ML20076A808

Response Opposing Util & NRC Motions for Summary Disposition of Carolina Environ Study Group Contention 18/Palmetto Alliance 44.Matl Facts Do Not Relate to Reactor Ability to Withstand Stress.Affirmation of Svc Encl

ML20076A808
Person / Time
Site:	Catawba
Issue date:	08/15/1983
From:	Jeffrey Riley CAROLINA ENVIRONMENTAL STUDY GROUP
To:	Atomic Safety and Licensing Board Panel
References
NUDOCS 8308180338
Download: ML20076A808 (16)

Text

August 15, 1983 i,

1 UNITED STATES OF AMERICA NUCLEAR REGULATORY C010iISSION SOCKETED USNRC l

BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 8

17 A1059 In the Matter of

)

)

DUKE POWER COMPANY, et al.

)

Docket Nos. 50 f1Ker ncur.-

l t

)

50411&'FG 4.5

{

E d' (Catawba Nuclear Station,

)

t Units 1 and 2)

)

CESG'S OPPOSITION TO APPLICANT AND STAFF MOTIONS FOR

SUMMARY

DISPOSITION OF CESG CONTENTION 18/PALVETTO hh.

Applicant and NRC staff have filed motions for the summary disposition of CESG's Contention 18/ Palmetto Alliance 144 (MA and 1

MS respectively).

These motions state both the contention as initially filed and as " clarified" by CESG and admitted by the

~

Board.

The sense of Contention 18 is that whether the Applicant observes or fails to observe the regulations pertainin5 to reactor embrittlnment which are in effect there will be a hazard of reactor breach to which emtrittlement will be a contributor.

'CESG. entreats the Board to scrutinize the pleadings in regard to meaning in the world of experience.

Observance of a regulation will only give a satisfactory outcome if the regulation is adequate.

Frequently, projected outcomes under regulation are represented as facts.

Adequate regulations are not always observed.

The thousands of LER's and hundreds of I and E Information Notices and Bulletins, g g Power Reactor Events, etc., demonstrate that it is one thing to

$8 prescribe an intent, another to realize it.

In the present instance

-va SO the question is, do the regulations give reasonable assurance that o

gox

$g a Catawba reactor will not breach't Intervenor will seek to show mo P

that the although the regulatory lanSuaEe appears to provide this 88 a close analysis of the factors entering into reactor mao assurance, breach indicates thts assurance is more apparent than real.

'9503

CESG will cddress the alleged material facts asserted by the several arguments; present naterial facts Applicant and Staff; which do not support summary disposition and a discussion.

MATERIAL FACTS ALLEGED BY APPLICANT Cf material facts designated A through N by Applicant, CESG differs only with I and N.

Intervenors have not raised by affidavit or otherwise I.

special circumstances of a "special safety significance" relating-showing "that application directly to Catawba that make a prima facie

. [ Appendices G and H of 10 CFB Part 50) would not serve the of...

purposes for which.

[they were) adopted."

10 CFR $2.758 Applicant objects to Intervenorts " challenge to the regulations,"

There is indeed a special ci.rcumstance implicit in CESG's MA p. 10.

Heactor breach raising this matter of special safety significance.

The would have grave consequences, NUREG-2236, Table C-1, p. C-3 assertion of this contention is, in the real world, CESG's one It is beyond CESG's fiscal opportunity for a bite at the apple.

and other capabilities to raise a generic challenge to current Catawba is the plant on our embrittlement (or other) regulations.

The very special circumstance is that if we are to be doorstep.

able to raise valid technical concerns, which I believe we are able t

to do, concerning the Catawba' plant it must be here and now and in the face of existing regulations.

assertion,.Section II C 3 a of N.

Contrary to Intervenors' capsule requirements Appendix H to 10 CFR Part 50, providing specimen for plants wigh estimated end-of-life reference temp l

Staff's response to Palmetto Interogetory 36, Dec. 15, 1982, t

states' that the conditions of the subject section are met by the adjusted reference temperatures for Catawba reactors.

end of life

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o These temperatures "will be less than limit conditions of Para 5raph.

II C 3 a of Appendix H, 10 CFR Part 50."

In a response to Palmetto Intero5atory 25 of the same date Staff states that Catawba was constructed according to ASME codes dated 1971, 1972, 1967 and 1966.

The 100 F RT was in effe t during this period.

NDT KATERIAL FACTS ALLEGED BY STAFF Of material facts 1 through 20 as designated by Staff, CESG differs with nine as follows:

1.

CESG Contention 18 (Palmetto 44) claims that a safety hazard exists at Catawba because:

(1) the NRC's projection of the is nonconservNSIv,e,hich results amount of increase in reference temperature RT w

and (2) the from neutron irradiation damage, amount of reactor material degradation for the reactor vessels cannot be accurately measured.

Affidavit 92.

CESG contends that Staff.has not demonstrated that it has a sufficient information base to assert whether an RT Pfof,ection NDT is conservative or not; that the poor agreement between several methods and the large varicnce cssignable to individual values, the standard deviation with the Guthrie formula is 24; and the fact that the projected values apply to coupons rather than to the weakest spot in the reactor vessel make the projections not merely nonconservative, but worthless.

6.

Comparison of the projections using Reg. Guide 1.99 and the test results from Oconee shows that the actual increase in reference temperature has been well below that predicted, and therefore there has been no " unanticipated trapid increase in reference tenperature'

." at Oconee.

Affidavit, U 4 Staff does not define the time at which it made the high prediction for the Oconee RT Intervenor maintains that at NDT.

the time of licensing Oconee, and of promulgating the version of Appendices G cnd H to 10 CFr " art 50 then in effect, the mcgnitude of actual increase at that reactor and others was unanticipated.

The copper, nickel, phosphorus effect was not known.

_4_

11.

Appendix H, 10 CFR, Part 50 requires that all commercially operated reactor vessels have samples from their limiting materials placed in capsules which are then irradiated and subsequently withdrawn according to a schedule and tested to determine the amount of reactor vessel material embrittlement resulting from neutron irradiation damage.

Affidavit, 9 7.

It should be noted that Oconee capsules were irradiated at Crystal River, an exemption from the rule having been obtained.

See Staff response to Palmetto InteroEatory 35 a..and b. of Dec. 15, 1982.

"The Oconee reactors. surveillance programs are not in compliance with the in-vessel surveillance requirements of Appendix H, 10 CFR Part 50."

Details of the exemption granted are provided.

Further if one takes the bond of the weld to the reactor plate to be the limiting material, the Appendix H requirement is not met in that ks neither exposed to irradiation nor tested.

I 14 The combin'ation of prediction methods previously discussed and Applicant's reactor' vessel surveillance program will accurately determine the amount of reactor material degradation for the Catawba reactor vessel materials.

, Affidavit, M 7.

The Guthrie formula ascribes a standard deviation of 24 F to the RT values it arrives at.

To reach.a 95% level or conridence, NDT 48 F are added to the calculated value.

In a context were presumably was the si nal for an action, at one time exceedin5 a 100 F RT 5

NDT

~

the uncertainty associated with actually calculated RT V"1"**

NDT is enormous.

By no stretch of misuse of scientific parlance can accuracy be attributed to these heasurements.

The discrepancy between Reg. Guide 1.99 calculated RT

's and Guthrie formula NDT values further repudiates the attribution of accuracy, see MS Table I in which the difference for Catawba Unit 2 is 32 5 F exclusive of the contributions of variance.

15 The Staff ensures safe operation of the' reactor vessel during normal, anticipated upset,and test conditions by requiring the vessel to be operated within the operating limits of Appendix G, 10 CFR Part 50, which, in turn, are based upon the RTNDT ' #

limiting reactor vessel material.

Affidavit, H 8.

-5 The Staff does not refer to safe operation under accident conditions.

During a range of LOCA events the ECCS can pressurize the reactor with relatively cool water, violating the operating limits in a way not subject to operator. control..As a material fact this statement is misleading in that it misrepresents the ensuring of safe operation.

16.

Since the Catawba reactor vessel materials will have their RT accurat61y determined throughout the life of the plant, and thEh3taffwillusethehigher'oftheRT values produced by comparison of the projection methods and $he surveillance program for calculating operating limit curves (augmented by a safety factor of 2) the reactor vessels can be safely operated during normal, upset and test conditions.

Affidavit, T 8.

The accuracy imputed to RT estimates is discussed in 14 NDT foregoing.

The omission of LOCA conditions is discussed in 15 The omission of the weld metal / reactor plate specimens from the irradiation capsule is discussed in 11.

There is further, the absence of stress fatigue in the capsule specimens, a factor materially reducing the st'rength of the reactor vessel.

Fur'ther still, every real structure has a weak spot as the result of an accumulation of faults.

Capsule specimens in no way indicate the capability of the weak spot.

The reliability of a' safety factor of 2 has not been experimentally demonstrated.

17.

The Staff ensures safe operation of the reactor vessel during faulted and emergency conditions by requiring the vessel RT to comply with the screening criteria of Commission Report SECEEb2-465,

" Pressurized Thermal Shock," which states that "the risk from PTS events for reactor vessels with RT values less than the proposed screening criteria (2700F for axiaE 3 elds and 3000F for circumferential welds) is acceptable."

Affidavit,.U 9.

There is no connection between safe operation during faulted and emergency conditions and compliance with RT s reening riteria NDT of 270 F for axial wel's and 300 F for circumferential welds.

Nor d

is there a rational response to this allegation.

At best it appears to be a make-weight which includes a phrase pertaining to faulted

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. and emergency conditions.

19.

The upperbound 95% conridence RT ror Catawba units 1 and 2 reactor vessels are 1620F and 124.50 F,Nh$pectively; these values are well below the PTS screening criteria an.d indicate that the risk to the vessel during faulted and emergency conditions is acceptable.

Affidavit, t 10.

Alleged material fact 19 depends on the validity of alleged material fact 18.

As it is worthless, see foregoing, 19 is also worthless.

20.

Since"Appendiz G vessel operating limits will be based upon accurate measurements of reactor material degradation and conservative methods of predicting such degradation, there is reasonable assurance that the Cata'ba reactor vessels can and will be operated well within w

acceptable safety margins for material degradation.

Affadavit, U 11.

As stated foregoing, the measurements of coupon degradation are not accurate.

The material in the reactor is not monitored and' is subject to stress fatigue as well as radiation damage.

The weak spot is not identified nor quantified.

One conclusion of reasonable assurance that the reactor.will be operated "well within' acceptable safety margins for material de5rndation" is not supported by the available data.

And the potential for violating the limit curve under some LOCA conditions is 1 nored.

5 RELEVANT MATERIAL FACTS NOT ADVANCED BY APPLICANT OR STAFF

.1

. he inspection by the manufacturer, Klockner WA3ke' b, W R,-

~

1.

1.G., represented the Oconee-1 r,eacto'r vessel as free of flaws.

Applicant holds that a recent inspection of the reactor vessel revealed flaws, not cracks.

Applicanth response to CESG Interogatory 1, April 26, 1983 Staff responds that "[f] law indications identified therein were reported to have been produced during f abrication of the reactor vessel." May 10, 1983 filing.

2. " Changes in Oconee-1 reactor vessel have occurred during operation which, depending on word usage, are designated either cracks or flaws.

. 3 The NRC at the time of licensin5 Oconee units 1, 2, and 3 assumed the technical reasonableness of setting a limit to the end-of-life RT f 100 F.

NDT 4

The providing of six test specimen capsules for Catawba reactors is not an. expression of confidence that the RTNDT "i11 not rapidly increase.

For reactors in which it was expected the adjusted reference temperature at end of life would not exceed 100 F three capsules were required; where it was expected that 200 F would~not be exc.eeded, four capsules were required; where

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200 F was expected to be exceeded, "at least five surveillance capsules shall be provided."

Appendix H, 10.CFR 50, II 3 a, b, i

and c.

Revision of Jan.'1, 1978.

5 The Catawba. reactor vessels have required exemptions from 10 CFR 50 Appendix G Paragraphs III B 1, III B 1, III C 1, IV A 1, 4

IV r 3, and IV B and Appendix H II.C 3 SER 5.3 3. This includes the failing of vessel 1 to meet the reactor beltline material l

requirement of 75 foot pounds.

6.

The Oconee reactor surveillance program'is not in l

compliance.with the requirements of Appendix H, 10 CFR 50.

Staff response to Palmetto Interrogatory 35 b, Dec. 15, 1982.

It is, instead, a member of a Babcock and Wilcox Owner's Group.

The number of capsules initially placed, consistent with a NDT' refle ts an anticipation of a lower rate of increase l

lower RT in reference temperature than was actually experienced.

7 Reactor embrittlement has subsequently been perceived as a major problem.

"There have been hundreds of studies, documents, technical reports and treatises and volumes of testimony dealing with the subject matter of embrittlement."

Staff response to Palmetto Interrogatory 1, Dec. 15, 1982.

l__

O 8.

Investi5ation has disclosed that rapid embrittlement on irradiation of reactor vessel materials is associated with the levels of copper, nickel and phosphorus.

Staff response to Palmetto Interrogatory 3, Dec. 15, 1982 9.

The Staff believes that rapid embrittlement will not be s

a problem at Catawba'because the concentrations of copper, nickel and phosphorus will be lower tha'n at Oconee.

Staff response to Palmetto interrogatory 3, Dec. 15, 1982.

10.

Not enough time has passed nor experience been accumulated to confirm Staff's belief (Staff's language: "The staff believes.

.").

11 The Staff has not used a well delineated nor uniform A variety of methods is used approach in its references.to RTNDT.

to assay fracture toughness: Reg. Guide 1.99 method; Guthrie formula;

" fracture mechanics approach."

Response to Palmetto Interrogatory 21, Dec. 15, 1982.

12.

" Reactor Vessel Materials Toughness" is Unresolved Safety Issue Task A-11.

Staff response to Palmetto Interrogatory 21, Dec. 15, 1982.

Pressurized Thermal Shock is Unresolved Safety Issue A-49.

13 SER 5 3 1.3 14 Staff is not consistent in its interpretation and application of RT It is a material ~ property of an irradiated specimen.

NDT.

Response to 'CESG Interrogatory 3, May 18, 1983 But Staff also refers to the RT "at the vessel ID".

Response to Palmetto NDT j

Interrogatory 3, Dec. 15, 1982.

15 Capsules contain specimens of reactor plate material, of weld met'al, but not of weld to plate.

There is no testing of welds.

In this context the weld is the interface between plate and weld me?,al.

+ 16.

The weld metal / reactor plate interfaces are a most likely

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site of flaws and a most likely region for flaw or crack initiation It is the re5 on wherein the' attempt is made to 1

and propagation.

bridge a discontinuity.

17 Reactor breach can be initiated by the propagation of a linear flaw or crack.

18.

Crack propagation is the most likely mechanism of reactor

.e breach and the concern of 10 CFR 50 Appendices G and H.

None of the tests required under regulation deal with:

a) the reactor plate / weld metal-interface b) specimens experiencing fatigue representative of the cyclic heatup and cooldown of the reactor vessel for which they are a test surrogate.

c) fatigued specimens experiencing a stress gradient comparable to that*of a reactor in various states, including.cooldown or an out-of-limits LOCA cooldown.

19.

Fatigue is a desis'n determinant for reactor life.

It is put at 200 cycles for the Catawba reactor.

Staff response to CESG l

Interrogatory 7, May 10, 1983 20.

The stress level at the inner reactor vessel surface is critical in respect to crack (or linear flaw) growth and propa5ation.

There is no test measuring this property for fatigued, irradiated 1

reactor plate / weld met'al interface under conditions of stress simulating rapid cooldown of a pressurized reactor, the critical case.

21.

Applicant and Staff are not even in agreement in the simple matter of the effect of a notch on a stressed tensile specimen, Applicant holding that the notch increases stress concentration, Staff holding that it decreases it.

Responses to CESG Interrogatory m

l_ _ _ _

11.

Nor do they agree as to the response to notching.

Cess

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Interrogatory 10 Nor on the failure stress levels of notched and cracked materials.

Interrogatory 12.

22 The high variance in a series of Charpy V-not'h samples c

is reflected in the high variance of the adduced values of RTNDT*

In science: - and engineering, high precision signifies low variance.

Accuracy denotes both correctness and precision.

The standard deviation of a group of replicate measurements provides a characterization of variance.

The standard deviation for the Guthrie formula derivations of RT is said to be 24 F.

In a NDT technology in which a precision of less than 1 F is commonly

~obtained, a 24 F standard deviation is incompatible with the Staff's claims of " accurate' measurements".

MS 20.

23.

None of'.the tests prescribed by regulation take place near the temperature of an operating reactor vessel, about 600 F.

The tensile properties of reactor =aterials are known to decrease with increase in temperature.

ASME Boiler and Pressure Vessel Code, Section 3, Appendix I-Stress Tables.

DISCUSSION

~

Two kinds of issue have emerged in. connection with CESG Contention 18/ Palmetto Alliance 44; legal and technical.

In the main CESG does not challen8e the proposed licensin6 action on the basis of non-compliance with re5ulations, though it notes the many exemptions given the Applicant in regard to meeting Appendices,G and H, 5

. s It foregoing.

Applicant usually accedes to Staff requirements.

is the wsy to gain Staff sanction and support in the licensing However when a situation has been misjudged, or erroneously process.

increase with neutron anticipated as in the case of Oconee RTNDT fluence, resultin5 in an Applicant being in noncompliance, the Staff revises the re5ulation, Appendices G and H have just been amended, 48 Fed. Reg. 24008 (May 27, 1983) and amends the license, 119, 119 ahd 116 for units 1, 2 and 3 respectively, Oconee mmendments CESG, whose members live within 20 miles of the MS attachment.

is concerned with the world of sites of four lar6e reactors, and is, experience, as opposed to that of regulatory legalisms, accordingly focussing on technical matters and the substance and significance of the regulations.

that in'the matter of reactor embrittle-CESO's position is this:

ment the Staff did not at the time of early licensings, the Oconee This generation, know enough to provide adequate regulation.

deficiency in knowledge is at a cost to the utility and to the Subsequently it may be at a health and safety cost.

rate payers.

And we think our material ~ facts show that the Staff still does not know enough, nor sufficiently embody in the regulations such knowledge as will subsequently protect the utility and the public.

Our deposition of Applicant's' witness on reactor embrittlement In brought forth a disclaimer of any fundamental knowledge.

response to our technical interrogatories the.andErs were provided by the vendor, not the utility.

believed at the f

Staff can be reasonably perceived as havin5

)

time of licensing Oconee that the provisions of Section II C 3 a f

were realistic and would be met.

Experience has shown that a less than 100 F was wron5 projection of an end-of-life RTNDT l

. l This unrealistic projection was also made for Robinson-2, San Onofre-1, Maine Yankee, Palisades and Yankee-Rowe.

Staff has avoided shutting down these reactors, or requiring them to be annealed, by devising operating limit curves.

MS attachment Fig. 3 1.2-1A, 1B, and IC; 2A, 2B, and 20; and 3A, 3B, and 30.

Further, by use of the limit curve procedure, Staff contemplates permitting the operation of at beltline is 270 F for reactors in which the projected RTNDT axial welds and 300 F,for circumferential welds.

This is a bit like buying a new car and being told somewhat later by the manufacturer that you shouldn't drive over forty, or, to mention a specific case, brake hard.

Clearly the Staff proceeded in the past on the basis of inadequate information.

In our view it is still proceeding with inadequate information.

It is our hope that it will not take a reactor breach to establish this point.

1.

The test coupons do not provide information in regard to the weak'est point in the reactor.

If the reactor breaches the f ailure will start at this point..

It represents the far end of the distributions of properties and stresses.

It is the critical value.

The part that doesn't fail doesn't matter.

2.

The RT information bears only indirectly a) on the NDT characteristics of the. weakest spot in the reactor and b)'the conditions to which it may be e5 posed.

The. most adver$*e"[d$'bhich a reactor is likely to be exposed 3

is a pressurized condition during a LOCA in which ECCS water at a temperature no higher than 70 F is injected. 'Under these conditions 0

the operator is not controlling.

The limit curve will be violated.

4.,The stresses and stress gradients durin5 both pressurized and unpressurized heatup and cooldown fatigue the materials of the reactor.

The capacity to withstand stress is reduced by fatigue.

13.

An aged reactor which has both embrittled anc ratigued is particularly vulnerable to the stresses of LUCA events in which the reactor coolant pressure remains high.

5 Test coupons are not ratigued.

They provide no indication of the actual condition of the most vulnerable material in the reactor.

6.

The interface between reactor plate and weld metal is not included in the coupon test program nor specified in the regulation.

7 Althou5h coupons are said to experience about a factor 4 less neutron flux than the inner surface of the reactor there is no demonstration that this is sufficient to compensate for'the deficiencies mentioned foregoing.

Proj5ctioHs 6.

The imprecission and lack of accuracy of RTnDT is demonstrated by a comparison of Re5. Guide 1.99 and Guthrie formula values.

Lifetime RTNDT

. Method Catawba-1 Catawba-2 Ref.

Reg. G. 1.99 100 F 109 F MS, mat. fact 7 Guthrie 112 F 76 5 F MS, met, fact 6 95% confidence 162 F 124 5 F MS, mat. fact 19

• f. MS Table 1 c

9.

The NRC has not claimed that brittle fracture would be eliminated by 10 CPH 50 Appendices G and H, only that 'the probability

-j

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of rap, idly propair,ating fracture 1s minimized.". MA 111 A, p.,7 Uncertainties are identified in " determining (a) material properties, 'd (b) the effects of irradiation, (c) residual, steady state and transient stress, and (d) size of flaws", ibid.

These uncertainties have neither been quantified nor resolved.

The Commission admissions

)

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correspond to many but not all of Intervenor's concerns.

14 10.

The regulations transfer focus and emphasis to test results on coupons and away from the reactor vessel, where it properly should be.

MA p.12.

UUNCLUSION None of the material facts adduced by Applicant and Staff, with which Intervenor concurs, twelve of Applicant's fourteen, eleven of Staff's twe'nty, bear directly on the actual capacity of the Catawba reactors to withstand stresses under faulted conditions, including the effects of embrittlement and fatigue.

The remainin5 material facts alleged by Applicant and Staff have been controverted.

Intervenor has presented twenty-three material facts which support the conclusion that there is not an adequate technical basis for assuring the reactor will not be breached by stresses which it may oncounter during its operating lifetime.

This Board, accordingly, should dismiss Applicant'.s and Staff's motions for summary disposition of CESG Contentention 18/ Palmetto Alliance 44 gapectfull

ubmitted, L

'1. / -

esse L. Riley for CESG e

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION

'83 M U P3 3 l

BEFORE THE ATOMIC SAFETY AND LICENSING BDARD 0FH;E CF E t a.y.

DDCK5 igm 5 Sgg oRANCH In the Matter of DUKE POWER COP.PANY, ET AL.

Docket Nos. 50-413 50-414 (Catawta Nuclear Station, Units 1 and 2) i APPIRMATION OF SERVICE I hereby affirm that copies of "CESG'S OPPOSITION TO APPLICANT AND STAFF MOTIONS FOR

SUMMARY

DISPOSITION OF CESG CONTENTION 18/ PALM 2TTO $"

in the above captioned proceeding have been served 'on the followins by deposit in the United States mail, first class, or as indicated by an asterisk, by U.S. overnight mail this 15th day of August, 1983:

• James L. Kelley, Chainnan Robert Guild, Esq.

~

Administrative Judge Attorney for the Palmetto Alliance Atomic Safety and Licensing Board P. O. Box 12097 U.S. Nuclear Regulatory Comission Charleston, South Carolina 29412

. Washington, DC 20555 Palmetto Alliance Dr. A. Dixon Callihan 21351 Devine Street Administrative Judge Columbia, South Carolina 29205 Union Carbide Corporation P. O. Box Y Oak Ridge, TN 37830 Carole F. Kagan, Attorney Atomic Safety and Licensing Board Panel Dr. Richard F. Foster U.S. Nuclear Regulatory Commission Administrative Judge Washington, DC 20555 P. O. Box 4263 Sunriver, Oregon 97702 Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Comission Richard P. Wilson, Esq.

Washington, DC 20555 Assistant Attorney General P. O. Box 11549 Columbia, South Carolina 29211 William L. Porter, Esq.

Albert V. Carr, Esq.

J. Michael McGarry, III, Esq.

Ellen T. Ruff, Esq.

Debevoise and Libennan Duke Power Company 1200 17th Street, NW P. O. Box 33189 Washington, DC 20036 Charlotte, North Carolina 28242

e e

George E. Johnson, Esq.

Counsel for NRC Staff Office of the Executive Legal Director U.S. nuclear Regulatory Comdssion Washington, D.C.

20555 Docketing & Service Section Office of the Secretary U.S. Nuclear Regulatory Comission Washington, DC 20555 Atomic Safety and Licensing Appeal Board Panel U.S. Nuclear Regulatory Comission Washington, DC 20555

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