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 STAT ES OF AMERICANUCLEAR REG ULATORY COMMISSION Before the Com missionIn the Matter of

Entergy Nuclear Operations, I nc.(Palisades Nuc lear Plant)Operating License Amendment Reque st) Docket No. 50-255

)June 2, 2015

)) *****INTERVENORS' 10 C.F

.R. § 2.311( c

) NOTICE OF A PPEAL OF ATOMIC SAF ETY AND LICENSING BOARD'S DENIAL OF PETITION TO INTE RVENE AND REQUEST FOR A HEARING ON ENTERGY LICENSE AMENDMENT REQUEST FOR AUTHORIZATION TO IMPLEMENT 10 CFR

§ 50.61a AND BRIEF IN SUPPORTTerry J. Lodge (OH #0029271) 316 N. Michig an St., Ste. 520 Toledo, OH 43604-5627(419) 255-7552Fax (419) 255-7552tjlodge50@

yahoo.comCounsel for Petitioners TABLE OF CONTENTS Table of Authorities iiI. Introduction 1II. Factual and Proc edural Background 3A. The 1985 PTS Rule And Embrittlement Scre ening Program (

10 C.F.R.

§ 50.61) 3B. The Alternate PTS Rule And Embrittlement Sc reening Program (

10 C.F.R.

§ 50.61a) 7C. Invocation Of The Alternate PTS Rule 10D. Petitioners' Objec tions To Entergy License AmendmentRequest(LAR) Invoking Alternate PTS Rule 12III. Argument 18A. The ASLB Erroneously Found The De cision Allowi ng Entergy To Inv oke10 C.F.R. § 50.61a To Be Nondiscretionary 18B. 'Reasonable Assurance' Cannot Apply Alike To Two Regulati ons Addressing The Same Subjec t When One Is Deemed To Be W eaker Than The Other 20C. Variabil ities In Sister Plant Data Erroneously Allowed Inappropriate Comparisons 22IV. Conclusion 22Certifica te of Servic e 25-i-TABLE OF AUTHORITIE SCasesAmerGen Energy Co., LLC (Oyster Cree k Nuclear Generating Station), L BP-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), L BP-82-116, 16 NRC 1937, 1946 (1982) 21Matter of Entergy Nucle ar Generation Co., et al.

(Pilgrim Nuc lear Power Station),

50-293-LR (ASLB Oct. 16, 2006), 2006 WL 480114223Power Authority of the State of New Y ork, et al.

(James FitzP atrick Nuclear PowerPlant; Indian Point Nuclear Generating Unit 3), CL I-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. § 50.57 2010 C.F.R. § 50.61 1, 2, 3, 4, 6, 7, 8, 9, 12, 15, 16, 18, 20, 22 10 C.F.R. § 50.61a 1, 2, 3, 7, 8, 9, 10, 11, 12, 14, 15, 16, 18, 19, 20, 21, 22 10 C.F.R. § 50.90 1010 C.F.R. § 50.92 2, 13

-ii-UNITED STAT ES OF AMERICANUCLEAR REG ULATORY COMMISSION Before the Com missionIn the Matter of

Entergy Nuclear Operations, I nc.(Palisades Nuc lear Plant)Operating License Amendment Reque st) Docket No. 50-255

)June 2, 2015

))PETITIONERS' 10 C.F

.R. § 2.311( c

) NOTICE OF A PPEAL OF ATOMIC SAF ETY AND LICENSING BOARD'S DENIAL OF 'PETITION TO INTE RVENE AND REQUEST FOR A HEARING ON ENTERGY LICENSE AMENDMENT REQUEST FOR AUTHORIZATION TO IMPLEMENT 10 C.F.R

. § 50.61a' Beyond Nuclea r, Don't Waste Michig an, Michig an Safe Energy Future - Shoreline Chapter (

Shoreline), a nd the Nucle ar Energy Information Servic e (NEIS) (collec tively"Petitioners")

, by and throug h counsel, pursua nt to 10 C.F.R. § 2.311(c), he reby give notice of their appe al to the U.S. Nuclea r Regulatory Commission ("Commissi on") for review of the Atomic Safety and Licensing Board's ("ASLB") "Memorandum and O rder (Ruling on Petition to Intervene and Re quest for a Hearing", LBP-15-17 (

May 8, 2015) whe rein the A SLB deniedPetitioners' "Petition to I ntervene and for a Public Adjudica tion Hear ing of Entergy LicenseAmendment Reque st for Authorization to I mplement 10 CFR § 50.61a, 'A lternate FractureToughness Requireme nts for Protection Ag ainst Pressurized Therma l Shock Events.'"

According to 10 C.F.R. § 2.311( c), "

An order denying a petition to int ervene, and/or request for hea ring . . . is appea lable by the requestor/petitioner on the que stion as to whether the request and/or petition should have bee n granted." Petitioners intend to urg e on appe al that their petition to int ervene and request for a hearing should have be en granted. /s/ Terry J. Lodge Terry J. Lodge (OH #0029271) 316 N. Michig an St., Ste. 520 Toledo, OH 43604-5627(419) 255-7552Fax (419) 255-7552Tjlodge50@yahoo.comCounsel for Petitioners UNITED STAT ES OF AMERICANUCLEAR REG ULATORY COMMISSION Before the Com missionIn the Matter of

Entergy Nuclear Operations, I nc.(Palisades Nuc lear Plant)Operating License Amendment Reque st) Docket No. 50-255

)June 2, 2015

))BRIEF IN SUP PORT OF PETITIONERS'10 C.F.R. § 2.311( c

) APPEAL OF ATOMIC SAF ETY ANDLICENSING BOARD'S DENIAL OF

'PETITION TO INTE RVENEAND REQUEST FOR A HEARING ON ENTERGY LICENSE AMENDMENT REQUEST FOR AUTHORIZATION TO IMP LEMENT 10 C.F.R

. § 50.61a

'I. Introduction This proce eding concerns Enterg y Nuclear Operations, I nc.'s ("Entergy's") request toamend the ope rating license f or the Palisades nuc lear plant ("Palisade s"). Palisade s is a single pressurized wa ter reactor ("PWR") fac ility located on the eastern shore of Lake Michigan, fivemiles south of Sout h Haven, Michig an. The requested amendme nt would permit Enterg y touse an alternate method to evaluate the minimum fracture toughness require d by the Palisades reactor pressure vessel (RPV) to safe ly withstand a pre ssurized thermal shock (PTS) eve nt.That alter nate method is set for th in an ag ency regulation, "Alterna te fracture toughnessrequirements for prote ction against pressurized therma l shock eve nts." In an oper ating nuclearpower plant, the re actor vessel is continuously exposed to neutrons from fission rea ctionsoccurring inside the vessel. Ove r time, this neutron radia tion embrittles the RP V walls, making them less able to re sist fractur ing, i.e., "fracture toughness" de creases. If there is a flaw in a reactor vessel wall that is embrittled due to neutron e xposure, c ertain events ca n cause the flaw topropagate throug h the wall, re sulting in a bre ach of the RPV and a possible ac cident. Of significant concern is a pr essurized thermal shoc k, or "PTS," eve nt, which is "cha racterized by arapid cooling (i.e., thermal shock) of the interna l RPV surface and downcomer

, which may befollowed by repressurization of the RPV."

The possible trig gers of a PTS event include "

a pipe1break or stuck-ope n valve in the pr imary pressure circuit," or "a break of the main stea m line." 2On September 30, 2014, the NRC Staff (the Staff

) published notice of Entergy's LAR,and concluded that the L AR presents "no signif icant hazar ds considera tion" under 10 C.F

.R. §50.92( c)

. In response to the L AR notice, Petitioners filed the instant petition to intervene a ndrequest for a he aring. 3Division of Fuel, Eng ineering and Radiologic al Research, Office of Nuclear Regulatory1Research, Technical Basis for Revision of the Pressurized Ther mal Shock (PTS) Scree ningLimit in the PTS Rul e (10 CFR 50.61) Summary Report, NUREG-1806 at xi x (Aug. 2007), at http://www.nrc.gov/reading-rm/doc-collections/nureg s/staff/sr1806/v1/ (her einafter "AlternatePTS Rule Technical Basis Report")

. Id. at xix; see also "Alternate Fracture Toug hness Requireme nts for Protection Ag ainst2Pressurized Therma l Shock Events, Final Rule,"

75 Fed. Reg. 13, 14 (Jan. 4, 2010). During these sce narios, "the water level in the cor e drops a s a result of" depr essurization or leaks.

Alternate PTS Rul e Technical Basis Report at x ix. Emergency makeup wa ter is then adde d to thereactor cooling loop, either manually or automatica lly, to keep the r eactor core covered withwater. Id. As the make up water is much colder tha n the wate r in the re actor, a rapid cooling of the outside rea ctor wall results.

Id. For over-embrittled RPVs, the temperatur e shock "

could besufficient to init iate a running crack, which could pr opagate all the way through the vessel wa ll."Id. As the re actor is still producing heat, even in a shutdown mode, the RPV could re-pr

essurize, adding additional stress to the alre ady-propagating crack. See id. at xix, xxiv, xxv ("A majorcontributor to the risk-sig nificance of [cer tain PTS events]

is the return to full sy stem pressure "after cold make up water is introduced. This could occ ur, for example, when a stuc k-open va lverecloses)."Amended Petition t o Intervene and for a Public Adjudica tion Hear ing of Entergy3License Amendment Reque st for Authorization to I mplement 10 CFR §50.61a, 'A lternateFracture Toughness Requireme nts for Protection Ag ainst Pressurized Therma l Shock Events'"

Petitioners' statement of the ir contention is:

The licensing framework that the N RC is applying to allow Palisades to continue to operate until August 2017 include s both non-conser vative ana lytical cha nges andmathematica lly dubious comparisons to alleg edly similar "sister" re actor vessels.

Palisades' ne utron embrittlement dilemma continues to worse n as the plant a ges, andPalisades has re peatedly requested life extensions which have ig nored and deferredworsening embrittlement cha racteristics of the RPV for de cades. Presently

, Entergy plansto deviate f rom the re gulatory requirements of 10 C.F.R. § 50.61 to §50.61a (A lternateFracture Toughness Requireme nts). This new ame ndment reque st introduces fur ther non-conservative ana lytical assumptions into t he troubled f orty-three (43) year operationalhistory of Palisades. Enter gy's License Amendment Reque st (LAR) contains a nequivalent mar gins evaluation, which is an untr ied methodolog ical appr oach.Petitioners' hea ring request was re ferred to an Atomic Safety and Licensing Board forconsidera tion. Both Enterg y and the NRC Staff f iled answe rs opposing the Amende d Petition, t owhich Petitioners filed a reply. On Marc h 25, 2015, the B oard heard oral argument on standing and conte ntion admissibi lity, and on May 8, 2015, the ASL B issued its "Me morandum and O rder(Ruling on Petition to I ntervene and Re quest for a Hearing"), LBP-15-17 whe rein the A SLBdenied Petitioners' A mended Petition to I ntervene and for a Public Adjudica tion Hear ing. II. Factual and Proc edural Backgroun dA. The 1985 PTS Ru le And Em brittlement Screening Program (10 C.F.R. § 50.61)

In 1985, the NRC implemented a manda tory program to monitor PW R RPVs forembrittlement over time, c oupled with scre ening limits to prevent ove r-embrittled reac tors from operating. The prog ram to monitor PW R RPVs is describe d in 10 C.F.R. Part 50, Appendix H, 4(December 8, 2014) (

hereinafter "Amended Petition").

See "Analysis of Potential P ressurized The rmal Shock Events, F inal Rule," 50 F ed. Reg.429,937 (Jul y 23, 1985) (c reating the sc reening criteria); "Fracture Toughness and Surve illanceProgram Require ments, Final Rule," 38 F ed. Reg. 19,012 (Jul y 17, 1973) (c reating the pr ogramto monitor P WR RPVs).

and is titled "Reac tor Vesse l Material Surve illance Prog ram Require ments" (Surve illanceProgram). The purpose of the Surveillanc e Program "is to moni tor changes in the fr acturetoughness properties of ferritic materia ls [iron-base d metals, such as stee l] . . . which re sult from exposure of these ma terials to neutron irr adiation and the the rmal environme nt." The5Surveillance Program relies on phy sical mater ial samples, also known a s specimens, c apsules,or coupons, "

which are withdra wn periodically from the r eactor vessel."

The NRC must pre-6approve the sche dule for r emoving material samples from the r eactor vessel.

7The actual scr eening limits require d by Appendix H's Surveillance Prog ram formonitoring re actor pressure vessels ("

RPVs") for f racture toug hness are established in 10 C.F.R.

§ 50.61, entitled "F racture toug hness requirements for protection against pressurized therma lshock eve nts." Section 50.61 relies on da ta gathered from the Surveillance Program to ca lculatethe RPV wall's fra cture toughness, and compar es it with a safe ty limit that cannot be exceeded.8NRC regulations repre sent steel fr acture toughness as a temperature value, known as "reference temperature." The NRC Staff say s, "[r]efere nce temperature is the metric that the NRC uses to quantitatively assess brittleness, so these terms may be regarded as synonymous.Steel having a high 'reference temperature' also has a hig her degree of brittleness than stee l with10 C.F.R. Part 50, App. H(I

).5Id. The NRC's re gulations further r equire that the phy sical spec imens "be loc ated near6the inside vessel wa ll in the beltline reg ion so that the specimen irr adiation history duplicates, to the extent practica ble within the phy sical constra ints of the sy stem, the neutron spe ctrum,temperature history

, and maximum neutron flue nce experienced by the reactor vessel inner surface." Id. Part 50, App. H(I II)(B)(2).Id. Part 50, App. H(I II)(B)(3).7See id. § 50.61(c)

(2)(i).8 a low reference temperature." The ability of steel to re sist fractur e changes as a function of 9temperature; whe n steel is at hig h tempera tures, it can r etain its ductility and related ability toresist fra cturing from PTS events, eve n after extended per iods of neutron irr adiation. B ut at low temperatures, stee l is naturally brittle, and eve n unirradia ted steel c an potentially suffer brittle failure. The point at which stee l transitions from the hig h-temperature, fracture-resistant-state, 10to the low-temper ature, brittle state, is called the "RTNDT," or "Transition fra cture toughnessreference temperature," or more simply "reference temperature." As descr ibed by Staff11guidance documents, this transition point depends primarily on two fac tors materia l composition and cumulative ir radiation by high-energy neutrons.

As steel is exposed to more hig h-energy12neutrons (i.e

., its fluence inc reases), RTNDT increases concurrently. Thus, as fluenc e increases,1314John B. Giessner, D ivision of Reactor Projec ts, Summary of the Mar ch 19, 2013, Public 9Meeting Webinar Reg arding Palisades Nucle ar Plant, enc

l. 2 at 4 (Apr. 18, 2013)

(ADAMSAccession No. ML 13108A336) (he reinafter "Palisades Webinar"

).See Alternate PTS Rul e Technical Basis Report at x xxviii-xxxix (noting that with steel 10at high tempera tures "cleavage cannot occur

"). A "Cleavage fracture" is the ty pe of fractureassociate d with frac ture of br ittle materials.

See id. at xxxviii.Id. at xxxiv. "NDT" stands for Nil-D uctility Transition.

Id. at xxxi.11Id. at xx ("[T]ransition temperature s increase as a result of irr adiation damag e12throughout the opera tional life of the ve ssel."); id. § 2.1.3 (discussing the factors affectingfracture toughness); id. § 2.4.2 (limiti ng the fluenc e to only high-energy "fast" neutrons, whic hhave energies above one mega electron volt).

Fluence is the integ ral of the neutron flux over time. The ne utron flux i s the total 13distance tra versed by neutrons within a unit volume of mater ial within one unit of time. Ty picallythe unit volume is one cubic c entimeter a nd the unit time is one second. Thus the unit of ne utronflux is neutron-c entimeter/c entimeter(

cubed)-second, typically expressed as ne utrons/centimeter (squared)-second. See Samuel Glasstone and A lexander Sesonske, Nuc lear Reactor Engineering§ 2.118 (Va n Nostrand Reinhold Co. 1967).

See Alternate PTS Rul e Technical Basis Report § 2.4.1 (discussing the reference14temperature approach to char acterizing fr acture toughness in fer ritic materia ls).

the steel stay s brittle at highe r and hig her temperatures, and it is there fore more likely to fractureas a result of PTS events.

The NRC established scr eening limits in 10 C.F.R. § 50.61, which are the currentscreening criteria, to reduc e the risk that a PTS event will result in an RPV frac ture. The screening limits are expressed as tempe rature values. When the re ference temperature of a n RPVis above this scre ening limit, the RPV is considered to have an unreasonably high risk of fra cturefrom a PTS eve nt. The PTS "scre ening criterion" is 270°F for plates, for gings, and axial weld 15materials, and 300°F f or circumferential weld mater ials."16If the RTNDT values proje cted at specific a reas of the RPV for the e nd of life of the plant, known as RT PTS, surpass the Curr ent Screening Criteria, the lice nsee must submit a safe ty17analysis and obtain the appr oval of the O ffice of Nuclear Reactor Regulation to continue to operate. If that off ice does not approve c ontinued opera tion based on the lice nsee's safety18analysis, the licensee must request an oppor tunity to modify the RPV or rela ted reactor systemsSee 10 C.F.R. § 50.61(b)(

2). The c urrent screening criteria "

correspond to a limit of 5 x 1510-6 events/y ear on the annua l probability of developing a through-wall crack" in the RPV.

Alternate PTS Rul e Technical Basis Report at x x.10 C.F.R. § 50.61(b)(

2); see also 75 Fed. Reg. at 13 ("

The current PTS rule . . .

16establishes scr eening criteria below whic h the potential for a reactor vessel to fail due to a PTS event is dee med to be ac ceptably low").10 C.F.R. § 50.61(a)

(7) ("RTPTS means the r eference temperature, RTNDT, evaluated for17the [end of life] Fluenc e for each of the ve ssel beltline materia ls."); Alterna te PTS Rul eTechnical Basis Report § 11.2 ("

10 CFR 50.61 define s RTPTS as the maximum RTNDT of anyregion in the vessel (a region is an axi al weld, a circumferential weld, a plate

, or a forging)evaluated at the pe ak fluence occurring in that r egion").10 C.F.R. § 50.61(b)(

3)-(5).18 to "reduce the potential for f ailure of the reactor vessel due to PTS events."

19B. The Alt ernate PTS Rul e And Embrittlement Screening Program (10 C.F.R. § 50.61a)

While no reactor is expected to exceed the c urrent screening criteria e stablished in 10 C.F.R. § 50.61 during its 40 y ear operating lice nse, the Staff has noted that Palisades in pa rticularis one of the f irst plants likely to exceed them, as Palisade s' RPV is "constructed f rom some of the most irradiation-se nsitive materials in commerc ial reactor service today

." This conce rn, as20well as sig nificant a dvancements in failure a nalysis and materia ls knowledg e, prompted the NRCto reexamine the § 50.61 a pproach for projecting fracture toughness and the scr eening criteria.21In August 2007, the NRC iss ued NUREG-1806, "Te chnical Basis for Revision of the [PTS

]Screening Limit in the PTS Rul e (10 CFR 50.61)." Tha t report summarized the r esults of a five year study by the NRC, the purpose of which "w as, to deve lop the technica l basis for re vision of the Pressurized Ther mal Shock (PTS) Rule."

The report conc luded that throug h-wall cracks22were much ha rder to create in RPVs than initially thought, a nd occurred in fe wer circum-stances. The report thus rec ommended a mor e detailed approa ch to setting screening criteria23that would take into ac count the var ying conditions along diff erent parts of the Id. § 50.61(b)(

6).19Alternate PTS Rul e Technical Basis Report at x xii.20See "Alternate Fracture Toug hness Requireme nts for Protection Ag ainst Pressurized 21Thermal Shock Events, Proposed Rule," 72 F ed. Reg. 56,275, 56,276 (Oc

t. 3, 2007); Alternate PTS Rule Technical Basis Report at iii, x x-xxiii.Alternate PTS Rul e Technical Basis Report at x ix.22See id. at xx-xxiii.23 RPV. The report also re commended r emoving the "mar gin term" that ha d been inc luded in the 24current screening criteria to acc ount for unknown f actors, because essentially all factors are nowknown and a re effectively quantified.

25On Octobe r 3, 2007, the Staff published a notice of proposed r ulemaking

. The26rulemaking notice stated tha t the Alterna te PTS Rul e Technical Basis Report "conc lude[d] thatthe risk of throug h-wall cracking due to a PTS event is much lower tha n previously estimated,"

and that "[t]hi s finding indica tes that the scr eening criteria in 10 CFR 50.61 are unnecessarilyconservative." 27On January 4, 2010, the NRC issued the final rule, c reating 10 C.F

.R. § 50.61a. The Alternate PTS Rul e makes two important chang es. Section 50.61a re places the rela tively broad28current screening criteria (270°F for plate s, forgings, and axial weld materials, and 300°F forcircumferential weld mater ials) with more de tailed Alterna te Screening Criteria.

The Alter nate29Screening Criter ia consist of eig hteen diff erent reference temperature limits that depend on RPV Id. at xxv ("Specifically, we recommend a reference temperature for flaws oc curring24along axial weld fusion lines (RT AW or RTAW-MAX), anothe r for flaws oc curring in plate s or inforgings (RTPL or TRPL-MAX), and a third for fla ws occurring along circumferential weld fusion lines (RT CW or RTCW-MAX)").Id. at xxvii.2572 Fed. Reg. 56,275.

26Id. at 56,276.

27However, like the old rule, the new rule provides mea sures for ongoing reporting, 1028C.F.R.§ 50.61a(d)

(1), and mitigation proc esses for licensee s if they project they will exceed (orthey do exceed) the Alterna te PTS Rul e's screening criteria. I

d. § 50.61a(d)

(2)-(7).75 Fed. Reg. at 18.29 wall thickness and the part of the RPV under consider ation. The Alter nate PTS Rule also 30changes how lice nsees derive proje cted reference temperatures for the c omponents of their RPVs. Section 50.61a re lies on a proba bilistic "embrittlement model" to predict f uture31reference temperatures across the RPV, which is then verif ied by existing surve illance da ta in aprocess called the "

consistency check." Section 50.61, by contrast, continuously integrates32surveillance data into future embrittlement projec tions. In the final rule making notice, the 33Commission concluded that the ne w "estimation procedure s provide a be tter (compared to theexisting regulation) method for e stimating the fr acture toughness of re actor vessel mater ials over the lifetime of the pla nt." The fina l rulemaking notice stated tha t the Alterna te PTS Rul e34"provides reasonable a ssurance that license es operating below the sc reening criteria c ould endure a PTS event without fra cture of vesse l materials, thus assuring integrity of the re actor pressure vessel." Furthermore, the final rule making stated that "[t]

he final rule will not significa ntly3510 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 license e perform a "consistency check"33of its embrittlement model ag ainst available surveillance data), and Alter nate PTS Rule Technical Basis Report § 3.1.1 (The Alternate PTS Rul e is desig ned to "e nable all commercia lPWR licensees to a ssess the state of the ir RPVs relative to such a new criterion without the nee dto make new material property measurements," instead using "only information that is curr entlyavailable

."), with 10 C.F.R. § 50.61(c)(

2)(i) (requiring that "plant-spe cific sur veillance data must be integrated into the RT NDT estimate")

, and Alter nate PTS Rule Technic al Basis Report § 2.4.2 (Under the Curre nt PTS Rule, material sa mples "fr om RPV surveillanc e programs provide theempirical basis to establish embrittlement trend cur ves . . . .")

.75 Fed. Reg. at 18.34Id. at 22.35 increase the probability or conse quences of accidents, re sult in chang es being made in the ty pesof any effluents that may be released off site, or r esult in a signif icant incr ease in occupa tional or public radia tion exposure."36C. Invocation Of The Alt ernate PTS Rul eTo take a dvantage of the Alternate PTS Rul e, a lice nsee must re quest approva l from the NRC Office of Nuclear Reactor Regulation, in accor dance with the proce dures for submitting a license a mendment under 10 C.F.R. § 50.90. The a pplication must contain: (i) under Sec tion50.61a(f), the proje cted embrittlement refe rence temperatures along various portions of the RPV, from now to a future point, compare d to the Alterna te Screening Criteria; and (

ii) under Section 50.61a(e

), an assessment of flaw s in the RPV.

In calculating e mbrittlement refe rence37temperatures under Section 50.61a(f

), a lice nsee must ca lculate ne utron flux t hrough the RPV "using a methodolog y that has bee n benchma rked to e xperimenta l measure ments and with quantified unc ertainties and possible biases."

From that point, the licensee must establish 38RTNDT(U) for various key points along the RPV. Then a licensee uses a se ries of e quations and 39charts provided in the rule to c reate an embrittlement model. That model projec ts the ref erencetemperatures for various par ts of the RPV at the end of life of the plant, known in the ne w rule asId.3610 C.F.R. § 50.61a(c

)(1)-(2). Unde r Section 50.61a, the licensee must separa tely37examine for fla ws in the rea ctor vessel. Id. § 50.61a(c

)(2). The analysis of flaws in the Palisades RPV is not in dispute in thi s proceeding.Id. § 50.61a(f

).38Id. § 50.61a(f

)(4). RTNDT(U) is the nil-ductility reference temperature for the RPV39material in the annea led state, be fore the reactor was operational. I

d. If measured values are notavailable

, a license e can use a se t of generic mean va lues. Id. § 50.61a(f

)(4)(i), (ii).

RTMAX-X. The embrittlement model allows for ca lculations of RT MAX-X across the RPV using 40probabilistic analy ses, without having to rely on measure d data. The RTMAX-X values are41compared to the Alter nate Screening Criteria to dete rmine whe ther the RPV is safe to opera te.42Importantly

, as calculations of RT MAX-X are made analytically, without directly incorpora tingsurveillance data, lice nsees have to ver ify that their ca lculations at the time of the a pplication match up with surveillanc e data. To do so, licensee s have to pe rform the "consistenc y check"43of their c alculations for specific materials against "hea t-specific surveillanc e data that are collected as part of 10 CFR Part 50, App. H, surve illance pr ograms." The purpose of the c heck44is to "determine if the surveillanc e data show a sig nificantly different trend tha n theembrittlement model predic ts." The check includes three statistical analy ses that compar e the45model's inputs, fluence and mater ial proper ties, with the model's output, refe rence temperature.46Id. § 50.61a(f

)(1)-(3). "RTMAX-X is the equivalent ter m for RTPTS in 10 CFR 50.61a."

40"Proposed Rulemaking

- Alterna te Fracture Toughness Requireme nts for Protection Ag ainstPressurized Therma l Shock Events" (RI N 3150-AI 01), 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 r egulatory history of the Alter nate PTS Rule and associa ted44draft guidance indica tes that unce rtainty in surveillance data mea surements may be a concern,which license es' applications should address.

See id. at 16-17 (discussing potential conce rnswith variability in surveillance data); "Regulatory Guidance on the Alterna te Pressure d ThermalShock Rule," Dra ft Regulatory Guide DG-1299 at 12 (Mar

. 2015) (he reinafter "DG-1299") ("Theinput variables to [the equations comprising the consistency check] are subjec t to variability andare often ba sed on limited data," pa rticularly fluence).10 C.F.R. § 50.61a(f

)(6)(i)(B).4575 Fed. Reg. at 16 ("

The NRC is modify ing the final rule to include three statistical tests 46to determine the signific ance of the diff erences between heat-specific surve illance da ta and the The consistency check is require d "[i]f three or more sur veillance data points measur ed at thre eor more dif ferent neutron f luences exist for a spe cific material." 47In the eve nt the embrittlement model deviate s from the phy sical samples ove r the limits specified in the reg ulation, the licensee must submi t additional evaluations and se ek approvalfor the de viations from the Dire ctor of the Office of Nuclear Reactor Regulation. 48D. Petitioners' Objections To Ent ergy License AmendmentRequest (LAR) Invoking A lternate PTS Ru leOn September 30, 2014, notice wa s published in the Fede ral Register of Enter gy's49intentions of seeking amendment of the oper ating license of Palisades Nucle ar Plant to allow implementation of an a lternative me thod of ca lculation of the de gree of embrittlement of the Palisades nuclea r reactor pressure vessel. The 10 C.F.R. § 50.61 scre ening criteria, to which Palisades supposedly adhered, define a limiting leve l of embrittlement bey ond which plant operation cannot continue without furthe r evaluation. The switch to the use of 10 CFR § 50.61a will chang e how fracture toug hness of the r eactor vessel is deter mined, moving f rom ananalytical to a proba bilistic risk assessment method. Ente rgy's proposed "

no signific ant hazards" determination, required by 10 C.F.R. § 50.91(a)

, conclude d that the proposed c hange will not involve a sig nificant incr ease in the probability or conse quences of an a ccident previously embrittlement trend c urve"). The consistency check compares the mea n and slope of the embrittlement model curve against surveillance data, as well as che cks to confir m that outliers fall within acc eptable r esidual value s provided in the re gulation. 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 conclude d that the proposed c hange does not c reate the possibility of50a new or diffe rent type of accident from any accident previously evaluated. The utility 51maintained, also, that the pr oposed cha nge would not involve a sig nificant re duction in a marg inof safety. In light of Ente rgy's analysis, the NRC S taff concluded that "

the three standards of 5210 CFR 50.92(c) are satisfied. Ther efore, the NRC staff proposes to dete rmine that the amendment r equest involves no sig nificant ha zards considera tion."53When the Palisades RPV was bra nd new, its ref erence temperature-nil ductility transition (RT-ndt) wa s at 40 deg rees F. By the early 1980s, NRC had wea kened Palisade s' screeningcriteria - and the re st of the U.S. pressurized wa ter reactors' - to 200 deg rees F, whic h is closer to the operating temperature of Palisade s, which is around 550 de grees F. Thus if the E mergencyCore Cooling Sy stem ("ECCS") pumps too-cold wa ter into the 550 deg rees F reactor pressure vessel and c ools it t oo quickly down to 200 deg rees F (or, later, 270 or 300 degrees), there instantaneously arises a serious potential for a fracture of the RPV, which would be a ve rysignificant reactor accident. When the PW R safety system repr essurizes the RPV, the metal ca n'ttake it any more, and fractures. It breaks, either by major cr acking or actual fragmentation, presumably at the point of a f law in the RPV.

As noted, 200 deg rees F was merely an early retreat from r egulation. The cr iteria we relater relaxed to 270 deg rees F for axial/vertical welds, and to 300 de grees F for welds of a Id. at 58815.

50Id.51Id.52Id.53 circumferential/horiz ontal orientation. And throug h it all, Palis ades and/or the NRC have projected, again and a gain that the new PTS screening criteria would be e xceeded by a predictedfuture da te. These dates ha ve been 1995; 1999; September 2001; 2004; 2007; 2014; April 2017; and August 2017. On or nea r those date s, Palisades or the NRC has said, the a llowable boundar ybeyond which lies the risk of disa ster will be cr ossed. Eac h time, though, the da te of heightenedvulnerability to this t ype of disaste r has routinely slipped back f urther into the f uture. In the many years since the e arly indicators of e mbrittlement in it s first opera tionaldecade, Palisades ha s gained notorie ty as one of the nation's most-embrittled re actors. In its May 19, 1995 NRC Gener ic Letter 1992-001, Supplement 1, the NRC Staff per mitted Palis ades to54operate until late 1999, observing that it had "re viewed the other PWR vessels and, based upon currently available information, believe s that the Palisades vesse l will reac h the PTS scree ningcriteria by late 1999, before any other PW R." (Empha sis added).

Id.Petitioners' objections to the ASL B relied in larg e part on the expert opinion of nuclear engineer Arnold Gunde rsen (see "Declaration of Arnold Gunde rsen," hereinafter "GundersenDeclaration") that the a nalysis provided to the NRC by Entergy is inadequate and relies upon unsupported assumptions which wa rrant a hearing as to whethe r Entergy should be allowed to switch over to 10 C.F

.R. § 50.61a. Petitioners urg ed the possibility exists that significa nt hazards associate d with implementation of the alterna tive calc ulation method under 10 C.F.R. § 50.61a may occur, caused by materially-underestimated prospe cts of a se vere loss-of-c oolant acc ident(LOCA) involving the reactor. ADAMS No. ML 031070449.

54 Arnold Gunde rsen state d that "Almost half of the initial capsules [coupon samples]

installed 43 y ears ago still remain inside the embrittled nuclea r reactor" and tha t if the NRC allows Enterg y to postpone the next P alisades c oupon sampling until 2019, "then no a ccuratecurrent assessment of Palisades' seve re embrittlement condition ex ists." Gunderse n Declarationp. 8, ¶ 21. Gunde rsen opined tha t § 50.61 is analy tical in nature

, while § 50.61a a uthorizes probabilistic risk assessment, and tha t the discretionar y availability of § 50.61a unde r thecircumstances c annot be use d as a substitute for sc ientific investig ation. Id. at p. 9, ¶ 24.3.

Gundersen obser ved (id. at p. 3, ¶ 8) tha t "Continued opera tion of the Palisades nucle ar powerplant without analy zing the coupon de signated to be sampled more than seve n years ago meansthat Enterg y may be operating Palisades as a test according to 10 C.F.R. § 50.59." (Emphasis in original).Petitioners' expert further alleged that the unde rlying data from other supposedly comparative nucle ar plants assessing ductility of their RPVs is not legitimate: "The NRC hasallowed Palisade s to compare itself to rea ctors of dispar ate designs from other ve ndors, built in different years and oper ating at diverse power levels." G undersen Declaration at ¶ 24.2. These plants, which he sa ys "thus far have not e xhibited significa nt signs of r eactor metal embrittle-ment," ar e poor comparables because:. . . the dra matically different nuclea r core design and oper ational power characteristics make a n accurate comparison imposs ible. The diff erence between the Westinghouse nuc lear cores and the Combustion Enginee ring nuclear core impacts the neutron flux on each r eactor vessel, thus making an accurate compar ison of neutron bombardment a nd embrittlement impossibl e.Id. at p. 10, ¶ 27.

The core objection raised by Petitioners' filing is that the 10 C.F

.R. § 50.61a alterna tive to § 50.61 allows Enterg y to substit ute various estimates of the status of the RPV for a ctual data investiga tion and analy sis. Those § 50.61a proje ctions are attained, a mong othe r means, byaveraging data on reactor vessels fr om other nucle ar power plants, to arrive a t a projec tion of the current status of the Palisades RPV. Enterg y's recourse to the alterna te approach, accompanied asit is by delibera te non-testing of metal c oupons from the RPV for 16 y ears (2003-2019) c an beunderstood only if one a ssumes that Enterg y does not want to know wha t physical testing mightattain by way of useful data about the tr ue 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 Combust ion Engineering nuclear core impacts the neutron flux on each reactorvessel, thus making an ac curate comparison of neutron bombardment and embrittlementimpossible.Id. at p. 10, ¶ 27.

A good exa mple of a false comparison is found in Structural Integrit y 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 regula tory 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 dis similarRPVs. Ssuch false comparisons, Gundersen says, "significantly dilute Palisades' embrittlement calculations." Id. at p. 11, ¶ 28.

He adds: "This rogue compara tive data is not sound scientificmethodology and cl early 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 "mathemati callyimplausible 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 t heeffective 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. Argum entA. The AS LB Erroneously Foun d The Dec ision Allowing EntergyTo Invoke 10 C.F.R. § 50.61a To Be Nondiscretion aryThe Atomic Safe ty and Licensing Board generally denied the Petition, holding that:

Petitioners appar ently want the B oard to pre clude Ente rgy from relying on Sec tion50.61a to avoid mee ting the r equirements of Section 50.61, but it is j ust such a "de via-tion" that Section 50.61a author izes. The evident pur pose of the Alternate PTS Ru le's"Alternat e Fracture Toughn ess Requirem ents" is to pr ovide an alternative to satisfying the more deman ding requi rements of Section 50.61

. Therefore, Petitioners are insubstance a sking tha t the Boa rd prohibit what Section 50.61a a llows. Under 10 C.F

.R. §2.335, we may not consider suc h a conte ntion except unde r specific conditions not present here.(Emphasis supplied). L BP-15-17 at 29.

The Licensing Board's reasoning is flawed; it involves two distinct considerations. Even assuming arguendo that the NRC can pr omulgate an alternative r egulation that is weake r than the other, and afford a choice of laws to nuclea r utility operators, that position say s nothing a bout thediscretionar y nature of the NRC Direc tor of Nuc lear Reactor Regulation over whe ther to allow a particula r applica nt to invoke 10 C.F.R. § 50.61a. The A SLB ruled, in essenc e, that if the paperwork is prope rly completed, the substantive issue -

whether to allow Enterg y to move to 10 C.F.R. § 50.61a - is esse ntially irrelevant, is to be automatica lly allowed, a nd that the NRC Staff's r egulatory hand must be stay ed. This dog matic stance is appare nt in severa l ASLBstatements. For example, the ASL B adopted Enterg y's argument that "a c ontention asserting thatdifferent analysis or technique should be utilized is inadmiss ible beca use it indirectly attacks the Commission's reg ulations." L BP-15-17 at 33. Petitioners wer e advocating, not for usag e of adifferent technique to be used, but that that the Dire ctor of N RR should have disc retionarilyconsidere d whether a superior "reasonable assurance" of protec tion of public health and sa fety would be der ived from r ejecting Ente rgy's request to invoke § 50.61a.

This is because 10 C.F.R. § 50.61a cle arly contemplates a discretionar y determination by the Director of N RR. See, for example, § 50.61a(

c)(1) (RTMAX-X values a ssessment "must specify the base s for the pr ojected va lue of RT MAX-X for each reactor vessel be ltline material, including the assumptions reg arding future pla nt operation"); § 50.61a( c

)(2) ("Each license eshall perf orm an examination and an a ssessment of flaw s in the rea ctor vessel beltline as re quiredby paragraph (e) of this section" - a nd (e) requires disclosure of te sts perfor med but, ag ain,detailed e xplanation of the me thodology underlying NDE uncertainties assumptions, and55adjustments must be disclosed. This is merely a recognition that even objective da ta, onceinterpreted, may be examined to asce rtain the objec tivity or inappropr iate bias whic h may haveoccurred in the me ans of analysis which have be en applied to it. Where the re is discre tion vested in the reg ulator, diffe rences of opinion, interpre tation, and expert ana lysis are le gitimate bases for challenging the decision bec ause the decision is potentially arrived at in an a dversarial manne

r. This principle is also obvious in § 50.61a(f)

(7), whic h require s that "The lice nsee shallreport any information that sig nificantly influence s the RTMAX-X value to the Dir ector inaccordance with the re quirements of pa ragraphs (c)(1) and (d)(1) of this section." The requirement clea rly introduces subjec tive judgme nt and selec tion among dif ferent conditions or findings into the decision of wha t data is to be provided to the D irector 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 c ritique of the mea ns by which the § 50.61a investig ation was c onducted, a nd the wea knesses or bia ses in the under lying data,assumptions and manipulations of information ca nnot be construe d as a frontal assa ult on the regulatory citadel, but must instead be se en, for purposes of the admissibili ty determination, as an exposition of the flaws c aused by straying away from knowa ble scienc

e. Petitioners' c ritique was not answer ed by any experts on behalf of the NRC Staff or Enter gy. Petitioners articulate dchallenges to the propose d exercise of discretion by the Dire ctor of N uclear Reactor Regulationand should be a ccorded a hearing to provide more e vidence.The Commission s hould take note that the a gency regulations contain a "

pressurized thermal shock r egulatory relief valve" for situations wher e a nuclear utility cannot mee t even the flaccid threshold of 10 C.F

.R. § 50.61a, by means of w hich the Dire ctor of N RR may allow an embrittled rea ctor to oper ate beyond the PTS scree ning criteria.

See slide show, "Te chnical Briefon Regulatory Guidance on the Alterna tive PTS R ule (10 C.F.R. § 50.61a

)," Official Transcriptof Proceedings, ADAMS No. ML 14321A542, at p. 242/268 of .pdf:

Use of 10 CF R 50.61a PTS screening criteria requires submittal for re view andapproval by Director, NRR.

For plants that do not satisfy PTS Screening Criteria, plant-spe cific PTS assessment is requir ed.Must be submitt ed for review and approval by Director, NRR.

Guidance is not provided for this case.

Subsequent requir ements (i.e., after submittal) are defined in para graph (d) of 10CFR 50.61a. (Empha sis suppli ed).B. 'Reasonable Assu rance' Cannot A pply Alike To Two Regulat ions Addressing The Same Subject When One Is Deemed To Be Weaker Than The Other When the ASL B referred to the 10 C.F.R. § 50.61 require ments as "more demanding "than the "A lternate Fracture Toughness Requireme nts," the B oard agreed that the "e vident purpose" of 10 C.F.R. § 50.61a is to wea ken the r egulatory rigor over nuclear utiliti es withserious RPV ductility problems. Petitioners sug gest that substitut ion of a strong er standa rd whichofficially provides "r easonable assura nce" of public protec tion with an admittedly weaker onealso "reasonably assured" to be pr

otective, is legally anomalous.

56Section 182a of the Atomic Energ y Act states that a reactor operating license must include "te chnical specifica tions" that include, inter alia

, "the spe cific characteristics of the facility, and such othe r informa tion as the Commis sion may, by rule or r egulation, deem necessary in order to e nable it to find that the utiliz ation . . . of spec ial nuclea r material . . . will provide a dequate protection to the health and saf ety of the public."

42 U.S.C. § 2232(a). The general requirement for operating lice nses, 10 C.F.R. § 50.57(a

)(3), require s a finding ofreasonable a ssurance of operation without endang ering the health and safe ty of the public.

Duke57Power Co.

(Catawba Nuclear Station, Units 1 & 2), L BP-82-116, 16 NRC 1937, 1946 (1982). I nthis procee ding, Entergy must demonstrate that it satisfies the "r easonable assura nce standard" bya preponderance of the evidenc

e. Reasonable a ssurance "is not susce ptible to formalistic quantifica tion or mecha nistic application. Rather, w hether the reasonable assurance standard ismet is based upon sound tec hnical judg ment applied on a c ase-by-case basis."

AmerGen EnergyCo., LLC (Oyster Cree k Nuclear Generating Station), L BP-07-17, 66 NRC 327, 340 (2007),

The "reasonable a ssurance" finding of 10 C.F.R. § 50.61a is found at 75 F ed. Reg. at 22.56"(a) Pursuant to § 50.56, an ope rating license ma y be issued by the Commiss ion, up to 57the full term author ized by § 50.51, upon finding that:(1) ***;

(2) ***;(3) There is reasonable assurance (i) that the ac tivities authorized by the oper ating licensecan be conducte d without endang ering the health and safe ty of the public. . ."

.

aff'd, CLI-09-07, 69 NRC 235, 263 (2009) (

rejecting an argument that rea sonable a ssuranceshould be quantified with 95% c onfidence). To consider a strong er regulation and a we aker oneto be on the same footing when it comes to providing reasonable a ssurance is logicallyinconsistent, as illustrated by this very case. Palisades contains the w orst-embrittled re actorpressure vessel in the United States. Posed a c hoice between a tougher, physical testing

-basedregulatory regime, or a w eaker, projec tive method of asse ssing RPV ductility

, owners of theworst-embrittled reac tor have chosen the less-protec tive regulations. Bec ause they are lessprotective, and g iven the enor mous discretion vested in the Dir ector of Nuclear ReactorRegulation to decide on a case-by-case basis wha t terms and conditions should be imposed under 10 C.F.R. § 50.61a, a he aring is necessary to resolve f actual issues in li ne with re gulatoryexpectations. The ASL B's candor shows that the alter native re gulation exi sts merely to provide Entergy with "re asonable assurance" of being able to oper ate Palisade s in disreg ard of thedestructive te sting oblig ations of 10 C.F.R. § 50.61 and in der ogation of the binding requirementof reasonable assurance that the public's hea lth and safe ty will be the priority for protection.

C. Variabilities In S ister Plant Data Erron eously Allow ed Inappropriate Com parisonsThe ASLB treated Petitioners' obje ctions to the invalidity of sister plant data as attempts to suggest regulatory parameters whic h exceed the r equirements of 10 C.F.R. § 50.61a. B ut Petitioners have pr eviously argued that the c onsiderable discretion ac corded the Dire ctor of N RRto allow invocation of § 50.61a should be construe d as lending relevance to their apples/ora ngesquibbling. F urther, 10 C.F

.R. § 50.61a(f)

(6)(i) requir es that "(

A) The surveillance material mustMAX-X be a heat-specific match f or one or more of the materials for which RT is being calculated."Petitioners' expert Gunder sen attested to the la ck of proof that the meta ls from the var ious RPVs match. This conc lusion was not rebutted by any expert evidenc e from either the N RC Staff norEntergy. The Licensing Board's implicit finding that the me tals compare d in the sister plants workup we re "of the appropriate chemical composition" (L BP-15-17 at 41) wa s seriously challenged by Petitioners' expert witness. Nor did Enter gy or the NRC Staff re fute Gunde rsen'sobservation that (noted at p. 17 infra) that there is "extraordinary variability between the neutron flux across the nuclea r core in this Combus tion Engine ering reactor" because of a "flux variation of as much a s 300% betwe en the 45-degree segment and the 75-degree segment," and c oncluding it was "mathe matically implausible that a 20% devia tion is poss ible when the ne utron flux i tselfvaries by 300%." G undersen Declaration p. 12, ¶ 34. Perhaps § 50.61a is the culmination of decades of lea rning about embrittlement, but it stil l cannot dispense w ith huge variations inneutron flux in P alisades, a lone. The A SLB imprope rly rejected this portion of Petitioners' contention.

IV. Conclusion The threshold admissibi lity requirements of NRC's contention rule should not be turne dinto a "for tress to deny intervention."

Power Authority of the State of New Y ork, et al.

(JamesFitzPatrick Nuclea r Power Plant; I ndian Point Nuclear Generating Unit 3), CL I-00-22, 52 NRC 266, 295 (2000). The re is no re quirement that the pe titioners' substantive ca se be made at the contention stag

e. Matter of Entergy Nucle ar Generation Co., et al.

(Pilgrim Nuc lear PowerStation), 50-293-L R (ASLB Oct. 16, 2006), 2006 WL 4801142 at (NRC) 85. The Commissionhas explained that the re quirement a t § 2.309(f)(

1)(v) "does not c all upon the interve nor to make its case a t [the contention] stage of the proc eeding, but rather to indicate wha t facts or expert opinions, be it one fac t or opinion or many

, of which it is awa re at that point in t ime which provide the ba sis for its contention."

Pilgrim at 84. The a dmissibility requirement "generally isfulfilled when the sponsor of an othe rwise acceptable conte ntion provides a brie f recitation of the factors underly ing the contention or re ferences to documents and texts t hat provide suc hreasons." Id.WHEREFORE

, the adve rse determinations of the Atomic Safe ty and Licensing Board inLBP-15-17 should be reve rsed and the matter r emanded to the AL SB for an evide ntiary hearing.Respectfully submitted, /s/ Terry J. Lodge Terry J. Lodge (OH #0029271) 316 N. Michig an St., Ste. 520 Toledo, OH 43604-5627(419) 255-7552Fax (419) 255-7552Tjlodge50@yahoo.comCounsel for Petitioners UNITED STAT ES OF AMERICANUCLEAR REG ULATORY COMMISSION Before the Com missionIn the Matter of Entergy Nuclear Operations, I nc.(Palisades Nuc lear Plant)Operating License Amendment Reque st)Docket No. 50-255

) June 2, 2015

)) *****CERTIFICATE OF SERVICE I hereby certify that copies of the foregoing "PETITIONERS' 10 C.F.R. § 2.311( c)

NOTICE OF APPEAL OF ATOMIC SAFETY AN D LICENSING BOARD'S DENI AL OF'PETITION TO INTERVENE A ND REQUEST F OR A HEARI NG ON ENTERGY LICENSEAMENDMENT REQUEST FOR AU THORIZATION TO IMPLEMENT 10 C.F.R. § 50.61a' "and the a ccompanying "BRIEF IN SUPPORT" were served by me upon the par ties to this proceeding via the NRC's Elec tronic Information Exchang e system this 2nd day of June, 2015.

/s/ Terry J. Lodge Terry J. Lodge (OH #0029271) 316 N. Michig an St., Ste. 520 Toledo, OH 43604-5627(419) 255-7552Fax (419) 255-7552Tjlodge50@yahoo.comCounsel for Petitioners