ML15153B263

Notice of 10 CFR 2.311 Appeal and Brief in Support

ML15153B263
Person / Time
Site:	Palisades
Issue date:	06/02/2015
From:	Lodge T J Beyond Nuclear, Don't Waste Michigan, Michigan Safe Energy Future - Shoreline Chapter (MSEF), Nuclear Energy Information Service
To:	NRC/OCM
SECY RAS
References
50-255-LA, ASLBP 15-936-03-LA-BD01, RAS 27884
Download: ML15153B263 (30)

Text

UNITED STATES OF AMERICANUCLEAR REGULATORY COMMISSIONBefore the CommissionIn the Matter of: Entergy Nuclear Operations, Inc.(Palisades Nuclear Plant)Operating License Amendment Request) Docket No. 50-255)June 2, 2015)) *****INTERVENORS' 10 C.F.R. § 2.311( c) NOTICE OF APPEAL OF ATOMIC SAFETY AND LICENSING BOARD'S DENIALOF PETITION TO INTERVENE AND REQUEST FOR A HEARINGON ENTERGY LICENSE AMENDMENT REQUEST FOR AUTHORIZATIONTO IMPLEMENT 10 CFR § 50.61a AND BRIEF IN SUPPORTTerry J. Lodge (OH #0029271)316 N. Michigan St., Ste. 520Toledo, OH 43604-5627(419) 255-7552Fax (419) 255-7552tjlodge50@yahoo.comCounsel for Petitioners TABLE OF CONTENTSTable of Authorities iiI. Introduction 1II. Factual and Procedural Background 3A. The 1985 PTS Rule And Embrittlement Screening Program (10 C.F.R. § 50.61) 3B. The Alternate PTS Rule And Embrittlement Screening Program (10 C.F.R.§ 50.61a) 7C. Invocation Of The Alternate PTS Rule 10D. Petitioners' Objections To Entergy License AmendmentRequest(LAR) Invoking Alternate PTS Rule 12III. Argument 18A. The ASLB Erroneously Found The Decision Allowing Entergy To Invoke10 C.F.R. § 50.61a To Be Nondiscretionary 18B. 'Reasonable Assurance' Cannot Apply Alike To Two Regulations AddressingThe Same Subject When One Is Deemed To Be Weaker Than The Other 20C. Variabilities In Sister Plant Data Erroneously Allowed InappropriateComparisons 22IV. Conclusion 22Certificate of Service 25-i-TABLE OF AUTHORITIESCasesAmerGen Energy Co., LLC (Oyster Creek Nuclear Generating Station), LBP-07-17, 66 NRC 327, 340 (2007), aff'd, CLI-09-07, 69 NRC 235, 263 (2009) 21Duke Power Co. (Catawba Nuclear Station, Units 1 & 2), LBP-82-116, 16 NRC 1937,1946 (1982)21Matter of Entergy Nuclear Generation Co., et al. (Pilgrim Nuclear Power Station), 50-293-LR (ASLB Oct. 16, 2006), 2006 WL 480114223Power Authority of the State of New York, et al. (James FitzPatrick Nuclear PowerPlant; Indian Point Nuclear Generating Unit 3), CLI-00-22, 52 NRC 266, 295 (2000) 23Statutes42 U.S.C. § 2232(a)20Regulations10 C.F.R. § 2.309 2310 C.F.R. § 2.311 110 C.F.R. § 50.572010 C.F.R. § 50.61 1, 2, 3, 4, 6, 7, 8, 9, 12, 15, 16, 18, 20, 2210 C.F.R. § 50.61a 1, 2, 3, 7, 8, 9, 10, 11, 12, 14, 15, 16, 18, 19, 20, 21, 2210 C.F.R. § 50.901010 C.F.R. § 50.92 2, 13-ii-UNITED STATES OF AMERICANUCLEAR REGULATORY COMMISSIONBefore the CommissionIn the Matter of:Entergy Nuclear Operations, Inc.(Palisades Nuclear Plant)Operating License Amendment Request) Docket No. 50-255)June 2, 2015))PETITIONERS' 10 C.F.R. § 2.311( c) NOTICE OF APPEAL OF ATOMIC SAFETY AND LICENSING BOARD'S DENIALOF 'PETITION TO INTERVENE AND REQUEST FOR A HEARINGON ENTERGY LICENSE AMENDMENT REQUEST FOR AUTHORIZATIONTO IMPLEMENT 10 C.F.R. § 50.61a'Beyond Nuclear, Don't Waste Michigan, Michigan Safe Energy Future - ShorelineChapter (Shoreline), and the Nuclear Energy Information Service (NEIS) (collectively"Petitioners"), by and through counsel, pursuant to 10 C.F.R. § 2.311(c), hereby give notice oftheir appeal to the U.S. Nuclear Regulatory Commission ("Commission") for review of theAtomic Safety and Licensing Board's ("ASLB") "Memorandum and Order (Ruling on Petition toIntervene and Request for a Hearing", LBP-15-17 (May 8, 2015) wherein the ASLB deniedPetitioners' "Petition to Intervene and for a Public Adjudication Hearing of Entergy LicenseAmendment Request for Authorization to Implement 10 CFR § 50.61a, 'Alternate FractureToughness Requirements for Protection Against Pressurized Thermal Shock Events.'" According to 10 C.F.R. § 2.311( c), "An order denying a petition to intervene, and/orrequest for hearing . . . is appealable by the requestor/petitioner on the question as to whether therequest and/or petition should have been granted." Petitioners intend to urge on appeal that their petition to intervene and request for a hearing should have been granted. /s/ Terry J. Lodge Terry J. Lodge (OH #0029271)316 N. Michigan St., Ste. 520Toledo, OH 43604-5627(419) 255-7552Fax (419) 255-7552Tjlodge50@yahoo.comCounsel for Petitioners UNITED STATES OF AMERICANUCLEAR REGULATORY COMMISSIONBefore the CommissionIn the Matter of:Entergy Nuclear Operations, Inc.(Palisades Nuclear Plant)Operating License Amendment Request) Docket No. 50-255 )June 2, 2015))BRIEF IN SUPPORT OF PETITIONERS'10 C.F.R. § 2.311( c) APPEAL OF ATOMIC SAFETY ANDLICENSING BOARD'S DENIAL OF 'PETITION TO INTERVENEAND REQUEST FOR A HEARING ON ENTERGY LICENSE AMENDMENTREQUEST FOR AUTHORIZATION TO IMPLEMENT 10 C.F.R. § 50.61a'I. IntroductionThis proceeding concerns Entergy Nuclear Operations, Inc.'s ("Entergy's") request toamend the operating license for the Palisades nuclear plant ("Palisades"). Palisades is a singlepressurized water reactor ("PWR") facility located on the eastern shore of Lake Michigan, fivemiles south of South Haven, Michigan. The requested amendment would permit Entergy touse an alternate method to evaluate the minimum fracture toughness required by the Palisadesreactor pressure vessel (RPV) to safely withstand a pressurized thermal shock (PTS) event.That alternate method is set forth in an agency regulation, "Alternate fracture toughnessrequirements for protection against pressurized thermal shock events." In an operating nuclearpower plant, the reactor vessel is continuously exposed to neutrons from fission reactionsoccurring inside the vessel. Over time, this neutron radiation embrittles the RPV walls, makingthem less able to resist fracturing, i.e., "fracture toughness" decreases. If there is a flaw in a reactor vessel wall that is embrittled due to neutron exposure, certain events can cause the flaw topropagate through the wall, resulting in a breach of the RPV and a possible accident. Ofsignificant concern is a pressurized thermal shock, or "PTS," event, which is "characterized by arapid cooling (i.e., thermal shock) of the internal RPV surface and downcomer, which may befollowed by repressurization of the RPV." The possible triggers of a PTS event include "a pipe1break or stuck-open valve in the primary pressure circuit," or "a break of the main steam line." 2On September 30, 2014, the NRC Staff (the Staff) published notice of Entergy's LAR,and concluded that the LAR presents "no significant hazards consideration" under 10 C.F.R. §50.92( c). In response to the LAR notice, Petitioners filed the instant petition to intervene andrequest for a hearing. 3Division of Fuel, Engineering and Radiological Research, Office of Nuclear Regulatory1Research, Technical Basis for Revision of the Pressurized Thermal Shock (PTS) ScreeningLimit in the PTS Rule (10 CFR 50.61) Summary Report, NUREG-1806 at xix (Aug. 2007), athttp://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1806/v1/ (hereinafter "AlternatePTS Rule Technical Basis Report"). Id. at xix; see also "Alternate Fracture Toughness Requirements for Protection Against2Pressurized Thermal Shock Events, Final Rule," 75 Fed. Reg. 13, 14 (Jan. 4, 2010). Duringthese scenarios, "the water level in the core drops as a result of" depressurization or leaks.Alternate PTS Rule Technical Basis Report at xix. Emergency makeup water is then added to thereactor cooling loop, either manually or automatically, to keep the reactor core covered withwater. Id. As the makeup water is much colder than the water in the reactor, a rapid cooling ofthe outside reactor wall results. Id. For over-embrittled RPVs, the temperature shock "could besufficient to initiate a running crack, which could propagate all the way through the vessel wall."Id. As the reactor is still producing heat, even in a shutdown mode, the RPV could re-pressurize,adding additional stress to the already-propagating crack. See id. at xix, xxiv, xxv ("A majorcontributor to the risk-significance of [certain PTS events] is the return to full system pressure"after cold makeup water is introduced. This could occur, for example, when a stuck-open valverecloses)."Amended Petition to Intervene and for a Public Adjudication Hearing of Entergy3License Amendment Request for Authorization to Implement 10 CFR §50.61a, 'AlternateFracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events'"

Petitioners' statement of their contention is:The licensing framework that the NRC is applying to allow Palisades to continueto operate until August 2017 includes both non-conservative analytical changes andmathematically dubious comparisons to allegedly similar "sister" reactor vessels.Palisades' neutron embrittlement dilemma continues to worsen as the plant ages, andPalisades has repeatedly requested life extensions which have ignored and deferredworsening embrittlement characteristics of the RPV for decades. Presently, Entergy plansto deviate from the regulatory requirements of 10 C.F.R. § 50.61 to §50.61a (AlternateFracture Toughness Requirements). This new amendment request introduces further non-conservative analytical assumptions into the troubled forty-three (43) year operationalhistory of Palisades. Entergy's License Amendment Request (LAR) contains anequivalent margins evaluation, which is an untried methodological approach.Petitioners' hearing request was referred to an Atomic Safety and Licensing Board forconsideration. Both Entergy and the NRC Staff filed answers opposing the Amended Petition, towhich Petitioners filed a reply. On March 25, 2015, the Board heard oral argument on standingand contention admissibility, and on May 8, 2015, the ASLB issued its "Memorandum and Order(Ruling on Petition to Intervene and Request for a Hearing"), LBP-15-17 wherein the ASLBdenied Petitioners' Amended Petition to Intervene and for a Public Adjudication Hearing. II. Factual and Procedural BackgroundA. The 1985 PTS Rule And Embrittlement Screening Program (10 C.F.R. § 50.61)In 1985, the NRC implemented a mandatory program to monitor PWR RPVs forembrittlement over time, coupled with screening limits to prevent over-embrittled reactors fromoperating. The program to monitor PWR RPVs is described in 10 C.F.R. Part 50, Appendix H,4(December 8, 2014) (hereinafter "Amended Petition").See "Analysis of Potential Pressurized Thermal Shock Events, Final Rule," 50 Fed. Reg.429,937 (July 23, 1985) (creating the screening criteria); "Fracture Toughness and SurveillanceProgram Requirements, Final Rule," 38 Fed. Reg. 19,012 (July 17, 1973) (creating the programto monitor PWR RPVs).

and is titled "Reactor Vessel Material Surveillance Program Requirements" (SurveillanceProgram). The purpose of the Surveillance Program "is to monitor changes in the fracturetoughness properties of ferritic materials [iron-based metals, such as steel] . . . which result fromexposure of these materials to neutron irradiation and the thermal environment." The5Surveillance Program relies on physical material samples, also known as specimens, capsules,or coupons, "which are withdrawn periodically from the reactor vessel." The NRC must pre-6approve the schedule for removing material samples from the reactor vessel.7The actual screening limits required by Appendix H's Surveillance Program formonitoring reactor pressure vessels ("RPVs") for fracture toughness are established in 10 C.F.R.§ 50.61, entitled "Fracture toughness requirements for protection against pressurized thermalshock events." Section 50.61 relies on data gathered from the Surveillance Program to calculatethe RPV wall's fracture toughness, and compares it with a safety limit that cannot be exceeded.8NRC regulations represent steel fracture toughness as a temperature value, known as"reference temperature." The NRC Staff says, "[r]eference temperature is the metric that theNRC uses to quantitatively assess brittleness, so these terms may be regarded as synonymous.Steel having a high 'reference temperature' also has a higher degree of brittleness than steel with10 C.F.R. Part 50, App. H(I).5Id. The NRC's regulations further require that the physical specimens "be located near6the inside vessel wall in the beltline region so that the specimen irradiation history duplicates, tothe extent practicable within the physical constraints of the system, the neutron spectrum,temperature history, and maximum neutron fluence experienced by the reactor vessel innersurface." Id. Part 50, App. H(III)(B)(2).Id. Part 50, App. H(III)(B)(3).7See id. § 50.61(c)(2)(i).8 a low reference temperature." The ability of steel to resist fracture changes as a function of9temperature; when steel is at high temperatures, it can retain its ductility and related ability toresist fracturing from PTS events, even after extended periods of neutron irradiation. But at lowtemperatures, steel is naturally brittle, and even unirradiated steel can potentially suffer brittlefailure. The point at which steel transitions from the high-temperature, fracture-resistant-state,10to the low-temperature, brittle state, is called the "RTNDT," or "Transition fracture toughnessreference temperature," or more simply "reference temperature." As described by Staff11guidance documents, this transition point depends primarily on two factors material compositionand cumulative irradiation by high-energy neutrons. As steel is exposed to more high-energy12neutrons (i.e., its fluence increases), RTNDT increases concurrently. Thus, as fluence increases,1314John B. Giessner, Division of Reactor Projects, Summary of the March 19, 2013, Public9Meeting Webinar Regarding Palisades Nuclear Plant, encl. 2 at 4 (Apr. 18, 2013) (ADAMSAccession No. ML13108A336) (hereinafter "Palisades Webinar").See Alternate PTS Rule Technical Basis Report at xxxviii-xxxix (noting that with steel10at high temperatures "cleavage cannot occur"). A "Cleavage fracture" is the type of fractureassociated with fracture of brittle materials. See id. at xxxviii.Id. at xxxiv. "NDT" stands for Nil-Ductility Transition. Id. at xxxi.11Id. at xx ("[T]ransition temperatures increase as a result of irradiation damage12throughout the operational life of the vessel."); id. § 2.1.3 (discussing the factors affectingfracture toughness); id. § 2.4.2 (limiting the fluence to only high-energy "fast" neutrons, whichhave energies above one mega electron volt).Fluence is the integral of the neutron flux over time. The neutron flux is the total13distance traversed by neutrons within a unit volume of material within one unit of time. Typicallythe unit volume is one cubic centimeter and the unit time is one second. Thus the unit of neutronflux is neutron-centimeter/centimeter(cubed)-second, typically expressed as neutrons/centimeter(squared)-second. See Samuel Glasstone and Alexander Sesonske, Nuclear Reactor Engineering§ 2.118 (Van Nostrand Reinhold Co. 1967).See Alternate PTS Rule Technical Basis Report § 2.4.1 (discussing the reference14temperature approach to characterizing fracture toughness in ferritic materials).

the steel stays brittle at higher and higher temperatures, and it is therefore more likely to fractureas a result of PTS events.The NRC established screening limits in 10 C.F.R. § 50.61, which are the currentscreening criteria, to reduce the risk that a PTS event will result in an RPV fracture. Thescreening limits are expressed as temperature values. When the reference temperature of an RPVis above this screening limit, the RPV is considered to have an unreasonably high risk of fracturefrom a PTS event. The PTS "screening criterion" is 270°F for plates, forgings, and axial weld15materials, and 300°F for circumferential weld materials."16If the RTNDT values projected at specific areas of the RPV for the end of life of the plant,known as RTPTS, surpass the Current Screening Criteria, the licensee must submit a safety17analysis and obtain the approval of the Office of Nuclear Reactor Regulation to continue tooperate. If that office does not approve continued operation based on the licensee's safety18analysis, the licensee must request an opportunity to modify the RPV or related reactor systemsSee 10 C.F.R. § 50.61(b)(2). The current screening criteria "correspond to a limit of 5 x1510-6 events/year on the annual probability of developing a through-wall crack" in the RPV.Alternate PTS Rule Technical Basis Report at xx.10 C.F.R. § 50.61(b)(2); see also 75 Fed. Reg. at 13 ("The current PTS rule . . .16establishes screening criteria below which the potential for a reactor vessel to fail due to a PTSevent is deemed to be acceptably low").10 C.F.R. § 50.61(a)(7) ("RTPTS means the reference temperature, RTNDT, evaluated for17the [end of life] Fluence for each of the vessel beltline materials."); Alternate PTS RuleTechnical Basis Report § 11.2 ("10 CFR 50.61 defines RTPTS as the maximum RTNDT of anyregion in the vessel (a region is an axial weld, a circumferential weld, a plate, or a forging)evaluated at the peak fluence occurring in that region").10 C.F.R. § 50.61(b)(3)-(5).18 to "reduce the potential for failure of the reactor vessel due to PTS events."19B. The Alternate PTS Rule And Embrittlement ScreeningProgram (10 C.F.R. § 50.61a)While no reactor is expected to exceed the current screening criteria established in 10C.F.R. § 50.61 during its 40 year operating license, the Staff has noted that Palisades in particularis one of the first plants likely to exceed them, as Palisades' RPV is "constructed from some ofthe most irradiation-sensitive materials in commercial reactor service today." This concern, as20well as significant advancements in failure analysis and materials knowledge, prompted the NRCto reexamine the § 50.61 approach for projecting fracture toughness and the screening criteria.21In August 2007, the NRC issued NUREG-1806, "Technical Basis for Revision of the [PTS]Screening Limit in the PTS Rule (10 CFR 50.61)." That report summarized the results of a fiveyear study by the NRC, the purpose of which "was, to develop the technical basis for revision ofthe Pressurized Thermal Shock (PTS) Rule." The report concluded that through-wall cracks22were much harder to create in RPVs than initially thought, and occurred in fewer circum-stances. The report thus recommended a more detailed approach to setting screening criteria23that would take into account the varying conditions along different parts of the Id. § 50.61(b)(6).19Alternate PTS Rule Technical Basis Report at xxii.20See "Alternate Fracture Toughness Requirements for Protection Against Pressurized21Thermal Shock Events, Proposed Rule," 72 Fed. Reg. 56,275, 56,276 (Oct. 3, 2007); AlternatePTS Rule Technical Basis Report at iii, xx-xxiii.Alternate PTS Rule Technical Basis Report at xix.22See id. at xx-xxiii.23 RPV. The report also recommended removing the "margin term" that had been included in the24current screening criteria to account for unknown factors, because essentially all factors are nowknown and are effectively quantified.25On October 3, 2007, the Staff published a notice of proposed rulemaking. The26rulemaking notice stated that the Alternate PTS Rule Technical Basis Report "conclude[d] thatthe risk of through-wall cracking due to a PTS event is much lower than previously estimated,"and that "[t]his finding indicates that the screening criteria in 10 CFR 50.61 are unnecessarilyconservative." 27On January 4, 2010, the NRC issued the final rule, creating 10 C.F.R. § 50.61a. TheAlternate PTS Rule makes two important changes. Section 50.61a replaces the relatively broad28current screening criteria (270°F for plates, forgings, and axial weld materials, and 300°F forcircumferential weld materials) with more detailed Alternate Screening Criteria. The Alternate29Screening Criteria consist of eighteen different reference temperature limits that depend on RPVId. at xxv ("Specifically, we recommend a reference temperature for flaws occurring24along axial weld fusion lines (RTAW or RTAW-MAX), another for flaws occurring in plates or inforgings (RTPL or TRPL-MAX), and a third for flaws occurring along circumferential weld fusionlines (RTCW or RTCW-MAX)").Id. at xxvii.2572 Fed. Reg. 56,275.26Id. at 56,276.27However, like the old rule, the new rule provides measures for ongoing reporting, 1028C.F.R.§ 50.61a(d)(1), and mitigation processes for licensees if they project they will exceed (orthey do exceed) the Alternate PTS Rule's screening criteria. Id. § 50.61a(d)(2)-(7).75 Fed. Reg. at 18.29 wall thickness and the part of the RPV under consideration. The Alternate PTS Rule also30changes how licensees derive projected reference temperatures for the components of theirRPVs. Section 50.61a relies on a probabilistic "embrittlement model" to predict future31reference temperatures across the RPV, which is then verified by existing surveillance data in aprocess called the "consistency check." Section 50.61, by contrast, continuously integrates32surveillance data into future embrittlement projections. In the final rulemaking notice, the33Commission concluded that the new "estimation procedures provide a better (compared to theexisting regulation) method for estimating the fracture toughness of reactor vessel materials overthe lifetime of the plant." The final rulemaking notice stated that the Alternate PTS Rule34"provides reasonable assurance that licensees operating below the screening criteria could endurea PTS event without fracture of vessel materials, thus assuring integrity of the reactor pressurevessel." Furthermore, the final rulemaking stated that "[t]he final rule will not significantly3510 C.F.R. § 50.61a(g) tbl. 1.30See Id. § 50.61a(f), (f)(6)(B)(ii).31Id. 32Compare id. § 50.61a(f)(6)(i) (requiring that a licensee perform a "consistency check"33of its embrittlement model against available surveillance data), and Alternate PTS RuleTechnical Basis Report § 3.1.1 (The Alternate PTS Rule is designed to "enable all commercialPWR licensees to assess the state of their RPVs relative to such a new criterion without the needto make new material property measurements," instead using "only information that is currentlyavailable."), with 10 C.F.R. § 50.61(c)(2)(i) (requiring that "plant-specific surveillance data mustbe integrated into the RTNDT estimate"), and Alternate PTS Rule Technical Basis Report § 2.4.2(Under the Current PTS Rule, material samples "from RPV surveillance programs provide theempirical basis to establish embrittlement trend curves . . . .").75 Fed. Reg. at 18.34Id. at 22.35 increase the probability or consequences of accidents, result in changes being made in the typesof any effluents that may be released off site, or result in a significant increase in occupational orpublic radiation exposure."36C. Invocation Of The Alternate PTS RuleTo take advantage of the Alternate PTS Rule, a licensee must request approval from theNRC Office of Nuclear Reactor Regulation, in accordance with the procedures for submitting alicense amendment under 10 C.F.R. § 50.90. The application must contain: (i) under Section50.61a(f), the projected embrittlement reference temperatures along various portions of theRPV, from now to a future point, compared to the Alternate Screening Criteria; and (ii) underSection 50.61a(e), an assessment of flaws in the RPV. In calculating embrittlement reference37temperatures under Section 50.61a(f), a licensee must calculate neutron flux through the RPV"using a methodology that has been benchmarked to experimental measurements and withquantified uncertainties and possible biases." From that point, the licensee must establish 38RTNDT(U) for various key points along the RPV. Then a licensee uses a series of equations and39charts provided in the rule to create an embrittlement model. That model projects the referencetemperatures for various parts of the RPV at the end of life of the plant, known in the new rule asId.3610 C.F.R. § 50.61a(c)(1)-(2). Under Section 50.61a, the licensee must separately37examine for flaws in the reactor vessel. Id. § 50.61a(c)(2). The analysis of flaws in the PalisadesRPV is not in dispute in this proceeding.Id. § 50.61a(f).38Id. § 50.61a(f)(4). RTNDT(U) is the nil-ductility reference temperature for the RPV39material in the annealed state, before the reactor was operational. Id. If measured values are notavailable, a licensee can use a set of generic mean values. Id. § 50.61a(f)(4)(i), (ii).

RTMAX-X. The embrittlement model allows for calculations of RTMAX-X across the RPV using40probabilistic analyses, without having to rely on measured data. The RTMAX-X values are41compared to the Alternate Screening Criteria to determine whether the RPV is safe to operate.42Importantly, as calculations of RTMAX-X are made analytically, without directly incorporatingsurveillance data, licensees have to verify that their calculations at the time of the applicationmatch up with surveillance data. To do so, licensees have to perform the "consistency check"43of their calculations for specific materials against "heat-specific surveillance data that arecollected as part of 10 CFR Part 50, App. H, surveillance programs." The purpose of the check44is to "determine if the surveillance data show a significantly different trend than theembrittlement model predicts." The check includes three statistical analyses that compare the45model's inputs, fluence and material properties, with the model's output, reference temperature.46Id. § 50.61a(f)(1)-(3). "RTMAX-X is the equivalent term for RTPTS in 10 CFR 50.61a."40"Proposed Rulemaking - Alternate Fracture Toughness Requirements for Protection AgainstPressurized Thermal Shock Events" (RIN 3150-AI01), SECY-07-0104 (June 25, 2007)See supra note 34.41See 10 C.F.R. § 50.61a(c)(3).42Id. § 50.61a(f)(6)(i).4375 Fed. Reg. at 16. The regulatory history of the Alternate PTS Rule and associated44draft guidance indicates that uncertainty in surveillance data measurements may be a concern,which licensees' applications should address. See id. at 16-17 (discussing potential concernswith variability in surveillance data); "Regulatory Guidance on the Alternate Pressured ThermalShock Rule," Draft Regulatory Guide DG-1299 at 12 (Mar. 2015) (hereinafter "DG-1299") ("Theinput variables to [the equations comprising the consistency check] are subject to variability andare often based on limited data," particularly fluence).10 C.F.R. § 50.61a(f)(6)(i)(B).4575 Fed. Reg. at 16 ("The NRC is modifying the final rule to include three statistical tests46to determine the significance of the differences between heat-specific surveillance data and the The consistency check is required "[i]f three or more surveillance data points measured at threeor more different neutron fluences exist for a specific material." 47In the event the embrittlement model deviates from the physical samples over the limitsspecified in the regulation, the licensee must submit additional evaluations and seek approvalfor the deviations from the Director of the Office of Nuclear Reactor Regulation. 48D. Petitioners' Objections To Entergy License AmendmentRequest (LAR) Invoking Alternate PTS RuleOn September 30, 2014, notice was published in the Federal Register of Entergy's49intentions of seeking amendment of the operating license of Palisades Nuclear Plant to allowimplementation of an alternative method of calculation of the degree of embrittlement of thePalisades nuclear reactor pressure vessel. The 10 C.F.R. § 50.61 screening criteria, to whichPalisades supposedly adhered, define a limiting level of embrittlement beyond which plantoperation cannot continue without further evaluation. The switch to the use of 10 CFR § 50.61awill change how fracture toughness of the reactor vessel is determined, moving from ananalytical to a probabilistic risk assessment method. Entergy's proposed "no significant hazards"determination, required by 10 C.F.R. § 50.91(a), concluded that the proposed change will notinvolve a significant increase in the probability or consequences of an accident previouslyembrittlement trend curve"). The consistency check compares the mean and slope of theembrittlement model curve against surveillance data, as well as checks to confirm that outliersfall within acceptable residual values provided in the regulation. See 10 C.F.R. §50.61a(f)(6)(ii)-(v).10 C.F.R. § 50.61a(f)(6)(i)(B).47Id. § 50.61a(f)(6)(vi).4879 Fed. Reg. 58812 (September 30, 2014)49 evaluated. Entergy further concluded that the proposed change does not create the possibility of50a new or different type of accident from any accident previously evaluated. The utility51maintained, also, that the proposed change would not involve a significant reduction in a marginof safety. In light of Entergy's analysis, the NRC Staff concluded that "the three standards of5210 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that theamendment request involves no significant hazards consideration."53When the Palisades RPV was brand new, its reference temperature-nil ductility transition(RT-ndt) was at 40 degrees F. By the early 1980s, NRC had weakened Palisades' screeningcriteria - and the rest of the U.S. pressurized water reactors' - to 200 degrees F, which is closer tothe operating temperature of Palisades, which is around 550 degrees F. Thus if the EmergencyCore Cooling System ("ECCS") pumps too-cold water into the 550 degrees F reactor pressurevessel and cools it too quickly down to 200 degrees F (or, later, 270 or 300 degrees), thereinstantaneously arises a serious potential for a fracture of the RPV, which would be a verysignificant reactor accident. When the PWR safety system repressurizes the RPV, the metal can'ttake it any more, and fractures. It breaks, either by major cracking or actual fragmentation,presumably at the point of a flaw in the RPV.As noted, 200 degrees F was merely an early retreat from regulation. The criteria werelater relaxed to 270 degrees F for axial/vertical welds, and to 300 degrees F for welds of aId. at 58815. 50Id.51Id.52Id.53 circumferential/horizontal orientation. And through it all, Palisades and/or the NRC haveprojected, again and again that the new PTS screening criteria would be exceeded by a predictedfuture date. These dates have been 1995; 1999; September 2001; 2004; 2007; 2014; April 2017;and August 2017. On or near those dates, Palisades or the NRC has said, the allowable boundarybeyond which lies the risk of disaster will be crossed. Each time, though, the date of heightenedvulnerability to this type of disaster has routinely slipped back further into the future. In the many years since the early indicators of embrittlement in its first operationaldecade, Palisades has gained notoriety as one of the nation's most-embrittled reactors. In its May19, 1995 NRC Generic Letter 1992-001, Supplement 1, the NRC Staff permitted Palisades to54operate until late 1999, observing that it had "reviewed the other PWR vessels and, based uponcurrently available information, believes that the Palisades vessel will reach the PTS screeningcriteria by late 1999, before any other PWR." (Emphasis added). Id.Petitioners' objections to the ASLB relied in large part on the expert opinion of nuclearengineer Arnold Gundersen (see "Declaration of Arnold Gundersen," hereinafter "GundersenDeclaration") that the analysis provided to the NRC by Entergy is inadequate and relies uponunsupported assumptions which warrant a hearing as to whether Entergy should be allowed toswitch over to 10 C.F.R. § 50.61a. Petitioners urged the possibility exists that significant hazardsassociated with implementation of the alternative calculation method under 10 C.F.R. § 50.61amay occur, caused by materially-underestimated prospects of a severe loss-of-coolant accident(LOCA) involving the reactor. ADAMS No. ML031070449.54 Arnold Gundersen stated that "Almost half of the initial capsules [coupon samples]installed 43 years ago still remain inside the embrittled nuclear reactor" and that if the NRCallows Entergy to postpone the next Palisades coupon sampling until 2019, "then no accuratecurrent assessment of Palisades' severe embrittlement condition exists." Gundersen Declarationp. 8, ¶ 21. Gundersen opined that § 50.61 is analytical in nature, while § 50.61a authorizesprobabilistic risk assessment, and that the discretionary availability of § 50.61a under thecircumstances cannot be used as a substitute for scientific investigation. Id. at p. 9, ¶ 24.3.Gundersen observed (id. at p. 3, ¶ 8) that "Continued operation of the Palisades nuclear powerplant without analyzing the coupon designated to be sampled more than seven years ago meansthat Entergy may be operating Palisades as a test according to 10 C.F.R. § 50.59." (Emphasis inoriginal).Petitioners' expert further alleged that the underlying data from other supposedlycomparative nuclear plants assessing ductility of their RPVs is not legitimate: "The NRC hasallowed Palisades to compare itself to reactors of disparate designs from other vendors, built indifferent years and operating at diverse power levels." Gundersen Declaration at ¶ 24.2. Theseplants, which he says "thus far have not exhibited significant signs of reactor metal embrittle-ment," are poor comparables because:. . . the dramatically different nuclear core design and operational powercharacteristics make an accurate comparison impossible. The difference between theWestinghouse nuclear cores and the Combustion Engineering nuclear core impacts theneutron flux on each reactor vessel, thus making an accurate comparison of neutronbombardment and embrittlement impossible.Id. at p. 10, ¶ 27. The core objection raised by Petitioners' filing is that the 10 C.F.R. § 50.61a alternative to § 50.61 allows Entergy to substitute various estimates of the status of the RPV for actual datainvestigation and analysis. Those § 50.61a projections are attained, among other means, byaveraging data on reactor vessels from other nuclear power plants, to arrive at a projection of thecurrent status of the Palisades RPV. Entergy's recourse to the alternate approach, accompanied asit is by deliberate non-testing of metal coupons from the RPV for 16 years (2003-2019) can beunderstood only if one assumes that Entergy does not want to know what physical testing mightattain by way of useful data about the true state of affairs within the Palisades RPV. As Petitioners' expert, Arnold Gundersen objected to the specific comparable nuclear reactorvessels cited by Entergy to comply with § 50.61a, pointing out that "The NRC has allowed Palisades tocompare itself to reactors of disparate designs from other vendors, built in different years and operatingat diverse power levels." Gundersen Declaration at ¶ 24.2. These plants, which he said "thus far havenot exhibited significant signs of reactor metal embrittlement," are poor comparables because:. . . the dramatically different nuclear core design and operational power characteristicsmake an accurate comparison impossible. The difference between the Westinghouse nuclearcores and the Combustion Engineering nuclear core impacts the neutron flux on each reactorvessel, thus making an accurate comparison of neutron bombardment and embrittlementimpossible.Id. at p. 10, ¶ 27. A good example of a false comparison is found in Structural Integrity Associates, Inc.'s ReportNo. 0901132.401, Revision 0, "Evaluation of Surveillance Data for Weld Heat No. W5214 forApplication to Palisades PTS Analysis," ADAMS No. ML110060693. This document was part of thetechnical basis for the PTS safety risk regulatory rollback of PTS screening criteria, from January 2014to April 2017 at Limiting Beltline Weld W5214. "Similar Sister Plant" proxies were used whichinvolved the inappropriate averaging of 11 sample surveillance capsules/coupons from very dissimilarRPVs. Ssuch false comparisons, Gundersen says, "significantly dilute Palisades' embrittlement calculations." Id. at p. 11, ¶ 28. He adds: "This rogue comparative data is not sound scientificmethodology and clearly places the operations of the Palisades NPP in the experimental test venue,possibly as delineated in 10 CFR 50.59." Id. at p. 11, ¶ 29. The most serious analytical problem in using sister plants data "is the extraordinary difficultycomparing data from four separate plants while still maintaining one standard deviation (1ó) or 20%between all the data. According to the Palisades Reactor Pressure Vessel Fluence Evaluation, onestandard deviation is required, however there has never been a discussion of how this was achievedbetween the four sister units." Gundersen Declaration at p. 11, ¶ 30. While "[a] 1ó analysis appears tobe binding within the Palisades data, . . . the NRC lowers the bar when comparing data from similar sisterplants that are included in Entergy's analysis of the Palisades reactor vessel without requiring the same1ó variance with Palisades." Id. at p. 12, ¶ 32. Gundersen added: "There can be no assurance that the20% error band at Palisades encompasses the 20% error band at the Robinson or Indian Point plants. Tocompare this different data without assurance that the 1ó variance from each plant overlaps the otherplants lacks scientific validity." Id. at p. 12, ¶ 33. Gundersen further found that there is "extraordinary variability between the neutron flux acrossthe nuclear core in this Combustion Engineering reactor" because of a "flux variation of as much as300% between the 45-degree segment and the 75-degree segment," calling it "mathematicallyimplausible that a 20% deviation is possible when the neutron flux itself varies by 300%." Id. at p. 12, ¶34. In sum, he noted that:The Westinghouse Analysis delineates that a 20% variation is mandatory, yet theeffective fluence variability can be as high as 300%, therefore, the analytical data does notsupport relicensure without destructive testing and complete embrittlement analysis of additionalcapsule samples.Id. at p. 16, ¶ 39.

III. ArgumentA. The ASLB Erroneously Found The Decision Allowing EntergyTo Invoke 10 C.F.R. § 50.61a To Be NondiscretionaryThe Atomic Safety and Licensing Board generally denied the Petition, holding that:Petitioners apparently want the Board to preclude Entergy from relying on Section50.61a to avoid meeting the requirements of Section 50.61, but it is just such a "devia-tion" that Section 50.61a authorizes. The evident purpose of the Alternate PTS Rule's"Alternate Fracture Toughness Requirements" is to provide an alternative to satisfyingthe more demanding requirements of Section 50.61. Therefore, Petitioners are insubstance asking that the Board prohibit what Section 50.61a allows. Under 10 C.F.R. §2.335, we may not consider such a contention except under specific conditions notpresent here.(Emphasis supplied). LBP-15-17 at 29.The Licensing Board's reasoning is flawed; it involves two distinct considerations. Evenassuming arguendo that the NRC can promulgate an alternative regulation that is weaker than theother, and afford a choice of laws to nuclear utility operators, that position says nothing about thediscretionary nature of the NRC Director of Nuclear Reactor Regulation over whether to allow aparticular applicant to invoke 10 C.F.R. § 50.61a. The ASLB ruled, in essence, that if thepaperwork is properly completed, the substantive issue - whether to allow Entergy to move to 10C.F.R. § 50.61a - is essentially irrelevant, is to be automatically allowed, and that the NRCStaff's regulatory hand must be stayed. This dogmatic stance is apparent in several ASLBstatements. For example, the ASLB adopted Entergy's argument that "a contention asserting thatdifferent analysis or technique should be utilized is inadmissible because it indirectly attacks theCommission's regulations." LBP-15-17 at 33. Petitioners were advocating, not for usage of adifferent technique to be used, but that that the Director of NRR should have discretionarilyconsidered whether a superior "reasonable assurance" of protection of public health and safety would be derived from rejecting Entergy's request to invoke § 50.61a. This is because 10 C.F.R. § 50.61a clearly contemplates a discretionary determination bythe Director of NRR. See, for example, § 50.61a( c)(1) (RTMAX-X values assessment "mustspecify the bases for the projected value of RTMAX-X for each reactor vessel beltline material,including the assumptions regarding future plant operation"); § 50.61a( c)(2) ("Each licenseeshall perform an examination and an assessment of flaws in the reactor vessel beltline as requiredby paragraph (e) of this section" - and (e) requires disclosure of tests performed but, again,detailed explanation of the methodology underlying NDE uncertainties assumptions, and55adjustments must be disclosed. This is merely a recognition that even objective data, onceinterpreted, may be examined to ascertain the objectivity or inappropriate bias which may haveoccurred in the means of analysis which have been applied to it. Where there is discretion vestedin the regulator, differences of opinion, interpretation, and expert analysis are legitimate bases forchallenging the decision because the decision is potentially arrived at in an adversarial manner. This principle is also obvious in § 50.61a(f)(7), which requires that "The licensee shallreport any information that significantly influences the RTMAX-X value to the Director inaccordance with the requirements of paragraphs (c)(1) and (d)(1) of this section." Therequirement clearly introduces subjective judgment and selection among different conditions orfindings into the decision of what data is to be provided to the Director of NRR. § 50.61a says in part: "The methodology to account for NDE-related uncertainties must be55based on statistical data from the qualification tests and any other tests that measure the differencebetween the actual flaw size and the NDE [no-destructive examination] detected flaw size. Licenseeswho adjust their test data to account for NDE-related uncertainties to verify conformance with the valuesin Tables 2 and 3 shall prepare and submit the methodology used to estimate the NDE uncertainty, thestatistical data used to adjust the test data and an explanation of how the data was analyzed for reviewand approval by the Director in accordance with paragraphs (c)(2) and (d)(2) of this section."

Hence for Petitioners to provide their expert's critique of the means by which the §50.61a investigation was conducted, and the weaknesses or biases in the underlying data,assumptions and manipulations of information cannot be construed as a frontal assault on theregulatory citadel, but must instead be seen, for purposes of the admissibility determination, as anexposition of the flaws caused by straying away from knowable science. Petitioners' critique wasnot answered by any experts on behalf of the NRC Staff or Entergy. Petitioners articulatedchallenges to the proposed exercise of discretion by the Director of Nuclear Reactor Regulationand should be accorded a hearing to provide more evidence.The Commission should take note that the agency regulations contain a "pressurizedthermal shock regulatory relief valve" for situations where a nuclear utility cannot meet even theflaccid threshold of 10 C.F.R. § 50.61a, by means of which the Director of NRR may allow anembrittled reactor to operate beyond the PTS screening criteria. See slide show, "Technical Briefon Regulatory Guidance on the Alternative PTS Rule (10 C.F.R. § 50.61a)," Official Transcriptof Proceedings, ADAMS No. ML14321A542, at p. 242/268 of .pdf:Use of 10 CFR 50.61a PTS screening criteria requires submittal for review andapproval by Director, NRR.For plants that do not satisfy PTS Screening Criteria, plant-specific PTSassessment is required.Must be submitted for review and approval by Director, NRR.Guidance is not provided for this case.Subsequent requirements (i.e., after submittal) are defined in paragraph (d) of 10CFR 50.61a. (Emphasis supplied).B. 'Reasonable Assurance' Cannot Apply Alike To Two Regulations AddressingThe Same Subject When One Is Deemed To Be Weaker Than The OtherWhen the ASLB referred to the 10 C.F.R. § 50.61 requirements as "more demanding"than the "Alternate Fracture Toughness Requirements," the Board agreed that the "evident purpose" of 10 C.F.R. § 50.61a is to weaken the regulatory rigor over nuclear utilities withserious RPV ductility problems. Petitioners suggest that substitution of a stronger standard whichofficially provides "reasonable assurance" of public protection with an admittedly weaker onealso "reasonably assured" to be protective, is legally anomalous. 56Section 182a of the Atomic Energy Act states that a reactor operating license mustinclude "technical specifications" that include, inter alia, "the specific characteristics of thefacility, and such other information as the Commission may, by rule or regulation, deemnecessary in order to enable it to find that the utilization . . . of special nuclear material . . . willprovide adequate protection to the health and safety of the public." 42 U.S.C. § 2232(a). Thegeneral requirement for operating licenses, 10 C.F.R. § 50.57(a)(3), requires a finding ofreasonable assurance of operation without endangering the health and safety of the public. Duke57Power Co. (Catawba Nuclear Station, Units 1 & 2), LBP-82-116, 16 NRC 1937, 1946 (1982). Inthis proceeding, Entergy must demonstrate that it satisfies the "reasonable assurance standard" bya preponderance of the evidence. Reasonable assurance "is not susceptible to formalisticquantification or mechanistic application. Rather, whether the reasonable assurance standard ismet is based upon sound technical judgment applied on a case-by-case basis." AmerGen EnergyCo., LLC (Oyster Creek Nuclear Generating Station), LBP-07-17, 66 NRC 327, 340 (2007),The "reasonable assurance" finding of 10 C.F.R. § 50.61a is found at 75 Fed. Reg. at 22.56"(a) Pursuant to § 50.56, an operating license may be issued by the Commission, up to57the full term authorized by § 50.51, upon finding that:(1) ***; (2) ***;(3) There is reasonable assurance (i) that the activities authorized by the operating licensecan be conducted without endangering the health and safety of the public. . .".

aff'd, CLI-09-07, 69 NRC 235, 263 (2009) (rejecting an argument that reasonable assuranceshould be quantified with 95% confidence). To consider a stronger regulation and a weaker oneto be on the same footing when it comes to providing reasonable assurance is logicallyinconsistent, as illustrated by this very case. Palisades contains the worst-embrittled reactorpressure vessel in the United States. Posed a choice between a tougher, physical testing-basedregulatory regime, or a weaker, projective method of assessing RPV ductility, owners of theworst-embrittled reactor have chosen the less-protective regulations. Because they are lessprotective, and given the enormous discretion vested in the Director of Nuclear ReactorRegulation to decide on a case-by-case basis what terms and conditions should be imposed under10 C.F.R. § 50.61a, a hearing is necessary to resolve factual issues in line with regulatoryexpectations. The ASLB's candor shows that the alternative regulation exists merely to provideEntergy with "reasonable assurance" of being able to operate Palisades in disregard of thedestructive testing obligations of 10 C.F.R. § 50.61 and in derogation of the binding requirementof reasonable assurance that the public's health and safety will be the priority for protection. C. Variabilities In Sister Plant Data Erroneously Allowed Inappropriate ComparisonsThe ASLB treated Petitioners' objections to the invalidity of sister plant data as attemptsto suggest regulatory parameters which exceed the requirements of 10 C.F.R. § 50.61a. But Petitioners have previously argued that the considerable discretion accorded the Director of NRRto allow invocation of § 50.61a should be construed as lending relevance to their apples/orangesquibbling. Further, 10 C.F.R. § 50.61a(f)(6)(i) requires that "(A) The surveillance material mustMAX-X be a heat-specific match for one or more of the materials for which RT is being calculated."Petitioners' expert Gundersen attested to the lack of proof that the metals from the various RPVs match. This conclusion was not rebutted by any expert evidence from either the NRC Staff norEntergy. The Licensing Board's implicit finding that the metals compared in the sister plantsworkup were "of the appropriate chemical composition" (LBP-15-17 at 41) was seriouslychallenged by Petitioners' expert witness. Nor did Entergy or the NRC Staff refute Gundersen'sobservation that (noted at p. 17 infra) that there is "extraordinary variability between the neutronflux across the nuclear core in this Combustion Engineering reactor" because of a "flux variationof as much as 300% between the 45-degree segment and the 75-degree segment," and concludingit was "mathematically implausible that a 20% deviation is possible when the neutron flux itselfvaries by 300%." Gundersen Declaration p. 12, ¶ 34. Perhaps § 50.61a is the culmination ofdecades of learning about embrittlement, but it still cannot dispense with huge variations inneutron flux in Palisades, alone. The ASLB improperly rejected this portion of Petitioners'contention.IV. ConclusionThe threshold admissibility requirements of NRC's contention rule should not be turnedinto a "fortress to deny intervention." Power Authority of the State of New York, et al. (JamesFitzPatrick Nuclear Power Plant; Indian Point Nuclear Generating Unit 3), CLI-00-22, 52 NRC266, 295 (2000). There is no requirement that the petitioners' substantive case be made at thecontention stage. Matter of Entergy Nuclear Generation Co., et al. (Pilgrim Nuclear PowerStation), 50-293-LR (ASLB Oct. 16, 2006), 2006 WL 4801142 at (NRC) 85. The Commissionhas explained that the requirement at § 2.309(f)(1)(v) "does not call upon the intervenor to makeits case at [the contention] stage of the proceeding, but rather to indicate what facts or expertopinions, be it one fact or opinion or many, of which it is aware at that point in time which provide the basis for its contention." Pilgrim at 84. The admissibility requirement "generally isfulfilled when the sponsor of an otherwise acceptable contention provides a brief recitation of thefactors underlying the contention or references to documents and texts that provide suchreasons." Id.WHEREFORE, the adverse determinations of the Atomic Safety and Licensing Board inLBP-15-17 should be reversed and the matter remanded to the ALSB for an evidentiary hearing.Respectfully submitted, /s/ Terry J. Lodge Terry J. Lodge (OH #0029271)316 N. Michigan St., Ste. 520Toledo, OH 43604-5627(419) 255-7552Fax (419) 255-7552Tjlodge50@yahoo.comCounsel for Petitioners UNITED STATES OF AMERICANUCLEAR REGULATORY COMMISSIONBefore the CommissionIn the Matter ofEntergy Nuclear Operations, Inc.(Palisades Nuclear Plant)Operating License Amendment Request)Docket No. 50-255) June 2, 2015)) *****CERTIFICATE OF SERVICEI hereby certify that copies of the foregoing "PETITIONERS' 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 LICENSEAMENDMENT REQUEST FOR AUTHORIZATION TO IMPLEMENT 10 C.F.R. § 50.61a'"and the accompanying "BRIEF IN SUPPORT" were served by me upon the parties to thisproceeding via the NRC's Electronic Information Exchange system this 2nd day of June, 2015. /s/ Terry J. Lodge Terry J. Lodge (OH #0029271)316 N. Michigan St., Ste. 520Toledo, OH 43604-5627(419) 255-7552Fax (419) 255-7552Tjlodge50@yahoo.comCounsel for Petitioners-25-