ML103190494

NextEra Energy Seabrook, Llc’S Answer Opposing the Petition to Intervene and Request for Hearing of Friends of the Coast and the New England Coalition

ML103190494
Person / Time
Site:	Seabrook
Issue date:	11/15/2010
From:	Hamrick S NextEra Energy Seabrook
To:	Atomic Safety and Licensing Board Panel
SECY RAS
References
RAS 19111, 50-443-LR, ASLBP 10-906-02-LR-BD01
Download: ML103190494 (151)

Text

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of

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NextEra Energy Seabrook, LLC

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Docket No.

50-443-LR

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(Seabrook Station)

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ASLBP No. 0-906-02-LR (Operating License Renewal)

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NextEra Energy Seabrook, LLCs Answer Opposing The Petition to Intervene and Request for Hearing of Friends of the Coast and the New England Coalition November 15, 2010

i TABLE OF CONTENTS Page I.

INTRODUCTION................................................................................................1 II.

BACKGROUND..................................................................................................2 III.

THE PETITION IS NOT TIMELY......................................................................3 IV.

PETITIONERS HAVE NOT DEMONSTRATED STANDING OR PROFFERED AN ADMISSIBLE CONTENTION.............................................4 A.

Petitioners Have Not Demonstrated Standing......................................................4 B.

Legal Standards for Contention Admissibility.....................................................6

1. Petitioner Must Specifically State the Issue of Law or Fact to Be Raised...........8

2. Petitioner Must Explain the Basis for the Contention..........................................8

3. Contentions Must Be Within the Scope of the Proceeding...................................9

4.

Contentions Must Raise a Material Issue.............................................................10

5. Contentions Must Be Supported by Adequate Factual Information or Expert Opinion.................................................................................................................10

6. Contentions Must Raise a Genuine Dispute of Material Law or Fact..................12 C.

Overview of Reactor License Renewal.................................................................14

1. Scope of Safety Issues in License Renewal Proceedings.....................................15

2. Aging Management and the GALL Report..........................................................17

3. Severe Accident Mitigation Alternatives in License Renewal Proceedings.........19 D.

Petitioners Have Not Proffered An Admissible Contention.................................24

1. Contention 1 - Inaccessible Cables Is Inadmissible........................................25

a. Contention 1 Lacks Adequate Factual or Expert Support............................26

b. Contention 1 Fails to Raise a Genuine Dispute With the LRA....................28

c. Contention 1 Fails to Raise A Material Issue................................................35

d. Contention 1 Raises Issues Beyond the Scope of License Renewal..............39

e. Contention 1 is Moot.......................................................................................41

2.

Contention 2 - Transformers Is Inadmissible....................................................43

a. Transformers are Active Components Not Subject to Aging Management Review........................................................................................43

b. Contention 2 Lacks the Requisite Factual or Expert Support...........................46

ii

3. Contention 3 - Buried, Below Ground, or Hard-to-Access Piping Is Inadmissible.....................................................................................................48

a. Contention 3 Is Beyond the Scope of a License Renewal Proceeding...........49

b. Contention 3 Lacks the Requisite Factual or Expert Support........................53

c. Contention 3 Fails to Show a Genuine Dispute on a Material Issue................58

d. Contention 3 is Moot......................................................................................61

4. Contention 4 - Severe Accident Cost is Inadmissible.....................................65

a. Contention 4A is Inadmissible.......................................................................70

b. Contention 4B is Inadmissible........................................................................73

i. No Mitigation Analysis is Required for Spent Fuel Pool Accidents........................................................................73 ii. Petitioners Challenge to the MAAP Code is Inadmissible......................75

c. Contention 4C is Inadmissible........................................................................77

d. Contention 4D is Inadmissible.........................................................................78

e. Contention 4E is Inadmissible........................................................................89

i. Decontamination and Cleanup Costs.........................................................89 ii. Health Costs..............................................................................................94 iii. Myriad Other Economic Costs.................................................................99

f. Contention 4F is Inadmissible.......................................................................100

5. Petitioners CLB Challenges are Not Material....................................................103 V.

SELECTION OF HEARING PROCEDURES.....................................................105 VI.

CONCLUSION.....................................................................................................106

November 15, 2010 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of

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NextEra Energy Seabrook, LLC

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Docket No.

50-443-LR

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(Seabrook Station)

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ASLBP No. 0-906-02-LR (Operating License Renewal)

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NextEra Energy Seabrook, LLCs Answer Opposing The Petition to Intervene and Request for Hearing of Friends of the Coast and the New England Coalition I.

INTRODUCTION NextEra Energy Seabrook, LLC (NextEra or Applicant) hereby submits this answer (Answer) opposing the Request for a Public Hearing and Petition to Intervene (Petition) filed by Friends of the Coast and the New England Coalition (collectively the Petitioners) in this proceeding on October 21, 2010. The Petition should be denied because Petitioners have failed to propose an admissible contention.

The Nuclear Regulatory Commissions (NRC or Commission) regulations and case law clearly set forth the requirements that a petitioner must satisfy in order to propose an admissible contention. As this Answer describes more fully below, the Commissions current pleading standards were designed to raise the threshold for the admission of contentions. The purpose of these intentionally strict admissibility requirements is to ensure that hearings, if required, would focus on concrete issues that

2 are relevant to the proceeding and that are supported by some factual and legal foundation. Each of Petitioners Contentions fails to reach the required threshold, falling short of any number of the applicable pleading standards. Accordingly, the Board should reject all of Petitioners Contentions and deny their request for hearing.

II.

BACKGROUND NextEra submitted its application requesting renewal of Operating License NPF-86 for the Seabrook Station (the Application or LRA) by [[letter::L-2010-101, Comment (1) of Larry Nicholson, on Behalf of Florida Power & Light Co., on Proposed NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants; Branch Technical Position (BTP) 7-19, Guidance for .|letter dated May 25, 2010]]. The NRC Staff conducted a sufficiency review, found the Application acceptable for docketing, and published notice of an opportunity for hearing in the Federal Register.

Notice of Acceptance for Docketing of the Application and Notice of Opportunity for Hearing Regarding Renewal of Facility Operating License No. NPF-86 for an Additional 20-Year Period; Nextera Energy Seabrook, LLC; Seabrook Station, Unit 1, 75 Fed. Reg.

42,462 (July 21, 2010) (Hearing Notice). The Hearing Notice permitted any person whose interest may be affected to file a request for hearing and petition for leave to intervene within 60 days. 75 Fed. Reg. at 42,463. On September 17, 2010, the Secretary of the Commission granted the State of New Hampshire, Beyond Nuclear, Friends of the Coast and the New England Coalition an extension of time to file intervention petitions, until October 20, 2010. Petitioners filed their Petition on October 21, 2010.

The Hearing Notice directs that any petition shall set forth with particularity the interest of the petitioner and how that interest may be affected, and must also set forth the specific contentions sought to be litigated. 75 Fed. Reg. at 42,463. It also states:

Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the requestor/petitioner shall provide a brief explanation of the bases of each contention and a concise statement of the

3 alleged facts or the expert opinion that supports the contention on which the requestor/petitioner intends to rely in proving the contention at the hearing.

The requestor/petitioner must also provide references to those specific sources and documents of which the requestor/petitioner is aware and on which the requestor/petitioner intends to rely to establish those facts or expert opinion. The requestor/petitioner must provide sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact.

Contentions shall be limited to matters within the scope of the action under consideration. The contention must be one that, if proven, would entitle the requestor/petitioner to relief. A requestor/petitioner who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party.

Id.

III.

THE PETITION IS NOT TIMELY Having already been granted a 30-day extension of time to file their Petition, Petitioners filed it a day lateon October 21, 2010. Petitioners later sought a retroactive request for extension of time (Request), in which they explained that they waited until after 10 p.m. on October 20 to begin the process of uploading thirteen separate documents to the NRCs E-Filing System and encountered technical difficulties that could not be resolved in a timely fashion. Request at 1-2. Petitioners served courtesy copies of their pleadings by e-mail shortly after midnight on October 21, but did not serve their documents via the E-Filing system until much later that day. The Hearing Notice states that [t]o be timely, an electronic filing must be submitted to the E-Filing system no later than 11:59 p.m. Eastern Time on the due date. 75 Fed. Reg. 42,464. See also 10 C.F.R. § 2.306(b)(4)(2). The Hearing Notice also explains that an NRC help desk is available to assist participants who encounter technical difficulties, but is only available until 8 p.m. Id.

4 By waiting until 10 p.m. on the day the Petition was due to be filed, Petitioners assumed the risk that they may encounter technical difficulties that would require the assistance of the NRCs help desk. Moreover, the Hearing Notice explains that [n]on-timely filings will not be entertained absent a determination by the presiding officer that the petition or request should be granted or the contentions should be admitted, based on a balancing of the factors specified in 10 C.F.R. § 2.309(c)(1)(i)-(viii). Id. Petitioners have not addressed these factors, either in their Petition or in their late-filed Request.

Accordingly, Petitioners have failed to show good cause (§ 2.309(c)(1)(i)) for their failure to file on time or that their non-timely filing should be entertained. The Commission generally extend[s] some latitude to pro se litigants, but they still are expected to comply with [its] procedural rules. South Carolina Electric and Gas Co.

(Summer Nuclear Station, Units 2 and 3), CLI-10-01, 71 NRC ___, (2010) (slip op. at 5).

IV.

PETITIONERS HAVE NOT DEMONSTRATED STANDING OR PROFFERED AN ADMISSIBLE CONTENTION To be admitted as parties to this proceeding, Petitioners must demonstrate standing and submit at least one admissible contention. 10 C.F.R. § 2.309(a). As discussed below, Petitioners have not demonstrated standing or proposed any admissible contentions. Therefore, the Petition should be denied.

A.

Petitioners Have Not Demonstrated Standing In order to obtain a hearing before the NRC, a petitioner must demonstrate its standing and file at least one admissible contention. See 10 C.F.R. § 2.309(a), (d). The burden of demonstrating standing rests with the petitioner. Commonwealth Edison Co.

(Zion Nuclear Power Station, Units 1 and 2), CLI-00-5, 51 NRC 90, 98 (2000). To establish standing, the petitioner must plead the nature of the [petitioners] right under

5 the Act to be made a party to the proceeding[,]... the nature and extent of [the petitioners] property, financial or other interest in the proceeding; and [t]he possible effect of any decision or order that may be issued in the proceeding on the [petitioners]

interest. 10 C.F.R. § 2.309(d)(1).

Petitioners assert representational standing. An organization that seeks representational standing must demonstrate how at least one of its members may be affected by the licensing action (as a result of the members activities on or near the site),

must identify that member by name and address, and must show (preferably by affidavit) that the organization is authorized to request a hearing on behalf of that member.

Vermont Yankee Nuclear Power Corp. & Amergen Vermont, LLC (Vermont Yankee Nuclear Power Station), CLI-00-20, 52 NRC 151, 163 (2000) (citing GPU Nuclear, Inc.

(Oyster Creek Nuclear Generating Station), CLI-00-6, 51 NRC 193, 202 (May 3, 2000)).

Petitioners have presented several declarations of individuals who purport to:

(a) live in close proximity to Seabrook; and (b) authorize Petitioners to request a hearing on their behalf. However, not one of the member declarations is signed. Several of the declarants names are signed using a cursive script font, which is clearly not a signature.1 The Teed Declaration does not even make this effort and simply leaves a blank signature line. Without valid handwritten or electronic signatures, this Board cannot determine whether Petitioners are, in fact, authorized to represent the individuals identified in the various member declarations.2 By submitting defective member 1

The proffered expert declaration of Paul Blanch also utilizes this technique.

2 NRC regulations provide for an alternate method of electronic signature in 10 C.F.R. § 2.304(d) that must be used for any signature other than that of the individual performing the E-Filing function, including an affiant. Petitioners have failed to comply with this requirement.

6 declarations, Petitioners have failed to meet their burden to demonstrate representational standing.

B.

Legal Standards for Contention Admissibility The Commissions contention admissibility rules are strict by design.

Dominion Nuclear Connecticut, Inc. (Millstone Nuclear Power Station, Units 2 & 3),

CLI-01-24, 54 NRC 349, 358 (2001) (citing Duke Energy Corp. (Oconee Nuclear Station, Units 1, 2, and 3), CLI-99-11, 49 NRC 328, 334 (1999)). While federal courts permit considerably less-detailed notice pleading, the Commission requires far more to plead a contention. Private Fuel Storage, L.L.C. (Independent Spent Fuel Storage Installation), LBP-01-39, 54 NRC 497, 505 (2001); see also Fansteel, Inc. (Muskogee, Oklahoma Site) CLI-03-13, 58 NRC 195, 203 (2003). 10 C.F.R. § 2.714 (now § 2.309) was amended in 1989 to raise the threshold for the admission of contentions. Rules of Practice for Domestic Licensing Proceedings - Procedural Changes in the Hearing Process, 54 Fed. Reg. 33,168 (Aug. 11, 1989) (Final Rule). These rules were toughened... because in prior years licensing boards had admitted and litigated numerous contentions that appeared to be based on little more than speculation.

Millstone, CLI-01-24, 54 NRC at 358. Under the NRCs Rules of Practice, a protestant does not become entitled to an evidentiary hearing merely on request, or on a bald or conclusory allegation that such a dispute exists. The protestant must make a minimal showing that material facts are in dispute, thereby demonstrating that an inquiry in depth is appropriate. 54 Fed. Reg. at 33,171 (quoting Conn. Bankers Assn v. Bd. of Governors, 627 F.2d 245, 251 (D.C. Cir. 1980)).

7 Accordingly, a petition must set forth with particularity the contentions sought be raised. 10 C.F.R. § 2.309(f)(1). Petitioners must provide a clear statement as to the basis for the contentions and [submit] supporting information and references to specific documents and sources that establish the validity of the contention.

USEC, Inc.

(American Centrifuge Plant), CLI-06-9, 63 NRC 433, 437 (2006) (citing Arizona Public Service Co. (Palo Verde Nuclear Generating Station, Units 1, 2, and 3), CLI-91-12, 34 NRC 149, 155-56 (1991)). Specifically, for each contention, the petition must:

(i) Provide a specific statement of the issue of law or fact to be raised or controverted; (ii) Provide a brief explanation of the basis for the contention; (iii) Demonstrate that the issue raised in the contention is within the scope of the proceeding; (iv) Demonstrate that the issue raised in the contention is material to the findings the NRC must make to support the action that is involved in the proceeding; (v) Provide a concise statement of the alleged facts or expert opinions which support the requestors/petitioners position on the issue and on which the petitioner intends to rely at hearing, together with references to the specific sources and documents on which the requestor/petitioner intends to rely to support its position on the issue; and (vi) [P]rovide sufficient information to show that a genuine dispute exists with the applicant/licensee on a material issue of law or fact.

10 C.F.R. § 2.309(f)(1). Contentions that do not satisfy each of these six requirements must be rejected. Progress Energy Carolinas, Inc. (Shearon Harris Nuclear Power Plant, Units 2 and 3), CLI-09-8, 69 NRC 317, 324 (2009).

The petitioner bears the burden of proffering contentions that meet the NRCs pleading requirements. See Baltimore Gas & Electric Co. (Calvert Cliffs Nuclear Power Plant, Units 1 and 2), CLI-98-14, 48 NRC 39, 41 (1998). Licensing boards are not to overlook deficiencies in contentions or to assume the existence of missing information.

Palo Verde, CLI-91-12, 34 NRC at 155. In other words, [a] contentions proponent, not

8 the licensing board, is responsible for formulating the contention and providing the necessary information to satisfy the basis requirement for the admission of contentions.

Statement of Policy on Conduct of Adjudicatory Proceedings, CLI-98-12, 48 NRC 18, 22 (1998) (1998 Policy Statement). The requirements are discussed in detail below.

1.

Petitioner Must Specifically State the Issue of Law or Fact to Be Raised Each contention must provide a specific statement of the issue of law or fact to be raised or controverted. 10 C.F.R. § 2.309(f)(1)(i). To be admissible, a contention must explain, with specificity, particular safety or legal reasons requiring rejection of the contested [application]. Millstone, CLI-01-24, 54 NRC at 359-60. Moreover, the Commission has explained that Petitioners must articulate at the outset the specific issues they wish to litigate as a prerequisite to gaining formal admission as parties.

Oconee, CLI-99-11, 49 NRC at 338.

2.

Petitioner Must Explain the Basis for the Contention In addition, petitioners must provide a brief explanation of the basis for the contention. 10 C.F.R. § 2.309(f)(1)(ii). A petitioner must provide the licensing board with sufficient foundation to warrant further exploration. Public Service Co. of New Hampshire (Seabrook Station, Units 1 & 2), ALAB-942, 32 NRC 395, 428 (1990)

(footnote omitted). In other words, a petitioner must provide some sort of minimal basis indicating the potential validity of the contention. 54 Fed. Reg. at 33,170. While licensing boards generally admit contentions for litigation rather than bases, the Commission has recognized that [t]he reach of a contention necessarily hinges upon its terms coupled with its stated bases. Entergy Nuclear Generation Co. & Entergy Nuclear Operations, Inc., (Pilgrim Nuclear Power Station) CLI-10-11, 71 NRC __, __

9 (slip op. at 28) (2010) (emphasis in original) (citing Public Service Co. of New Hampshire (Seabrook Station, Units 1 and 2), ALAB-899, 28 NRC 93, 97 (1988), affd sub nom. Mass. v. NRC, 924 F.2d 311 (D.C. Cir.), cert. denied, 502 U.S. 899 (1991)).

Therefore, the lack of an adequate basis is sufficient grounds for rejecting a proposed contention.

3.

Contentions Must Be Within the Scope of the Proceeding Petitioners must also demonstrate that the issue raised in the contention is within the scope of the proceeding. 10 C.F.R. § 2.309(f)(1)(iii). The scope of this proceeding for which this licensing board has been delegated jurisdiction was set forth in the Commissions Hearing Notice. See Duke Power Co. (Catawba Nuclear Station, Units 1 and 2), ALAB-825, 22 NRC 785, 790-91 (1985). The Hearing Notice explained that the Licensing Board would consider NextEras Application for a renewed operating license under Part 54 for Seabrook. 75 Fed. Reg. at 42,462. Licensing boards are delegates of the Commission and so may exercise only those powers which the Commission has given [them]. Public Service Co. of Indiana (Marble Hill Nuclear Generating Station, Units 1 and 2), ALAB-316, 3 NRC 167, 170 (1976) (footnote omitted); accord Portland General Electric Co. (Trojan Nuclear Plant), ALAB-534, 9 NRC 287, 289-90 n.6 (1979).

Any contention that falls outside the specified scope of this proceeding is inadmissible.

Any contention that challenges an NRC rule is outside the scope of the proceeding because no rule or regulation of the Commission... is subject to attack... in any adjudicatory proceeding. See 10 C.F.R. § 2.335(a); see also Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. (Vermont Yankee Nuclear Power Station); Entergy Nuclear Generation Company & Entergy Nuclear

10 Operations, Inc. (Pilgrim Nuclear Power Station), CLI-07-3, 65 NRC 13, 18 n.15 (2007).

Petitioners may not demand an adjudicatory hearing to attack generic NRC requirements or regulations, or to express generalized grievances about NRC policies. Oconee, CLI-99-11, 49 NRC at 334. Contentions seeking to impose requirements in addition to those contained in Commission regulations impermissibly challenge those regulations.

Long Island Lighting Co. (Shoreham Nuclear Power Station, Unit 1), CLI-87-12, 26 NRC 383, 395 (1987); see also Metropolitan Edison Co. (Three Mile Island Nuclear Station, Unit No. 1), LBP-83-76, 18 NRC 1266, 1273 (1983) (When a Commission regulation permits the use of a particular analysis or technique, a contention asserting that a different analysis or technique should be used is an impermissible challenge to the regulation).

4.

Contentions Must Raise a Material Issue Petitioners must further demonstrate that the issue raised in the contention is material to the findings the NRC must make to support the action that is involved in the proceeding.3 10 C.F.R. § 2.309(f)(1)(iv). Admissible contentions must explain, with specificity, particular safety or legal reasons requiring rejection of the contested

[application]. Millstone, CLI-01-24, 54 NRC at 359-60. The Commission has defined a material issue as one where resolution of the dispute would make a difference in the outcome of the licensing proceeding. 54 Fed. Reg. at 33,172 (emphasis added).

5.

Contentions Must Be Supported by Adequate Factual Information or Expert Opinion Each contention must also [p]rovide a concise statement of the alleged facts or expert opinions which support [the petitioners] position on the issue and on which [the petitioner] intends to rely at hearing, together with references to the specific sources and 3

The standards defining the findings that the NRC must make to support issuance of a renewed license in this proceeding are set forth in 10 C.F.R. § 54.29.

11 documents on which [the petitioner] intends to rely to support its position in the issue.

10 C.F.R. § 2.309 (f)(1)(v). The petitioner bears the burden of coming forward with a sufficient factual basis indicating that a further inquiry is appropriate. Yankee Atomic Electric Co. (Yankee Nuclear Power Station), CLI-96-7, 43 NRC 235, 249 (1996) (citing Final Rule, 54 Fed. Reg. at 33,171 (requiring some factual basis for the contention)).

Under this standard, a petitioner is obligated to provide the [technical] analyses and expert opinion or other information showing why its bases support its contention.

Georgia Institute of Technology (Georgia Tech Research Reactor, Atlanta, Georgia),

LBP-95-6, 41 NRC 281, 305, vacated in part and remanded on other grounds, CLI-95-10, 42 NRC 1, affd in part, CLI-95-12, 42 NRC 111 (1995). Where a petitioner has failed to do so, the [Licensing] Board may not make factual inferences on [the]

petitioners behalf. Id. (citing Palo Verde, CLI-91-12, 34 NRC at 149). See also Private Fuel Storage, L.L.C. (Independent Spent Fuel Storage Installation), LBP-98-7, 47 NRC 142, 180 (1998) (a bald assertion that a matter ought to be considered or that a factual dispute exists... is not sufficient; rather, a petitioner must provide documents or other factual information or expert opinion to support a contentions proffered bases)

(citations omitted). A mere reference to documents does not provide an adequate basis for a contention.

Baltimore Gas & Electric Co. (Calvert Cliffs Nuclear Power Plant, Units 1 and 2), CLI-98-25, 48 NRC 325, 348 (1998). A petitioners failure to present the factual information or expert opinions necessary to support its contention adequately requires that the contention be rejected. Yankee, CLI-96-7, 43 NRC at 262; Palo Verde, CLI-91-12, 34 NRC at 155.

12 The Commission has made clear that conclusory statements, even when provided by an expert, are insufficient to demonstrate that further inquiry is appropriate. USEC (American Centrifuge Plant), CLI-06-10, 63 NRC 451, 472 (2006) ([A]n expert opinion that merely states a conclusion (e.g., the application is deficient, inadequate, or wrong) without providing a reasoned basis or explanation for that conclusion is inadequate because it deprives the Board of the ability to make the necessary, reflective assessment of the opinion. (footnote omitted)).

This requirement must be met at the outset. A contention is not to be admitted where an intervenor has no facts to support its position and where the intervenor contemplates using discovery or cross-examination as a fishing expedition which might produce relevant supporting facts. 54 Fed. Reg. at 33,171. The Rules of Practice bar contentions where petitioners have what amounts only to generalized suspicions, hoping to substantiate them later, or simply a desire for more time and more information in order to identify a genuine material dispute for litigation. Duke Energy Corp. (McGuire Nuclear Station, Units 1 and 2; Catawba Nuclear Station, Units 1 and 2), CLI-03-17, 58 NRC 419, 424 (2003).

6.

Contentions Must Raise a Genuine Dispute of Material Law or Fact Finally, each contention must provide sufficient information to show that a genuine dispute exists with the applicant... on a material issue of law or fact.

10 C.F.R. § 2.309 (f)(1)(vi). The NRCs pleading standards require a petitioner to read the pertinent portions of the combined license application and supporting documents, including the safety information required by 10 CFR 54.21 and the Environmental Report (ER), state the applicants position and the petitioners opposing view, and explain why

13 it has a disagreement with the applicant. 54 Fed. Reg. at 33,171; Millstone, CLI-01-24, 54 NRC at 358. Contentions must be based on documents or other information available at the time the petition is filed. 10 C.F.R. § 2.309(f)(2). Indeed, a petitioner has an ironclad obligation to examine the publicly available documentary material pertaining to the facility in question with sufficient care to enable the petitioner to uncover any information that could serve as the foundation for a specific contention. Neither Section 189a of the Atomic Energy Act nor [the corresponding Commission regulation] permits the filing of a vague, unparticularized contention, followed by an endeavor to flesh it out through discovery against the applicant or Staff.

54 Fed. Reg. at 33,170 (quoting Duke Power Co. (Catawba Nuclear Station, Units 1 & 2),

ALAB-687, 16 NRC 460, 468 (1982), vacated in part on other grounds, CLI-83-19, 17 NRC 1041 (1983)). The obligation to make specific reference to relevant facility documentation applies with special force to the application and a contention should be rejected if it inaccurately describes an applicants proposed actions or ignores or misstates the content of the licensing documents. See, e.g., Carolina Power & Light Co.

(Shearon Harris Nuclear Power Plant, Units 1 and 2), LBP-82-119A, 16 NRC 2069, 2076 (1982); Duke Power Co. (Catawba Nuclear Station, Units 1 and 2), LBP-82-107A, 16 NRC 1791, 1804 (1982); Philadelphia Electric Co. (Limerick Generating Station, Units 1 and 2), LBP-82-43A, 15 NRC 1423, 1504-05 (1982).

If the petitioner does not believe that a licensing request and supporting documentation addresses a relevant issue, the petitioner is to explain why the application is deficient. 54 Fed. Reg. at 33,170; Palo Verde, CLI-91-12, 34 NRC at 156. A contention that does not directly controvert a position taken by the applicant in the license renewal application is subject to dismissal. See Texas Utilities Electric Co. (Comanche

14 Peak Steam Electric Station, Unit 2), LBP-92-37, 36 NRC 370, 384 (1992). An allegation that some aspect of a license renewal application is inadequate does not give rise to a genuine dispute unless it is supported by facts and a reasoned statement of why the application is unacceptable in some material respect. Florida Power & Light Co.

(Turkey Point Nuclear Generating Plant, Units 3 and 4), LBP-90-16, 31 NRC 509, 521 &

n.12 (1990).

As set forth below, none of Petitioners Contentions complies with the Commissions standards.

C.

Overview of Reactor License Renewal As discussed above, 10 C.F.R. § 2.309(f)(1)(iii) requires the issue raised in a contention to be within the scope of the proceeding. The Commission has stated that

[a]djudicatory hearings in individual license renewal proceedings will share the same scope of issues as our NRC Staff review, for our hearing process (like [the] Staffs review) necessarily examines only the questions our safety rules make pertinent.

Florida Power & Light Co. (Turkey Point Nuclear Generating Plant, Units 3 and 4), CLI-01-17, 54 NRC 3, 10 (2001). See also Final Rule, Nuclear Power Plant License Renewal; Revisions, 60 Fed. Reg. 22,461, 22,482 n.2 (May 8, 1995) (1995 Final Rule). Under 10 C.F.R. Part 54, the NRC Staff conducts a technical review of the LRA to assure that public health and safety requirements are satisfied. Under 10 C.F.R. Part 51, the NRC Staff completes an environmental review for license renewal, focusing upon the potential impacts of an additional 20 years of nuclear power plant operation. Admissible contentions must address these areas of the NRC Staffs review, which are further described below.

15

1.

Scope of Safety Issues in License Renewal Proceedings Part 54 limits the review of LRAs to matters relevant to the extended period of operation requested by the applicant, which are not reviewed on a continuing basis under existing NRC inspection and oversight processes. The safety review is limited to the plant systems, structures, and components (as defined in 10 C.F.R. § 54.4) that require an aging management review (AMR) for the period of extended operation or are subject to an evaluation of time-limited aging analyses (TLAA). See 10 C.F.R. §§ 54.21(a) and (c), 54.29, and 54.30. Aging management programs (AMPs) are relied upon to demonstrate reasonable assurance that the effects of aging will be adequately managed during the period of extended operation, as required by 10 C.F.R. § 54.21(a)(3). The Hearing Notice describes the scope of the safety portion of this proceeding by describing the findings the NRC must make prior to issuance of a renewed license:

In accordance with 10 CFR 54.29, the NRC may issue a renewed license on the basis of its review if it finds that actions have been identified and have been or will be taken with respect to: (1) Managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified as requiring aging management review, and (2) time-limited aging analyses that have been identified as requiring review, such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the current licensing basis (CLB) and that any changes made to the plants CLB will comply with the Act and the Commissions regulations.

Hearing Notice, 75 Fed. Reg. at 42,463.

The NRCs license renewal regulations reflect a distinction between aging management issues and ongoing regulatory issues (e.g., security and emergency planning). With the exception of aging management issues, the NRC considers its

16 ongoing regulatory process to be adequate to ensure that the CLB of operating plants provides and maintains an acceptable level of safety. See 10 C.F.R. § 54.30; see also Final Rule, Nuclear Power Plant License Renewal, 56 Fed. Reg. 64,943, 64,946 (Dec. 13, 1991) (1991 Final Rule). As the Commission has explained:

[CLB is] a term of art comprehending the various Commission requirements applicable to a specific plant that are in effect at the time of the license renewal application

.... The [CLB] represents an evolving set of requirements and commitments for a specific plant that are modified as necessary over the life of a plant to ensure continuation of an adequate level of safety. 60 Fed. Reg. at 22,473. It is effectively addressed and maintained by ongoing agency oversight, review, and enforcement.

Turkey Point, CLI-01-17, 54 NRC at 9.

Because a full reassessment of these issues at the license renewal stage would be both unnecessary and wasteful, the NRCs license renewal review focuses upon those potential detrimental effects of aging that are not routinely addressed by ongoing regulatory oversight programs. Id. at 7. Thus, license renewal reviews are not intended to duplicate the Commissions ongoing review of operating reactors. Id. (quoting 1991 Final Rule, 56 Fed. Reg.at 64,946.

Accordingly, the potential detrimental effects of aging essentially defines the scope of... license renewal proceedings. Dominion Nuclear Connecticut, Inc.

(Millstone Nuclear Power Station, Units 2 and 3), CLI-04-36, 60 NRC 631, 637 (2004).

Accordingly, the Commission has specifically limited its license renewal safety review to the matters specified in 10 C.F.R. §§ 54.21 and 54.29(a)(2), which focus on the management of aging of certain systems, structures and components, and the review of TLAAs. See Turkey Point, CLI-01-17, 54 NRC at 7-8; Duke Energy Corp. (McGuire

17 Nuclear Station, Units 1 and 2), CLI-02-26, 56 NRC 358, 363 (2002). Specifically, applicants must demonstrate how their programs will be effective in managing the effects of aging during the proposed period of extended operation. Turkey Point, CLI-01-17, 54 NRC at 8 (quoting 1995 Final Rule, 60 Fed. Reg. at 22,462).

2.

Aging Management and the GALL Report The GALL Report contains the NRC Staffs generic evaluation of existing programs and documents the technical bases for determining the adequacy of those programs, with or without modification, in order to effectively manage the effects of aging during the period of extended plant operation. It is based on a systematic compilation of plant aging information and the evaluation of program attributes for managing the effects of aging on systems, structures and components for license renewal.

GALL Rev. 1 (NUREG 1801) at 1-3. The GALL Report explains that many existing programs are adequate to manage the aging effects for particular structures or components for license renewal without change. Id. at 1. For other programs, the GALL Report recommends augmentation for license renewal. Id. at 4. The GALL Report was developed at the Commissions direction with considerable public involvement, and its issuance was approved by the Commission, using essentially the same procedures typically employed in rulemaking proceedings.4 As an NRC guidance document, the GALL Report should be afforded special weight. See Private Fuel Storage, L.L.C.

(Independent Spent Fuel Storage Installation), CLI-01-22, 54 NRC 255, 264 (2001) 4 The NRC Staff developed the GALL Report in response to the Commissions direction to provide a basis for evaluating the adequacy of aging management programs for license renewal. GALL Rev. 1 at 1, 4; Memorandum from A. Vietti-Cook to W. Travers, Staff Requirements - SECY-99-148 - Credit for Existing Programs for License Renewal (Aug. 27, 1999) (ADAMS Accession No. ML003751930).

18 (Where the NRC develops a guidance document to assist in compliance with applicable regulations, it is entitled to special weight (footnote omitted)).

The Commission allows license renewal applicants to reference the GALL Report in order to demonstrate that the programs at its facility correspond to those reviewed and approved therein.

AmerGen Energy Co. (Oyster Creek Nuclear Generating Station),

CLI-08-23, 68 NRC 461, 467-68 (2008)). But the NRC does not allow applicants to simply cite the GALL Report. Instead:

If an applicant takes credit for a program in GALL, it is incumbent on the applicant to ensure that the plant program contains all the elements of the referenced GALL program.

In addition, the conditions at the plant must be bounded by the conditions for which the GALL program was evaluated.

The above verifications must be documented on-site in an auditable form. The applicant must include a certification in the license renewal application that the verifications have been completed.

GALL Rev. 1 at 3.5 As a result, assuming the plant conditions are bounded by those in the GALL Report and the plant program contains the elements of the GALL program, the license renewal applicants use of an [AMP] identified in the GALL Report constitutes reasonable assurance that it will manage the targeted aging effect during the renewal period. Oyster Creek, CLI-08-23, 68 NRC at 468 (emphasis added). The Commission recently reiterated this point, stating that a commitment to implement an AMP that the NRC finds is consistent with the GALL Report constitutes one acceptable method for 5

The Seabrook LRA includes this certification on pages B-4 and B-5. Referencing a program described in the GALL Report does not insulate a program from an adequately supported challenge at hearing. Vermont Yankee, CLI-10-17, 72 NRC at __ (slip op. at 47).

19 compliance with 10 C.F.R. § 54.21(c)(1)(iii).6 Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. (Vermont Yankee Nuclear Power Station),

CLI-10-17, 72 NRC __, __ (July 8, 2010) (slip op. at 44). Even then, the NRC Staff still reviews the LRA and its supporting documentation and conducts inspections and onsite audits to verify the information in the application, including the consistency of AMPs with GALL. Oyster Creek, CLI-08-23, 68 NRC at 468.

3.

Severe Accident Mitigation Alternatives in License Renewal Proceedings NEPA requires agencies considering a major federal action to provide a detailed statement on any adverse environmental effects which cannot be avoided should the proposal be implemented. 42 U.S.C. § 4332(2)(C)(ii). NEPA does not explicitly require any mitigation analysis, but the Council on Environmental Quality, and later, the Supreme Court, recognized that [i]mplicit in [this] demand... is an understanding that the EIS will discuss the extent to which adverse effects can be avoided. Robertson v.

Methow Valley Citizens Council, 490 U.S. 332, 351 (1989) (citations omitted).

Nonetheless, while Methow Valley requires a mitigation analysis, it made clear that NEPA imposes no substantive requirement that mitigation measures actually be taken.

Id. at 353, n.16. Accordingly, to ensure that environmental concerns have been sufficiently evaluated, an EIS must contain a reasonably complete discussion of measures to mitigate impacts. Id. at 351-52.

6 In Vermont Yankee, the Commission dealt specifically with aging management of components subject to a TLAA under 10 C.F.R. § 54.21(c)(1)(iii). However, in Oyster Creek, the Commission made clear that the same approach could also be taken to AMPs intended to demonstrate compliance with 10 C.F.R. § 54.21(a)(3). Oyster Creek, CLI-08-23, 68 NRC at 468 (the license renewal applicants use of an aging management program identified in the GALL Report constitutes reasonable assurance that it will manage the targeted aging effect during the renewal period); see also Vermont Yankee, CLI-10-17, 72 NRC at __ n. 85 (slip op. at 21). As the Commission stated in Vermont Yankee, section 54.21(c)(1)(iii) tracks the language of section 54.21(a)(3). CLI-10-17, 72 NRC at __ (slip op. at 20).

20 Although the NRC has argued that severe accidents7 are remote and speculative (see Limerick Ecology Action v. NRC, 869 F.2d 719, 739-40 (3rd Cir. 1989)), the NRCs Generic Environmental Impact Statement for License Renewal of Nuclear Plants (NUREG-1437) (GEIS) includes a generic evaluation of severe accident impacts and the technical basis for the NRCs codified conclusion that 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. 10 C.F.R. Part 51, Subpart A, Appendix B, Table B-1 (regarding severe accidents); GEIS at 5-12 to 5-106.

The severe accident analysis in the GEIS provides an estimated prediction of environmental impacts of severe accidents for all plants, using 95 percent upper confidence bounding values.

Entergy Nuclear Generation Co. and Entergy Nuclear Operations, Inc. (Pilgrim Nuclear Power Station), CLI-10-11, 71 NRC at __ (March 26, 2010) (slip op. at 38) (citing GEIS at 5-113, 5-115). The GEIS analysis also includes a discussion of the uncertainties associated with the likelihood of accident sequences and the estimates of environmental consequences, including the uncertainty in atmospheric dispersion modeling of radioactive plume transport, and acknowledges that plume dispersion may be influenced by the terrain surrounding the plant. Id. (citing GEIS at 5-100 to 5-106; 5-26). As the Commission explained in Pilgrim, [b]ecause the GEIS provides a severe accident impacts analysis that envelopes the potential impacts at all 7

Generally, the NRC categorizes accidents as design basis (i.e., the plant is designed specifically to accommodate these) or severe (i.e., those involving multiple failures of equipment or function and, therefore, whose likelihood is generally lower than design-basis accidents but where consequences may be higher). GEIS at 5-1.

21 existing plants, the environmental impacts of severe accidents during the license renewal term already have been addressed generically in bounding fashion. Id.

However, the NRC was unable to provide a generic conclusion in the GEIS regarding severe accident mitigation. GEIS at 5-113. NRC regulations therefore require license renewal applicants to provide a consideration of alternatives to mitigate severe accidents. 10 C.F.R. § 51.53(c)(3)(ii)(L). This is known as a severe accident mitigation alternatives, or SAMA, analysis. SAMAs refer to safety enhancements such as a new hardware item or procedure intended to reduce the risk of severe accidents. SAMA reviews ensure that any plant changes - in hardware, procedures, or training - that have a potential for significantly improving severe accident safety performance are identified and assessed. Duke Energy Corp. (McGuire Nuclear Station, Units 1 and 2; Catawba Nuclear Station, Units 1 and 2), CLI-02-17, 56 NRC 1, 5 (2002). The SAMA analysis does not replace the Commissions generic evaluation of severe accidents in the GEIS or the codified finding in Appendix B to Part 51. Pilgrim, CLI-10-11, 71 NRC at __ (slip op. at 37). It is not a site-specific severe accident analysis, but instead is simply a site-specific mitigation analysis. Id. at 38 (emphasis in original).

NEPA demands no fully developed plan or detailed examination of specific measures which will be employed to mitigate adverse environmental effects. Id. at 38 (quoting Methow Valley, 490 U.S. at 353)). In fact, the Commission has not proscribed by rule the requirements of an acceptable license renewal SAMA analysis. See Final Rule, Environmental Review for Renewal of Nuclear Power Plant Operating Licenses, 61 Fed. Reg. 28,467, 28,481 (June 5, 1996) (1996 Final Rule). Instead, it will review each severe accident mitigation consideration provided by a license renewal applicant on

22 its merits and determine whether it constitutes a reasonable consideration of severe accident mitigation alternatives. Id. at 28,481-82 (emphasis added).

Consistent with NEPAs rule of reason, the concept of alternatives must be bounded by some notion of feasibility. Catawba/McGuire, CLI-02-17, 56 NRC at 9 (citing Vermont Yankee Nuclear Power Corp. v. NRDC, 435 U.S. 519, 551 (1978)).

Accordingly, SAMA analysis is rooted in a cost-benefit assessment. Id. at 5. Whether a SAMA is worthwhile to implement is determined by a weighing of the costs to implement the SAMA with the reduction in risks to public health, occupational health, and offsite and onsite property.8 Id. at 7-8. Thus, it is unreasonable to trigger full adjudicatory proceedings based merely upon a suggested SAMA under circumstances in which the Petitioners have done nothing to indicate the approximate relative cost and benefit of the SAMA. Id. at 9.

The SAMA analysis presented in NextEras LRA follows a standard approach of evaluating the costs and benefits of particular SAMAs.9 First, it determined the risk of a severe accident through a Level 1 and Level 2 probabilistic risk assessment (PRA).

ER at F-9. The Level 1 PRA models internal and external initiating events, determining the contribution to core damage frequency (CDF) and the dominant initiating events.

Id. at F-12. The Level 2 analysis extends the Level 1 analysis to the release category 8

In promulgating its environmental rules for reactor license renewal, the Commission explained that it is unlikely that any site-specific consideration of [SAMAs] 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. 1996 Final Rule, 61 Fed. Reg. at 28,481. Instead, the Commission expected that any cost-beneficial changes generally would be procedural and programmatic fixes, with any hardware changes being only minor in nature and few in number. Id.

9 NextEras SAMA analysis follows the guidance in NEI-05-01, Severe Accident Mitigation Alternatives (SAMA) Analysis Guidance Document, November 2005. The NRC has endorsed this document and encouraged licensees to follow this approach. See Final License Renewal Interim Staff Guidance, LR-ISG-2006-03: Staff Guidance for Preparing Severe Accident Mitigation Alternatives Analyses, August 2, 2007.

23 potential for the Level 1 core damage end states. Id. at F-31. It determines release frequency, severity, and timing based on the Level 1 PRA, containment performance, and accident progression analyses.

NextEras Level 3 PRA model then determines off-site dose and economic impacts of severe accidents based on the Level 1 and 2 PRA results together with atmospheric transport, mitigating actions, dose accumulation, and economic analyses. Id.

at F-52. The off-site consequences of a severe accident are calculated using the MELCOR Accident Consequence Code System (MACCS2), Version 1.13.1.10 Id. The Level 3 PRA addresses plant-specific release data including the time-dependent nuclide distribution of releases and release frequencies, the behavior of the population during a release (evacuation parameters), and site-specific meteorology to simulate the probability distribution of impact risks (both exposures and economic effects) to the surrounding 50-mile radius population. Id.

For Seabrook, the total cost of severe accident risk (consequence times probability), and thus the maximum benefit of any particular SAMA, is $818,721. Id. at F-71. NextEra prepared a list of 191 SAMA candidates by reviewing the major contributors to CDF and population dose. Id. at F F-95. Following a two-step screening process (see ER at F-96, F-125), NextEra identified two SAMA candidates that are potentially cost-beneficial. Id. at F-186 - F-187.

10 The MAACS2 code is documented in NUREG/CR-6613, Code Manual for MACCS2: Volumes 1 and 2, which the ER references at F-52.

24 D.

Petitioners Have Not Proffered An Admissible Contention As discussed below, Petitioners have not proffered an admissible contention. At the outset, it should be noted that NextEra recently submitted a supplement to its LRA that materially affects Contentions 1 and 3. This supplement is briefly described below.

Over the past two years, the NRC has been engaged in a public process to revise the GALL Report for a second time. This revision (GALL Rev. 2) will incorporate lessons learned from the reviews of other LRAs, operating experience obtained following the issuance of GALL Rev. 1, and other public comments.11 These lessons learned led the NRC to revise or replace several GALL Rev. 1 AMPs, notably XI.M34, Buried Piping and Tanks Inspection and XI.E3 Inaccessible Medium-Voltage Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements. In the fall of 2010, the NRC publicly released a draft final version of GALL Rev. 2 (ADAMS Accession No. ML1026660219) in anticipation of an October 22, 2010 meeting of the Advisory Committee on Reactor Safeguards (ACRS). GALL Rev. 2 is scheduled to be published in final form in December 2010.12 NextEra has closely followed the NRCs efforts toward completion of GALL Rev. 2 and, in consultation with the NRC Staff, determined that a supplement to the Seabrook LRA was warranted in order to bring it in line with the scheduled revision to GALL. Accordingly, on October 29, 2010, NextEra submitted a supplement to its LRA, which reflected amendments to two aging management programs, the Buried Piping and 11 A preliminary version of GALL Rev. 2 was posted on the NRC public web page on December 23, 2009. The draft revision of the GALL Rev. 2 was further refined and issued for public comment on May 18, 2010. The NRC also held public meetings with stakeholders to facilitate dialogue and to discuss comments. The staff subsequently took into consideration comments received (see NUREG-1950) and incorporated the dispositions of those comments in the final version of GALL Rev. 2.

12 See http://www.nrc.gov/reactors/operating/licensing/renewal/guidance/updated-guidance.html

25 Tanks Inspection Program (LRA at B.2.1.22) and the Inaccessible Medium Voltage Cables Not Subject to 10 CFR 50.49 EQ Requirements Program (Non-EQ Inaccessible Medium-Voltage Cables Program) (LRA at B.2.1.34). Letter from P. Freeman, NextEra to NRC Document Control Desk, Supplement to the NextEra Energy Seabrook, LLC Seabrook Station License Renewal Application (October 29, 2010) (LRA Supplement),

attached hereto as Exhibit 1.

As shown below, Contentions 1 and 3 are inadmissible regardless of NextEras amendment of the two programs in question because they fail to meet the NRCs stringent pleading requirements. However, to the extent the Board finds either of these two contentions to be admissible, NextEra submits that they both have been mooted by its amendment of the LRA, which addresses the very issues raised in Contentions 1 and 3.

1.

Contention 1 - Inaccessible Cables Is Inadmissible In Contention 1, Petitioners assert:

The license renewal application for Seabrook Station fails to comply with the requirements of 10 C.F.R. §§ 54.21(a) and 54.29 because applicant has not proposed an adequate or sufficiently specific plan for aging management of non-environmentally qualified inaccessible electrical cables and wiring for which such aging management is required.

Without an adequate plan for aging management of non-environmentally qualified inaccessible electrical cables protection of public health and safety cannot be assured.

Pet. at 10-11.

Petitioners allege that NextEra has not demonstrated that the effects of aging will be adequately managed, because NextEra has failed to (1) identify the location and extent of Non-EQ Inaccessible Cables; (2) provide access to referenced documents; (3) address

26 specific recommendations from the referenced Sandia report; (4) provide a technical basis to support life extension using the existing cables without an aging management plan; and (5) provide a technical basis justifying differences between programs for aging management of accessible and inaccessible cables. Pet. at 12.

a.

Contention 1 Lacks Adequate Factual or Expert Support Contention 1 is copied, nearly verbatim, from New York State Contention 6, which challenged the Indian Point LRA.13 See New York State Notice of Intention to Participate and Petition to Intervene (Nov. 30, 2007) at 92-100 (ADAMS Accession No. ML073400187) (New York Petition). Instead of performing an independent review of the Seabrook LRA, Petitioners have simply removed the references to the Indian Point LRA from New York State Contention 6 and added a single block quotation from the Seabrook LRA. See Petition at 14. As a result, Petitioners fail to provide sufficient information to demonstrate the existence of a genuine dispute with this particular LRA in contravention of 10 C.F.R. § 2.309(f)(1)(vi), which requires petitioners to review the application at issue and identify specific deficiencies.

Contention 1 also suffers the related flaw of failing to include a concise statement of alleged facts or expert opinion in support of its claims. 10 C.F.R. § 2.309(f)(1)(v). The only support offered for Contention 1 is an unsigned declaration of Paul Blanch, a former nuclear industry engineer, that is not cited anywhere in the Petition and with respect to this Contention does not address the Seabrook LRA. See Pet. at 10-20. Indeed, Mr.

Blanch does not even claim to have read the Seabrook LRA. See Blanch Decl. at 4 (I 13 The New England Coalition (NEC) also recently copied New York State Contention 6 in an unsuccessful attempt to reopen the Vermont Yankee license renewal proceeding. [NECs] Motion to Reopen the Hearing and for the Admission of New Contentions (Aug. 20, 2010) (ADAMS Accession No. ML102420042).

27 have reviewed Vermont Yankees License Renewal Application and the subsequent submittals by Entergy to renew the operating licenses for Indian Point Unit 2 and Unit 3.

I have also reviewed pertinent sections of the NRCs Safety Evaluation Report [for Vermont Yankee] dated May 2008 (NUREG 1907)). Perhaps because he has not read the Seabrook LRA, Mr. Blanch claims, incorrectly, that his diligent review of the LRA and the NRC Staffs SER finds no such Time Limited Aging Analysis (TLAA) or Aging Management Program (AMP) for electrical cables. Id. at 7. Of course, the NRC Staff has yet to publish an SER for Seabrook and the Seabrook LRA contains both a TLAA for the Environmental Qualification of Electrical Components (LRA § 4.4) and three distinct aging management programs for electrical components, including the Non-EQ Inaccessible Medium-Voltage Cables Program (LRA App. B.2.1.34).14 This clearly incorrect claim is the only challenge Mr. Blanch raises to the Seabrook LRAs discussion of Non-EQ inaccessible cables. Accordingly, the Blanch Declaration is patently insufficient to support admission of Contention 1.

Because Mr. Blanchs declaration makes no effort to address the relevant AMPs in the Seabrook LRA, the few vague claims in the Contention itself pertaining to the Non-EQ Inaccessible Medium-Voltage Cables Program are entirely unsupported. Thus, 14 NEC made this same misleading statement in its recent attempt to introduce this contention in the Vermont Yankee proceeding, as that Board explained:

At one point, NEC seems to be arguing that the LRA contains no AMP that addresses the subject of age related degradation of safety-related electrical cables. Blanch Declaration at 8 (A diligent review of the LRA... finds no such [TLAA or AMP]). This is patently incorrect because the LRA contains an AMP for such cables. Entergy Answer at 27, Entergy Declaration at 1-3. Thus, we do not examine this issue.

Likewise, given that the LRA contains an AMP, there is no need for a TLAA on the same subject. See 10 C.F.R. § 54.21(c)(1)(i)-(iii). Thus, we do not need to analyze the TLAA prong of Contention 7.

Entergy Nuclear Vermont Yankee, L.L.C., and Entergy Nuclear Operations, Inc. (Vermont Yankee Nuclear Power Station), LBP-10-19, 72 NRC __, n. 18 (slip op at 21).

28 there is no support for NECs claim that it defies engineering logic to limit this AMP to cable subjected to system voltage more than 25 percent of the time.15 Pet. at 14 ¶13.

Petitioners fail to offer any engineering logic of their own in support of this bald claim, as the Blanch Declaration does not address either the energization threshold or the number of cables that would be excluded by this provision.

Likewise, NEC provides no support whatsoever for questioning the two-year maximum interval for inspecting for water collection. Pet. at 15 ¶17. Similarly, NEC provides no support for its claim that [t]here are no testing methods available to adequately assure the submerged or previously submerged cables will perform their

[intended] functions... Pet. at 14 ¶15.

Accordingly, Petitioners fail to provide sufficient factual assertions or expert opinion to demonstrate a genuine, material dispute in these aspects of Contention 1.

10 C.F.R. § 2.309(f)(1)(v) and (vi).

b.

Contention 1 Fails to Raise a Genuine Dispute With the LRA Much of Contention 1 either expresses agreement with the LRA or simply mischaracterizes it. Neither approach is sufficient to demonstrate the existence of genuine dispute with the applicant. 10 C.F.R. § 2.309(f)(1)(vi).

15 This limitation was taken directly from GALL Rev. 1 on which the AMP was based and is consistent. As the GALL Report explains, this AMP was intended to address an aging effect called water treeing that is voltage dependent: When an energized medium-voltage cable (2 kV to 35 kV) is exposed to wet conditions for which it is not designed, water treeing or a decrease in the dielectric strength of the conductor insulation can occur. This can potentially lead to electrical failure. GALL Rev. 1 at XI E-7.

The formation and growth of water trees varies directly with operating voltage. Water treeing is much less prevalent in 4kV cables than those operated at 13 or 33kV. Id. at XI E-9.

The GALL Report also stated As additional operating experience is obtained, lessons learned can be used to adjust the program, as needed. In light of the NRC Staffs recommendations in the draft GALL Rev. 2 to extend this program to low voltage cables, NextEra has revised this AMP so that it now applies to both low and medium voltage cable and is no longer limited to cable energized more than 25% of the time. Thus, even if Petitioners claim had been supported, it is moot.

29 GALL Rev. 1,Section XI.E3 provides an acceptable program for managing the effects of aging on Inaccessible Medium-Voltage Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements. This program applies to inaccessible (in vaults, conduits, troughs, or directly buried) medium-voltage cables within the scope of license renewal that are exposed to significant moisture (defined as periodic exposures to moisture that last for more than a few days, i.e., cable in standing water). GALL Rev. 1 at XI.E-7. GALL Rev. 1 calls for periodic actions to prevent cable exposure to significant moisture (inspecting manholes for water collection and draining water as needed). Id. at XI.E-7 to XI.E-8. Because these actions cannot guarantee that non-EQ inaccessible cables will not be submerged, the XI.E3 program also calls for testing of the conductor insulation in order to provide further assurance that the cable will perform its intended function. Id. at XI.E-7. GALL Rev. 1 calls for the use of a proven test for detecting deterioration of the insulation system due to wetting, such as power factor, partial discharge, or polarization index, as described in EPRI TR-103834-P1-2, or other testing that is state-of-the-art at the time the test is performed. Id.

Petitioners focus much of their contention on the need for an AMP for inaccessible cables. For example, Petitioners claim that [i]naccessible cables are significantly more likely to fail or experience undetected failures due to submergence and moisture accumulation.16 Pet. at 13 ¶11. Petitioners also argue, without support, that

[m]ost of the inaccessible cables at Seabrook are not specified to operate in a submerged environment. Pet. at 14 ¶14. But the NRC, in preparing the GALL Report, recognized that unplanned wetting or submergence may occur, potentially leading to cable failure.

16 Petitioners do not indicate what this sentence is meant to compare - more likely than what?

Regardless of what Petitioners mean, Petitioners offer no support for this claim, which is not addressed in the Blanch Declaration.

30 GALL Rev. 1 at XI E-7. Accordingly, the GALL Report recommends an aging management program, XI.E3, to manage those conditions. And NextEras LRA includes a Non-EQ Inaccessible Medium-Voltage Cables Program that is consistent with GALL Rev. 1, in order to manage this particular aging mechanism. See LRA at B-180-183.

As the Vermont Yankee Board recently explained, while the total preclusion of wetting or submergence below grade cables might be ideal, it does not appear that the mere existence of such wetting or submergence is automatically significant. Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. (Vermont Yankee Nuclear Power Station), LBP-10-19, 72 NRC __, __ (slip op. at 25-26). In fact, the Board reasoned, the potential for such wetting or submergence seems to be assumed, otherwise there would be no need for an AMP to manage it. Id. at 26. As a result, Petitioners claim that cables may become submerged is in agreement with the LRA and so does not demonstrate the existence of a genuine dispute with the applicant. 10 C.F.R.

§ 2.309(f)(1)(vi).

Petitioners also claim that NextEra failed to address specific recommendations from the recently issued Brookhaven report funded by the NRC and titled Essential Elements of an Electric Cable Condition Monitoring Program NUREG/CR-7000. Pet.

at 12 ¶8. NUREG/CR-7000 provides a number of general best practices for cable monitoring. However, Petitioners have failed to identify any provision of NUREG/CR-7000 that contradicts or is not already included in LRA App. B.2.1.34. The Blanch Declaration mentions NUREG/CR-7000 to show that there has been concern about environmental stressors on cables, which, again, is the aging effect the AMP was

31 developed to address. See Blanch Decl. at 9, n.2. Once more, this claim fails to demonstrate a genuine dispute with the applicant. 10 C.F.R. § 2.309(f)(1)(vi).

Similarly, Mr. Blanch quotes from the Executive Summary of NUREG/CR-7000 which states that in-service testing of safety-related systems can demonstrate the function of the cables under test conditions but does not provide specific information on the status of cable aging degradation processes nor the physical integrity and dielectric strength of its insulation and jacket materials. Blanch Decl. at 9 ¶ 26 (quoting NUREG/CR-7000 at xi). This portion of NUREG/CR-7000 refers to in-service systems testing conducted under current operating licenses. However, the Non-EQ Inaccessible Medium-Voltage Cables Program does not rely on the in-service systems testing to which NUREG/CR-7000 refers but instead requires a proven test that will provide an indication of the condition of the conductor insulation. LRA at B-181; see also GALL Rev. 1 at XI.E-7. The LRA and GALL Rev. 1 state that this type of testing includes power factor, partial discharge, or polarization index testing. Id. NUREG/CR-7000 identifies these same tests as methods that may be used to monitor the condition of insulation. See NUREG/CR-7000 at 3-4 to 3-7. Neither Petitioners nor Mr. Blanch address the specific testing required by the AMP, or explain why it is inadequate. Thus, Petitioners provide no information demonstrating a genuine material dispute with the testing specified in the LRA.

Where Petitioners do not demonstrate their agreement with the LRA, they routinely misstate its contents, again failing to demonstrate the existence of a genuine dispute with the applicant. For instance, Petitioners claim that there is no technical basis provided to support life extension using the existing cables without an aging management

32 plan. Pet. at 12 ¶9. But the LRA provides an AMP, which is the subject of this contention, a fact that Petitioners simply ignore.17 Further, Petitioners claim that NextEra has not demonstrated that the effects of aging will be adequately managed such that the intended functions will be maintained for the SSCs identified for pressurized water reactors in Table 1 of the GALL Report. Pet. at 12 ¶4. But that table provides the NRC Staffs Summary of Aging Management Programs for the Reactor Coolant System Evaluated in Chapter IV of the GALL Report and is not applicable to Non-EQ Cables.18 See GALL Rev. 1, Vol. 1, at 7.

Petitioners also claim that the Seabrook program is improperly limited to a program with a maximum time between inspections of no more than two years. Pet. at 15

¶17. According to Petitioners, [e]xperience indicates that not all inaccessible cables are capable of inspection via manholes, which leaves open the questions of how many cycles of wetting and drying (and freezing?) the insulation of a given cable may be expected to undergo in two years, and the potential effect on operability of the anticipated wet/dry cycles. Id. Petitioners again ignore the contents of the LRA, which explains:

The Seabrook Station program includes periodic inspections of manholes containing in-scope medium voltage cables. The inspection focuses on water collection in cable manholes, and draining water, as needed. The frequency of manhole inspections for accumulated water and subsequent pumping will be based on inspection results. The objective of the inspections is to keep the cables from becoming submerged thereby minimizing their exposure to significant moisture. To meet this objective, 17 As another example of Petitioners inaccurate pleading, they claim that [t]he Applicant has failed to provide a copy of its referenced Non-EQ Insulated Cables and Connections Program with the application. Pet. at 12. But the Seabrook LRA nowhere references a Non-EQ Insulated Cables and Connections Program.

18 Instead, Vol. 1, Table 6 of the GALL Report, Summary of Aging Management Programs for the Electrical Components Evaluated in Chapter VI of the GALL Report, is applicable to Non-EQ Cables.

NUREG-1801, Vol. 1, at 93. Item 4 of Table 6 is addressed in LRA Table 3.6.2-1 and Appendix B.2.1.34.

33 adjustments in inspection frequency may be required. The maximum time between inspections will be no more than two years.

LRA at B-181 (emphasis added). In other words, the program does not rely simply on manhole inspections, but also aims to prevent submergence by draining water. Id. The LRA also explains that more frequent inspections will be undertaken, if necessary. Id.

This argument misrepresents the LRA and so fails to present a genuine dispute with the applicant.19 Petitioners claim that NextEra failed to address specific recommendations from the referenced Sandia Report (SAND96-0344),20 which they claim contains recommendations related to the management of aging of cables and terminations. Pet at 12 ¶7.

See also Pet. at 16 ¶18 (claiming that the LRA does not commit to recommendations in SAND96-034). The Sandia Cable Aging Management Guideline (SAND96-0344) provides the technical basis for the GALL Report,Section XI.E3 program. See GALL Rev. 1 at XI.E-7. Appendix B.2.1.34 of the LRA clearly states that Seabrooks Non-EQ Inaccessible Medium-Voltage Cables Program is consistent with NUREG-1801 XI.E3, with no exceptions, and it also explains that the development of the B.2.1.34 program considers the technical information and guidance in [SAND96-0344].

LRA at B-181.21 Thus, this claim does not demonstrate a genuine dispute with the 19 Regardless, the Blanch Declaration does not address these claims and they are not supported by any factual assertions. See Blanch Decl. at 4-11. Accordingly, these claims are simply Petitioners unsupported speculation.

20 SAND96-0344, Aging Management Guideline for Commercial Nuclear Power Plants - Electrical Cable and Terminations, prepared by Sandia National Laboratories for the U.S. Department of Energy, September 1996.

21 Almost all of the language on page 18 and the top half of page 19, even the portions that are not presented in block quotes, is taken directly from the GALL Report, Vol. 2, page XI E-7 (Section XI.E3).

The medium voltage cable program in LRA Appendix B.2.1.34 certifies that it is consistent with the GALL Report,Section XI.E3, which references SAND96-0344. Neither the GALL Report nor any NRC

34 applicant. 10 C.F.R. § 2.309(f)(1)(vi). In any event, Petitioners fail to specify which recommendations they believe were not addressed.

Petitioners also claim that NextEra has not provided access to referenced documents that are not publicly available (e.g., EPRI TR-103834-P1-2 and EPRI TR-109619). Pet. at 12 ¶6. Petitioners claim to have searched for these documents using ADAMS, CITRIX, BRS, GOOGLE, and the EPRI website,the same searches performed by the State of New York in Indian Point - and were unable to find the documents. Compare id. with New York Petition at 93-94. EPRI TR-109619,22 which is referenced in the LRA, is publicly available in ADAMS (Accession No. ML003727052).

EPRI TR-103834-P1-223 is also referenced in the LRA because it is referenced in the XI.E3 program in the GALL Report. However, Petitioners offer no reason why they need to access EPRI TR-103834-P1-2, which is listed in the GALL program and the LRA only to identify representative testing methods that may be used. See GALL Rev. 1 at XI.E XI.E-8. The testing methods listed in the GALL and the LRA are known methods that are described in various publicly available documents, including very specifically in NUREG/CR-7000 cited by Mr. Blanch.

Petitioners also claim that the AMP description is vague and there are no discussions in the LRA that the recommendations of NUREG/CR-5643 have been addressed. Pet. at 15 ¶17. To the contrary, the LRA clearly states that the development regulation requires absolute adherence to referenced guidance documents without regard to other guidance.

SAND96-0344 was published in September 1996, and there have been numerous studies and inspection methods developed since SAND96-0344 was published, which is the reason that the NRC does not require a commitment to a specific method.

22 EPRI TR-109619, Guideline for the Management of Adverse Localized Equipment Environments, Electric Power Research Institute, Palo Alto, CA, June 1999.

23 EPRI TR-103834-P1-2, Effects of Moisture on the Life of Power Plant Cables, Electric Power Research Institute, Palo Alto, CA, August 1994.

35 of the program considered the technical information and guidance in NUREG/CR-5643.24 Regardless, Petitioners do not identify what specific insights from NUREG/CR-5643 they believe need to be but were not incorporated.

Finally, Petitioners claim that there is no basis to justify differences between programs for aging management of accessible cables and inaccessible cables. Pet. at 12

¶10. These two aging issues are covered by two different programs that have different criteria and requirements. The AMP for non-EQ accessible cables is provided in Appendix B.2.1.32, Electrical Cables and Connections Not Subject to 10 CFR 50.49 EQ Requirements, which is consistent with GALL Rev. 1,Section XI.E1. Petitioners fail to identify what differences between the programs they believe need to be justified.

c.

Contention 1 Fails to Raise A Material Issue Contention 1 challenges an AMP that is consistent with the GALL Report. The Board in Indian Point admitted New York State Contention 6 because the Board did not comprehend how a commitment to develop a program can demonstrate that the effects of aging will be adequately managed. Entergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 and 3), LBP-08-13, 68 NRC 43, 86 (2008) (emphasis in original). But as the Commission later explained in Vermont Yankee, a commitment to implement an AMP that the NRC finds is consistent with the GALL Report constitutes one acceptable method for demonstrating that the effects of aging will be adequately managed. CLI-10-17, 72 NRC at __ (slip op. at 44); see also Oyster Creek, CLI-08-23, 68 NRC at 468 (use of a GALL program constitutes reasonable assurance that it will manage the targeted aging effect during the renewal period). Because NextEra has committed to implement an AMP that is consistent with GALL and has certified that its 24 NUREG/CR-5643, Insights Gained From Aging Research (Feb. 1992).

36 program contains the ten elements of the GALL program and that the plant conditions are bounded by those evaluated in the GALL Report, NextEra has demonstrated reasonable assurance that the aging effects will be adequately managed so that the intended functions of the relevant SSCs will be maintained during the period of extended operation.

Petitioners have not directly challenged the GALL program or NextEras consistency with the GALL program, and so have failed to demonstrate that Contention 1 raises a material issue. 10 C.F.R. § 2.309(f)(1)(iv).

Petitioners make a number of claims about accidents that also fail to raise a material issue. The NRC has established nuclear station EQ requirements in 10 C.F.R.

§ 50.49, which requires that an EQ program be established to demonstrate that certain electrical components located in harsh plant environments (that is, those areas of the plant that could be subject to the harsh environmental effects of a loss of coolant accident (LOCA), high energy line breaks (HELBs) or post-LOCA environment) are qualified to perform their safety function in those harsh environments after the effects of inservice aging. These EQ analyses are considered to be a TLAA, which is addressed in section 4.4 of the Seabrook LRA.

The EQ requirements in 10 C.F.R. § 50.49 do not apply to equipment located in a mild environment. 10 C.F.R. § 50.49(c). A mild environment is an environment that would at no time be significantly more severe than the environment that would occur during normal plant operations, including anticipated operational occurrences. Id. As the NRC explained in describing the GALL Reports XI.E3 program, [s]ince they are not subject to the environmental qualification requirements of 10 CFR 50.49, the electrical cables covered by this aging management program are either not exposed to

37 harsh accident conditions or are not required to remain functional during or following an accident to which they are exposed. GALL Rev. 1 at XI E-7.

But Petitioners claim that [t]here are no testing methods available to adequately assure the submerged or previously submerged cables will perform their functions for the duration of the postulated accident. Pet. at 14 ¶15. The Blanch Declaration asserts that:

a cable circuit with undetected damaged or degraded insulation could pass an in-service functional test, but still fail unexpectedly when called upon to operate under anticipated environmental conditions or the severe stresses encountered during a design basis event (i.e., fully loaded equipment, more extreme environmental conditions, extended operation in a heavily loaded state).

Blanch Decl. at 9-10 ¶27 (emphasis added). Mr. Blanchs statement expresses concern about cables that are called upon to perform under harsh environmental conditions during an accident. However, the only cables that are subject to this AMP are those that are not subject to the EQ requirements under 10 C.F.R. § 50.49 and so are either not exposed to harsh conditions following a postulated accident or are not required to remain functional following an accident. See GALL Rev. 1 at XI E-7.25 Accordingly, Mr. Blanchs assertions about the impact of the severe stress and extreme environmental conditions of an accident do not raise a material issue with respect to these cables.

Petitioners also claim that [t]he failure to properly manage aging of the Non-EQ Inaccessible Cables could result in the loss of safety related cables and buses that supply emergency power to safety equipment including Station Blackout (SBO) loads, service water motors/pumps, safety injection pumps, and other electrical loads. Pet. at 11 ¶2.

25 Moreover, the Non-EQ Inaccessible Medium-Voltage Cables Program does not rely on inservice functional tests, but instead uses tests to provide an indication of the condition of the conductor insulation. LRA at B-181.

38 Petitioner provides no basis for such a claim. As demonstrated above, the LRA addresses aging management of Non-EQ Inaccessible Medium-Voltage Cables in a manner consistent with the GALL Report. Moreover, Petitioners claim is factually incorrect because there are no Station Blackout or safety injection pump cables within the scope of the AMP.

Finally, Mr. Blanch criticizes the application for addressing electrical cables as a commodity group and not identifying the location for each relevant cable. Blanch Decl.

at 13, ¶3826; see also Pet. at 12, ¶5. Mr. Blanch asserts that the LRA should provide drawings so that reviewers can identify location of cables that may be subjected to moisture and submergence. Id. However, as the GALL Report explains, [e]lectrical cables and their required terminations (i.e., connections) are typically reviewed as a single commodity.27 GALL Rev. 1 at VI A-1; see also NUREG 1800, Rev. 1, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, at 2.1-5, 2.1-13. Petitioners provide no support for their assertion that required information 26 This discussion is included with Mr. Blanchs discussion of transformers, but refers to submerged cables and appears to be misplaced.

27 The basis for including structures or components in a single commodity group can be such characteristics as similar design, similar materials of construction, similar aging management practices and similar environments. Also, the License Renewal SRP explains that applicants may use a plant spaces approach to determined the applicable environment for electrical components:

Under the plant spaces approach, an applicant would identify all passive, long-lived electrical equipment within a specified plant space as subject to an AMR, regardless of whether these components perform any intended functions. For example, an applicant could identify all passive, long-lived electrical equipment located within the turbine building (plant space) to be subject to an AMR for license renewal. In the subsequent AMR, the applicant would evaluate the environment of the turbine building to determine the appropriate aging management activities for this equipment.

SRP at 2.5.1. Therefore, aging effects for cables can be reviewed without identifying the precise location of every cable in the LRA.

39 is missing. The information requested by the Standard Review Plan for license renewals is provided in the LRA.28 Petitioners have failed to demonstrate that their concerns related to the aging management of inaccessible cables not subject to EQ requirements of 10 C.F.R. § 50.49 are material to the findings the NRC must make. 10 C.F.R. § 2.309(f)(1)(iv).

d.

Contention 1 Raises Issues Beyond the Scope of License Renewal As the Commission has explained, [u]nderlying the renewal regulations is the principle that each nuclear power plant has a plant-specific licensing basis that must be maintained during the renewal term in the same manner and to the same extent as during the original licensing term. Entergy Nuclear Generation Co. and Entergy Nuclear Operations, Inc. (Pilgrim Nuclear Power Station), CLI-10-14, 71 NRC __, __ (slip op. at

4) (2010) (citing 1995 Final Rule, 60 Fed. Reg. at 22,464); see also 10 C.F.R. § 54.30.

The Commissions license renewal rules did not throw open the full gamut of provisions in a plants current licensing basis to re-analysis during the license renewal review.

Turkey Point, CLI-01-17, 54 NRC at 5. Instead, the CLB is effectively addressed and maintained by ongoing agency oversight, review, and enforcement. Id. But Petitioners 28 In accordance with 10 C.F.R. § 54.4, the scoping and screening of electrical systems is described in LRA Section 2.5 (Scoping and Screening Results: Electrical and Instrumentation and Control (I&C)

Systems/Commodity Groups), which discusses components subject to aging management review. All electrical insulated cables and connections not subject to the EQ requirements of 10 CFR 50.49 are included in a single commodity group that is within the scope of license renewal (LRA at 2.5-3) and is subject to an aging management review (LRA at 2.5-6). This commodity group includes non-EQ cables and connections, connectors, electrical splices, fuse holders, terminal blocks, power cables, control cables, instrument cables, insulated cables and communication cables. LRA at 2.5-6. LRA Table 2.5.4-1 explains that these components have an intended function of electrical continuity. LRA at 2.5-7. LRA Table 3.6.2-1 provides the results of the aging management review for these cables, indicates that the aging effect requiring management is [l]ocalized damage and breakdown of insulation leading to electrical failure, and explains that the aging effect will be managed by the Inaccessible Medium Voltage Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements program (B.2.1.34). LRA at 3.6-15.

40 have raised a number of current operational issues that are not within the scope of this license renewal proceeding.

For example, Petitioners claim that [c]ables experiencing periodic submergence must be replaced with cables designed to operate in the environment to which they may be exposed, citing General Design Criteria (GDC) III and IV and NUREG-7000. Pet.

at 15 ¶16.29 Similarly, Petitioners claim that [m]ost of the inaccessible cables at Seabrook are not specified to operate in a submerged environment therefore operation of these cables is a clear violation of many NRC regulations including 10 CFR 50 Appendix A and B. Pet. at 14 ¶14. These unfounded assertions that Seabrook is not in compliance with its CLB do not raise aging management issues, but instead present matters relevant to current operations that are beyond the scope of this license renewal proceeding. See 10 C.F.R. § 54.30(b).

Petitioners also claim that NextEra does not incorporate certain measures from the NRCs Generic Letter 2007-01, Inaccessible or Underground Power Cable Failures That Disable Accident Mitigation Systems or Cause Plant Transients (GL 2007-01). Pet. at

16. GL 2007-01 does not address aging management and is unrelated to license renewal.

GL 2007-01 (ADAMS Accession No. ML070360665). The purpose of GL 2007-01 was to determine the number and types of cable failures experienced by nuclear power plants.

Accordingly, GL 2007-01 requested licensees provide the types of cable testing currently being performed. The letter did not request that licensees establish any additional cable programs and did not provide recommendations for improvements to existing cable 29 The Blanch Declaration contradicts this assertion by acknowledging that an effective AMP can account for the potential physical degradation effects of submergence in water on electrical cables. Blanch Decl. at 7, ¶20.

41 programs. NextEras response to GL 2007-01 was submitted in May 2007, and therefore, is part of the CLB, and is not litigable in this proceeding.30 e.

Contention 1 is Moot To the extent that Petitioners raise issues that will be covered by GALL Rev. 2 and the Board determines that those issues are otherwise admissible, Petitioners Contention has been mooted by NextEras recent supplement to its LRA to incorporate a revised AMP in Appendix B.2.1.34, titled Inaccessible Power Cables Not Subject to 10 CFR 50.49 EQ Requirements (Non-EQ Inaccessible Power Cables Program).

Petitioners challenge the original AMP, which is consistent with GALL Rev. 1, because it defined significant voltage exposure as being subjected to system voltage for more than 25% of the time. Pet. at 14. The new program eliminates this 25% threshold and applies to cables exposed to significant moisture regardless of the frequency of eneergization, thus making this issue moot.31 LRA Supplement, Encl. 2 at 6. Where a contention is superseded by the subsequent issuance of licensing-related documentsthe contention must be disposed of or modified. Duke Energy Corp.

(McGuire Nuclear Station, Units 1 and 2; Catawba Nuclear Station, Units 1 and 2),

CLI-02-28, 56 NRC 373, 382 (2002) (footnote omitted). Where a contention alleges the 30 Letter from J.A. Stall, Senior Vice President, Nuclear and Chief Nuclear Officer, to NRC Document Control Desk, Response to NRC Generic Letter 2007-01 Inaccessible or Underground Power Cable Failures that Disable Accident Mitigation Systems or Cause Plant Transients (May 8, 2007)

(ADAMS Accession No. ML071290579). The response explained that there was no history of failure of cables within the scope of the maintenance rule at Seabrook and also provided a description of testing and monitoring methods applicable at that time. Also, GL 2007-01 is referenced in the operating experience section of the LRA. See LRA at B-183.

31 The revised AMP provided in LRA Supplement includes a number of other significant changes.

Most notably, its scope is no longer limited to medium-voltage cables, and now includes low voltage power cables as well. LRA Supplement, Encl. 2 at 2, 6. As a result, the name of the program has been revised to Inaccessible Power Cables Not Subject to 10 CFR 50.49 EQ Requirements. Id. The revised program also reduces the maximum time between manhole inspections from two years to one year and the frequency of cable testing from at least every ten years to at least every six years. Id.

42 omission of particular information or an issue from an application, and the information is later supplied by the applicant... the contention is moot. Id. at 383 (footnote omitted).

In addition, Petitioners challenge whether manhole inspections with a maximum frequency of every two years will be sufficient to manage aging effects on cable insulation. Pet. at 15, ¶17. The revised AMP has reduced the frequency of manhole inspections to at least once per year and adds event-driven inspections. LRA Supplement, Encl. 2 at 2, 6. This challenge is now moot.

In sum, the LRA addresses aging management for Non-EQ Inaccessible Medium-Voltage Cables in accordance with the NRC regulations and guidance. As demonstrated above, Petitioners allegations are full of inaccuracies and misconceptions of NRC regulations and guidance. As a result, Petitioners have raised issues that are beyond the scope of this proceeding, contrary to 10 C.F.R. § 2.309(f)(1)(iii), raised issues that are not material, contrary to § 2.309(f)(1)(iv), failed to provide support for their position, contrary to § 2.309(f)(1)(v), and failed to provide sufficient information to show that a genuine dispute exists with regard to a material issue of law or fact, contrary to

§ 2.309(f)(1)(vi). Moreover, to the extent Contention 1 raises issues that are not addressed by GALL Rev. 1, it has been mooted by NextEras submittal of an amendment, committing to implement a revised AMP that is consistent with the recent draft of GALL Rev. 2 and addresses Petitioners concerns. Therefore, Petitioners Contention 1 must be dismissed in its entirety.

43

2.

Contention 2 - Transformers Is Inadmissible In Contention 2, Petitioners assert:

The LRA for Seabrook violates 10 C.F.R. §§ 54.21(a) and 54.29 because it fails to include an aging management plan for each electrical transformer whose proper function is important to plant safety Pet. at 20.

Contention 2 is inadmissible because it is not supported by any basis or support indicating a genuine, material dispute with the applicant. Once again, Petitioners have taken a contention that was admitted in the Indian Point proceeding and copied it nearly verbatim, without performing a sufficient review of the Seabrook LRA, resulting in quotations in the Petition with no relationship to the Seabrook LRA.32 See New York State Contention 8, New York State Petition at 103-05.

a.

Transformers are Active Components Not Subject to Aging Management Review 10 C.F.R. § 54.4 defines the plant SSCs within the scope of the NRCs license renewal rule, but 10 C.F.R. § 54.21(a)(1) then limits the structures and components subject to an aging management review to those structures and components that perform an intended function... without moving parts or without a change in configuration or properties.33 These are considered passive components. Pilgrim, CLI-10-14, 71 NRC at __ (slip op. at 5). Active components, by contrast, are not subject to an aging management review because [e]xisting regulatory programs, including required 32 For instance, in paragraph 4 on page 21, Petitioners reference Appendix A, Page A-35 which describes a

Structures Monitoring Program that includes a

program for monitoring transformer/switchyard support structures. In paragraph 5, Petitioners assert that the LRA also discusses the need for an AMP for transformer support structures. Neither of these quotations is contained in the Seabrook LRA.

33 The identification of this subset of components subject to review is screening. See NUREG-1800, Rev. 1, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, at 2.1-1, 2.1-9.

44 maintenance programs, can be expected to directly detect the effects of aging on active functions. Id. (citing 1995 Final Rule, 60 Fed. Reg. at 22,472).

In its Statements of Consideration accompanying the revised license renewal rule, the Commission concluded that a change in configuration or properties should be interpreted to include a change in state. 1995 Final Rule, 60 Fed. Reg. at 22,477. As an example, the Commission explained that a transistor can change its state, and so would be considered an active component. Id. In 10 C.F.R. § 54.21(a)(1)(i), the Commission provided additional examples of electrical components that are screened out of license renewal review as active components (including transistors, batteries, breakers, relays, switches, power inverters, circuit boards, battery chargers, and power supplies) but stated that the active component exclusion is not limited to these examples.

Subsequent NRC Guidance implementing 10 C.F.R. § 54.21(a)(1) specifically determined that transformers are active components and are excluded from aging management review. The NRC has determined that:

Transformers perform their intended function through a change in state by stepping down voltage from higher to lower value, stepping up voltage to a higher value, or providing isolation to a load. Transformers perform their intended function through a change of state similar to switchgear, power supplies, battery chargers, and power inverters, which have been excluded in §54.21(a)(1)(i) from an aging management review. Any degradation of the transformers ability to perform its intended function is readily monitorable by a change in the electrical performance of the transformer and associated circuits.

Trending electrical parameters measured during transformer surveillance and maintenance such as Doble test results, and advanced monitoring methods such as infrared thermography, and electrical circuit characterization and diagnosis provide a direct indication of the performance of the transformer. Therefore, transformers are not subject to an aging management review.

45 Letter from C. Grimes, NRC License Renewal Project Directorate, to D. Walters, NEI, Determination of Aging Management Review for Electrical Components (Sept. 19, 1997) at 2 (Grimes Letter). This NRC position is included in Appendix C to NEI 10, Industry Guidelines for Implementing the Requirements of 10 CFR Part 54 - The License Renewal Rule (Rev. 6, June 2005) (ADAMS Accession No. ML051860406).34 Similarly, the NRC Standard Review Plan indicates that transformers are not passive components subject to aging management review under Section 54.21(a)(1)(i). 35 See NUREG-1800 at 2.1-23, line 104.

Because transformers are active components, Petitioners have failed to show that their claim that NextEras LRA omitted an AMP for transformers is a material issue, i.e.,

one whose resolution would make a difference in the outcome of the licensing proceeding. 54 Fed. Reg. at 33,172. In fact, neither Petitioners (Pet. at 22 ¶9) nor Mr.

Blanch (Blanch Decl. at 12 ¶36 (the licensee has not provided for any AMP to assure

???????)) can offer any explanation of how a transformers intended function could be compromised by the lack of an AMP. This inability is likely due to the fact that transformer performance can be monitored and so transformers do not require aging management. In explaining why it created the distinction between active and passive components, the Commission stated:

a pump or valve has moving parts, an electrical relay can change its configuration, and a battery changes its electrolyte properties when discharging. Therefore, the 34 NEI-95-10 is endorsed by NRC Regulatory Guide 1.188 Rev. 1.

35 A similar contention was filed in the Prairie Island license renewal proceeding. See Northern States Power Co. (Prairie Island Nuclear Generating Plant, Units 1 and 2), LBP-08-26, 68 NRC 905, 946 (2008). After the applicants Answer explained why transformers are considered by the NRC to be active components, the Petitioner acknowledged the NRCs position and withdrew its contention. Id.

46 performance or condition of these components is readily monitored and would not be captured by this description.

1995 Final Rule, 60 Fed. Reg. at 22,477 (emphasis added). And as the Grimes Letter explained, degradation of a transformers function can be readily monitored by a change in the electrical performance of the transformer and associated circuits. Grimes Letter at

2. Also, trending electrical parameter measurements provides a direct indication of the performance of the transformer. Id. The fact that transformer performance can be monitored and trended, as explained in the Grimes Letter, shows that they are appropriately considered to be active components under the framework set forth in the Commissions Statements of Consideration.

b.

Contention 2 Lacks the Requisite Factual or Expert Support Petitioners fail to provide any explanation why transformers should be considered passive components under the rules. Instead, they simply make conclusory and contradictory claims without any support or reasoning. As a threshold matter, both Petitioners and Mr. Blanch contradict their own positions by admitting that transformers are active devices.... Pet. at 22 ¶9; Blanch Decl. at 12 ¶36. But they then assert without explanation or support that transformers contain no moving parts and do not undergo a change of properties or state. Pet. at 22 ¶8: Blanch Decl. at 11 ¶28.

If Petitioners and Mr. Blanch do believe transformers to be passive components, neither provides any support for such conclusion. Neither explains why transformers do not have a change in configuration or propertiesthey simply assert that transformers are passive components. But conclusory statements, even when provided by an expert, are insufficient to demonstrate that further inquiry is appropriate. USEC, CLI-06-10, 63 NRC at 472. Petitioners must do more than simply state that transformers function without a

47 change in configuration or properties, they must explain why they believe that to be the case.

Finally, Petitioners quote, without attribution, an NRC letter dated April 1, 2002, Staff Guidance on Scoping of Equipment Relied on to Meet the Requirements of the Station Blackout (SBO) Rule (10 CFR 50.63) for License Renewal (10 CFR 54.4(a)(3))

(ADAMS Accession No. ML02090464). Pet. at 22 ¶10. The letter explains that the plant system portion of the offsite power system used to connect the plant to the offsite power source should be included within the scope of license renewal, a path that typically includes switchyard circuit breakers that connect to the offsite system power transformers (startup transformers), the transformers themselves... Pet. at 22 ¶10. As the letters title indicates, it addresses the scope of license renewal under 10 C.F.R. § 54.4, not whether any particular components are screened out under 10 C.F.R. § 54.21(a)(1). See also Indian Point, LBP-08-13, 68 NRC at 88.

In sum, Petitioners provide no basis (other than confused and conclusory statements) to dispute the NRCs determination that transformers are active components.

Moreover, Petitioners make no attempt to address the NRCs Statements of Consideration or interpretive guidance that is available on the NRC website. Petitioners have raised issues that are not material, contrary to § 2.309(f)(1)(iv), have not provided support for their assertions, contrary to § 2.309(f)(1)(v), and have not provided sufficient information to show that a genuine dispute exists with regard to a material issue of law or fact, contrary to § 2.309(f)(1)(vi). Contention 2 is inadmissible.36 36 Petitioners also argue that they cannot determine which transformers are within the scope of license renewal because Seabrooks updated final safety analysis report is not publicly available. Pet. at 22,

¶8. NextEra submits updates to its FSAR to the NRC within six months of the end of each refueling outages. 10 C.F.R. § 50.71(e)(4). The latest revision currently available in ADAMS is Revision 11, dated

48 3.

Contention 3 - Buried, Below Ground, or Hard-to-Access Piping Is Inadmissible In Contention 3, Petitioners assert:

The aging management plan contained in the license renewal application violates 10 C.F.R. §§ 54.21 and 54.29(a) because it does not provide adequate inspection and monitoring for corrosion, structural

failure, degradation, or leaks in all buried systems, structures, and components that may convey or contain radioactively-contaminated water or other fluids and/or may be important to safety.

Pet. at 22-23 ¶1.

For the reasons set forth by the Licensing Board in Prairie Island, Contention 3 is inadmissible. See Prairie Island, LBP-08-23, 68 NRC at 944-45. In Prairie Island, as here, the petitioners filed a contention borrowed almost entirely from New York State Contention 5 in the Indian Point proceeding. Compare New York State Petition at 80.

Contrary to the Boards decision to admit the contention in Indian Point, the Prairie Island Board rejected the contention, finding it impermissibly broad because it was aimed at all buried systems, structures, and components that may convey or contain radioactively-contaminated water but the petitioner did not identify any specific buried components that may contain radioactive fluids. Prairie Island, LBP-08-26, 68 NRC at 944. The Prairie Island Board also noted that the petitioner failed to dispute with fact or expert opinion the Applicants assertion that none of the buried pipes within the scope of license renewal contain radioactive water. Id. at 944-45. Finally, the Board ruled that the petitioners claims regarding the plants monitoring and leak prevention programs May 8, 2007. See ML071430583, et al. Regardless, because transformers are screened out as active components, their precise location is not material.

49 have no relevance to aging management and so are beyond the scope of this proceeding.

Id. at 945.

Further, to the extent that this Board finds portions of Contention 3 to be adequately stated and supported, they are now moot. The revised buried piping AMP NextEra recently submitted to the NRC addresses many of the specific claims Petitioners raise in Contention 3, rendering it moot.

a. Contention 3 Is Beyond the Scope of a License Renewal Proceeding Contrary to the wide breadth of Contention 3, the license renewal rule does not encompass all systems, structures and components that may contain or convey water, radioactively-contaminated water, and/or other fluids. See Pet. at 23 ¶1. The scope of the NRCs license renewal regulations is carefully prescribed in 10 C.F.R. § 54.4. That provision limits the scope of 10 C.F.R. Part 54 to (1) safety related SSCs relied on to maintain the integrity of the reactor coolant pressure boundary, to shut down the reactor and maintain it in a safe condition, and to prevent or mitigate the consequences of reactor accidents; (2) non-safety related SSCs whose failure could prevent such safety-related systems from accomplishing their intended function; and (3) other nuclear power plant SSCs relied on to comply with specific Commission rules. Not one of these criteria involves protection against leaks of radioactively contaminated water.

In fact, several plant systems and components that may contain or convey radioactively contaminated water are not within this defined scope of 10 C.F.R. Part 54.37 37 The Commission specifically denied a petition for rulemaking filed by the Union of Concerned Scientists that would have revised the scope of the license renewal rule to cover liquid and gaseous radioactive waste management systems. 66 Fed. Reg. 65,141 (Dec. 18, 2001). The Commission denied the petition because (1) liquid and gaseous radioactive waste management systems are not involved in design and licensing basis events considered for license renewal, and (2) the existing regulatory process is acceptable for maintaining the performance of the radioactive waste systems throughout the period of

50 Petitioners broadly worded contention would, however, include such systems in its claims of inadequate aging management, a direct challenge to the Commissions determination that such systems are not covered by the license renewal rules. In so doing, the Contention impermissibly challenges the Commissions regulation. 10 C.F.R.

§ 2.335(a).

Petitioners argue that leaks and corrosion of buried piping compromise their ability to perform their intended function. Petition at 24 ¶6. Petitioners do not identify the intended function to which they refer, but that intended function cannot be the prevention of leaks of radioactively contaminated water. The only issue for license renewal is whether covered systems and components will continue to perform their intended safety functions or other license renewal function as specified in 10 C.F.R.

§ 54.4(a)(1)-(3), during the license renewal period. Entergy Nuclear Generation Co. and Entergy Nuclear Operations, Inc. (Pilgrim Nuclear Power Station), LBP-07-12, 66 NRC 113, 129 (2007). Intended functions are defined in 10 C.F.R. § 54.4(b) as those functions that are the bases for including them within the scope of license renewal as specified in 10 C.F.R. § 54.4(a)(1)-(3). The original version of the buried piping AMP in the LRA indicated that there are seven systems within the scope of license renewal that contain buried piping. LRA at B-125. Prevention of leakage of radioactive liquids into the environment is not the intended function for any of these pipes.38 Whether an AMP extended operation in order to keep exposures to radiation at the current levels below regulatory limits consistent with the conclusions made in the applicable regulations. Id.

38 There are three intended functions listed in the LRA for piping in these seven systems:

(1) pressure boundary; (2) structural integrity (attached); and (3) leakage boundary (spatial). These intended functions are defined in the LRA at pp. 2.1-29 and 2.1-30. The seven systems and their specific intended functions for piping are listed on the following LRA pages: (1) the auxiliary boiler system (LRA at 2.3-60); (2) the control building air handling system (id. at 2.3-112); (3) the diesel generator system (id.

51 created to address a different intended function also prevents the release of radioactivity into groundwater is not a material issue. See Nuclear Management Co., LLC (Monticello Nuclear Generating Plant), LBP-05-31, 62 NRC 735, 754 (2005) (rejecting claims of inadequate radiation monitoring and asserted need for new monitoring techniques).

Instead, such concerns are covered by the plants operational monitoring programs, which, as the Commission has made clear, are not within the scope of license renewal. License renewal reviews are not intended to duplicate the Commissions ongoing reviews of operating reactors. Turkey Point, CLI-01-17, 54 NRC at 7; see also Entergy Nuclear Generation Co. and Entergy Nuclear Operations, Inc. (Pilgrim Nuclear Power Station), LBP-06-23, 64 NRC 257, 274-77 (2006); 1991 Final Rule, 56 Fed. Reg.

at 64,946; Monticello, LBP-05-31, 62 NRC at 754. In fact, radiation monitoring programs... are subject to ongoing regulatory oversight... and, therefore, are beyond the scope of [a license renewal] proceeding. Id. (footnote omitted); see also Pilgrim, LBP-07-12, 66 NRC at 130 n.81 (monitoring of radiological releases, or determinations of how leakage could harm health or the environment, are not legitimately in dispute here, because they do not relate to aging and/or because they are addressed as part of ongoing regulatory processes) (citation omitted).

Despite this clear precedent, Petitioners claim that there is no adequate monitoring to determine if and when leakage from these systems, structures and components occurs (Pet. at 23 ¶1) and that deficiencies in leak detection could at 2.3.133); (4) the fire protection system (id. at 2.3.-151); (5) the plant floor drain system (id. at 2.3-195);

(6) the service water system (id. at 2.3-241); and (7) the condensate system (id. at 2.3-298).

Mr. Blanch identities seven systems that he claims contain piping and may contain radioactive water. Blanch Decl. at 15 ¶43. The list (apparently copied from Indian Point) only partially corresponds to the list of Seabrook systems that contain buried piping and includes systems that are not identified in the Seabrook LRA (e.g., security generator, ECCS, auxiliary feedwater).

52 endanger the safety and welfare of the public and therefore are within the scope of a re-licensing hearing. Pet. at 24 ¶5. To the contrary, this claim is beyond the scope of this proceeding because monitoring for radiological releases is a CLB issue that is beyond the scope of license renewal.

While a similar contention alleging the need to prevent leakage of radioactive liquids from buried piping was admitted in the Pilgrim proceeding (LBP-06-23, 64 NRC at 310-15), the Board in that case reformulated the Contention, and the Commission subsequently suggested that such reformulation was not appropriate. Accordingly, the Pilgrim case should not be followed.

In that case, the Board acknowledged that prevention of leaks per se is not a stated objective of any relevant aging management program, and made clear that issues concerned with monitoring of radiological releases, or determination of how leakage could harm health or the environment, are not legitimately in dispute here, because they do not relate to aging and/or because they are addressed as part of ongoing regulatory processes. LBP-07-12, 66 NRC at 129. However, the Board reformulated the contention from one focused on radiological leakage to one challenging whether Entergys aging management program for buried pipes and tanks are adequate on their own, without need of any leak detection devices... to assure that the pipes and tanks in question will perform their intended functions and thereby protect public health and safety. Id. at 128. The Pilgrim Board reasoned that this claim was implicitly raised in the original radiological leakage contention. Id. at 129.

In ruling on the Pilgrim petitioners petition for review of the Boards initial decision in that case, the Commission made clear that such a reformulation was not

53 appropriate. CLI-10-14, 71 NRC __, n. 80 (slip op. at 19). The Commission explained that it has urged the licensing boards to exercise caution when reformulating contentions. Id. While a licensing board may reformulate contentions to eliminate extraneous issues or to consolidate issues for a more efficient proceeding... a board should not add material not raised by a petitioner in order to render a contention admissible. Crow Butte Resources, Inc. (North Trend Expansion Area), CLI-09-12, 69 NRC 535, 552-53 (2009) (citing Andrew Siemaszko, CLI-06-16, 63 NRC 708, 720-21 (2006)). The Commission in Pilgrim encouraged the licensing boards to adhere to this standard when reformulating contentions. In order to admit Petitioners Contention 3 in this proceeding, the Board would need to engage in the same reformulation of the contention that the Commission criticized in Pilgrim.

Because Contention 3: (1) focuses on monitoring of radioactive leakage; and (2) seeks to include all underground piping and tanks regardless of whether they perform an intended function as defined in 10 C.F.R. § 54.4(b), it fails to comply with 10 C.F.R.

§ 2.309(f)(1)(iii), which requires that a contention must raise an issue that is within the scope of the proceeding.

b.

Contention 3 Lacks the Requisite Factual or Expert Support Contention 3 also fails to provide factual or expert support and so fails to provide sufficient information to demonstrate the existence of a genuine dispute with the applicant. 10 C.F.R. §§ 2.309(f)(1)(v), (vi). Where New York State Contention 5 in Indian Point was supported by the declarations of Dr. Rudolf Hausler39 and Tim Rice,40 39 According to his declaration in Indian Point, Dr. Hausler made his recommendations on the basis of my education and experience as a corrosion engineer with proven expertise... in chemistry, physical chemistry, electrochemistry, corrosion-chemistry, -processes, -mechanisms, and -phenomenology, failure analysis, corrosion modeling and management (inhibitors and other chemical additives), and system

54 Petitioners here have simply deleted the references that New York State provided to the Hausler and Rice Declarations.41 For instance, in paragraph 3 on page 23, Petitioners deleted a reference to the Rice Declaration in support of the statement that plant pipes may contain radioactive water.42 On page 25, in paragraphs 8, 9, 10, 11, and 12, Petitioners deleted references to the Hausler Declaration, leaving no cited support for their assertions regarding test frequency (¶¶ 8, 9), baseline conditions (¶ 10), cathodic protection (¶ 11), and NACE corrosion control standards (¶ 12). Petitioners also replaced paragraph 13 in its entirety, in which New York State respectfully refer[red] to and incorporate[d] the accompanying Declarations of Rudolf Hausler, Ph.D., and Timothy Rice. See New York State Petition at 84.

Instead of these Declarations, Petitioners attach the unsigned Blanch Declaration, which Petitioners do not ever cite in Contention 3. See Pet. at 22-33. Mr. Blanch states in his Declaration that he has a bachelors degree in electrical engineering, but provides no evidence that he has any specialized knowledge about either corrosion or radiological or environmental protection. See Blanch Decl. at 2. Instead, he has simply restated New analysis for corrosion management. New York State - Notice of Intention to Participate and Petition to Intervene, Supporting Declarations and Exhibits, Volume I of II (ADAMS Accession No. ML073400205)

(New York State Exhibits), Hausler Decl. at 2.

40 According to his declaration in Indian Point, Mr. Rice is an Environmental Radiation Specialist for the New York State Department of Environmental Conservation. New York State Exhibits, Rice Decl.

at 1.

41 While Petitioners deleted most of the references to the Indian Point application, several remain.

See e.g., Pet. at 32 ¶21. The references to Appendix A.3.1.5 and A-46 do not correspond to any discussion of buried piping in the Seabrook LRA. Further, the quoted passage stating that buried components are inspected when excavated during maintenance, does not appear in the Seabrook LRA. Pet. at 32 ¶21. To present an admissible contention, petitioners must demonstrate a genuine dispute with this application by citing specific portions of the application... that the petitioner disputes. 10 C.F.R. § 2.309(f)(1)(vi).

42 Petitioners added a statement in paragraph 3 that this water may be in excess of EPA drinking water limits, which did not appear in the New York State contention and for which Petitioners offer no support.

55 York Contention 5 from Indian Point as his own declaration. Even if Mr. Blanch is an expert on corrosion and radiological/environmental protection (which Petitioners have not shown), his bald and conclusory statements would not be sufficient to support the admission of a contention. Conclusory statements, even when provided by an expert, are insufficient to demonstrate that further inquiry is appropriate. USEC, CLI-06-10, 63 NRC at 472.

Moreover, as was the case in Prairie Island, there are no buried components within the scope of the license renewal rule at Seabrook that contain radioactive liquids.

Petitioners do not show that any of the buried piping that is within the scope of the license renewal rule contains radioactive liquid. Instead, they simply offer the speculation of Mr. Blanch, who claims that certain pipes and tanks may contain radioactive water, either by design or through a structural or system failure. Blanch Decl. at 15 ¶43. This assertion is pure speculation. Speculation, even by an expert, is insufficient to allow the admission of a proffered contention. Southern Nuclear Operating Company (Vogtle Electric Generating Plant Units 3 and 4), LBP-09-03, 69 NRC 139, 153 (2009) (citing Fansteel, Inc., CLI-03-13, 58 NRC at 203). Therefore, Petitioners do not show the existence of any genuine dispute with the applicant, contrary to 10 C.F.R.

§ 2.309(f)(1)(vi).

In a further effort to support their Contention, Petitioners present a list of events at other plants in which radioactively contaminated water was released into the ground.

Pet. at 26 ¶ 14. Petitioners claim that [o]ne common aspect of many of these leaks is that they have been discovered by happenstance and that they usually have gone undetected for an extended period of time thereby permitting increasingly larger amounts

56 of contaminated water to enter the ground (or air) around the facilities. Pet. at 30 ¶17.

However, Petitioners make no showing that any of these events is relevant to buried components within the scope of license renewal at Seabrook or to the adequacy of the programs managing the aging of such components at Seabrook. Therefore, the incidents at other plants provide no factual basis for the contention. A brief review of some of the incidents referred to in the Petition demonstrate the lack of relevance to the license renewal issues presented here:



The document cited by Petitioners (Pet. at 26-27) regarding the Dresden Nuclear Power Plant states that the leakage event was not the cause of the elevated levels of tritium. See NRC Preliminary Listing of Events Involving Tritium Leaks (Mar.

28, 2006), ML060930382 (licensees other monitoring results and an independent hydrology study do not appear to support that the elevated levels of tritium in that well were from the 2004 [Condensate Storage Tank] pipe leakage.) at 4.



The document cited by Petitioners (Pet. at 27) regarding Palo Verde Nuclear Generating Station does not claim that buried pipes or tanks containing radioactive fluids are the source of the tritium contamination found in water onsite. See Follow Up for Tritium Contamination Found In Water Onsite (March 17, 2006), ML060760584 (The apparent cause or source of the elevated tritium levels in the test holes has not been found/determined to date and is still under investigation by the licensee.) at 1.



The document cited by Petitioners (Pet. at 28) regarding Catawba Nuclear Power Station does not discuss the source of tritium discovered. Nothing in this document suggests that in-scope license renewal buried pipes and tanks are the source of the tritium. NRC Preliminary Notification of Event or Unusual Occurrence, PNO-II-07-012, Onsite Groundwater Tritium Contamination (Oct.

2007), ML072850013. [The petition incorrectly cites the ADAMS accession number of this document.]



The document cited by Petitioners (Pet. at 28) regarding Quad Cities Nuclear Power Station only states that underground piping was being examined as a potential source. NRC Preliminary Notification of Event or Unusual Occurrence, PNO-III-08-011, Tritium Leakage (Oct. 11, 2007), ML072890262.



The document cited by Petitioners (Pet. at 28) regarding Byron Nuclear Power Station explains that the leaking portion of the pipe goes from basins to the ground and that rust was found on the sections of piping between the basins and

57 the ground. It does not state that the leak was in the buried portion of the pipe.

NRC Preliminary Notification of Event or Unusual Occurrence, PNO-III-07-012, Both Units at Byron Shut Down Due to a Leak in Pipe (October 23, 2007),

ML072960109.

Petitioners also allege spent fuel pool leaks at Indian Point (the spent fuel pool is not a buried pipe) and make general allegations to elevated tritium at Indian Point. Pet. at 29-

30. While these references may have had some relevance to the original version of this contention at Indian Point, they have none here. See New York State Petition at 87.

Incidents at other plants generally are not probative for consideration at an individual plant absent some showing of relevance or similarity. See Pilgrim, LBP-07-12, 66 NRC at 130.

Petitioners allegations of four-and five-year-old cases of elevated tritium (which were recent when first cited in the Indian Point proceeding three years ago) does not demonstrate the existence of a genuine dispute with the LRA. Reports of leakage at other plants provide no basis to support a claim that in-scope systems at Seabrook with underground piping are likely to leak radioactive fluids or that Seabrooks aging management plan for underground piping is inadequate. Indeed, as noted above, none of the buried pipes at Seabrook within the scope of license renewal contain radioactively contaminated water. Petitioners present no indication that these plants were implementing an AMP for underground piping when the leaks occurred. Thus, there is no basis to suggest that the program described in the LRA is deficient. The reported leaks at other plants provide no basis for the claims in Contention 3, contrary to 10 C.F.R.

§ 2.309(f)(1)(v).

58

c.

Contention 3 Fails to Show a Genuine Dispute on a Material Issue Petitioners also fail to provide sufficient information to demonstrate the existence of a genuine dispute on an issue of material fact or law, contrary to 10 C.F.R.

§ 2.309(f)(1)(vi). Petitioners argue that NextEras LRA consists of no preventative measures... Pet. at 24-25, ¶8. But the original Buried Piping and Tanks Inspection Program that Petitioners Contention addressed was based on, and consistent with, the GALL Rev. 1 program, XI.M34, which, like all GALL programs, included a Preventive Actions element. See GALL Rev. 1, at XI M-111. It specifically stated that it relies upon existing preventive measures such as coating or wrapping to protect the piping from contacting the aggressive soil environment. Id.; see also LRA at B-125. And the revised AMP that NextEra has submitted to address GALL Rev. 2 addresses an even greater range of preventive actions.43 Petitioners claim that NextEras LRA consists of... no leak tests any more frequently than every ten years.... (Pet. at 24-25, ¶8) similarly fails to address and show a genuine dispute with the LRA. Contrary to Petitioners assertion, the LRA does not address leak teststhe Buried Piping and Tanks Inspection Program is an inspection program. See GALL Rev. 1, at XI M-111. Further, the original AMP provided for inspections more frequently than every ten years, if an opportunity for such inspection were presented, such as a piping modification or periodic maintenance that requires digging around the pipe to provide for visual inspection44 (LRA at B-125); and the 43 GALL Rev. 2 includes a substantially new buried piping program, entitled Buried, Underground, and Limited-Access Piping and Tanks. GALL Rev. 2 at XI M41-1.

44 Petitioners mischaracterize opportunistic inspections as allowing assessment of pipe conditions without excavation. Pet. at 32, ¶21. Inspection of buried pipe would require excavation regardless of whether it was planned or opportunistic.

59 revised AMP now commits to a more much detailed and frequent inspection program based on risk ranking of components. LRA Supplement Encl. 1 at 10-12.

Petitioners only specific (and accurate) reference to the Seabrook LRA is in the 25th numbered paragraph of Contention 3, where they include a block quote from the introduction to the LRAs program description. Pet. at 33 (citing LRA at B-125).

However, Petitioners fail to raise a genuine dispute with this quoted portion of the LRA.

Petitioners claim that the AMP is limited to external surfaces only and only piping containing iron. Pet. at 33, ¶26. The original Buried Piping and Tanks Inspection Program in the Seabrook LRA was consistent with XI.M034 in GALL Rev. 1 (see LRA at B-126), which was only intended to manage external corrosion of steel components.

Petitioners ignore the existence of other AMPs intended to address internal corrosion, such as the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Surfaces (LRA App. B.2.1.25).

Nevertheless, Petitioners fail to offer any support for their belief that internal corrosion needs to be age-managed or that materials other than steel (Petitioners say iron) need to be age-managed. See Pet. at 33. Mr. Blanch argues that the scope of the program should be expanded to include inter alia, stainless steel pipes. Blanch Decl. at 14, ¶ 40. However, he fails to acknowledge that Seabrook took an exception from the GALL program, specifically to expand its scope to include stainless steel pipes. See LRA at B-126. Nor does he state with specificity any pipes made of other materials that he believes should fall within the scope of the program.

Petitioners also claim that [m]ost tanks containing radioactive materials and/or perform functions within the scope of 10 CFR 54.4 are partially buried and not addressed

60 by this particular AMP, specifically identifying the following tanks: (RWST, CST, Spent Fuel Pool, Waste Tanks etc.). Pet. at 33, ¶ 27. Once more, this claim has no factual support and the Blanch Declaration does not address it. The LRA explains that there are no buried tanks at Seabrook within the scope of license renewal. LRA at B-125.

Moreover, each of the components identified by Petitioners, with the exception of Waste Tanks, a phrase that does not appear in the LRA, are subject to other AMPs.

Mr. Blanch argues that the LRA references mechanical scoping drawings that are not publicly available. See Blanch Decl. at 13-14, ¶39. However, by letter dated June 1, 2010, NextEra provided the NRC with the drawings illustrating the in-scope mechanical systems, structures, and components that are subject to aging management review, all of which are publicly available.45 Regardless, Petitioners do not identify any regulation requiring a license renewal applicant to provide such drawings, so this issue is not material.

Petitioners also reference a paper by David Lochbaum that claims that aging related problems in nuclear plants increase as the plant ages. Pet. at 30-31, ¶18.

NextEra acknowledges that buried pipes can leak due to corrosion and other effects of aging, which is why they are covered by an AMP. And of course, the general concept of aging effects that Lochbaum addresses is the reason why the NRC requires AMPs in the first place. This discussion does not present a genuine dispute with the LRA. 10 C.F.R.

§ 2.309(f)(1)(vi).

45 See Letter from P. Freeman, NextEra, to NRC Document Control Desk, Seabrook Station -

Supporting Information for NRC Review of Application for Renewed Operating License (June 1, 2010)

(ADAMS Accession No. ML101620330). The drawings are available at ADAMS accession numbers ML101620331, ML101620333, ML101620334, ML101620337, and ML101620329.

61 Lastly, Petitioners claim that the LRA does not appear to consider the current revision of GALL (NUREG 1800) with respect to pipes and tanks. Pet. at 33, ¶ 28. But Seabrooks original AMP certified its consistency with GALL Rev. 1, the then current and still current version of the GALL Report. And as discussed below, NextEra has kept pace with changes to the GALL, incorporating a revision to this program to be consistent with the applicable program in the soon-to-be-published Revision 2 of the GALL Report.

Petitioners have failed to demonstrate the existence of a genuine dispute with the LRA on a material issue of law or fact. 10 C.F.R. § 2.309(f)(1)(vi).

d.

Contention 3 is Moot To the extent the Board determines that portions of Contention 3 are admissible, they have been mooted by NextEras submittal of a revised Buried Piping and Tanks Inspection Program in its recent supplement to the LRA.46 The new program was created in response to industry operating experience and the NRCs effort towards the creation of GALL Rev. 2. Where a contention is superseded by the subsequent issuance of licensing-related documentsthe contention must be disposed of or modified.

McGuire/Catawba CLI-02-28, 56 NRC at 382. Where a contention alleges the omission of particular information or an issue from an application, and the information is later supplied by the applicant... the contention is moot. Id. at 383 (footnote omitted).

Many of the specific challenges raised (though not adequately supported) in Contention 3 have been directly addressed in the revised AMP, as discussed below, rendering Contention 3 moot.

46 Due to changes in the GALL Rev. 2, and an oversight in the original application, the new program identifies four additional systems that will be covered by the Buried Piping and Tanks Inspection Program.

See LRA Supplement, Encl. 1 at 7 (adding the instrument air system, auxiliary steam condensate system, auxiliary steam heating system, and feedwater system). Still, no buried pipes in these systems have an intended function related to prevention of leakage of radioactive liquid.

62 In paragraphs 8 and 11, Petitioners assert that the AMP contains no preventative measures and specifically no cathodic protection. Pet. at 25. The revised AMP includes specific discussion of preventive measures available and taken to minimize the effects of aging on buried and underground piping from corrosion, cracking, and changes in material properties including verification of: (1) the effectiveness of the plants cathodic protection system, (2) adequacy of backfill materials, and (3) integrity of coatings and wrappings. LRA Supplement, Encl. 1 at 7-8. Such preventive measures are described as they apply to different materials of construction, including steel, stainless steel, and polymers. Id. at 12-13. Accordingly, Petitioners claim that Seabrooks LRA does not address preventative measures, including cathodic protection, is now moot.

In paragraphs 8 and 9, Petitioners assert that there are no leak tests any more frequently than every 10 years unless, by happenstance (opportunistic), the opportunity to look at a pipe arises for some other reason, and [t]here should be regular and frequent inspections of all components that contain radioactive water. Pet. at 25. Assuming this phrase is referring to inspections as opposed to tests, the revised AMP provides three tables that show the number of inspections required during each 10 year interval during the period of extended operation for buried piping, underground piping, and submerged piping. LRA Supplement, Encl. 1 at 12-13. The tables show that the number of inspections per ten-year period will be based on the material of the pipe, type of material transported through the pipe, and preventative actions taken (adequacy of the backfill, cathodic protection and applied coatings). Id. The number of inspections ranges from one to eight times per ten-year-period based on these variables. Id. If Petitioners did mean to challenge testing frequency, the revised LRA allows for hydrostatic testing, in

63 accordance with 49 C.F.R. Part 195, in place of excavation, on an interval not to exceed five years. Id. at 9. Another new alternative is flow testing of fire mains, which may be performed on an interval not to exceed one year. Id. at 10. Accordingly, Petitioners claim that Seabrooks LRA calls for leak tests no more frequently than every ten years is now moot.

In paragraph 10, Petitioners assert that the AMP is deficient because it does not provide any evaluation of the baseline conditions of buried systems. Pet. at 25. The revised AMP explains that [a]t least one opportunistic or directed (focused) inspection will be performed for each piping material within the scope of this program within 10 years prior to entering the period of extended operation. LRA Supplement, Encl. 1 at 5 (emphasis added). Accordingly, Petitioners claim that Seabrooks LRA does not provide for a baseline evaluation prior to entering the period of extended operation is now moot.

In paragraph 12, Petitioners assert that the LRA makes no commitment to comply with the National Association of Corrosion Engineers (NACE) corrosion control standards. Pet. at 25. However, the revised AMP specifically states that [t]he AWWA C203 specification meets the requirements of NACE SP0169-2007, Standard Practice, Control of External Corrosion on Underground or Submerged Metallic Piping Systems, Table 1, and [t]he cathodic protection system meets the NACE recommendations for pipe-to-soil potential as defined by NACE SP0169-2007, Standard Practice, Control of External Corrosion on Underground or Submerged Metallic Piping Systems. LRA Supplement, Encl. 1 at 7, 8. Accordingly, Petitioners claim that Seabrooks LRA does not commit to comply with NACE corrosion control standards is now moot.

64 In paragraph 13, Petitioners assert that the LRA contains no assurance that the backfill of buried pipes and tanks is consistent with SP0169-2007 section 5.2.3. Pet. at

26. However, the revised AMP states, [b]ackfill is consistent with SP0169-2007 section 5.2.3. LRA Supplement, Encl. 1 at 14. Accordingly, Petitioners claim that Seabrooks LRA is not consistent with SP0169-2007 section 5.2.3 is now moot.

In paragraph 26, Petitioners assert that Seabrooks LRA is deficient because it is limited to external surfaces only. Pet. at 33. However, the revised AMP provides for internal inspection where appropriate in lieu of external excavations. LRA Supplement, Encl. 1 at 11. Accordingly, Petitioners claim that Seabrooks LRA is limited to external surfaces is now moot.

Also in paragraph 26, Petitioners assert that Seabrooks LRA is deficient because it is limited to piping containing iron. Pet. at 33; see also Blanch Decl. at 14 ¶40. While this was factually incorrect with regard to the original AMP, the revised AMP shows clearly that it applies to piping made of various materials, including steel, stainless steel, and polymer. LRA Supplement, Encl. 1 at 12-13. Accordingly, Petitioners claim that the scope of materials covered by Seabrooks LRA is unduly limited is now moot.

In sum, Petitioners have raised issues that are beyond the scope of a license renewal proceeding, contrary to § 2.309(f)(1)(iv), have raised issues that are not material, contrary to § 2.309(f)(1)(iv), have not provided support for their assertions, contrary to

§ 2.309(f)(1)(v), and have not provided sufficient information to show that a genuine dispute exists with regard to a material issue of law or fact, contrary to § 2.309(f)(1)(vi).

Moreover, to the extent Contention 3 raises issues that are not addressed by GALL Rev.

1, it has been mooted by NextEras submittal of an amendment, committing to implement

65 an AMP that is consistent with the recent draft of GALL Rev. 2. Therefore, Petitioners Contention 3 must be dismissed in its entirety.

4.

Contention 4 - Severe Accident Cost is Inadmissible In Contention 4, Petitioners assert:

the Environmental Report is inadequate because it underestimates the true cost of a severe accident at Seabrook Station in violation of 10 C.F.R.

51.53 (C)(3)(II)(L) and further analysis by the applicant is called for.

Pet. at 33-34. Contention 4 is inadmissible for three fundamental reasons.

First, Petitioners fail to demonstrate that it raises a genuine, material dispute.

10 C.F.R. § 2.309(vi). Petitioners argument boils down to their assertion that there are better methods available for determining the offsite dose consequence in the SAMA analysis. But because it is subject to NEPAs rule of reason, the pertinent question for a SAMA analysis is not whether there are plainly better models or whether the analysis can be further refined, but rather whether the selected methodology is reasonable.

Pilgrim, CLI-10-11, 71 NRC at __ (slip op. at 37). Unless it looks genuinely plausible that inclusion of an additional factor or use of other assumptions or models may change the cost-benefit conclusions for the SAMA candidates evaluated, no purpose would be served to further refine the SAMA analysis, whose goal is only to determine what safety enhancements are cost-effective to implement. Id. (slip op. at 39) (emphasis added).

Thus, in order to demonstrate that their concerns raise a genuine, material dispute with the applicant, Petitioners must provide sufficient information to show that, if their proposed refinements were incorporated, it is genuinely plausible that the Seabrook cost-benefit conclusions may change.

66 But Petitioners freely admit that they have not even attempted to meet this standard.

See Pet. at 76 (Petitioners do not offer examples of how the cost benefit equation might have been skewed in favor of no mitigation.). Instead, Petitioners simply assert that [t]he dramatic minimization of costs by NextEra are such that it should be obvious that many SAMAS would be cost effective if the described defects in the analysis were addressed. Id. at 76-77 (emphasis in original); see also Pet. at 72 ([i]t seems clear that a number of additional SAMAs that were previously rejected by the applicants methodology will now become cost effective.). Such unsupported and conclusory speculation is patently insufficient to show that it is genuinely plausible that a cost-benefit conclusion in the SAMA analysis may change. Accordingly, Petitioners Contention 4 does not show that any purpose would be served by further refinement of NextEras SAMA analysis. In short, Petitioners have not shown that the issues they raise are material to the NRCs review (10 C.F.R. § 2.309(f)(1)(iv)) i.e., issues whose resolution would make a difference in the outcome of the licensing proceeding. 54 Fed.

Reg. at 33,172.

Second, to make this necessary showing, Petitioners must rely on allegations of fact or expert opinion. 10 C.F.R. § 2.309(f)(1)(v). The use of probabilistic methodologies such as the MACCS2 code requires substantial technical and specialized expertise. But Petitioners criticisms of the MACCS2 code are not supported by any expert opinion (or by specific references to the technical literature that may contain relevant expert opinion). Petitioners reference a number of studies on severe accident risk, dispersal patterns, and other relevant topics, but those studies are not explained in a manner that supports admission of the contention. Parties must clearly identify evidence

67 on which they rely... with reference to a specific point. The Commission cannot be faulted for not having searched for a needle that may be in a haystack. Public Service Co. of New Hampshire (Seabrook Station, Units 1 and 2), CLI-89-03, 29 NRC 234, 241 (1989). As the Commission recently explained, [m]erely citing to pages in diverse reports without any additional explanation or other obvious link to the SAMA analysis is insufficient to raise a genuine material dispute for hearing. Pilgrim, CLI-10-11, 71 NRC at __n.121 (slip op. at 31). As a result, Petitioners have not provided sufficient information to show that a genuine dispute exists as to NextEras SAMA analysis.

10 C.F.R. § 2.309(f)(1)(vi).

Third, Contention 4 is focused mainly on uncertainty and alleges that greater precision in NextEras SAMA analysis is required. However, in this area, uncertainty is unavoidable and it is inarguable that precise predictions of such complex phenomena are not possible. But NEPA does not call for certainty or precision, but an estimate of anticipated (not unduly speculative) impacts. Louisiana Energy Services, L.P. (National Enrichment Facility), CLI-05-20, 62 NRC 523, 536 (2005) (emphasis added). The SAMA analysis necessarily relies upon modeling and only attempts to simulate the probability of impact risks. ER at F-52. As a result, the Commission explained in Pilgrim that [u]ltimately, NEPA requires the NRC to provide a reasonable mitigation alternatives analysis, containing reasonable estimates, including, where appropriate, full disclosures of any known shortcomings in available methodology, disclosure of incomplete or unavailable information and significant uncertainties, and a reasoned evaluation of whether and to what extent these or other considerations credibly could or would alter the... SAMA analysis conclusions on which SAMAs are cost-beneficial to

68 implement. Entergy Nuclear Generation Co. and Entergy Nuclear Operations, Inc.

(Pilgrim Nuclear Power Station), CLI-10-22, 72 NRC __, __ (Aug. 27, 2010) (slip op. at 9-10).47 Thus, merely claiming that the SAMA analysis contains uncertainty or methodological limitations does not raise a genuine dispute with the application because it has already disclosed and accounted for this uncertainty. Consistent with Pilgrim, NextEras SAMA analysis acknowledges that the inputs to PRA cannot be known with complete certainty, and that there is the possibility that the actual plant risk is greater than the mean values used. ER at F-158. Meanwhile, the GEIS, which the ER is intended to supplement, explains that despite technical advances, large uncertainties in the results of these [probabilistic severe accident risk] analyses remain, including uncertainties associated with the likelihood of the accident sequences and containment failure modes leading to the release categories, the source terms for the release categories, and the estimates of environmental consequences. GEIS at 5-100. Such uncertainties are manifest in modeling the atmospheric transport of radioactivity in gaseous and particulate states and the actual transport, diffusion, and deposition or fallout that would occur during an accident. Id. at 5-101. According to the GEIS, this uncertainty can result in overestimates or underestimates of both early and later effects (health and economic). Id.

47 In making this statement, the Commission relied upon numerous precedents, including: Laguna Greenbelt, Inc. v. United States Dept. of Transp., 42 F.3d 517, 528 (9th Cir. 1994); Lands Council v.

McNair, 537 F.3d 981, 1001-02 (9th Cir. 2008); Communities Inc. v. Busey, 956 F.2d 619, 626 (6th Cir.

1992); Or. Natural Res. Council Fund v. Goodman, 503 F.3d 884, 897 (9th Cir. 2007); Village of Bensonville v. FAA, 457 F.3d 52, 71 (D.C. Cir. 2006); N.J. Dept. of Envtl Prot. v. NRC, 561 F.3d 132, 144 (3d Cir. 2009); Salmon River Concerned Citizens v. Robertson, 32 F.3d 1346, 1358-60 (9th Cir. 1994);

Sierra Club v. United States Dep't. of Transp., 310 F.Supp.2d 1168, 1188 (D. Nev. 2004); San Francisco Baykeeper v. United States Army Corps of Engrs, 219 F.Supp.2d 1001, 1013-1016 (N.D. Calif. 2002)).

69 NextEra performed a number of sensitivity analyses to account for uncertainty both in CDF determination and in atmospheric and evacuation inputs modeling. See ER at F-158-59. The sensitivity analyses of the MACCS2 inputs included varying the annual meteorological data set, release height, release heat, wake effects, evacuation speed, evacuation preparation time, evacuation warning time, fraction of population evacuating, and meteorology in a 40-50 mile ring segment. Id. As a result of these sensitivity analyses, NextEra chose to use in its baseline analysis the meteorological data set that resulted in the maximum dose and cost risk (id. at F-158), and also assumed rainfall within the 40-50 mile ring segment to force a conservatively large deposition and exposure. Id. at 160. NextEra determined that none of the evaluated changes to the other input parameters would increase the accident risk by more than 4%. Id. NextEras ER thus addressed the impacts of these sensitivity analyses on the model results and provided a reasoned evaluation of whether and to what extent these considerations credibly could or would alter the SAMA cost-benefit conclusions. Id.

Petitioners refer to only one of NextEras sensitivity analyses, which accounts for the uncertainty in core damage frequency. See Pet. at 75. But Petitioners do not raise a specific dispute with that analysis, claiming only that it is not convincing. Id. And Petitioners completely ignore the meteorological and evacuation-based sensitivity analyses, thus failing to demonstrate the existence of a genuine dispute with the applicant in accordance with 10 C.F.R. § 2.309(f)(1)(vi), which requires petitioners to identify specific portions of the application with which they disagree.

70

a.

Contention 4A is Inadmissible In Contention 4A, Petitioners assert that:

NextEras use of probabilistic modeling underestimated the deaths, injuries, and economic impact likely from a severe accident by multiplying consequence values, irrespective of their amount, with very low probability numbers, the consequence figures appeared minimal.

Pet. at 37.

Petitioners assertion that a SAMA analysis should focus solely on mitigation of consequences without regard to the likelihood of their occurrence is contrary to the Commissions regulations as well as to the fundamental tenets of SAMA analysis. The NRC performs SAMA analyses in response to a mandate of the Third Circuit in Limerick Ecology Action, 869 F.2d 719. The Limerick Court explained that a SAMA analysis necessarily centers on the evaluation of risk (the likelihood of occurrence times the severity of the consequences). 869 F.2d at 738. Any serious evaluation of the costs and benefits of proposed alternatives to mitigate severe accidents must account for risk. See id. at 738-39. Every SAMA described in the ER is an alternative to mitigate severe accidents, but whether those SAMAs would have any real benefit necessarily requires consideration of risk. See Catawba/McGuire, CLI-02-17, 56 NRC at 9.

A similar challenge to the use of probabilistic modeling was recently brought in the Pilgrim license renewal proceeding. There, the Licensing Board deemed such a challenge to be inadmissible because the use of probabilistic risk assessment and modeling is obviously accepted and standard practice in SAMA analyses. See Pilgrim, LBP-06-23, 64 NRC at 340. The Commission made a similar point at a later stage in that proceeding, explaining that the SAMA analysis assesses whether and to what extent the probability-weighted consequences of the analyzed severe accident sequences would

71 decrease if a specific SAMA were implemented at a particular facility... [and] therefore is a probabilistic risk assessment analysis. Pilgrim, CLI-10-11, 71 NRC at __(slip op.

at 3) (emphases added).

Moreover, Petitioners challenge to the use of probabilistic modeling is inadmissible because it amounts to a challenge to the NRCs generic and probabilistic determination of the environmental impact of severe accidents. This finding is codified as follows:

SMALL. The 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.

However, alternatives to mitigate severe accidents must be considered for all plants that have not considered such alternatives. See § 51.53(c)(3)(ii)(L).

10 C.F.R. Part 51, Subpart A, Appendix B, Table B-1, Issue 76 (emphasis added).

Thus, the SAMA analysis is a mitigation analysis that is meant to complete the NRCs evaluation of severe accident risk, for which the impact determination is pre-existing, probabilistic, and codified by rule. A challenge asserting that this supplemental site-specific mitigation analysis must ignore risk and focus only on consequence necessarily implies that the underlying codified impact analysis improperly considered risk. In fact, Petitioners explicitly make this connection between their challenge and the NRCs probabilistic impact determination. See Pet. at 38, n.6 (contrary to the NRC the impacts of severe accidents are unlikely to be small (emphasis in original). Such a challenge is impermissible. 10 C.F.R. § 2.335(a).

Petitioners also argue that probabilistic methods are inappropriate for any decision regarding adequate protection. Pet. at 40. Leaving aside the merit of this broad

72 claim, it is simply not relevant to NextEras SAMA analysis, which is performed under NEPA and is unrelated to the NRCs obligation under the Atomic Energy Act to assure the adequate protection of the public health and safety. Under the NRCs license renewal rules, the NRCs safety review and findings are limited to managing the effects of age-related degradation and time-limited aging analyses. 10 C.F.R. § 54.29; Turkey Point, CLI-01-17, 54 NRC at 7-8.48 Finally, Petitioners argue that the use of probabilistic analysis is improper for considering intentional malevolent acts. Pet. at 40 (citing Edwin Lyman, Chernobyl on the Hudson? The Health and Economic Impacts of a Terrorist Attack at the Indian Point Nuclear Plant, at 16 (Sept. 2004)). The Commission has considered and rejected the argument that the SAMA analysis must consider intentional acts of terrorism. Pilgrim, CLI-10-14, 71 NRC at __(slip op. at 37). The Commission concluded that NEPA imposes no legal duty on the NRC to consider intentional malevolent acts... in conjunction with commercial power reactor license renewal applications. Id. (citing AmerGen Energy Co. (Oyster Creek Nuclear Generating Station), CLI-07-8, 65 NRC 124, 129 (2007); affd N.J. Dept of Envtl. Prot. v. NRC, 561 F.3d 132, 137-44 (3rd Cir.

2009)). Regardless, in the GEIS, the NRC performed a discretionary analysis of 48 For the same reason, Petitioners miscited references to the Licensing Board decision in the Turkey Point license renewal proceeding (Pet. at 38-39) are inapposite. The contentions being considered in that case were unrelated to SAMA. Florida Power & Light Co. (Turkey Point Nuclear Generating Plant, Units 3 and 4), LBP-01-6, 53 NRC 138, 158-161 (2001). Rather, the decision related to contentions alleging that a PRA should be performed to assess the probability that age-related degradation would cause multiple component failure during a hurricane (id. at 158), that a PRA should be performed to assess whether emergency planning requirements and dose limits would be met (id. at 159), and that an analysis of severe accidents should be performed (without raising any SAMA issue) (id. at 160-61). The Licensing Board found that these types of probabilistic evaluation were not required by the license renewal rules. Id.

at 158-61. Because the petitioner in the Turkey Point case never raised a SAMA contention, the Board never addressed whether a SAMA analysis should be focused on risk. It should also be noted that the Licensing Boards rulings on these contentions were not appealed and were therefore not reviewed by the Commission, contrary to the implications of Petitioners citations. See Turkey Point, CLI-01-17, 54 NRC at 5 (indicating that only one of the petitioners in that proceeding had appealed).

73 terrorist acts in connection with license renewal, and concluded that the core damage and radiological release from such acts would be no worse than the damage and release expected from internally initiated events. Id. (citing Oyster Creek, CLI-07-8, 65 NRC at 131). Thus, Petitioners claims regarding the need to address terrorism in a SAMA analysis do not raise a material issue. 10 C.F.R. § 2.309(f)(1)(iv).

b.

Contention 4B is Inadmissible In Contention 4B, Petitioners assert that:

The SAMA analysis for Seabrook minimizes the potential amount of radioactive release in a severe accident.

Pet. at 41.

Petitioners argue that NextEras SAMA analysis minimized the amount of radioactivity released in a severe accident by: (1) ignoring spent fuel pool accidents; and (2) using a source term to estimate severe accident consequences that is based on radionuclide release fractions generated by the MAAP code. Pet. at 41.

i.

No Mitigation Analysis is Required for Spent Fuel Pool Accidents The Commission has repeatedly held that SAMAs do not encompass spent fuel pool accidents. Pilgrim, CLI-10-11, 71 NRC at __(slip op. at 24); see also Pilgrim, CLI-10-14, 71 NRC at __(slip op. at 30-32); Vermont Yankee/Pilgrim, CLI-07-3, 65 NRC at 19-21; Turkey Point, CLI-01-17, 54 NRC at 21-23.

Part 51 designates the environmental impacts of on-site spent fuel as a Category 1 issue. Pilgrim, CLI-10-14, 71 NRC at __ (slip op. at 30); 10 C.F.R. Part 51, subpart A, Appendix B, Table B-1. This means that the need for mitigation alternatives within the context of [license] renewal... has been considered, and the Commission concludes that its regulatory requirements already in place provide adequate mitigation incentives for

74 on-site storage of spent fuel. Id. (citing GEIS at 6-92). Therefore, it makes obvious sense that no discussion of mitigation alternatives is needed in a license renewal application for a Category 1 issue because for all issues designated as category one the Commission has concluded that [generically] that additional site-specific mitigation alternatives are unlikely to be beneficial.49 Vermont Yankee/Pilgrim, CLI-07-03, 65 NRC at 21 (citing Turkey Point, CLI-01-17, 54 NRC at 21-22).

In the face of this precedent, Petitioners argue, without any factual or expert support, that the offsite cost risk of a pool fire is substantially higher than the offsite cost of a release from a core-damage accident. Pet. at 41. However, the NRC recently reviewed this issue in its consideration of a rulemaking petition and found the risk of beyond design-basis accidents (DBAs) in [spent fuel pools]... to be several orders of magnitude below those involving the reactor core. Attorney General of Massachusetts, Attorney General of California; Denial of Petitions for Rulemaking, 73 Fed. Reg. 46,204, 46,207 (Aug. 8, 2008)) (affd sub nom. New York v. NRC, 589 F.3d 551, 555 (2nd Cir.

2009) (per curiam)). Citing this decision, the Commission stated in Pilgrim that a SAMA that addresses [spent fuel pool] accidents would not be expected to have a significant impact on total risk for the site because the spent fuel pool accident risk level is less than that for a reactor accident. CLI-10-14, 71 NRC at __ (slip op. at 16)

(citing 73 Fed. Reg. at 46,207-08, 46,211-12).

Petitioners also cite a study prepared by Dr. Gordon Thompson, Risks of Pool Storage of Spent Fuel at Pilgrim Nuclear Power Station and Vermont Yankee, A Report 49 The Commission has also considered and rejected Petitioners argument (Pet. at 43-44) that Chapter 6 of the GEIS deals only with normal operations, and not accident conditions. Pilgrim, CLI-10-14, (slip op. at 32-34) (Chapter six clearly is not limited to discussing only normal operations, but also discusses potential accidents and other non-routine events.).

75 for the Massachusetts Attorney General, dated May 2006. Pet. at 42. Petitioners rely on the Thompson study to argue that interactions between the spent fuel pool and the reactor need to be studied in the context of severe accidents. Id. This report was prepared specifically for Vermont Yankee and Pilgrim and includes plant-specific discussions of those reactors spent fuel pools. Petitioners have not shown that it has any bearing on Seabrook. Regardless, both Licensing Boards presented with the Thompson study found that it did not present a litigable issue for the reason discussed aboveSAMAs need not address spent fuel pool accidents. See Pilgrim, LBP-06-23, 64 NRC at 288 and Vermont Yankee, LBP-06-20, 64 NRC 131, 155 (2006); (both affd Vermont Yankee/Pilgrim, CLI-07-03, 65 NRC 13).

ii.

Petitioners Challenge to the MAAP Code is Inadmissible Petitioners also challenge NextEras use of the Modular Accident Analysis Progression (MAAP) code to generate the source term for severe accidents in Contention 4B. Pet. at 44. Petitioners argue that source terms generated by the MAAP code are consistently smaller than source terms generated by NUREG-1465.50 Id.

Petitioners offer no fact or expert opinion in support of this contention, as required by 10 C.F.R. § 2.309(f)(1)(v), only their own unsupported speculation. In fact, this contention is copied almost verbatim from a report presented in support of Riverkeeper EC-2, a contention in the Indian Point license renewal proceeding. See Edwin Lyman, A Critique of the Radiological Consequence Assessment Conducted in Support of the Indian Point Severe Accident Mitigation Alternatives Analysis (Nov. 2007) 50 Source term refers to a fission product release from the reactor core. Dominion Nuclear Connecticut, Inc. (Millstone Nuclear Power Station, Unit 2), CLI-03-14, 58 NRC 207, 209, (2003). It is a result of the magnitude and mix of the radionuclides released from the fuel, their physical and chemical properties, and the timing of their release. Id.

76 (the Lyman Report), attached to Riverkeeper, Inc.s Request for Hearing and Petition to Intervene in Indian Point License Renewal Proceeding (November 30, 2007)

(Riverkeeper Petition) (ADAMS Accession No. ML073410093).51 Dr. Lyman performed his own independent SAMA analysis for Indian Point, using the MACCS2 code to conduct an independent evaluation of severe accident consequences for the highest-impact severe accident scenario. Lyman Decl. at 2. His results indicated that the licensees SAMA analysis underestimated consequences, in part due to the particular source term used. Id.

Here, Petitioners have simply copied the Lyman Report and replaced references to Indian Point with references to Seabrook. Compare Pet. at 44 with Lyman Report at 2-3.

But such an analysis cannot simply be copied and applied to a different plant. It is, on its face, a plant-specific analysis. Without expert opinion or a similar plant-specific SAMA analysis, Petitioners cannot show that their claims, as applied to Seabrook, are based on anything other than their own uninformed speculation. Accordingly, Petitioners assertions regarding the source term used in the NextEra SAMA analysis is not admissible because it does not provide sufficient information to demonstrate the existence of a genuine dispute with the applicant. 10 C.F.R. § 2.309(f)(1)(vi).

Regardless, even with Dr. Lymans ostensibly expert, plant-specific analysis, Riverkeeper Contention EC-2 was not admitted in Indian Point. LBP-08-13, 68 NRC at 185. In that case, the NRC Staff explained that NUREG-1465 addresses only releases into containment and assumes that containment remains intact but leaks. Id. Therefore, the NRC Staff explained and the Board concluded that the NUREG-1465 methodology 51 Dr. Lymans declaration explained that he holds a Ph.D. in physics, had studied nuclear reactor for fifteen years, and was generally familiar with NRC technical studies of accident risks posed by nuclear power plants.

77 does not apply in the scenario in which Petitioners would like to apply it, that of early energetic containment breach. Id. In any case, the NRC Staff also explained that the MAAP code includes the scenario raised by Riverkeeper (and thus here, by Petitioners),

along with other accident scenarios weighted in proportion to their probabilities of occurrence. Id. Accordingly, the Board found the Riverkeeper contention to be inadmissible. Id. As with Indian Point, the Seabrook SAMA analysis also considers a range of accident scenarios, including large early containment failures. See ER at F-32.

Accordingly, Petitioners challenge to the use of MAAP-generated source terms is unsupported and fails to demonstrate the existence of a genuine dispute with the applicant. 10 C.F.R. § 2.309(f)(1)(v, vi).

c.

Contention 4C is Inadmissible In Contention 4C, Petitioners assert that:

The SAMA analysis for Seabrook uses an outdated and inaccurate proxy to perform its SAMA analysis, the MACCS2 computer program.

Pet. at 46.

Petitioners do not raise any specific challenge to NextEras SAMA analysis in Contention 4C. Instead, they make only general and unsupported assertions about the MAACS2 code - that it was not QAd and that it was established for research and not licensing purposes. Pet. at 46. But Petitioners do not explain why quality assurance requirements would apply to a model used for a NEPA analysis. Under Appendix B to Part 50, quality assurance requirements apply to activities affecting the safety-related functions of structures, systems, and components. A SAMA analysis is not safety related

78 and so is not subject to Appendix B. This claim does not raise a genuine, material dispute with the applicant.

Petitioners also state that there is no explanation of exactly how [MACCS2]

works. Pet. at 46. But elsewhere in Contention 4 Petitioners reference and quote from the MACCS2 Users Guide, which explains how it works. See Pet. at 51 n.26, 62.52 Petitioners fail to explain why NextEra would be required to explain how the industry standard code functions when that information is already publicly available. The rest of Contention 4C includes conclusory and unsupported challenges to the MACCS2 code-specifically that its cost formula and assumptions underestimate the costs of a severe accident. See Pet. at 46-7. Without any support for these assertions, they fail to demonstrate the existence of a genuine, material dispute with the applicant. Contention 4C is not admissible.

d.

Contention 4D is Inadmissible In Contention 4D, Petitioners challenge the:

use of an inappropriate air dispersion model, the straight-line Gaussian plume, and meteorological data inputs that did not accurately predict the geographic dispersion and deposition of radionuclides at Seabrooks coastal location.

Pet. at 47. At its core, Contention 4D is an objection to the ATMOS atmospheric dispersion model, which is a module of the MACCS2 code used by NextEra in its SAMA 52 Chanin, D.I.; Young, M.L. (1997 & 1998). Code Manual for MACCS2: Volume 1, User's Guide, SAND97-0594, NUREG/CR-6613, Sandia National Laboratories. As Petitioners have cited this document and put it before the Board, the entire document is subject to scrutiny, both as to those portions that support Petitioners assertions and those that do not. See, e.g., Southern Nuclear Operating Co., (Early Site Permit for Vogtle Site), LBP-07-3, 65 NRC 237, 254 (2007).

79 analysis.53 This general challenge to the adequacy of the MACCS2 code does not provide grounds for admissible contention for two reasons.

First, Petitioners fail to provide adequate support for their assertions. 10 C.F.R.

§§ 2.309(f)(1)(v) and (vi). Petitioners argue that, in determining the geographic concentration of radionuclides released in a severe accident, NextEra used an atmospheric dispersion model that is inappropriate for Seabrooks coastal site. Pet. at 47.

This model, ATMOS, is a steady-state, straight-line Gaussian plume model that is incorporated, or embedded, in the MACCS2 code. Id. at 47-8. Petitioners argue that the plume model underestimated the area likely to be affected in a severe accident and the dose likely to be received in those areas. Id. at 48. Petitioners argue that the use of a straight-line steady-state model cannot account for variations in wind or for sea breezes.

Id. at 49. They seek to require NextEra to re-perform the modeled transport and deposition using a site-appropriate variable plume model such as AERMOD or CALPUFF. Id. at 47.

Petitioners cite to a masters thesis entitled Eastern Massachusetts Sea Breeze Study (Pet. Ex. B, the Sea Breeze Study) for the purpose of showing that the sea breeze is a real phenomenon. But the Sea Breeze Study does not describe the effect of the sea breeze on the dispersal of a radiological plume, much less address how it would affect NextEras SAMA analysis. Petitioners have provided sufficient information to show that the sea breeze is a real phenomenon, but have provided no evidence, no 53 The MACCS2 code contains a meteorological atmospheric dispersion module (called ATMOS) that uses a straight-line Gaussian plume dispersion model. The ATMOS module is used to predict the transport, dispersion, and deposition of radiological material following a severe accident. Other modules in the MACCS2 code (called EARLY and CHRONC) use the ATMOS dispersion modeling results to calculate expected accident consequences (e.g., from radiological doses and land contamination) and complete the SAMA cost-benefit risk analysis. Pilgrim, CLI-10-22, 72 NRC at __, (slip op. at 7); see also MACCS2 Users Guide, NUREG/CR-6613, cited by Petitioners at 51, n.26 and 62.

80 allegations of fact or expert opinion, as to the effect that a sea breeze would have on the cost-benefit conclusions in NextEras SAMA analysis. Petitioners must show that it is genuinely plausible that inclusion of an additional factor or use of other assumptions or models may change the cost-benefit conclusions for the SAMA candidates evaluated.

Pilgrim, CLI-10-11 71 NRC at __ (slip op. at 39). Petitioners have attempted no such showing. Instead, they simply conclude, with no basis or support, that NextEra, by utilizing a straight-line Gaussian plume model, ignores sea breeze and underestimates consequence and is non-conservative. Pet. at 50.

But as the Board stated in the Pilgrim proceeding:

for a fact to be material with regard to the SAMA analysis, it must be a fact which can reasonably be expected to impact the Staffs conclusion that any particular mitigation alternative may (or may not) be cost effective. Mr. Egans vague conclusory statement that the approach used in MACCS2 to modeling changing and uncertain meteorological patterns has caused the Applicant to draw incorrect cost-benefit conclusions fails entirely to address whether the errors he suggests are present would (or even could) cause the results to be less conservative or, in fact, to be non-conservative.

Entergy Nuclear Generation Co. and Entergy Nuclear Operations, Inc. (Pilgrim Nuclear Power Station), LBP-07-13, 66 NRC 131, 152 n.22 (2007) (revd, CLI-10-11, 71 NRC at

__ (slip op. at 18, 21)).54 Here, Petitioners have at least alleged that the modeling shortcomings led to a non-conservative result, but their allegation is unsupported by any expert or review of NextEras SAMA analysis. See Pet. at 50. As the Pilgrim Board implied, and the GEIS stated explicitly, the methodological shortcomings are as likely to 54 The Commission reversed LBP-07-13, finding generously (see Additional Views of Commissioner Svinicki, slip op. at 40) that the petitioners vague expert declaration was sufficient to withstand summary disposition. CLI-10-11, 71 NRC __ (slip op. at 18, 21). Here, Petitioners claims are not supported by any expert opinion whatsoever.

81 result in an overly conservative result. See GEIS at 5-101. Petitioners conclusory (and in this case, entirely unsupported) assertion that the results are not conservative is insufficient to demonstrate the existence of a genuine dispute with the applicant.

Second, Petitioners fail to show that they have raised a genuine, material dispute.

10 C.F.R. § 2.309(f)(1)(vi). NEPA allows agencies to select their own methodology as long as that methodology is reasonable. Pilgrim, CLI-10-11, 71 NRC at __ (slip op. at

37) (citing Town of Winthrop v. FAA, 535 F.3d 1, 13 (1st Cir. 2008)). In LR-ISG-2006-03, the NRC specifically recommends that applicants for license renewal follow the guidance provided in NEI 05-01, Rev. A, when preparing SAMA analyses.55 NEI 05-01, in turn, indicates that use of MACCS2 in an applicants SAMA analysis is acceptable.56 As a result, the Commission has explained that MACCS2 is widely used and accepted as an appropriate tool for conducting SAMA analyses. Pilgrim, CLI-10-11 71 NRC at __ (slip op. at 17); see also Pilgrim, LBP-07-13, 66 NRC 131, 141 (2007) (the development of MACCS2 was sponsored by the NRC and it is the current standard for performing SAMA analysis). Similarly, the GEIS states that MACCS2 is the current, state-of-the-art computer code for assessing risks associated with postulated severe reactor accidents. GEIS at 5-33.

Because the NRC can rely upon its own reasonable methodology, questions such as whether there are plainly better atmospheric dispersion models or whether the SAMA analysis can be refined further are not proper lines of inquiry for a contention 55 Letter to J. Riley (NEI) from P. Kuo (NRC NRR), encl. at 1 (Aug. 2, 2007) (Final License Renewal Interim Staff Guidance LR-ISG-2006-03: Staff Guidance for Preparing Severe Accident Mitigation Alternatives (SAMA) Analyses). (ADAMS Accession No. ML071640133).

56 NEI 05-01, Severe Accident Mitigation Alternatives (SAMA) Analysis Guidance Document, Rev. A, at 13 (Nov. 2005) (ADAMS Accession No. ML060530203).

82 challenging an applicants SAMA analysis. Pilgrim, CLI-10-11, 71 NRC at __ (slip op.

at 37). Nevertheless, the models embedded in the MAACS2 code are not immune from a challenge that is otherwise adequately supported. Id. (slip op. at 17). But in order to proffer an admissible challenge to the straight-line Gaussian plume model in the ATMOS component of the MACCS2 code, a petitioner must show that by altering the model, it is genuinely plausible that the cost-benefit conclusions may be changed. Id. at 39. To make that showing, Petitioners must be able to point to some alternative model capable of performing that task. To that end, Petitioners assert that NextEra should simply replace the ATMOS module in MACCS2 with AERMOD or CALPUFF, which they believe can adequately account for the sea breeze and other environmental factors. Pet. at 47.

However, in a later Pilgrim decision, the Commission identified the difficulty with simply replacing the ATMOS module as Petitioners propose:

Notably, there are practical constraints on the degree to which the meteorological modeling can be altered in the MACCS2 code, which is the most current, established code for NRC SAMA analysis. As Pilgrim Watch states, the straight-line Gaussian plume model is embedded in the MACCS2 code. Therefore, it is not possible simply to plug in and run a different atmospheric dispersion model in the MACCS2 code to see if the SAMA cost-benefit conclusions change. The three modules (ATMOS, EARLY, and CHRONC) in the MACCS2 code are integral parts of the code.

As we earlier emphasized, NEPA requirements are tempered by a practical rule of reason. An environmental impact statement is not intended to be a research document. If relevant or necessary meteorological data or modeling methodology prove to be unavailable, unreliable, inapplicable, or simply not adaptable for evaluating the SAMA analysis cost-benefit conclusions, there may be no way to assess, through mathematical or precise model-to-model comparisons, how alternate meteorological models would change the SAMA analysis results. Some

83 assessments may necessarily be qualitative, based simply on expert opinion.

Pilgrim, CLI-10-22, 72 NRC at __ (slip op. at 9) (emphases added) (footnotes omitted).57 Thus, the straight-line Gaussian ATMOS model cannot simply be replaced without replacing the MACCS2 code itself. Petitioners do not point to any other code capable of replacing the MACCS2 code and so make no showing that use of some other code is reasonable or feasible.

Even assuming that the other models could be plugged in to MACCS2 and that these other models might yield more accurate predictions, Petitioners do not provide adequate information to show that NextEras use of a different code or code component would materially alter the results of its SAMA analysis. Indeed, Petitioners make no reference to the Application, or to any of the specific SAMAs described therein.

Petitioners, therefore, have not met their burden to identify any specific deficiency in the SAMA analysis.

Further, Petitioners provide no explanation of why sea breezes are not already adequately accounted for in the meteorological data used in the Seabrook analysis. As the ER indicates, the meteorological data inputted into and analyzed by MACCS2 included 8,760 hourly recordings of wind direction, wind speed, atmospheric stability, and accumulated precipitation over a year. ER at F-64. Petitioners state that a sea breeze would draw contaminants across the land and inland, penetrating inland 20-40 miles. Pet.

57 Petitioners acknowledge that ATMOS is embedded in MACCS2. Pet at 47.

84 at 49. But the straight-line Gaussian plume model in the MACCS2 code will treat any recorded wind that blows inland as continuing inland.58 In addition to the sea breeze, Petitioners argue that the MAACS2 code is inappropriate because it fails to account for the behavior of plumes over water and for terrain effects (Pet. at 50), and inappropriately uses meteorological inputs from a single station for a single year (Pet. at 53). As discussed below, these claims are not adequately supported to support admission of a contention.

In support of their claims regarding the behavior of plumes over water, Petitioners cite to two unnamed papers ([Zanger et al.; Angevin et al. 2006]) that they neither provide nor discuss.59 Pet. at 50. Instead, in a footnote, Petitioners reference page 11 of a Report to The Massachusetts Attorney General on the Potential Consequences of a Spent-Fuel-Pool Fire at the Pilgrim or Vermont Yankee Nuclear Power Plant, by Dr. Jan Beyea, submitted by petitioner Massachusetts in the Pilgrim and Vermont Yankee proceedings.60 On that page, Dr. Beyea says:

58 That some plumes may initially head out to sea and then be drawn back (Pet. at 49) would simply mean that there would be more time for dispersion before the plume moves inland. Moreover, for every change in direction associated with a sea breeze (winds blowing on shore during the day when the land becomes warmer than the water), there will also be an opposite change in direction associated with a land breeze (winds blowing offshore during the night). See Sea Breeze Study (Pet. Ex. B) at 10. In any event, the frequency of wind blowing inland or out to sea is reflected in the hourly meteorological data used in the Seabrook analysis.

59 These names seem to refer to an article cited in footnote 21. See Pet. at 48. (citing Journal of Applied Meteorology and Climatology 2006; 45: 137-154; Modeling of the Coastal Boundary Layer and Pollutant Transport in New England, Wayne M. Angevine, Michael Tjernstrm and Mark Žagar). But this article addresses pollutant dispersal and does not appear to discuss concentration of plumes. Petitioners do not clearly identify the evidence on which they rely with reference to a specific point, and instead impermissibly seek to require the Board and other parties to search for a needle that may be in a haystack. Seabrook, CLI-89-03, 29 NRC at 241. Merely citing to pages in diverse reports without any additional explanation or other obvious link to the SAMA analysis is insufficient to raise a genuine material dispute for hearing. Pilgrim, CLI-10-11, 71 NRC at __ n.121 (slip op. at 31).

60 Petitioners provide an ADAMS Accession number for Dr. Beyeas declaration, but not the report they cite. The report is available at ADAMS Accession No. ML061640065.

85 I have not been able to incorporate new understanding of the flow of air over and around the New England Coastline that has been achieved in recent years. Still, this new knowledge should be taken into account in EISs for coastal facilities. Releases from Pilgrim headed initially out to sea will remain tightly concentrated due to reduced turbulence until winds blow the puffs back over land (Zagar et al.),

(Angevine et al. 2006). This can lead to hot spots of radioactivity in unexpected locations (Angevine et al.

2004). Dismissing radioactivity blowing out to sea is inappropriate. Reduction of turbulence on transport from Pilgrim across the water to Boston should also be studied.

Although incorporating such meteorological understanding into a PSA or equivalent at Pilgrim would not be likely to make more that a factor of two difference in risk, the change could bring more SAMAs into play and would be significant in an absolute sense, when combined with the increase arising from incorporation of new values of radiation dose conversion coefficients (discussed below).

The program CALPUFF (Scire et al. 2000) has the capability to account for reduced turbulence over ocean water and could be used in sensitivity studies to see how important the phenomenon is at Pilgrim.

Beyea Report at 11 (emphases added).

Thus, Petitioners claims regarding radioactive hot spots are based on a report discussing a different reactor by a claimed expert who had not performed any analysis to show that accounting for his concerns would impact that reactors SAMA analysis cost-benefit conclusions. And Dr. Beyea even admitted that doing so would not lead to large difference in risk, arguing only that such reanalysis could bring more SAMAs into play. This is a far cry from showing that it is genuinely plausible that modifying the Seabrook model would result in a change to NextEras cost-benefit model.61 In fact, the Commission has criticized this particular passage from the Beyea Report, noting that it 61 ER Table F.7-1 shows that increasing the benefit by a factor of two, as Dr. Beyea hypothesizes, would not change the cost-benefit conclusions of any of the SAMAs that were determined to be non-cost beneficial because the benefit of each of those SAMAs is less than half of its cost. See ER at F-128 -

F-157.

86 simply calls for further study. Pilgrim, CLI-10-11, 71 NRC at __ (slip op. at 25, n.97).

Furthermore, Dr. Beyea proposed the use of the CALPUFF atmospheric dispersal model, which the Commission has already explained cannot be simply plugged in to the MACCS2 code. Pilgrim, CLI-10-22, 71 NRC at __ (slip op. at 9).

As to terrain effects, Petitioners argue that a steady-state, straight-line model is inappropriate and cannot account for changes in terrain. Pet. at 51. Petitioners cite to 40 C.F.R. Part 51, Appendix W, Guideline on Air Quality Models to argue that a Gaussian plume model is inappropriate. Id. Petitioners note that Appendix A to Appendix W lists EPAs preferred models and ATMOS, the air dispersion model in the MACCS2 code, is not on the list. Id. But this EPA guidance is not intended to address radiological modeling from a severe reactor accident. Instead it addresses modeling of hazardous pollutants under the Clean Air Act. See 40 C.F.R. Part 51 App. A to App. W at A.0 (1) (This appendix summarizes key features of refined air quality models preferred for specific regulatory applications). Petitioners have not shown why the EPAs need to follow a specific identifiable plume to determine compliance with National Ambient Air Quality Standards would be relevant to NextEras modeling of an assumed radiological plume for a NEPA cost-benefit review, which requires only a reasonable evaluation. Further, Appendix W explains that a preferred model is one that has been found to work better than others, not necessarily that other models are unreasonable. 40 C.F.R. Part 51 App. W at 3.1.1(b). As the Commission stated in Pilgrim, the question is not whether there are plainly better models. CLI-10-11, 71 NRC at __ (slip op. at 37).

87 Petitioners also cite a MACCS2 Guidance Report, which they do not include as an exhibit or provide with a reference link, in their argument on terrain effects.62 Pet. at 52; see also id. at 59. This document explains that the code does not model dispersal less than 100 feet from the source, which Petitioners claim without any support to mean that somehow the resuspension of contaminants is ignored. Id. But cited portion of the MACCS2 Guidance Report makes no such assertion. See MACCS2 Guidance at 3-8.

This section of the MACCS2 Guidance Report does express concern about the wake effects of nearby buildings. Id. NextEra provided a specific sensitivity analysis to account for the impacts of wake effects caused by nearby buildings. ER at F-159. The sensitivity analysis showed that risk is not sensitive to the effects of nearby buildings. Id.

Petitioners ignore this sensitivity analysis in the application, thus failing to demonstrate the existence of a genuine, material dispute.

The cited page of the MACCS2 Guidance does say that the code works best where there is minimal variation in terrain and as a result, there is inherent conservatism (and simplicity) if the environs have significant nearby buildings, tall vegetation, or grade variations not taken into account in the dispersion parameterization. MACCS2 Guidance at 3-8. Thus, the document Petitioners cite to support their terrain effects claim actually shows that the straight-line modeling limitation serves to increase conservatism. Moreover, Petitioners ignore a sensitivity analysis that accounts for variations in release height by considering straight-line releases at ground height and at 25%, 50%, 75%, and 100% (the baseline case) of containment height. See ER at F-159.

62 U.S. Department of Energy, MACCS2 Computer Code Application Guidance for Documented Safety

Analysis, Final Report (June 2004)

(MACCS2 Guidance).

Available at:

http://www.doeal.gov/SWEIS/DOEDocuments/064%20DOE%202004d%20MACCS2.pdf?bcsi_scan_6F88 D07203151B30=0&bcsi_scan_filename=064%20DOE%202004d%20MACCS2.pdf

88 This sensitivity analysis showed that the maximum risk is seen with the maximum height of the release (the baseline case); with the decrease in release height, close-in deposition increased, but the larger population farther away would be affected by a depleted plume.

Id. Petitioners have failed to address this sensitivity analysis, much less show that it did not reasonably account for any variation in terrain. Thus, this claim fails to demonstrate the existence of a genuine, material dispute with the application.

Petitioners also reference a number of NRC regulatory guides and similar documents to suggest that the straight line Gaussian plume model cannot account for complex terrain effects. Pet. at 53-57. But these statements are made in a completely different context than a SAMA analysis and either relate to emergency planning (i.e.

using a model during an actual event to predict deposition from a specific plume under specific meteorological conditions)63 or calculating the maximum exposure of individual at a plants Exclusion Area boundary. None of these documents suggest that MACCS2 cannot be reasonably used to calculate mean, annual consequences for use in a cost benefit analysis.

Finally, Petitioners claim that the meteorological inputs (e.g., wind speed, wind direction, atmospheric stability and mixing heights) NextEra used are based on data it collected at a single, on-site anemometer for a single year, 2005. Pet. at 53. According to Petitioners, measurement data from one station will definitely not suffice to define the sea breeze or capture variability. Id. Once again, Petitioners provide no factual or expert support for their assertion and fail to explain why data collected at NextEras site 63 For instance, RTM-96 (Pet. at 54) explains that its purpose is estimating the possible consequences of different kinds of radiological accidents. The resulting estimates will help officials determine or confirm where to recommend protective actions to the public. These methods should be used only by trained personnel who can interpret the calculations, table, and figures in this document. RTM-96 at pdf page 17 (ADAMS Accession No. ML062560259).

89 weather facility, located near the coast, would not reflect the sea breeze. Accordingly, it cannot provide the basis for an admissible contention. 10 C.F.R. § 2.309(f)(1)(v).

Moreover, NextEras ER explains that it examined five years of data, and chose 2005 because it resulted in the maximum dose and cost risk, thus adding to the conservatism of the analysis. ER at F-64, F-158. Petitioners do not reference or challenge this portion of the application. In any case, the MACCS2 Users Guide (NUREG/CR-6613), cited in both the ER and by Petitioners, explains that a single years worth of meteorological data is what the MACCS2 code accepts. NUREG/CR-6613 at 1-5, 5-31, A-1.

e.

Contention 4E is Inadmissible In Contention 4E, Petitioners challenge the:

use of inputs that minimized and inaccurately reflected the economic consequences of a severe accident, including decontamination costs, cleanup costs and health costs, and that either minimized or ignored a host of other costs.

Pet. at 61.

i.

Decontamination and Cleanup Costs Petitioners argue that in place of MACCS2, NextEra should incorporate, for example, the analytical framework contained in the 1996 Sandia National laboratories report concerning site restoration costs. Pet. at 66. At the outset, this claim lacks specificity. 10 C.F.R. § 2.309(f)(1)(i). Petitioners fail to describe the analytical framework of the Site Restoration Study that they believe should be incorporated, fail to explain the differences between the Site Restoration Study and NextEras SAMA analysis, and fail to explain how that framework should be used to revise NextEras SAMA analysis. See Pet. at 62-67.

90 According to Petitioners, the Sandia Site Restoration Study64 shows that earlier estimates (such as incorporated in WASH-1400 and up through and including MACCS2) of decontamination costs are incorrect because they examined fallout from the explosion of nuclear weapons that produce large particle sizes and high mass loadings. Pet. at 66.

This claim is inadmissible for a number of reasons. First, it is factually incorrect-the MACCS2 code (published in 1997) is actually one year newer than the 1996 Site Restoration Study. Second, the Site Restoration Study only indicates that decontamination data may not be applicable to a plutonium dispersal accident (the subject of the Site Restoration Study) and makes no such assertion with respect to a reactor accident. In fact, it specifically indicates that there is applicable data pertaining to reactor accidents.

Very few experiments have been conducted under conditions that approximate those of the [plutonium dispersal] accidents under consideration. The vast majority of the available data is focused on nuclear explosions or reactor accidents where chemistry, mass loadings, and particle size differs greatly from what would be expected in a plutonium-dispersal accident.

Site Restoration Study at 5-7 (emphasis added).

But more importantly, Petitioners do not explain why NextEra should base its analysis on plutonium dispersal in a nuclear weapons accident. Petitioners point to a number of perceived deficiencies with WASH-1400, which they claim to be the basis of the MACCS2 cleanup cost model. See Pet. at 62-63. But Petitioners do not provide any basis to show that the analytical framework of the Site Restoration Study would be any more appropriate. Instead, they simply state, in conclusory fashion, that [a]lthough there 64 David Chanin, Walt Murfin, Site Restoration: Estimation of Attributable Costs from Plutonium-Dispersal Accidents, SAND96-0957 (May 1996).

91 would be many differences [between a plutonium-dispersal accident and] a nuclear reactor accident, the methodology and conclusions to estimate costs are directly useful.

Id. at 66.

Crucially, Petitioners fail to explain how the information in the Site Restoration Study is relevant, if at all, to the nature and purpose of NextEras SAMA analysis.

Petitioners provide no method for applying the unidentified Site Restoration Study framework to a SAMA analysis and do not explain how the referenced information relates to the specific inputs or assumptions that were entered into the MACCS2 code to evaluate the off-site consequences of a severe accident at Seabrook. But to raise a genuine, material dispute they must show it looks genuinely plausible that inclusion of an additional factor or use of other assumptions or models may change the cost-benefit conclusions for the SAMA candidates evaluated. Pilgrim, CLI-10-11, 71 NRC at __,

(slip op. at 39) (emphasis added). Petitioners fall far short of this standard and fail to even identify what specific assumptions or models should be changed.

For instance, Petitioners quote from page 7-10 of the MACCS2 Users Guide to show that that the MACCS2 code relies upon WASH-1400 for its economic cost model.

Pet. at 62. Petitioners then reference the Site Restoration Study (without citation) to argue that relying on the WASH-1400 framework will underestimate cost. Id. See also id. at 66 (earlier decontamination cost estimates are incorrect because they examined fallout from nuclear weapon explosions that produce large particle sizes and high mass loadings); Site Restoration Study at 2-9. But the Site Restoration Study criticized only a

92 specific portion of WASH-1400, its use of a large decontamination factor (DF).65 The Site Restoration Study explains:

Prior to the 1986 Chernobyl accident, reactor accident risk assessments in the U.S. and Europe relied heavily on the economic cost model of WASH-1400... The use of a DF of 20 in WASH-1400 was apparently based on contemporary guidance documents for anticipated recovery actions following nuclear explosions of warfare. Nuclear explosions produce fallout with large particles and high mass loadings on surfaces. The DF of 20 was widely used in planning documents addressing such events.

Site Restoration Study at 2-9 (emphasis added).

But, contrary to Petitioners allegations, use of the MACCS2 code does not require or imply the use of a DF of 20. The MAACS2 Users Guide explains that the code can accommodate several different decontamination strategies (decontamination levels), each of which would reduce the resulting dose by what the Users Guide calls a dose reduction factor. NUREG/CR-6613 at 7-9. The Users Guide suggests the use of two decontamination levels. Id. at 7-10. And the page immediately following that cited by Petitioners shows the dose reduction factor as an input to the code with suggested values of 3 and 15 for the two decontamination levels. NUREG/CR-6613 at 7-11 (Variable Name: DSRFCT); see also id., App. C, Sample Problem A, at C-32, line item 12.66 Thus, Petitioners have failed to show that the Site Restoration Studys 65 The MACCS2 Users Guide explains that the decontamination factor (DF) (input variable DSRFCT, described in Section 7.5) is a linear scaling factor by which the doses are reduced.

NUREG/CR-6613 at 7-3. This computation is performed to account for decontamination activities that would be taken during the long-term period to reduce doses to acceptable levels. Id. at 7-9. A DF of 20 means that contamination is reduced by a factor of 20 (i.e., 95% of the radioactive material is removed).

Site Restoration Study at 2-9 n.8; see also NUREG/CR-6613 at 7-10, 7-11.

66 As Petitioners have cited this document and put it before the Board, the entire document is subject to scrutiny, both as to those portions that support Petitioners assertions and those that do not. See, e.g.,

Vogtle Early Site Permit, LBP-07-3, 65 NRC at 254.

93 criticism of WASH-1400 has any bearing on the MACCS2 code, much less on NextEras SAMA analysis.

The Commission recently examined a similar contention in Pilgrim and criticized the Petitioners for not drawing a connection between the Site Restoration Study and any perceived shortcoming of the applicants SAMA analysis:

Repeatedly, as we examined Pilgrim Watchs evidence (when it had any) on economic costs, we could not discern any direct connection to the Pilgrim SAMA cost-benefit results. For example, as support for a claim that clean-up costs are underestimated, Pilgrim Watch cites to a page in a Sandia National Laboratories report. See, e.g., Petition for Review at 18; Pilgrim Watch Initial Brief at 12 (citing to SAND96-0957, Site Restoration:

Estimation of Attributable Costs from Plutonium-Dispersal Accidents (May 1996)); see also Pilgrim Watch Initial Brief at 21.

But the cited page merely states that after the Chernobyl accident it became recognized that decontamination of urban areas and particularly porous surfaces can be very difficult, although the acknowledged difficulties of the Chernobyl clean-up may largely have been due to poor training, lack of equipment, and a nearly complete break-down in leadership. Pilgrim Watch provided no specific argument of error in the SAMA cost-benefit analysis calculations or conclusions. Merely citing to pages in diverse reports without any additional explanation or other obvious link to the SAMA analysis is insufficient to raise a genuine material dispute for hearing.

Pilgrim, CLI-10-11, 71 NRC at ___, n.121 (slip op. at 31).

The Commissions criticism in Pilgrim is directly applicable here. Petitioners have cited to numerous studies, including the same Site Restoration Study cited in Pilgrim, but have provided no specific argument of error in NextEras SAMA cost-benefit calculations and so have failed to demonstrate the existence of a genuine, material dispute. 10 C.F.R. § 2.309(f)(1)(vi). Much of Contention 4F simply asserts that the costs of a reactor accident would be very high. See Pet. at 65 (citing studies by Reichmuth and

94 Luna); at 66 (cleanup estimates from the Site Restoration Study of approximately

$300,000,000 per km); and at 67 (severe reactor accident costs could greatly exceed worst case plutonium dispersal accident). But merely asserting that the consequence may be great without also addressing risk, fails to show that Petitioners claims are relevant to the question at hand, whether they have identified specific inputs, assumptions, or models that can be used that would alter the probabilistic cost-benefit conclusions.

Petitioners make several additional arguments that are without any basis or support. For instance, Petitioners claim to know, without explanation, that certain decontamination methods (plowing and fire hosing) would not be allowed by federal and local authorities. Pet. at 64. But as their citation to the Users Guide shows, plowing and fire hosing were evaluated in order to add conservatism to the codeit assumes that the decontamination of farmland using these or similar methods would reduce direct exposure doses to farmers without reducing uptake of radioactivity by root systems. See Pet. at 62. If more stringent decontamination methods were required, the resulting dose would presumably be even lower than estimated in the SAMA analysis. Petitioners also assert that weapons explosions result in non-penetrating radiation, that weapons debris can be easily swept up while contamination from a reactor accident could not, that weapons-related contamination could be shipped to Utah or the Nevada Test Site, and that forests, shorelines, and wetlands cannot be cleaned up. Id. at 63-64. Petitioners provide no technical reference or expert opinion to support any of these assertions.

ii.

Health Costs In the second part of Contention 4F, Petitioners argue that the population dose conversion factor of $2000/person-rem used by NextEra to estimate the cost of the health

95 effects generated by radiation exposure is based on a deeply flawed analysis and seriously underestimates the cost of the health consequences of severe accidents. Pet. at

68. According to Petitioners, use of this conversion factor is inappropriate because it (i) does not take into account the significant loss of life associated with early fatalities from acute radiation exposure that could result from some severe accident scenarios; and (ii) underestimates the generation of stochastic health effects by failing to take into account the fact that some members of the public exposed to radiation after a severe accident will receive doses above the threshold level for application of a dose-and dose-rate reduction effectiveness factor (DDREF). Id.

Once again, Petitioners have based their claim on a contention filed in the Indian Point proceeding, and once again they have deleted all references to the evidentiary support provided in that case. See Riverkeeper Petition at 71-72. As a result, Petitioners assertions again are simply unsupported speculation and inadequate to provide the basis for an admissible contention. 10 C.F.R. § 2.309(f)(1)(v).

With regard to the first factor (early fatalities), Petitioners claim that the

$2000/person-rem conversion factor is intended to represent only stochastic health effects (e.g. cancer), and not deterministic health effects, including early fatalities that result from very high doses to particular individuals. Pet. at 69. Petitioners maintain that, for some of the severe accident scenarios evaluated by NextEra, large numbers of early fatalities could occur, representing a significant fraction of the total number of projected fatalities, both early and latent. Id. But Petitioners provide no support for this statement.

The Riverkeeper contention was supported by the report of an ostensible expert, Dr.

Edwin Lyman, who had reviewed the applicants SAMA analysis and performed his own

96 independent evaluation. See Riverkeeper Petition at 72. His conclusions with respect to Indian Point cannot simply be copied and pasted into a contention addressing a different reactor with a different SAMA analysis. But that is exactly what Petitioners have done here.

See Pet. at 70 (claiming, without explanation, to have estimated the potential number of early fatalities resulting from a severe accident at Seabrook).

As to the second factor (cost conversion factor), Petitioners assert: we estimate that considerable numbers of people would receive doses above the threshold level for application of a DDREF factor of 2. Pet. at 70. Again, Petitioners replaced a reference in the Riverkeeper contention to Dr. Lymans report based on his dose calculations at Indian Point with their own claim to have estimated the dose impacts. Compare Riverkeeper Petition at 72 with Pet. at 70. Petitioners cannot simply claim to estimate dose impacts at Seabrook with no supporting documentation. Instead, they must support their assertions with fact or expert opinion. Here, they have done neither and so this claim is inadmissible.

Based on their estimation, Petitioners conclude that a single cost conversion factor, based on a DDREF of 2, is not appropriate. Pet. at 70. Instead, Petitioners propose that a better way to evaluate the cost equivalent is simply to sum the total number of early fatalities and latent cancer fatalities, as computed by the MACCS2 code, and multiply them by the $3 million figure [the NRC Staffs estimate for the statistical value of a life]. Id. at 71. Contrary to Petitioners assertions, the NRC specifically recommends that license renewal applicants use a $2,000 per person-rem conversion factor in the cost-benefit component of their SAMA analyses. The use of a $2,000 per person-rem conversion factor is consistent with guidance set forth in NEI 05-01, which

97 the NRC endorsed in ISG-LR-2006-03. This guidance has been used by other license renewal applicants with the approval of the NRC.67 That value represents a longstanding NRC regulatory practice and guidance that is not limited to license renewal.68 Petitioners also reference a study from 1982, eight years before Seabrook began licensed operations, which they claim shows that, based on 1970 census data, the number of cancer deaths from a severe accident at Seabrook would be 6,000 with 7,000 early fatalities and 27,000 early injuries. Pet. at 71. But Petitioners do not provide the study or a link to access it. See id. Nor do Petitioners provide any discussion of the relevance of this consequence study to the probabilistic, risk-based SAMA analysis or how this information could be incorporated into the model in such a way as to make a difference in the cost-benefit conclusions.

Confusingly, Petitioners twice claim that NextEras SAMA analysis did not consider cancer incidence (Pet. at 71, 72) immediately after they argued that the SAMA was limited to stochastic effects like cancer. See Pet. at 69. This contradictory claim is unaccompanied by any factual or expert support or references to the SAMA analysis.

Petitioners also argue that the SAMA analysis failed to account for indirect costs such as losses in time and economic productivity and liability. Pet. at 72. Again, Petitioners provide no support for these claims and fail to tie them to specific deficiencies in the application.

67 See NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants, Supplement 42, Regarding Duane Arnold Energy Center (Oct. 2010) at F F-29.

68 The monetary worth of $2000 per person-rem is a standard valuation for comparison purposes recommended by NUREG/BR-0058, Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory Commission (Sept. 2004).

98 Finally, Petitioners argue that NextEras evacuation time input data into the code were unrealistically low and unsubstantiated; and that if correct evacuation times and assumptions regarding evacuation had been used, the analysis would show far fewer will evacuate in a timely manner, increasing health-related costs. Pet. at 72. This allegation ignores the evacuation time estimate sensitivity analyses NextEra performed and described in its ER. See ER at F-159. NextEra provides four different evacuation time sensitivity analyses, each showing that there would be little or minor impact to dose or economic cost. Id.

Evacuation Speed - Baseline case evacuation speed is based on the Seabrook Station Radiological Emergency Plan evaluation considering adverse weather conditions, projected to 2050. Two evacuation sensitivity cases were performed to determine the impact of evacuation speed assumptions. One sensitivity case used one-half the base case evacuation speed and the second sensitivity case doubled the base case evacuation speed. Insight gained:

Dose risk increases as evacuation speed decreases. Change in dose risk not significant.

Evacuation Preparation Time - Baseline case preparation time is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> based on the MACCS2 sample problem A.

Sensitivity cases considered one half the baseline time to prepare for evacuation and a doubling of the baseline time.

Insight gained: Changing the preparation time had a minor effect on most accident category risks; a slightly larger effect was noted on late containment release categories with risk concentrated near the release.

Evacuation Warning Time - Baseline case emergency declaration time is dependent on the accident progression.

Sensitivity cases considered one half the baseline time to warn to evacuate (declaration of general emergency) and a doubling of the baseline time. Insight gained: Similar behavior as changes in evacuation preparation time.

Fraction of Population Evacuating - The baseline case for population evacuation considers 95% percent of the population within 10 miles of the plant evacuating and 5

99 percent not evacuating. This is judged conservative relative to the NUREG 1150 study, which assumed evacuation of 99.5 percent of the population within the emergency planning zone. Release category SE-3 is identified as a risk-dominant release category. An important contributor to SE-3 is a seismically-induced severe accident event. A sensitivity case was performed which conservatively assumed that the population does not evacuate for the SE-3 release category. Insight gained: Assumed no evacuation for release category SE-3 results in a small increase to the overall total accident dose-risk, no change to economic risk.

ER at F-159. Thus, consistent with Pilgrim, NextEra identified evacuation time inputs as a potential area of uncertainty and provided a reasoned evaluation of the impact of that uncertainty on its cost-benefit conclusions. CLI-10-22, 72 NRC at __ (slip op. at 9-10).

Without challenging the evacuation sensitivity analyses in the ER, Petitioners cannot show that their concerns about evacuation speeds or the percentage of the population that would evacuate represent a genuine, material dispute with the application. 10 C.F.R.

§ 2.309(f)(1)(vi).

iii.

Myriad Other Economic Costs In the third part of Contention 4E, Petitioners allege that NextEra failed to include a myriad of other economic costs including the business value of property and costs of job training, unemployment costs, and litigation. Pet. at 73. They also claim that NextEras assumed value of non-farm wealth (which Petitioners do not reference) appeared not justified by Banker and Tradesman sales figures. Id. Petitioners also allege that NextEra underestimated the value of farm property (again, without referencing that value) by ignoring its development value and ignoring the fact that farm assessments are intentionally low. Id. In addition to failing to reference the application (farm and non-farm wealth estimates are provided on page F-58), Petitioners also fail to provide any

100 factual allegations or expert support for their assertions.69 Accordingly, these claims cannot form the basis for an admissible contention.

In summary, Proposed Contention 4E must be dismissed because it lacks adequate factual or expert support and fails to establish a genuine dispute with NextEra on a material issue of law or fact, all contrary to 10 C.F.R. § 2.309(f)(1)(iv), (v), and (vi).

f.

Contention 4F is Inadmissible In Contention 4F, Petitioners challenge the:

Use of inappropriate statistical analysis of the data -

specifically the Applicant chose to follow NRC practice, not NRC regulation, regarding SAMA analyses by using mean consequence values instead of, for example, 95 percentile values.

Pet. at 74.

Petitioners argue that NextEra failed to consider uncertainties in its consequence calculation resulting from meteorological variations by using only mean values for population does and offsite economic cost estimates. Id. Here, Petitioners cite LRA, Appendix E, 2.10. Id. However, the Seabrook SAMA analysis is provided in Attachment F to the ER, which has neither a page 2.10 nor a section 2.10. Thus, Petitioners have failed to refer to specific portions of the application that they dispute, in contravention of 10 C.F.R. § 2.309(f)(1)(vi).

Petitioners claim that NextEras SAMA analysis should be based on the 95th percentile of the risk uncertainty distribution as an appropriate upper confidence bound.

Pet. at 75. But Petitioners then dismiss NextEras consideration of uncertainty without 69 Petitioners claims that the values of farm-and non-farm wealth fail to account for impacts to infrastructure and multiplier effects are similarly unsupported. See Pet. at 67.

101 acknowledging that it accomplishes what Petitioners appear to wantcomparison of the results with the 95th percentile consequence value. The ER explains:

Because the inputs to PRA cannot be known with complete certainty, there is the possibility that the actual plant risk is greater than the mean values used in the evaluation of the SAMA described in the previous sections. To consider this uncertainty, a sensitivity analysis was performed in which an uncertainty factor was applied to the frequencies calculated by the PRA and the subsequent upper bound (UB) benefits were calculated based upon the mean risk values multiplied by this uncertainty factor. The uncertainty factor applied is the ratio of the 95th percentile value of the CDF from the PRA uncertainty analysis to the mean value of the CDF. For Seabrook Station, the 95th percentile value of the CDF is 2.75E-05/yr; therefore, the uncertainty factor is 1.90. Table F.8-1 provides the benefit results from each of the sensitivities for each of the SAMA cases evaluated.

ER at F-158. See also ER at Table F.8-1 (Benefit at UB column). Petitioners argue that absent any specifics this approach at proof is not convincing. Pet. at 75. But this is not an attempt at proofit is an attempt to compare the results using the mean value for core damage frequency to those that would be calculated using the 95th percentile upper bound by using an uncertainty factor derived from a ratio of the mean CDF value to the 95th percentile CDF. Again, consistent with Pilgrim, NextEra identified the use of the mean consequence value as a potential source of uncertainty and provided a reasoned evaluation of the impact of that uncertainty on its cost-benefit conclusions. CLI-10-22, 72 NRC at __ (slip op. at 9-10). Petitioners have not provided any information that would demonstrate the existence of a genuine dispute with NextEras reasoned evaluation of uncertainty. 10 C.F.R. § 2.309(f)(1)(vi).

Once again, this is an issue that was first raised in Indian Point. There, it was accompanied by the Lyman Report, which included a plant-specific SAMA analysis that

102 attempted to show changes to the cost-benefit conclusion that would result by utilizing the 95th percentile value for CDF. See Riverkeeper Petition at 70. The Indian Point Board rejected this contention, because the petitioners made no showing that the applicant failed to meet a regulatory requirement. Indian Point, LBP-08-13, 68 NRC at 183-84.

In fact, there is no requirement for license renewal applicants to base their SAMA analysis upon consequence values at the 95th percentile consequence level that the NRC chose to analyze the environmental impacts of severe accidents in the GEIS. Pilgrim, CLI-10-11, 71 NRC at __ (slip op. at 39). As a NEPA analysis, the SAMA evaluation need not incorporate a worst-case impacts analysis. Id. (slip op. at 38). Such an analysis would distort[] the decisionmaking process by overemphasizing highly speculative harms. Methow Valley, 490 U.S. at 356; see also Private Fuel Storage, L.L.C.

(Independent Spent Fuel Storage Installation) CLI-02-25, 56 NRC 340, 354 (2002)

(analyzing worst-case scenarios is unproductive and unnecessarily uses the agencies limited resources). Instead, NRC practice is to utilize the mean values of the consequence distributions for each postulated release scenario or category[,] which are then multiplied by the estimated frequency of occurrence of specific accident scenarios to determine population dose risk and offsite economic cost risk for each type of accident sequence studied.70 Pilgrim, CLI-10-11, 71 NRC at __ (slip op. at 38-9). A SAMA analysis thus averages potential consequences to generate an expected value. Id.

Petitioners claim in Contention 4F that NextEra must rely on consequence values at the 95th percentile amounts to a claim that NextEra must perform a worst-case analysis, but 70 NEI-05-01 is based on NUREG/BR-0184, Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory Commission, which explains that best estimates should be made in terms of the mean or expected value. NUREG/BR-0184 at 23.

103 because such a claim is not required, Petitioners fail to demonstrate that the issue they seek to litigate is material 10 C.F.R. § 2.309(f)(1)(iv).

In summary, Contention 4 addresses a highly technical issue without any expert support specifically tied to NextEras application. It improperly challenges the use of probabilistic modeling in a SAMA analysis. It also fails to identify what specific changes to NextEras SAMA modeling should be made, fails to show that wholesale substitution of the MACCS2 code or of any of its integral components is possible, and fails to show that any of their generalized concerns with the use of the MACCS2 code necessarily resulted in any reduction in conservatism in NextEras SAMA analysis. Further, Contention 4 does not acknowledge where NextEra has recognized the uncertainty and methodological shortcomings inherent in a SAMA analysis and provided a reasoned evaluation of their impact on the cost-benefit conclusions. For all of these reasons, Petitioners have failed to demonstrate the existence of a genuine, material dispute with the SAMA analysis in NextEras ER and Contention 4 is inadmissible.

5.

Petitioners CLB Challenges are Not Material At the conclusion of the Petition, Petitioners raise two issues related to the CLB.

These claims are not contentions and so need not be addressed by the Board. See 10 C.F.R. § 2.309(a), (f). Regardless, these concerns are unfounded.

First, Petitioners argue that NextEra is not currently managing aging in an adequate manner. Pet. at 78. Petitioners acknowledge that CLB issues are beyond the scope of this proceeding but claim that NextEra may not be adequately managing aging during the period of currently licensed operation, making it difficult to show that it will have the ability to do so during the period of extended operations. Id. Petitioners allege,

104 but do not specify, potential failures related to transformers, safety-related cables, and buried piping and leaks of radioactive liquids. Id. However, Petitioners provide no evidence of any such inadequacies and earlier in their Petition admitted that it was not aware of any such instances at Seabrook, at least with respect to buried piping. Pet. at 26 (It is not the petitioners intent to imply that such events have occurred at Seabrook Station). Regardless, claims of this nature have been specifically removed from the scope of this proceeding:

(a) If the reviews required by § 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, then the licensee shall take measures under its current license, as appropriate, to ensure that the intended function of those systems, structures or components will be maintained in accordance with the CLB throughout the term of its current license.

(b) The licensees compliance with the obligation under Paragraph (a) of this section to take measures under its current license is not within the scope of the license renewal review.

10 C.F.R. § 54.30.

Second, Petitioners complain that the CLB is not currently compiled into one easily accessible location. Id. at 78-9. The Commission considered this issue in the 1995 Final License Renewal rule in response to public comments suggesting that the CLB should be compiled in a manner accessible to the public. 1995 Final Rule, 60 Fed.

Reg. at 22,474. The Commission disagreed with the comment:

The Commission continues to believe that a prescriptive requirement to compile the CLB is not necessary.

Furthermore, submission of documents for the entire CLB is not necessary for the Commission's review of the renewal application.

105 The definition of CLB in §54.3(a) states that a plants CLB

consists, in

part, of a

licensees written commitments... that are docketed... Because these documents have already been submitted to the NRC and are in the docket files for the plant, they are not only available to the NRC for use in the renewal review, they are also available for public inspection and copying in the Commissions public document rooms.

Id. Accordingly, Petitioners suggestion that the NRC should require a compilation of the CLB is contrary to the NRCs reasoned consideration of this issue.

V.

SELECTION OF HEARING PROCEDURES Commission rules require the Atomic Safety and Licensing Board designated to rule on a petition to intervene to determine and identify the specific procedures to be used for the proceeding pursuant to 10 C.F.R. §§ 2.310 (a)-(h). 10 C.F.R. § 2.310. The regulations are explicit that proceedings for the... grant... of licenses subject to

[10 C.F.R. Part 52] may be conducted under the procedures of subpart L. 10 C.F.R.

§ 2.310(a). The regulations permit the presiding officer to use the procedures in 10 C.F.R. Part 2, Subpart G (Subpart G) in certain circumstances. 10 C.F.R.

§ 2.310(d). It is the proponent of the contentions, however, who has the burden of demonstrating by reference to the contention and bases provided and the specific procedures in subpart G of this part, that resolution of the contention necessitates resolution of material issues of fact which may be best determined through the use of the identified procedures. 10 C.F.R. § 2.309(g). Petitioners did not address the selection of hearing procedures in their Petition and, therefore, did not satisfy their burden to demonstrate why Subpart G procedures should be used in this proceeding. Accordingly, any hearing arising from the Petition should be governed by the procedures of Subpart L.

106 VI.

CONCLUSION For all of the foregoing reasons, the Petition should be denied.

Respectfully Submitted, Signed (electronically) by Steven C. Hamrick Mitchell S. Ross Antonio Fernández NextEra Energy Seabrook, LLC 700 Universe Blvd.

Juno Beach, Florida 33408 Telephone: 561-691-7126 Facsimile: 561-691-7135 E-mail: mitch.ross@fpl.com antonio.fernandez@fpl.com Steven C. Hamrick NextEra Energy Seabrook, LLC 801 Pennsylvania Avenue, N.W. Suite 220 Washington, DC 20004 Telephone: 202-349-3496 Facsimile: 202-347-7076 E-mail: steven.hamrick@fpl.com Counsel for NextEra Energy Seabrook, LLC November 15, 2010

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of

)

)

NextEra Energy Seabrook, LLC

)

Docket No.

50-443-LR

)

(Seabrook Station)

)

)

ASLBP No. 0-906-02-LR (Operating License Renewal)

)

CERTIFICATE OF SERVICE I hereby certify that copies of the foregoing NextEra Energy Seabrook LLCs Answer Opposing The Petition to Intervene and Request for Hearing of Friends of the Coast and the New England Coalition, were provided to the Electronic Information Exchange for service to those individuals listed below and others on the service list in this proceeding, this 15th day of November, 2010.

Administrative Judge Paul S. Ryerson, Esq., Chair Atomic Safety and Licensing Board Mail Stop T-3 F23 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Email: psr1@nrc.gov Administrative Judge Dr. Michael Kennedy Atomic Safety and Licensing Board Mail Stop T-3 F23 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Email: michael.kennedy@nrc.gov Administrative Judge Dr. Richard E. Wardwell Atomic Safety and Licensing Board Mail Stop T-3 F23 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Email: richard.wardwell@nrc.gov Secretary Attn: Rulemakings and Adjudications Staff Mail Stop O-16 C1 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 hearingdocket@nrc.gov

2 Office of Commission Appellate Adjudication Mail Stop O-16 C1 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 E-mail: OCAAMAIL@nrc.gov Mary Spencer, Esq.

Catherine E. Kanatas, Esq.

Maxwell C. Smith, Esq.

Emily L. Monteith, Esq.

Megan Wright, Esq.

Office of the General Counsel Mail Stop O-15 D21 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 E-mail: mary.baty@nrc.gov Raymond Shadis New England Coalition Post Office Box 98 Edgecomb, Maine 04556 E-mail: shadis@prexar.com Kurt Ehrenberg New Hampshire Sierra Club 40 N. Main Street Concord, NH 03301 E-mail: Kurt.Ehrenberg@sierraclub.org Paul Gunter, Reactor Oversight Project Beyond Nuclear 6930 Carroll Avenue, Suite 400 Takoma Park, MD 20912 E-mail: paul@beyondnuclear.org Doug Bogen Executive Director Seacoast Anti-Pollution League PO Box 1136 Portsmouth, NH 03802 E-mail: bogen@metrocast.net Signed (electronically) by Steven C. Hamrick Steven C. Hamrick

Exhibit 1 to NextEra's Answer to NEC/Friends of the Coast Petition