ML102940558

2010/10/20-Friends of the Coast and New England Coalition Petition for Leave to Intervene, Request for Hearing, and Admission of Contentions

ML102940558
Person / Time
Site:	Seabrook
Issue date:	10/20/2010
From:	Shadis R New England Coalition
To:	NRC/SECY
SECY RAS
Shared Package
ML102940545	List: ML102940546 ML102940547 ML102940548 ML102940549 ML102940550 ML102940551 ML102940552 ML102940553 ML102940554 ML102940556 ML102940557 ML102940558
References
License Renewal 6, RAS 18957, 50-443-LR
Download: ML102940558 (83)

Text

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE SECRETARY In the Matter of FPL Energy Seabrook, LLC (NextEra, Inc) October 20, 2010 (Seabrook Station, Unit 1 - License Renewal Application)

Docket No. 50-443 FRIENDS OF THE COAST and NEW ENGLAND COALITION PETITION FOR LEAVE TO INTERVENE, REQUEST FOR HEARING, AND ADMISSION OF CONTENTIONS Raymond Shadis Pro se Representative Post Office Box 98 Edgecomb, Maine 04556

TABLE OF CONTENTS INTRODUCTION I. FRIENDS OF THE COAST/NECHAS STANDING.

II. ADMISSION OF CONTENTIONS - APPICABLE LEGAL STANDARDS III. FRIENDS OF THE COAST/NEC SUBMITS ADMISSIBLE CONTENTIONS A. CONTENTION ONE - INACCESSIBLE CABLES Safety Contention Supported by Fact and Expert Testimony B. CONTENTION TWO- TRANSFORMERS Safety Contention Supported by Evidence and Expert Testimony C. CONTENTION THREE - BURIED, BELOW-GROUND, OR HARD-TO-ACCESS PIPING A Safety-related Contention Supported by Evidence and Expert Testimony D. CONTENTION FOUR- SEVERE ACCIDENT COAST UNDERESTIMATED An Environmentally-related Contention Supported by Evidence.

IV. CONCLUSION V. ATTACHMENTS

LIST OF ATTACHMENTS I. DECLARATIONS OF REPRESENTATIVE MEMBERS A. Friends of the Coast Members Attachment 1- Diane Teed- Newburyport, Massachusetts Attachment 2- Peter Kellman- North Berwick, Maine Attachment 3- Deborah Breen- Newburyport, Massachusetts Attachment 4- Sandra Gavutis - Kensington, New Hampshire Attachment 5- Deborah Grinnel - West Newbury, Massachusetts B. New England Coalition Member - Karen Stewart, Rye, New Hampshire II. DECLARATION OF EXPERT WITNESS PAUL M. BLANCH REGARDING CONTENTIONS ONE,TWO, THREE - Declaration of Paul M. Blanch III. ATTACHMENTS TO CONTENTION FOUR Attachment A: Kamiar Jamali, Use of Risk Measures in Design and Licensing Future Reactors, Reliability Engineering and System Safety 95 (2010) 935-943 Attachment B: Thorp, Jennifer E., Eastern Massachusetts Sea Breeze Study, Thesis Submitted to Plymouth State University in Partial Fulfillment of the Requirements for the Degree of Master of Science in Applied Meteorology, May 2009; Colby Jr, F.P.., 2004: Simulation of the New England Sea Breeze Attachment C: Economic Consequences of a Rad/Nuc attack: Cleanup Standards Significantly Affect Cost, Barbara Reichmuth, Steve Short, Tom Wood, Fred Rutz, Debbie Swartz, Pacific Northwest National laboratory, 2005 Attachment D: Survey of Costs Arising From Potential Radionuclide Scattering Events, Robert Luna, Sandia National laboratories, WM2008 Conference, February 24-28, 2008, Phoenix AZ

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE SECRETARY In the Matter of FPL Energy Seabrook, LLC (NextEra, Inc) October 20, 2010 (Seabrook Station, Unit 1 - License Renewal Application)

Docket No. 50-443 FRIENDS OF THE COAST and NEW ENGLAND COALITION PETITION FOR LEAVE TO INTERVENE, REQUEST FOR HEARING, AND ADMISSION OF CONTENTIONS Friends of the Coast - Opposing Nuclear Pollution (Friends of the Coast) and New England Coalition, Inc. (NEC), pursuant to 10 C.F.R. § 2.309 (d) and (e), jointly (herein as, Friends of the Coast/NEC)1, in accord with the terms of the U.S. Nuclear Regulatory Commissions (NRC) 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 2, as amended by Order of the Secretary, submit this petition to intervene in NRCs process of the application of 1

As specified in 10 CFR 2.309, if two or more requestors/petitioners seek to co-sponsor a contention or propose substantially the same contention, the requestors/petitioners must jointly designate a representative who shall have the authority to act for the requestors/petitioners with respect to that contention.

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[Federal Register: July 21, 2010 (Volume 75, Number 139)][Notices] [Page 42462-42464]

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Florida Power and Light, LlC (NextEra Energy, Inc.) to renew its operating license for Seabrook Station for twenty years beyond the current expiration date in 2030.

Friends of the Coast/NEC also requests a hearing under 10 C.F.R. § 2.309(a).

Pursuant to 10 C.F.R. § 2.309(f), Friends of the Coast/NEC should be granted leave to intervene because it has standing, and it herein submits four admissible contentions germane to public health and safety and protection of the natural and human environment.

I. FRIENDS OF THE COAST/NECHAS STANDING.

The standing requirements for Nuclear Regulatory Commission (NRC) adjudicatory proceedings derive from the Atomic Energy Act (AEA), which requires the NRC to provide a hearing "upon the request of any person whose interest may be affected by the proceeding."

42 U.S.C. 2239(a)(1)(A).

Friends of the Coast is a non-profit membership organization incorporated for the public good in the State of Maine since 1995. Friends of the Coast is initially incorporated as Earth Day Commitment, dba/ Friends of the Coast - Opposing Nuclear Pollution.

NEC is a Vermont not-for-profit corporation (formerly the New England Coalition on Nuclear Pollution) incorporated in Vermont since 1971. From its inception NEC's purpose has been, and remains, to oppose nuclear hazards and advocate for sustainable energy alternatives to nuclear power.

Friends of the Coast/NEC has numerous members that reside in the immediate vicinity Seabrook Station and throughout New England; said members concrete and particularized interests will be directly affected by this proceeding. Thus Friends of the Coast/NEC asserts standing and seeks a hearing through representative members whose declarations are 2

attached hereto as Attachments1-6.

A. Friends of the Coast/NEC qualifies for discretionary intervention per 10 C.F.R. § 2.309 (e).

Friends of the Coast/NEC's participation may reasonably be expected to assist in developing a sound record. It is very well versed in the field of nuclear energy and safety.

Indeed, Friends of the Coast/NEC has participated in numerous NRC proceedings. The nature of Friends of the Coast/NEC's interests is not only its members' (and its own) real interests, but the public interest.

Friends of the Coast/NEC and its members can also provide local insight that cannot be provided by the Applicant or other potential parties. Its members are Seabrook stations neighbors. And, as established by this pleading, this proceeding may have significant effects on NEC and its members.

It therefore qualifies for discretionary intervention. 10 C.F.R. § 2.309 (e).

B. Friends of the Coast/NEC has Representational Standing.

As stated above, the purpose of Friends of the Coast/NEC is to oppose nuclear hazards and advocate for sustainable energy alternatives to nuclear power and hence, germane to this proceeding. The attached declarations demonstrate that NEC has numerous members that reside within Seabrook Stations affected vicinity and whose particular interests are directly affected by this matter.

Members providing declarations for the purpose of representational standing 3 include: Diane Teed- Newburyport, Massachusetts; Peter Kellman- North Berwick, Maine; Deborah Breen- Newburyport, Massachusetts; Sandra Gavutis - Kensington, New Hampshire; Deborah Grinnel - West Newbury, Massachusetts; Karen Stewart, Rye, 3 See, Members Declarations, Attachments 1-6.

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New Hampshire Many members live less than fifty miles from Seabrook Station, and subject to radiological contamination, evacuation, loss of property, or other harms in the event of any mishap at the plant. Id. and many members live less than ten miles from Seabrook Station and are within its Emergency Planning Zone making the potential for NextEras next error all that much more palpable.

Members also use and enjoy the segment of the New Hampshire, Maine, and Massachusetts seacoast adjacent to Seabrook Station for social activities, work, recreation, and the gathering of natural provender. Indeed, there exists a tradition of natural resources stewardship in the region that predates our nations founding. The Gulf of Maine; New Hampshire, and Massachusetts littoral and estuarine waters are the receiving waters for any continued thermal or radiologically contaminated discharge.

An organization has standing to sue on behalf of its members when a member would have standing to sue in his or her own right, the interests at issue are germane to the organization's purpose, and participation of the individual is not necessary to the claim or requested relief. Hunt v. Washington State Apple Advertising Comm n, 432 U.S. 333, 343 (1977).

As the Commission has applied this standard, an individual demonstrates an interest in a reactor licensing proceeding sufficient to establish standing by showing that his or her residence is within the geographical area that might be affected by an accidental release of fission products. This "proximity approach" presumes that the elements of standing are satisfied if an individual lives within the zone of possible harm from the source of radioactivity. See Virginia Elec. And Power Co., 9 NRC 54, 56 (1979)("close proximity [to a 4

facility] has always been deemed to be enough, standing alone, to establish the requisite interest" to confer standing). The Commission's "rule of thumb" in reactor licensing proceedings is that "persons who reside or frequent the area within a 50-mile radius of the facility" are presumed to have standing. Sequoyah Fuels Corp., 40 NRC 64, 75 n.22 (1994);

See also, Duke Energy Corp., 48 NRC 381, 385 n.1 (1998).

As is demonstrated by the above discussion and attached declarations, the members represented by Friends of the Coast/NEC would all have standing in their own right. The issues are germane to purposes of Friends of the Coast/ NEC. And, the individual participation of the members is not necessary to the claims or requested relief.

C. Friends of the Coast/NEC Meets Prudential Standing Requirements.

In addition, Courts have created a prudential standing requirement that a plaintiffs interests fall within the "zone of interests" protected by the statute on which the claim is based.

Bennett v. Spear, 520 U.S. 154, 162 (1997). The Atomic Energy Act and NEPA, the statutes at issue here, protect the same interests held by Friends of the Coast/NEC's members and furthered by Friends of the Coast/NEC's purpose.

II. ADMISSION OF CONTENTIONS - APPICABLE LEGAL STANDARDS Proposed contentions must satisfy six requirements of 10 C.F.R. § 2.309(f)(1). This rule is intended to ensure that "full adjudicatory hearings are triggered only by those able to proffer at least some minimal factual and legal foundation in support of their contentions."

Duke Energy Corp. (Oconee Nuclear Station, Units 1, 2, and 3), 49 N.R.C. 328, 334 (1999)(emphasis added). Sections (1) through (6) below summarize the requirements of Section 2.309(f)(1).

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1. Specifically State the Issue of Law or Fact to Be Raised Section 2.309(f)(i) requires "a specific statement of the issue of law or fact to be raised or controverted."

2. Briefly Explain the Basis for the Contention Section 2.309(f)(ii) requires "a brief explanation of the, basis for the contention."

3. Contentions Must Be Within the Scope of the Proceeding Section 2.309(f)(iii) requires petitioner to "demonstrate that the issue raised in the contention is within the scope of the proceeding."

a. Scope of Environmental Review The scope of the NRC's environmental review in the context of a license renewal proceeding is defined by 10 CFR Part 51 and by NRC's "Generic Environmental Impact Statement for License Renewal of Nuclear Plants" (NUREG-1437 (May 1996).

Some environmental issues are resolved generically for all plants, and such issues classified in 10 C.F.R. Part 51, Subpart A, Appendix B as "Category 1" issues are normally beyond the scope of a license renewal hearing. In the Matter of Florida Power & Light Company (Turkey Point Nuclear Generating Plant, Units 3 and 4), 54 NRC 3, 15; 10 CFR § 51.53(c)(3)(i). The remaining issues in Appendix B, which are designated as "Category 2" issues, are issues for which (1) the applicant must make a plant-specific analysis of environmental impacts in its Environmental Report , 10 CRF § 51.53(c)(3)(ii), and (2) the NRC Staff must prepare a supplemental Environmental Impact Statement, 10 C.F.R. § 51.95(c). Contentions concerning Category 2 issues are within the scope of license renewal proceedings. Turkey Point Nuclear Generating Plant, Units 3 and 4, 54 NRC at 11-13.

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b. Scope of Safety/ Aging Management Review 10 C.F.R. 54.4 sets forth the scope of review4 concerning safety issues in a license renewal proceeding.' The safety review "is confined to matters relevant to the extended period of operations requested by the applicant," and focuses on the plant systems, structures, and components "that will require an aging management review for the period of extended operation," or "are subject to an evaluation of time-limited aging analyses." Duke Energy Corp. (McGuire Nuclear Station, Units 1 and 2; Catawba Nuclear Station, Units 1, 2 and 3), 56 N.R.C. 358, 363-64 (2002).

The NRC has emphasized that the level of inspection and testing related to age-management over the extended license term is one of the core issues addressed by the license renewal proceeding.

Part 54 centers the license renewal reviews on the most significant overall safety concern posed by extended reactor operation the detrimental effects of aging. By its very nature, the aging of materials 'becomes important principally during the period of extended operation beyond the initial 40-year license term,' . . . . Adverse aging effects can result from material fatigue, oxidation, erosion, corrosion, . . . and shrinkage. Such age-related 4

This rule reads in relevant part:

§ 54.4 Scope.

(a) Plant systems, structures, and components within the scope of this part are --

(1) Safety-related systems, structures, and components which are those relied upon to remain functional during and following design-basis events (as defined in 10 CFR 50.49 (b)(1)) to ensure the following functions --

The integrity of the reactor coolant pressure boundary; The capability to shut down the reactor and maintain it in a safe shutdown condition; or The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to those referred to in § 50.34(a)(1), § 50.67(b)(2), or § 100.11 of this chapter, as applicable.

(2) All non-safety-related systems, structures, and components whose failure could prevent satisfactory accomplishrnent of any of the functions identified in paragraphs (a)(1)(i), (ii), or (iii) of this section.

(3) All systems, structures, and components relied on in safety analyses or plant evaluations to perform a function that demonstrates compliance with the Commission's regulations for fire protection (10 CFR 50.48),

environmental qualification (10 CFR 50.49), pressurized thermal shock (10 CFR 50.61), anticipated transients without scram (10 CFR 50.62), and station blackout (10 CFR 50.63).

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degradation can affect a number of reactor and auxiliary systems, . . . Indeed, a host of individual components and structures are at issue. See 10 C.F.R. 54.21(a)(1)(i).

Left unmitigated, the effects of aging can overstress equipment, unacceptably reduce safety margins, and lead to the loss of required plant functions, including the capability . . .

to otherwise prevent or mitigate the consequences of accidents with a potential for offsite exposures.

Accordingly, Part 54 requires renewal applicants to demonstrate how their programs will be effective in managing the effects of aging during the proposed period of extended operation. . . . Applicants must identify any additional actions, i.e. maintenance, replacement of parts, etc., that will need to be taken to manage adequately the detrimental effects of aging. Adverse aging effects generally are manifested gradually and thus can be detected by programs that ensure sufficient inspections and testing.

Turkey Point Nuclear Generating Plant, Units 3 and 4, 54 N.R.C. 3, 7-8 (2001)(internal citations omitted).

4. Contentions Must Raise a Material Issue Section 2.309(f)(iv) requires "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."

5. Contentions Must Be Supported by Facts or Expert Opinions Section 2.309(f)(v) requires "a concise statement of the alleged facts or expert opinion which support the petitioner's 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 petitioner intends to rely to support its position on the issue." An intervenor is not required to prove its case at the contention filing stage: "the factual support necessary to show that a 8

genuine dispute exists need not be in affidavit or formal evidentiary form and need not be of the quality as that is necessary to withstand a summary disposition motion." Statement of Policy on Conduct of Adjudicatory Proceedings, 48 N.R.C. 18, 22 n.1 (1998), citing, Rules of Practice for Domestic Licensing Proceedings Procedural Changes in the Hearing Process, Final Rule, 54 F.R. 33168, 33171 (Aug. 11, 1989). Rather, petitioner must make "a minimal showing that the material facts are in dispute, thereby demonstrating that an inquiry in depth is appropriate." In Gulf States Utilities Co., 40 NRC 43, 51 (1994),

citing, Rules of Practice for Domestic Licensing Proceedings Procedural Changes in the Hearing Process, Final Rule, 54 F.R. 33168, 33171 (Aug. 11, 1989).

6. Contentions Must Raise a Genuine Dispute of Material Law or Fact Section 2.309(f)(vi) requires that a petitioner provide sufficient information to show that a genuine dispute exists with the applicant/licensee on a material issue of law or fact. This information must include references to specific portions of the application (including the applicant's environmental report and safety report) that the petitioner disputes and the supporting reasons for each dispute, or, if the petitioner believes that the application fails to contain information on a relevant matter as required by law, the identification of each failure and the supporting reasons for the petitioners belief.

All that is needed is "a minimal showing that the material facts are in dispute, thereby demonstrating that an inquiry in depth is appropriate." In Gulf States Utilities Co.,

40 NRC 43, 51 (1994), citing, Rules of Practice for Domestic Licensing Proceedings Procedural Changes in the Hearing Process, Final Rule, 54 F.R. 33168, 33171 (Aug. 11, 1989).

Because the NRC rules have made adjudicatory hearings in license renewals 9

discretionary, and the requirements the rules place on citizens wishing to have their concerns addressed in a hearing are burdensome, the Licensing Board should take care not to exclude a party who is raising valid and significant safety concerns that are relevant to the renewal.

In Cincinnati Gas & Electric Co. (William H. Zimmer Nuclear Power Station),

ALAB-305, 3 NRC 8, 12 (1976) the Board stated that since a mandatory hearing is not required at the operating license stage, Licensing Boards should take the utmost care to assure that the one good contention rule is met in such a situation because, absent successful intervention, no hearing need be held.

In addition, a Licensing Board should not address the merits of a contention when addressing admissibility. Public Service Co. of New Hampshire (Seabrook Station, Units 1 and 2), LBP-82-106, 16 NRC 1649, 1654 (1982).

The basis for a contention may not be undercut, and the contention thereby excluded, through an attack on the credibility of the expert who provided the basis for the contention. Cleveland Electric Illuminating Co. (Perry Nuclear Power Plant, Units 1 and 2), LBP-82-98, 16 NRC 1459, 1466 (1982).

III. FRIENDS OF THE COAST/NEC SUBMITS ADMISSIBLE CONTENTIONS.

In accordance with the foregoing cited regulations, rulings and criteria, Friends of the Coast/NEC submits the following four contentions:

A. CONTENTION ONE - INACCESSIBLE CABLES A Technical/Safety-related Contention Supported by Fact and Expert Testimony 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 10

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.

BASIS

1. Failure to properly manage aging of Non-environmentally-qualified (Non-EQ)

Inaccessible Medium-Voltage Cables may challenge:

a. the integrity of the reactor coolant pressure boundary;

b. the capability to shut down the reactor and maintain it in a safe shutdown condition; or

c. the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to those referred to in § 50.34(a)(1), § 50.67(b)(2), or § 100.11.

2. The 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 required to meet the requirements of 10 C.F.R. § 54.4.

3. Consequence of failures of Non-EQ Inaccessible Cables may result in accidents beyond the Design Basis Accidents resulting in exposures to the public exceeding 10 C.F.R. § 100 limits.

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4. The Applicant has not demonstrate(d) that the effects of aging will be adequately managed so that the intended function(s) will be maintained consistent with the CLB for the period of extended operation, 10 C.F.R. § 54.21(a)(3) for those SSCs identified for Pressurized Water Reactors in Table 1 of NUREG 1801.

5. The Applicant has failed to identify the location and extent of Non-EQ Inaccessible Cables in use at Seabrook.

6. The Applicant has failed to provide access to referenced documents that are not publicly available (e.g., EPRI TR-103834-P1-2 and EPRI TR-109619). A computer search has been conducted by Friends of the Coast/NEC of all publicly available documents using ADAMS, CITRIX, BRS, GOOGLE and the EPRI web site and the search has not located these referenced documents. It is not possible to fully evaluate the adequacy of the AMP without these references.

The Applicant has failed to provide a copy of its referenced Non-EQ Insulated Cables and Connections Program with the application.

7. The Applicant has failed to address specific recommendations from the referenced Sandia report (SAND96-0344).

8. The Applicant has 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

9. There is no technical basis provided to support life extension using the existing cables without an aging management plan.

10. There is no technical basis provided to justify differences between programs for aging management of accessible cables and inaccessible cables. 10 C.F.R. § 54.21(a)(3).

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11. Inaccessible cables are significantly more likely to fail or experience undetected failures due to submergence and moisture accumulation.

SUPPORTING EVIDENCE

12. The Seabrook Aging Management for medium voltage cables 5 is described on page B-180/181 of the LRA: B.2.1.34 INACCESSIBLE MEDIUM-VOLTAGE CABLES NOT SUBJECT TO 10 CFR 50.49 EQ REQUIREMENTS Program Description The Inaccessible Medium-Voltage Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Program is a new program that will manage the aging effects of localized damage and breakdown of insulation leading to electrical failure of inaccessible medium voltage cables due to adverse localized environments caused by exposure to significant moisture and voltage.

Seabrook Station defines an adverse localized environment for medium voltage cables as exposure to moisture for more than a few days while energized at the system voltage for more than 25 percent of the time. [Emphasis added]

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. [Emphasis added] 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, adjustments in inspection frequency may be required. The maximum time between inspections will be no more than two years. The first inspections will be completed prior to entering the period of extended operation.

In addition to periodic manhole inspections, in-scope, medium-voltage cables exposed to significant moisture and energized at the system voltage for more that 25 percent of the time [Emphasis added]are tested to provide an indication of the condition of the conductor insulation. The specific type of test performed will be determined prior to the initial test, and is 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, Effects of Moisture on the Life of Power Plant Cables or other testing that is state-of-the-art at the time the test is performed. Cable testing will be performed prior to entering the period of extended operation and at least every 10 years thereafter.

Development of this program considers the technical information and guidance provided in the following:

a. NUREG/CR-5643, Insights Gained From Aging Research

b. IEEE Std. P1205, IEEE Guide for Assessing, Monitoring and Mitigating Aging Effects on Class 1E Equipment Used in Nuclear Power Generating Stations 5 FPL/NextEra Seabrook does not propose any APM to manage any cables normally 13

c. SAND96-0344, Aging Management Guidelines for Commercial Nuclear Power Plants - Electrical Cable and Terminations

d. EPRI TR-109619, Guideline for the Management of Adverse Localized Equipment Environments Seabrook Station defines significant moisture as periodic exposures to moisture that last more than a few days (e.g., cable in standing water). Seabrook Station considers periodic exposures to moisture that last less than a few days

[Emphasis added] (i.e., normal rain and drain) as not being significant.

Significant voltage exposure is defined as being subjected to system voltage for more than twenty-five percent [Emphasis added]of the time. The Seabrook Station program includes periodic actions taken to prevent cables from being exposed to significant moisture, such as inspecting for water collection (and draining if needed) in manholes that contain in-scope inaccessible medium-voltage cables. The Seabrook Station program acceptance criteria for the electrical cable test is defined by the specific type of test performed and the specific cable tested. If water is found in manholes, the water will be drained and the inspection frequency will be increased. Unacceptable tests or inspections will be entered into the Corrective Action Program. The corrective action will include an engineering evaluation when the cable testing test acceptance criteria are not met to determine the acceptability of the cable to perform its intended function consistent with the current licensing basis. The evaluation will also consider the significance of the test results, the operability of the component, the reportability of the event, the extent of the concern, the potential root causes for not meeting the test acceptance criteria, the corrective actions required, and the likelihood of recurrence. The corrective action process will include a determination as to whether the same condition or situation is applicable to other inaccessible, in-scope, medium-voltage cables.

13. The use of the words energized at the system voltage for more that 25 percent of the time defies any engineering logic. This statement by itself eliminates most vital cables, such as those supplying power from and to emergency equipment, from any aging management program.

14. Most 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.

15. There 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.

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16. Cables experiencing periodic submergence must be replaced with cables designed to operate in the environment to which they may be exposed. See 10 CFR 50 Appendix A GDC III and IV, Also NUREG-7000.

17. The Seabrook program description is impermissibly vague. NUREG/CR-5643 contains recommendations for detecting degradation of cables; however there are no discussions in the LRA that these recommendations have been addressed. A review of all documents supplied as part of the LRA has failed to identify which cables are encompassed by the AMP. No details are provided explaining the Non-EQ Inaccessible Medium-Voltage Cable Program except that it appears to be limited to . . .

The maximum time between inspections will be no more than two years Experience indicates that not all inaccessible cables are capable of inspection via manholes. This 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. Clearly the LRA AMP has not bounded the problem.

18. NUREGs 1800 and 1801 contain extensive discussions about Aging Management for Electrical Cables and Terminations. The NUREGs reference a study conducted by Sandia National Laboratory (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, and sponsored by the Department of Energy and EPRI. The Sandia study contains numerous recommendations related to the management of aging of cables and terminations with specific emphasis on 10 C.F.R. Part 54 and meeting the requirements of the regulation.

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Section 6 of the Sandia report contains eighteen (18) pages of recommendations and conclusions as to aging management for cables and terminations. The LRA Appendices A and B fail to address or commit to any of the specific recommendations of SAND96-0344. Page 6.4 of the Sandia study states:

No currently available technique was identified as being effective at monitoring the electrical aging of medium-voltage power cable. Some methods may be effective at detecting severe electrical degradation or monitoring certain types of degradation (such as thermal aging); however, correlation of these measurements with the expended or remaining life of these cables has not been demonstrated.

This program includes: A representative sample of accessible insulated cables and connections within the scope of license renewal will be visually inspected for cable and connection jacket surface anomalies such as embrittlement, discoloration, cracking or surface contamination. The technical basis for sampling will be determined using EPRI document TR-109619, Guideline for the Management of Adverse Localized Equipment Environments.

The NRC Staff has recognized the importance of an aging management program for cables, even if they are inaccessible. The U.S. Nuclear Regulatory Commission has issued Generic Letter 2007-01: Inaccessible or Underground Power Cable Failures That Disable Accident Mitigation Systems or Cause Plant Transients (February 7, 2007):

Electrical cables in nuclear power plants are usually located in dry environments, but some cables are exposed to moisture from condensation and wetting in inaccessible locations such as buried conduits, cable trenches, cable troughs, above ground and underground duct banks, underground vaults, and direct-buried installations. The cable insulation goes through gradual degradation due to a variety of reasons.

The Generic Letter emphasizes the potential importance of these cables and evidence that inaccessible cables important to safety have failed even before the end of their useful life:

The design criteria require that cables, which are routed underground, be capable of performing their function when subjected to anticipated environmental conditions such as moisture or flooding. Further, the design should minimize the probability of power interruption when transferring power between sources. The cable failures that could disable risk-significant equipment are expected to have monitoring programs to demonstrate that the cables can perform their safety 16

function when called on. However, the recent industry cable failure data indicates a trend in unanticipated failures of underground/inaccessible cables that are important to safety.

The Generic Letter suggests several procedures that could be implemented to address the degradation of inaccessible cables:

Some licensees have detected cable degradation prior to failures through techniques for measuring and trending the condition of cable insulation.

Licensees can assess the condition of cable insulation with reasonable confidence using one or more of the following testing techniques: partial discharge testing, time domain reflectometry, dissipation factor testing, and very low frequency AC testing.

Licensees can replace faulty cables during scheduled refueling outages prior to cable failure that would challenge plant safety.

None of these measures are included in the Seabrook AMP for Non-EQ Inaccessible Medium-Voltage Cables.

The Staff has documented failures in medium-voltage cables that were inaccessible and for which routine inspections were insufficient to detect the hazard before the cable failed. For example, at Davis-Besse Nuclear Power Station, in determining the root cause of the medium-voltage cable failure, the licensee theorized that:

water in the conduit gradually penetrated the outer neoprene cable jacket, migrated through the cloth binder tape just inside the jacket and through the various layers of the cable construction, and finally penetrated the ethylene propylene rubber (EPR) insulation by osmosis. The water seeping into the cable layers likely contained impurities that precipitated in the outer region of the EPR.

Because the conductor was off-centered, precipitation of these impurities presumably disturbed the electric field in the jacket material. The accompanying observed cracking and conversion of the jacket material to carbon may have released additional impurities that would have further degraded the cable.

Breakdown of the insulation would be most concentrated in the regions of the highest electric field intensity produced by the current in the conductor.

However, this scenario has not been confirmed.

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(NRC Information Notice 2002-12: Submerged Safety-related Electrical Cables )

3/21/02) at 2).

Finally, the GALL Report specifically notes the dangers present in the failure to have an effective AMP for medium voltage cables:

The purpose of the aging management program described herein is to provide reasonable assurance that the intended functions of inaccessible medium-voltage cables that are not subject to the environmental qualification requirements of 10 C.F.R. § 50.49 and are exposed to adverse localized environments caused by moisture while energized will be maintained consistent with the current licensing basis through the period of extended operation. An adverse localized environment is a condition in a limited plant area that is significantly more severe than the specified service environment for the cable. An adverse variation in environment is significant if it could appreciably increase the rate of aging of a component or have an immediate adverse effect on operability.

The GALL program considers the technical information and guidance provided in NUREG/CR-5643, IEEE Std. P1205, SAND96-0344, and EPRI TR-109619. In this aging management program periodic actions are taken to prevent cables from being exposed to significant moisture, such as inspecting for water collection in cable manholes, and draining water, as needed. The above actions are not sufficient to assure that water is not trapped elsewhere in the raceways. For example, if duct bank conduit has low points in the routing, there could be potential for long-term submergence at these low points. In addition, concrete raceways may crack due to soil settling over a long period of time and manhole covers may not be watertight. Additionally, in certain areas, the water table is high in seasonal cycles and therefore, the raceways may get refilled soon after purging.

Furthermore, potential uncertainties exist with water trees even when duct banks are sloped with the intention to minimize water accumulation. Experience has shown that insulation degradation may occur if the cables are exposed to 100 percent relative 18

humidity. The above periodic actions are necessary to minimize the potential for insulation degradation.

In addition to above periodic actions, in-scope, medium-voltage cables exposed to significant moisture and significant voltage are tested to provide an indication of the condition of the conductor insulation. The specific type of test performed will be determined prior to the initial test, and is to be 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.

NUREG-1801Vol. 2, Rev. 1, September 2005, at XI E-7 (emphasis added).

19. None of these additional measures are specifically included in the Seabrook AMP for Non-EQ Inaccessible Medium-Voltage Cables.

20. The NRC has recently issued NUREG/CR-7000 conducted by Brookhaven National Laboratory. This study is best summarized by the Abstract:

For more than 20 years the NRC has sponsored research studying the aging degradation, condition monitoring and environmental qualification, and testing practices for electric cables and cable accessories used in nuclear power plants.

The essential elements for an effective cable condition monitoring program presented in this report are based upon the results of the NRCs electric cable and equipment research programs, industry guidance and standards, and the experience and observations of others who have studied or conducted electric cable condition monitoring and qualification testing. The program methodology presented herein provides guidance on the selection of cables to be included in the program, characterization and monitoring of cable operating environments and stressors, selection of the most effective and practical condition monitoring techniques, documentation and review of cable condition monitoring testing and inspection results, and the periodic review and assessment of cable condition and operating environments.

21. Finally this NRC-sponsored study states:

FOREWORD Electric cables are one of the most important components in a nuclear plant because they provide the power needed to operate safety-related equipment and to transmit signals to and from the various controllers used to perform safety operations in the plant. In spite of their importance, cables typically receive little attention because they are considered passive, long-lived components that have proven to be very reliable over the years.

22. The NRC study concludes with specific recommendations:

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5.2 Recommendations This research study developed recommendations for a comprehensive cable condition monitoring program consisting of nine essential elements that consolidate a core program of periodic CM inspections and tests, together with the results of in-service testing, environmental monitoring and management activities, and the incorporation of cable-related operating experience. The recommended nine essential elements of the cable CM program are listed in Table 5.1 with a summary of the purpose and expected result for each element of the program.

A comprehensive cable condition monitoring program consisting of the nine essential elements listed in Table 5.1 can address the shortcomings of indirectly demonstrating cable integrity and function through in-service testing of systems and components.

23. With respect to adequate assurance of public health and safety and to comply with the above referenced guidance, FPL Energy/NextEra Seabrook must either replace all cables (and splices) that have been exposed to submergence or develop a comprehensive aging management program to preclude moisture and adequately test all cables that have been exposed to an environment for which it was not designed.

B. CONTENTION TWO- TRANSFORMERS A Technical Safety-related Contention Supported by Evidence and Expert Testimony 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 FOR PLANT SAFETY.

BASIS

1. There are numerous electrical transformers that perform a function described in

§§ 54.4(a)(1)/(2) and (3). Transformers function without moving parts or without a change in configuration or properties as defined in that regulation.

2. Failure to properly manage aging of Electrical Transformers may compromise:

a. The integrity of the reactor coolant pressure boundary; 20

b. The capability to shut down the reactor and maintain it in a safe shutdown condition; or

c. The capability to prevent or mitigate the consequences of accidents, which could result in potential offsite exposures comparable to those referred to in §§ 50.34(a)(1), 50.67(b)(2), or § 100.11 of this chapter, as applicable. 10 C.F.R. §§ 54.4(a)(1)(2) and (3).

3. The consequence of failures of Electrical Transformers may result in accidents beyond the Design Basis Accidents resulting in exposures to the public exceeding 10 C.F.R. § 100 limits.

4. Failure to properly manage aging of electrical transformers could result in loss of emergency power to safety equipment and vital busses, including all station blackout loads. Appendix A, Page A-35 of the UFSAR supplement describes a Structures Monitoring Program that includes a program for monitoring transformer/switchyard support structures yet there is no APM described for transformers within the scope of 10 C.F.R. § 54.21(a)(1)(I).

5. The LRA also discusses the need for an AMP for transformer support structures based on the criterion of 10 CFR § 54.4(a)(3).

SUPPORTING EVIDENCE

6. The role of most of the transformers in providing power for safety functions is normally described in Chapter 8 of the UFSAR. The Seabrook LRA provides an FSAR supplement as required by 10 CFR 54.21.

7. While other License renewal applications contained a copy of relevant sections of the UFSAR, Seabrook did not provide a copy and only reference applicable sections of 21

the UFSAR.

8. Without a copy of the UFSAR it is not possible to identify all of the transformers within the scope of 10 CFR 54.4, however it is well known that many transformers perform functions described in 10 CFR 54 and are passive devices in that they contain no moving parts and do not undergo a change of properties or state.

9. Transformers are active devices within the scope of 10 CFR 54.4 yet the licensee has not provided any AMP to assure

10. For purposes of the license renewal rule, the staff has determined that the plant system portion of the offsite power system that is used to connect the plant to the offsite power source should be included within the scope of the rule. This path typically includes switchyard circuit breakers that connect to the offsite system power transformers (startup transformers), the transformers themselves, the intervening overhead or underground circuits between circuit breaker and transformer and transformer and onsite electrical system, and the associated control circuits and structures. Ensuring that the appropriate offsite power system long-lived passive structures and components that are part of this circuit path are subject to an AMR will assure that the bases underlying the SBO requirements are maintained over the period of extended license.

C. CONTENTION THREE - BURIED, BELOW-GROUND, OR HARD-TO-ACCESS PIPING A Technical/Safety-related Contention Supported by Evidence and Expert Testimony 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 22

AND/OR MAY BE IMPORTANT FOR PLANT SAFETY.

BASIS

1. The Aging Management program proposed in the license renewal application for Seabrook is inadequate because: (1) it does not provide for adequate inspection of all systems, structures, and components that may contain or convey water, radioactively-contaminated water, and/or other fluids; (2) there is no adequate leak prevention or detection programs designed to replace such systems, structures, and components before leaks occur; and (3) there is no adequate monitoring to determine if and when leakage from these systems, structures, and components occurs. (4) There is no identification within the LRA of the specific piping systems and tanks covered by this AMP.

2. In order to renew its licenses for another 20 years, 10 C.F.R. § 54.21 requires Seabrook to demonstrate that for each system, structure, and component included within the scope of Part 54.4 the effects of aging will be adequately managed for the period of extended operation. 10 C.F.R. § 54.21 specifically includes "piping" as one of the systems, structures and components included within Part 54. The transfer canal between a reactor and an associated spent fuel pool is another system, structure, or component that falls within Part 54.

3. Pipes perform a critical role in the following systems: (1) safety injection; (2) service water (SW); (3) fire protection; (4) diesel fuel oil; (5) security generator; (6)

ECCS and (7) auxiliary feedwater and other systems within the scope of 10 CFR 54.4.

These pipes and tanks- whether by design or a structural or system failure within the nuclear power station - may contain radioactive water in excess of EPA drinking water limits.

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4. In addition, the refueling water cavities, and spent fuel pool, transfer canals that connect each unit's reactor core with the unit's associated spent fuel pool are included within in the scope of Part 54's systems, structures, and components. See 10 C.F.R. § 54.21(a)(1)(I). These transfer canals contain radioactive water during refuelings.

5. Deficiencies in the Aging Management Plan that encompass the detection of corrosion or leaks in underground buried pipes and tanks, the transfer canals, and essential service water systems could endanger the safety and welfare of the public and are therefore within the scope of a re-licensing hearing. In addition, deficiencies in the Aging Management Plan concerning the detection of leaks or corrosion in other systems, structures, and components containing radioactive water could endanger the safety and welfare of the public and therefore also are within the scope of a re-licensing hearing.

6. Recent events around the United States and the world - as well as at the Seabrook Nuclear Power Station - have demonstrated that various aging piping systems have experienced leaks and/or corrosion. These leaks and corrosion threaten the integrity of such systems and compromise their ability to achieve their intended function. The existence of these leaks demonstrates that aging management of the piping systems is absolutely essential for extended operation of Seabrook.

7. In addition, reports have also confirmed that leaks of underground pipes and tanks can result in the release of significant amounts of radioactive materials into the groundwater or the atmosphere. Exposure to this radiation can threaten human health.

8. Despite the substantial evidence of the dangers of underground leaks from pipes, the LRA fails to include a comprehensive program of leak detection and prevention. Rather, the Applicants aging management program for pipes consists of no 24

preventative measures and 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. There is substantial evidence that such a laissez-faire inspection program will be ineffective at prevention or early detection of leaks from pipes that carry radioactive water or are otherwise important for plant safety.

9. Inspections that might only occur every ten years are insufficient if there is a potential leak of radioactive water from corroded components that could be migrating off-site. "Opportunistic inspections" that might occur no more often than ten years give the appearance that the matter of discovering leaks is being left to chance. There should be regular and frequent inspections of all components that contain radioactive water in this aging plant, including all weld junctures.

10. Seabrooks License Renewal Application and proposed Aging Management Plan are deficient because they do not provide any evaluation of the baseline conditions of buried systems or their many weld junctures, nor do they provide any support for postulated or typical corrosion rates within the facility.

11. The LRA contains no plan or discussion of cathodic protection or other methods to prevent leaks from occurring. Prevention is the best protection against leakage from pipes. 49 CFR 195 provides reasonable requirements for the protection of buries pipes for the transportation industry, yet Seabrook and the NRC have failed to consider the lessons learned from these important requirements to the protect the public from the release of hazardous materials to the environment.

12. Seabrook makes no commitment to comply with the National Association of Corrosion Engineers (NACE) corrosion control standards.

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13. There is no assurance that the backfill of buried pipes and tanks is consistent with SP0169-2007 section 5.2.3.

SUPPORTING EVIDENCE

14. Industry experience plays heavily into maintenance development, regulation, oversight, and decision-making for both regulators and licensees. In licensing proceedings, it is quite common for licensees to invoke industry experience in lieu of performing site-specific analyses. At the reactor oversight level, it is also not uncommon that NRC will chastise a licensee with a failed structure or component for not having heeded industry experience. In the case of buried and hard-to-access pipes and tanks

,over the last decade a series of events, occurring at a quickening pace and with increasing magnitude, have raised serious questions about whether nuclear facilities are in compliance with federal regulations governing the release of radioactive materials into the environment. A number of events have occurred where radioactively contaminated water has leaked into the ground from spent fuel pools, underground pipes and potentially from other systems and components, and remained undetected for as long as 12 years. It is not the petitioners intent to imply that such events have occurred at Seabrook Station, but that based on industry experience Seabrook Station should not gloss over the very real likelihood that similar events will occur at Seabrook with increasing frequency as the plant grows older and more so as the plant enters the extended period of operation. Below is listed a sampling of recent events:

• In August 2004, the owner of the Dresden Nuclear Power Plant in Illinois discovered an underground leak from the condensate storage tank piping. Tritium 26

levels in onsite ground water monitoring wells were as high as 1,700,000 picocuries per liter. A survey of neighboring private wells revealed tritium contamination in at least one well above background levels (approximately 1,000 picocuries per liter). See NRC, Preliminary Listing of Events Involving Tritium Leaks (Mar. 28, 2006), ML060930382.

• In December 2005, tritium was detected in a drinking water well at a home near the Braidwood Nuclear Plant in Illinois. The "initial evaluation indicated that the tritium in the groundwater was a result of past leakage from a pipe which carries normally non-radioactive circulating water discharge to the Kankakee River, about five miles from the site. Several millions [sic] gallons of water leaked from the discharge pipe in 1998 and 2000." See NRC Preliminary Notification of Event or Unusual Occurrence PNO-RIII-05-016A, "Potential Off-site Migration of Tritium Contamination (Update)" (December 7, 2005), ML053410293.

• In March 2006, a leak was discovered at Palo Verde Nuclear Generating Station in Arizona. See NRC Preliminary Notification of Event or Unusual Occurrence, PNO-IV-06-001, "Followup For Tritium Contamination Found In Water Onsite" (March 17, 2006), ML060760584. An analysis of the ground water revealed tritium levels of 71,400 picocuries/Liter (pCi/L). Id.

• The Arizona Republic reported on March 4, 2006 that, "Arizona Public Service Co. discovered radioactive water near a maze of underground pipes at the Palo Verde Nuclear Generating Stationand tests confirmed that the water contains more than three times the acceptable amount of tritium." Radioactive Water Found at Palo Verde, Ken Alltucker, The Arizona Republic (Mar. 4, 2006).

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• In October 2007, high levels of tritium were detected in the groundwater under the Catawba Nuclear Power Station located in York, South Carolina. At one groundwater monitoring well, the tritium measured 42,000 pCi/L. See NRC Preliminary Notification of Event or Unusual Occurrence, PNO-II-07-012, Onsite Groundwater Tritium Contamination" (October 11, 2007), ML073111396.

• That same week, high levels of tritium were discovered in the groundwater at the Quad Cities Nuclear Power Station located in Warrenville, Illinois. The tritium levels measure up to 800,000 picocuries per liter. See NRC Preliminary Notification of Event or Unusual Occurrence, PNO-III-08-011, "Tritium Leakage" (October 11, 2007), ML072890262. "Underground piping from the condensate water storage tank is being examined as a possible 50-286-LR 86

• Seven days later, on October 19, 2007, a leak was discovered in piping within the essential service water system that serviced both reactors at the Byron Nuclear Power Station located in Byron, Illinois. See 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.

• The NRC then announced that had begun a special inspection at the Byron Nuclear Power Station to review the circumstances surrounding the corrosion of piping in the equipment cooling water system and subsequent leak in one pipe.

"As a result of the leakage, reactor operators shut both reactors down on Friday, Oct. 19, to repair the leak and inspect similar pipes. The pipes carry water from the plant where it is used for cooling of essential safety equipment back to basins 28

under fan-driven cooling towers." See NRC Press Release, III-07-24, "NRC Begins Special Inspection at Byron Nuclear Station to Review Corrosion and Leakage of Equipment Cooling Water Pipe" (October 23, 2007), ML072960643.

• Similar leaks have been detected at other nuclear power plants in New Jersey (Salem) and Connecticut (Haddam) as well as the spent fuel pool at the Brookhaven National Laboratory on Long Island. See NRC Office of Nuclear Reactor Regulation, "Spent Fuel Pool Leakage To Onsite Groundwater," NRC Information Notice 2004-05, March 3, 2004 (Salem, New Jersey, Nuclear Power Generating Station); NRC Office of Nuclear Reactor Regulation, "Ground-Water Contamination Due to Undetected Leakage of Radioactive Water," NRC Information Notice 2006-13, July 10, 2006 (discussing leaks at Haddam Neck and other nuclear power plants); Recent tritium leaks at Vermont Yankee AOG System and numerous others. General Accounting Office, Information on the Tritium Leak and Contractor Dismissal at the Brookhaven National Laboratory (GAO/RCED-98-26) November 1997. These NRC and GAO documents are incorporated herein by reference.

• In September 2005, during planned excavation adjacent to the IP2 spent fuel pool, Indian Point discovered cracks in the concrete wall caused by shrinkage during the concrete curing process that leaked spent fuel pool water. Upon further investigation, the licensee determined that groundwater underlying portions of the Indian Point Nuclear Power Station site was contaminated with tritium due to possible leakage from the spent fuel pool or other on-site sources. On February 27, 2006, a sample showed tritium contamination levels of 30,000 pCi/L at a 29

location close to the Hudson River. See Indian Point Nuclear Generating Unit 2 -

NRC Special Inspection Report No. 05000247/2005011 (March 16, 2007)

ML060750842.

• On March 21, 2006, Indian Point announced that samples taken from an on-site monitoring well located near the Hudson River also showed detectable levels of strontium-90; Indian Point also has identified elevated levels of nickel-63 and cesium in groundwater under the Indian Point Nuclear Power Station. See Jim Fitzgerald, High Levels of Strontium-90 Found in Indian Point Groundwater, Associated Press, Mar. 21, 2006; Greg Clary, Indian Point Leak of Radioactive Element Spreads, Poughkeepsie Journal News, Mar. 22, 2006; E-mail from Donald Croulet of Indian Point to James Noggle of USNRC, regarding H-3 sources IPEC-RL-Comments-1" (Dec. 12, 2005), ML061000598.

17.. One common aspect of many of these leaks -- around the nation 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 of contaminated water to enter the ground (or air) around the facilities. See NRC, Liquid Radioactive Release Lessons Learned Task Force Final Report, Sept. 1, 2006, at ii, ML071420239.

18. The older the structure in question, the more likely it is for leakages to occur.

see David Lochbaum, U.S. Nuclear Plants in the 21st Century: the Risk of a Lifetime, Union of Concerned Scientists (May 2004).

To describe the likelihood of aging related problems in nuclear plants, Lochbaum uses the "Bathtub Curve," which was developed by National Aeronautics and Space 30

Administration (NASA) scientists studying statistically the lifetimes of both living and non-living things. Using Reliability-Centered Maintenance As The Foundation For An Efficient And Reliable Overall Maintenance Strategy, NASA (2001).

The curve, which is a graph of failure rate versus age, shows that after a relatively stable (bottom of the bathtub) period in the middle life of the subject, a steep rise in age-related failures occurs towards the end of its life.

The right-hand side of the curve, labeled Region C," is the wear-out phase. Due to aging, it takes less stress to cause failure in this phase. Thus, the chances of failure increase with time spent in Region C. U.S. Nuclear Plants in the 21st Century, at 4.

The renewal period would be this Region C, wear-out phase. "As reactors approach or enter Region C [the wear-out phase] and become more vulnerable to failure, aging management programs monitor the condition of equipment and structures so as to affect repairs or replacements before minimum safety margins are compromised.

Unfortunately, age-related degradation is being found too often by failures than by condition-monitoring activities."

19. This is especially true at Seabrook where the buried systems, structures, and components have been under the ground for 25 years or longer. Under the two-step licensing process, these buried components were installed well before the facility received its operation license.

20. The presence of radioactive fluid in buried pipes or similar systems, structures, or components is significant because a recent study shows that radioactive water carried in underground pipes of an aging plant can speed up corrosion of already-worn pipes.

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Nuclear power plants emit radiation and particles across a range of energies. This radiation can cause accelerated corrosion in critically important parts of the plant, which can lead to efficiency and safety problems.

Gamma rays and neutrons have energies high enough to cause changes in the interior metallic structure resulting in accelerated damage. Consequently these types of radiation and the best alloys to use to mitigate their effects have been extensively researched and their findings applied. Low energy radiation affects metal structures in a different way, but can still cause appreciable and expensive corrosion. Low energy radiation degrades the passive oxide layers that protect metals. Without this protective layer the metals are easily corroded. G. Bellanger, Corrosion Induced by Low Energy Radionuclides: Modeling of Tritium and Its Radiolytic and Decay Products Formed in Nuclear Installations (Elsevier Publications, 2006), ISBN 0 08 0445101. Such structural changes and degradation can be prevented through monitoring and inspection.

21. The Applicant describes the inspection and aging management programs for underground pipes and tanks at Seabrook in Appendix A and B of its License Renewal Application. Appendix A.3.1.5. entitled "Buried Pipes and Tanks Inspection Program" states that "buried components are inspected when excavated during maintenance." See LRA, at A-46.

The LRA also states that if "trending" identifies a susceptible location, the areas with a history of corrosion might have an additional inspection, an alternative coating or a replacement. "Focused inspections" will be performed within 10 years of the license renewal unless an "opportunistic inspection" which allows assessment of pipe condition without excavation, occurs within the ten-year period. See LRA, at A-46.

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25. In the LRA, section B.2.1.22 BURIED PIPING AND TANKS INSPECTION Program Description Seabrook describes its program as follows:

The Seabrook Station Buried Piping and Tanks Inspection Program is a new program that will manage the aging effects of loss of material due to general pitting, crevice, and microbiologically influenced corrosion from the external surfaces of buried steel (including cast iron) and stainless steel components.

Although the program refers to buried tanks as well as piping, Seabrook Station has no buried tanks in scope for license renewal. The systems that contain buried piping that credit this program are the Auxiliary Boiler, Control Building Air Handling, Condensate, Plant Floor Drain, Diesel Generator, Fire Protection and Service Water systems.

26. It appears that this program is limited to external surfaces only and only piping containing iron. Many pipes and tanks contain other structural materials and liners that appear to be exempt to this AMP.

27. Most tanks containing radioactive materials and/or perform functions within the scope of 10 CFR 54.4 are partially buried and not addressed by this particular AMP.

(RWST, CST, Spent Fuel Pool, Waste Tanks etc.)

28. The LRA does not appear to consider the current revision of GALL (NUREG 1800) with respect to pipes and tanks.

D. CONTENTION FOUR- SEVERE ACCIDENT COAST UNDERESTIMATED An Environmentally-related Contention Supported by Evidence.

If a hearing is granted, Friends of the Coast/NEC intends to bring forward expert testimony in support of this contention during succeeding stages of this proceeding. Key aspects of Contention Four are discussed individually as D-A, D-B, D-C, etc., below.

CONTENTION FOUR: THE ENVIRONMENTAL REPORT IS INADEQUATE 33

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.

1. CONTENTION FOUR IS WITHIN THE SCOPE OF THESE PROCEEDINGS Under 10 CFR §2.309, a petitioner is required to show that the issue raised in the contention is within the scope of the proceeding. The National Environmental Policy Act, NEPA, 42 USC § 4332, is the basic charter for protection of the environment. 40 CFR § 1500.1(a). Its fundamental purpose is to help public officials make decisions that are based on understanding of environmental consequences, and take decisions that protect, restore and enhance the environment. 40 CFR § 1500.1(c). The NRC regulations implementing NEPA for Nuclear Plant license renewals are in 10 CFR § 51(c) Operating license renewal stage.

In its application for license renewal of Seabrook, Next Era Energy was required under 10 CFR § 51 to provide an analysis of the impacts on the environment that will result if it is allowed to continue beyond the initial license. The primary method by which NEPA ensures that its mandate is met is the action-forcing requirement for preparation of an EIS. Robertson v. Methow Valley, 490 U.S. at 348-49 (1989).

The environmental impacts that must be considered in an EIS include those which are reasonably foreseeable and have catastrophic consequences, even if their probability of occurrence is low. 40 CFR §1502.22(b)(1). The fact that the likelihood of an impact may not be easily quantifiable is not an excuse for failing to address it in an EIS. NRC regulations require that to the extent that there are important qualitative 34

considerations or factors that cannot be quantified, these considerations or factors will be discussed in qualitative terms. 10 CFR§51.71.

The regulation governing licensing renewals requires the Applicant for renewal to submit an Environmental Report. 10 CFR 51.53(c)(1). The NRC then uses the ER to prepare an EIS or Environmental Assessment, although it has an independent obligation to evaluate and be responsible for the reliability of the information. 10 CFR §51.70.

In a petition for intervention, contentions that seek compliance with NEPA must be based on the applicants Environmental Report (ER). 10 CFR§2.309(f)(2). Under 10 CFR §51 (c)(3)(ii) the plant is required to provide an ER that contains analyses of the environmental impacts of the proposed action, including the impacts of refurbishment activities, if any, associated with license renewal and the impacts of operation during the renewal term for those issues identified as Category 2 issues in Appendix B to subpart A of that part. Under 10 CFR §51(c)(ii)(L) if the staff has not previously considered severe accident mitigation alternatives for the applicant's plant in an environmental impact statement or related supplement or in an environmental assessment, a consideration of alternatives to mitigate severe accidents must be provided. Severe Accidents are a Category 2 issue in Subpart B to Appendix A of section 51, which states 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. Contentions implicating Category 2 issues ordinarily are deemed to be within the scope of license renewal proceedings. See Turkey Point at 11-13. As Seabrook did not consider 35

mitigation alternatives for accidents in the environmental impact statement of its original licensing, this issue is within the scope of this proceeding.

2. THE ISSUE RAISED IN THE CONTENTION IS MATERIAL 10 CFR 2.309(f)(iv) requires that the Petitioner 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. In discussing the materiality requirement, the Atomic Safety and Licensing Board considering the license renewal for Millstone Nuclear Power Station stated In order to be admissible, the regulations require that all contentions assert an issue of law or fact that is material to the outcome of a licensing proceeding; that is, the subject matter of the contention must impact the grant or denial of a pending license application. Where a contention alleges a deficiency or error in the application, the deficiency or error must have some independent health and safety significance. In the Matter of Dominion Nuclear Connecticut, Inc. (Millstone Nuclear Power Station, Units 2 and 3) Docket Nos. 50-336-LR, 50-423-LR ASLBP No. 04-824-01-LR July 28, 2004, p.

7. See Private Fuel Storage, L.L.C. (Independent Spent Fuel Storage Installation), LBP-98-7, 47 NRC 142, 179-80 (1998), affd in part, CLI-98-13, 48 NRC 26 (1998). The deficiency highlighted in this contention has enormous independent health and safety significance. By underestimating the cost of a severe accident in its SAMA analysis Next Era incorrectly discounts possible mitigation alternatives. This could have enormous implications for public health and safety because a potentially cost effective mitigation alternative might not be considered that could prevent or reduce the impacts of that accident. Petitioners allege the Environmental Reports SAMA analysis is deficient and the deficiency could significantly impact health and safety.

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3. THERE IS A SUBSTANTIAL BASIS FOR THE CONTENTION The ER is required to include a consideration of alternatives to mitigate severe accidents (SAMA). 10 C.F.R. 51.53 (c)(3)(ii)(L). That analysis depends on an accurate calculation of the cost of a severe accident in order to have a base line against which to measure proposed mitigation measures. However, NextEras SAMA analysis for Seabrook instead minimized costs likely to be incurred in a severe accident so as mitigation to reduce risk appeared not to be justified.

Each of the following, individually and together with one or more of the others, improperly minimized costs likely to result in a severe accident:

a. 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.

b. Minimization of the potential amount of radioactive material released in a severe accident.

c. Use of an outdated and inaccurate proxy, the MACCS2 computer program, to perform its SAMA analysis.

d. 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.

e. 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.

f. 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.

C-A NEXTERAS USE OF PROBABILISTIC MODELING UNDERESTIMATED THE TRUE CONSEQUENCES OF A SEVERE ACCIDENT 37

BASIS The regulatory requirement that nuclear plants perform a SAMA analysis states: 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 6. Appendix B to Subpart A of 10 CFR §51.53. In other words, even though the probability of a severe accident is so low that the impacts can be considered small, all plants must still consider alternatives to mitigate the consequences of those accidents.

In its ER, Entergy estimated the severe accident risk by using the Probabilistic Safety Analysis (PSA) Model and a Level 3 model developed by the MACCS2 code.

Using this method, the application states that Risk is defined as the product of consequence and frequency of accidental release. Application ER E.4.20. In using the PSA Model to estimate risk, Seabrook was following standard NRC and industry practice. However practice is not a regulation.

In the license renewal proceeding for Turkey Point, the board used the following interpretation of the regulations to dismiss the Petitioners concerns about particular severe accidents. It stated, . . . the commissions environmental regulations in 10 C.F.R. Part 51 do not require probabilistic risk assessments. Section 51.53(c) lists the information the Applicant must include in its environmental report, and a probabilistic risk analysis of multiple failures is not specified. Likewise sections 51.71(d) and 51.95(c) set forth the requirements the agency must follow in preparing the draft and final SEIS 6 Petitioners contend, contrary to NRC, that the societal and economic impacts from severe accidents are unlikely to be small for all plants and simply appear so by the use of methods that minimized consequences as set forth in this Motion.

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for the Turkey Point license renewal, and nowhere do those provisions require the preparation of a probabilistic risk analysis of multiple failures. Turkey Point, supra at 23-24. It went on to say, . . . section 51.53(c) does not require the Applicant broadly to consider severe accident risks. Rather, it only requires the Applicant to consider severe accident mitigation alternatives (SAMA). 10 C.F.R. § 51.53(c)(3)(ii)(L). Id. at 26.

While in that instance the licensing board used this argument to reject the Petitioners contention related to Emergency Preparedness, the boards reading of the regulatory requirement is also instructive here. It would make no sense for the NRC to require Severe Accident Mitigation Analysis if an Applicant could simply multiply all consequences of an accident by extremely low probability and thus reject all possible mitigation as too costly.

It is widely recognized that probabilistic modeling can underestimate the deaths, injuries, and economic impact likely from a severe accident. By multiplying high consequence values with low probability numbers, the consequence figures appear far less startling. For example a release that would cause 100,000 cancer fatalities would only appear to cause 1 cancer fatality per year if the associated probability of the release were 1/100,000 per year. This issue was central to a New York case, Indian Point Special Proceeding, US Nuclear Regulatory Commission, Atomic Safety and Licensing Board, Recommendations to the Commission, October 24, 1983, p. 107. Before the proceeding, the NRC ruled that all testimony on accident consequences must also contain a discussion of accident probabilities. In its decision, the three-judge ASLB panel concluded that the Commission should not ignore the potential consequences of severe-consequence accidents by always multiplying those consequences by low probability values. Further 39

Kamiar Jamali (DOE Project Manager for Code Manual for MACCS2) Use of Risk Measures in Design and Licensing Future Reactors in Reliability Engineering and System Safety 95 (2010) 935-943 (Attachment A) makes clear that PRA uncertainties are so large and so unknowable that it is a huge mistake to use a single number coming from them for any decision regarding adequate protection. Examples of these uncertainties include probabilistic quantification of single and common-cause hardware or software failures, occurrence of certain physical phenomena, human errors of omission and commission, magnitudes of source terms, radionuclide release and transport, atmospheric dispersion, biological effects of radiation, dose calculations, and many others. (Jamali, Pg., 935)

In addition, in his report on the likely consequences of an accident at the Indian Point Nuclear Plant, Dr. Edwin S. Lyman (Union of Concerned Scientists, Senior Scientist) stresses that intentional acts represent a class of accidents that should not be considered using probabilistic modeling. Accident probabilities are not relevant for scenarios that are intentionally caused by sabotage. Severe releases resulting from the simultaneous failure of multiple safety systems, while very unlikely if left up to chance, are precisely the outcomes sought by terrorists seeking to maximize the impact of their attack. Thus the most unlikely accident sequences may well be the most likely sabotage sequences. Edwin S. Lyman, PhD, Chernobyl on the Hudson? The Health and Economic Impacts of a Terrorist Attack at the Indian Point Nuclear Plant, Union of Concerned Scientists, p. 16 (September, 2004) 7. NextEra failed to model intentional acts 7 Available on internet at:

http://www.ucsusa.org/nuclear_power/nuclear_power_risk/sabotage_and_attacks_on_reactors/impacts-of-a-terrorist-attack.html 40

in its analysis of external events. ER E.F.2. 8 C-B THE SAMA ANALYSIS FOR SEABROOK MINIMIZES THE POTENTIAL AMOUNT OF RADIOACTIVE RELEASE IN A SEVERE ACCIDENT BASIS The SAMA analysis for Seabrook minimized the potential amount of radioactive releases in a potential severe accident at Seabrook Station by: (1) not considering a severe accident in the spent fuel pool, either alone or in combination with a reactor core accident; and (2) using a source term to estimate the consequences of the most severe accidents with early containment failure based on radionuclide release fractions generated by the MAAP code (a proprietary industry code that has not been validated by NRC), which are smaller for key radionuclides than the release fractions specified in NRC guidance such as NUREG-1465 and its recent reevaluation for high-burnup fuel. 9 Therefore the source term used by NextEra results in lower consequences than would be obtained from NUREG-1465 release fractions and release durations.

1. SAMA Analysis of Spent Fuel Risks Is Required By NRC Regulations NextEra did not consider a severe accident from the spent fuel pool at Seabrook resulting from either human error, mechanical failure or an act of malice, although such accidents are reasonably foreseeable. The offsite cost risk of a pool fire is substantially higher than the offsite cost of a release from a core-damage accident. Further, SAMAs designed to avoid or mitigate conventional accidents may be different than SAMAs 8 Application ER E.F2 The external hazards evaluated are internal fires, external floods, and seismic events only.

9 L. Soffer, et al. U.S. Nuclear Regulatory Commission, Accident Source Terms for Light-Water Nuclear Power Plants: Final Report, NUREG-1465, February 1995; Energy Research, Inc., Accident Source Terms for Light-Water Nuclear Power Plants: High-Burnup and MOX Fuels: Final Report, ERI/NRC 02-202, November 2002.

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designed to avoid or mitigate spent fuel accidents. Moreover, the radiological consequences of a spent-fuel-pool fire are significantly different from the consequences of a core-damage accident.

Further NextEra did not consider the potential interactions between the pool and the reactor in the context of severe accidents at Seabrook. There the spent-fuel pool is located outside but immediately adjacent to the reactors containment and shares some essential support systems with the reactor. There could be at least three types of interactions between the pool and reactor. 10 First, a pool fire and a core-damage accident could occur together, with a common cause. For example, a severe earthquake could cause leakage of water from the pool, while also damaging the reactor and its supporting systems to such an extent that a core-damage accident occurs. Second, the high radiation field produced by a pool fire could initiate or exacerbate an accident at the reactor by precluding the presence and functioning of operating personnel. Third, the high radiation field produced by a core-damage accident could initiate or exacerbate a pool fire, again by precluding the presence and functioning of operating personnel. Many core-damage sequences would involve the interruption of cooling to the pool, which would call for the presence of personnel to provide makeup water or spray cooling of exposed fuel. The third type of interaction was considered in a license-amendment proceeding in regard to expansion of spent-fuel-pool capacity at the Harris nuclear power plant. Such accidents are conceivable and would result in a very high magnitude of release.

Although 10 C.F.R. § 51.53(c)(3)(ii)(L), does not provide a definition of severe 10 Dr. Gordon Thompson, Risks of Pool Storage of Spent Fuel at Pilgrim Nuclear Power Station and Vermont Yankee, A Report for the Massachusetts Attorney General by IRSS, May 2006, Pgs., 12, 16. NRC Electronic Library, Adams Accession Number ML061630088 42

accidents the GEIS 11 which provides the factual background for the SAMA requirement in the regulations, does define a severe accident. According to Section 5.2.1 of NUREG 1437 General Characteristics of Accidents, the term accident refers to any unintentional event outside the normal plant operational envelope that results in a release or the potential for release of radioactive materials into the environment and severe

[includes] 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 . . . (emphasis added). This section recognizes the potential for a severe accident in which there are releases substantially in excess of permissible limits for normal operation. 12 Section 5 focuses on potential consequences to determine whether or not a potential accident is severe - and thus within the scope of a Severe Accident Mitigation Analysis. The question is not whether the source of the Severe Accident is the first or second largest inventory of radioactive materials. 13 Perhaps NextEra confused Section 6 of the GEIS with Section 5. Section 6 deals with normal operations (see, for example, 11 See NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants (May 1960)

[hereinafter GEIS]; Final Rule, Environmental Review for Renewal of Nuclear Power Plant Operating

Licenses, 61 Fed., Reg., 28, 467 (June 5, 1960, amended by 61 Fed. Reg. 66, 537 (Dec. 18, 1996); 10 C.F.R. Pt. 51, Subpart A, Appendix B n.1) 12 The term "accident" refers to any unintentional event outside the normal plant operational envelope that results in a release or the potential for release of radioactive materials into the environment. Generally, the U.S. Nuclear Regulatory Commission (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), for which plants are analyzed to determine their response. The predominant focus in environmental assessments is on events that can lead to releases substantially in excess of permissible limits for normal operation. Normal release limits are specified in the NRC's regulations (10 C.F.R. Part 20 and 10 C.F.R. Part 50, Appendix A). GEIS, 5.2.1, Italics added 13 Due to 40 years of operations, the inventory of radioactive materials in Seabrooks spent fuel pool will be many times over that in its reactor core. The statement in sub-section 5.2.1.1 that the spent fuel storage pool contains the second largest inventory after the reactor core does not now nor will it apply to Seabrook.

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section 6.1: Accidental releases could conceivable result in releases that would cause moderate or large radiological impacts. Such conditions are beyond the scope of regulations controlling normal operations. (Emphasis added).

Section 5, not Section 6, deals with severe accidents. Nothing in Section 5 excludes severe accidents involving what at Seabrook Station is the largest inventory of radioactive materials - the spent fuel pool.

2. Source Terms Used By NextEra to Estimate the Consequences of Severe Accidents The source terms used by NextEra to estimate the consequences of severe accidents (radionuclide release fractions generated by the Modular Accident Analysis Progression, MAAP 14) code, has not been validated by NRC. They are consistently smaller for key radionuclides than the release fractions specified in NUREG-1465 and its recent revision for high-burnup fuel. The source term used results in lower consequences than would be obtained from NUREG-1465 release fractions and release durations.

It has been previously observed that MAAP generates lower release fractions than those derived and used by NRC in studies such as NUREG-1150. A Brookhaven National Laboratory study that independently analyzed the costs and benefits of one SAMA in the license renewal application for the Catawba and McGuire plants noted that the collective dose results reported by the applicant for early failures seemed less by a factor between 3 and 4 than those found for NUREG-1150 early failures for comparable scenarios. The difference in health risk was then traced to differences between [the applicants definitions of the early failure release classes] and the release classes from NUREG-1150 for comparable scenarios the NUREG-1150 release fractions for the important radionuclides are about a factor of 4 higher than the ones used in the Duke PRA. The Duke results were obtained using the Modular Accident Analysis Package (MAAP) 14 See, for example, ER. E. F-32, F-45-48 44

code, while the NUREG-1150 results were obtained with the Source Term Code Package [NRCs state-of-the-art methodology for source term analysis at the time of NUREG-1150] and MELCOR. Apparently the differences in the release fractions are primarily attributable to the use of the different codes in the two 15 analyses.

Thus the use of source terms generated by MAAP, a proprietary industry code that has not been independently validated by NRC, appears to lead to anomalously low consequences when compared to source terms generated by NRC staff. In fact, NRC has been aware of this discrepancy for at least two decades. In the draft Reactor Risk Reference Document (NUREG-1150, Vol. 1), NRC noted that for the Zion plant (a four-loop PWR), that comparisons made between the Source Term Code Package results and MAAP results indicated that the MAAP estimates for environmental release fractions were significantly smaller. It is very difficult to determine the precise source of the differences observed, however, without performing controlled comparisons for identical boundary conditions and input data. 16 We are unaware of NRC having performed such comparisons.

The NUREG-1465 source term was also reviewed by an expert panel in 2002, which concluded that it was generally applicable for high-burnup fuel. 17 This and other insights by the panel on the NUREG-1465 source term are being used by the NRC in radiological consequence assessments for the ongoing analysis of nuclear power plant vulnerabilities. 18 15 J. Lehner et al., Benefit Cost Analysis of Enhancing Combustible Gas Control Availability at Ice Condenser and Mark III Containment Plants, Final Letter Report, Brookhaven National Laboratory, Upton, NY, December 23, 2002, p. 17. ADAMS Accession Number ML031700011.

16 U.S. NRC, Reactor Risk Reference Document: Main Report, Draft for Comment, NUREG-1150, Volume 1, February 1987, p. 5-14.

17 J. Schaperow, U.S. NRC, memorandum to F. Eltawila, Radiological Source Terms for High-Burnup and MOX Fuels, December 13, 2002.

18 J. Schaperow (2002), op cit.

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In light of this, it is clear that Next Era should not have used a MAAP-generated source terms in its SAMA analysis.

C-C THE SAMA ANALYSIS FOR SEABROOK USES AN OUTDATED AND INACCURATE PROXY TO PERFORM ITS SAMA ANALYSIS, THE MACCS2 COMPUTER PROGRAM BASIS The MACCS2 Code: The Applicants SAMA analysis uses MELCOR Accident Consequence Code System (MACCS2) computer program. 19 There is no NRC regulation requiring the use of that code, or any other particular code. It was a choice by NextEra and the wrong choice, certainly without considerably updating it. The code is not QAd 20

- the codes MACCS & MACCS2 were developed for research purposes not licensing purposes -for that reason they were not held to the QA requirements of NQA-a (American Society of Mechanical Engineering, QA Program Requirements for Nuclear Facilities, 1994). Rather they were developed using following the less rigorous QA guidelines of ANSI/ANS 10.4. [American Nuclear Standards Institute and American Nuclear Society, Guidelines for the Verification and Validation of Scientific and Engineering Codes for the Nuclear Industry, ANSI/ANS 10.4, La Grange Park, IL (1987). ] A further defect of the code is that there is no explanation of exactly how it works - its assumptions and bases for those assumptions- how it interacts with long-term dose accumulation models. The cost formula and assumptions contained in the MACCS2 19 ER.E E, Attachment F, F.3.4 20 Chanin, D.I. (2005), "The Development of MACCS2: Lessons Learned," [written for:] EFCOG Safety Analysis Annual Workshop Proceedings, Santa Fe, NM, April 29-May 5, 2005. Available online at: Full text: the development of maccs2.pdf (154 KB), revised 12/17/2009.

http://chaninconsulting.com/index.php?resume.

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underestimate the costs likely to be incurred as a result of a severe accident, as explained further below. The cost formula and assumptions contained in the MACCS2 underestimate the costs likely to be incurred as a result of a severe accident, explained in greater detail further below. As an example, the code incorrectly models doses in the codes EARLY and CHRONC modules. In CHRONC (7 days after the accident to 30 years) the code incorrectly assumes the indoor dose is essentially zero; whereas in reality, the indoor dose at this stage of the accident becomes equivalent to the outdoor dose. If correctly modeled, the indoor dose would increase by a factor of 2-4.

C-D. 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.

BASIS In determining the geographic concentration of radionuclides released in a severe accident, NextEra used an atmospheric dispersion model not appropriate for Seabrook Stations coastal site. They used a steady-state, straight-line Gaussian plume model that is incorporated, or embedded, in the MACCS2 code. The plume model underestimated the area likely to be affected in a severe accident and the dose likely to be received in those areas. Instead, NextEra should have modeled transport and deposition using a site-appropriate variable plume model such as AERMOD or CALPUFF. Meteorological research performed at coastal sites, including along the coast of Massachusetts, support our contention 21.

21 For example: Miller, Samuel T.K.; Keim, Barry; Synoptic-Scale Controls on the Sea Breeze of the 47

a. The straight-line Gaussian plume model assumes that a released radioactive plume travels in a steady -state straight-line, i.e., the plume functions much like a beam from a flashlight. The MACCS2 code used by Applicant is based upon this straight-line, steady-state model; it also assumes meteorological conditions that are steady in time and uniform spatially across the study region. However, site specific research at Seabrooks location shows that the assumption of a steady-state, straight-line plume is inappropriate

- winds are variable and dose will be more concentrated than modeled and extend over a larger area.

The accuracy of a straight-line steady-state Gaussian air dispersion model decreaces with distance from the source of the release. For that reason, EPA does not approve of use of a straight-line Gaussian plume to predict the dispersion of a pollutant beyond 32 miles. Therefore the Applicants use of the ATMOS model to predict dispersion in a 50-mile radius of the plant, an area which includes the highest population concentrations, 22 is unacceptable. Appendix E (2.2) says that, There are two metropolitan areas within 50 miles of the site; Manchester, New Hampshire (31 miles west-northwest), and Boston, Massachusetts (41 miles south-southwest).

Central New England Coast, AMS Journal Online, Volume 18, Issue 2 (April 2003), available on line at:

http://journals.ametsoc.org/doi/full/10.1175/1520-0434%282003%29018%3C0236%3ASCOTSB%3E2.0.CO%3B2; Angevine, Wayne; Trainer, Michael; McKeen, Stuart; Berkowitz, Carl; Mesoscale Meteorology of the New England Coast. Gulf of Maine and Nova Scotia: Overview, JOURNAL OF GEOPHYSICAL RESEARCH, VOL. 101, NO. D22, PP. 28,893-28,901, 1996 doi:10.1029/95JD03271, available on line at:

http://www.agu.org/pubs/crossref/1996/95JD03271.shtml; Thorp, Jennifer E., Eastern Massachusetts Sea Breeze Study, Thesis Submitted to Plymouth State University in Partial Fulfillment of the Requirements for the Degree of Master of Science in Applied Meteorology, May 2009; Colby Jr, F.P.., 2004: Simulation of the New England Sea Breeze (Attachment B): The effect of grid spacing. WEA Forecast., 19, 277-285; 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, available on line at:

• http://journals.ametsoc.org/doi/pdf/10.1175/JAM2333.1 22 Appendix E, Section 2.1, Figure 2.1-2 50 mile Vicinity Map shows urban areas shaded 48

NextEras straight-line, steady-state Gaussian plume model does not allow consideration for the fact that the winds for a given time period may be spatially varying, and it ignores the presences of sea breeze circulations which dramatically alter air flow patterns. Because of these failings the straight-line Gaussian plume model is not appropriate for the Seabrooks coastal location.

The nearby presence of the ocean greatly affects atmospheric dispersion processes and is of great importance to estimating the consequences in terms of human lives and health effects of any radioactive releases from the facility, and that the transport, diffusion, and deposition of airborne species emitted along a shoreline can be influenced by mesoscale atmospheric motions. These cannot be adequately simulated using a Gaussian plume model.

b. The Sea Breeze Effect: The sea breeze effect, ignored by NextEras model, is a critical feature to consider at Seabrooks coastal location. In a sea breeze winds heading initially out to sea are drawn back on shore when the land becomes warmer than the water - sometimes penetrating inland here 20-40 miles. 23 The reverse occurs as the land cools. Sea breeze pulls the plume down towards the land surface increasing dose to the population. If the same meteorological conditions (strong solar insolation, low synoptic-scale winds) that are conducive to the formation of sea breezes at a coastal site occurred at a non coastal location, the resulting vertical thermals developing over a pollution source would carry contaminants aloft. In contrast, at a coastal site, the sea breeze would draw contaminants across the land and inland subjecting the population to 23 See, for example, attached document, Thorp, Jennifer E., Eastern Massachusetts Sea Breeze Study, Thesis Submitted to Plymouth State University in Partial Fulfillment of the Requirements for the Degree of Master of Science in Applied Meteorology, May 2009, Attachment B 49

potentially. Straight-line Gaussian plume are thereby non-conservative. Next- Era by ignoring this important and well-documented coastal phenomena underestimates consequence.

The presence of a sea breeze circulation changes the wind directions, wind speeds and turbulence intensities both spatially and temporally through out its entire area of influence. The classic reference Meteorology and Atomic Energy, (Section 2-3.5 )

(Slade, 1968) 24 succinctly comments on the importance of sea breeze circulations as, The sea breeze is important to diffusion studies at seaside locations because of the associated changes in atmospheric stability, turbulence and transport patterns. Moreover its almost daily occurrence at many seaside locations during the warmer seasons results in significant differences in diffusion climatology over rather short distances.

c. Behavior of Plumes over Water: NextEras Gaussian plume model appears to assume that plumes blowing out to sea would have no impact. However a plume over water, rather than being rapidly dispersed, will remain tightly concentrated due to the lack of turbulence, and will remain concentrated until winds blow it onto land [Zager et al.; Angevine et al. 2006]. This can lead to hot spots of radioactivity in places along the coast, certainly to Boston or north up the NH and Maine coasts bringing larger doses over a greater geographic area than modeled and with high population concentrations. 25

d. Terrain Effects: According to the Seabrook License Renewal Application, The terrain varies from hilly to mountainous except along the coast. (ER. E, Section 2-10, 24

1. Slade, David, Meteorology and atomic energy, 1968.Prepared by Air Resources Laboratories, et al. For the Division of Reactor Development and Technology, US AEC 25 In addition to Angevine, Miller and Thorp see: Jan Beyea, Ph.D., 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, May 25, Pg., 11, NRC Electronic Library, Adams Accession No. ML061640329 50

pg., 2-70) ATMOS does not allow consideration of the fact that the winds for any given period of time may be spatially varying. The 1997 User Guide for MACCS2, SAND 97-0594 26 makes the point: The atmospheric model included in the code does not model the impact of terrain effects on atmospheric dispersion. Terrain effects can have a highly complex impact on wind field patterns and plume dispersion. Wind blowing inland will experience the frictional effects of the surface which decrease speed and direction. EPA has recognized that geographical variations can generate local winds and circulations, and modify the prevailing ambient winds and circulations and that assumptions of steady-state straight-line transport both in time and space are inappropriate. [EPA Guidelines on Air Quality Models (Federal Register Nov. 9, 2005, Section 7.2.8, Inhomogeneous Local Winds, italics added EPA's November 9, 2005 modeling Guideline (Appendix A to Appendix W) lists EPA's "preferred model; the Gaussian plume model used by NextEra (ATMOS) is not on the list. EPA recommends that CALPUFF, a non-straight-line model, be used for dispersion beyond 50 Km. 27 The essential difference between the models that EPA recommends for dispersion studies and the two-generation-old Gaussian plume model (ATMOS) used by NextEra is more than determining where a plume will likely to go. Major improvements in the simulation of vertical dispersion rates have been made in the EPA models by recognizing the importance of surface conditions on turbulence rates as a function of height above the ground (or ocean) surfaces. We know that turbulence rates and wind speeds vary greatly 26 Chanin, D.I., and M.L. Young, Code Manual for MACCS2:Volume 1, Users Guide, SAND97-0594 Sandia National Laboratories, Albuquerque, NM, (1997), available on line at:

http://www.doeal.gov/SWEIS/OtherDocuments/481%20MACCS2%20Vol%201.pdf 27 Appendix A to Appendix W to 40 CFR Part 51, EPA Revision to the Guideline on Air Quality Models:

Adoption of a Preferred General Purpose (Flat and Complex Terrain) Dispersion Model and Other Revisions; Final Rule, November 9, 2005. http://www.epa.gov/ scram001/guidance/guide/appw_05.pdf.

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as a function of height above a surface depending upon whether the surface is rough or smooth (trees vs. over water transport) (Roughness), how effectively the surface reflects or absorbs incoming solar radiation (Albedo) and the degree that the surface converts latent energy in moisture into thermal energy (Bowen ratio). These parameters are included in the AERMOD and CALPUFF models and determine the structure of the temperature, wind speed and turbulent mixing rate profiles as a function of height above the ground. NextEras ATMOS model does not include these parameters. This is an especially important deficiency when modeling facilities located along coastlines, such as Seabrook.

Additionally, the MACCS2 Guidance Report, June 2004, 28 itself warns that the code does not model dispersion close to the source (less than 100 meters from the source). Thereby ignoring resuspension of contamination blowing offsite and affecting deposition in offsite communties and adding to costs.

The fact that the MACCS2s ATMOS model was inappropriate for use at Seabrook Station should have been apparent to NextEra from reading the MACCS2 Guidance Report, June 2004, referenced directly above. It additionally warned that the code should be applied with caution at distances greater than ten to fifteen miles, especially if meteorological conditions are likely to be different from those at the source of release.

There are large potentially affected population concentrations more than 10-15 miles from Seabrook - for example: Haverhill is 15 miles southwest of the site and is decribed in ER, 2.1 as the closest population center (defined in 10 CFR [Reactor site Criteria]

as densely populated center with 20,000 or more); and as already mentioned, the 28 MACCS2 Guidance Report June 2004 Final Report page 3-8:3.2 Phenomenological Regimes of Applicability 52

metropolitan area of Boston is 41 miles away and Manchester 31 miles distant. Furthur the MACCS2 Guidance Report, June 2004 said that, Gaussian models are inherently flat-earth models, and perform best over regions where there is minimal variation in terrain. Next Eras description of the Seabrook site says that the, [t]opography consists of rolling forested hills interspersed with urban areas. [ER, 2-1]

e. Input Data: Another significant defect in Applicants model is that its meteorological inputs (e.g., wind speed, wind direction, atmospheric stability and mixing heights) into the MACCS2 are based on data collected by Applicant at a single, on-site anemometer for a single year, 2005. (LRA, Appendix E, F.3.4.5) Measurement data from one station will definitely not suffice to define the sea breeze or capture variability.

f. Government and Independent Studies: Support Petitioners claim that a straight line Gaussian plume model cannot account for the effects of complex terrain on the dispersion of pollutants from a source. Therefore its use is inappropriate for use for Seabrooks analysis to determine the potential area of impact and deposition in a severe accident. For example:

NRC 1972: NRC Regulatory Guide 123 (Safety Guide 23) On Site Meteorological Programs 1972, states that, "at some sites, due to complex flow patterns in non-uniform terrain, additional wind and temperature instrumentation and more comprehensive programs may be necessary.

1977: NRC began to question the feasibility of using straight line Gaussian plume models for complex terrain. See U.S.NRC, 1977, Draft for Comment Reg. Guide 1.111 at 53

1c (pages 1.111-9 to 1.111-10) 1983: In January 1983, NRC Guidance [ NUREG-0737, Supplement 1 Clarification of TMI Action Plan Requirements," January 1983 Regulatory Guide 1.97- Application to Emergency Response Facilities; 6.1 Requirements],suggested that changes in on-site meteorological monitoring systems would be warranted if they have not provided a reliable indication of monitoring conditions that are representative within the 10-mile plume exposure EPZ.

1996: The NRC acknowledged the inadequacy of simple straight-line Gaussian plume models to predict air transport and dispersion of a pollutant released from a source in a complex terrain when it issued RTM-96, Response Technical Manual, which contains simple methods for estimating possible consequences of various radiological accidents.

In the glossary of that document, the NRCs definition of Gaussian plume dispersion model states that such models have important limitations, including the inability to deal well with complex terrain. NUREG/BR-0150, Vol.1 Rev.4, Section Q; ADAMS Accession Number ML062560259, 2004: A NRC research paper, Comparison of Average Transport and Dispersion Among a Gaussian, A Two- Dimensional and a Three-Dimensional Model, Lawrence Livermore National Laboratory, October, 2004 at 2. (Livermore Report) had an important caveat added to the Reports summary about the scientific reliability of the use of a straight-line Gaussian model in complex terrains:

. . . [T]his study was performed in an area with smooth or favorable terrain and persistent winds although with structure in the form of low-level nocturnal jets and severe storms. In regions with complex terrain, particularly if the surface wind direction changes with height, caution should be used.

Livermore Report at 72 (Emphasis added) 54

2005: In December, 2005, as part of a cooperative program between the governments of the United States and Russia to improve the safety of nuclear power plants designed and built by the former Soviet Union, the NRC issued a Procedures Guide for a Probabilistic Risk, related to a Russian Nuclear Power Station. The Guide, prepared by the Brookhaven National Laboratory and NRC staff, explained that atmospheric transport of released material is carried out assuming Gaussian plume dispersion, which is generally valid for flat terrain. However, the Guide the caveat that in specific cases of plant location, such as, for example, a mountainous area or a valley, more detailed dispersion models may have to be considered. Kalinin VVER-1000 Nuclear power Station Unit 1 PRA, Procedures Guide for a Probabilistic Risk Assessment, NUREG/CR- 6572, Rev. 1 at 3-114; excerpt attached as Exhibit 8, full report available at http://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr6572.

2007: NRC revised their Regulatory Guide 1.23, Meteorological Monitoring Programs for Nuclear Power Plants. On page 11, the section entitled Special Considerations for Complex Terrain Sites says that, At some sites, because of complex flow patterns in nonuniform terrain, additional wind and temperature instrumentation and more comprehensive programs may be necessary. For example, the representation of circulation for a hill-valley complex or a site near a large body of water may need additional measuring points to determine airflow patterns and spatial variations of atmospheric stability. Occasionally, the unique diffusion characteristics of a particular site may also warrant the use of special meteorological instrumentation and/or studies.

The plants operational meteorological monitoring program should provide an adequate basis for atmospheric transport and diffusion estimates within the plume exposure 55

emergency planning zone [i.e., within approximately 16 kilometers (10 miles)]. 29 These excerpts from Regulatory Guide 1.23 demonstrate that the NRC recognizes there are certain sites, such as those located in coastal areas, like Seabrook, that multiple meteorological data input sources are needed for appropriate air dispersion modeling. Not simply one or two meteorological towers onsite. Since the straight-line Gaussian plume model is incapable of handling complex flow patterns and meteorological data input from multiple locations, Regulatory Guide 1.23 demonstrates NRCs recognition that it should not be used at any site with complex terrain.

2009: NRC made a presentation to the National Radiological Emergency Planning Conference 30 concluded that the straight-line Gaussian plume models cannot accurately predict dispersion in a complex terrain and are therefore scientifically defective for that purpose

[ADAMS - ML091050226, ML091050257, and ML091050269 (page references used here refer to the portion attached, Part 2, ML091050257).] Most reactors, if not all, are located in complex terrains. In the presentation, NRC said that the most limiting aspect of the basic Gaussian Model, is its inability to evaluate spatial and temporal differences in model inputs [Slide 28]. Spatial refers to the ability to represent impacts on the plume after releases from the site e.g., plume bending to follow a river valley or sea breeze circulation. Temporal refers to the ability of the model to reflect data changes over time, e.g., change in release rate and meteorology [Slide 4]. Because the basic Gaussian model is non-spatial, it cannot account for the effect of terrain on the trajectory of the plume -

29 For example, if the comparison of the primary and supplemental meteorological systems indicates convergence in a lake breeze setting, then a keyhole protective action recommendation (e.g., evacuating a 2-mile radius) 30 Ibid 56

that is, the plume is assumed to travel in a straight line regardless of the surrounding terrain. Therefore, it cannot, for example, curve a plume around mountains or follow a river valley. NRC 2009 Presentation, Slide 33. NextEra describes Seabrooks 50-mile area, Appendix E (Section 2.10) as, The terrain varies from hilly to mountainous except along the coast. Further NRC says that it cannot account for transport and diffusion in coastal sites subject to the sea breeze. NextEra at Appendix E (Section 2.10) acknowledges that there is a sea breeze during spring and summer. The NRC says that the sea breeze causes the plume to change direction caused by differences in temperature of the air above the water versus that above the land after sunrise. If the regional wind flow is light, a circulation will be established between the two air masses. At night, the land cools faster, and a reverse circulation (weak) may occur [Slide 43]. Turbulence causes the plume to be drawn to ground level [Slide 44]. The presentation goes on to say that, Additional meteorological towers may be necessary to adequately model sea breeze sites [Slide 40].

Significantly, the NRC 2009 Presentation then discussed the methods of more advanced models that can address terrain impact on plume transport, including models in which emissions from a source are released as a series of puffs, each of which can be carried separately by the wind, (NRC 2009 Presentation Slides 35, 36). This modeling method is similar to CALPUFF. Licensees are not required, however, to use these models in order to more accurately predict where the plume will travel to base protective action recommendations.

EPA Likewise, EPA recognized the need for complex models. For example: EPAs 2005 57

Guideline on Air Quality Models says in Section 7.2.8 Inhomogenous Local Winds that, In very rugged hilly or mountainous terrain, along coastlines, or near large land use variations, the characterization of the winds is a balance of various forces, such that the assumptions of steady-state straight line transport both in time and space are inappropriate. (Fed.

Reg., 11/09/05).

EPA goes on to say that, In special cases described, refined trajectory air quality models can be applied in a case-by-case basis for air quality estimates for such complex non-steady-state meteorological conditions. This EPA Guideline also references an EPA 2000 report, Meteorological Monitoring Guidance for Regulatory Model Applications, EPA-454/R-99-005, February 2000. Section 3.4 of this Guidance for coastal Locations, discusses the need for multiple inland meteorological monitoring sites, with the monitored parameters dictated by the data input needs of particular air quality models.

EPA concludes that a report prepared for NRC 31 provides a detailed discussion of considerations for conducting meteorological measurement programs at coastal sites, reactors on large bodies of water. Most important, EPA's November 2005 Modeling Guideline (Appendix A to Appendix W) lists EPA's "preferred models and the use of straight line Gaussian plume model, called ATMOS, is not listed. Sections 6.1 and 6.2.3 discuss that the Gaussian model is not capable of modeling beyond 50 km (32 miles) and the basis for EPA to recommend CALPUFF, a non - straight line model. 32 DOE DOE, too, recognizes the limitations of the straight-line Gaussian plume model. They say 31 Raynor, G.S.P. Michael, and S. SethuRaman, 1979, Recommendations for Meteorological Measurement Programs and Atmospheric Diffusion Prediction Methods for Use at Coastal Nuclear Reactor Sites.

NUREG/CR-0936, U.S. Nuclear Regulatory Commission, Washington, DC 32 http://www.epa.gov/scram001/guidance/guide/appw_05.pdf 58

for example that Gaussian models are inherently flat-earth models, and perform best over regions of transport where there is minimal variation in terrain. Because of this, there is inherent conservatism (and simplicity) if the environs have a significant nearby buildings, tall vegetation, or grade variations not taken into account in the dispersion parameterization. 33 National Research Council Tracking and Predicting The Atmospheric Dispersion of Hazardous Material Releases Implications for Homeland Security, Committee on the Atmospheric Dispersion of Hazardous Material Releases Board on Atmospheric Sciences and Climate Division on Earth and Life Studies, National Research Council of the National Academies, 2003. The report discusses how the analytical Gaussian models were used in the 1960s and tested against limited field experiments in flat terrain areas performed in earlier decades.

In the 1970s the US passed the Clean Air Act which required the use of dispersion models to estimate the air quality impacts of emissions sources for comparison to regulatory limits. This resulted in the development and testing of advanced models for applications in complex terrain settings such as in mountainous or coastal areas. In the 1980s, further advances were made with Lagrangian puff models and with Eulerian grid models. Gaussian models moved beyond the simple use of sets of dispersion coefficients to incorporate Monin-Obukhov and other boundary layer similarity measures which are the basis of contemporary EPA models used for both short range and long range transport applications. Helped enormously by advances in computer technologies, in the 1990s, significant advances were made in numerical weather prediction models and also further 33 the MACCS2 Guidance Report June 2004 Final Report, page 3-8:3.2 Phenomenological Regimes of Applicability 59

improve dispersion models through the incorporation of field experiment results and improved boundary layer parameterization. The decade starting with the year 2000 has seen improved resolution of meteorological models such as MM5 and the routine linkage of meteorological models with transport and dispersion models as exemplified by the real time forecasts of detailed fine grid weather conditions available to the public at Olympic events. Computational Fluid Dynamics (CFD) models which involve very fine grid numerical simulations of turbulence and fluid flow began to see applications in atmospheric dispersion studies. The next decade will see routine application of CFD techniques to complex flows associated with emergency response needs.

The nuclear industry does not show evidence of keeping up with these technological advances. For use in modeling air quality concentrations, the NRC uses straight-line Gaussian dispersion algorithms that date back to the 1960s. Complex flow situations such as those associated with flow around high terrain features or that would incorporate sea breeze circulations are not simulated. For emergency response applications, the NRC does not seem to require any advanced modeling to be installed at nuclear power plants.

Atmospheric Scientists & Meteorologists For over three decades atmospheric scientists and meteorologists have been identifying problems in the use of models similar to ATMOS for such settings. Example:

Steven R. Hanna, Gary A. Briggs, Rayford P. Hosker, Jr., National Oceanic and Atmospheric Administration, Atmospheric Turbulence and Diffusion Laboratory, Handbook on Atmospheric Diffusion (1982)).

The inability of a simple Gaussian plume model to accurately predict air transport and 60

dispersion in complex terrains is such a basic flaw that it is discussed in a textbook for a college-level introductory course in environmental science and engineering (Environmental Science and Engineering, J. Glynn Henry & Gary W. Heinke, (Prentice-Hall 1989) at 528 (Chapter 13 authored by William J. Moroz). In listing the assumptions that are made to develop a simple straight line Gaussian plume model, the textbook warns that:

The equation is to be used over relatively flat, homogeneous terrain. It should not be used routinely in coastal or mountainous areas, in any area where building profiles are highly irregular, or where the plume travels over warm bare soil and then over colder snow or ice covered surfaces.

g. NextEra used NRCs practice of using mean consequence values in their SAMA analysis, resulting in averaging of potential consequences that minimized the findings and conclusions on the meteorological modeling: This is discussed below, Section F.

C-D 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.

Basis The ER is required to include a consideration of alternatives to mitigate severe accidents (SAMA). 10 CFR 51.53(c )(30(ii)(L) That analysis depends upon an accurate calculation of the cost of a severe accident in order to have a base line against which to measure proposed mitigation measures. NextEra, instead, severely minimized decontamination and clean-up costs, health costs (that includes inaccurately modeling 61

evacuation time estimates), and minimized and ignored a myriad of other economic costs that belong in a SAMA analysis.

1. Decontamination And Clean Up Costs The SAMA analysis for Seabrook Station Unit 1 uses the outdated and inaccurate MACCS2 code to calculate decontamination and clean up costs. (Appendix E, F.4.1) The cost formula used in the MACCS2 underestimates costs likely to be incurred as a result of a dispersion of radiation. Therefore NextEras SAMA analysis significantly underestimates the costs associated with an accident.

The MACCS2 Decontamination Plan is described in part in the Code Manual for MACCS2: Volume I, Users Guide (NUREG/CR-6613, Vol. 1) Prepared by D. Chanin and M.I. Young, May 1998. Section 7.5 Decontamination Plan describes some of the assumptions. It says at 7-10 that, Many decontamination processes (e.g., plowing, fire hosing) reduce groundshine and resuspension doses by washing surface contamination down into the ground. Since these processes may not move contamination out of the root zone, the WASH-1400 based economic cost model of MACCS2 assumes that farmland decontamination reduces direct exposure doses to farmers without reducing uptake of radioactivity by root systems.

Thus decontamination of farmland does not reduce the ingestion doses produced by the consumption of crops that are contaminated by root uptake.

Simply from this section of the document, it becomes clear what is wrong. For example:

(1) It says the economic cost model, is based on WASH-1400; WASH-1400, in turn, was based on clean up after a nuclear explosion. However, cleanup after a nuclear bomb explosion is not comparable to clean up after a nuclear reactor accident and assuming so will underestimate cost. Nuclear explosions result in larger-sized radionuclide particles; reactor accidents release small sized particles. Decontamination is far less effective, or 62

even possible, for small particle sizes. Nuclear reactor releases range in size from a fraction of a micron to a couple of microns; whereas nuclear bomb explosions fallout is much larger- particles that are ten to hundreds of microns. These small nuclear reactor releases can get wedged into small cracks and crevices of buildings making clean up extremely difficult or impossible.

WASH-1400s referenced nuclear weapon clean up experiments involved cleaning up fallout involving large mass loading where the there was a small amount of radioactive material in a large mass of dirt and demolished material. Only the bottom layer will be in contact with the soil and the massive amount of debris can be swept up with brooms or vacuums resulting in a relatively effective, quick and cheap cleanup that would not be the case with a nuclear reactors fine particulate.

Third a weapon explosion results in non-penetrating radiation so that workers only require basic respiration and skin protection. This allows for cleaning up soon after the event. In contrast a reactor release involves gamma radiation and there is no gear to protect workers from gamma radiation. Therefore cleanup cannot be expedited and decontamination is less effective with the passage of time.

Also ignored is radioactive waste disposal. In a weapons event, the waste could be shipped to Utah or to the Nevada Test Site. The Greater- than- Class C waste expected in a reactor accident would not have a repository likely available to receive such a large quantity of material in the foreseeable future. Also, the costs incurred for safeguarding the wastes and preventing their being re-suspended are not accounted for in the model.

Even optimistically assuming a repository becoming available, (Utah site is approximately one-square mile and the volume of waste from a severe accident at 63

Seabrook would likely require a larger facility) it seems unlikely that there would be a sufficient quantity of transport containers and communities not objecting to the hazardous materials going over their roads and through their communities.

(2) The Users Guide described decontamination processes as plowing and fire hosing. We know that CERLA, EPA and local authorities would not allow use of those methods. Fire hosing and plowing does not decontaminate, it simply moves the contamination from one place to another -only to reappear again later in groundwater, resuspended into the air, or in food. Therefore cleanup will take far longer, be more expensive and its success (defined as returning to pre-accident status) unlikely.

Apparently missing from consideration is that forests, wetlands and shorelines cannot realistically be cleanup and decontaminated. The area within 50-miles of Seabrook Station consists of miles of beaches, rivers, wetlands, forests and park land (see, for example, ER, E. 2.1. Figures 2.1-1 and 2.1-2)

Additionally, urban areas will be considerably more expensive and time consuming to decontaminate and clean than rural areas. A map ( Application, ER. Figure 2-6-1) clearly shows numerous water and urban areas within 50-miles.

The US Department of Homeland Security has commissioned studies for the economic consequences of a Rad/Nuc attack and although much more deposition would occur in reactor accident, magnifying consequences and costs, there are important lessons to be learned from these studies.

Barbara Reichmuths study, Economic Consequences of a Rad/Nuc attack: Cleanup 64

Standards Significantly Affect Cost, 2005, 34 Table 1 Summary Unit Costs for D &D (Decontamination and Decommissioning) Building Replacement and Evacuation Costs provides estimates for different types of areas from farm or range land to high density urban areas. Reichmuths study also points out that the economic consequences of a Rad/Nuc event are highly dependent on cleanup standards. Cleanup costs generally increase dramatically for standards more stringent than 500 mrem/yr; however currently a cleanup standard is not agreed upon by NRC and EPA and appears to range from 15 mrem/yr to 5 rem/yr.

The General Accounting Office (GAO) reports that the current EPA and NRC cleanup standards differ and these differences have implications for both the pace and ultimate cost of cleanup. 35 NextEra does not appear to account for this issue.

A similar study was done by Robert Luna, Survey of Costs Arising from Potential Radionuclide Scattering Events. 36 Luna concluded that, the expenditures needed to recover from a successful attack using an RDD type device are likely to be significant from the standpoint of resources available to local or state governments. Even a device that contaminates an area of a few hundred acres (a square kilometer) to a level that requires modest remediation is likely to produce costs ranging from $10M to $300M or more depending on the intensity of commercialization, population density, and details of land use in the area. (Luna, Pg., 6)

Therefore a severe accident at Seabrook is likely to result in huge costs; costs not accounted for by NextEra, because of the type and magnitude of radionuclides released in comparison with a RDD type device.

34 Economic Consequences of a Rad/Nuc attack: Cleanup Standards Significantly Affect Cost Barbara Reichmuth, Steve Short, Tom Wood, Fred Rutz, Debbie Swartz, Pacific Northwest National laboratory, 2005(Attachment C) 35 GAO, radiation Standards Scientific Basis Inconclusive, and EPA and NRC Disagreement Continues, June 2004 36 Survey of Costs Arising From Potential Radionuclide Scattering Events, Robert Luna, Sandia National laboratories, WM2008 Conference, February 24-28, 2008, Phoenix AZ (Attachment D) 65

In place of the outdated decontamination costs figure in the MACCS2 code, the SAMA analysis for Seabrook should incorporate, for example, the analytical framework contained in the 1996 Sandia National laboratories report concerning site restoration costs 37 as well as studies examining Chernobyl and RDD type devices.

The Sandia Site restoration study analyzed the expected financial costs for cleaning up and decontaminating a mixed-use urban land and Midwest farm and range land. The study was commissioned by DOE to estimate activities likely to be involved in the decontamination of an accident involving the dispersal of plutonium. Although there would be many differences in a nuclear reactor accident, the methodology and conclusions to estimate costs are directly useful.

The study recognized 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 nuclear explosion of nuclear weapons that produce large particle sizes and high mass loadings.

For an extended decontamination and remediation operation in a mixed-use urban area with an average population density, Site restoration predicted a cleanup cost of

$311,000,000 per square km using offsite disposal and $309,000,000 per square km using on-site disposal. (Site restoration, Pg., 6-5)

The costs would be much higher for example for the metropolitan areas of Boston, Manchester, Portsmouth, and Portland considering that they are tourist, educational, transportation, and financial centers. The economic losses stemming from the stigma 37 Site Restoration: Estimation of Attributable Costs from Plutonium-Dispersal Accidents, SAND96-0957, David Chanin, Walt Murfin, UC-502, (May 1996), available on line at:

http://chaninconsulting.com/index.php?resume 66

effects of a severe accident would be staggering. The Sandia Site restoration study further says, In comparing the numbers of cancer health effects that could result from a plutonium-dispersal accident to those that could result from a severe accident at a commercial nuclear power plant, it is readily apparent that the health consequences and costs of a severe reactor accident could greatly exceed the consequences of even a worst- case plutonium-dispersal accident because the quantities of radioactive material in nuclear weapons are a small fraction of the quantities present in an operating nuclear power plant.

(Site restoration, Pg., 2-3, 2-4)

NextEra lists under decontamination costs (LRA, Appendix E, Attachment F, F.3.4.2) the costs of farm and non-farm decontamination and the value of farm and nonfarm wealth. However nowhere is there a discussion of the loss of, and costs to remediate the economic infrastructure that make business, tourism and other economic activity possible. Economic infrastructure is the basic physical and organizational structures needed for the operation of a society or enterprise, or the services and facilities necessary for an economy to function. The term typically refers to the technical structures that support a society, such as roads, water supply, sewers, power grids telecommunications, and so forth. Viewed functionally, infrastructure facilitates the production of goods and services; for example, roads enable the transport of raw materials to a factory, and also for the distribution of finished products to markets. Also, the term may also include basic social services such as schools and hospitals.

NextEra also appears to ignore the indirect economic effects or the multiplier effects. For example, depending on the business done inside the building contaminated, the regional and national economy could be negatively impacted. A resulting decrease in the areas real estate prices, tourism, and 67

commercial transactions could have long-term negative effects on the regions economy.

NextEra must be required to take all of these real cleanup costs into account.

NextEras SAMA analysis fails to do so and grossly underestimates costs making mitigations not appear cost effective.

2. Health Costs Health costs are an important part of economic consequences. NextEras life lost value is much too low. EPA values a life lost at $6.1 million (U.S.E.P.A., 1997, The Benefits and Costs of the Clean Air Act, 1970 to 1990, Report to US Congress (October),

pages 44-45). The current ER assigns a value of $2000 per person rem (ER.E, Attachment F, F.4)

The population dose conversion factor of $2000/person-rem used by NextEra to estimate the cost of the health effects generated by radiation exposure is based on a deeply flawed analysis and seriously underestimates the cost of the health consequences of severe accidents.

NextEra underestimates the population-dose related costs of a severe accident by relying inappropriately on a $2000/person-rem conversion factor. NextEra use of the 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 of the severe accident scenarios included in NextEras risk analysis; 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 68

effectiveness factor (DDREF).

The $2000/person-rem conversion factor is intended to represent the cost associated with the harm caused by radiation exposure with respect to the causation of stochastic health effects, that is, fatal cancers, nonfatal cancers, and hereditary effects. 38 The value was derived by NRC staff by dividing the Staffs estimate for the value of a statistical life, $3 million (presumably in 1995 dollars, the year the analysis was published) by a risk coefficient for stochastic health effects from low-level radiation of 7x10-4/person-rem, as recommended in Publication No. 60 of the International Commission on Radiological Protection (ICRP). (This risk coefficient includes nonfatal stochastic health effects in addition to fatal cancers.) But the use of this conversion factor in NextEras SAMA analysis is inappropriate in two key respects. As a result NextEra underestimates the health-related costs associated with severe accidents.

First, the $2000/person-rem conversion factor is specifically intended to represent only stochastic health effects (e.g. cancer), and not deterministic health effects including early fatalities which could result from very high doses to particular individuals. 39 However, for some of the severe accident scenarios evaluated by NextEra at Seabrook, we estimate that large numbers of early fatalities could occur representing a significant fraction of the total number of projected fatalities, both early and latent. This is consistent with the findings of the Generic Environmental Impact Statement for License Renewal of Nuclear Plants (NUREG-1437). 40 Therefore, it is inappropriate to use a conversion factor that does not include deterministic effects.

38 U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Reassessment of NRCs Dollar Per Person-Rem Conversion Factor Policy, NUREG-1530, 1995, p. 12 39 U.S. NRC (1995), op cit., p. 1.

40 U.S. NRC, Generic Environmental Impact Statement for License Renewal of Nuclear Plants, NUREG-1437, Vol. 1, May 1996, Table 5.5.

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According to NRCs guidance, the NRC believes that regulatory issues involving deterministic effects and/or early fatalities would be very rare, and can be addressed on a case-specific basis, as the need arises. 41 Based on our estimate of the potential number of early fatalities resulting from a severe accident at Seabrook Station, this is certainly a case where this need exists.

Second, the $2000/person-rem factor, as derived by NRC, also underestimates the total cost of the latent cancer fatalities that would result from a given population dose because it assumes that all exposed persons receive dose commitments below the threshold at which the dose and dose-rate reduction factor (DDREF) (typically a factor of

2) should be applied. However, for certain severe accident scenarios at Seabrook evaluated by NextEra, we estimate that considerable numbers of people would receive doses high enough so that the DDREF should not be applied. 42 This means, essentially, that for those individuals, a one-rem dose would be worth more because it would be more effective at cancer induction than for individuals receiving doses below the threshold. To illustrate, if a group of 1000 people receive doses of 30 rem each over a short period of time (population dose 30,000 person-rem), 30 latent cancer fatalities would be expected, associated with a cost of $90 million, using NRCs estimate of $3 million per statistical life and a cancer risk coefficient of 1x10-3/person-rem. If a group of 100,000 people received doses of 0.3 rem each (also a population dose of 30,000 person-rem), a DDREF of 2 would be applied, and only 15 latent cancer fatalities would be expected, at a cost of $45 million. Thus a single cost conversion factor, based on a DDREF of 2, is not appropriate when some members of an exposed population receive 41 U.S. NRC, Reassessment of NRCs Dollar Per Person-Rem Conversion Factor Policy (1995), op cit., p.

13.

42 The default value of the DDREF threshold is 20 rem in the MACCS2 code input 70

doses for which a DDREF would not be applied.

A better way to evaluate the cost equivalent of the health consequences resulting from a severe accident is simply to sum the total number of early fatalities and latent cancer fatalities, as computed by the MACCS2 code, and multiply by the $3 million figure.

Again, we do not believe it is reasonable to distinguish between the loss of a statistical life and the loss of a deterministic life when calculating the cost of health effects.

Another way to explain why NextEras estimates of how many lives might be lost are too low is to look at the 1982 Sandia National Laboratory report, using 1970 census data, that estimated the number of cancer deaths at Seabrook in a severe accident to be 6,000; early fatalities 7,000; and early injuries 27,000. Peak fatalities were estimated by CRAC to occur within 20 miles of Seabrook; and peak injuries to occur with 65 miles of Seabrook from a core melt. (CRAC 2, Sandia, 1982 43) The population of the affected area, no matter what model is used, has greatly increased during the intervening almost 40 years; SAMAs project forward to 2050 based on projected demographics. NextEra estimated the population within 50-miles (2050) to total 5185206. (LRA, Appendix E, Section F.3.4.1, Table F.3.4.1-1) Further CRAC was based on old, and now outdated, dose response models.

In the SAMA, cancer incidence was not considered; neither were the many other potential health effects from exposure in a severe radiological event (National Academy of Sciences, BEIR VII Report, 2005).

NextEras cost-benefit analysis ignored a marked increase in the value of cancer mortality risk per unit of radiation at low doses (2-3 rem average), as shown by recent 43 Calculation of Reactor Accident Consequences, U.S. Nuclear Power Plants (CRAC-2), Sandia National Laboratory, 1982 71

studies published on radiation workers (Cardis et al. 2005 44) and by the Techa River cohort (Krestina et al (2005 45). Both studies give similar values for low dose, protracted exposure, namely (1) cancer death per Sievert (100 rem). According to the results of the study by Cardis et al. and use of the risk numbers derived from the Techa River cohort the SAMA analyses prepared for Seabrook needs to be redone. It seems clear that a number of additional SAMAs that were previously rejected by the applicants methodology will now become cost effective.

Cancer incidence and the other many health effects from exposure to radiation in a severe radiological event (National Academy of Sciences, BEIR VII Report, 2005) must be considered; they were not. Neither did NextEra appear to consider indirect costs.

Medical expenditures are only one component of the total economic burden of cancer.

The indirect costs include losses in time and economic productivity and liability resulting from radiation health related illness and death.

Petitioners examination of NextEras Emergency Response analysis (LRA, Appendix E, Section F.3.4.4) shows that the Applicants 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. NextEra failed to reference specific KLD-type actual time estimates, instead references the paper plan, 44 Elizabeth Cardis, Risk of cancer risk after low doses of ionising radiation: retrospective cohort study in 15 countries. British Medical Journal (2005) 331:77. Available on line at:

http://www.bioone.org/doi/abs/10.1667/RR1443.1?cookieSet=1&prevSearch=

45 Krestinina LY, Preston DL, Ostroumova EV, Degteva MO, Ron E, Vyushkova OV, et al.

2005.Protracted radiation exposure and cancer mortality in the Techa River cohort. Radiation Research 164(5):602-611. Available on line at: http://www.bioone.org/doi/abs/10.1667/RR3452.1 72

Seabrook Station Radiological Emergency Response Plan, Rev. 56, July 2008. No indication is provided, for example, that the following site-specific variables that would slow response time were taken into consideration in the analysis: shadow evacuation; evacuation time estimates during inclement weather coinciding with high traffic periods such as commuter traffic, peak commute time, holidays, summer beach/holiday traffic; notification delay delays because notification is largely based on sirens that cannot be heard in doors above normal ambient noise with windows closed or air conditioning systems operating. The Applicant (ER E., F-160) claims that they assumed no evacuation of the population in a seismically induced severe accident and found only a small increase to the overall total accident dose risk and no change in economic risk. Petitioners find that sensitivity studies do not add useful information if the primary model is flawed, as we have shown is true.

3. A myriad of other economic costs were underestimated or totally ignored by the applicant that when added together would in all likelihood add up collectively to a significant amount. For example, NextEra did not appear to include in their economic cost estimates the business value of property and the incurred costs such as costs required from job retraining, unemployment payments, and inevitable litigation.

They used an assumed value of non-farm wealth that appeared not justified by review of Banker and Tradesmen sales figures. NextEra appear to underestimate Farm Value, for example, by not considering the value of the farm property for development purposes as opposed to agricultural; and farm land assessments are intentionally very low to encourage farming and open space.

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C-E 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 Basis NextEra fails to consider the uncertainties in its consequence calculation resulting from meteorological variations by only using mean values (LRA, Appendix E, 2.10) for population dose and offsite economic cost estimates.

Dr. Edwin S. Lyman, Senior Staff Scientist, Union of Concerned Scientists report commissioned by Riverkeeper, Inc., November 2007, A Critique of the Radiological Consequence Assessment Conducted in Support of the Indian Point Severe Accident Mitigation Alternatives Analysis 46 provides valuable lessons to apply to Seabrooks SAMA.

The consequence calculation, as carried out by the MACCS2 code, generates a series of results based on random sampling of a years worth of weather data. The code provides a statistical distribution of the results. We find, based on calculations done at other reactors such as Indian Point, that the ratio of the 95th percentile to the mean of this distribution is typically a factor of 3 to 4 for outcomes such as early fatalities, latent cancer fatalities and off-site economic consequences.

NextEra admits (LRA, Appendix E, F.8.2- Uncertainty) that, the inputs to the PRA cannot be known with complete certainty, there is a possibility that the actual plant risk is greater than the mean values used in the evaluation of the SAMA described in the previous sections.

46 Report available at NRC Electronic Library, Adams Accession Number ML073410093 74

Kamiar Jamali 47 (Use of risk in measures in design and licensing of future reactors, Reliability Engineering and Safety System 95 (2010) 935-943 www.elsevier,com/locate/ress) makes the same observation. He says that, It is well- known that quantitative results of PRAs, in particular, are subject to various types of uncertainties. Examples of these uncertainties include probabilistic quantification of single and common cause hardware or software failures, occurrence of certain physical phenomena, human errors of omission or commission, magnitudes of source terms, radionuclide release and transport, atmospheric dispersion, biological effects of radiation, dose calculations, and many others. (935).

Despite warning, NextEra describes an unconvincing sensitivity analysis (ER, E.F.8.2- Uncertainty) that they claim resolves the issue. They report, absent any specifics of the study, that to consider the uncertainty, a sensitivity analysis was performed in which an uncertainty factor was applied to the frequencies calculated by the PRA and in subsequent upper bound (UB) benefits were calculated based upon the mean risk multiplied by the this uncertainty factor. The uncertainty factor applied to 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 2.75 E-05/yr; therefore the uncertainty factor is 1.90. NextEras approach at proof is not convincing.

Petitioners contend that Seabrooks SAMA cost-benefit evaluation should be based on the 95th percentile of the meteorological distribution to be consistent with the approach taken in the License Renewal GEIS, which refers repeatedly to the 95th percentile of the risk uncertainty distribution as an appropriate upper confidence bound 47 Kamiar Jamali, DOE Project Manager for Code Manual for MACCS2: Vol. 1, Users Guide (NUREG/CR 6613/SAND 97-0594, Vol.1; DOE Project Manager for Code Manual for MACCS2: Vol. 2, Preprocessor Codes COMIDA A2, FGRDCF, DCF2 (NUREG/CR 6613/SAND 97-0594, Vol. 2); member of the working group for DOE Standard Guidance for Preparation DOE 5480.22(TSR) and DOE 5480.23 (SAR) Implementation Plans, November 1994.

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in order not to underestimate potential future environmental impacts. 48

4. Summary The SAMA analysis included in the Seabrook Environmental Report is incomplete.

NextEras SAMA analysis for Seabrook instead minimized costs likely to be incurred in a severe accident so as mitigation to reduce risk appeared not to be justified by: (1)

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. (2) Minimization of the potential amount of radioactive material released in a severe accident. (3) Use of an outdated and inaccurate proxy, the MACCS2 computer program, to perform its SAMA analysis. (4) 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. (5) 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.

(6) 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.

Petitioners do not offer examples of how this cost benefit equation might have been skewed in favor of no mitigation. The dramatic minimization of costs by NextEra are such that it should be obvious that many SAMAs would be cost effective if the described 48 U.S. NRC, Generic Environmental Impact Statement for License Renewal of Nuclear Plants, NUREG-1437, Vol. 1, May 1996, Section 5.3.3.2.1 76

defects in the analysis were addressed. 49 In Duke Energy Corp., at 13, the licensee argued that NEPA could not require it to implement any particular SAMA, regardless of the how the cost benefit calculations come out, and therefore there was no remedy possible for the Petitioners. But the board rejected this argument, saying While NEPA does not require agencies to select particular options, it is intended to foster both informed decision-making and informed public participation, and thus to ensure the agency does not act upon incomplete information, only to regret its decision after it is too late to correct (citing Louisiana Energy Services (Claiborne Enrichment Center), CLI-98-3, 47 NRC 77, 88 (1998)). It then said if further analysis is called for, that in itself is a valid and meaningful remedy under NEPA. In this contention, Petitioners point to a material deficiency in the Application that the Applicant has drastically under counted the costs of a severe accident that could have led to erroneously rejecting mitigation alternatives and a requirement for further analysis could produce a very different outcome of the proceeding.

IV. CLB vs PEO In considering the foregoing proposed contentions the Commission is requested to also consider the interplay between current design basis (CLB) and aging management through the period of extended operation (PEO). Any suggestion on the part of intervenors that the licensee is not presently in compliance with NRC regulations or operating outside of its design basis is generally rebuffed as outside the scope of license 49 ER 4.21, Pg., 4-41,

Conclusion:

The SAMA analysis identified two SAMA candidates that are potentially cost beneficial: SAMA 157-use of a portable generator to charge station battery; and SAMA 165-install hose adapter and valve to enhance alignment efficiency of fire water to the refueling water tank.

Neither of these SAMA candidates is age-related and therefore does not need to be implemented as part of license renewal pursuant to 10 CFR 54. ER E. Table F.7-1: SAMA 157, cost =$30K; SAMA 165, cost=50K.

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renewal review; matter for enforcement through the Reactor Oversight Process (Program) or non-adjudicatory means. However, the licensee prefaces his LRA with the following:

NextEra Energy Seabrook does not propose to construct or alter any production or utilization facility in connection with this renewal application.

The current licensing basis (CLB) will be continued and maintained throughout the period of extended operation (PEO). Section 1.1 .at 21 The license is promising to keep on, keeping on If the license is not presently adequately managing the aging of critical systems, structures, and components vital to safety (such as safety-related cables, FW and SW pipes, and transformers) and protection of the environment (such as below-grade or inaccessible pipes carrying radionuclides) how can any reasonable reviewer ignore this status quo, when attempting to evaluate the effectiveness of proposed AMPs that are similar or identical? Friends of the Coast/NEC now emphasizes that its contentions do not challenge the existing CLB. Friends of the Coast/NEC could say that with more assurance, and respond to the LRA with more specificity and particularity, if they could access the CLB in its entirety 50, but that is under present circumstances impossible. It is not contained within the LRA. It is not accessible on ADAMS nor in any public library without an exhaustive forensic, time and cost prohibitive, exercise.

50

§ 54.3 Definitions.

(a) As used in this part current licensing basis (CLB) is the set of NRC requirements applicable to a specific plant and a licensee's written commitments for ensuring compliance with and operation within applicable NRC requirements and the plant-specific design basis (including all modifications and additions to such commitments over the life of the license) that are docketed and in effect. The CLB includes the NRC regulations contained in 10 CFR parts 2, 19, 20, 21, 26, 30, 40, 50, 51, 52, 54, 55, 70, 72, 73, 100 and appendices thereto; orders; license conditions; exemptions; and technical specifications.

Note: It also includes the plant-specific design-basis information defined in 10 CFR 50.2 as documented in the most recent final safety analysis report (FSAR) as required by 10 CFR 50.71 and the licensee's commitments remaining in effect that were made in docketed licensing correspondence such as licensee responses to NRC bulletins, generic letters, and enforcement actions, as well as licensee commitments documented in NRC safety evaluations or licensee event reports.

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Yet it is the plants core documentation.

We find it troubling that in todays electronic age it is not possible for petitioners to get onto the NRCs public site or the ADAMS document management system and find the CLB for each plant clearly laid out in a folder with hyperlinks to each separate document. If the NRC must compile this information to continually monitor the compliance of a facility with the regulations, then presumably someone has already done so. If the CLB has not been compiled, then effective regulation is beyond hope.

If the CLB has not been compiled in one easy to access location, how can the public be assured that the NRC is adequately monitoring the facility? It is imperative for purposes of fair and meaningful public participation that the CLB be available to the public.

Friends of the Coast/NEC believes that the Rule on License Renewal technology has advanced to a point where it would be possible for the NRC to make this information available to the public. This simple act would foster a level of transparency that would be very helpful in the license renewal process.

Friends of the Coast/NEC urges the Commission to consider an organized CLB compilation, if not as a prerequisite to filing a LRA, then as a first requisite for applicant disclosures that are normally mandated in LRA Sub-part L Proceedings.

V. CONCLUSION AND REQUEST FOR RELIEF NextEra's application should be denied. Alternatively, Friends of the Coast and New England Coalition seek protection of their interests through an ASLB Order requiring, as prerequisite to license renewal, that NextEra cure the inadequacies in its application as described 79

above so as to provide assurance of public health and safety.

In addition, Seabrook must operate within the requirements of 10 CFR 50 and 54 and the NRC must provide reasonable assurance to the public that the plant is in compliance with all NRC regulations.

Further, Friends of the Coast and New England Coalition request that the Board order that, if and when Next Era cures the inadequacies in its application, Next Era shall then resubmit the relevant portions of its application with appropriate notice and opportunity for adjudication by the ASLB and the parties.

October 20, 2010 Friends of the Coast New England Coalition by

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Raymond Shadis New England Coalition Post Office Box 98 Edgecomb, Maine 04556 207-882-7801 shadis@prexar.com 80