ML15313A433

Commission Memorandum and Order CLI-15-22

ML15313A433
Person / Time
Site:	Palisades
Issue date:	11/09/2015
From:	Annette Vietti-Cook NRC/SECY
To:
SECY RAS
References
50-255-LA, ASLBP 15-936-03-LA-BD01, CLI-15-22, RAS 28487
Download: ML15313A433 (17)

Text

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION COMMISSIONERS:

Stephen G. Burns, Chairman Kristine L. Svinicki William C. Ostendorff Jeff Baran In the Matter of:

ENTERGY NUCLEAR OPERATIONS, INC. Docket No. 50-255-LA (Palisades Nuclear Plant)

CLI-15-22 MEMORANDUM AND ORDER This proceeding stems from Entergy Nuclear Operations, Inc.s application to amend the operating license for the Palisades Nuclear Plant to allow for the use of an alternate method provided in our regulations for evaluating the fracture toughness of certain reactor pressure vessels. In LBP-15-17, the Atomic Safety and Licensing Board denied a petition to intervene and request for hearing challenging that application from Beyond Nuclear, Dont Waste Michigan, Michigan Safe Energy FutureShoreline Chapter, and the Nuclear Energy Information Service (collectively, Petitioners).1 As discussed below, we affirm the Boards decision.

I. PROCEDURAL AND TECHNICAL BACKGROUND As a general matter, Petitioners challenge the regulatory scheme that the NRC adopted to protect certain reactors from pressurized thermal shock events. Due to the highly technical 1 LBP-15-17, 81 NRC 753 (2015).

nature of Entergys license amendment request, we provide first a short summary of that regulatory scheme and the pressurized thermal shock phenomenon itself. A more detailed discussion of the technical issues and the relevant regulatory history is available in the Boards decision and in the Federal Register notices for our pressurized thermal shock regulations 10 C.F.R §§ 50.61 and 50.61a.2 The reactor pressure vessel in an operating pressurized water reactor is continuously exposed to neutron radiation from the fission reaction occurring inside the vessel, which over the life of the reactor causes embrittlement of the pressure vessel walls and decreases the vessels fracture toughness.3 Fracture toughness, which depends on the vessels chemical composition and its cumulative exposure to neutron radiation, is a measurement of a reactor pressure vessels ability to withstand a pressurized thermal shock eventwhere rapid cooling of the reactor vessel internal surface causes thermal stress on the reactor vessel, potentially leading to a breach of the reactor vessel wall.4 Recognizing the need to monitor fracture toughness, the NRC adopted regulatory requirements for pressurized water reactor licensees to implement programs to monitor the embrittlement of reactor pressure vessels.5 These surveillance programs provided material property data necessary to implement a regulatory scheme to protect reactor pressure vessels from failure due to pressurized thermal shock. In doing so, the NRC established valuescalled 2 See id. at 761-68; Analysis of Potential Pressurized Thermal Shock Events, Final Rule, 50 Fed. Reg. 29,937 (July 23, 1985) (1985 PTS Rule); Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, Final Rule, 75 Fed.

Reg. 13 (Jan. 4, 2010) (2010 PTS Rule).

3 1985 PTS Rule, 50 Fed. Reg. at 29,938.

4 Id.

5 See Fracture Toughness and Surveillance Program Requirements, Final Rule, 38 Fed. Reg.

19,012 (July 17, 1973) (establishing the program to monitor reactor pressure vessels in 10 C.F.R. pt. 50, app. H, Reactor Vessel Material Surveillance Program Requirements).

screening criteriafor fracture toughness. For materials with values above these screening criteria, licensees prepare fracture-toughness calculations to determine whether additional detailed plant-specific evaluations and possibly modifications to existing equipment, systems, and procedures are necessary.6 For materials with values below these screening criteria, no further analysis is required.

The screening criteria set temperature limits that measure the temperature range at which the reactor pressure vessels transition from a crack-resistant to brittle state occurs.7 Our regulations use a temperature valueknown as the reference temperature, which is defined as [t]he point at which steel transitions from the high-temperature, fracture-resistant-state, to the low-temperature, brittle state.8 The reference temperature provides a quantitative assessment of the toughness of the reactor pressure vessela higher reference temperature reflects a higher degree of brittleness.9 As the reactor experiences greater cumulative exposure to neutron radiation over its operating life, the reactor pressure vessel will become brittle at higher temperatures, thus increasing the likelihood that the vessel could fracture during a pressurized thermal shock event.10 6 1985 PTS Rule, 50 Fed. Reg. at 29,938.

7 See id.

8 LBP-15-17, 81 NRC at 762 (citing John B. Giessner, Division of Reactor Projects, Summary of the March 19, 2013, Public Meeting Webinar Regarding Palisades Nuclear Plant (Apr. 18, 2013), encl. 2 at 4 (ADAMS accession no. ML13108A336)) (internal citations omitted); id. at 762-63 (citing Division of Fuel, Engineering and Radiological Research, Office of Nuclear Regulatory Research, Technical Basis for Revision of the Pressurized Thermal Shock (PTS)

Screening Limit in the PTS Rule (10 CFR 50.61) Summary Report, NUREG-1806 (Aug. 2007),

at xxxiv (ML072830074 (package)) (quotation marks omitted) (Technical Basis)).

9 Id. at 762.

10 Id. at 763.

Since the promulgation of these screening criteria in 1985, the state of science and engineering knowledge in this area has increased dramatically; as a result, the NRC has determined that the risk of through-wall cracking due to a [pressurized thermal shock] event is much lower than previously estimated.11 Due in part to this improved understanding of the risk associated with such an event, the NRC issued a new rule, 10 C.F.R. § 50.61a.12 The new rule employs an updated embrittlement model to predict future reference temperatures across the

[reactor pressure vessel], which is then verified by existing surveillance data.13 In adopting the new rule, the NRC determined that, compared to the requirements of 10 C.F.R. § 50.61, the updated estimation procedures provide a better method for estimating the fracture toughness of reactor vessel materials over the lifetime of the plant.14 Moreover, the NRC concluded that the final rule provides reasonable assurance that licensees operating below the screening criteria could endure a [pressurized thermal shock] event without fracture of vessel materials, thus assuring integrity of the reactor pressure vessel.15 Licensees seeking to use the updated methodology in 10 C.F.R. § 50.61a must submit a license amendment request under 10 C.F.R. § 50.90.16 The request must provide information that includes: (1) calculations of the values of the material properties that characterize the reactor vessels resistance to fracture; (2) an examination and assessment of flaws discovered by American Society of Mechanical Engineers Code inspections; and (3) a comparison of the 11 2010 PTS Rule, 75 Fed. Reg. at 13. [T]he screening criteria in [10 C.F.R.] § 50.61 are unnecessarily conservative and may impose unnecessary burden on some licensees. Id.

12 Id.

13 LBP-15-17, 81 NRC at 765.

14 2010 PTS Rule, 75 Fed. Reg. at 18.

15 Id. at 22.

16 Id. at 18.

material property reference temperature values against the applicable screening criteria.17 To verify the calculations used to support the license amendment request, licensees must compare their calculations to heat-specific surveillance datathis verification is called the consistency check.18 The surveillance data used in the consistency check need not come from the same reactor pressure vessel that is the subject of the license amendment request.19 When it is available, licensees are required to use data from other reactors, called sister-plant data, provided the data are from material samples of the same heat20 as the material in the vessel for the reactor that is the subject of the license amendment request.21 In accordance with this rule, Entergy submitted a license amendment application for Palisades that provided the information requested in 10 C.F.R. § 50.61a.22 Petitioners thereafter requested a hearing challenging the application.23 In their hearing request, Petitioners expressed general concerns about the licensing framework that establishes the requirements to address pressurized thermal shock events.

Petitioners asserted particularly that the implementation of the new requirements in 10 C.F.R. 17 Id.; see 10 C.F.R. § 50.61a(a)(2)-(6).

18 LBP-15-17, 81 NRC at 766-67 (citing 10 C.F.R. § 50.61a(f)(6)(i) and 2010 PTS Rule, 75 Fed.

Reg. at 16).

19 See 10 C.F.R. §§ 50.61a(a)(10) and (f)(6)(i) (defining Surveillance data to include data from other plants and requiring licensees to consider this data under certain circumstances).

20 Here, the heat of a sample pertains to its material composition.

21 See id; see also 2010 PTS Rule, 75 Fed. Reg. at 16.

22 License Amendment Request to Implement 10 C.F.R. § 50.61a, Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events (July 29, 2014) (ML14211A524) (License Amendment Request). The Staff has yet to reach a decision on the license amendment request.

23 Amended Petition to Intervene and for a Public Adjudication Hearing of Entergy License Amendment Request for Authorization to Implement 10 CFR § 50.61a Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events (Dec. 8, 2014).

§ 50.61a at Palisades will introduce further non-conservative analytical assumptions into the troubled forty-three year operational history of Palisades.24 In reviewing Petitioners contention, the Board evaluated three bases provided by Petitioners to support their contention.25 Specifically, the Board evaluated Petitioners claims that the licensee cannot:

(1) provide reasonable assurance under 10 C.F.R. § 50.61a without obtaining or using additional data from the Palisades reactor pressure vessel; (2) consider, under 10 C.F.R.

§ 50.61a(f)(6), sister-plant data in addition to Palisades surveillance data; and (3) account for spatial variability in fluence across a reactor, given the current surveillance data available.26 The Board considered each of these claims as a separate contention.27 In each instance, the Board found that Petitioners either did not satisfy our contention admissibility standards or had impermissibly challenged our regulations.28 The Board therefore denied Petitioners request for hearing.29 Petitioners appealed.30 Entergy and the Staff oppose Petitioners appeal. 31 24 Intervenors 10 C.F.R. § 2.311(c) Notice of Appeal of Atomic Safety and Licensing Boards Denial of Petition to Intervene and Request for a Hearing on Entergy License Amendment Request for Authorization to Implement 10 CFR § 50.61a and Brief in Support (June 2, 2015),

at 3 (Appeal).

25 LBP-15-17, 81 NRC at 779.

26 Id.

27 Id. at 780.

28 See id. at 783-84, 789.

29 See id. at 792.

30 See Appeal.

31 Entergys Answer Opposing Petitioners Appeal of LBP-15-17 (June 29, 2015); NRC Staff Answer to Appeal of LBP-15-17 by Beyond Nuclear, Dont Waste Michigan, Michigan Safe Energy FutureShoreline Chapter, and the Nuclear Energy Information Service (June 29, 2015) (Staff Answer).

II. DISCUSSION Petitioners have appealed the Boards decision under 10 C.F.R. § 2.311(c), which provides for an appeal as of right on the question whether a request for hearing or petition to intervene should have been granted. Their appeal falls squarely within this rule. Our decision today assesses whether the Board erred in denying the petition to intervene and request for hearing for failure to proffer an admissible contention.32 In ruling on Petitioners appeal, we will defer to the Boards rulings on contention admissibility absent an error of law or abuse of discretion.33 As discussed below, we find no Board error and affirm the Boards decision.

Our strict by design contention admissibility standards focus our hearing process on disputes that can be resolved in adjudication.34 The Board provided an extensive discussion of the contention admissibility requirements that we do not repeat here.35 On appeal, Petitioners present three arguments challenging the Boards contention admissibility decision, any of which they contend is sufficient to overturn the Boards denial.

First, Petitioners argue that the Board did not consider the discretionary authority of the NRC Staff, specifically, the Director of the Office of Nuclear Reactor Regulation, over whether to allow a particular applicant to invoke 10 C.F.R. § 50.61a.36 Second, Petitioners challenge the NRCs regulatory approach to the fracture toughness of reactor pressure vessels, arguing that the NRC cannot reasonably maintain two regulations that address the same topic when the 32 The Board found that Petitioners had demonstrated standing. LBP-15-17, 81 NRC at 769-76.

We do not address that ruling here.

33 See, e.g., AmerGen Energy Co., LLC (Oyster Creek Nuclear Generating Station), CLI-06-24, 64 NRC 111, 121 (2006).

34 Dominion Nuclear Connecticut, Inc. (Millstone Nuclear Power Station, Unit 3), CLI-08-17, 68 NRC 231, 233 (2008) (citations omitted).

35 LBP-15-17, 81 NRC at 777-78.

36 Appeal at 18.

requirements of one regulation are assertedly weaker than the other.37 And third, Petitioners challenge the Boards determination regarding the portion of their contention related to the licensees use of sister-plant data to satisfy the requirements of 10 C.F.R. § 50.61a.38 Petitioners first two points are related. Before the Board, Petitioners argued that 10 C.F.R. § 50.61a clearly contemplates a discretionary determination by the Director of [the Office of Nuclear Reactor Regulation] wherein the Staff has some power to say no to the license amendment request.39 Similarly, Petitioners argued that the Board has the authority to direct the licensee to follow 10 C.F.R. § 50.61 instead of § 50.61a.40 In essence, Petitioners requested that the Board impose additional requirements on licensees by imposing new restrictions on the use of 10 C.F.R. § 50.61a that go beyond the requirements in the regulation.41 The Board rejected this argument, finding that 10 C.F.R. § 2.335 prohibits the Board from considering such a contention except under specific conditions not present here.42 On appeal, Petitioners rephrase their challenge to 10 C.F.R. § 50.61a and claim that the Board found that if the paperwork is properly completed, then the licensee is automatically allowed 37 Id. at 20.

38 See id. at 22-23.

39 Petitioners Combined Reply in Support of Amended Petition to Intervene and for a Public Adjudication Hearing of Entergy License Amendment Request for Authorization to Implement 10 CFR §50.61a, Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events (Jan. 20, 2015), at 3-4; Tr. at 132.

40 Tr. at 18.

41 See, e.g., id. at 34-35 (where Petitioners note that the problem is that there isnt a threshold, there isnt a limbo stick over which Entergy must leap to qualify to use

[10 C.F.R. 50.61a].).

42 See LBP-15-17, 81 NRC at 779-80 (Petitioners apparently want the Board to preclude Entergy from relying on Section 50.61a to avoid meeting the requirements of Section 50.61, but it is just such a deviation that Section 50.61a authorizes.).

to invoke 10 C.F.R. § 50.61a.43 At bottom, however, Petitioners have not raised a genuine dispute with the application because they have not challenged the license amendment request itself.

A license amendment request must provide sufficient documentation and analysis to show that the licensee has complied with the relevant requirements, thereby demonstrating that the amended license will continue to provide reasonable assurance of adequate protection of public health and safety.44 As we stated when we adopted the 2010 PTS Rule, licensees would only be allowed to implement § 50.61a upon NRC approvalmere submission of a license amendment request is not sufficient to take advantage of the new regulation.45 Petitioners argue that the NRC should consider whether rejecting the license amendment request would provide a superior reasonable assurance of protection of public health and safety.46 In making this request, Petitioners do not demonstrate an error of law or abuse of discretion on the part of the Board. Rather, they would have the NRC impose a standard that goes beyond what our regulations require. Petitioners do not refute the Boards observationwith which we agreethat [w]hen the Commission has determined that 43 Appeal at 18.

44 See, e.g., Duke Energy Co. (Catawba Nuclear Station, Units 1 and 2), LBP-04-32, 60 NRC 713, 720-21 (2004). The legal standards that apply in this proceeding are found in various NRC regulations. First, under 10 C.F.R. § 50.90, whenever a holder of a license wishes to amend the license, including technical specifications in the license, an application for amendment must be filed, fully describing the changes desired. Under 10 C.F.R. § 50.92(a),

determinations on whether to grant an applied-for license amendment are to be guided by the considerations that govern the issuance of initial licenses or construction permits to the extent applicable and appropriate. Both the common standards for licenses and construction permits in 10 C.F.R. § 50.40(a), and those specifically for issuance of operating licenses in 10 C.F.R.

§ 50.57(a)(3), provide that there must be reasonable assurance that the activities at issue will not endanger the health and safety of the public. Id.

45 2010 PTS Rule, 75 Fed. Reg. at 18.

46 Appeal at 18-19.

compliance with a regulation is sufficient to provide for reasonable assurance of public health and safety, a licensing board cannot impose requirements that exceed those in the regulation.47 In license amendment matters such as this, to approve the amendment the Staff must verify that reasonable assurance has been demonstrated by evaluating the licensees technical documentation to ensure that the requested amendment complies with our regulationsin this case, principally 10 C.F.R. § 50.61a. If the Staff determines that the licensee has satisfied our regulatory requirements, it then issues the requested license amendment. It is true that the Staff will apply engineering judgment as one consideration among many in determining whether the applicable regulatory requirements are satisfied. But once the Staff is satisfied that all such requirements are met, it is obliged to approve the request. And as we have noted in the past, regardless of whether a hearing request is granted, the NRC Staff performs a full safety review of every license amendment requestincluding the one at issue hereand no request is approved until all necessary public health and safety findings have been made.48 Petitioners claim that the standard in 10 C.F.R. § 50.61 is stronger than the standard in section 50.61a and that therefore it is legally anomalous for the NRC to conclude that both provide reasonable assurance.49 Here again, Petitioners neither articulate Board error nor raise a genuine dispute with the license amendment request. The existence of two regulations that provide alternative methods to demonstrate reasonable assurance is not, in and of itself, an anomaly. Petitioners did not provide a basis for their contention that two regulations cannot adequately address the same technical issue. The regulations at issue here, 10 C.F.R.

§§ 50.61 and 50.61a, prescribe alternate methods by which a licensee can demonstrate 47 LBP-15-17, 81 NRC at 789.

48 Dominion Nuclear Connecticut, Inc. (Millstone Nuclear Power Station, Unit 3), CLI-08-17, 68 NRC 231, 234 (2008).

49 Appeal at 20-21.

reasonable assurance of protection against pressurized thermal shock events. Through our rulemaking process, we have found that compliance with either regulation is sufficient to demonstrate reasonable assurance with respect to the fracture toughness of a reactor pressure vessel.50 At the time the NRC adopted the 2010 PTS Rule, we stated that after the promulgation of 10 C.F.R. § 50.61a the rule at 10 C.F.R. § 50.61 would remain in place.

Section 50.61a would not be required, but could be used by current licensees at their option.51 Thus, contrary to Petitioners claim, the NRC expressly put both regulations in place because compliance with either 10 C.F.R. § 50.61 or 10 C.F.R. § 50.61a provides reasonable assurance of a reactor pressure vessels ability to endure a pressurized thermal shock event.

Petitioners also argued before the Board that the licensee should be required to test metal coupons from inside the Palisades reactor pressure vessel.52 The Board found that in asking for testing of additional samples, Petitioners are asking the Board to demand more than

[10 C.F.R. § 50.61a] requires.53 The Board correctly found that 10 C.F.R. § 50.61a does not require a licensee to collect additional surveillance data from the subject plant; it requires the licensee to use any data that demonstrates the embrittlement trends for the materials, including surveillance programs at other plants with or without a surveillance program integrated under 10 C.F.R. Part 50, Appendix H.54 Petitioners argument that additional testing 50 Compare 1985 PTS Rule, 50 Fed. Reg. at 29,937 with 2010 PTS Rule, 75 Fed. Reg. at 13.

51 2010 PTS Rule, 75 Fed. Reg. at 22.

52 See, e.g., Petition at 11-12, 14-15.

53 LBP-15-17, 81 NRC at 783.

54 10 C.F.R. § 50.61a(a)(10); see also LBP-15-17, 81 NRC at 782-83.

should be required to demonstrate compliance with 10 C.F.R. § 50.61a is therefore an impermissible challenge to 10 C.F.R. § 50.61a.55 These claims fundamentally challenge the regulatory scheme that allows for the submission, NRC review and, if appropriate, approval of the license amendment request.

Challenges to our regulatory scheme are not permitted in adjudicatory proceedings absent a waiver.56 Petitioners have neither submitted a waiver request nor addressed the waiver criteria in 10 C.F.R. § 2.335.57 Moreover, the information in Petitioners pleadings does not support the granting of a waiver here.

Next, Petitioners claim that the Board erred in not admitting the portion of their contention challenging the use of sister-plant data in the license amendment request.58 The requirements in 10 C.F.R. § 50.61a require material samples used in a licensees consistency check to be from the same heat; there is no requirement that the sample come from the reactor pressure vessel subject to the license amendment request.59 Petitioners argue that the license amendment request lacked proof that the metals from the various [reactor pressure vessels] match.60 The Board found that [Petitioners expert] admits that the sister[-]plant data and Palisades samples are similar. [T]heir argument is without support and contradicts the 55 Petitioners raised a similar issue related to the frequency of surveillance testing at Palisades in a separate challenge to a different license amendment request. See Entergy Nuclear Operations, Inc. (Palisades Nuclear Plant), LBP-15-20, 81 NRC 829 (granting a hearing), revd, CLI-15-23, 82 NRC __ (Nov. 9, 2015) (slip op.).

56 10 C.F.R. § 2.335.

57 See Dominion Nuclear Connecticut, Inc. (Millstone Nuclear Power Station, Units 2 and 3),

CLI-05-24, 62 NRC 551, 559-60 (2005).

58 Appeal at 22-23.

59 10 C.F.R. § 50.61a(f)(6)(i)(A); see also LBP-15-17, 81 NRC at 788.

60 Appeal at 22-23.

statement of their expert.61 Petitioners do not challenge the Boards conclusion that there is no reason to doubt that the sister[-]plant material samples are the same heat or composition compared to the materials in the Palisades reactor pressure vessel.62 Petitioners here repackage their argument before the Board without asserting Board error, and we find none.

This portion of the contention is inadmissible for the reasons given by the Board.

Finally, Petitioners raise a variation of their argument below that NRC guidance supports a 20% deviation limit between data obtained from sister plants and Palisades, which would be mathematically implausible to satisfy, given a flux variation between sister plants that varies by 300%.63 The Board found that Petitioners argument lacked support because Petitioners had relied on a limit described in an NRC regulatory guidethe guidance in questionthat pertains to fluence modeling within a single reactor, rather than to a comparison of data between a particular reactor and sister plants.64 Thus, the Board found that the regulatory guide did not support Petitioners claims, and Petitioners therefore did not submit an admissible contention.65 61 LBP-15-17, 81 NRC at 791-92. Moreover, the Board noted, and we agree, that Petitioners inappropriately first raised this argument in their reply. Id.; see Louisiana Energy Services, L.P.

(National Enrichment Facility), CLI-04-25, 60 NRC 223, 225 (2004).

62 LBP-15-17, 81 NRC at 792.

63 Appeal at 23 (internal citations and quotation marks omitted).

64 LBP-15-17, 81 NRC at 789; see Office of Nuclear Regulatory Research, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, Regulatory Guide 1.190 (Mar. 2001), at 3, 31 (ML010890301) (explaining that the uncertainty of the fluence in the reactor pressure vessel must be 20% (1 ) or less for calculating various reference temperatures for the purposes of complying with 10 C.F.R. § 50.61.).

65 LBP-15-17, 81 NRC at 789; 10 C.F.R. § 2.309(f)(1)(v). Further, Petitioners focus on the variation in flux between sister plants is immaterial. With respect to this point, the Board correctly noted that [a]ny variation in flux is captured in the materials fluence measurement, because fluence is the integral of flux over time. Under [10 C.F.R. § 50.61(a)(f)(6)(i)], when the fluence of a material sample is known it must be used in the consistency check if it is of the appropriate chemical composition. The regulations consistency check does not rely on information that is unique to a particular [reactor pressure vessel], but instead on the chemical

Moreover, in considering this issue, the Board found that the use of a material sample in the consistency check is not dependent on its location inside [a reactor pressure vessel], or which [reactor pressure vessel] it comes from. From the standpoint of the consistency check, a material sample of the same fluence and material type is no different whether obtained from the Palisades [reactor pressure vessel] or a sister[-]plant [reactor pressure vessel].66 In other words, the Board found that Petitioners claims constituted an improper challenge to the requirements in 10 C.F.R. § 50.61a, which is prohibited by 10 C.F.R. § 2.335.67 Nothing in Petitioners appeal challenges these findings. The Board appropriately reviewed the support provided for the contention and determined that it did not apply to the circumstances presented here.68 In sum, Petitioners fundamentally challenge 10 C.F.R. § 50.61a, which is not permissible in this license amendment proceeding. In addition, Petitioners have not identified an adequately supported, genuine dispute with the license amendment application. For these reasons, we find that the Board appropriately denied Petitioners hearing request.

properties and fluence of the material samples. LBP-15-17, 81 NRC at 788 (citing 10 C.F.R.

§ 50.61a, equations 5-7); see also Staff Answer at 16 n.74 (explaining that differences in fluence are accounted for in the analysis.).

66 LBP-15-17, 81 NRC at 788.

67 Id.

68 See, e.g., USEC Inc. (American Centrifuge Plant), CLI-06-10, 63 NRC 451, 457 (2006)

(licensing boards are expected to examine cited materials to verify that they do, in fact, support a contention.) (citations omitted).

III. CONCLUSION Petitioners have identified no error of law or abuse of discretion on the part of the Board in LBP-15-17. For the foregoing reasons and for the reasons given by the Board, we affirm the Boards decision in LBP-15-17.

IT IS SO ORDERED.

For the Commission NRC SEAL

/RA/ .

Annette L. Vietti-Cook Secretary of the Commission Dated at Rockville, Maryland, this 9th day of November, 2015.

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of )

)

ENTERGY NUCLEAR OPERATIONS, INC. ) Docket No. 50-255-LA (Entergy) )

)

(Palisades Nuclear Plant) )

)

CERTIFICATE OF SERVICE I hereby certify that copies of the foregoing COMMISSION MEMORANDUM AND ORDER CLI-15-22 have been served upon the following persons by Electronic Information Exchange.

U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Office of Commission Appellate Adjudication Office of the Secretary of the Commission Mail Stop: O-7H4 Mail Stop O-16C1 Washington, DC 20555-0001 Washington, DC 20555-0001 ocaamail@nrc.gov Hearing Docket hearingdocket@nrc.gov U.S. Nuclear Regulatory Commission. U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Board Panel Office of the General Counsel Mail Stop: T-3F23 Mail Stop O-15D21 Washington, DC 20555-0001 Washington, DC 20555-0001 Mary Spencer, Esq.

Ronald M. Spritzer, Chair David Roth, Esq.

Administrative Judge Joseph Lindell, Esq.

ronald.spritzer@nrc.gov Sherwin Turk, Esq.

Anita Ghosh, Esq.

Gary S. Arnold, Daniel Straus, Esq.

Administrative Judge John Tibbetts, Paralegal gary.arnold@nrc.gov Edward Williamson, Esq.

mary.spencer@nrc.gov Thomas J. Hirons david.roth@nrc.gov Administrative Judge joseph.lindell@nrc.gov thomas.hirons@nrc.gov sherwin.turk@nrc.gov anita.ghosh@nrc.gov Sachin Desai, Law Clerk daniel.straus@nrc.gov sachin.desai@nrc.gov john.tibbetts@nrc.gov edward.williamson@nrc.gov Nicole Pepperl, Law Clerk nicole.pepperl@nrc.gov

Docket No. 50-255-LA COMMISSION MEMORANDUM AND ORDER CLI-15-22 Morgan, Lewis & Bockius LLP Beyond Nuclear, Dont Waste Michigan, Michigan 1111 Pennsylvania Avenue, NW Safe Energy Future - Shoreline Chapter, Nuclear Washington, DC 20004 Energy Information Service The Petitioners Raphael P. Kuyler, Esq. 316 N. Michigan St., Street, 520 Paul Bessette, Esq. Toledo, OH 43604-5627 Ryan Lighty, Esq. Terry Lodge, Counsel for the Petitioners rkuyler@morganlewis.com tjlodge50@yahoo.com pbessette@morganlewis.com rlighty@morganlewis.com Entergy William B. Glew, Jr., Esq.

440 Hamilton Avenue White Plains, NY 10601 wglew@entergy.com

[Original signed by Brian Newell ]

Office of the Secretary of the Commission Dated at Rockville, Maryland this 9th day of November, 2015 2

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION COMMISSIONERS:

Stephen G. Burns, Chairman Kristine L. Svinicki William C. Ostendorff Jeff Baran In the Matter of:

ENTERGY NUCLEAR OPERATIONS, INC. Docket No. 50-255-LA (Palisades Nuclear Plant)

CLI-15-22 MEMORANDUM AND ORDER This proceeding stems from Entergy Nuclear Operations, Inc.s application to amend the operating license for the Palisades Nuclear Plant to allow for the use of an alternate method provided in our regulations for evaluating the fracture toughness of certain reactor pressure vessels. In LBP-15-17, the Atomic Safety and Licensing Board denied a petition to intervene and request for hearing challenging that application from Beyond Nuclear, Dont Waste Michigan, Michigan Safe Energy FutureShoreline Chapter, and the Nuclear Energy Information Service (collectively, Petitioners).1 As discussed below, we affirm the Boards decision.

I. PROCEDURAL AND TECHNICAL BACKGROUND As a general matter, Petitioners challenge the regulatory scheme that the NRC adopted to protect certain reactors from pressurized thermal shock events. Due to the highly technical 1 LBP-15-17, 81 NRC 753 (2015).

nature of Entergys license amendment request, we provide first a short summary of that regulatory scheme and the pressurized thermal shock phenomenon itself. A more detailed discussion of the technical issues and the relevant regulatory history is available in the Boards decision and in the Federal Register notices for our pressurized thermal shock regulations 10 C.F.R §§ 50.61 and 50.61a.2 The reactor pressure vessel in an operating pressurized water reactor is continuously exposed to neutron radiation from the fission reaction occurring inside the vessel, which over the life of the reactor causes embrittlement of the pressure vessel walls and decreases the vessels fracture toughness.3 Fracture toughness, which depends on the vessels chemical composition and its cumulative exposure to neutron radiation, is a measurement of a reactor pressure vessels ability to withstand a pressurized thermal shock eventwhere rapid cooling of the reactor vessel internal surface causes thermal stress on the reactor vessel, potentially leading to a breach of the reactor vessel wall.4 Recognizing the need to monitor fracture toughness, the NRC adopted regulatory requirements for pressurized water reactor licensees to implement programs to monitor the embrittlement of reactor pressure vessels.5 These surveillance programs provided material property data necessary to implement a regulatory scheme to protect reactor pressure vessels from failure due to pressurized thermal shock. In doing so, the NRC established valuescalled 2 See id. at 761-68; Analysis of Potential Pressurized Thermal Shock Events, Final Rule, 50 Fed. Reg. 29,937 (July 23, 1985) (1985 PTS Rule); Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, Final Rule, 75 Fed.

Reg. 13 (Jan. 4, 2010) (2010 PTS Rule).

3 1985 PTS Rule, 50 Fed. Reg. at 29,938.

4 Id.

5 See Fracture Toughness and Surveillance Program Requirements, Final Rule, 38 Fed. Reg.

19,012 (July 17, 1973) (establishing the program to monitor reactor pressure vessels in 10 C.F.R. pt. 50, app. H, Reactor Vessel Material Surveillance Program Requirements).

screening criteriafor fracture toughness. For materials with values above these screening criteria, licensees prepare fracture-toughness calculations to determine whether additional detailed plant-specific evaluations and possibly modifications to existing equipment, systems, and procedures are necessary.6 For materials with values below these screening criteria, no further analysis is required.

The screening criteria set temperature limits that measure the temperature range at which the reactor pressure vessels transition from a crack-resistant to brittle state occurs.7 Our regulations use a temperature valueknown as the reference temperature, which is defined as [t]he point at which steel transitions from the high-temperature, fracture-resistant-state, to the low-temperature, brittle state.8 The reference temperature provides a quantitative assessment of the toughness of the reactor pressure vessela higher reference temperature reflects a higher degree of brittleness.9 As the reactor experiences greater cumulative exposure to neutron radiation over its operating life, the reactor pressure vessel will become brittle at higher temperatures, thus increasing the likelihood that the vessel could fracture during a pressurized thermal shock event.10 6 1985 PTS Rule, 50 Fed. Reg. at 29,938.

7 See id.

8 LBP-15-17, 81 NRC at 762 (citing John B. Giessner, Division of Reactor Projects, Summary of the March 19, 2013, Public Meeting Webinar Regarding Palisades Nuclear Plant (Apr. 18, 2013), encl. 2 at 4 (ADAMS accession no. ML13108A336)) (internal citations omitted); id. at 762-63 (citing Division of Fuel, Engineering and Radiological Research, Office of Nuclear Regulatory Research, Technical Basis for Revision of the Pressurized Thermal Shock (PTS)

Screening Limit in the PTS Rule (10 CFR 50.61) Summary Report, NUREG-1806 (Aug. 2007),

at xxxiv (ML072830074 (package)) (quotation marks omitted) (Technical Basis)).

9 Id. at 762.

10 Id. at 763.

Since the promulgation of these screening criteria in 1985, the state of science and engineering knowledge in this area has increased dramatically; as a result, the NRC has determined that the risk of through-wall cracking due to a [pressurized thermal shock] event is much lower than previously estimated.11 Due in part to this improved understanding of the risk associated with such an event, the NRC issued a new rule, 10 C.F.R. § 50.61a.12 The new rule employs an updated embrittlement model to predict future reference temperatures across the

[reactor pressure vessel], which is then verified by existing surveillance data.13 In adopting the new rule, the NRC determined that, compared to the requirements of 10 C.F.R. § 50.61, the updated estimation procedures provide a better method for estimating the fracture toughness of reactor vessel materials over the lifetime of the plant.14 Moreover, the NRC concluded that the final rule provides reasonable assurance that licensees operating below the screening criteria could endure a [pressurized thermal shock] event without fracture of vessel materials, thus assuring integrity of the reactor pressure vessel.15 Licensees seeking to use the updated methodology in 10 C.F.R. § 50.61a must submit a license amendment request under 10 C.F.R. § 50.90.16 The request must provide information that includes: (1) calculations of the values of the material properties that characterize the reactor vessels resistance to fracture; (2) an examination and assessment of flaws discovered by American Society of Mechanical Engineers Code inspections; and (3) a comparison of the 11 2010 PTS Rule, 75 Fed. Reg. at 13. [T]he screening criteria in [10 C.F.R.] § 50.61 are unnecessarily conservative and may impose unnecessary burden on some licensees. Id.

12 Id.

13 LBP-15-17, 81 NRC at 765.

14 2010 PTS Rule, 75 Fed. Reg. at 18.

15 Id. at 22.

16 Id. at 18.

material property reference temperature values against the applicable screening criteria.17 To verify the calculations used to support the license amendment request, licensees must compare their calculations to heat-specific surveillance datathis verification is called the consistency check.18 The surveillance data used in the consistency check need not come from the same reactor pressure vessel that is the subject of the license amendment request.19 When it is available, licensees are required to use data from other reactors, called sister-plant data, provided the data are from material samples of the same heat20 as the material in the vessel for the reactor that is the subject of the license amendment request.21 In accordance with this rule, Entergy submitted a license amendment application for Palisades that provided the information requested in 10 C.F.R. § 50.61a.22 Petitioners thereafter requested a hearing challenging the application.23 In their hearing request, Petitioners expressed general concerns about the licensing framework that establishes the requirements to address pressurized thermal shock events.

Petitioners asserted particularly that the implementation of the new requirements in 10 C.F.R. 17 Id.; see 10 C.F.R. § 50.61a(a)(2)-(6).

18 LBP-15-17, 81 NRC at 766-67 (citing 10 C.F.R. § 50.61a(f)(6)(i) and 2010 PTS Rule, 75 Fed.

Reg. at 16).

19 See 10 C.F.R. §§ 50.61a(a)(10) and (f)(6)(i) (defining Surveillance data to include data from other plants and requiring licensees to consider this data under certain circumstances).

20 Here, the heat of a sample pertains to its material composition.

21 See id; see also 2010 PTS Rule, 75 Fed. Reg. at 16.

22 License Amendment Request to Implement 10 C.F.R. § 50.61a, Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events (July 29, 2014) (ML14211A524) (License Amendment Request). The Staff has yet to reach a decision on the license amendment request.

23 Amended Petition to Intervene and for a Public Adjudication Hearing of Entergy License Amendment Request for Authorization to Implement 10 CFR § 50.61a Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events (Dec. 8, 2014).

§ 50.61a at Palisades will introduce further non-conservative analytical assumptions into the troubled forty-three year operational history of Palisades.24 In reviewing Petitioners contention, the Board evaluated three bases provided by Petitioners to support their contention.25 Specifically, the Board evaluated Petitioners claims that the licensee cannot:

(1) provide reasonable assurance under 10 C.F.R. § 50.61a without obtaining or using additional data from the Palisades reactor pressure vessel; (2) consider, under 10 C.F.R.

§ 50.61a(f)(6), sister-plant data in addition to Palisades surveillance data; and (3) account for spatial variability in fluence across a reactor, given the current surveillance data available.26 The Board considered each of these claims as a separate contention.27 In each instance, the Board found that Petitioners either did not satisfy our contention admissibility standards or had impermissibly challenged our regulations.28 The Board therefore denied Petitioners request for hearing.29 Petitioners appealed.30 Entergy and the Staff oppose Petitioners appeal. 31 24 Intervenors 10 C.F.R. § 2.311(c) Notice of Appeal of Atomic Safety and Licensing Boards Denial of Petition to Intervene and Request for a Hearing on Entergy License Amendment Request for Authorization to Implement 10 CFR § 50.61a and Brief in Support (June 2, 2015),

at 3 (Appeal).

25 LBP-15-17, 81 NRC at 779.

26 Id.

27 Id. at 780.

28 See id. at 783-84, 789.

29 See id. at 792.

30 See Appeal.

31 Entergys Answer Opposing Petitioners Appeal of LBP-15-17 (June 29, 2015); NRC Staff Answer to Appeal of LBP-15-17 by Beyond Nuclear, Dont Waste Michigan, Michigan Safe Energy FutureShoreline Chapter, and the Nuclear Energy Information Service (June 29, 2015) (Staff Answer).

II. DISCUSSION Petitioners have appealed the Boards decision under 10 C.F.R. § 2.311(c), which provides for an appeal as of right on the question whether a request for hearing or petition to intervene should have been granted. Their appeal falls squarely within this rule. Our decision today assesses whether the Board erred in denying the petition to intervene and request for hearing for failure to proffer an admissible contention.32 In ruling on Petitioners appeal, we will defer to the Boards rulings on contention admissibility absent an error of law or abuse of discretion.33 As discussed below, we find no Board error and affirm the Boards decision.

Our strict by design contention admissibility standards focus our hearing process on disputes that can be resolved in adjudication.34 The Board provided an extensive discussion of the contention admissibility requirements that we do not repeat here.35 On appeal, Petitioners present three arguments challenging the Boards contention admissibility decision, any of which they contend is sufficient to overturn the Boards denial.

First, Petitioners argue that the Board did not consider the discretionary authority of the NRC Staff, specifically, the Director of the Office of Nuclear Reactor Regulation, over whether to allow a particular applicant to invoke 10 C.F.R. § 50.61a.36 Second, Petitioners challenge the NRCs regulatory approach to the fracture toughness of reactor pressure vessels, arguing that the NRC cannot reasonably maintain two regulations that address the same topic when the 32 The Board found that Petitioners had demonstrated standing. LBP-15-17, 81 NRC at 769-76.

We do not address that ruling here.

33 See, e.g., AmerGen Energy Co., LLC (Oyster Creek Nuclear Generating Station), CLI-06-24, 64 NRC 111, 121 (2006).

34 Dominion Nuclear Connecticut, Inc. (Millstone Nuclear Power Station, Unit 3), CLI-08-17, 68 NRC 231, 233 (2008) (citations omitted).

35 LBP-15-17, 81 NRC at 777-78.

36 Appeal at 18.

requirements of one regulation are assertedly weaker than the other.37 And third, Petitioners challenge the Boards determination regarding the portion of their contention related to the licensees use of sister-plant data to satisfy the requirements of 10 C.F.R. § 50.61a.38 Petitioners first two points are related. Before the Board, Petitioners argued that 10 C.F.R. § 50.61a clearly contemplates a discretionary determination by the Director of [the Office of Nuclear Reactor Regulation] wherein the Staff has some power to say no to the license amendment request.39 Similarly, Petitioners argued that the Board has the authority to direct the licensee to follow 10 C.F.R. § 50.61 instead of § 50.61a.40 In essence, Petitioners requested that the Board impose additional requirements on licensees by imposing new restrictions on the use of 10 C.F.R. § 50.61a that go beyond the requirements in the regulation.41 The Board rejected this argument, finding that 10 C.F.R. § 2.335 prohibits the Board from considering such a contention except under specific conditions not present here.42 On appeal, Petitioners rephrase their challenge to 10 C.F.R. § 50.61a and claim that the Board found that if the paperwork is properly completed, then the licensee is automatically allowed 37 Id. at 20.

38 See id. at 22-23.

39 Petitioners Combined Reply in Support of Amended Petition to Intervene and for a Public Adjudication Hearing of Entergy License Amendment Request for Authorization to Implement 10 CFR §50.61a, Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events (Jan. 20, 2015), at 3-4; Tr. at 132.

40 Tr. at 18.

41 See, e.g., id. at 34-35 (where Petitioners note that the problem is that there isnt a threshold, there isnt a limbo stick over which Entergy must leap to qualify to use

[10 C.F.R. 50.61a].).

42 See LBP-15-17, 81 NRC at 779-80 (Petitioners apparently want the Board to preclude Entergy from relying on Section 50.61a to avoid meeting the requirements of Section 50.61, but it is just such a deviation that Section 50.61a authorizes.).

to invoke 10 C.F.R. § 50.61a.43 At bottom, however, Petitioners have not raised a genuine dispute with the application because they have not challenged the license amendment request itself.

A license amendment request must provide sufficient documentation and analysis to show that the licensee has complied with the relevant requirements, thereby demonstrating that the amended license will continue to provide reasonable assurance of adequate protection of public health and safety.44 As we stated when we adopted the 2010 PTS Rule, licensees would only be allowed to implement § 50.61a upon NRC approvalmere submission of a license amendment request is not sufficient to take advantage of the new regulation.45 Petitioners argue that the NRC should consider whether rejecting the license amendment request would provide a superior reasonable assurance of protection of public health and safety.46 In making this request, Petitioners do not demonstrate an error of law or abuse of discretion on the part of the Board. Rather, they would have the NRC impose a standard that goes beyond what our regulations require. Petitioners do not refute the Boards observationwith which we agreethat [w]hen the Commission has determined that 43 Appeal at 18.

44 See, e.g., Duke Energy Co. (Catawba Nuclear Station, Units 1 and 2), LBP-04-32, 60 NRC 713, 720-21 (2004). The legal standards that apply in this proceeding are found in various NRC regulations. First, under 10 C.F.R. § 50.90, whenever a holder of a license wishes to amend the license, including technical specifications in the license, an application for amendment must be filed, fully describing the changes desired. Under 10 C.F.R. § 50.92(a),

determinations on whether to grant an applied-for license amendment are to be guided by the considerations that govern the issuance of initial licenses or construction permits to the extent applicable and appropriate. Both the common standards for licenses and construction permits in 10 C.F.R. § 50.40(a), and those specifically for issuance of operating licenses in 10 C.F.R.

§ 50.57(a)(3), provide that there must be reasonable assurance that the activities at issue will not endanger the health and safety of the public. Id.

45 2010 PTS Rule, 75 Fed. Reg. at 18.

46 Appeal at 18-19.

compliance with a regulation is sufficient to provide for reasonable assurance of public health and safety, a licensing board cannot impose requirements that exceed those in the regulation.47 In license amendment matters such as this, to approve the amendment the Staff must verify that reasonable assurance has been demonstrated by evaluating the licensees technical documentation to ensure that the requested amendment complies with our regulationsin this case, principally 10 C.F.R. § 50.61a. If the Staff determines that the licensee has satisfied our regulatory requirements, it then issues the requested license amendment. It is true that the Staff will apply engineering judgment as one consideration among many in determining whether the applicable regulatory requirements are satisfied. But once the Staff is satisfied that all such requirements are met, it is obliged to approve the request. And as we have noted in the past, regardless of whether a hearing request is granted, the NRC Staff performs a full safety review of every license amendment requestincluding the one at issue hereand no request is approved until all necessary public health and safety findings have been made.48 Petitioners claim that the standard in 10 C.F.R. § 50.61 is stronger than the standard in section 50.61a and that therefore it is legally anomalous for the NRC to conclude that both provide reasonable assurance.49 Here again, Petitioners neither articulate Board error nor raise a genuine dispute with the license amendment request. The existence of two regulations that provide alternative methods to demonstrate reasonable assurance is not, in and of itself, an anomaly. Petitioners did not provide a basis for their contention that two regulations cannot adequately address the same technical issue. The regulations at issue here, 10 C.F.R.

§§ 50.61 and 50.61a, prescribe alternate methods by which a licensee can demonstrate 47 LBP-15-17, 81 NRC at 789.

48 Dominion Nuclear Connecticut, Inc. (Millstone Nuclear Power Station, Unit 3), CLI-08-17, 68 NRC 231, 234 (2008).

49 Appeal at 20-21.

reasonable assurance of protection against pressurized thermal shock events. Through our rulemaking process, we have found that compliance with either regulation is sufficient to demonstrate reasonable assurance with respect to the fracture toughness of a reactor pressure vessel.50 At the time the NRC adopted the 2010 PTS Rule, we stated that after the promulgation of 10 C.F.R. § 50.61a the rule at 10 C.F.R. § 50.61 would remain in place.

Section 50.61a would not be required, but could be used by current licensees at their option.51 Thus, contrary to Petitioners claim, the NRC expressly put both regulations in place because compliance with either 10 C.F.R. § 50.61 or 10 C.F.R. § 50.61a provides reasonable assurance of a reactor pressure vessels ability to endure a pressurized thermal shock event.

Petitioners also argued before the Board that the licensee should be required to test metal coupons from inside the Palisades reactor pressure vessel.52 The Board found that in asking for testing of additional samples, Petitioners are asking the Board to demand more than

[10 C.F.R. § 50.61a] requires.53 The Board correctly found that 10 C.F.R. § 50.61a does not require a licensee to collect additional surveillance data from the subject plant; it requires the licensee to use any data that demonstrates the embrittlement trends for the materials, including surveillance programs at other plants with or without a surveillance program integrated under 10 C.F.R. Part 50, Appendix H.54 Petitioners argument that additional testing 50 Compare 1985 PTS Rule, 50 Fed. Reg. at 29,937 with 2010 PTS Rule, 75 Fed. Reg. at 13.

51 2010 PTS Rule, 75 Fed. Reg. at 22.

52 See, e.g., Petition at 11-12, 14-15.

53 LBP-15-17, 81 NRC at 783.

54 10 C.F.R. § 50.61a(a)(10); see also LBP-15-17, 81 NRC at 782-83.

should be required to demonstrate compliance with 10 C.F.R. § 50.61a is therefore an impermissible challenge to 10 C.F.R. § 50.61a.55 These claims fundamentally challenge the regulatory scheme that allows for the submission, NRC review and, if appropriate, approval of the license amendment request.

Challenges to our regulatory scheme are not permitted in adjudicatory proceedings absent a waiver.56 Petitioners have neither submitted a waiver request nor addressed the waiver criteria in 10 C.F.R. § 2.335.57 Moreover, the information in Petitioners pleadings does not support the granting of a waiver here.

Next, Petitioners claim that the Board erred in not admitting the portion of their contention challenging the use of sister-plant data in the license amendment request.58 The requirements in 10 C.F.R. § 50.61a require material samples used in a licensees consistency check to be from the same heat; there is no requirement that the sample come from the reactor pressure vessel subject to the license amendment request.59 Petitioners argue that the license amendment request lacked proof that the metals from the various [reactor pressure vessels] match.60 The Board found that [Petitioners expert] admits that the sister[-]plant data and Palisades samples are similar. [T]heir argument is without support and contradicts the 55 Petitioners raised a similar issue related to the frequency of surveillance testing at Palisades in a separate challenge to a different license amendment request. See Entergy Nuclear Operations, Inc. (Palisades Nuclear Plant), LBP-15-20, 81 NRC 829 (granting a hearing), revd, CLI-15-23, 82 NRC __ (Nov. 9, 2015) (slip op.).

56 10 C.F.R. § 2.335.

57 See Dominion Nuclear Connecticut, Inc. (Millstone Nuclear Power Station, Units 2 and 3),

CLI-05-24, 62 NRC 551, 559-60 (2005).

58 Appeal at 22-23.

59 10 C.F.R. § 50.61a(f)(6)(i)(A); see also LBP-15-17, 81 NRC at 788.

60 Appeal at 22-23.

statement of their expert.61 Petitioners do not challenge the Boards conclusion that there is no reason to doubt that the sister[-]plant material samples are the same heat or composition compared to the materials in the Palisades reactor pressure vessel.62 Petitioners here repackage their argument before the Board without asserting Board error, and we find none.

This portion of the contention is inadmissible for the reasons given by the Board.

Finally, Petitioners raise a variation of their argument below that NRC guidance supports a 20% deviation limit between data obtained from sister plants and Palisades, which would be mathematically implausible to satisfy, given a flux variation between sister plants that varies by 300%.63 The Board found that Petitioners argument lacked support because Petitioners had relied on a limit described in an NRC regulatory guidethe guidance in questionthat pertains to fluence modeling within a single reactor, rather than to a comparison of data between a particular reactor and sister plants.64 Thus, the Board found that the regulatory guide did not support Petitioners claims, and Petitioners therefore did not submit an admissible contention.65 61 LBP-15-17, 81 NRC at 791-92. Moreover, the Board noted, and we agree, that Petitioners inappropriately first raised this argument in their reply. Id.; see Louisiana Energy Services, L.P.

(National Enrichment Facility), CLI-04-25, 60 NRC 223, 225 (2004).

62 LBP-15-17, 81 NRC at 792.

63 Appeal at 23 (internal citations and quotation marks omitted).

64 LBP-15-17, 81 NRC at 789; see Office of Nuclear Regulatory Research, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, Regulatory Guide 1.190 (Mar. 2001), at 3, 31 (ML010890301) (explaining that the uncertainty of the fluence in the reactor pressure vessel must be 20% (1 ) or less for calculating various reference temperatures for the purposes of complying with 10 C.F.R. § 50.61.).

65 LBP-15-17, 81 NRC at 789; 10 C.F.R. § 2.309(f)(1)(v). Further, Petitioners focus on the variation in flux between sister plants is immaterial. With respect to this point, the Board correctly noted that [a]ny variation in flux is captured in the materials fluence measurement, because fluence is the integral of flux over time. Under [10 C.F.R. § 50.61(a)(f)(6)(i)], when the fluence of a material sample is known it must be used in the consistency check if it is of the appropriate chemical composition. The regulations consistency check does not rely on information that is unique to a particular [reactor pressure vessel], but instead on the chemical

Moreover, in considering this issue, the Board found that the use of a material sample in the consistency check is not dependent on its location inside [a reactor pressure vessel], or which [reactor pressure vessel] it comes from. From the standpoint of the consistency check, a material sample of the same fluence and material type is no different whether obtained from the Palisades [reactor pressure vessel] or a sister[-]plant [reactor pressure vessel].66 In other words, the Board found that Petitioners claims constituted an improper challenge to the requirements in 10 C.F.R. § 50.61a, which is prohibited by 10 C.F.R. § 2.335.67 Nothing in Petitioners appeal challenges these findings. The Board appropriately reviewed the support provided for the contention and determined that it did not apply to the circumstances presented here.68 In sum, Petitioners fundamentally challenge 10 C.F.R. § 50.61a, which is not permissible in this license amendment proceeding. In addition, Petitioners have not identified an adequately supported, genuine dispute with the license amendment application. For these reasons, we find that the Board appropriately denied Petitioners hearing request.

properties and fluence of the material samples. LBP-15-17, 81 NRC at 788 (citing 10 C.F.R.

§ 50.61a, equations 5-7); see also Staff Answer at 16 n.74 (explaining that differences in fluence are accounted for in the analysis.).

66 LBP-15-17, 81 NRC at 788.

67 Id.

68 See, e.g., USEC Inc. (American Centrifuge Plant), CLI-06-10, 63 NRC 451, 457 (2006)

(licensing boards are expected to examine cited materials to verify that they do, in fact, support a contention.) (citations omitted).

III. CONCLUSION Petitioners have identified no error of law or abuse of discretion on the part of the Board in LBP-15-17. For the foregoing reasons and for the reasons given by the Board, we affirm the Boards decision in LBP-15-17.

IT IS SO ORDERED.

For the Commission NRC SEAL

/RA/ .

Annette L. Vietti-Cook Secretary of the Commission Dated at Rockville, Maryland, this 9th day of November, 2015.

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of )

)

ENTERGY NUCLEAR OPERATIONS, INC. ) Docket No. 50-255-LA (Entergy) )

)

(Palisades Nuclear Plant) )

)

CERTIFICATE OF SERVICE I hereby certify that copies of the foregoing COMMISSION MEMORANDUM AND ORDER CLI-15-22 have been served upon the following persons by Electronic Information Exchange.

U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Office of Commission Appellate Adjudication Office of the Secretary of the Commission Mail Stop: O-7H4 Mail Stop O-16C1 Washington, DC 20555-0001 Washington, DC 20555-0001 ocaamail@nrc.gov Hearing Docket hearingdocket@nrc.gov U.S. Nuclear Regulatory Commission. U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Board Panel Office of the General Counsel Mail Stop: T-3F23 Mail Stop O-15D21 Washington, DC 20555-0001 Washington, DC 20555-0001 Mary Spencer, Esq.

Ronald M. Spritzer, Chair David Roth, Esq.

Administrative Judge Joseph Lindell, Esq.

ronald.spritzer@nrc.gov Sherwin Turk, Esq.

Anita Ghosh, Esq.

Gary S. Arnold, Daniel Straus, Esq.

Administrative Judge John Tibbetts, Paralegal gary.arnold@nrc.gov Edward Williamson, Esq.

mary.spencer@nrc.gov Thomas J. Hirons david.roth@nrc.gov Administrative Judge joseph.lindell@nrc.gov thomas.hirons@nrc.gov sherwin.turk@nrc.gov anita.ghosh@nrc.gov Sachin Desai, Law Clerk daniel.straus@nrc.gov sachin.desai@nrc.gov john.tibbetts@nrc.gov edward.williamson@nrc.gov Nicole Pepperl, Law Clerk nicole.pepperl@nrc.gov

Docket No. 50-255-LA COMMISSION MEMORANDUM AND ORDER CLI-15-22 Morgan, Lewis & Bockius LLP Beyond Nuclear, Dont Waste Michigan, Michigan 1111 Pennsylvania Avenue, NW Safe Energy Future - Shoreline Chapter, Nuclear Washington, DC 20004 Energy Information Service The Petitioners Raphael P. Kuyler, Esq. 316 N. Michigan St., Street, 520 Paul Bessette, Esq. Toledo, OH 43604-5627 Ryan Lighty, Esq. Terry Lodge, Counsel for the Petitioners rkuyler@morganlewis.com tjlodge50@yahoo.com pbessette@morganlewis.com rlighty@morganlewis.com Entergy William B. Glew, Jr., Esq.

440 Hamilton Avenue White Plains, NY 10601 wglew@entergy.com

[Original signed by Brian Newell ]

Office of the Secretary of the Commission Dated at Rockville, Maryland this 9th day of November, 2015 2