyses that are based on an explicitly assumed 40-year plant life (for example, aspects of the reactor vessel design).Pursuantto 10 CFR 54.21 (c)(1), a license renewal applicant is required to provide a list of TLAAs, as defined in 10 CFR 54.3. The area relating to the identification of TLAAs is reviewed.TLAAs may have developed since issuance of a plant's operating license. As indicated in 10 CFR 54.30, the adequacy of the plant's CLB, which includes TLAAs, is not an area within the scope of the license renewal review. Any question regarding the adequacy of the CLB must be addressed under the backfit rule (10 CFR 50.109) and is separate from the license renewal process.In addition, pursuant to 10 CFR 54.21 (c)(2), an applicant must provide a list of plant-specific exemptions granted under 10 CFR 50.12 that are based on TLAAs. However, the initial license renewal applicants have found no such exemptions for their plants.It-is an applicant's option to include more analyses than those required by 10 CFR 54.21 (c)(1).The staff should focus its review to confirm that the applicant did not omit any TLAAs, as defined in 10 CFR 54.3.Pursuant to 10 CFR 54.21(d), each application is to include a FSAR Supplement summary description for each TLAA that is identified in accordance with 10 CFR 54.3.4.1.2 Acceptance Criteria The acceptance criteria for the areas of review described in Subsection 4.1.1 of this review plan section delineate acceptable methods for meeting the requirements of the NRC's regulations in 10 CFR 54.21 (c)(1). For the applicant's list of exemptions to be acceptable, the staff should have reasonable assurance that there has been no omission of TLAAs from that list.Pursuant to 10 CFR 54.3, TLAAs are those licensee calculations and analyses that: 1. Involve systems, structures, and components within the scope of license renewal, as delineated in 10 CFR 54.4(a);2. Consider the effects of aging;3. Involve time-limited assumptions defined by the current operating term, for example, 40 years;4. Were determined to be relevant by the licensee in making a safety determination; September 2005 4.1-1 NUREG-1800, Rev. 1 4.7 OTHER PLANT-SPECIFIC TIME-LIMITED AGING ANALYSES Review Responsibilities Primary- Branch responsible for the TLAA issues Secondary

-Other branches responsible for systems, as appropriate 4.7.1 Areas of Review There are certain plant-specific safety analyses that may have been based on an explicitly assumed 40-year plant life (for example, aspects of the reactor vessel design) and may, therefore, be time-limited aging analyses (TLAAs.) Pursuant to 10 CFR 54.21 (c), a license renewal applicant is required to evaluate TLAAs. The definition of TLAAs is provided in 10 CFR 54.3 and in Section 4.1 of this standard review plan.TLAAs may have evolved since issuance of a plant's operating license, and are plant-specific.

As indicated in 10 CFR 54.30, the adequacy of the plant's CLB, which includes TLAAs, is not an area within the scope of the license renewal review. Any question regarding the adequacy of the CLB must be addressed under the backfit rule (10 CFR 50.109) and is separate from the license renewal process.License renewal reviews focus on the period of extended operation.

Pursuant to 10 CFR 54.30, if the reviews required by 10 CFR 54.21(a) or (c) show that there is not reasonable assurance during the current license term that licensed activities will be conducted in accordance with the CLB, the licensee is required to take measures under its current license to ensure that the intended function of those systems, structures, or components will be maintained in accordance with the CLB throughout the term of the current license. The adequacy of the measures for the term of the current license is not within the scope of the license renewal review.Pursuant to 10 CFR 54.21(c), an applicant must provide a listing of TLAAs and plant-specific exemptions that are based. on TLAAs. The staff reviews the applicant's identification of TLAAs and exemptions separately, following the guidance in Section 4.1 of this standard review plan.Based on lessons learned in the review of the initial license renewal applications, the staff has developed review procedures for the evaluation of certain TLAAs. If an applicant identifies these TLAAs as applicable to its plant, the staff reviews them separately, following the guidance in Sections 4.2 through 4.6. The reviewer reviews other TLAAs that are identified by the applicant, following.the generic guidance in this section. For particular systems, the reviewers from branches responsible for those systems may be requested to assist in the review, as appropriate.

The following areas relating to a TLAA are reviewed: 4.7.1.1 Time-Limited Aging Analysis The evaluation of the TLAA for the period of extended operation is reviewed.4.7.1.2 FSAR Supplement The FSAR supplement summarizing the evaluation of the TLAA for the period of extended operation in accordance with 10 CFR 54.21(d) is reviewed.Septembet 2005 4.7-1 NUREG-1800, Rev. 1

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