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. 

(Palisades Nuclear Plant) 

Docket No. 50-255-LA 

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, Don’t Waste 

Michigan, Michigan Safe Energy Future—Shoreline Chapter, and the Nuclear Energy 

Information Service (collectively, Petitioners).1 As discussed below, we affirm the Board’s 

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). 
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nature of Entergy’s 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 Board’s 

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 

vessel’s fracture toughness.3 Fracture toughness, which depends on the vessel’s chemical 

composition and its cumulative exposure to neutron radiation, is a measurement of a reactor 

pressure vessel’s ability to withstand a pressurized thermal shock event—where “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 values—called 

 

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”). 
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screening criteria—for 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 vessel’s transition from a crack-resistant to brittle state occurs.7 Our 

regulations use a temperature value—known 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 vessel—a 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. 
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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 vessel’s 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. 
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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 data”—this 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 “heat”20 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). 
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§ 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 Board’s 

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 Entergy’s Answer Opposing Petitioners’ Appeal of LBP-15-17 (June 29, 2015); NRC Staff 

Answer to Appeal of LBP-15-17 by Beyond Nuclear, Don’t Waste Michigan, Michigan Safe 

Energy Future—Shoreline Chapter, and the Nuclear Energy Information Service (June 29, 

2015) (Staff Answer). 
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II. DISCUSSION 

Petitioners have appealed the Board’s 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 Board’s 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 Board’s 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 Board’s contention 

admissibility decision, any of which they contend is sufficient to overturn the Board’s 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 

NRC’s 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. 
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requirements of one regulation are assertedly “weaker” than the other.37 And third, Petitioners 

challenge the Board’s determination regarding the portion of their contention related to the 

licensee’s 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 isn’t a 

threshold, there isn’t 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.”). 
- 9 - 

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 approval—mere 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 Board’s 

observation—with which we agree—that “[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. 
- 10 - 

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 licensee’s technical 

documentation to ensure that the requested amendment complies with our regulations—in 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 request—including the one at issue here—and 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. 
- 11 - 

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 vessel’s 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.
- 12 - 

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 licensee’s 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), rev’d, 

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. 
- 13 - 

statement of their expert.”61 Petitioners do not challenge the Board’s 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 guide—the guidance in question—that 

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 σ) 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 material’s 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 regulation’s consistency check does not rely on 

information that is unique to a particular [reactor pressure vessel], but instead on the chemical 
- 14 - 

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). 
- 15 - 

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 

Board’s 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 

Office of Commission Appellate Adjudication 

Mail Stop: O-7H4 

Washington, DC 20555-0001 

ocaamail@nrc.gov

U.S. Nuclear Regulatory Commission 

Office of the Secretary of the Commission 

Mail Stop O-16C1 

Washington, DC 20555-0001 

Hearing Docket 

hearingdocket@nrc.gov

U.S. Nuclear Regulatory Commission. 

Atomic Safety and Licensing Board Panel 

Mail Stop: T-3F23 

Washington, DC 20555-0001 

Ronald M. Spritzer, Chair 

Administrative Judge 

ronald.spritzer@nrc.gov

Gary S. Arnold, 

Administrative Judge 

gary.arnold@nrc.gov

Thomas J. Hirons 

Administrative Judge 

thomas.hirons@nrc.gov

Sachin Desai, Law Clerk 

sachin.desai@nrc.gov 

Nicole Pepperl, Law Clerk 

nicole.pepperl@nrc.gov 

U.S. Nuclear Regulatory Commission 

Office of the General Counsel 

Mail Stop O-15D21 

Washington, DC 20555-0001 

Mary Spencer, Esq. 

David Roth, Esq. 

Joseph Lindell, Esq. 

Sherwin Turk, Esq. 

Anita Ghosh, Esq. 

Daniel Straus, Esq. 

John Tibbetts, Paralegal 

Edward Williamson, Esq. 

mary.spencer@nrc.gov

david.roth@nrc.gov

joseph.lindell@nrc.gov

sherwin.turk@nrc.gov

anita.ghosh@nrc.gov 

daniel.straus@nrc.gov 

john.tibbetts@nrc.gov

edward.williamson@nrc.gov
Docket No. 50-255-LA 

COMMISSION MEMORANDUM AND ORDER CLI-15-22

2

Morgan, Lewis & Bockius LLP 

1111 Pennsylvania Avenue, NW 

Washington, DC 20004 

Raphael P. Kuyler, Esq. 

Paul Bessette, Esq. 

Ryan Lighty, Esq. 

rkuyler@morganlewis.com

pbessette@morganlewis.com

rlighty@morganlewis.com 

Beyond Nuclear, Don’t Waste Michigan, Michigan 

Safe Energy Future – Shoreline Chapter, Nuclear 

Energy Information Service “The Petitioners” 

316 N. Michigan St., Street, 520 

Toledo, OH 43604-5627 

Terry Lodge, Counsel for the Petitioners 

tjlodge50@yahoo.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